Simulations towards optimization of a neutron/anti-neutron oscillation experiment at the European Spallation Source

NASA Astrophysics Data System (ADS)

Frost, Matthew; Kamyshkov, Yuri; Castellanos, Luis; Klinkby, Esben; US NNbar Collaboration

2015-04-01

The observation of Neutron/Anti-neutron oscillation would prove the existence of Baryon Number Violation (BNV), and thus an explanation for the dominance of matter over anti-matter in the universe. The latest experiments have shown the oscillation time to be greater than 8.6 x 107 seconds, whereas current theoretical predictions suggest times on the order of 108 to 109 seconds. A neutron oscillation experiment proposed at the European Spallation Source (ESS) would provide sensitivity of more than 1000 times previous experiments performed, thus providing a result well-suited to confirm or deny current theory. A conceptual design of the proposed experiment will be presented, as well as the optimization of key experiment components using Monte-Carlo simulation methods, including the McStas neutron ray-trace simulation package. This work is supported by the Organized Research Units Program funded by The University of Tennessee, Knoxville Office of Research and Engagement.

Population-based metaheuristic optimization in neutron optics and shielding design

NASA Astrophysics Data System (ADS)

DiJulio, D. D.; Björgvinsdóttir, H.; Zendler, C.; Bentley, P. M.

2016-11-01

Population-based metaheuristic algorithms are powerful tools in the design of neutron scattering instruments and the use of these types of algorithms for this purpose is becoming more and more commonplace. Today there exists a wide range of algorithms to choose from when designing an instrument and it is not always initially clear which may provide the best performance. Furthermore, due to the nature of these types of algorithms, the final solution found for a specific design scenario cannot always be guaranteed to be the global optimum. Therefore, to explore the potential benefits and differences between the varieties of these algorithms available, when applied to such design scenarios, we have carried out a detailed study of some commonly used algorithms. For this purpose, we have developed a new general optimization software package which combines a number of common metaheuristic algorithms within a single user interface and is designed specifically with neutronic calculations in mind. The algorithms included in the software are implementations of Particle-Swarm Optimization (PSO), Differential Evolution (DE), Artificial Bee Colony (ABC), and a Genetic Algorithm (GA). The software has been used to optimize the design of several problems in neutron optics and shielding, coupled with Monte-Carlo simulations, in order to evaluate the performance of the various algorithms. Generally, the performance of the algorithms depended on the specific scenarios, however it was found that DE provided the best average solutions in all scenarios investigated in this work.

Research and development of a dedicated collimator for 14.2 MeV fast neutrons for imaging using a D-T generator

NASA Astrophysics Data System (ADS)

Sabo-Napadensky, I.; Weiss-Babai, R.; Gayer, A.; Vartsky, D.; Bar, D.; Mor, I.; Chacham-Zada, R.; Cohen, M.; Tamim, N.

2012-06-01

One of the main problems in neutron imaging is the scattered radiation that accompanies the direct neutrons that reach the imaging detectors and affect the image quality. We have developed a dedicated collimator for 14.2 MeV fast neutrons. The collimator optimizes the amount of scattered radiation to primary neutrons that arrive at the imaging plane. We have used different materials within the collimator in order to lower the scattered radiation that arrives at the scanned object. The image quality and the signal to noise ratios that are measured show that a mixture of BORAX (Na2B4O7ṡ10H2O) and water in the experimental beam collimator give the best results. We have used GEANT4 to simulate the collimator performance, the simulations predict the optimized material looking on the ratios of the scattered to primary neutrons that contribute in the detector. We present our experimental setup, report the results of the experimental and related simulation studies with neutrons beam generated by a 14.2 MeV D-T neutron generator.

Development of a NRSE Spectrometer with the Help of McStas - Application to the Design of Present and Future Instruments

NASA Astrophysics Data System (ADS)

Kredler, L.; Häußler, W.; Martin, N.; Böni, P.

The flux is still a major limiting factor in neutron research. For instruments being supplied by cold neutrons using neutron guides, both at present steady-state and at new spallation neutron sources, it is therefore important to optimize the instrumental setup and the neutron guidance. Optimization of neutron guide geometry and of the instrument itself can be performed by numerical ray-tracing simulations using existing open-access codes. In this paper, we discuss how such Monte Carlo simulations have been employed in order to plan improvements of the Neutron Resonant Spin Echo spectrometer RESEDA (FRM II, Germany) as well as the neutron guides before and within the instrument. The essential components have been represented with the help of the McStas ray-tracing package. The expected intensity has been tested by means of several virtual detectors, implemented in the simulation code. Comparison between simulations and preliminary measurements results shows good agreement and demonstrates the reliability of the numerical approach. These results will be taken into account in the planning of new components installed in the guide system.

SU-F-T-183: Design of a Beam Shaping Assembly of a Compact DD-Based Boron Neutron Capture Therapy System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hsieh, M; Liu, Y; Nie, L

Purpose: To design a beam shaping assembly (BSA) to shape the 2.45-MeV neutrons produced by a deuterium-deuterium (DD) neutron generator and to optimize the beam output for boron neutron capture therapy of brain tumors Methods: MCNP is used for this simulation study. The simulation model consists of a neutron surface source that resembles an actual DD source and is surrounded by a BSA. The neutron source emits 2.45-MeV neutrons isotropically. The BSA is composed of a moderator, reflector, collimator and filter. Various types of materials and geometries are tested for each component to optimize the neutron output. Neutron characteristics aremore » measured with an 2×2×2-cm{sup 3} air-equivalent cylinder at the beam exit. The ideal BSA is determined by evaluating the in-air parameters, which include epithermal neutron per source neutron, fast neutron dose per epithermal neutron, and photon dose per epithermal neutron. The parameter values are compared to those recommended by the IAEA. Results: The ideal materials for reflector and thermal neutron filter were lead and cadmium, respectively. The thickness for reflector was 43 cm and for filter was 0.5 mm. At present, the best-performing moderator has 25 cm of AlF{sub 3} and 5 cm of MgF{sub 2}. This layout creates a neutron spectrum that has a peak at approximately 10 keV and produces 1.35E-4 epithermal neutrons per source neutron per cm{sup 2}. Additional neutron characteristics, fast neutrons per epithermal neutron and photon per epithermal neutron, are still under investigation. Conclusion: Working is ongoing to optimize the final layout of the BSA. The neutron spectrum at the beam exit window of the final configuration will have the maximum number of epithermal neutrons and limited photon and fast neutron contaminations within the recommended values by IAEA. Future studies will also include phantom experiments to validate the simulation results.« less

Applications of Elpasolites as a Multimode Radiation Sensor

NASA Astrophysics Data System (ADS)

Guckes, Amber

This study consists of both computational and experimental investigations. The computational results enabled detector design selections and confirmed experimental results. The experimental results determined that the CLYC scintillation detector can be applied as a functional and field-deployable multimode radiation sensor. The computational study utilized MCNP6 code to investigate the response of CLYC to various incident radiations and to determine the feasibility of its application as a handheld multimode sensor and as a single-scintillator collimated directional detection system. These simulations include: • Characterization of the response of the CLYC scintillator to gamma-rays and neutrons; • Study of the isotopic enrichment of 7Li versus 6Li in the CLYC for optimal detection of both thermal neutrons and fast neutrons; • Analysis of collimator designs to determine the optimal collimator for the single CLYC sensor directional detection system to assay gamma rays and neutrons; Simulations of a handheld CLYC multimode sensor and a single CLYC scintillator collimated directional detection system with the optimized collimator to determine the feasibility of detecting nuclear materials that could be encountered during field operations. These nuclear materials include depleted uranium, natural uranium, low-enriched uranium, highly-enriched uranium, reactor-grade plutonium, and weapons-grade plutonium. The experimental study includes the design, construction, and testing of both a handheld CLYC multimode sensor and a single CLYC scintillator collimated directional detection system. Both were designed in the Inventor CAD software and based on results of the computational study to optimize its performance. The handheld CLYC multimode sensor is modular, scalable, low?power, and optimized for high count rates. Commercial?off?the?shelf components were used where possible in order to optimize size, increase robustness, and minimize cost. The handheld CLYC multimode sensor was successfully tested to confirm its ability for gamma-ray and neutron detection, and gamma?ray and neutron spectroscopy. The sensor utilizes wireless data transfer for possible radiation mapping and network?centric deployment. The handheld multimode sensor was tested by performing laboratory measurements with various gamma-ray sources and neutron sources. The single CLYC scintillator collimated directional detection system is portable, robust, and capable of source localization and identification. The collimator was designed based on the results of the computational study and is constructed with high density polyethylene (HDPE) and lead (Pb). The collimator design and construction allows for the directional detection of gamma rays and fast neutrons utilizing only one scintillator which is interchangeable. For this study, a CLYC-7 scintillator was used. The collimated directional detection system was tested by performing laboratory directional measurements with various gamma-ray sources, 252Cf and a 239PuBe source.

Comparison of Image Restoration Methods for Lunar Epithermal Neutron Emission Mapping

NASA Technical Reports Server (NTRS)

McClanahan, T. P.; Ivatury, V.; Milikh, G.; Nandikotkur, G.; Puetter, R. C.; Sagdeev, R. Z.; Usikov, D.; Mitrofanov, I. G.

2009-01-01

Orbital measurements of neutrons by the Lunar Exploring Neutron Detector (LEND) onboard the Lunar Reconnaissance Orbiter are being used to quantify the spatial distribution of near surface hydrogen (H). Inferred H concentration maps have low signal-to-noise (SN) and image restoration (IR) techniques are being studied to enhance results. A single-blind. two-phase study is described in which four teams of researchers independently developed image restoration techniques optimized for LEND data. Synthetic lunar epithermal neutron emission maps were derived from LEND simulations. These data were used as ground truth to determine the relative quantitative performance of the IR methods vs. a default denoising (smoothing) technique. We review and used factors influencing orbital remote sensing of neutrons emitted from the lunar surface to develop a database of synthetic "true" maps for performance evaluation. A prior independent training phase was implemented for each technique to assure methods were optimized before the blind trial. Method performance was determined using several regional root-mean-square error metrics specific to epithermal signals of interest. Results indicate unbiased IR methods realize only small signal gains in most of the tested metrics. This suggests other physically based modeling assumptions are required to produce appreciable signal gains in similar low SN IR applications.

Applying Monte-Carlo simulations to optimize an inelastic neutron scattering system for soil carbon analysis

USDA-ARS?s Scientific Manuscript database

Computer Monte-Carlo (MC) simulations (Geant4) of neutron propagation and acquisition of gamma response from soil samples was applied to evaluate INS system performance characteristic [sensitivity, minimal detectable level (MDL)] for soil carbon measurement. The INS system model with best performanc...

Design of a setup for 252Cf neutron source for storage and analysis purpose

NASA Astrophysics Data System (ADS)

Hei, Daqian; Zhuang, Haocheng; Jia, Wenbao; Cheng, Can; Jiang, Zhou; Wang, Hongtao; Chen, Da

2016-11-01

252Cf is a reliable isotopic neutron source and widely used in the prompt gamma ray neutron activation analysis (PGNAA) technique. A cylindrical barrel made by polymethyl methacrylate contained with the boric acid solution was designed for storage and application of a 5 μg 252Cf neutron source. The size of the setup was optimized with Monte Carlo code. The experiments were performed and the results showed the doses were reduced with the setup and less than the allowable limit. The intensity and collimating radius of the neutron beam could also be adjusted through different collimator.

Complex optimization for big computational and experimental neutron datasets

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bao, Feng; Oak Ridge National Lab.; Archibald, Richard

Here, we present a framework to use high performance computing to determine accurate solutions to the inverse optimization problem of big experimental data against computational models. We demonstrate how image processing, mathematical regularization, and hierarchical modeling can be used to solve complex optimization problems on big data. We also demonstrate how both model and data information can be used to further increase solution accuracy of optimization by providing confidence regions for the processing and regularization algorithms. Finally, we use the framework in conjunction with the software package SIMPHONIES to analyze results from neutron scattering experiments on silicon single crystals, andmore » refine first principles calculations to better describe the experimental data.« less

Complex optimization for big computational and experimental neutron datasets

DOE PAGES

Bao, Feng; Oak Ridge National Lab.; Archibald, Richard; ...

2016-11-07

Here, we present a framework to use high performance computing to determine accurate solutions to the inverse optimization problem of big experimental data against computational models. We demonstrate how image processing, mathematical regularization, and hierarchical modeling can be used to solve complex optimization problems on big data. We also demonstrate how both model and data information can be used to further increase solution accuracy of optimization by providing confidence regions for the processing and regularization algorithms. Finally, we use the framework in conjunction with the software package SIMPHONIES to analyze results from neutron scattering experiments on silicon single crystals, andmore » refine first principles calculations to better describe the experimental data.« less

On the use of a (252Cf-3He) assembly for landmine detection by the neutron back-scattering method.

PubMed

Elsheikh, N; Viesti, G; ElAgib, I; Habbani, F

2012-04-01

Experiments were carried out to optimize the performance of the neutron back-scattering (NBS) technique in landmine detection using an assembly consisting of three different layers placed above a (252)Cf neuron source, producing about 10(4)s(-1), in conjunction with a (3)He detector. The assembly was optimized experimentally. The selected assembly configuration was then examined against different (252)Cf stand-off distances and mine burial depths using dummy landmines. Furthermore, Monte Carlo simulations were performed to study the effect of the shield when a (252)Cf source in the range 10(4)-10(7)s(-1) was employed, and to optimize the geometry for future prototypes. Copyright © 2012 Elsevier Ltd. All rights reserved.

Study on optimization of multiionization-chamber system for BNCT.

PubMed

Fujii, T; Tanaka, H; Maruhashi, A; Ono, K; Sakurai, Y

2011-12-01

In order to monitor stability of doses from the four components such as thermal, epi-thermal, fast neutron and gamma-ray during BNCT irradiation, we are developing a multiionization-chamber system. This system is consisted of four kinds of ionization chamber, which have specific sensitivity for each component, respectively. Since a suitable structure for each chamber depends on the energy spectrum of the irradiation field, the optimization study of the chamber structures for the epi-thermal neutron beam of cyclotron-based epi-thermal neutron source (C-BENS) was performed by using a Monte Carlo simulation code "PHITS" and suitable chamber-structures were determined. Copyright © 2011 Elsevier Ltd. All rights reserved.

New generation high performance in situ polarized 3He system for time-of-flight beam at spallation sources.

PubMed

Jiang, C Y; Tong, X; Brown, D R; Glavic, A; Ambaye, H; Goyette, R; Hoffmann, M; Parizzi, A A; Robertson, L; Lauter, V

2017-02-01

Modern spallation neutron sources generate high intensity neutron beams with a broad wavelength band applied to exploring new nano- and meso-scale materials from a few atomic monolayers thick to complicated prototype device-like systems with multiple buried interfaces. The availability of high performance neutron polarizers and analyzers in neutron scattering experiments is vital for understanding magnetism in systems with novel functionalities. We report the development of a new generation of the in situ polarized 3 He neutron polarization analyzer for the Magnetism Reflectometer at the Spallation Neutron Source at Oak Ridge National Laboratory. With a new optical layout and laser system, the 3 He polarization reached and maintained 84% as compared to 76% in the first-generation system. The polarization improvement allows achieving the transmission function varying from 50% to 15% for the polarized neutron beam with the wavelength band of 2-9 Angstroms. This achievement brings a new class of experiments with optimal performance in sensitivity to very small magnetic moments in nano systems and opens up the horizon for its applications.

Dosimetry in Thermal Neutron Irradiation Facility at BMRR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, J. P.; Holden, N. E.; Reciniello, R. N.

Radiation dosimetry for Neutron Capture Therapy (NCT) has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF) of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR). In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; includingmore » (1) in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2) out of core moderator remodeling, done by replacing thicker D 2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3) beam shutter upgrade to reduce strayed neutrons and gamma dose, (4) beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5) beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates) to reduce prompt gamma and fast neutron doses, (6) sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7) holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4 - 7 % lower than the statistical mean derived from the Monte-Carlo modeling (5% uncertainty). The dose rate measured by ion chambers was 6 - 10 % lower than the output tallies (7% uncertainty). The detailed dosimetry that was performed at the TNIF for the NCT will be described.« less

Dosimetry in Thermal Neutron Irradiation Facility at BMRR

NASA Astrophysics Data System (ADS)

Hu, J.-P.; Holden, N. E.; Reciniello, R. N.

2016-02-01

Radiation dosimetry for Neutron Capture Therapy (NCT) has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF) of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR). In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1) in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2) out of core moderator remodeling, done by replacing thicker D2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3) beam shutter upgrade to reduce strayed neutrons and gamma dose, (4) beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5) beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates) to reduce prompt gamma and fast neutron doses, (6) sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7) holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4-7% lower than the statistical mean derived from the Monte-Carlo modeling (5% uncertainty). The dose rate measured by ion chambers was 6-10% lower than the output tallies (7% uncertainty). The detailed dosimetry that was performed at the TNIF for the NCT will be described.

A method to optimize the shield compact and lightweight combining the structure with components together by genetic algorithm and MCNP code.

PubMed

Cai, Yao; Hu, Huasi; Pan, Ziheng; Hu, Guang; Zhang, Tao

2018-05-17

To optimize the shield for neutrons and gamma rays compact and lightweight, a method combining the structure and components together was established employing genetic algorithms and MCNP code. As a typical case, the fission energy spectrum of 235 U which mixed neutrons and gamma rays was adopted in this study. Six types of materials were presented and optimized by the method. Spherical geometry was adopted in the optimization after checking the geometry effect. Simulations have made to verify the reliability of the optimization method and the efficiency of the optimized materials. To compare the materials visually and conveniently, the volume and weight needed to build a shield are employed. The results showed that, the composite multilayer material has the best performance. Copyright © 2018 Elsevier Ltd. All rights reserved.

MINER - A Mobile Imager of Neutrons for Emergency Responders

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goldsmith, John E. M.; Brennan, James S.; Gerling, Mark D

2014-10-01

We have developed a mobile fast neutron imaging platform to enhance the capabilities of emergency responders in the localization and characterization of special nuclear material. This mobile imager of neutrons for emergency responders (MINER) is based on the Neutron Scatter Camera, a large segmented imaging system that was optimized for large-area search applications. Due to the reduced size and power requirements of a man-portable system, MINER has been engineered to fit a much smaller form factor, and to be operated from either a battery or AC power. We chose a design that enabled omnidirectional (4π) imaging, with only a ~twofoldmore » decrease in sensitivity compared to the much larger neutron scatter cameras. The system was designed to optimize its performance for neutron imaging and spectroscopy, but it does also function as a Compton camera for gamma imaging. This document outlines the project activities, broadly characterized as system development, laboratory measurements, and deployments, and presents sample results in these areas. Additional information can be found in the documents that reside in WebPMIS.« less

Design optimization for a wearable, gamma-ray and neutron sensitive, detector array with directionality estimation

NASA Astrophysics Data System (ADS)

Ayaz-Maierhafer, Birsen; Britt, Carl G.; August, Andrew J.; Qi, Hairong; Seifert, Carolyn E.; Hayward, Jason P.

2017-10-01

In this study, we report on a constrained optimization and tradeoff study of a hybrid, wearable detector array having directional sensing based upon gamma-ray occlusion. One resulting design uses CLYC detectors while the second feasibility design involves the coupling of gamma-ray-sensitive CsI scintillators and a rubber LiCaAlF6 (LiCAF) neutron detector. The detector systems' responses were investigated through simulation as a function of angle in a two-dimensional plane. The expected total counts, peak-to-total ratio, directionality performance, and detection of 40 K for accurate gain stabilization were considered in the optimization. Source directionality estimation was investigated using Bayesian algorithms. Gamma-ray energies of 122 keV, 662 keV, and 1332 keV were considered. The equivalent neutron capture response compared with 3 He was also investigated for both designs.

Characterization of a neutron imaging setup at the INES facility

NASA Astrophysics Data System (ADS)

Durisi, E. A.; Visca, L.; Albertin, F.; Brancaccio, R.; Corsi, J.; Dughera, G.; Ferrarese, W.; Giovagnoli, A.; Grassi, N.; Grazzi, F.; Lo Giudice, A.; Mila, G.; Nervo, M.; Pastrone, N.; Prino, F.; Ramello, L.; Re, A.; Romero, A.; Sacchi, R.; Salvemini, F.; Scherillo, A.; Staiano, A.

2013-10-01

The Italian Neutron Experimental Station (INES) located at the ISIS pulsed neutron source (Didcot, United Kingdom) provides a thermal neutron beam mainly used for diffraction analysis. A neutron transmission imaging system was also developed for beam monitoring and for aligning the sample under investigation. Although the time-of-flight neutron diffraction is a consolidated technique, the neutron imaging setup is not yet completely characterized and optimized. In this paper the performance for neutron radiography and tomography at INES of two scintillator screens read out by two different commercial CCD cameras is compared in terms of linearity, signal-to-noise ratio, effective dynamic range and spatial resolution. In addition, the results of neutron radiographies and a tomography of metal alloy test structures are presented to better characterize the INES imaging capabilities of metal artifacts in the cultural heritage field.

Optimization of hybrid-type instrumentation for Pu accountancy of U/TRU ingot in pyroprocessing.

PubMed

Seo, Hee; Won, Byung-Hee; Ahn, Seong-Kyu; Lee, Seung Kyu; Park, Se-Hwan; Park, Geun-Il; Menlove, Spencer H

2016-02-01

One of the final products of pyroprocessing for spent nuclear fuel recycling is a U/TRU ingot consisting of rare earth (RE), uranium (U), and transuranic (TRU) elements. The amounts of nuclear materials in a U/TRU ingot must be measured as precisely as possible in order to secure the safeguardability of a pyroprocessing facility, as it contains the most amount of Pu among spent nuclear fuels. In this paper, we propose a new nuclear material accountancy method for measurement of Pu mass in a U/TRU ingot. This is a hybrid system combining two techniques, based on measurement of neutrons from both (1) fast- and (2) thermal-neutron-induced fission events. In technique #1, the change in the average neutron energy is a signature that is determined using the so-called ring ratio method, according to which two detector rings are positioned close to and far from the sample, respectively, to measure the increase of the average neutron energy due to the increased number of fast-neutron-induced fission events and, in turn, the Pu mass in the ingot. We call this technique, fast-neutron energy multiplication (FNEM). In technique #2, which is well known as Passive Neutron Albedo Reactivity (PNAR), a neutron population's changes resulting from thermal-neutron-induced fission events due to the presence or absence of a cadmium (Cd) liner in the sample's cavity wall, and reflected in the Cd ratio, is the signature that is measured. In the present study, it was considered that the use of a hybrid, FNEM×PNAR technique would significantly enhance the signature of a Pu mass. Therefore, the performance of such a system was investigated for different detector parameters in order to determine the optimal geometry. The performance was additionally evaluated by MCNP6 Monte Carlo simulations for different U/TRU compositions reflecting different burnups (BU), initial enrichments (IE), and cooling times (CT) to estimate its performance in real situations. Copyright © 2015 Elsevier Ltd. All rights reserved.

Development of a reliable estimation procedure of radioactivity inventory in a BWR plant due to neutron irradiation for decommissioning

NASA Astrophysics Data System (ADS)

Tanaka, Ken-ichi; Ueno, Jun

2017-09-01

Reliable information of radioactivity inventory resulted from the radiological characterization is important in order to plan decommissioning planning and is also crucial in order to promote decommissioning in effectiveness and in safe. The information is referred to by planning of decommissioning strategy and by an application to regulator. Reliable information of radioactivity inventory can be used to optimize the decommissioning processes. In order to perform the radiological characterization reliably, we improved a procedure of an evaluation of neutron-activated materials for a Boiling Water Reactor (BWR). Neutron-activated materials are calculated with calculation codes and their validity should be verified with measurements. The evaluation of neutron-activated materials can be divided into two processes. One is a distribution calculation of neutron-flux. Another is an activation calculation of materials. The distribution calculation of neutron-flux is performed with neutron transport calculation codes with appropriate cross section library to simulate neutron transport phenomena well. Using the distribution of neutron-flux, we perform distribution calculations of radioactivity concentration. We also estimate a time dependent distribution of radioactivity classification and a radioactive-waste classification. The information obtained from the evaluation is utilized by other tasks in the preparatory tasks to make the decommissioning plan and the activity safe and rational.

The automatic neutron guide optimizer guide_bot

NASA Astrophysics Data System (ADS)

Bertelsen, Mads

2017-09-01

The guide optimization software guide_bot is introduced, the main purpose of which is to reduce the time spent programming when performing numerical optimization of neutron guides. A limited amount of information on the overall guide geometry and a figure of merit describing the desired beam is used to generate the code necessary to solve the problem. A generated McStas instrument file performs the Monte Carlo ray-tracing, which is controlled by iFit optimization scripts. The resulting optimal guide is thoroughly characterized, both in terms of brilliance transfer from an idealized source and on a more realistic source such as the ESS Butterfly moderator. Basic MATLAB knowledge is required from the user, but no experience with McStas or iFit is necessary. This paper briefly describes how guide_bot is used and some important aspects of the code. A short validation against earlier work is performed which shows the expected agreement. In addition a scan over the vertical divergence requirement, where individual guide optimizations are performed for each corresponding figure of merit, provides valuable data on the consequences of this parameter. The guide_bot software package is best suited for the start of an instrument design project as it excels at comparing a large amount of different guide alternatives for a specific set of instrument requirements, but is still applicable in later stages as constraints can be used to optimize more specific guides.

Characterization of the new neutron imaging and materials science facility IMAT

NASA Astrophysics Data System (ADS)

Minniti, Triestino; Watanabe, Kenichi; Burca, Genoveva; Pooley, Daniel E.; Kockelmann, Winfried

2018-04-01

IMAT is a new cold neutron imaging and diffraction instrument located at the second target station of the pulsed neutron spallation source ISIS, UK. A broad range of materials science and materials testing areas will be covered by IMAT. We present the characterization of the imaging part, including the energy-selective and energy-dispersive imaging options, and provide the basic parameters of the radiography and tomography instrument. In particular, detailed studies on mono and bi-dimensional neutron beam flux profiles, neutron flux as a function of the neutron wavelength, spatial and energy dependent neutron beam uniformities, guide artifacts, divergence and spatial resolution, and neutron pulse widths are provided. An accurate characterization of the neutron beam at the sample position, located 56 m from the source, is required to optimize collection of radiographic and tomographic data sets and for performing energy-dispersive neutron imaging via time-of-flight methods in particular.

The performance of the upgraded Los Alamos Neutron Source

NASA Astrophysics Data System (ADS)

Ito, Takeyasu; LANL UCN Source Collaboration

2017-09-01

Los Alamos National Laboratory has been operating an ultracold (UCN) source based on a solid deuterium (SD2) UCN converter driven by spallation neutrons for over 10 years. It has recently been successfully upgraded, by replacing the cryostat that contains the cold neutron moderator, SD2 volume, and vertical UCN guide. The horizontal UCN guide that transports UCN out of the radiation shield was also replaced. The new design reflects lessons learned from the 10+ year long operation of the previous version of the UCN source and is optimized to maximize the cold neutron flux at the SD2 volume, featuring a close coupled cold neutron moderator, and maximize the transport of the UCN to experiments. During the commissioning of the upgraded UCN source, data were collected to measure its performance, including cold neutron spectra as a function of the cold moderator temperature, and the UCN density in a vessel outside the source. In this talk, after a brief overview of the design of the upgraded source, the results of the performance tests and comparison to prediction will be presented. This work was funded by LANL LDRD.

Development of the prototype pneumatic transfer system for ITER neutron activation system.

PubMed

Cheon, M S; Seon, C R; Pak, S; Lee, H G; Bertalot, L

2012-10-01

The neutron activation system (NAS) measures neutron fluence at the first wall and the total neutron flux from the ITER plasma, providing evaluation of the fusion power for all operational phases. The pneumatic transfer system (PTS) is one of the key components of the NAS for the proper operation of the system, playing a role of transferring encapsulated samples between the capsule loading machine, irradiation stations, counting stations, and disposal bin. For the validation and the optimization of the design, a prototype of the PTS was developed and capsule transfer tests were performed with the developed system.

Modelisation and distribution of neutron flux in radium-beryllium source (226Ra-Be)

NASA Astrophysics Data System (ADS)

Didi, Abdessamad; Dadouch, Ahmed; Jai, Otman

2017-09-01

Using the Monte Carlo N-Particle code (MCNP-6), to analyze the thermal, epithermal and fast neutron fluxes, of 3 millicuries of radium-beryllium, for determine the qualitative and quantitative of many materials, using method of neutron activation analysis. Radium-beryllium source of neutron is established to practical work and research in nuclear field. The main objective of this work was to enable us harness the profile flux of radium-beryllium irradiation, this theoretical study permits to discuss the design of the optimal irradiation and performance for increased the facility research and education of nuclear physics.

Simulation study of accelerator based quasi-mono-energetic epithermal neutron beams for BNCT.

PubMed

Adib, M; Habib, N; Bashter, I I; El-Mesiry, M S; Mansy, M S

2016-01-01

Filtered neutron techniques were applied to produce quasi-mono-energetic neutron beams in the energy range of 1.5-7.5 keV at the accelerator port using the generated neutron spectrum from a Li (p, n) Be reaction. A simulation study was performed to characterize the filter components and transmitted beam lines. The feature of the filtered beams is detailed in terms of optimal thickness of the primary and additive components. A computer code named "QMNB-AS" was developed to carry out the required calculations. The filtered neutron beams had high purity and intensity with low contamination from the accompanying thermal, fast neutrons and γ-rays. Copyright © 2015 Elsevier Ltd. All rights reserved.

FY17 Status Report on NEAMS Neutronics Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, C. H.; Jung, Y. S.; Smith, M. A.

2017-09-30

Under the U.S. DOE NEAMS program, the high-fidelity neutronics code system has been developed to support the multiphysics modeling and simulation capability named SHARP. The neutronics code system includes the high-fidelity neutronics code PROTEUS, the cross section library and preprocessing tools, the multigroup cross section generation code MC2-3, the in-house meshing generation tool, the perturbation and sensitivity analysis code PERSENT, and post-processing tools. The main objectives of the NEAMS neutronics activities in FY17 are to continue development of an advanced nodal solver in PROTEUS for use in nuclear reactor design and analysis projects, implement a simplified sub-channel based thermal-hydraulic (T/H)more » capability into PROTEUS to efficiently compute the thermal feedback, improve the performance of PROTEUS-MOCEX using numerical acceleration and code optimization, improve the cross section generation tools including MC2-3, and continue to perform verification and validation tests for PROTEUS.« less

Design of an epithermal column for BNCT based on D D fusion neutron facility

NASA Astrophysics Data System (ADS)

Durisi, E.; Zanini, A.; Manfredotti, C.; Palamara, F.; Sarotto, M.; Visca, L.; Nastasi, U.

2007-05-01

Boron Neutron Capture Therapy (BNCT) is currently performed on patients at nuclear reactors. At the same time the international BNCT community is engaged in the development of alternative facilities for in-hospital treatments. This paper investigates the potential of a novel high-output D-D neutron generator, developed at Lawrence Berkeley National Laboratory (CA, USA), for BNCT. The simulation code MCNP-4C is used to realize an accurate study of the epithermal column in view of the treatment of deep tumours. Different materials and Beam Shaping Assemblies (BSA) are investigated and an optimized configuration is proposed. The neutron beam quality is defined by the standard free beam parameters, calculated averaging over the collimator aperture. The results are discussed and compared with the performances of other facilities.

Optimization of 6LiF:ZnS(Ag) scintillator light yield using GEANT4

NASA Astrophysics Data System (ADS)

Yehuda-Zada, Y.; Pritchard, K.; Ziegler, J. B.; Cooksey, C.; Siebein, K.; Jackson, M.; Hurlbut, C.; Kadmon, Y.; Cohen, Y.; Ibberson, R. M.; Majkrzak, C. F.; Maliszewskyj, N. C.; Orion, I.; Osovizky, A.

2018-06-01

A new cold neutron detector has been developed at the NIST Center for Neutron Research (NCNR) for the CANDoR (Chromatic Analysis Neutron Diffractometer or Reflectometer) project. Geometric and performance constraints dictate that this detector be exceptionally thin (∼ 2 mm). For this reason, the design of the detector consists of a 6LiF:ZnS(Ag) scintillator with embedded wavelength shifting (WLS) fibers. We used the GEANT4 package to simulate neutron capture and light transport in the detector to optimize the composition and arrangement of materials to satisfy the competing requirements of high neutron capture probability and light production and transport. In the process, we have developed a method for predicting light collection and total neutron detection efficiency for different detector configurations. The simulation was performed by adjusting crucial parameters such as the scintillator stoichiometry, light yield, component grain size, WLS fiber geometry, and reflectors at the outside edges of the scintillator volume. Three different detector configurations were fabricated and their test results were correlated with the simulations. Through this correlation we have managed to find a common photon threshold for the different detector configurations which was then used to simulate and predict the efficiencies for many other detector configurations. New detectors that have been fabricated based on simulation results yielding the desired sensitivity of 90% for 3.27 meV (5 Å) cold neutrons. The simulation has proven to be a useful tool by dramatically reducing the development period and the required number of detector prototypes. It can be used to test new designs with different thicknesses and different target neutron energies.

Optimisation of the manufacturing process of tritide and deuteride targets used for neutron production

NASA Astrophysics Data System (ADS)

Monnin, Carole; Bach, Pierre; Tulle, Pierre Alain; van Rompay, Marc; Ballanger, Anne

2002-03-01

As a neutron tube manufacturer, SODERN is now in charge of manufacturing tritium targets for accelerators, in cooperation with CEA/DAM/DTMN in Valduc. Specific deuterium and tritium targets are manufactured on request, according to the requirements of the users, starting from titanium targets on copper substrates, and going to more sophisticated devices. The range of possible uses is wide, including thin targets for neutron calibration, thick targets with controlled loading of deuterium and tritium, rotating targets or large size rotating targets for higher lifetimes. The activity of the targets ranges from 3.7×10 10 to 3.7×10 13 Bq (1-1000 Ci), the diameter being up to 30 cm. Sodern and the CEA/Valduc centre have developed different technologies for tritium target manufacture, allowing the selection of the best configuration for each kind of use. In order to optimize the production of high energy neutrons, the performance of tritide and deuteride titanium targets made by different processes has been studied experimentally by bombardment with 120 and 350 kV deuterons provided by electrostatic accelerators. It is then possible to optimize either neutron output or lifetime and stability or thermal behaviour. The importance of the deposit evaporation conditions on the efficiency of neutron emission is clearly demonstrated, as well as the thermomechanical stability of the Ti thin film under deuteron bombardment. The main parameters involved in the target performance are discussed from a thermodynamical approach.

Performance characteristics of the MIT fission converter based epithermal neutron beam.

PubMed

Riley, K J; Binns, P J; Harling, O K

2003-04-07

A pre-clinical characterization of the first fission converter based epithermal neutron beam (FCB) designed for boron neutron capture therapy (BNCT) has been performed. Calculated design parameters describing the physical performance of the aluminium and Teflon filtered beam were confirmed from neutron fluence and absorbed dose rate measurements performed with activation foils and paired ionization chambers. The facility currently provides an epithermal neutron flux of 4.6 x 10(9) n cm(-2) s(-1) in-air at the patient position that makes it the most intense BNCT source in the world. This epithermal neutron flux is accompanied by very low specific photon and fast neutron absorbed doses of 3.5 +/- 0.5 and 1.4 +/- 0.2 x 10(-13) Gy cm2, respectively. A therapeutic dose rate of 1.7 RBE Gy min(-1) is achievable at the advantage depth of 97 mm when boronated phenylalanine (BPA) is used as the delivery agent, giving an average therapeutic ratio of 5.7. In clinical trials of normal tissue tolerance when using the FCB, the effective prescribed dose is due principally to neutron interactions with the nonselectively absorbed BPA present in brain. If an advanced compound is considered, the dose to brain would instead be predominately from the photon kerma induced by thermal neutron capture in hydrogen and advantage parameters of 0.88 Gy min(-1), 121 mm and 10.8 would be realized for the therapeutic dose rate, advantage depth and therapeutic ratio, respectively. This study confirms the success of a new approach to producing a high intensity, high purity epithermal neutron source that attains near optimal physical performance and which is well suited to exploit the next generation of boron delivery agents.

TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

NASA Astrophysics Data System (ADS)

Powers, Jeffrey J.

2011-12-01

This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importance of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated at a system power level of 2000 MWth, took about 3.5 years to reach full plateau power, and was capable of an End of Plateau burnup of 38.7 %FIMA if considering just the neutronic constraints in the system design; however, fuel performance constraints led to a maximum credible burnup of 12.1 %FIMA due to a combination of internal gas pressure and irradiation effects on the TRISO materials (especially PyC) leading to SiC pressure vessel failures. The optimal neutron spectrum for the thorium-fueled blanket options evaluated seemed to favor a hard spectrum (low but non-zero neutron multiplier thicknesses and high TRISO packing fractions) in terms of neutronic performance but the fuel performance constraints demonstrated that a significantly softer spectrum would be needed to decrease the rate of accumulation of fast neutron fluence in order to improve the maximum credible burnup the system could achieve.

Vacuum System Upgrade for Extended Q-Range Small-Angle Neutron Scattering Diffractometer (EQ-SANS) at SNS

DOE PAGES

Stone, Christopher M.; Williams, Derrick C.; Price, Jeremy P.

2016-09-23

The Extended Q-Range Small-Angle Neutron Scattering Diffractometer (EQ-SANS) instrument at the Spallation Neutron Source (SNS), Oak Ridge, Tennessee, incorporates a 69m3 detector vessel with a vacuum system which required an upgrade with respect to performance, ease of operation, and maintenance. The upgrade focused on improving pumping performance as well as optimizing system design to minimize opportunity for operational error. This upgrade provided the following practical contributions: Reduced time required to evacuate from atmospheric pressure to 2mTorr from 500-1,000 minutes to 60-70 minutes Provided turn-key automated control with a multi-faceted interlock for personnel and machine safety.

Vacuum System Upgrade for Extended Q-Range Small-Angle Neutron Scattering Diffractometer (EQ-SANS) at SNS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stone, Christopher M.; Williams, Derrick C.; Price, Jeremy P.

The Extended Q-Range Small-Angle Neutron Scattering Diffractometer (EQ-SANS) instrument at the Spallation Neutron Source (SNS), Oak Ridge, Tennessee, incorporates a 69m3 detector vessel with a vacuum system which required an upgrade with respect to performance, ease of operation, and maintenance. The upgrade focused on improving pumping performance as well as optimizing system design to minimize opportunity for operational error. This upgrade provided the following practical contributions: Reduced time required to evacuate from atmospheric pressure to 2mTorr from 500-1,000 minutes to 60-70 minutes Provided turn-key automated control with a multi-faceted interlock for personnel and machine safety.

Sensitivity studies of beam directionality, beam size, and neutron spectrum for a fission converter-based epithermal neutron beam for boron neutron capture therapy.

PubMed

Sakamoto, S; Kiger, W S; Harling, O K

1999-09-01

Sensitivity studies of epithermal neutron beam performance in boron neutron capture therapy are presented for realistic neutron beams with varying filter/moderator and collimator/delimiter designs to examine the relative importance of neutron beam spectrum, directionality, and size. Figures of merit for in-air and in-phantom beam performance are calculated via the Monte Carlo technique for different well-optimized designs of a fission converter-based epithermal neutron beam with head phantoms as the irradiation target. It is shown that increasing J/phi, a measure of beam directionality, does not always lead to corresponding monotonic improvements in beam performance. Due to the relatively low significance, for most configurations, of its effect on in-phantom performance and the large intensity losses required to produce beams with very high J/phi, beam directionality should not be considered an important figure of merit in epithermal neutron beam design except in terms of its consequences on patient positioning and collateral dose. Hardening the epithermal beam spectrum, while maintaining the specific fast neutron dose well below the inherent hydrogen capture dose, improves beam penetration and advantage depth and, as a desirable by-product, significantly increases beam intensity. Beam figures of merit are shown to be strongly dependent on beam size relative to target size. Beam designs with J/phi approximately 0.65-0.7, specific fast neutron doses of 2-2.6x10(-13) Gy cm2/n and beam sizes equal to or larger than the size of the head target produced the deepest useful penetration, highest therapeutic ratios, and highest intensities.

Lithium target performance evaluation for low-energy accelerator-based in vivo measurements using gamma spectroscopy.

PubMed

Aslam; Prestwich, W V; McNeill, F E

2003-03-01

The operating conditions at McMaster KN Van de Graaf accelerator have been optimized to produce neutrons via the (7)Li(p, n)(7)Be reaction for in vivo neutron activation analysis. In a number of earlier studies (development of an accelerator based system for in vivo neutron activation analysis measurements of manganese in humans, Ph.D. Thesis, McMaster University, Hamilton, ON, Canada; Appl. Radiat. Isot. 53 (2000) 657; in vivo measurement of some trace elements in human Bone, Ph.D. Thesis. McMaster University, Hamilton, ON, Canada), a significant discrepancy between the experimental and the calculated neutron doses has been pointed out. The hypotheses formulated in the above references to explain the deviation of the experimental results from analytical calculations, have been tested experimentally. The performance of the lithium target for neutron production has been evaluated by measuring the (7)Be activity produced as a result of (p, n) interaction with (7)Li. In contradiction to the formulated hypotheses, lithium target performance was found to be mainly affected by inefficient target cooling and the presence of oxides layer on target surface. An appropriate choice of these parameters resulted in neutron yields same as predicated by analytical calculations.

Monte Carlo simulation of moderator and reflector in coal analyzer based on a D-T neutron generator.

PubMed

Shan, Qing; Chu, Shengnan; Jia, Wenbao

2015-11-01

Coal is one of the most popular fuels in the world. The use of coal not only produces carbon dioxide, but also contributes to the environmental pollution by heavy metals. In prompt gamma-ray neutron activation analysis (PGNAA)-based coal analyzer, the characteristic gamma rays of C and O are mainly induced by fast neutrons, whereas thermal neutrons can be used to induce the characteristic gamma rays of H, Si, and heavy metals. Therefore, appropriate thermal and fast neutrons are beneficial in improving the measurement accuracy of heavy metals, and ensure that the measurement accuracy of main elements meets the requirements of the industry. Once the required yield of the deuterium-tritium (d-T) neutron generator is determined, appropriate thermal and fast neutrons can be obtained by optimizing the neutron source term. In this article, the Monte Carlo N-Particle (MCNP) Transport Code and Evaluated Nuclear Data File (ENDF) database are used to optimize the neutron source term in PGNAA-based coal analyzer, including the material and shape of the moderator and neutron reflector. The optimized targets include two points: (1) the ratio of the thermal to fast neutron is 1:1 and (2) the total neutron flux from the optimized neutron source in the sample increases at least 100% when compared with the initial one. The simulation results show that, the total neutron flux in the sample increases 102%, 102%, 85%, 72%, and 62% with Pb, Bi, Nb, W, and Be reflectors, respectively. Maximum optimization of the targets is achieved when the moderator is a 3-cm-thick lead layer coupled with a 3-cm-thick high-density polyethylene (HDPE) layer, and the neutron reflector is a 27-cm-thick hemispherical lead layer. Copyright © 2015 Elsevier Ltd. All rights reserved.

Development of a 6LiF/ZnS-based Neutron Multiplicity Counter

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stave, Sean C.; Behling, Richard S.; Bernacki, Bruce E.

2016-10-06

Abstract–Neutron multiplicity counters are used in safeguards to provide rapid assay of samples which contain an unknown amount of plutonium in a potentially unknown configuration. A project at PNNL is using regular and nickel-quenched 6LiF/ZnS neutron-scintillator sheets and wavelength shifting plastic for light pipes in place of 3He. A combination of laboratory and modeling work predicts a LiF/ZnS-based system to be able to match or exceed the performance of the best 3He-based systems available. Also, the Ni-quenched material is expected to allow for improved neutron/gamma-ray discrimination at twice the event rate relative to the non-Ni-quenched LiF/ZnS. A new system basedmore » on the LiF/ZnS material is under construction and components are being used to optimize the detection efficiency and neutron/gamma-ray discrimination properties. Components of the new system are partially constructed and undergoing performance testing utilizing high-speed digitizers with field programmable gate arrays to perform the neutron/gamma-ray discrimination. The expected performance of the full-scale system is expected to be nearly the same as for 3He-based systems and is due for completion in 2016.« less

Monte-Carlo Application for Nondestructive Nuclear Waste Analysis

NASA Astrophysics Data System (ADS)

Carasco, C.; Engels, R.; Frank, M.; Furletov, S.; Furletova, J.; Genreith, C.; Havenith, A.; Kemmerling, G.; Kettler, J.; Krings, T.; Ma, J.-L.; Mauerhofer, E.; Neike, D.; Payan, E.; Perot, B.; Rossbach, M.; Schitthelm, O.; Schumann, M.; Vasquez, R.

2014-06-01

Radioactive waste has to undergo a process of quality checking in order to check its conformance with national regulations prior to its transport, intermediate storage and final disposal. Within the quality checking of radioactive waste packages non-destructive assays are required to characterize their radio-toxic and chemo-toxic contents. The Institute of Energy and Climate Research - Nuclear Waste Management and Reactor Safety of the Forschungszentrum Jülich develops in the framework of cooperation nondestructive analytical techniques for the routine characterization of radioactive waste packages at industrial-scale. During the phase of research and development Monte Carlo techniques are used to simulate the transport of particle, especially photons, electrons and neutrons, through matter and to obtain the response of detection systems. The radiological characterization of low and intermediate level radioactive waste drums is performed by segmented γ-scanning (SGS). To precisely and accurately reconstruct the isotope specific activity content in waste drums by SGS measurement, an innovative method called SGSreco was developed. The Geant4 code was used to simulate the response of the collimated detection system for waste drums with different activity and matrix configurations. These simulations allow a far more detailed optimization, validation and benchmark of SGSreco, since the construction of test drums covering a broad range of activity and matrix properties is time consuming and cost intensive. The MEDINA (Multi Element Detection based on Instrumental Neutron Activation) test facility was developed to identify and quantify non-radioactive elements and substances in radioactive waste drums. MEDINA is based on prompt and delayed gamma neutron activation analysis (P&DGNAA) using a 14 MeV neutron generator. MCNP simulations were carried out to study the response of the MEDINA facility in terms of gamma spectra, time dependence of the neutron energy spectrum, neutron flux distribution. The validation of the measurements simulations with Mont-Carlo transport codes for the design, optimization and data analysis of further P&DGNAA facilities is performed in collaboration with LMN CEA Cadarache. The performance of the prompt gamma neutron activation analysis (PGNAA) for the nondestructive determination of actinides in small samples is investigated. The quantitative determination of actinides relies on the precise knowledge of partial neutron capture cross sections. Up to today these cross sections are not very accurate for analytical purpose. The goal of the TANDEM (Trans-uranium Actinides' Nuclear Data - Evaluation and Measurement) Collaboration is the evaluation of these cross sections. Cross sections are measured using prompt gamma activation analysis facilities in Budapest and Munich. Geant4 is used to optimally design the detection system with Compton suppression. Furthermore, for the evaluation of the cross sections it is strongly needed to correct the results to the self-attenuation of the prompt gammas within the sample. In the framework of cooperation RWTH Aachen University, Forschungszentrum Jülich and the Siemens AG will study the feasibility of a compact Neutron Imaging System for Radioactive waste Analysis (NISRA). The system is based on a 14 MeV neutron source and an advanced detector system (a-Si flat panel) linked to an exclusive converter/scintillator for fast neutrons. For shielding and radioprotection studies the codes MCNPX and Geant4 were used. The two codes were benchmarked in processing time and accuracy in the neutron and gamma fluxes. Also the detector response was simulated with Geant4 to optimize components of the system.

The neutronic design and performance of the Indiana University Cyclotron Facility (IUCF) Low Energy Neutron Source (LENS)

NASA Astrophysics Data System (ADS)

Lavelle, Christopher M.

Neutron scattering research is performed primarily at large-scale facilities. However, history has shown that smaller scale neutron scattering facilities can play a useful role in education and innovation while performing valuable materials research. This dissertation details the design and experimental validation of the LENS TMR as an example for a small scale accelerator driven neutron source. LENS achieves competitive long wavelength neutron intensities by employing a novel long pulse mode of operation, where the neutron production target is irradiated on a time scale comparable to the emission time of neutrons from the system. Monte Carlo methods have been employed to develop a design for optimal production of long wavelength neutrons from the 9Be(p,n) reaction at proton energies ranging from 7 to 13 MeV proton energy. The neutron spectrum was experimentally measured using time of flight, where it is found that the impact of the long pulse mode on energy resolution can be eliminated at sub-eV neutron energies if the emission time distribution of neutron from the system is known. The emission time distribution from the TMR system is measured using a time focussed crystal analyzer. Emission time of the fundamental cold neutron mode is found to be consistent with Monte Carlo results. The measured thermal neutron spectrum from the water reflector is found to be in agreement with Monte Carlo predictions if the scattering kernels employed are well established. It was found that the scattering kernels currently employed for cryogenic methane are inadequate for accurate prediction of the cold neutron intensity from the system. The TMR and neutronic modeling have been well characterized and the source design is flexible, such that it is possible for LENS to serve as an effective test bed for future work in neutronic development. Suggestions for improvements to the design that would allow increased neutron flux into the instruments are provided.

TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

DOE Office of Scientific and Technical Information (OSTI.GOV)

Powers, Jeffrey James

2011-11-30

This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importancemore » of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated at a system power level of 2000 MW th, took about 3.5 years to reach full plateau power, and was capable of an End of Plateau burnup of 38.7 %FIMA if considering just the neutronic constraints in the system design; however, fuel performance constraints led to a maximum credible burnup of 12.1 %FIMA due to a combination of internal gas pressure and irradiation effects on the TRISO materials (especially PyC) leading to SiC pressure vessel failures. The optimal neutron spectrum for the thorium-fueled blanket options evaluated seemed to favor a hard spectrum (low but non-zero neutron multiplier thicknesses and high TRISO packing fractions) in terms of neutronic performance but the fuel performance constraints demonstrated that a significantly softer spectrum would be needed to decrease the rate of accumulation of fast neutron fluence in order to improve the maximum credible burnup the system could achieve.« less

Simulation of a suite of generic long-pulse neutron instruments to optimize the time structure of the European Spallation Source.

PubMed

Lefmann, Kim; Klenø, Kaspar H; Birk, Jonas Okkels; Hansen, Britt R; Holm, Sonja L; Knudsen, Erik; Lieutenant, Klaus; von Moos, Lars; Sales, Morten; Willendrup, Peter K; Andersen, Ken H

2013-05-01

We here describe the result of simulations of 15 generic neutron instruments for the long-pulsed European Spallation Source. All instruments have been simulated for 20 different settings of the source time structure, corresponding to pulse lengths between 1 ms and 2 ms; and repetition frequencies between 10 Hz and 25 Hz. The relative change in performance with time structure is given for each instrument, and an unweighted average is calculated. The performance of the instrument suite is proportional to (a) the peak flux and (b) the duty cycle to a power of approximately 0.3. This information is an important input to determining the best accelerator parameters. In addition, we find that in our simple guide systems, most neutrons reaching the sample originate from the central 3-5 cm of the moderator. This result can be used as an input in later optimization of the moderator design. We discuss the relevance and validity of defining a single figure-of-merit for a full facility and compare with evaluations of the individual instrument classes.

Basic requirements and parameter optimization for boron neutron capture therapy of extracorporeal irradiated and auto-transplanted organs.

PubMed

Wortmann, Birgit; Knorr, Jürgen

2012-08-01

In 2001 and 2003, at the University of Pavia, Italy, boron neutron capture therapy (BNCT) has been successfully used in the treatment of hepatic colorectal metastases (Pinelli et al., 2002; Zonta et al., 2006). The treatment procedure (TAOrMINA protocol) is characterised by the auto-transplantation and extracorporeal irradiation of the liver using a thermal neutron beam. The clinical use of this approach requires well founded data and an optimized irradiation facility. In order to start with this work and to decide upon its feasibility at the research reactor TRIGA Mainz, basic data and requirements have been considered (Wortmann, 2008). Computer calculations using the ATTILA (Transpire Inc. 2006) and MCNP (LANL, 2005) codes have been performed, including data from conventional radiation therapy, from the TAOrMINA approach, resulting in reasonable estimations. Basic data and requirements and optimal parameters have been worked out, especially for use at an optimized TRIGA irradiation facility (Wortmann, 2008). Advantages of the extracorporeal irradiation with auto-transplantation and the potential of an optimized irradiation facility could be identified. Within the requirements, turning the explanted organ over by 180° appears preferable to a whole side source, similar to a permanent rotation of the organ. The design study and the parameter optimization confirm the potential of this approach to treat metastases in explanted organs. The results do not represent actual treatment data but a first estimation. Although all specific values refer to the TRIGA Mainz, they may act as a useful guide for other types of neutron sources. The recommended modifications (Wortmann, 2008) show the suitability of TRIGA reactors as a radiation source for BNCT of extracorporeal irradiated and auto-transplanted organs. Copyright © 2012 Elsevier Ltd. All rights reserved.

Reusable shielding material for neutron- and gamma-radiation

NASA Astrophysics Data System (ADS)

Calzada, Elbio; Grünauer, Florian; Schillinger, Burkhard; Türck, Harald

2011-09-01

At neutron research facilities all around the world radiation shieldings are applied to reduce the background of neutron and gamma radiation as far as possible in order to perform high quality measurements and to fulfill the radiation protection requirements. The current approach with cement-based compounds has a number of shortcomings: "Heavy concrete" contains a high amount of elements, which are not desired to obtain a high attenuation of neutron and/or gamma radiation (e.g. calcium, carbon, oxygen, silicon and aluminum). A shielding material with a high density of desired nuclei such as iron, hydrogen and boron was developed for the redesign of the neutron radiography facility ANTARES at beam tube 4 (located at a cold neutron source) of FRM-II. The composition of the material was optimized by help of the Monte Carlo code MCNP5. With this shielding material a considerable higher attenuation of background radiation can be obtained compared to usual heavy concretes.

Intensity modulated neutron radiotherapy optimization by photon proxy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Snyder, Michael; Hammoud, Ahmad; Bossenberger, Todd

2012-08-15

Purpose: Introducing intensity modulation into neutron radiotherapy (IMNRT) planning has the potential to mitigate some normal tissue complications seen in past neutron trials. While the hardware to deliver IMNRT plans has been in use for several years, until recently the IMNRT planning process has been cumbersome and of lower fidelity than conventional photon plans. Our in-house planning system used to calculate neutron therapy plans allows beam weight optimization of forward planned segments, but does not provide inverse optimization capabilities. Commercial treatment planning systems provide inverse optimization capabilities, but currently cannot model our neutron beam. Methods: We have developed a methodologymore » and software suite to make use of the robust optimization in our commercial planning system while still using our in-house planning system to calculate final neutron dose distributions. Optimized multileaf collimator (MLC) leaf positions for segments designed in the commercial system using a 4 MV photon proxy beam are translated into static neutron ports that can be represented within our in-house treatment planning system. The true neutron dose distribution is calculated in the in-house system and then exported back through the MATLAB software into the commercial treatment planning system for evaluation. Results: The planning process produces optimized IMNRT plans that reduce dose to normal tissue structures as compared to 3D conformal plans using static MLC apertures. The process involves standard planning techniques using a commercially available treatment planning system, and is not significantly more complex than conventional IMRT planning. Using a photon proxy in a commercial optimization algorithm produces IMNRT plans that are more conformal than those previously designed at our center and take much less time to create. Conclusions: The planning process presented here allows for the optimization of IMNRT plans by a commercial treatment planning optimization algorithm, potentially allowing IMNRT to achieve similar conformality in treatment as photon IMRT. The only remaining requirements for the delivery of very highly modulated neutron treatments are incremental improvements upon already implemented hardware systems that should be readily achievable.« less

Optimization of Thermal Neutron Converter in SiC Sensors for Spectral Radiation Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krolikowski, Igor; Cetnar, Jerzy; Issa, Fatima

2015-07-01

Optimization of the neutron converter in SiC sensors is presented. The sensors are used for spectral radiation measurements of thermal and fast neutrons and optionally gamma ray at elevated temperature in harsh radiation environment. The neutron converter, which is based on 10B, allows to detect thermal neutrons by means of neutron capture reaction. Two construction of the sensors were used to measure radiation in experiments. Sensor responses collected in experiments have been reproduced by the computer tool created by authors, it allows to validate the tool. The tool creates the response matrix function describing the characteristic of the sensors andmore » it was used for detailed analyses of the sensor responses. Obtained results help to optimize the neutron converter in order to increase thermal neutron detection. Several enhanced construction of the sensors, which includes the neutron converter based on {sup 10}B or {sup 6}Li, were proposed. (authors)« less

Development of a Neutron Long Counter Detector for (α, n) Cross Section Measurements at Ohio University

NASA Astrophysics Data System (ADS)

Brandenburg, Kristyn; Meisel, Zach; Brune, Carl R.; Massey, Thomas; Soltesz, Doug; Subedi, Shiv

2017-01-01

The origin of the elements from roughly zinc-to-tin (30 < Z < 50) has yet to be determined. The neutron-rich neutrino driven wind of core collapse supernova (CCSN) is a proposed site for the nucleosynthesis of these elements. However, a significant source of uncertainty exists in elemental abundance yields from astrophysics model calculations due to the uncertainty for (α , n) reaction rates, as most of the relevant cross sections have yet to be measured. We are developing a neutron long counter tailored to measure neutrons for (α , n) reaction measurements performed at The Ohio University Edwards Accelerator Laboratory. The detector design will be optimized using the Monte-Carlo N-Particle transport code (MCNP6). Details of the optimization process, as well as the present status of the detector design will be provided. The plans for first (α , n) cross section measurements will also be briefly discussed. This work was supported in part by the US Department of Energy under Grant Number DE-FG02-88ER40387.

Investigating a multi-purpose target for electron linac based photoneutron sources for BNCT of deep-seated tumors

NASA Astrophysics Data System (ADS)

Masoudi, S. Farhad; Rasouli, Fatemeh S.

2015-08-01

Recent studies in BNCT have focused on investigating appropriate neutron sources as alternatives for nuclear reactors. As the most prominent facilities, the electron linac based photoneutron sources benefit from two consecutive reactions, (e, γ) and (γ, n). The photoneutron sources designed so far are composed of bipartite targets which involve practical problems and are far from the objective of achieving an optimized neutron source. This simulation study deals with designing a compact, optimized, and geometrically simple target for a photoneutron source based on an electron linac. Based on a set of MCNPX simulations, tungsten is found to have the potential of utilizing as both photon converter and photoneutron target. Besides, it is shown that an optimized dimension for such a target slows-down the produced neutrons toward the desired energy range while keeping them economy, which makes achieving the recommended criteria for BNCT of deep-tumors more available. This multi-purpose target does not involve complicated designing, and can be considered as a significant step toward finding application of photoneutron sources for in-hospital treatments. In order to shape the neutron beam emitted from such a target, the beam is planned to pass through an optimized arrangement of materials composed of moderators, filters, reflector, and collimator. By assessment with the recommended in-air parameters, it is shown that the designed beam provides high intensity of desired neutrons, as well as low background contamination. The last section of this study is devoted to investigate the performance of the resultant beam in deep tissue. A typical simulated liver tumor, located within a phantom of human body, was subjected to the irradiation of the designed spectrum. The dosimetric results, including evaluated depth-dose curves and carried out in-phantom parameters show that the proposed configuration establishes acceptable agreement between the appropriate neutron intensity, and penetrating deep in tissue in a reasonable treatment time.

Final Stage in the Design of a Boron Neutron Capture Therapy facility at CEADEN, Cuba

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cabal, F. Padilla; Martin, G.

A neutron beam simulation study is carried out to determine the most suitable neutron energy for treatment of shallow and deep-seated brain tumors in the context of Boron Neutron Capture Therapy (BNCT). Two figures-of-merit, the therapeutic gain and the neutron fluence are utilized as beam assessment parameters. An irradiation cavity is used instead of a parallel beam port for the therapy. Calculations are performed using the MCNP5 code. After the optimization of our beam-shaper a study of the dose distribution in the head, neck, tyroids, lungs and upper and middle spine had been made. The therapeutic gain is increased whilemore » the current required for one hour treatment is decreased in comparison with the trading prototypes of NG used for BNCT.« less

The design of a multisource americium-beryllium (Am-Be) neutron irradiation facility using MCNP for the neutronic performance calculation.

PubMed

Sogbadji, R B M; Abrefah, R G; Nyarko, B J B; Akaho, E H K; Odoi, H C; Attakorah-Birinkorang, S

2014-08-01

The americium-beryllium neutron irradiation facility at the National Nuclear Research Institute (NNRI), Ghana, was re-designed with four 20 Ci sources using Monte Carlo N-Particle (MCNP) code to investigate the maximum amount of flux that is produced by the combined sources. The results were compared with a single source Am-Be irradiation facility. The main objective was to enable us to harness the maximum amount of flux for the optimization of neutron activation analysis and to enable smaller sample sized samples to be irradiated. Using MCNP for the design construction and neutronic performance calculation, it was realized that the single-source Am-Be design produced a thermal neutron flux of (1.8±0.0007)×10(6) n/cm(2)s and the four-source Am-Be design produced a thermal neutron flux of (5.4±0.0007)×10(6) n/cm(2)s which is a factor of 3.5 fold increase compared to the single-source Am-Be design. The criticality effective, k(eff), of the single-source and the four-source Am-Be designs were found to be 0.00115±0.0008 and 0.00143±0.0008, respectively. Copyright © 2014 Elsevier Ltd. All rights reserved.

The compact neutron spectrometer at ASDEX Upgrade.

PubMed

Giacomelli, L; Zimbal, A; Tittelmeier, K; Schuhmacher, H; Tardini, G; Neu, R

2011-12-01

The first neutron spectrometer of ASDEX Upgrade (AUG) was installed in November 2008. It is a compact neutron spectrometer (CNS) based on a BC501A liquid scintillating detector, which can simultaneously measure 2.45-MeV and 14-MeV neutrons emitted from deuterium (D) plasmas and γ radiation. The scintillating detector is coupled to a digital pulse shape discrimination data acquisition (DPSD) system capable of count rates up to 10(6) s(-1). The DPSD system can operate in acquisition and processing mode. With the latter n-γ discrimination is performed off-line based on the two-gate method. The paper describes the tests of the CNS and its installation at AUG. The neutron emission from the D plasma measured during a discharge with high auxiliary heating power was used to validate the CNS performance. The study of the optimal settings for the DPSD data processing to maximize the n-γ discrimination capability of the CNS is reported. The CNS measured both 2.45-MeV and 14-MeV neutrons emitted in AUG D plasmas with a maximum count rate of 5.4 × 10(5) s(-1) (>10 times higher than similar spectrometers previously achieved) with an efficiency of 9.3 × 10(-10) events per AUG neutron.

Monte Carlo simulations of neutron-scattering instruments using McStas

NASA Astrophysics Data System (ADS)

Nielsen, K.; Lefmann, K.

2000-06-01

Monte Carlo simulations have become an essential tool for improving the performance of neutron-scattering instruments, since the level of sophistication in the design of instruments is defeating purely analytical methods. The program McStas, being developed at Risø National Laboratory, includes an extension language that makes it easy to adapt it to the particular requirements of individual instruments, and thus provides a powerful and flexible tool for constructing such simulations. McStas has been successfully applied in such areas as neutron guide design, flux optimization, non-Gaussian resolution functions of triple-axis spectrometers, and time-focusing in time-of-flight instruments.

SWAN - Detection of explosives by means of fast neutron activation analysis

NASA Astrophysics Data System (ADS)

Gierlik, M.; Borsuk, S.; Guzik, Z.; Iwanowska, J.; Kaźmierczak, Ł.; Korolczuk, S.; Kozłowski, T.; Krakowski, T.; Marcinkowski, R.; Swiderski, L.; Szeptycka, M.; Szewiński, J.; Urban, A.

2016-10-01

In this work we report on SWAN, the experimental, portable device for explosives detection. The device was created as part of the EU Structural Funds Project "Accelerators & Detectors" (POIG.01.01.02-14-012/08-00), with the goal to increase beneficiary's expertise and competencies in the field of neutron activation analysis. Previous experiences and budged limitations lead toward a less advanced design based on fast neutron interactions and unsophisticated data analysis with the emphasis on the latest gamma detection and spectrometry solutions. The final device has been designed as a portable, fast neutron activation analyzer, with the software optimized for detection of carbon, nitrogen and oxygen. SWAN's performance in the role of explosives detector is elaborated in this paper. We demonstrate that the unique features offered by neutron activation analysis might not be impressive enough when confronted with practical demands and expectations of a generic homeland security customer.

Testing Moderating Detection Systems with {sup 252}Cf-Based Reference Neutron Fields

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hertel, Nolan E.; Sweezy, Jeremy; Sauber, Jeremiah S.

Calibration measurements were carried out on a probe designed to measure ambient dose equivalent in accordance with ICRP Pub 60 recommendations. It consists of a cylindrical {sup 3}He proportional counter surrounded by a 25-cm-diameter spherical polyethylene moderator. Its neutron response is optimized for dose rate measurements of neutrons between thermal energies and 20 MeV. The instrument was used to measure the dose rate in four separate neutron fields: unmoderated {sup 252}Cf, D{sub 2}O-moderated {sup 252}Cf, polyethylene-moderated {sup 252}Cf, and WEP neutron howitzer with {sup 252}Cf at its center. Dose equivalent measurements were performed at source-detector centerline distances from 50 tomore » 200 cm. The ratio of air-scatter- and room-return-corrected ambient dose equivalent rates to ambient dose equivalent rates calculated with the code MCNP are tabulated.« less

Studies on Flat Sandwich-type Self-Powered Detectors for Flux Measurements in ITER Test Blanket Modules

NASA Astrophysics Data System (ADS)

Raj, Prasoon; Angelone, Maurizio; Döring, Toralf; Eberhardt, Klaus; Fischer, Ulrich; Klix, Axel; Schwengner, Ronald

2018-01-01

Neutron and gamma flux measurements in designated positions in the test blanket modules (TBM) of ITER will be important tasks during ITER's campaigns. As part of the ongoing task on development of nuclear instrumentation for application in European ITER TBMs, experimental investigations on self-powered detectors (SPD) are undertaken. This paper reports the findings of neutron and photon irradiation tests performed with a test SPD in flat sandwich-like geometry. Whereas both neutrons and gammas can be detected with appropriate optimization of geometries, materials and sizes of the components, the present sandwich-like design is more sensitive to gammas than 14 MeV neutrons. Range of SPD current signals achievable under TBM conditions are predicted based on the SPD sensitivities measured in this work.

Lujan Center Mark-IV Target Neutronics Design Internal Review Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lisowski, Paul W.; Gallmeier, Franz; Guber, Klaus

The 1L Target Moderator Reflector System (TMRS) at the Lujan Center will need to be replaced before the CY 2020 operating cycle. A Physics Division design team investigated options for improving the overall target performance for nuclear science research with minimal reduction in performance for materials science. This review concluded that devoting an optimized arrangement of the Lujan TMRS upper tier to nuclear science and using the lower tier for materials science can achieve those goals. This would open the opportunity for enhanced nuclear science research in an important neutron energy range for NNSA. There will be no other facilitymore » in the US that will compete in the keV energy range provided flight paths and instrumentation are developed to take advantage of the neutron flux and resolution.« less

Treatment Planning for Accelerator-Based Boron Neutron Capture Therapy

NASA Astrophysics Data System (ADS)

Herrera, María S.; González, Sara J.; Minsky, Daniel M.; Kreiner, Andrés J.

2010-08-01

Glioblastoma multiforme and metastatic melanoma are frequent brain tumors in adults and presently still incurable diseases. Boron Neutron Capture Therapy (BNCT) is a promising alternative for this kind of pathologies. Accelerators have been proposed for BNCT as a way to circumvent the problem of siting reactors in hospitals and for their relative simplicity and lower cost among other advantages. Considerable effort is going into the development of accelerator-based BNCT neutron sources in Argentina. Epithermal neutron beams will be produced through appropriate proton-induced nuclear reactions and optimized beam shaping assemblies. Using these sources, computational dose distributions were evaluated in a real patient with diagnosed glioblastoma treated with BNCT. The simulated irradiation was delivered in order to optimize dose to the tumors within the normal tissue constraints. Using Monte Carlo radiation transport calculations, dose distributions were generated for brain, skin and tumor. Also, the dosimetry was studied by computing cumulative dose-volume histograms for volumes of interest. The results suggest acceptable skin average dose and a significant dose delivered to tumor with low average whole brain dose for irradiation times less than 60 minutes, indicating a good performance of an accelerator-based BNCT treatment.

Treatment Planning for Accelerator-Based Boron Neutron Capture Therapy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Herrera, Maria S.; Gonzalez, Sara J.; Minsky, Daniel M.

2010-08-04

Glioblastoma multiforme and metastatic melanoma are frequent brain tumors in adults and presently still incurable diseases. Boron Neutron Capture Therapy (BNCT) is a promising alternative for this kind of pathologies. Accelerators have been proposed for BNCT as a way to circumvent the problem of siting reactors in hospitals and for their relative simplicity and lower cost among other advantages. Considerable effort is going into the development of accelerator-based BNCT neutron sources in Argentina. Epithermal neutron beams will be produced through appropriate proton-induced nuclear reactions and optimized beam shaping assemblies. Using these sources, computational dose distributions were evaluated in a realmore » patient with diagnosed glioblastoma treated with BNCT. The simulated irradiation was delivered in order to optimize dose to the tumors within the normal tissue constraints. Using Monte Carlo radiation transport calculations, dose distributions were generated for brain, skin and tumor. Also, the dosimetry was studied by computing cumulative dose-volume histograms for volumes of interest. The results suggest acceptable skin average dose and a significant dose delivered to tumor with low average whole brain dose for irradiation times less than 60 minutes, indicating a good performance of an accelerator-based BNCT treatment.« less

Assessment of Laser-Driven Pulsed Neutron Sources for Poolside Neutron-based Advanced NDE – A Pathway to LANSCE-like Characterization at INL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roth, Markus; Vogel, Sven C.; Bourke, Mark Andrew M.

A variety of opportunities for characterization of fresh nuclear fuels using thermal (~25meV) and epithermal (~10eV) neutrons have been documented at Los Alamos National Laboratory. They include spatially resolved non-destructive characterization of features, isotopic enrichment, chemical heterogeneity and stoichiometry. The LANSCE spallation neutron source is well suited in neutron fluence and temporal characteristics for studies of fuels. However, recent advances in high power short pulse lasers suggest that compact neutron sources might, over the next decade, become viable at a price point that would permit their consideration for poolside characterization on site at irradiation facilities. In a laser-driven neutron sourcemore » the laser is used to accelerate deuterium ions into a beryllium target where neutrons are produced. At this time, the technology is new and their total neutron production is approximately four orders of magnitude less than a facility like LANSCE. However, recent measurements on a sub-optimized system demonstrated >10 10 neutrons in sub-nanosecond pulses in predominantly forward direction. The compactness of the target system compared to a spallation target may allow exchanging the target during a measurement to e.g. characterize a highly radioactive sample with thermal, epithermal, and fast neutrons as well as hard X-rays, thus avoiding sample handling. At this time several groups are working on laser-driven neutron production and are advancing concepts for lasers, laser targets, and optimized neutron target/moderator systems. Advances in performance sufficient to enable poolside fuels characterization with LANSCE-like fluence on sample within a decade may be possible. This report describes the underlying physics and state-of-the-art of the laser-driven neutron production process from the perspective of the DOE/NE mission. It also discusses the development and understanding that will be necessary to provide customized capability for characterization of irradiated fuels. Potential operational advantages compared to a spallation neutron source include reduced shielding complexity, reduced energy requirements, and a production target free of fission products. Contributors to this report include experts in laser-driven neutron production (Roth, Fernandez), laser design (Haefner, Siders, Leemans), laser target design (Glenzer), spallation target/moderator design (Mocko), neutron instrumentation and characterization applications (Vogel, Bourke).« less

A novel small-angle neutron scattering detector geometry

PubMed Central

Kanaki, Kalliopi; Jackson, Andrew; Hall-Wilton, Richard; Piscitelli, Francesco; Kirstein, Oliver; Andersen, Ken H.

2013-01-01

A novel 2π detector geometry for small-angle neutron scattering (SANS) applications is presented and its theoretical performance evaluated. Such a novel geometry is ideally suited for a SANS instrument at the European Spallation Source (ESS). Motivated by the low availability and high price of 3He, the new concept utilizes gaseous detectors with 10B as the neutron converter. The shape of the detector is inspired by an optimization process based on the properties of the conversion material. Advantages over the detector geometry traditionally used on SANS instruments are discussed. The angular and time resolutions of the proposed detector concept are shown to satisfy the requirements of the particular SANS instrument. PMID:24046504

Optimization of {sup 6}LiF:ZnS(Ag) Scintillator Light Yield Using Geant4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yehuda-Zada, Y.; Ben-Gurion University; Pritchard, K.

2015-07-01

Neutrons provide an effective tool to probe materials structure. Neutron diffraction is a method to determine the atomic and magnetic structure of a material based on neutron scattering. By this method a collimated incident beam of thermal neutrons heat the examined sample and based on the obtained diffraction pattern information on the structure of the material is provided. Research for developing a novel cold neutron detector for Chromatic Analysis Neutron Diffractometer Or Reflectometer (CANDOR) is underway at the NIST center for neutron research. The system unique design is aimed to provide over ten times fold faster analysis of materials thanmore » conventional system. In order to achieve the fast analysis a large number of neutron detectors is required. A key design constraint for this detector is the thickness of the neutron sensitive element. This is met using {sup 6}LiF:ZnS(Ag) scintillation material with embedded wavelength shifting (WLS) fibers conducting scintillation light to silicon photomultiplier photo-sensors. The detector sensitivity is determined by both the neutron capture probability ({sup 6}Li density) and the detectable light output produced by the ZnS(Ag) ionization, the latter of which is hindered by the fluorescence absorption of the scintillation light by the ZnS. Tradeoffs between the neutron capture probability, stimulated light production and light attenuation for determining the optimal stoichiometry of the {sup 6}LiF and ZnS(Ag) as well as the volume ratio of scintillator and fiber. Simulations performed using the GEANT4 Monte Carlo package were made in order to optimize the detector design. GEANT4 enables the investigation of the neutron interaction with the detector, the ionization process and the light transfer process following the nuclear process. The series of conversions required for this detector were modelled: - A cold neutron enters the sensor and is captured by {sup 6}Li in the scintillator mixture ({sup 6}Li (n,α) {sup 3}H reaction). The study of investigating the capture process probability for neutron energy of 5.1 meV to 2.27 meV (4 - 6 A) is presented. - Alpha particles and tritons travel for a few microns in the scintillation material (α ∼0.007 mm, T ∼0.04 mm) losing energy and ionizing the ZnS. The mean free path of the two particles in each of the component materials and the complete compound was investigated. - The ionization of the ZnS(Ag) scintillation material produces blue light photons with luminescence wavelength of 450 nm. The amount of light output produced for different grain sizes of ZnS is discussed. - A large portion of the scintillation photons are reabsorbed during their passage through the scintillation material. - The blue photons that reach the WLS fibers are absorbed by fluorescent dye and are re-emitted as green photons, conducted by the fiber to the SiPM photo-sensor. This work presents the CANDOR unique design and its design constrains, the results measured by the ultra-thin {sup 6}LiF:ZnS(Ag)-based neutron detector versus the simulation results for several binder concentrations. The light measurement attenuation results along with the measured stopping power were utilized to predict the sensitivity results of configuration with different ZnS grain size, weight ratios and fibers geometry (number and location). The simulations enable to optimize the final sensor design. This design successfully achieved both the high gamma rejection with a sensitive and accurate neutron event detection of 80 percent. (authors)« less

Dual-Chamber/Dual-Anode Proportional Counter Incorporating an Intervening Thin-Foil Solid Neutron Converter

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boatner, Lynn A; Neal, John S; Blackston, Matthew A

2012-01-01

A dual-chamber/dual-anode gas proportional counter utilizing thin solid 6LiF or 10B neutron converters coated on a 2-micon-thick Mylar film that is positioned between the two counter chambers and anodes has been designed, fabricated, and tested using a variety of fill gases including naturally abundant helium. In this device, neutron conversion products emitted from both sides of the coated converter foil are detected rather than having half of the products absorbed in the wall of a conventional tube type counter where the solid neutron converter is deposited on the tube wall. Geant4-based radiation transport calculations were used to determine the optimummore » neutron converter coating thickness for both isotopes. Solution methods for applying these optimized-thickness coatings on a Mylar film were developed that were carried out at room temperature without any specialized equipment and that can be adapted to standard coating methods such as silk screen or ink jet printing. The performance characteristics of the dual-chamber/dual-anode neutron detector were determined for both types of isotopically enriched converters. The experimental performance of the 6LiF converter-based detector was described well by modeling results from Geant4. Additional modeling studies of multiple-foil/multiple-chamber/anode configurations addressed the basic issue of the relatively longer absorption range of neutrons versus the shorter range of the conversion products for 6LiF and 10B. Combined with the experimental results, these simulations indicate that a high-performance neutron detector can be realized in a single device through the application of these multiple-foil/solid converter, multiple-chamber detector concepts.« less

Evaluation of tritium release properties of advanced tritium breeders

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hoshino, T.; Ochiai, K.; Edao, Y.

2015-03-15

Demonstration power plant (DEMO) fusion reactors require advanced tritium breeders with high thermal stability. Lithium titanate (Li{sub 2}TiO{sub 3}) advanced tritium breeders with excess Li (Li{sub 2+x}TiO{sub 3+y}) are stable in a reducing atmosphere at high temperatures. Although the tritium release properties of tritium breeders are documented in databases for DEMO blanket design, no in situ examination under fusion neutron (DT neutron) irradiation has been performed. In this study, a preliminary examination of the tritium release properties of advanced tritium breeders was performed, and DT neutron irradiation experiments were performed at the fusion neutronics source (FNS) facility in JAEA. Consideringmore » the tritium release characteristics, the optimum grain size after sintering is <5 μm. From the results of the optimization of granulation conditions, prototype Li{sub 2+x}TiO{sub 3+y} pebbles with optimum grain size (<5 μm) were successfully fabricated. The Li{sub 2+x}TiO{sub 3+y} pebbles exhibited good tritium release properties similar to the Li{sub 2}TiO{sub 3} pebbles. In particular, the released amount of HT gas for easier tritium handling was higher than that of HTO water. (authors)« less

A comparison of back propagation and Generalized Regression Neural Networks performance in neutron spectrometry.

PubMed

Martínez-Blanco, Ma Del Rosario; Ornelas-Vargas, Gerardo; Solís-Sánchez, Luis Octavio; Castañeda-Miranada, Rodrigo; Vega-Carrillo, Héctor René; Celaya-Padilla, José M; Garza-Veloz, Idalia; Martínez-Fierro, Margarita; Ortiz-Rodríguez, José Manuel

2016-11-01

The process of unfolding the neutron energy spectrum has been subject of research for many years. Monte Carlo, iterative methods, the bayesian theory, the principle of maximum entropy are some of the methods used. The drawbacks associated with traditional unfolding procedures have motivated the research of complementary approaches. Back Propagation Neural Networks (BPNN), have been applied with success in neutron spectrometry and dosimetry domains, however, the structure and learning parameters are factors that highly impact in the networks performance. In ANN domain, Generalized Regression Neural Network (GRNN) is one of the simplest neural networks in term of network architecture and learning algorithm. The learning is instantaneous, requiring no time for training. Opposite to BPNN, a GRNN would be formed instantly with just a 1-pass training on the development data. In the network development phase, the only hurdle is to optimize the hyper-parameter, which is known as sigma, governing the smoothness of the network. The aim of this work was to compare the performance of BPNN and GRNN in the solution of the neutron spectrometry problem. From results obtained it can be observed that despite the very similar results, GRNN performs better than BPNN. Copyright © 2016 Elsevier Ltd. All rights reserved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lafleur, Adrienne M.; Ulrich, Timothy J. II; Menlove, Howard O.

Objective is to investigate the use of Passive Neutron Albedo Reactivity (PNAR) and Self-Interrogation Neutron Resonance Densitometry (SINRD) to quantify fissile content in FUGEN spent fuel assemblies (FAs). Methodology used is: (1) Detector was designed using fission chambers (FCs); (2) Optimized design via MCNPX simulations; and (3) Plan to build and field test instrument in FY13. Significance was to improve safeguards verification of spent fuel assemblies in water and increase sensitivity to partial defects. MCNPX simulations were performed to optimize the design of the SINRD+PNAR detector. PNAR ratio was less sensitive to FA positioning than SINRD and SINRD ratio wasmore » more sensitive to Pu fissile mass than PNAR. Significance was that the integration of these techniques can be used to improve verification of spent fuel assemblies in water.« less

Neutron counter based on beryllium activation

NASA Astrophysics Data System (ADS)

Bienkowska, B.; Prokopowicz, R.; Scholz, M.; Kaczmarczyk, J.; Igielski, A.; Karpinski, L.; Paducha, M.; Pytel, K.

2014-08-01

The fusion reaction occurring in DD plasma is followed by emission of 2.45 MeV neutrons, which carry out information about fusion reaction rate and plasma parameters and properties as well. Neutron activation of beryllium has been chosen for detection of DD fusion neutrons. The cross-section for reaction 9Be(n, α)6He has a useful threshold near 1 MeV, which means that undesirable multiple-scattered neutrons do not undergo that reaction and therefore are not recorded. The product of the reaction, 6He, decays with half-life T1/2 = 0.807 s emitting β- particles which are easy to detect. Large area gas sealed proportional detector has been chosen as a counter of β-particles leaving activated beryllium plate. The plate with optimized dimensions adjoins the proportional counter entrance window. Such set-up is also equipped with appropriate electronic components and forms beryllium neutron activation counter. The neutron flux density on beryllium plate can be determined from the number of counts. The proper calibration procedure needs to be performed, therefore, to establish such relation. The measurements with the use of known β-source have been done. In order to determine the detector response function such experiment have been modeled by means of MCNP5-the Monte Carlo transport code. It allowed proper application of the results of transport calculations of β- particles emitted from radioactive 6He and reaching proportional detector active volume. In order to test the counter system and measuring procedure a number of experiments have been performed on PF devices. The experimental conditions have been simulated by means of MCNP5. The correctness of simulation outcome have been proved by measurements with known radioactive neutron source. The results of the DD fusion neutron measurements have been compared with other neutron diagnostics.

Commissioning optically stimulated luminescence in vivo dosimeters for fast neutron therapy.

PubMed

Young, Lori A; Yang, Fei; Woodworth, Davis; McCormick, Zephyr; Sandison, George

2016-01-01

Clinical in vivo dosimeters intended for use with photon and electron therapies have not been utilized for fast neutron therapy because they are highly susceptible to neutron damage. The objective of this work was to determine if a commercial optically stimulated luminescence (OSL) in vivo dosimetry system could be adapted for use in fast neutron therapy. A 50.5 MeV fast neutron beam generated by a clinical neutron therapy cyclotron was used to irradiate carbon doped aluminum oxide (Al2O3:C) optically simulated luminescence dosimeters (OSLDs) in a solid water phantom under standard calibration conditions, 150 cm SAD, 1.7 cm depth, and 10.3 × 10.0 cm field size. OSLD fading and electron trap depletion studies were performed with the OSLDs irradiated with 20 and 50 cGy and monitored over a 24-h period to determine the optimal time for reading the dosimeters during calibration. Four OSLDs per group were calibrated over a clinical dose range of 0-150 cGy. OSLD measurement uncertainties were lowered to within ±2%-3% of the expected dose by minimizing the effect of transient fading that occurs with neutron irradiation and maintaining individual calibration factors for each dosimeter. Dose dependent luminescence fading extended beyond the manufacturer's recommended 10 min period for irradiation with photon or electron beams. To minimize OSL variances caused by inconsistent fading among dosimeters, the observed optimal time for reading the OSLDs postirradiation was between 30 and 90 min. No field size, wedge factor, or gantry angle dependencies were observed in the OSLDs irradiated by the studied fast neutron beam. Measurements demonstrated that uncertainties less than ±3% were attainable in OSLDs irradiated with fast neutrons under clinical conditions. Accuracy and precision comparable to clinical OSL measurements observed with photons can be achieved by maintaining individual OSLD calibration factors and minimizing transient fading effects.

Commissioning optically stimulated luminescence in vivo dosimeters for fast neutron therapy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Young, Lori A., E-mail: layoung@uw.edu; Sandison, George; Yang, Fei

Purpose: Clinical in vivo dosimeters intended for use with photon and electron therapies have not been utilized for fast neutron therapy because they are highly susceptible to neutron damage. The objective of this work was to determine if a commercial optically stimulated luminescence (OSL) in vivo dosimetry system could be adapted for use in fast neutron therapy. Methods: A 50.5 MeV fast neutron beam generated by a clinical neutron therapy cyclotron was used to irradiate carbon doped aluminum oxide (Al{sub 2}O{sub 3}:C) optically simulated luminescence dosimeters (OSLDs) in a solid water phantom under standard calibration conditions, 150 cm SAD, 1.7more » cm depth, and 10.3 × 10.0 cm field size. OSLD fading and electron trap depletion studies were performed with the OSLDs irradiated with 20 and 50 cGy and monitored over a 24-h period to determine the optimal time for reading the dosimeters during calibration. Four OSLDs per group were calibrated over a clinical dose range of 0–150 cGy. Results: OSLD measurement uncertainties were lowered to within ±2%–3% of the expected dose by minimizing the effect of transient fading that occurs with neutron irradiation and maintaining individual calibration factors for each dosimeter. Dose dependent luminescence fading extended beyond the manufacturer’s recommended 10 min period for irradiation with photon or electron beams. To minimize OSL variances caused by inconsistent fading among dosimeters, the observed optimal time for reading the OSLDs postirradiation was between 30 and 90 min. No field size, wedge factor, or gantry angle dependencies were observed in the OSLDs irradiated by the studied fast neutron beam. Conclusions: Measurements demonstrated that uncertainties less than ±3% were attainable in OSLDs irradiated with fast neutrons under clinical conditions. Accuracy and precision comparable to clinical OSL measurements observed with photons can be achieved by maintaining individual OSLD calibration factors and minimizing transient fading effects.« less

Experimental evidence of impact ignition: 100-fold increase of neutron yield by impactor collision.

PubMed

Azechi, H; Sakaiya, T; Watari, T; Karasik, M; Saito, H; Ohtani, K; Takeda, K; Hosoda, H; Shiraga, H; Nakai, M; Shigemori, K; Fujioka, S; Murakami, M; Nagatomo, H; Johzaki, T; Gardner, J; Colombant, D G; Bates, J W; Velikovich, A L; Aglitskiy, Y; Weaver, J; Obenschain, S; Eliezer, S; Kodama, R; Norimatsu, T; Fujita, H; Mima, K; Kan, H

2009-06-12

We performed integrated experiments on impact ignition, in which a portion of a deuterated polystyrene (CD) shell was accelerated to about 600 km/s and was collided with precompressed CD fuel. The kinetic energy of the impactor was efficiently converted into thermal energy generating a temperature of about 1.6 keV. We achieved a two-order-of-magnitude increase in the neutron yield by optimizing the timing of the impact collision, demonstrating the high potential of impact ignition for fusion energy production.

Calculations to Support On-line Neutron Spectrum Adjustment by Measurements with Miniature Fission Chambers in the JSI TRIGA Reactor

NASA Astrophysics Data System (ADS)

Kaiba, Tanja; Radulović, Vladimir; Žerovnik, Gašper; Snoj, Luka; Fourmentel, Damien; Barbot, LoÏc; Destouches, Christophe AE(; )

2018-01-01

Preliminary calculations were performed with the aim to establish optimal experimental conditions for the measurement campaign within the collaboration between the Jožef Stefan Institute (JSI) and Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA Cadarache). The goal of the project is to additionally characterize the neutron spectruminside the JSI TRIGA reactor core with focus on the measurement epi-thermal and fast part of the spectrum. Measurements will be performed with fission chambers containing different fissile materials (235U, 237Np and 242Pu) covered with thermal neutron filters (Cd and Gd). The changes in the detected signal and neutron flux spectrum with and without transmission filter were studied. Additional effort was put into evaluation of the effect of the filter geometry (e.g. opening on the top end of the filter) on the detector signal. After the analysis of the scoping calculations it was concluded to position the experiment in the outside core ring inside one of the empty fuel element positions.

Optimal moderator materials at various proton energies considering photon dose rate after irradiation for an accelerator-driven ⁹Be(p, n) boron neutron capture therapy neutron source.

PubMed

Hashimoto, Y; Hiraga, F; Kiyanagi, Y

2015-12-01

We evaluated the accelerator beam power and the neutron-induced radioactivity of (9)Be(p, n) boron neutron capture therapy (BNCT) neutron sources having a MgF2, CaF2, or AlF3 moderator and driven by protons with energy from 8 MeV to 30 MeV. The optimal moderator materials were found to be MgF2 for proton energies less than 10 MeV because of lower required accelerator beam power and CaF2 for higher proton energies because of lower photon dose rate at the treatment position after neutron irradiation. Copyright © 2015 Elsevier Ltd. All rights reserved.

Monte Carlo calculation for the development of a BNCT neutron source (1eV-10KeV) using MCNP code.

PubMed

El Moussaoui, F; El Bardouni, T; Azahra, M; Kamili, A; Boukhal, H

2008-09-01

Different materials have been studied in order to produce the epithermal neutron beam between 1eV and 10KeV, which are extensively used to irradiate patients with brain tumors such as GBM. For this purpose, we have studied three different neutrons moderators (H(2)O, D(2)O and BeO) and their combinations, four reflectors (Al(2)O(3), C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation showed that the best obtained assembly configuration corresponds to the combination of the three moderators H(2)O, BeO and D(2)O jointly to Al(2)O(3) reflector and two filter Cd+Bi optimize the spectrum of the epithermal neutron at 72%, and minimize the thermal neutron to 4% and thus it can be used to treat the deep tumor brain. The calculations have been performed by means of the Monte Carlo N (particle code MCNP 5C). Our results strongly encourage further studying of irradiation of the head with epithermal neutron fields.

Neutron spectroscopy with scintillation detectors using wavelets

NASA Astrophysics Data System (ADS)

Hartman, Jessica

The purpose of this research was to study neutron spectroscopy using the EJ-299-33A plastic scintillator. This scintillator material provided a novel means of detection for fast neutrons, without the disadvantages of traditional liquid scintillation materials. EJ-299-33A provided a more durable option to these materials, making it less likely to be damaged during handling. Unlike liquid scintillators, this plastic scintillator was manufactured from a non-toxic material, making it safer to use, as well as easier to design detectors. The material was also manufactured with inherent pulse shape discrimination abilities, making it suitable for use in neutron detection. The neutron spectral unfolding technique was developed in two stages. Initial detector response function modeling was carried out through the use of the MCNPX Monte Carlo code. The response functions were developed for a monoenergetic neutron flux. Wavelets were then applied to smooth the response function. The spectral unfolding technique was applied through polynomial fitting and optimization techniques in MATLAB. Verification of the unfolding technique was carried out through the use of experimentally determined response functions. These were measured on the neutron source based on the Van de Graff accelerator at the University of Kentucky. This machine provided a range of monoenergetic neutron beams between 0.1 MeV and 24 MeV, making it possible to measure the set of response functions of the EJ-299-33A plastic scintillator detector to neutrons of specific energies. The response of a plutonium-beryllium (PuBe) source was measured using the source available at the University of Nevada, Las Vegas. The neutron spectrum reconstruction was carried out using the experimentally measured response functions. Experimental data was collected in the list mode of the waveform digitizer. Post processing of this data focused on the pulse shape discrimination analysis of the recorded response functions to remove the effects of photons and allow for source characterization based solely on the neutron response. The unfolding technique was performed through polynomial fitting and optimization techniques in MATLAB, and provided an energy spectrum for the PuBe source.

SECOND TARGET STATION MODERATOR PERFORMANCE WITH A ROTATING TARGET

DOE Office of Scientific and Technical Information (OSTI.GOV)

Remec, Igor; Gallmeier, Franz X; Rennich, Mark J

2016-01-01

Oak Ridge National Laboratory manages and operates the Spallation Neutron Source and the High Flux Isotope Reactor, two of the world's most advanced neutron scattering facilities. Both facilities are funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Science, and are available to researchers from all over the world. Delivering cutting edge science requires continuous improvements and development of the facilities and instruments. The SNS was designed from the outset to accommodate an additional target station, or Second Target Station (STS), and an upgraded accelerator feeding proton beams to STS and the existing First Targetmore » Station (FTS). Upgrade of the accelerator and the design and construction of STS are being proposed. The presently considered STS configuration is driven with short (<1 s) proton pulses at 10 Hz repetition rate and 467 kW proton beam power, and is optimized for high intensity and high resolution long wavelength neutron applications. STS will allow installation of 22 beamlines and will expand and complement the current national neutron scattering capabilities. In 2015 the STS studies were performed for a compact tungsten target; first a stationary tungsten plate target was analyzed to considerable details and then dropped in favor of a rotating target. For both target options the proton beam footprint as small as acceptable from mechanical and heat removal aspects is required to arrive at a compact-volume neutron production zone in the target, which is essential for tight coupling of target and moderators and for achieving high-intensity peak neutron fluxes. This paper will present recent STS work with the emphasis on neutronics and moderator performance.« less

Measuring and monitoring KIPT Neutron Source Facility Reactivity

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cao, Yan; Gohar, Yousry; Zhong, Zhaopeng

2015-08-01

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on developing and constructing a neutron source facility at Kharkov, Ukraine. The facility consists of an accelerator-driven subcritical system. The accelerator has a 100 kW electron beam using 100 MeV electrons. The subcritical assembly has k eff less than 0.98. To ensure the safe operation of this neutron source facility, the reactivity of the subcritical core has to be accurately determined and continuously monitored. A technique which combines the area-ratio method and the flux-to-current ratio method is purposed to determine themore » reactivity of the KIPT subcritical assembly at various conditions. In particular, the area-ratio method can determine the absolute reactivity of the subcritical assembly in units of dollars by performing pulsed-neutron experiments. It provides reference reactivities for the flux-to-current ratio method to track and monitor the reactivity deviations from the reference state while the facility is at other operation modes. Monte Carlo simulations are performed to simulate both methods using the numerical model of the KIPT subcritical assembly. It is found that the reactivities obtained from both the area-ratio method and the flux-to-current ratio method are spatially dependent on the neutron detector locations and types. Numerical simulations also suggest optimal neutron detector locations to minimize the spatial effects in the flux-to-current ratio method. The spatial correction factors are calculated using Monte Carlo methods for both measuring methods at the selected neutron detector locations. Monte Carlo simulations are also performed to verify the accuracy of the flux-to-current ratio method in monitoring the reactivity swing during a fuel burnup cycle.« less

Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

NASA Astrophysics Data System (ADS)

Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

2016-01-01

Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

Testing of Photomultiplier Tubes in a Magnetic Field

NASA Astrophysics Data System (ADS)

Waldron, Zachary; A1 Collaboration

2016-09-01

The A1 collaboration at MAMI in Mainz, Germany has designed a neutron detector that can be used in experiments to measure the electric form factor of the neutron. They will measure elastic scattering from the neutron, using the polarized electron beam from MAMI at A1's experimental hall. The detector will be composed of two walls of staggered scintillator bars which will be read out by photomultiplier tubes (PMT), connected to both ends of each scintillator via light guides. The experiment requires a magnetic field with strength of 1 Tesla, 2m away from the first scintillator wall. The resulting fringe field is sufficient to disrupt the PMTs, despite the addition of Mu Metal shielding. The effects of the fringe field on these PMTs was tested to optimize the amplification of the PMTs. A Helmholtz Coil was designed to generate a controlled magnetic field with equivalent strength to the field that the PMTs will encounter. The PMTs were read out using a multi-channel analyzer, were tested at various angles relative to the magnetic field in order to determine the optimal orientation to minimize signal disruption. Tests were also performed to determine: the neutron detector response to cosmic radiation; and the best method for measuring a magnetic field's strength in two dimensions. National Science Foundation Grant No. IIA-1358175.

Silicon Photomultipliers for Compact Neutron Scatter Cameras

NASA Astrophysics Data System (ADS)

Ruch, Marc L.

The ability to locate and identify special nuclear material (SNM) is critical for treaty verification and emergency response applications. SNM is used as the nuclear explosive in a nuclear weapon. This material emits neutrons, either spontaneously or when interrogated. The ability to form an image of the neutron source can be used for characterization and/or to confirm that the item is a weapon by determining whether its shape is consistent with that of a weapon. Additionally, treaty verification and emergency response applications might not be conducive to non-portable instruments. In future weapons treaties, for example, it is unlikely that host countries will make great efforts to facilitate large, bulky, and/or fragile inspection equipment. Furthermore, inspectors and especially emergency responders may need to access locations not easily approachable by vehicles. Therefore, there is a considerable need for a compact, human-portable neutron imaging system. Of the currently available neutron imaging technologies, only neutron scatter cameras (NSCs) can be made truly compact because aperture-based imagers, and time-encoded imagers, rely on large amounts of materials to modulate the neutron signal. NSCs, in contrast, can be made very small because most of the volume of the imager can be filled with active detector material. Also, unlike other neutron imaging technologies, NSCs have the inherent ability to act as neutron spectrometers which gives them an additional means of identifying a neutron source. Until recently, NSCs have relied on photomultiplier tubes (PMT) readouts, which are bulky and fragile, require high voltage, and are very sensitive to magnetic fields. Silicon photomultipliers (SiPMs) do not suffer from these drawbacks and are comparable to PMTs in many respects such as gain, and cost with better time resolution. Historically, SiPMs have been too noisy for these applications; however, recent advancements have greatly reduced this issue and they have now been shown to be viable alternatives to PMTs for neutron detection applications. In this thesis, the development of a handheld NSC based on SiPMs coupled to stilbene bars is presented. An algorithm for performing image reconstruction with this type of device is detailed. Prototype design optimization is achieved using a series of simulations and the construction of the optimized prototype is described. The device is calibrated through a series of collimated measurements, backscatter-gated measurements, and a time-of-flight measurement. Experimental imaging and spectroscopic results are presented for a measurement of a Cf-252 spontaneous fission source. Simulated detector response, based on measurements performed with components of the design, demonstrates that fission sources of different sizes would be distinguishable. Notably, a significant quantity of plutonium can be confidently distinguished from a point neutron source.

Optimization of beam shaping assembly based on D-T neutron generator and dose evaluation for BNCT

NASA Astrophysics Data System (ADS)

Naeem, Hamza; Chen, Chaobin; Zheng, Huaqing; Song, Jing

2017-04-01

The feasibility of developing an epithermal neutron beam for a boron neutron capture therapy (BNCT) facility based on a high intensity D-T fusion neutron generator (HINEG) and using the Monte Carlo code SuperMC (Super Monte Carlo simulation program for nuclear and radiation process) is proposed in this study. The Monte Carlo code SuperMC is used to determine and optimize the final configuration of the beam shaping assembly (BSA). The optimal BSA design in a cylindrical geometry which consists of a natural uranium sphere (14 cm) as a neutron multiplier, AlF3 and TiF3 as moderators (20 cm each), Cd (1 mm) as a thermal neutron filter, Bi (5 cm) as a gamma shield, and Pb as a reflector and collimator to guide neutrons towards the exit window. The epithermal neutron beam flux of the proposed model is 5.73 × 109 n/cm2s, and other dosimetric parameters for the BNCT reported by IAEA-TECDOC-1223 have been verified. The phantom dose analysis shows that the designed BSA is accurate, efficient and suitable for BNCT applications. Thus, the Monte Carlo code SuperMC is concluded to be capable of simulating the BSA and the dose calculation for BNCT, and high epithermal flux can be achieved using proposed BSA.

ALTERNATIVES TO HELIUM-3 FOR NEUTRON MULTIPLICITY DETECTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ely, James H.; Siciliano, Edward R.; Swinhoe, Martyn T.

Collaboration between the Pacific Northwest National Laboratory (PNNL) and the Los Alamos National Laboratory (LANL) is underway to evaluate neutron detection technologies that might replace the high-pressure helium (3He) tubes currently used in neutron multiplicity counter for safeguards applications. The current stockpile of 3He is diminishing and alternatives are needed for a variety of neutron detection applications including multiplicity counters. The first phase of this investigation uses a series of Monte Carlo calculations to simulate the performance of an existing neutron multiplicity counter configuration by replacing the 3He tubes in a model for that counter with candidate alternative technologies. Thesemore » alternative technologies are initially placed in approximately the same configuration as the 3He tubes to establish a reference level of performance against the 3He-based system. After these reference-level results are established, the configurations of the alternative models will be further modified for performance optimization. The 3He model for these simulations is the one used by LANL to develop and benchmark the Epithermal Neutron Multiplicity Counter (ENMC) detector, as documented by H.O. Menlove, et al. in the 2004 LANL report LA-14088. The alternative technologies being evaluated are the boron-tri-fluoride-filled proportional tubes, boron-lined tubes, and lithium coated materials previously tested as possible replacements in portal monitor screening applications, as documented by R.T. Kouzes, et al. in the 2010 PNNL report PNNL-72544 and NIM A 623 (2010) 1035–1045. The models and methods used for these comparative calculations will be described and preliminary results shown« less

Hierarchical optimization for neutron scattering problems

DOE PAGES

Bao, Feng; Archibald, Rick; Bansal, Dipanshu; ...

2016-03-14

In this study, we present a scalable optimization method for neutron scattering problems that determines confidence regions of simulation parameters in lattice dynamics models used to fit neutron scattering data for crystalline solids. The method uses physics-based hierarchical dimension reduction in both the computational simulation domain and the parameter space. We demonstrate for silicon that after a few iterations the method converges to parameters values (interatomic force-constants) computed with density functional theory simulations.

Hierarchical optimization for neutron scattering problems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bao, Feng; Archibald, Rick; Bansal, Dipanshu

In this study, we present a scalable optimization method for neutron scattering problems that determines confidence regions of simulation parameters in lattice dynamics models used to fit neutron scattering data for crystalline solids. The method uses physics-based hierarchical dimension reduction in both the computational simulation domain and the parameter space. We demonstrate for silicon that after a few iterations the method converges to parameters values (interatomic force-constants) computed with density functional theory simulations.

Correlation between radiation dose and histopathological findings in patients with gliblastoma treated with boron neutron capture therapy (BNCT).

PubMed

Kageji, T; Mizobuchi, Y; Nagahiro, S; Nakagawa, Y; Kumada, H

2014-06-01

The purpose of this study was to clarify the correlation between the radiation dose and histopathological findings in patients with glioblastoma multiforme (GBM) treated with boron neutron capture therapy (BNCT). Histopathological studies were performed on specimens from 8 patients, 3 had undergone salvage surgery and 5 were autopsied. For histopathological cure of GBM at the primary site, the optimal minimal dose to the gross tumor volume (GTV) and the clinical target volume (CTV) were 68Gy(w) and 44Gy(w), respectively. Copyright © 2014. Published by Elsevier Ltd.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ashenfelter, J.; Jaffe, D.; Diwan, M. V.

A meter-long, 23-liter EJ-309 liquid scintillator detector has been constructed to study the light collection and pulse-shape discrimination performance of elongated scintillator cells for the PROSPECT reactor antineutrino experiment. The magnitude and uniformity of light collection and neutron-gamma discrimination power in the energy range of antineutrino inverse beta decay products have been studied using gamma and spontaneous fission calibration sources deployed along the cell axis. We also study neutron-gamma discrimination and light collection abilities for differing PMT and reflector configurations. As a result, key design features for optimizing MeV-scale response and background rejection capabilities are identified.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osovizky, A.; Rotem Industries Ltd, Rotem Industrial Park; University of Maryland, College park, Maryland

A Chromatic Analysis Neutron Diffractometer Or Reflectometer (CANDOR) is under development at the NIST Center for Neutron Research (NCNR). The CANDOR neutron sensor will rely on scintillator material for detecting the neutrons scattered by the sample under test. It consists of {sup 6}LiF:ZnS(Ag) scintillator material into which wavelength shifting (WLS) fibers have been embedded. Solid state photo-sensors (silicon photomultipliers) coupled to the WLS fibers are used to detect the light produced by the neutron capture event ({sup 6}Li (n,α) {sup 3}H reaction) and ionization of the ZnS(Ag). This detector configuration has the potential to accomplish the CANDOR performance requirements formore » efficiency of 90% for 5 A (3.35 meV) neutrons with high gamma rejection (10{sup 7}) along with compact design, affordable cost and materials availability. However this novel design includes challenges for precise neutron detection. The recognizing of the neutron signature versus the noise event produce by gamma event cannot be easy overcome by pulse height discrimination obstacle as can be achieved with {sup 3}He gas tube. Furthermore the selection of silicon photomultipliers (SiPM) as the light sensor maintains the obstacle of dark noise that does not exist when a photomultiplier tube is coupled to the scintillator. A proper selection of SiPM should focus on increasing the output signal and reducing the dark noise in order to optimize the detection sensitivity and to provide a clean signal pulse shape discrimination. The main parameters for evaluation are: - Quantum Efficiency (QE) - matching the SiPM peak QE with the peak transmission wavelength emission of the WLS. - Recovery time - a short recovery time is preferred to minimize the pulse width beyond the intrinsic decay time of the scintillator crystal (improves the gamma rejection based output pulse shape (time)). - Diode dimensions -The dark noise is proportional to the diode active area while the signal is provided by the WLS fibers; therefore the diode area should ideally be only minimally larger than fiber bundle area. - Low dark noise - it is desirable to minimize the dark noise during the pulse integration period so as to minimize the background for pulse shape discrimination. - Photon Detection Efficiency - it is desirable to increase the SiPM PDE in order to enhance light collection. This will increase the likelihood of detecting neutron events with lower light production and will present a cleaner raw signal for pulse shape discrimination. We will present the SiPM optimization process and studies of dark noise and gamma and neutron sensitivity as a function of bias voltage and operating temperature that have enabled us to optimize the detector sensitivity and gamma rejection. The gamma rejection performance goal requires to overcome the challenge of discriminating between the light signature accepted by neutron event to the one received by the noise. In addition there is a huge variation between the number of light photons that reaching the WLS fibers for different neutron events caused by the heavy ions energy losses prior to ionizing the ZnS(Ag) and the high light attenuation of the scintillation mixture. This variation in the light signal along with the long decay time of the ZnS(Ag) (tens of microseconds) can cause double counting of the same neutron event in the case of high light output signature or preventing the detection of low sequential light output signature neutron event. We will presents the algorithm developed for {sup 6}LiF:ZnS(Ag) sensor readout and the results achieved by an off-line analysis by Matlab software code that successfully achieved both the high gamma rejection with a sensitive and accurate neutron event detection. (authors)« less

Target design optimization for an electron accelerator driven subcritical facility with circular and square beam profiles.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gohar, M. Y. A; Sofu, T.; Zhong, Z.

2008-10-30

A subcritical facility driven by an electron accelerator is planned at the Kharkov Institute of Physics and Technology (KIPT) in Ukraine for medical isotope production, materials research, training, and education. The conceptual design of the facility is being pursued through collaborations between ANL and KIPT. As part of the design effort, the high-fidelity analyses of various target options are performed with formulations to reflect the realistic configuration and the three dimensional geometry of each design. This report summarizes the results of target design optimization studies for electron beams with two different beam profiles. The target design optimization is performed viamore » the sequential neutronic, thermal-hydraulic, and structural analyses for a comprehensive assessment of each configuration. First, a target CAD model is developed with proper emphasis on manufacturability to provide a basis for separate but consistent models for subsequent neutronic, thermal-hydraulic, and structural analyses. The optimizations are pursued for maximizing the neutron yield, streamlining the flow field to avoid hotspots, and minimizing the thermal stresses to increase the durability. In addition to general geometric modifications, the inlet/outlet channel configurations, target plate partitioning schemes, flow manipulations and rates, electron beam diameter/width options, and cladding material choices are included in the design optimizations. The electron beam interactions with the target assembly and the neutronic response of the subcritical facility are evaluated using the MCNPX code. the results for the electron beam energy deposition, neutron generation, and utilization in the subcritical pile are then used to characterize the axisymmetric heat generation profiles in the target assembly with explicit simulations of the beam tube, the coolant, the clad, and the target materials. Both tungsten and uranium are considered as target materials. Neutron spectra from tungsten and uranium are very similar allowing the use of either material in the subcritical assembly without changing its characteristics. However, the uranium target has a higher neutron yield, which increases the neutron flux of the subcritical assembly. Based on the considered dimensions and heat generation profiles, the commercial CFD software Star-CD is used for the thermal-hydraulic analysis of each target design to satisfy a set of thermal criteria, the most limiting of which being to maintain the water temperature 50 below the boiling point. It is found that the turbulence in the inlet channels dissipates quickly in narrow gaps between the target plates and, as a result, the heat transfer is limited by the laminar flow conditions. On average, 3-D CFD analyses of target assemblies agree well with 1-D calculations using RELAP (performed by KIPT). However, the recirculation and stagnation zones predicted with the CFD models prove the importance of a 3-D analysis to avoid the resulting hotspots. The calculated temperatures are subsequently used for the structural analysis of each target configuration to satisfy the other engineering design requirements. The thermo-structural calculations are performed mostly with NASTRAN and the results occasionally compared with the results from MARC. Both, NASTRAN and MARC are commercially available structural-mechanics analysis software. Although, a significant thermal gradient forms in target elements along the beam direction, the high thermal stresses are generally observed peripherally around the edge of thin target disks/plates. Due to its high thermal conductivity, temperatures and thermal stresses in tungsten target are estimated to be significantly lower than in uranium target. The deformations of the target disks/plates are found to be insignificant, which eliminate concerns for flow blockages in narrow coolant channels. Consistent with the specifications of the KIPT accelerator to be used in this facility, the electron beam power is 100-kW with electron energy in the range of 100 to 200 MeV. As expected, the 100 MeV electrons deposit their energy faster while the 200-MeV electrons spread their energy deposition further along the beam direction. However in that electron energy range, the energy deposition profiles near the beam window require very thin target plates/disks to limit the temperatures and thermal stresses.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Budden, B. S.; Stonehill, L. C.; Warniment, A.

In this study, a new class of elpasolite scintillators has garnered recent attention due to the ability to perform as simultaneous gamma spectrometers and thermal neutron detectors. Such a dual-mode capability is made possible by pulse-shape discrimination (PSD), whereby the emission waveform profiles of gamma and neutron events are fundamentally unique. To take full advantage of these materials, we have developed the Compact Advanced Readout Electronics for Elpasolites (CAREE). This handheld instrument employs a multi-channel PSD-capable ASIC, custom micro-processor board, front-end electronics, power supplies, and a 2 in. photomultiplier tube for readout of the scintillator. The unit is highly configurablemore » to allow for performance optimization amongst a wide sample of elpasolites which provide PSD in fundamentally different ways. We herein provide an introduction to elpasolites, then describe the motivation for the work, mechanical and electronic design, and preliminary performance results.« less

Very High Reflectivity Supermirrors And Their Applications

NASA Astrophysics Data System (ADS)

Mezei, F.

1989-01-01

Very high reflectivity (some 95 % or better) supermirrors, with cut-off angles up to 2 times the critical angle of Ni coated simple total reflection neutron mirrors, can be produced using well established conventional deposition techniques. This performance makes applications involving multiple reflections and transmission geometries feasible, which in turn allow us to use more sophisticated neutron optical systems in order to optimize performance and minimize the amount a scarce supermirrors required. A key feature of several of these novel systems is the distribution of tasks between the several optical components achieving the desired performance by multiple action. The design and characteristics of a series of novel applications, such as polarizing cavities, collimators and guides, non-polarizing guides, beam compressors, deflectors and splitters (most of them tested or being implemented) are the main subjects of the present paper.

Characteristics comparison between a cyclotron-based neutron source and KUR-HWNIF for boron neutron capture therapy

NASA Astrophysics Data System (ADS)

Tanaka, H.; Sakurai, Y.; Suzuki, M.; Masunaga, S.; Kinashi, Y.; Kashino, G.; Liu, Y.; Mitsumoto, T.; Yajima, S.; Tsutsui, H.; Maruhashi, A.; Ono, K.

2009-06-01

At Kyoto University Research Reactor Institute (KURRI), 275 clinical trials of boron neutron capture therapy (BNCT) have been performed as of March 2006, and the effectiveness of BNCT has been revealed. In order to further develop BNCT, it is desirable to supply accelerator-based epithermal-neutron sources that can be installed near the hospital. We proposed the method of filtering and moderating fast neutrons, which are emitted from the reaction between a beryllium target and 30-MeV protons accelerated by a cyclotron accelerator, using an optimum moderator system composed of iron, lead, aluminum and calcium fluoride. At present, an epithermal-neutron source is under construction from June 2008. This system consists of a cyclotron accelerator, beam transport system, neutron-yielding target, filter, moderator and irradiation bed. In this article, an overview of this system and the properties of the treatment neutron beam optimized by the MCNPX Monte Carlo neutron transport code are presented. The distribution of biological effect weighted dose in a head phantom compared with that of Kyoto University Research Reactor (KUR) is shown. It is confirmed that for the accelerator, the biological effect weighted dose for a deeply situated tumor in the phantom is 18% larger than that for KUR, when the limit dose of the normal brain is 10 Gy-eq. The therapeutic time of the cyclotron-based neutron sources are nearly one-quarter of that of KUR. The cyclotron-based epithermal-neutron source is a promising alternative to reactor-based neutron sources for treatments by BNCT.

Evolution in boron-based GEM detectors for diffraction measurements: from planar to 3D converters

NASA Astrophysics Data System (ADS)

Albani, Giorgia; Perelli Cippo, Enrico; Croci, Gabriele; Muraro, Andrea; Schooneveld, Erik; Scherillo, Antonella; Hall-Wilton, Richard; Kanaki, Kalliopi; Höglund, Carina; Hultman, Lars; Birch, Jens; Claps, Gerardo; Murtas, Fabrizio; Rebai, Marica; Tardocchi, Marco; Gorini, Giuseppe

2016-11-01

The so-called ‘3He-crisis’ has motivated the neutron detector community to undertake an intense R&D programme in order to develop technologies alternative to standard 3He tubes and suitable for neutron detection systems in future spallation sources such as the European spallation source (ESS). Boron-based GEM (gas electron multiplier) detectors are a promising ‘3He-free’ technology for thermal neutron detection in neutron scattering experiments. In this paper the evolution of boron-based GEM detectors from planar to 3D converters with an application in diffraction measurements is presented. The use of 3D converters coupled with GEMs allows for an optimization of the detector performances. Three different detectors were used for diffraction measurements on the INES instrument at the ISIS spallation source. The performances of the GEM-detectors are compared with those of conventional 3He tubes installed on the INES instrument. The conceptual detector with the 3D converter used in this paper reached a count rate per unit area of about 25% relative to the currently installed 3He tube. Its timing resolution is similar and the signal-to-background ratio (S/B) is 2 times lower.

Indigenous design and development of multiPSD array for time of flight neutron spectrometer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Desai, Shraddha S., E-mail: ssdesai@barc.gov.in; Devan, Shylaja; Das, Amrita

Time of Flight neutron Spectrometer for inelastic neutron scattering studies is being commissioned at Dhruva reactor. Neutron pulse tagging is carried out using the Fermi chopper. Scattered neutrons at maximum possible scattered solid angle are needed to be detected along with the time information at detector. Precise and efficient detection system is essential for measurement of spatial and temporal distribution. Detection area is optimized to cover an angle of 70° in horizontal direction and 23° in vertical direction. Position sensitive detectors (PSDs) are designed with judicial use of precious {sup 3}He gas and a few PSDs with BF{sub 3} gasmore » as suitable alternative for {sup 3}He gas. An array of vertically arranged 1 m long, 50 PSDs, covers the arc length of 2.5 m and detection area 2.5 m{sup 2}. The design of the BF{sub 3} PSDs is supported by investigations on the gas purity, fill gas pressure, drift region, drift field and neutron absorption in cathode wall. Design details and performance of the PSDs are presented in the paper.« less

Neutron counter based on beryllium activation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bienkowska, B.; Prokopowicz, R.; Kaczmarczyk, J.

2014-08-21

The fusion reaction occurring in DD plasma is followed by emission of 2.45 MeV neutrons, which carry out information about fusion reaction rate and plasma parameters and properties as well. Neutron activation of beryllium has been chosen for detection of DD fusion neutrons. The cross-section for reaction {sup 9}Be(n, α){sup 6}He has a useful threshold near 1 MeV, which means that undesirable multiple-scattered neutrons do not undergo that reaction and therefore are not recorded. The product of the reaction, {sup 6}He, decays with half-life T{sub 1/2} = 0.807 s emitting β{sup −} particles which are easy to detect. Large areamore » gas sealed proportional detector has been chosen as a counter of β–particles leaving activated beryllium plate. The plate with optimized dimensions adjoins the proportional counter entrance window. Such set-up is also equipped with appropriate electronic components and forms beryllium neutron activation counter. The neutron flux density on beryllium plate can be determined from the number of counts. The proper calibration procedure needs to be performed, therefore, to establish such relation. The measurements with the use of known β–source have been done. In order to determine the detector response function such experiment have been modeled by means of MCNP5–the Monte Carlo transport code. It allowed proper application of the results of transport calculations of β{sup −} particles emitted from radioactive {sup 6}He and reaching proportional detector active volume. In order to test the counter system and measuring procedure a number of experiments have been performed on PF devices. The experimental conditions have been simulated by means of MCNP5. The correctness of simulation outcome have been proved by measurements with known radioactive neutron source. The results of the DD fusion neutron measurements have been compared with other neutron diagnostics.« less

Precision Neutron Time-of-Flight Detectors Provide Insight into NIF Implosion Dynamics

NASA Astrophysics Data System (ADS)

Schlossberg, David; Eckart, M. J.; Grim, G. P.; Hartouni, E. P.; Hatarik, R.; Moore, A. S.; Waltz, C. S.

2017-10-01

During inertial confinement fusion, higher-order moments of neutron time-of-flight (nToF) spectra can provide essential information for optimizing implosions. The nToF diagnostic suite at the National Ignition Facility (NIF) was recently upgraded to include novel, quartz Cherenkov detectors. These detectors exploit the rapid Cherenkov radiation process, in contrast with conventional scintillator decay times, to provide high temporal-precision measurements that support higher-order moment analyses. Preliminary measurements have been made on the NIF during several implosions and initial results are presented here. Measured line-of-sight asymmetries, for example in ion temperatures, will be discussed. Finally, advanced detector optimization is shown to advance accessible physics, with possibilities for energy discrimination, gamma source identification, and further reduction in quartz response times. Work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under contract DE-AC52-07NA27344.

Single-view 3D reconstruction of correlated gamma-neutron sources

DOE PAGES

Monterial, Mateusz; Marleau, Peter; Pozzi, Sara A.

2017-01-05

We describe a new method of 3D image reconstruction of neutron sources that emit correlated gammas (e.g. Cf- 252, Am-Be). This category includes a vast majority of neutron sources important in nuclear threat search, safeguards and non-proliferation. Rather than requiring multiple views of the source this technique relies on the source’s intrinsic property of coincidence gamma and neutron emission. As a result only a single-view measurement of the source is required to perform the 3D reconstruction. In principle, any scatter camera sensitive to gammas and neutrons with adequate timing and interaction location resolution can perform this reconstruction. Using a neutronmore » double scatter technique, we can calculate a conical surface of possible source locations. By including the time to a correlated gamma we further constrain the source location in three-dimensions by solving for the source-to-detector distance along the surface of said cone. As a proof of concept we applied these reconstruction techniques on measurements taken with the the Mobile Imager of Neutrons for Emergency Responders (MINER). Two Cf-252 sources measured at 50 and 60 cm from the center of the detector were resolved in their varying depth with average radial distance relative resolution of 26%. To demonstrate the technique’s potential with an optimized system we simulated the measurement in MCNPX-PoliMi assuming timing resolution of 200 ps (from 2 ns in the current system) and source interaction location resolution of 5 mm (from 3 cm). Furthermore, these simulated improvements in scatter camera performance resulted in radial distance relative resolution decreasing to an average of 11%.« less

Single-view 3D reconstruction of correlated gamma-neutron sources

DOE Office of Scientific and Technical Information (OSTI.GOV)

Monterial, Mateusz; Marleau, Peter; Pozzi, Sara A.

We describe a new method of 3D image reconstruction of neutron sources that emit correlated gammas (e.g. Cf- 252, Am-Be). This category includes a vast majority of neutron sources important in nuclear threat search, safeguards and non-proliferation. Rather than requiring multiple views of the source this technique relies on the source’s intrinsic property of coincidence gamma and neutron emission. As a result only a single-view measurement of the source is required to perform the 3D reconstruction. In principle, any scatter camera sensitive to gammas and neutrons with adequate timing and interaction location resolution can perform this reconstruction. Using a neutronmore » double scatter technique, we can calculate a conical surface of possible source locations. By including the time to a correlated gamma we further constrain the source location in three-dimensions by solving for the source-to-detector distance along the surface of said cone. As a proof of concept we applied these reconstruction techniques on measurements taken with the the Mobile Imager of Neutrons for Emergency Responders (MINER). Two Cf-252 sources measured at 50 and 60 cm from the center of the detector were resolved in their varying depth with average radial distance relative resolution of 26%. To demonstrate the technique’s potential with an optimized system we simulated the measurement in MCNPX-PoliMi assuming timing resolution of 200 ps (from 2 ns in the current system) and source interaction location resolution of 5 mm (from 3 cm). Furthermore, these simulated improvements in scatter camera performance resulted in radial distance relative resolution decreasing to an average of 11%.« less

Optimization and Characterization of a Novel Self Powered Solid State Neutron Detector

NASA Astrophysics Data System (ADS)

Clinton, Justin

There is a strong interest in detecting both the diversion of special nuclear material (SNM) from legitimate, peaceful purposes and the transport of illicit SNM across domestic and international borders and ports. A simple solid-state detector employs a planar solar-cell type p-n junction and a thin conversion layer that converts incident neutrons into detectable charged particles, such as protons, alpha-particles, and heavier ions. Although simple planar devices can act as highly portable, low cost detectors, they have historically been limited to relatively low detection efficiencies; ˜10% and ˜0.2% for thermal and fast detectors, respectively. To increase intrinsic detection efficiency, the incorporation of 3D microstructures into p-i-n silicon devices was proposed. In this research, a combination of existing and new types of detector microstructures were investigated; Monte Carlo models, based on analytical calculations, were constructed and characterized using the GEANT4 simulation toolkit. The simulation output revealed that an array of etched hexagonal holes arranged in a honeycomb pattern and filled with either enriched (99% 10B) boron or parylene resulted in the highest intrinsic detection efficiencies of 48% and 0.88% for thermal and fast neutrons, respectively. The optimal parameters corresponding to each model were utilized as the basis for the fabrication of several prototype detectors. A calibrated 252Cf spontaneous fission source was utilized to generate fast neutrons, while thermal neutrons were created by placing the 252Cf in an HDPE housing designed and optimized using the MCNP simulation software. Upon construction, thermal neutron calibration was performed via activation analysis of gold foils and measurements from a 6Li loaded glass scintillator. Experimental testing of the prototype detectors resulted in maximum intrinsic efficiencies of 4.5 and 0.12% for the thermal and fast devices, respectively. The prototype thermal device was filled with natural (19% 10B) boron; scaling the response to 99% 10B enriched boron resulted in an intrinsic efficiency of 22.5%, one of the highest results in the literature. A comparison of simulated and experimental detector responses demonstrated a high degree of correlation, validating the conceptual models.

Influence of beam efficiency through the patient-specific collimator on secondary neutron dose equivalent in double scattering and uniform scanning modes of proton therapy.

PubMed

Hecksel, D; Anferov, V; Fitzek, M; Shahnazi, K

2010-06-01

Conventional proton therapy facilities use double scattering nozzles, which are optimized for delivery of a few fixed field sizes. Similarly, uniform scanning nozzles are commissioned for a limited number of field sizes. However, cases invariably occur where the treatment field is significantly different from these fixed field sizes. The purpose of this work was to determine the impact of the radiation field conformity to the patient-specific collimator on the secondary neutron dose equivalent. Using a WENDI-II neutron detector, the authors experimentally investigated how the neutron dose equivalent at a particular point of interest varied with different collimator sizes, while the beam spreading was kept constant. The measurements were performed for different modes of dose delivery in proton therapy, all of which are available at the Midwest Proton Radiotherapy Institute (MPRI): Double scattering, uniform scanning delivering rectangular fields, and uniform scanning delivering circular fields. The authors also studied how the neutron dose equivalent changes when one changes the amplitudes of the scanned field for a fixed collimator size. The secondary neutron dose equivalent was found to decrease linearly with the collimator area for all methods of dose delivery. The relative values of the neutron dose equivalent for a collimator with a 5 cm diameter opening using 88 MeV protons were 1.0 for the double scattering field, 0.76 for rectangular uniform field, and 0.6 for the circular uniform field. Furthermore, when a single circle wobbling was optimized for delivery of a uniform field 5 cm in diameter, the secondary neutron dose equivalent was reduced by a factor of 6 compared to the double scattering nozzle. Additionally, when the collimator size was kept constant, the neutron dose equivalent at the given point of interest increased linearly with the area of the scanned proton beam. The results of these experiments suggest that the patient-specific collimator is a significant contributor to the secondary neutron dose equivalent to a distant organ at risk. Improving conformity of the radiation field to the patient-specific collimator can significantly reduce secondary neutron dose equivalent to the patient. Therefore, it is important to increase the number of available generic field sizes in double scattering systems as well as in uniform scanning nozzles.

Feasibility study of using laser-generated neutron beam for BNCT.

PubMed

Kasesaz, Y; Rahmani, F; Khalafi, H

2015-09-01

The feasibility of using a laser-accelerated proton beam to produce a neutron source, via (p,n) reaction, for Boron Neutron Capture Therapy (BNCT) applications has been studied by MCNPX Monte Carlo code. After optimization of the target material and its thickness, a Beam Shaping Assembly (BSA) has been designed and optimized to provide appropriate neutron beam according to the recommended criteria by International Atomic Energy Agency. It was found that the considered laser-accelerated proton beam can provide epithermal neutron flux of ∼2×10(6) n/cm(2) shot. To achieve an appropriate epithermal neutron flux for BNCT treatment, the laser must operate at repetition rates of 1 kHz, which is rather ambitious at this moment. But it can be used in some BNCT researches field such as biological research. Copyright © 2015 Elsevier Ltd. All rights reserved.

Focusing adaptive-optics for neutron spectroscopy at extreme conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simeoni, G. G., E-mail: ggsimeoni@outlook.com; Physics Department E13, Technical University of Munich, D-85748 Garching; Valicu, R. G.

2015-12-14

Neutron Spectroscopy employing extreme-conditions sample environments is nowadays a crucial tool for the understanding of fundamental scientific questions as well as for the investigation of materials and chemical-physical properties. For all these kinds of studies, an increased neutron flux over a small sample area is needed. The prototype of a focusing neutron guide component, developed and produced completely at the neutron source FRM II in Garching (Germany), has been installed at the time-of-flight (TOF) disc-chopper neutron spectrometer TOFTOF and came into routine-operation. The design is based on the compressed Archimedes' mirror concept for finite-size divergent sources. It represents a uniquemore » device combining the supermirror technology with Adaptive Optics, suitable for broad-bandwidth thermal-cold TOF neutron spectroscopy (here optimized for 1.4–10 Å). It is able to squeeze the beam cross section down to a square centimeter, with a more than doubled signal-to-background ratio, increased efficiency at high scattering angles, and improved symmetry of the elastic resolution function. We present a comparison between the simulated and measured beam cross sections, as well as the performance of the instrument within real experiments. This work intends to show the unprecedented opportunities achievable at already existing instruments, along with useful guidelines for the design and construction of next-generation neutron spectrometers.« less

Slow neutron total cross-section, transmission and reflection calculation for poly- and mono-NaCl and PbF2 crystals

NASA Astrophysics Data System (ADS)

Mansy, Muhammad S.; Adib, M.; Habib, N.; Bashter, I. I.; Morcos, H. N.; El-Mesiry, M. S.

2016-10-01

A detailed study about the calculation of total neutron cross-section, transmission and reflection from crystalline materials was performed. The developed computer code is approved to be sufficient for the required calculations, also an excellent agreement has been shown when comparing the code results with the other calculated and measured values. The optimal monochromator and filter parameters were discussed in terms of crystal orientation, mosaic spread, and thickness. Calculations show that 30 cm thick of PbF2 poly-crystal is an excellent cold neutron filter producing neutron wavelengths longer than 0.66 nm needed for the investigation of magnetic structure experiments. While mono-crystal filter PbF2 cut along its (1 1 1), having mosaic spread (η = 0.5°) and thickness 10 cm can only transmit thermal neutrons of the desired wavelengths and suppress epithermal and γ-rays forming unwanted background, when it is cooled to liquid nitrogen temperature. NaCl (2 0 0) and PbF2 (1 1 1) monochromator crystals having mosaic spread (η = 0.5°) and thickness 10 mm shows high neutron reflectivity for neutron wavelengths (λ = 0.114 nm and λ = 0.43 nm) when they used as a thermal and cold neutron monochromators respectively with very low contamination from higher order reflections.

Multi-source irradiation facility with improved space configuration for neutron activation analysis: Design optimization.

PubMed

Kotb, N A; Solieman, Ahmed H M; El-Zakla, T; Amer, T Z; Elmeniawi, S; Comsan, M N H

2018-05-01

A neutron irradiation facility consisting of six 241 Am-Be neutron sources of 30 Ci total activity and 6.6 × 10 7 n/s total neutron yield is designed. The sources are embedded in a cubic paraffin wax, which plays a dual role as both moderator and reflector. The sample passage and irradiation channel are represented by a cylindrical path of 5 cm diameter passing through the facility core. The proposed design yields a high degree of space symmetry and thermal neutron homogeneity within 98% of flux distribution throughout the irradiated spherical sample of 5 cm diameter. The obtained thermal neutron flux is 8.0 × 10 4 n/cm 2 .s over the sample volume, with thermal-to-fast and thermal-to-epithermal ratios of 1.20 and 3.35, respectively. The design is optimized for maximizing the thermal neutron flux at sample position using the MCNP-5 code. The irradiation facility is supposed to be employed principally for neutron activation analysis. Copyright © 2018 Elsevier Ltd. All rights reserved.

Optimizing the Construction of the A1 Collaboration Neutron Detector

NASA Astrophysics Data System (ADS)

Chinn, Edward; A1 Collaboration

2016-09-01

We report on the design and construction of a frame designed to optimize both the time efficiency and construction quality of the large scintillator elements These elements will be assembled to form a neutron detector for use by the A1 Collaboration at the Institute for Nuclear Physics in Mainz, Germany. The design had to provide adequate support for the 20 kg scintillator bars while gluing light guides and photomultiplier tubes to both sides of the bars using optical cement. The optical cement requires approximately 24 hours to dry and 100 bars have to be glued with this apparatus. To address each of these issues, several different prototypes were designed and reviewed. The selected apparatus minimized size to meet space constraints, with reduced material cost and provided the most time-efficient way to build the neutron detector. Once the schematic design was selected, we produced technical drawings in AutoDesk Inventor. Assembled the structure and completed gluing of the first batch of scintillators, in order to verify the performance. This apparatus was successful at producing high quality scintillators which were evaluated using cosmic rays. National Science Foundation Grant No. IIA-1358175.

Optimization of a mirror-based neutron source using differential evolution algorithm

NASA Astrophysics Data System (ADS)

Yurov, D. V.; Prikhodko, V. V.

2016-12-01

This study is dedicated to the assessment of capabilities of gas-dynamic trap (GDT) and gas-dynamic multiple-mirror trap (GDMT) as potential neutron sources for subcritical hybrids. In mathematical terms the problem of the study has been formulated as determining the global maximum of fusion gain (Q pl), the latter represented as a function of trap parameters. A differential evolution method has been applied to perform the search. Considered in all calculations has been a configuration of the neutron source with 20 m long distance between the mirrors and 100 MW heating power. It is important to mention that the numerical study has also taken into account a number of constraints on plasma characteristics so as to provide physical credibility of searched-for trap configurations. According to the results obtained the traps considered have demonstrated fusion gain up to 0.2, depending on the constraints applied. This enables them to be used either as neutron sources within subcritical reactors for minor actinides incineration or as material-testing facilities.

Handheld readout electronics to fully exploit the particle discrimination capabilities of elpasolite scintillators

DOE PAGES

Budden, B. S.; Stonehill, L. C.; Warniment, A.; ...

2015-06-10

In this study, a new class of elpasolite scintillators has garnered recent attention due to the ability to perform as simultaneous gamma spectrometers and thermal neutron detectors. Such a dual-mode capability is made possible by pulse-shape discrimination (PSD), whereby the emission waveform profiles of gamma and neutron events are fundamentally unique. To take full advantage of these materials, we have developed the Compact Advanced Readout Electronics for Elpasolites (CAREE). This handheld instrument employs a multi-channel PSD-capable ASIC, custom micro-processor board, front-end electronics, power supplies, and a 2 in. photomultiplier tube for readout of the scintillator. The unit is highly configurablemore » to allow for performance optimization amongst a wide sample of elpasolites which provide PSD in fundamentally different ways. We herein provide an introduction to elpasolites, then describe the motivation for the work, mechanical and electronic design, and preliminary performance results.« less

Handheld readout electronics to fully exploit the particle discrimination capabilities of elpasolite scintillators

NASA Astrophysics Data System (ADS)

Budden, B. S.; Stonehill, L. C.; Warniment, A.; Michel, J.; Storms, S.; Dallmann, N.; Coupland, D. D. S.; Stein, P.; Weller, S.; Borges, L.; Proicou, M.; Duran, G.; Kamto, J.

2015-09-01

A new class of elpasolite scintillators has garnered recent attention due to the ability to perform as simultaneous gamma spectrometers and thermal neutron detectors. Such a dual-mode capability is made possible by pulse-shape discrimination (PSD), whereby the emission waveform profiles of gamma and neutron events are fundamentally unique. To take full advantage of these materials, we have developed the Compact Advanced Readout Electronics for Elpasolites (CAREE). This handheld instrument employs a multi-channel PSD-capable ASIC, custom micro-processor board, front-end electronics, power supplies, and a 2 in. photomultiplier tube for readout of the scintillator. The unit is highly configurable to allow for performance optimization amongst a wide sample of elpasolites which provide PSD in fundamentally different ways. We herein provide an introduction to elpasolites, then describe the motivation for the work, mechanical and electronic design, and preliminary performance results.

Light collection and pulse-shape discrimination in elongated scintillator cells for the PROSPECT reactor antineutrino experiment

NASA Astrophysics Data System (ADS)

Ashenfelter, J.; Balantekin, B.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bowes, A.; Brodsky, J. P.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Commeford, K.; Davee, D.; Dean, D.; Deichert, G.; Diwan, M. V.; Dolinski, M. J.; Dolph, J.; Dwyer, D. A.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Goddard, B. W.; Green, M.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Jaffe, D. E.; Langford, T. J.; Littlejohn, B. R.; Martinez Caicedo, D. A.; McKeown, R. D.; Mendenhall, M. P.; Mueller, P.; Mumm, H. P.; Napolitano, J.; Neilson, R.; Norcini, D.; Pushin, D.; Qian, X.; Romero, E.; Rosero, R.; Saldana, L.; Seilhan, B. S.; Sharma, R.; Sheets, S.; Stemen, N. T.; Surukuchi, P. T.; Varner, R. L.; Viren, B.; Wang, W.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yeh, M.; Yen, Y. R.; Zangakis, G.; Zhang, C.; Zhang, X.

2015-11-01

A meter-long, 23-liter EJ-309 liquid scintillator detector has been constructed to study the light collection and pulse-shape discrimination performance of elongated scintillator cells for the PROSPECT reactor antineutrino experiment. The magnitude and uniformity of light collection and neutron-gamma discrimination power in the energy range of antineutrino inverse beta decay products have been studied using gamma and spontaneous fission calibration sources deployed along the cell axis. We also study neutron-gamma discrimination and light collection abilities for differing PMT and reflector configurations. Key design features for optimizing MeV-scale response and background rejection capabilities are identified.

Optimization of the Army’s Fast Neutron Moderator for Radiography

DTIC Science & Technology

2013-02-26

thermal neutron flux from a commercially available high-energy D-T neutron generator. This paper details the steps taken to increase exposure rates...experiment was to have increased thermal neutron flux rates and shorter exposure times than previously achieved. Additional technology developments...This reduced the thermalizing efficiency of the moderator at higher energies, resulted in a large loss of neutron flux at the image plane, and

Optimization of Shielding- Collimator Parameters for ING-27 Neutron Generator Using MCNP5

NASA Astrophysics Data System (ADS)

Hegazy, Aya Hamdy; Skoy, V. R.; Hossny, K.

2018-04-01

Neutron generators are now used in various fields. They produce only fast neutrons; D-D neutron generator produces 2.45 MeV neutrons and D-T produces 14.1 MeV neutrons. In order to optimize shielding-collimator parameters to achieve higher neutron flux at the investigated sample (The signal) with lower neutron and gamma rays flux at the area of the detectors, design iterations are widely used. This work was applied to ROMASHA setup, TANGRA project, FLNP, Joint Institute for Nuclear Research. The studied parameters were; (1) shielding-collimator material, (2) Distance between the shielding-collimator assembly first plate and center of the neutron beam, and (3) thickness of collimator sheets. MCNP5 was used to simulate ROMASHA setup after it was validated on the experimental results of irradiation of Carbon-12 sample for one hour to detect its 4.44 MeV characteristic gamma line. The ratio between the signal and total neutron flux that enters each detector was calculated and plotted, concluding that the optimum shielding-collimator assembly is Tungsten of 5 cm thickness for each plate, and a distance of 2.3 cm. Also, the ratio between the signal and total gamma rays flux was calculated and plotted for each detector, leading to the previous conclusion but the distance was 1 cm.

Accelerator-based epithermal neutron sources for boron neutron capture therapy of brain tumors.

PubMed

Blue, Thomas E; Yanch, Jacquelyn C

2003-01-01

This paper reviews the development of low-energy light ion accelerator-based neutron sources (ABNSs) for the treatment of brain tumors through an intact scalp and skull using boron neutron capture therapy (BNCT). A major advantage of an ABNS for BNCT over reactor-based neutron sources is the potential for siting within a hospital. Consequently, light-ion accelerators that are injectors to larger machines in high-energy physics facilities are not considered. An ABNS for BNCT is composed of: (1) the accelerator hardware for producing a high current charged particle beam, (2) an appropriate neutron-producing target and target heat removal system (HRS), and (3) a moderator/reflector assembly to render the flux energy spectrum of neutrons produced in the target suitable for patient irradiation. As a consequence of the efforts of researchers throughout the world, progress has been made on the design, manufacture, and testing of these three major components. Although an ABNS facility has not yet been built that has optimally assembled these three components, the feasibility of clinically useful ABNSs has been clearly established. Both electrostatic and radio frequency linear accelerators of reasonable cost (approximately 1.5 M dollars) appear to be capable of producing charged particle beams, with combinations of accelerated particle energy (a few MeV) and beam currents (approximately 10 mA) that are suitable for a hospital-based ABNS for BNCT. The specific accelerator performance requirements depend upon the charged particle reaction by which neutrons are produced in the target and the clinical requirements for neutron field quality and intensity. The accelerator performance requirements are more demanding for beryllium than for lithium as a target. However, beryllium targets are more easily cooled. The accelerator performance requirements are also more demanding for greater neutron field quality and intensity. Target HRSs that are based on submerged-jet impingement and the use of microchannels have emerged as viable target cooling options. Neutron fields for reactor-based neutron sources provide an obvious basis of comparison for ABNS field quality. This paper compares Monte Carlo calculations of neutron field quality for an ABNS and an idealized standard reactor neutron field (ISRNF). The comparison shows that with lithium as a target, an ABNS can create a neutron field with a field quality that is significantly better (by a factor of approximately 1.2, as judged by the relative biological effectiveness (RBE)-dose that can be delivered to a tumor at a depth of 6cm) than that for the ISRNF. Also, for a beam current of 10 mA, the treatment time is calculated to be reasonable (approximately 30 min) for the boron concentrations that have been assumed.

Boron-Coated Straw Collar for Uranium Neutron Coincidence Collar Replacement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Croft, Stephen; McElroy, Robert Dennis

The objective of this project was to design and optimize, in simulation space, an active neutron coincidence counter (or collar) using boron-coated straws (BCSs) as a non- 3He replacement to the Uranium Neutron Coincidence Collar (UNCL). UNCL has been used by the International Atomic Energy Agency (IAEA) and European Atomic Energy Community (Euratom) since the 1980s to verify the 235U content in fresh light water reactor fuel assemblies for safeguards purposes. This report documents the design and optimization of the BCS collar.

Modeling down-scattered neutron images from cryogenic fuel implosions at the National Ignition Facility

NASA Astrophysics Data System (ADS)

Raman, Kumar; Casey, Dan; Callahan, Debra; Clark, Dan; Fittinghoff, David; Grim, Gary; Hatchett, Steve; Hinkel, Denise; Jones, Ogden; Kritcher, Andrea; Seek, Scott; Suter, Larry; Merrill, Frank; Wilson, Doug

2016-10-01

In experiments with cryogenic deuterium-tritium (DT) fuel layers at the National Ignition Facility (NIF), an important technique for visualizing the stagnated fuel assembly is to image the 6-12 MeV neutrons created by scatters of the 14 MeV hotspot neutrons in the surrounding cold fuel. However, such down-scattered neutron images are difficult to interpret without a model of the fuel assembly, because of the nontrivial neutron kinematics involved in forming the images. For example, the dominant scattering modes are at angles other than forward scattering and the 14 MeV neutron fluence is not uniform. Therefore, the intensity patterns in these images usually do not correspond in a simple way to patterns in the fuel distribution, even for simple fuel distributions. We describe our efforts to model synthetic images from ICF design simulations with data from the National Ignition Campaign and after. We discuss the insight this gives, both to understand how well the models are predicting fuel asymmetries and to inform how to optimize the diagnostic for the types of fuel distributions being predicted. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

Neutron Radiography and Computed Tomography at Oak Ridge National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Raine, Dudley A. III; Hubbard, Camden R.; Whaley, Paul M.

1997-12-31

The capability to perform neutron radiography and computed tomography is being developed at Oak Ridge National Laboratory. The facility will be located at the High Flux Isotope Reactor (HFIR), which has the highest steady state neutron flux of any reactor in the world. The Monte Carlo N-Particle transport code (MCNP), versions 4A and 4B, has been used extensively in the design phase of the facility to predict and optimize the operating characteristics, and to ensure the safety of personnel working in and around the blockhouse. Neutrons are quite penetrating in most engineering materials and can be useful to detect internalmore » flaws and features. Hydrogen atoms, such as in a hydrocarbon fuel, lubricant or a metal hydride, are relatively opaque to neutron transmission. Thus, neutron based tomography or radiography is ideal to image their presence. The source flux also provides unparalleled flexibility for future upgrades, including real time radiography where dynamic processes can be observed. A novel tomography detector has been designed using optical fibers and digital technology to provide a large dynamic range for reconstructions. Film radiography is also available for high resolution imaging applications. This paper summarizes the results of the design phase of this facility and the potential benefits to science and industry.« less

Explosives (and other threats) detection using pulsed neutron interrogation and optimized detectors

NASA Astrophysics Data System (ADS)

Strellis, Dan A.; Elsalim, Mashal; Gozani, Tsahi

2011-06-01

We have previously reported results from a human-portable system using neutron interrogation to detect contraband and explosives. We summarized our methodology for distinguishing threat materials such as narcotics, C4, and mustard gas in the myriad of backgrounds present in the maritime environment. We are expanding our mission for the Domestic Nuclear Detection Office (DNDO) to detect Special Nuclear Material (SNM) through the detection of multiple fission signatures without compromising the conventional threat detection performance. This paper covers our initial investigations into using neutrons from compact pulsed neutron generators via the d(D,n)3He or d(T,n)α reactions with energies of ~2.5 and 14 MeV, respectively, for explosives (and other threats) detection along with a variety of gamma-ray detectors. Fast neutrons and thermal neutrons (after successive collisions) can stimulate the emission of various threat detection signatures. For explosives detection, element-specific gamma-ray signatures via the (n,n'γ) inelastic scattering reaction and the (n,'γ) thermal capture reaction are detected. For SNM, delayed gamma-rays following fission can be measured with the same detector. Our initial trade-off investigations of several gamma-ray detectors types (NaI, CsI, LaBr3, HPGe) for measuring gamma-ray signatures in a pulsed neutron environment for potential application in a human-portable active interrogation system are covered in this paper.

Application of Advanced Nuclear Emulsion Technique to Fusion Neutron Diagnostics

NASA Astrophysics Data System (ADS)

Nakayama, Y.; Tomita, H.; Morishima, K.; Yamashita, F.; Hayashi, S.; Cheon, MunSeong; Isobe, M.; Ogawa, K.; Naka, T.; Nakano, T.; Nakamura, M.; Kawarabayashi, J.; Iguchi, T.; Ochiai, K.

In order to measure the 2.5 MeV neutrons produced by DD nuclear fusion reactions, we have developed a compact neutron detector based on nuclear emulsion. After optimization of development conditions, we evaluated the response of the detector to an accelerator-based DD neutron source. The absolute efficiency at an energy of 2.5 MeV was estimated to be (4.1±0.2)×10-6 tracks/neutron.

Performance of coincidence-based PSD on LiF/ZnS Detectors for Multiplicity Counting

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robinson, Sean M.; Stave, Sean C.; Lintereur, Azaree

Abstract: Mass accountancy measurement is a nuclear nonproliferation application which utilizes coincidence and multiplicity counters to verify special nuclear material declarations. With a well-designed and efficient detector system, several relevant parameters of the material can be verified simultaneously. 6LiF/ZnS scintillating sheets may be used for this purpose due to a combination of high efficiency and short die-away times in systems designed with this material, but involve choices of detector geometry and exact material composition (e.g., the addition of Ni-quenching in the material) that must be optimized for the application. Multiplicity counting for verification of declared nuclear fuel mass involves neutronmore » detection in conditions where several neutrons arrive in a short time window, with confounding gamma rays. This paper considers coincidence-based Pulse-Shape Discrimination (PSD) techniques developed to work under conditions of high pileup, and the performance of these algorithms with different detection materials. Simulated and real data from modern LiF/ZnS scintillator systems are evaluated with these techniques and the relationship between the performance under pileup and material characteristics (e.g., neutron peak width and total light collection efficiency) are determined, to allow for an optimal choice of detector and material.« less

A novel neutron energy spectrum unfolding code using particle swarm optimization

NASA Astrophysics Data System (ADS)

Shahabinejad, H.; Sohrabpour, M.

2017-07-01

A novel neutron Spectrum Deconvolution using Particle Swarm Optimization (SDPSO) code has been developed to unfold the neutron spectrum from a pulse height distribution and a response matrix. The Particle Swarm Optimization (PSO) imitates the bird flocks social behavior to solve complex optimization problems. The results of the SDPSO code have been compared with those of the standard spectra and recently published Two-steps Genetic Algorithm Spectrum Unfolding (TGASU) code. The TGASU code have been previously compared with the other codes such as MAXED, GRAVEL, FERDOR and GAMCD and shown to be more accurate than the previous codes. The results of the SDPSO code have been demonstrated to match well with those of the TGASU code for both under determined and over-determined problems. In addition the SDPSO has been shown to be nearly two times faster than the TGASU code.

Micro-scale characterization of a CMOS-based neutron detector for in-phantom measurements in radiation therapy

NASA Astrophysics Data System (ADS)

Arbor, Nicolas; Higueret, Stephane; Husson, Daniel

2018-04-01

The CMOS sensor AlphaRad has been designed at the IPHC Strasbourg for real-time monitoring of fast and thermal neutrons over a full energy spectrum. Completely integrated, highly transparent to photons and optimized for low power consumption, this sensor offers very interesting characteristics for the study of internal neutrons in radiation therapy with anthropomorphic phantoms. However, specific effects related to the CMOS metal substructure and to the charge collection process of low energy particles must be carefully estimated before being used for medical applications. We present a detailed characterization of the AlphaRad chip in the MeV energy range using proton and alpha micro-beam experiments performed at the AIFIRA facility (CENBG, Bordeaux). Two-dimensional maps of the charge collection were carried out on a micro-metric scale to be integrated into a Geant4 Monte Carlo simulation of the system. The gamma rejection, as well as the fast and thermal neutrons separation, were studied using both simulation and experimental data. The results highlight the potential of a future system based on CMOS sensor for in-phantom neutron detection in radiation therapies.

Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bernard, J.A.

1989-09-01

This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power onmore » spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.« less

Low temperature magnetic structure of CeRhIn 5 by neutron diffraction on absorption-optimized samples

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fobes, David M.; Bauer, Eric Dietzgen; Thompson, Joe David

Here, two aspects of the ambient pressure magnetic structure of heavy fermion material CeRhIn 5 have remained under some debate since its discovery: whether the structure is indeed an incommensurate helix or a spin density wave, and what is the precise magnitude of the ordered magnetic moment. By using a single crystal sample optimized for hot neutrons to minimize neutron absorption by Rh and In, here we report an ordered moment ofmore » $$m=0.54(2)\\,{{\\mu}_{\\text{B}}}$$. In addition, by using spherical neutron polarimetry measurements on a similar single crystal sample, we have confirmed the helical nature of the magnetic structure, and identified a single chiral domain.« less

Low temperature magnetic structure of CeRhIn 5 by neutron diffraction on absorption-optimized samples

DOE PAGES

Fobes, David M.; Bauer, Eric Dietzgen; Thompson, Joe David; ...

2017-03-28

Here, two aspects of the ambient pressure magnetic structure of heavy fermion material CeRhIn 5 have remained under some debate since its discovery: whether the structure is indeed an incommensurate helix or a spin density wave, and what is the precise magnitude of the ordered magnetic moment. By using a single crystal sample optimized for hot neutrons to minimize neutron absorption by Rh and In, here we report an ordered moment ofmore » $$m=0.54(2)\\,{{\\mu}_{\\text{B}}}$$. In addition, by using spherical neutron polarimetry measurements on a similar single crystal sample, we have confirmed the helical nature of the magnetic structure, and identified a single chiral domain.« less

RESONEUT: A detector system for spectroscopy with (d,n) reactions in inverse kinematics

NASA Astrophysics Data System (ADS)

Baby, L. T.; Kuvin, S. A.; Wiedenhöver, I.; Anastasiou, M.; Caussyn, D.; Colbert, K.; Quails, N.; Gay, D.

2018-01-01

The RESONEUT detector setup is described, which was developed for resonance spectroscopy using (d,n) reactions with radioactive beams in inverse kinematics and at energies around the Coulomb barrier. The goal of experiments with this setup is to determine the spectrum and proton-transfer strengths of the low-lying resonances, which have an impact on astrophysical reaction rates. The setup is optimized for l = 0 proton transfers in inverse kinematics, for which most neutrons are emitted at backward angles with energies in the 80-300 keV range. The detector system is comprised of 9 p-terphenyl scintillators as neutron detectors, two annular silicon-strip detectors for light charged particles, one position-resolving gas ionization chamber for heavy ion detection, and a barrel of NaI-detectors for the detection of γ-rays. The detector commissioning and performance characteristics are described with an emphasis on the neutron-detector components.

The MCUCN simulation code for ultracold neutron physics

NASA Astrophysics Data System (ADS)

Zsigmond, G.

2018-02-01

Ultracold neutrons (UCN) have very low kinetic energies 0-300 neV, thereby can be stored in specific material or magnetic confinements for many hundreds of seconds. This makes them a very useful tool in probing fundamental symmetries of nature (for instance charge-parity violation by neutron electric dipole moment experiments) and contributing important parameters for the Big Bang nucleosynthesis (neutron lifetime measurements). Improved precision experiments are in construction at new and planned UCN sources around the world. MC simulations play an important role in the optimization of such systems with a large number of parameters, but also in the estimation of systematic effects, in benchmarking of analysis codes, or as part of the analysis. The MCUCN code written at PSI has been extensively used for the optimization of the UCN source optics and in the optimization and analysis of (test) experiments within the nEDM project based at PSI. In this paper we present the main features of MCUCN and interesting benchmark and application examples.

Design of a novel instrument for active neutron interrogation of artillery shells.

PubMed

Bélanger-Champagne, Camille; Vainionpää, Hannes; Peura, Pauli; Toivonen, Harri; Eerola, Paula; Dendooven, Peter

2017-01-01

The most common explosives can be uniquely identified by measuring the elemental H/N ratio with a precision better than 10%. Monte Carlo simulations were used to design two variants of a new prompt gamma neutron activation instrument that can achieve this precision. The instrument features an intense pulsed neutron generator with precise timing. Measuring the hydrogen peak from the target explosive is especially challenging because the instrument itself contains hydrogen, which is needed for neutron moderation and shielding. By iterative design optimization, the fraction of the hydrogen peak counts coming from the explosive under interrogation increased from [Formula: see text]% to [Formula: see text]% (statistical only) for the benchmark design. In the optimized design variants, the hydrogen signal from a high-explosive shell can be measured to a statistics-only precision better than 1% in less than 30 minutes for an average neutron production yield of 109 n/s.

Design of a novel instrument for active neutron interrogation of artillery shells

PubMed Central

Vainionpää, Hannes; Peura, Pauli; Toivonen, Harri; Eerola, Paula; Dendooven, Peter

2017-01-01

The most common explosives can be uniquely identified by measuring the elemental H/N ratio with a precision better than 10%. Monte Carlo simulations were used to design two variants of a new prompt gamma neutron activation instrument that can achieve this precision. The instrument features an intense pulsed neutron generator with precise timing. Measuring the hydrogen peak from the target explosive is especially challenging because the instrument itself contains hydrogen, which is needed for neutron moderation and shielding. By iterative design optimization, the fraction of the hydrogen peak counts coming from the explosive under interrogation increased from 53-7+7% to 74-10+8% (statistical only) for the benchmark design. In the optimized design variants, the hydrogen signal from a high-explosive shell can be measured to a statistics-only precision better than 1% in less than 30 minutes for an average neutron production yield of 109 n/s. PMID:29211773

Secondary Neutron Doses to Pediatric Patients During Intracranial Proton Therapy: Monte Carlo Simulation of the Neutron Energy Spectrum and its Organ Doses.

PubMed

Matsumoto, Shinnosuke; Koba, Yusuke; Kohno, Ryosuke; Lee, Choonsik; Bolch, Wesley E; Kai, Michiaki

2016-04-01

Proton therapy has the physical advantage of a Bragg peak that can provide a better dose distribution than conventional x-ray therapy. However, radiation exposure of normal tissues cannot be ignored because it is likely to increase the risk of secondary cancer. Evaluating secondary neutrons generated by the interaction of the proton beam with the treatment beam-line structure is necessary; thus, performing the optimization of radiation protection in proton therapy is required. In this research, the organ dose and energy spectrum were calculated from secondary neutrons using Monte Carlo simulations. The Monte Carlo code known as the Particle and Heavy Ion Transport code System (PHITS) was used to simulate the transport proton and its interaction with the treatment beam-line structure that modeled the double scattering body of the treatment nozzle at the National Cancer Center Hospital East. The doses of the organs in a hybrid computational phantom simulating a 5-y-old boy were calculated. In general, secondary neutron doses were found to decrease with increasing distance to the treatment field. Secondary neutron energy spectra were characterized by incident neutrons with three energy peaks: 1×10, 1, and 100 MeV. A block collimator and a patient collimator contributed significantly to organ doses. In particular, the secondary neutrons from the patient collimator were 30 times higher than those from the first scatter. These results suggested that proactive protection will be required in the design of the treatment beam-line structures and that organ doses from secondary neutrons may be able to be reduced.

Phillips-Tikhonov regularization with a priori information for neutron emission tomographic reconstruction on Joint European Torus

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bielecki, J.; Scholz, M.; Drozdowicz, K.

A method of tomographic reconstruction of the neutron emissivity in the poloidal cross section of the Joint European Torus (JET, Culham, UK) tokamak was developed. Due to very limited data set (two projection angles, 19 lines of sight only) provided by the neutron emission profile monitor (KN3 neutron camera), the reconstruction is an ill-posed inverse problem. The aim of this work consists in making a contribution to the development of reliable plasma tomography reconstruction methods that could be routinely used at JET tokamak. The proposed method is based on Phillips-Tikhonov regularization and incorporates a priori knowledge of the shape ofmore » normalized neutron emissivity profile. For the purpose of the optimal selection of the regularization parameters, the shape of normalized neutron emissivity profile is approximated by the shape of normalized electron density profile measured by LIDAR or high resolution Thomson scattering JET diagnostics. In contrast with some previously developed methods of ill-posed plasma tomography reconstruction problem, the developed algorithms do not include any post-processing of the obtained solution and the physical constrains on the solution are imposed during the regularization process. The accuracy of the method is at first evaluated by several tests with synthetic data based on various plasma neutron emissivity models (phantoms). Then, the method is applied to the neutron emissivity reconstruction for JET D plasma discharge #85100. It is demonstrated that this method shows good performance and reliability and it can be routinely used for plasma neutron emissivity reconstruction on JET.« less

Calibration of ITER Instant Power Neutron Monitors: Recommended Scenario of Experiments at the Reactor

NASA Astrophysics Data System (ADS)

Borisov, A. A.; Deryabina, N. A.; Markovskij, D. V.

2017-12-01

Instant power is a key parameter of the ITER. Its monitoring with an accuracy of a few percent is an urgent and challenging aspect of neutron diagnostics. In a series of works published in Problems of Atomic Science and Technology, Series: Thermonuclear Fusion under a common title, the step-by-step neutronics analysis was given to substantiate a calibration technique for the DT and DD modes of the ITER. A Gauss quadrature scheme, optimal for processing "expensive" experiments, is used for numerical integration of 235U and 238U detector responses to the point sources of 14-MeV neutrons. This approach allows controlling the integration accuracy in relation to the number of coordinate mesh points and thus minimizing the number of irradiations at the given uncertainty of the full monitor response. In the previous works, responses of the divertor and blanket monitors to the isotropic point sources of DT and DD neutrons in the plasma profile and to the models of real sources were calculated within the ITER model using the MCNP code. The neutronics analyses have allowed formulating the basic principles of calibration that are optimal for having the maximum accuracy at the minimum duration of in situ experiments at the reactor. In this work, scenarios of the preliminary and basic experimental ITER runs are suggested on the basis of those principles. It is proposed to calibrate the monitors only with DT neutrons and use correction factors to the DT mode calibration for the DD mode. It is reasonable to perform full calibration only with 235U chambers and calibrate 238U chambers by responses of the 235U chambers during reactor operation (cross-calibration). The divertor monitor can be calibrated using both direct measurement of responses at the Gauss positions of a point source and simplified techniques based on the concepts of equivalent ring sources and inverse response distributions, which will considerably reduce the amount of measurements. It is shown that the monitor based on the average responses of the horizontal and vertical neutron chambers remains spatially stable as the source moves and can be used in addition to the staff monitor at neutron fluxes in the detectors four orders of magnitude lower than on the first wall, where staff detectors are located. Owing to low background, detectors of neutron chambers do not need calibration in the reactor because it is actually determination of the absolute detector efficiency for 14-MeV neutrons, which is a routine out-of-reactor procedure.

The Macromolecular Neutron Diffractometer MaNDi at the Spallation Neutron Source

DOE PAGES

Coates, Leighton; Cuneo, Matthew J.; Frost, Matthew J.; ...

2015-07-18

The Macromolecular Neutron Diffractometer (MaNDi) is located on beamline 11B of the Spallation Neutron Source at Oak Ridge National Laboratory. Moreover, the instrument is a neutron time-of-flight wavelength-resolved Laue diffractometer optimized to collect diffraction data from single crystals. Finally, the instrument has been designed to provide flexibility in several instrumental parameters, such as beam divergence and wavelength bandwidth, to allow data collection from a range of macromolecular systems.

Designing of the 14 MeV neutron moderator for BNCT

NASA Astrophysics Data System (ADS)

Cheng, Dao-Wen; Lu, Jing-Bin; Yang, Dong; Liu, Yu-Min; Wang, Hui-Dong; Ma, Ke-Yan

2012-09-01

In boron neutron capture therapy (BNCT), the ratio of the fast neutron flux to the neutron flux in the tumor (RFNT) must be less than 3%. If a D-T neutron generator is used in BNCT, the 14 MeV neutron moderator must be optimized to reduce the RFNT. Based on the neutron moderation theory and the simulation results, tungsten, lead and diamond were used to moderate the 14 MeV neutrons. Satisfying RFNT of less than 3%, the maximum neutron flux in the tumor was achieved with a three-layer moderator comprised of a 3 cm thick tungsten layer, a 14 cm thick lead layer and a 21 cm thick diamond layer.

SINQ layout, operation, applications and R&D to high power

NASA Astrophysics Data System (ADS)

Bauer, G. S.; Dai, Y.; Wagner, W.

2002-09-01

Since 1997, the Paul Scherrer Institut (PSI) is operating a 1 MW class research spallation neutron source, named SINQ. SINQ is driven by a cascade of three accelerators, the final stage being a 590 MeV isochronous ring cyclotron which delivers a beam current of 1.8 mA at an rf-frequency of 51 MHz. Since for neutron production this is essentially a dc-device, SINQ is a continuous neutron source and is optimized in its design for high time average neutron flux. This makes the facility similar to a research reactor in terms of utilization, but, in terms of beam power, it is, by a large margin, the most powerful spallation neutron source currently in operation world wide. As a consequence, target load levels prevail in SINQ which are beyond the realm of existing experience, demanding a careful approach to the design and operation of a high power target. While the best neutronic performance of the source is expected for a liquid lead-bismuth eutectic target, no experience with such systems exists. For this reason a staged approach has been embarked upon, starting with a heavy water cooled rod target of Zircaloy-2 and proceeding via steel clad lead rods towards the final goal of a target optimised in both, neutronic performance and service life time. Experience currently accruing with a test target containing sample rods with different materials specimens will help to select the proper structural material and make dependable life time estimates accounting for the real operating conditions that prevail in the facility. In parallel, both theoretical and experimental work is going on within the MEGAPIE (MEGAwatt Pilot Experiment) project, a joint initiative by six European research institutions and JAERI (Japan), DOE (USA) and KAERI (Korea), to design, build, operate and explore a liquid lead-bismuth spallation target for 1MW of beam power, taking advantage of the existing spallation neutron facility SINQ.

Postsynthetic Functionalization of Mg-MOF-74 with Tetraethylenepentamine: Structural Characterization and Enhanced CO2 Adsorption.

PubMed

Su, Xiao; Bromberg, Lev; Martis, Vladimir; Simeon, Fritz; Huq, Ashfia; Hatton, T Alan

2017-03-29

Postsynthetic functionalization of magnesium 2,5-dihydroxyterephthalate (Mg-MOF-74) with tetraethylenepentamine (TEPA) resulted in improved CO 2 adsorption performance under dry and humid conditions. XPS, elemental analysis, and neutron powder diffraction studies indicated that TEPA was incorporated throughout the MOF particle, although it coordinated preferentially with the unsaturated metal sites located in the immediate proximity to the surface. Neutron and X-ray powder diffraction analyses showed that the MOF structure was preserved after amine incorporation, with slight changes in the lattice parameters. The adsorption capacity of the functionalized amino-Mg-MOF-74 (TEPA-MOF) for CO 2 was as high as 26.9 wt % versus 23.4 wt % for the original MOF due to the extra binding sites provided by the multiunit amines. The degree of functionalization with the amines was found to be important in enhancing CO 2 adsorption, as the optimal surface coverage improved performance and stability under both pure CO 2 and CO 2 /H 2 O coadsorption, and with partially saturated surface coverage, optimal CO 2 capacity could be achieved under both wet and dry conditions by a synergistic binding of CO 2 to the amines as well as metal centers.

Compact Short-Pulsed Electron Linac Based Neutron Sources for Precise Nuclear Material Analysis

NASA Astrophysics Data System (ADS)

Uesaka, M.; Tagi, K.; Matsuyama, D.; Fujiwara, T.; Dobashi, K.; Yamamoto, M.; Harada, H.

2015-10-01

An X-band (11.424GHz) electron linac as a neutron source for nuclear data study for the melted fuel debris analysis and nuclear security in Fukushima is under development. Originally we developed the linac for Compton scattering X-ray source. Quantitative material analysis and forensics for nuclear security will start several years later after the safe settlement of the accident is established. For the purpose, we should now accumulate more precise nuclear data of U, Pu, etc., especially in epithermal (0.1-10 eV) neutrons. Therefore, we have decided to modify and install the linac in the core space of the experimental nuclear reactor "Yayoi" which is now under the decommission procedure. Due to the compactness of the X-band linac, an electron gun, accelerating tube and other components can be installed in a small space in the core. First we plan to perform the time-of-flight (TOF) transmission measurement for study of total cross sections of the nuclei for 0.1-10 eV energy neutrons. Therefore, if we adopt a TOF line of less than 10m, the o-pulse length of generated neutrons should be shorter than 100 ns. Electronenergy, o-pulse length, power, and neutron yield are ~30 MeV, 100 ns - 1 micros, ~0.4 kW, and ~1011 n/s (~103 n/cm2/s at samples), respectively. Optimization of the design of a neutron target (Ta, W, 238U), TOF line and neutron detector (Ce:LiCAF) of high sensitivity and fast response is underway. We are upgrading the electron gun and a buncher to realize higher current and beam power with a reasonable beam size in order to avoid damage of the neutron target. Although the neutron flux is limited in case of the X-band electron linac based source, we take advantage of its short pulse aspect and availability for nuclear data measurement with a short TOF system. First, we form a tentative configuration in the current experimental room for Compton scattering in 2014. Then, after the decommissioning has been finished, we move it to the "Yayoi" room and perform the operation and measurement.

Performance Study of an aSi Flat Panel Detector for Fast Neutron Imaging of Nuclear Waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schumann, M.; Mauerhofer, E.; Engels, R.

Radioactive waste must be characterized to check its conformance for intermediate storage and final disposal according to national regulations. For the determination of radio-toxic and chemo-toxic contents of radioactive waste packages non-destructive analytical techniques are preferentially used. Fast neutron imaging is a promising technique to assay large and dense items providing, in complementarity to photon imaging, additional information on the presence of structures in radioactive waste packages. Therefore the feasibility of a compact Neutron Imaging System for Radioactive waste Analysis (NISRA) using 14 MeV neutrons is studied in a cooperation framework of Forschungszentrum Juelich GmbH, RWTH Aachen University and Siemensmore » AG. However due to the low neutron emission of neutron generators in comparison to research reactors the challenging task resides in the development of an imaging detector with a high efficiency, a low sensitivity to gamma radiation and a resolution sufficient for the purpose. The setup is composed of a commercial D-T neutron generator (Genie16GT, Sodern) with a surrounding shielding made of polyethylene, which acts as a collimator and an amorphous silicon flat panel detector (aSi, 40 x 40 cm{sup 2}, XRD-1642, Perkin Elmer). Neutron detection is achieved using a general propose plastic scintillator (EJ-260, Eljen Technology) linked to the detector. The thermal noise of the photodiodes is reduced by employing an entrance window made of aluminium. Optimal gain and integration time for data acquisition are set by measuring the response of the detector to the radiation of a 500 MBq {sup 241}Am-source. Detector performance was studied by recording neutron radiography images of materials with various, but well known, chemical compositions, densities and dimensions (Al, C, Fe, Pb, W, concrete, polyethylene, 5 x 8 x 10 cm{sup 3}). To simulate gamma-ray emitting waste radiographs in presence of a gamma-ray sources ({sup 60}Co, {sup 137}Cs, {sup 241}Am) were performed. A homemade algorithm was developed to determine a value which is related to the neutron absorption of the sample with the analysis of the raw detector data. The detector was placed 42 cm away from the neutron source. Distance between detector and the samples was 0.5 cm. At the sample position the fast neutron flux was estimated to 9x10{sup 3} n cm{sup -2} s{sup -1} for a neutron emission of 10{sup 8} n s{sup -1}. The acquisition time was 15 minutes. First neutron radiographs were successfully recorded despite the low detector efficiency and low neutron emission. Analysis of the data shows a correlation between the measured signal and determined neutron absorption. Thus discrimination between different materials of same thicknesses may be achieved. The measurements and results will be presented and discussed in details.« less

Synthesis and characterization of boron fenbufen and its F-18 labeled homolog for boron neutron capture therapy of COX-2 overexpressed cholangiocarcinoma.

PubMed

Yeh, Chun-Nan; Chang, Chi-Wei; Chung, Yi-Hsiu; Tien, Shi-Wei; Chen, Yong-Ren; Chen, Tsung-Wen; Huang, Ying-Cheng; Wang, Hsin-Ell; Chou, You-Cheng; Chen, Ming-Huang; Chiang, Kun-Chun; Huang, Wen-Sheng; Yu, Chung-Shan

2017-09-30

Boron neutron capture therapy (BNCT) is a binary therapy that employs neutron irradiation on the boron agents to release high-energy helium and alpha particles to kill cancer cells. An optimal response to BNCT depends critically on the time point of maximal 10 B accumulation and highest tumor to normal ratio (T/N) for performing the neutron irradiation. The aggressive cholangiocarcinoma (CCA) representing a liver cancer that overexpresses COX-2 enzyme is aimed to be targeted by COX-2 selective boron carrier, fenbufen boronopinacol (FBPin). Two main works were performed including: 1) chemical synthesis of FBPin as the boron carrier and 2) radiochemical labeling with F-18 to provide the radiofluoro congener, m-[ 18 F]fluorofenbufen ester boronopinacol (m-[ 18 F]FFBPin), to assess the binding affinity, cellular accumulation level and distribution profile in CCA rats. FBPin was prepared from bromofenbufen via 3 steps with 82% yield. The binding assay employed [ 18 F]FFBPin to compete FBPin for binding to COX-1 (IC 50 =0.91±0.68μM) and COX-2 (IC 50 =0.33±0.24μM). [ 18 F]FFBPin-derived 60-min dynamic PET scans predict the 10 B-accumulation of 0.8-1.2ppm in liver and 1.2-1.8ppm in tumor and tumor to normal ratio=1.38±0.12. BNCT was performed 40-55min post intravenous administration of FBPin (20-30mg) in the CCA rats. CCA rats treated with BNCT display more tumor reduction than that by NCT with respect of 2-[ 18 F]fluoro-2-deoxy glucose uptake in the tumor region of interest, 20.83±3.00% (n=12) vs. 12.83±3.79% (n=10), P=0.05. The visualizing agent [ 18 F]FFBPin resembles FBPin to generate the time-dependent boron concentration profile. Optimal neutron irradiation period is thus determinable for BNCT. A boron-substituted agent based on COX-2-binding features has been prepared. The moderate COX-2/COX-1 selectivity index of 2.78 allows a fair tumor selectivity index of 1.38 with a mild cardiovascular effect. The therapeutic effect from FBPin with BNCT warrants a proper COX-2 targeting of boron NSAIDs. Copyright © 2017. Published by Elsevier B.V.

Development and characterization of a high yield transportable pulsed neutron source with efficient and compact pulsed power system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Verma, Rishi, E-mail: rishiv9@gmail.com, E-mail: rishiv@barc.gov.in; Mishra, Ekansh; Dhang, Prosenjit

2016-09-15

The results of characterization experiments carried out on a newly developed dense plasma focus device based intense pulsed neutron source with efficient and compact pulsed power system are reported. Its high current sealed pseudospark switch based low inductance capacitor bank with maximum stored energy of ∼10 kJ is segregated into four modules of ∼2.5 kJ each and it cumulatively delivers peak current in the range of 400 kA–600 kA (corresponding to charging voltage range of 14 kV–18 kV) in a quarter time period of ∼2 μs. The neutron yield performance of this device has been optimized by discretely varying deuteriummore » filling gas pressure in the range of 6 mbar–11 mbar at ∼17 kV/550 kA discharge. At ∼7 kJ/8.5 mbar operation, the average neutron yield has been measured to be in the order of ∼4 × 10{sup 9} neutrons/pulse which is the highest ever reported neutron yield from a plasma focus device with the same stored energy. The average forward to radial anisotropy in neutron yield is found to be ∼2. The entire system is contained on a moveable trolley having dimensions 1.5 m × 1 m × 0.7 m and its operation and control (up to the distance of 25 m) are facilitated through optically isolated handheld remote console. The overall compactness of this system provides minimum proximity to small as well as large samples for irradiation. The major intended application objective of this high neutron yield dense plasma focus device development is to explore the feasibility of active neutron interrogation experiments by utilization of intense pulsed neutron sources.« less

Measurement of the scattering cross section of slow neutrons on liquid parahydrogen from neutron transmission

DOE PAGES

Grammer, K. B.; Alarcon, R.; Barrón-Palos, L.; ...

2015-05-08

Liquid hydrogen is a dense Bose fluid whose equilibrium properties are both calculable from first principles using various theoretical approaches and of interest for the understanding of a wide range of questions in many-body physics. Unfortunately, the pair correlation function g(r) inferred from neutron scattering measurements of the differential cross section dσ/dΩ from different measurements reported in the literature are inconsistent. We have measured the energy dependence of the total cross section and the scattering cross section for slow neutrons with energies between 0.43 and 16.1 meV on liquid hydrogen at 15.6 K (which is dominated by the parahydrogen component)more » using neutron transmission measurements on the hydrogen target of the NPDGamma collaboration at the Spallation Neutron Source at Oak Ridge National Laboratory. The relationship between the neutron transmission measurement we perform and the total cross section is unambiguous, and the energy range accesses length scales where the pair correlation function is rapidly varying. At 1 meV our measurement is a factor of 3 below the data from previous work. We present evidence that these previous measurements of the hydrogen cross section, which assumed that the equilibrium value for the ratio of orthohydrogen and parahydrogen has been reached in the target liquid, were in fact contaminated with an extra nonequilibrium component of orthohydrogen. Liquid parahydrogen is also a widely used neutron moderator medium, and an accurate knowledge of its slow neutron cross section is essential for the design and optimization of intense slow neutron sources. Furthermore, we describe our measurements and compare them with previous work.« less

Development and characterization of a high yield transportable pulsed neutron source with efficient and compact pulsed power system.

PubMed

Verma, Rishi; Mishra, Ekansh; Dhang, Prosenjit; Sagar, Karuna; Meena, Manraj; Shyam, Anurag

2016-09-01

The results of characterization experiments carried out on a newly developed dense plasma focus device based intense pulsed neutron source with efficient and compact pulsed power system are reported. Its high current sealed pseudospark switch based low inductance capacitor bank with maximum stored energy of ∼10 kJ is segregated into four modules of ∼2.5 kJ each and it cumulatively delivers peak current in the range of 400 kA-600 kA (corresponding to charging voltage range of 14 kV-18 kV) in a quarter time period of ∼2 μs. The neutron yield performance of this device has been optimized by discretely varying deuterium filling gas pressure in the range of 6 mbar-11 mbar at ∼17 kV/550 kA discharge. At ∼7 kJ/8.5 mbar operation, the average neutron yield has been measured to be in the order of ∼4 × 10 9 neutrons/pulse which is the highest ever reported neutron yield from a plasma focus device with the same stored energy. The average forward to radial anisotropy in neutron yield is found to be ∼2. The entire system is contained on a moveable trolley having dimensions 1.5 m × 1 m × 0.7 m and its operation and control (up to the distance of 25 m) are facilitated through optically isolated handheld remote console. The overall compactness of this system provides minimum proximity to small as well as large samples for irradiation. The major intended application objective of this high neutron yield dense plasma focus device development is to explore the feasibility of active neutron interrogation experiments by utilization of intense pulsed neutron sources.

Light collection and pulse-shape discrimination in elongated scintillator cells for the PROSPECT reactor antineutrino experiment

DOE PAGES

Ashenfelter, J.; Jaffe, D.; Diwan, M. V.; ...

2015-11-06

A meter-long, 23-liter EJ-309 liquid scintillator detector has been constructed to study the light collection and pulse-shape discrimination performance of elongated scintillator cells for the PROSPECT reactor antineutrino experiment. The magnitude and uniformity of light collection and neutron-gamma discrimination power in the energy range of antineutrino inverse beta decay products have been studied using gamma and spontaneous fission calibration sources deployed along the cell axis. We also study neutron-gamma discrimination and light collection abilities for differing PMT and reflector configurations. As a result, key design features for optimizing MeV-scale response and background rejection capabilities are identified.

Simultaneous polarized neutron reflectometry and anisotropic magnetoresistance measurements.

PubMed

Demeter, J; Teichert, A; Kiefer, K; Wallacher, D; Ryll, H; Menéndez, E; Paramanik, D; Steitz, R; Van Haesendonck, C; Vantomme, A; Temst, K

2011-03-01

A novel experimental facility to carry out simultaneous polarized neutron reflectometry (PNR) and anisotropic magnetoresistance (AMR) measurements is presented. Performing both techniques at the same time increases their strength considerably. The proof of concept of this method is demonstrated on a CoO/Co bilayer exchange bias system. Although information on the same phenomena, such as the coercivity or the reversal mechanism, can be separately obtained from either of these techniques, the simultaneous application optimizes the consistency between both. In this way, possible differences in experimental conditions, such as applied magnetic field amplitude and orientation, sample temperature, magnetic history, etc., can be ruled out. Consequently, only differences in the fundamental sensitivities of the techniques can cause discrepancies in the interpretation between the two. The almost instantaneous information obtained from AMR can be used to reveal time-dependent effects during the PNR acquisition. Moreover, the information inferred from the AMR measurements can be used for optimizing the experimental conditions for the PNR measurements in a more efficient way than with the PNR measurements alone.

Imaging of the Li spatial distribution within V 2O 5 cathode in a coin cell by neutron computed tomography

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Yuxuan; Chandran, K. S. Ravi; Bilheux, Hassina Z.

An understanding of Lithium (Li) spatial distribution within the electrodes of a Li-ion cell, during charge and discharge cycles, is essential to optimize the electrode parameters for increased performance under cycling. In this work, it is demonstrated that the spatial distribution of Li within Vanadium Pentoxide (V 2O 5) electrodes of a small coin cell can be imaged by neutron computed tomography. The neutron attenuation data has been used to construct the three-dimensional Li spatial images. Specifically, it is shown that there is sufficient neutron imaging contrast between lithiated and delithiated regions of V 2O 5 electrode making it possiblemore » to map Li distributions even in small electrodes with thicknesses <1 mm. The images reveal that the Li spatial distribution is inhomogeneous and a relatively higher C-rate leads to more non-uniform Li distribution after Li insertion. The non-uniform distribution suggests the limitation of Li diffusion within the electrode during the lithiation process under the relatively high cycling rates.« less

Optimization of a multi-channel parabolic guide for the material science diffractometer STRESS-SPEC at FRM II

NASA Astrophysics Data System (ADS)

Rebelo Kornmeier, Joana; Ostermann, Andreas; Hofmann, Michael; Gibmeier, Jens

2014-02-01

Neutron strain diffractometers usually use slits to define a gauge volume within engineering samples. In this study a multi-channel parabolic neutron guide was developed to be used instead of the primary slit to minimise the loss of intensity and vertical definition of the gauge volume when using slits placed far away from the measurement position in bulky components. The major advantage of a focusing guide is that the maximum flux is not at the exit of the guide as for a slit system but at the focal point relatively far away from the exit of the guide. Monte Carlo simulations were used to optimise the multi-channel parabolic guide with respect to the instrument characteristics of the diffractometer STRESS-SPEC at the FRM II neutron source. Also the simulations are in excellent agreement with experimental measurements using the optimised multi-channel parabolic guide at the neutron diffractometer. In addition the performance of the guide was compared to the standard slit setup at STRESS-SPEC using a single bead weld sample used in earlier round robin tests for residual strain measurements.

Imaging of the Li spatial distribution within V2O5 cathode in a coin cell by neutron computed tomography

NASA Astrophysics Data System (ADS)

Zhang, Yuxuan; Chandran, K. S. Ravi; Bilheux, Hassina Z.

2018-02-01

An understanding of Lithium (Li) spatial distribution within the electrodes of a Li-ion cell, during charge and discharge cycles, is essential to optimize the electrode parameters for increased performance under cycling. In this work, it is demonstrated that the spatial distribution of Li within Vanadium Pentoxide (V2O5) electrodes of a small coin cell can be imaged by neutron computed tomography. The neutron attenuation data has been used to construct the three-dimensional Li spatial images. Specifically, it is shown that there is sufficient neutron imaging contrast between lithiated and delithiated regions of V2O5 electrode making it possible to map Li distributions even in small electrodes with thicknesses <1 mm. The images reveal that the Li spatial distribution is inhomogeneous and a relatively higher C-rate leads to more non-uniform Li distribution after Li insertion. The non-uniform distribution suggests the limitation of Li diffusion within the electrode during the lithiation process under the relatively high cycling rates.

Imaging of the Li spatial distribution within V 2O 5 cathode in a coin cell by neutron computed tomography

DOE PAGES

Zhang, Yuxuan; Chandran, K. S. Ravi; Bilheux, Hassina Z.

2017-11-30

An understanding of Lithium (Li) spatial distribution within the electrodes of a Li-ion cell, during charge and discharge cycles, is essential to optimize the electrode parameters for increased performance under cycling. In this work, it is demonstrated that the spatial distribution of Li within Vanadium Pentoxide (V 2O 5) electrodes of a small coin cell can be imaged by neutron computed tomography. The neutron attenuation data has been used to construct the three-dimensional Li spatial images. Specifically, it is shown that there is sufficient neutron imaging contrast between lithiated and delithiated regions of V 2O 5 electrode making it possiblemore » to map Li distributions even in small electrodes with thicknesses <1 mm. The images reveal that the Li spatial distribution is inhomogeneous and a relatively higher C-rate leads to more non-uniform Li distribution after Li insertion. The non-uniform distribution suggests the limitation of Li diffusion within the electrode during the lithiation process under the relatively high cycling rates.« less

Optimization of small long-life PWR based on thorium fuel

NASA Astrophysics Data System (ADS)

Subkhi, Moh Nurul; Suud, Zaki; Waris, Abdul; Permana, Sidik

2015-09-01

A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% 233U & 2.8% 231Pa, 6% 233U & 2.8% 231Pa and 7% 233U & 6% 231Pa give low excess reactivity.

WARP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bergmann, Ryan M.; Rowland, Kelly L.

2017-04-12

WARP, which can stand for ``Weaving All the Random Particles,'' is a three-dimensional (3D) continuous energy Monte Carlo neutron transport code developed at UC Berkeley to efficiently execute on NVIDIA graphics processing unit (GPU) platforms. WARP accelerates Monte Carlo simulations while preserving the benefits of using the Monte Carlo method, namely, that very few physical and geometrical simplifications are applied. WARP is able to calculate multiplication factors, neutron flux distributions (in both space and energy), and fission source distributions for time-independent neutron transport problems. It can run in both criticality or fixed source modes, but fixed source mode is currentlymore » not robust, optimized, or maintained in the newest version. WARP can transport neutrons in unrestricted arrangements of parallelepipeds, hexagonal prisms, cylinders, and spheres. The goal of developing WARP is to investigate algorithms that can grow into a full-featured, continuous energy, Monte Carlo neutron transport code that is accelerated by running on GPUs. The crux of the effort is to make Monte Carlo calculations faster while producing accurate results. Modern supercomputers are commonly being built with GPU coprocessor cards in their nodes to increase their computational efficiency and performance. GPUs execute efficiently on data-parallel problems, but most CPU codes, including those for Monte Carlo neutral particle transport, are predominantly task-parallel. WARP uses a data-parallel neutron transport algorithm to take advantage of the computing power GPUs offer.« less

Optimization of 200 MWth and 250 MWt Ship Based Small Long Life NPP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fitriyani, Dian; Su'ud, Zaki

2010-06-22

Design optimization of ship-based 200 MWth and 250 MWt nuclear power reactors have been performed. The neutronic and thermo-hydraulic programs of the three-dimensional X-Y-Z geometry have been developed for the analysis of ship-based nuclear power plant. Quasi-static approach is adopted to treat seawater effect. The reactor are loop type lead bismuth cooled fast reactor with nitride fuel and with relatively large coolant pipe above reactor core, the heat from primary coolant system is directly transferred to watersteam loop through steam generators. Square core type are selected and optimized. As the optimization result, the core outlet temperature distribution is changing withmore » the elevation angle of the reactor system and the characteristics are discussed.« less

MOOSE Implementation of MAMBA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Galloway, Jack; Matthews, Topher

The development of MAMBA is targeted at capturing both core wide CRUD induced power shifts (CIPS) as well as pin-­level CRUD induced localized corrosion (CILC). Both CIPS and CILC require some sort of information from thermal-­hydraulic, neutronics, and fuel performance codes, although the degree of coupling is different for the two effects. Since CIPS necessarily requires a core-­wide power distribution solve, it requires tight coupling with a neutronics code. Conversely, CIPS tends to be an individual pin phenomenon, requiring tight coupling a fuel performance code. As efforts are now focused on coupling MAMBA within the VERA suite, a natural separationmore » has surfaced in which a FORTRAN rewrite of MAMBA is optimal for VERA integration to capture CIPS behavior, while a CILC focused calculation would benefit from a tight coupling with BISON, motivating a MOOSE version of MAMBA.« less

Hexagonal boron nitride neutron detectors with high detection efficiencies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maity, A.; Grenadier, S. J.; Li, J.

Here, neutron detectors fabricated from 10B enriched hexagonal boron nitride (h- 10BN or h-BN) epilayers have demonstrated the highest thermal neutron detection efficiency among solid-state neutron detectors to date at about 53%. In this work, photoconductive-like vertical detectors with a detection area of 1 × 1 mm 2 were fabricated from 50 μm thick free-standing h-BN epilayers using Ni/Au and Ti/Al bilayers as ohmic contacts. Leakage currents, mobility-lifetime (μτ) products under UV photoexcitation, and neutron detection efficiencies have been measured for a total of 16 different device configurations. The results have unambiguously identified that detectors incorporating the Ni/Au bilayer onmore » both surfaces as ohmic contacts and using the negatively biased top surface for neutron irradiation are the most desired device configurations. It was noted that high growth temperatures of h- 10BN epilayers on sapphire substrates tend to yield a higher concentration of oxygen impurities near the bottom surface, leading to a better device performance by the chosen top surface for irradiation than by the bottom. Preferential scattering of oxygen donors tends to reduce the mobility of holes more than that of electrons, making the biasing scheme with the ability of rapidly extracting holes at the irradiated surface while leaving the electrons to travel a large average distance inside the detector at a preferred choice. When measured against a calibrated 6LiF filled micro-structured semiconductor neutron detector, it was shown that the optimized configuration has pushed the detection efficiency of h-BN neutron detectors to 58%. These detailed studies also provided a better understanding of growth-mediated impurities in h-BN epilayers and their effects on the charge collection and neutron detection efficiencies.« less

Hexagonal boron nitride neutron detectors with high detection efficiencies

NASA Astrophysics Data System (ADS)

Maity, A.; Grenadier, S. J.; Li, J.; Lin, J. Y.; Jiang, H. X.

2018-01-01

Neutron detectors fabricated from 10B enriched hexagonal boron nitride (h-10BN or h-BN) epilayers have demonstrated the highest thermal neutron detection efficiency among solid-state neutron detectors to date at about 53%. In this work, photoconductive-like vertical detectors with a detection area of 1 × 1 mm2 were fabricated from 50 μm thick free-standing h-BN epilayers using Ni/Au and Ti/Al bilayers as ohmic contacts. Leakage currents, mobility-lifetime (μτ) products under UV photoexcitation, and neutron detection efficiencies have been measured for a total of 16 different device configurations. The results have unambiguously identified that detectors incorporating the Ni/Au bilayer on both surfaces as ohmic contacts and using the negatively biased top surface for neutron irradiation are the most desired device configurations. It was noted that high growth temperatures of h-10BN epilayers on sapphire substrates tend to yield a higher concentration of oxygen impurities near the bottom surface, leading to a better device performance by the chosen top surface for irradiation than by the bottom. Preferential scattering of oxygen donors tends to reduce the mobility of holes more than that of electrons, making the biasing scheme with the ability of rapidly extracting holes at the irradiated surface while leaving the electrons to travel a large average distance inside the detector at a preferred choice. When measured against a calibrated 6LiF filled micro-structured semiconductor neutron detector, it was shown that the optimized configuration has pushed the detection efficiency of h-BN neutron detectors to 58%. These detailed studies also provided a better understanding of growth-mediated impurities in h-BN epilayers and their effects on the charge collection and neutron detection efficiencies.

Hexagonal boron nitride neutron detectors with high detection efficiencies

DOE PAGES

Maity, A.; Grenadier, S. J.; Li, J.; ...

2018-01-23

Here, neutron detectors fabricated from 10B enriched hexagonal boron nitride (h- 10BN or h-BN) epilayers have demonstrated the highest thermal neutron detection efficiency among solid-state neutron detectors to date at about 53%. In this work, photoconductive-like vertical detectors with a detection area of 1 × 1 mm 2 were fabricated from 50 μm thick free-standing h-BN epilayers using Ni/Au and Ti/Al bilayers as ohmic contacts. Leakage currents, mobility-lifetime (μτ) products under UV photoexcitation, and neutron detection efficiencies have been measured for a total of 16 different device configurations. The results have unambiguously identified that detectors incorporating the Ni/Au bilayer onmore » both surfaces as ohmic contacts and using the negatively biased top surface for neutron irradiation are the most desired device configurations. It was noted that high growth temperatures of h- 10BN epilayers on sapphire substrates tend to yield a higher concentration of oxygen impurities near the bottom surface, leading to a better device performance by the chosen top surface for irradiation than by the bottom. Preferential scattering of oxygen donors tends to reduce the mobility of holes more than that of electrons, making the biasing scheme with the ability of rapidly extracting holes at the irradiated surface while leaving the electrons to travel a large average distance inside the detector at a preferred choice. When measured against a calibrated 6LiF filled micro-structured semiconductor neutron detector, it was shown that the optimized configuration has pushed the detection efficiency of h-BN neutron detectors to 58%. These detailed studies also provided a better understanding of growth-mediated impurities in h-BN epilayers and their effects on the charge collection and neutron detection efficiencies.« less

Influence of beam incidence and irradiation parameters on stray neutron doses to healthy organs of pediatric patients treated for an intracranial tumor with passive scattering proton therapy.

PubMed

Bonfrate, A; Farah, J; De Marzi, L; Delacroix, S; Hérault, J; Sayah, R; Lee, C; Bolch, W E; Clairand, I

2016-04-01

In scattering proton therapy, the beam incidence, i.e. the patient's orientation with respect to the beam axis, can significantly influence stray neutron doses although it is almost not documented in the literature. MCNPX calculations were carried out to estimate stray neutron doses to 25 healthy organs of a 10-year-old female phantom treated for an intracranial tumor. Two beam incidences were considered in this article, namely a superior (SUP) field and a right lateral (RLAT) field. For both fields, a parametric study was performed varying proton beam energy, modulation width, collimator aperture and thickness, compensator thickness and air gap size. Using a standard beam line configuration for a craniopharyngioma treatment, neutron absorbed doses per therapeutic dose of 63μGyGy(-1) and 149μGyGy(-1) were found at the heart for the SUP and the RLAT fields, respectively. This dose discrepancy was explained by the different patient's orientations leading to changes in the distance between organs and the final collimator where external neutrons are mainly produced. Moreover, investigations on neutron spectral fluence at the heart showed that the number of neutrons was 2.5times higher for the RLAT field compared against the SUP field. Finally, the influence of some irradiation parameters on neutron doses was found to be different according to the beam incidence. Beam incidence was thus found to induce large variations in stray neutron doses, proving that this parameter could be optimized to enhance the radiation protection of the patient. Copyright © 2016 Associazione Italiana di Fisica Medica. Published by Elsevier Ltd. All rights reserved.

Simultaneous determination of 76As, 122Sb and 153Sm in Chinese medicinal herbs by epithermal neutron activation analysis.

PubMed

Chen, Chien-Yi

2009-01-01

Optimal conditions for the simultaneous determination of As, Sb and Sm in Chinese medicinal herbs using epithermal neutron activation analysis were investigated. The minimum detectable concentrations of 76As, 122Sb and 153Sm in lichen and medicinal herbs depended on the weight of the irradiated sample, and irradiation and decay durations. Optimal conditions were obtained by wrapping the irradiated target with 3.2 mm borated polyethylene neutron filters, which were adopted to screen the original reactor fission neutrons and to reduce the background activities of 38Cl, 24Na and 42K. Twelve medicinal herbs, commonly consumed by Taiwanese children as a diuretic treatment, were analysed since trace elements, such as As and Sb, in these herbs may be toxic when consumed in sufficiently large quantities over a long period. Various amounts of medicinal herbs, standardised powder, lichen and tomato leaves were weighed, packed into polyethylene bags, irradiated and counted under different conditions. The results indicated that about 350 mg of lichen irradiated for 24 h and counted for 20 min following a 30-60 h decay period was optimal for irradiation in a 10(11)n/cm s epithermal neutron flux. The implications of the content of the studied elements in Chinese medicinal herbs are discussed.

Red China’s Capitalist Bomb: Inside the Chinese Neutron Bomb Program

DTIC Science & Technology

2015-01-01

developed an enhanced radiation weapon (ERW) but did not deploy it. ERWs, better known as “ neutron bombs,” are specialized nuclear weapons with...contemporary systems of concern. An ERW is a specialized nuclear weapon optimized to produce prompt radiation. Such a device emits neutrons with high...Council stated that China mastered “in succession the neutron bomb design technology and the nuclear weapon miniaturization technology.”10 This statement

Biodistribution of sodium borocaptate (BSH) for boron neutron capture therapy (BNCT) in an oral cancer model.

PubMed

Garabalino, Marcela A; Heber, Elisa M; Monti Hughes, Andrea; González, Sara J; Molinari, Ana J; Pozzi, Emiliano C C; Nievas, Susana; Itoiz, Maria E; Aromando, Romina F; Nigg, David W; Bauer, William; Trivillin, Verónica A; Schwint, Amanda E

2013-08-01

Boron neutron capture therapy (BNCT) is based on selective accumulation of ¹⁰B carriers in tumor followed by neutron irradiation. We previously proved the therapeutic success of BNCT mediated by the boron compounds boronophenylalanine and sodium decahydrodecaborate (GB-10) in the hamster cheek pouch oral cancer model. Based on the clinical relevance of the boron carrier sodium borocaptate (BSH) and the knowledge that the most effective way to optimize BNCT is to improve tumor boron targeting, the specific aim of this study was to perform biodistribution studies of BSH in the hamster cheek pouch oral cancer model and evaluate the feasibility of BNCT mediated by BSH at nuclear reactor RA-3. The general aim of these studies is to contribute to the knowledge of BNCT radiobiology and optimize BNCT for head and neck cancer. Sodium borocaptate (50 mg ¹⁰B/kg) was administered to tumor-bearing hamsters. Groups of 3-5 animals were killed humanely at nine time-points, 3-12 h post-administration. Samples of blood, tumor, precancerous pouch tissue, normal pouch tissue and other clinically relevant normal tissues were processed for boron measurement by optic emission spectroscopy. Tumor boron concentration peaked to therapeutically useful boron concentration values of 24-35 ppm. The boron concentration ratio tumor/normal pouch tissue ranged from 1.1 to 1.8. Pharmacokinetic curves showed that the optimum interval between BSH administration and neutron irradiation was 7-11 h. It is concluded that BNCT mediated by BSH at nuclear reactor RA-3 would be feasible.

Optimization and comprehensive characterization of metal hydride based hydrogen storage systems using in-situ Neutron Radiography

NASA Astrophysics Data System (ADS)

Börries, S.; Metz, O.; Pranzas, P. K.; Bellosta von Colbe, J. M.; Bücherl, T.; Dornheim, M.; Klassen, T.; Schreyer, A.

2016-10-01

For the storage of hydrogen, complex metal hydrides are considered as highly promising with respect to capacity, reversibility and safety. The optimization of corresponding storage tanks demands a precise and time-resolved investigation of the hydrogen distribution in scaled-up metal hydride beds. In this study it is shown that in situ fission Neutron Radiography provides unique insights into the spatial distribution of hydrogen even for scaled-up compacts and therewith enables a direct study of hydrogen storage tanks. A technique is introduced for the precise quantification of both time-resolved data and a priori material distribution, allowing inter alia for an optimization of compacts manufacturing process. For the first time, several macroscopic fields are combined which elucidates the great potential of Neutron Imaging for investigations of metal hydrides by going further than solely 'imaging' the system: A combination of in-situ Neutron Radiography, IR-Thermography and thermodynamic quantities can reveal the interdependency of different driving forces for a scaled-up sodium alanate pellet by means of a multi-correlation analysis. A decisive and time-resolved, complex influence of material packing density is derived. The results of this study enable a variety of new investigation possibilities that provide essential information on the optimization of future hydrogen storage tanks.

TREAT Neutronics Analysis and Design Support, Part I: Multi-SERTTA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Woolstenhulme, Nicolas E.; Hill, Connie M.

2016-08-01

Experiment vehicle design is necessary in preparation for Transient Reactor Test (TREAT) facility restart and the resumption of transient testing to support Accident Tolerant Fuel (ATF) characterization and other future fuels testing requirements. Currently the most mature vehicle design is the Multi-SERTTA (Static Environments Rodlet Transient Test Apparatuses), which can accommodate up to four concurrent rodlet-sized specimens under separate environmental conditions. Robust test vehicle design requires neutronics analyses to support design development, optimization of the power coupling factor (PCF) to efficiently maximize energy generation in the test fuel rodlets, and experiment safety analyses. Calculations were performed to support analysis ofmore » a near-final design of the Multi-SERTTA vehicle, the design process for future TREAT test vehicles, and establish analytical practices for upcoming transient test experiments. Models of the Multi-SERTTA vehicle containing typical PWR-fuel rodlets were prepared and neutronics calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Calculation of the PCF for reference conditions of a PWR fuel rodlet in clean water at operational temperature and pressure provided results between 1.10 and 1.74 W/g-MW depending on the location of the four Multi-SERTTA units with the stack. Basic changes to the Multi-SERTTA secondary vessel containment and support have minimal impact on PCF; using materials with less neutron absorption can improve expected PCF values, especially in the primary containment. An optimized balance is needed between structural integrity, experiment safety, and energy deposition in the experiment. Type of medium and environmental conditions within the primary vessel surrounding the fuel rodlet can also have a significant impact on resultant PCF values. The estimated reactivity insertion worth into the TREAT core is impacted more by the primary and secondary Multi-SERTTA vehicle structure with the experiment content and contained environment having a near negligible impact on overall system reactivity. Additional calculations were performed to evaluate the peak-to-average assembly powers throughout the TREAT core, as well as the nuclear heat generation for the various structural components of the Multi-SERTTA assembly. Future efforts include the evaluation of flux collars to shape the PCF for individual Multi-SERTTA units during an experiment such as to achieve uniformity in test unit environmental conditions impacted by the non-uniform axial flux/power profile of TREAT. Upon resumption of transient testing, experimental results from both the Multi-SERTTA and Multi-SERTTA-CAL will be compared against calculational results and methods for further optimization and design strategies.« less

SU-E-T-21: A D-D Based Neutron Generator System for Boron Neutron Capture Therapy: A Feasibility Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hsieh, M; Liu, Y; Nie, L

2015-06-15

Purpose: To investigate the feasibility of a deuterium-deuterium (DD) neutron generator for application in boron neutron capture therapy (BNCT) of brain cancer Methods: MCNP simulations were performed using a head phantom and a monoenergetic neutron source, which resembles the point source in a DD generator that emits 2.45-MeV neutrons. Source energies ranging from 5eV to 2.45MeV were simulated to determine the optimal treatment energy. The phantom consisted of soft tissue, brain tissue, skull, skin layer, and a brain tumor of 5 cm in diameter. Tumor depth was varied from 5–10 cm. Boron-10 concentrations of 10 ppm, 15 ppm, and 30more » ppm were used in the soft/brain tissues, skin, and tumor, respectively. The neutron flux required to deliver 60 Gy to the tumor as well as the normal tissue doses were determined. Results: Beam energies between 5eV and 10keV obtained doses with the highest dose ratios (3.3–25.9) between the tumor and the brain at various depths. The dose ratio with 2.45-MeV neutrons ranged from 0.8–6.6. To achieve the desired tumor dose in 40 minutes, the required neutron flux for a DD generator was between 8.8E10 and 5.2E11 n/s and the resulting brain dose was between 2.3 and 18 Gy, depending on the tumor depth. The skin and soft tissue doses were within acceptable tolerances. The boron-neutron interaction accounted for 54–58% of the total dose. Conclusion: This study shows that the DD neutron generator can be a feasible neutron source for BNCT. The required neutron flux for treatment is achievable with the current DD neutron technology. With a well-designed beam shaping assembly and treatment geometry, the neutron flux can be further improved and a 60-Gy prescription can be accurately delivered to the target while maintaining tolerable normal tissue doses. Further experimental studies will be developed and conducted to validate the simulation results.« less

Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors. Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hawari, Ayman; Ougouag, Abderrafi

2014-07-08

This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermalization is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can bemore » easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.« less

ANTS — a simulation package for secondary scintillation Anger-camera type detector in thermal neutron imaging

NASA Astrophysics Data System (ADS)

Morozov, A.; Defendi, I.; Engels, R.; Fraga, F. A. F.; Fraga, M. M. F. R.; Guerard, B.; Jurkovic, M.; Kemmerling, G.; Manzin, G.; Margato, L. M. S.; Niko, H.; Pereira, L.; Petrillo, C.; Peyaud, A.; Piscitelli, F.; Raspino, D.; Rhodes, N. J.; Sacchetti, F.; Schooneveld, E. M.; Van Esch, P.; Zeitelhack, K.

2012-08-01

A custom and fully interactive simulation package ANTS (Anger-camera type Neutron detector: Toolkit for Simulations) has been developed to optimize the design and operation conditions of secondary scintillation Anger-camera type gaseous detectors for thermal neutron imaging. The simulation code accounts for all physical processes related to the neutron capture, energy deposition pattern, drift of electrons of the primary ionization and secondary scintillation. The photons are traced considering the wavelength-resolved refraction and transmission of the output window. Photo-detection accounts for the wavelength-resolved quantum efficiency, angular response, area sensitivity, gain and single-photoelectron spectra of the photomultipliers (PMTs). The package allows for several geometrical shapes of the PMT photocathode (round, hexagonal and square) and offers a flexible PMT array configuration: up to 100 PMTs in a custom arrangement with the square or hexagonal packing. Several read-out patterns of the PMT array are implemented. Reconstruction of the neutron capture position (projection on the plane of the light emission) is performed using the center of gravity, maximum likelihood or weighted least squares algorithm. Simulation results reproduce well the preliminary results obtained with a small-scale detector prototype. ANTS executables can be downloaded from http://coimbra.lip.pt/~andrei/.

A Project of Boron Neutron Capture Therapy System based on a Proton Linac Neutron Source

NASA Astrophysics Data System (ADS)

Kiyanagi, Yoshikai; Asano, Kenji; Arakawa, Akihiro; Fukuchi, Shin; Hiraga, Fujio; Kimura, Kenju; Kobayashi, Hitoshi; Kubota, Michio; Kumada, Hiroaki; Matsumoto, Hiroshi; Matsumoto, Akira; Sakae, Takeji; Saitoh, Kimiaki; Shibata, Tokushi; Yoshioka, Masakazu

At present, the clinical trials of Boron Neutron Capture Therapy (BNCT) are being performed at research reactor facilities. However, an accelerator based BNCT has a merit that it can be built in a hospital. So, we just launched a development project for the BNCT based on an accelerator in order to establish and to spread the BNCT as an effective therapy in the near future. In the project, a compact proton linac installed in a hospital will be applied as a neutron source, and energy of the proton beam is planned to be less than about 10 MeV to reduce the radioactivity. The BNCT requires epithermal neutron beam with an intensity of around 1x109 (n/cm2/sec) to deliver the therapeutic dose to a deeper region in a body and to complete the irradiation within an hour. From this condition, the current of the proton beam required is estimated to be a few mA on average. Enormous heat deposition in the target is a big issue. We are aiming at total optimization of the accelerator based BNCT from the linac to the irradiation position. Here, the outline of the project is introduced and the moderator design is presented.

Neutron Tomography of a Fuel Cell: Statistical Learning Implementation of a Penalized Likelihood Method

NASA Astrophysics Data System (ADS)

Coakley, Kevin J.; Vecchia, Dominic F.; Hussey, Daniel S.; Jacobson, David L.

2013-10-01

At the NIST Neutron Imaging Facility, we collect neutron projection data for both the dry and wet states of a Proton-Exchange-Membrane (PEM) fuel cell. Transmitted thermal neutrons captured in a scintillator doped with lithium-6 produce scintillation light that is detected by an amorphous silicon detector. Based on joint analysis of the dry and wet state projection data, we reconstruct a residual neutron attenuation image with a Penalized Likelihood method with an edge-preserving Huber penalty function that has two parameters that control how well jumps in the reconstruction are preserved and how well noisy fluctuations are smoothed out. The choice of these parameters greatly influences the resulting reconstruction. We present a data-driven method that objectively selects these parameters, and study its performance for both simulated and experimental data. Before reconstruction, we transform the projection data so that the variance-to-mean ratio is approximately one. For both simulated and measured projection data, the Penalized Likelihood method reconstruction is visually sharper than a reconstruction yielded by a standard Filtered Back Projection method. In an idealized simulation experiment, we demonstrate that the cross validation procedure selects regularization parameters that yield a reconstruction that is nearly optimal according to a root-mean-square prediction error criterion.

Neutron detector using lithiated glass-scintillating particle composite

DOEpatents

Wallace, Steven [Knoxville, TN; Stephan, Andrew C [Knoxville, TX; Dai, Sheng [Knoxville, TN; Im, Hee-Jung [Knoxville, TN

2009-09-01

A neutron detector composed of a matrix of scintillating particles imbedded in a lithiated glass is disclosed. The neutron detector detects the neutrons by absorbing the neutron in the lithium-6 isotope which has been enriched from the natural isotopic ratio to a commercial ninety five percent. The utility of the detector is optimized by suitably selecting scintillating particle sizes in the range of the alpha and the triton. Nominal particle sizes are in the range of five to twenty five microns depending upon the specific scintillating particle selected.

Feasibility study of 235U and 239Pu characterization in radioactive waste drums using neutron-induced fission delayed gamma rays

NASA Astrophysics Data System (ADS)

Nicol, T.; Pérot, B.; Carasco, C.; Brackx, E.; Mariani, A.; Passard, C.; Mauerhofer, E.; Collot, J.

2016-10-01

This paper reports a feasibility study of 235U and 239Pu characterization in 225 L bituminized waste drums or 200 L concrete waste drums, by detecting delayed fission gamma rays between the pulses of a deuterium-tritium neutron generator. The delayed gamma yields were first measured with bare samples of 235U and 239Pu in REGAIN, a facility dedicated to the assay of 118 L waste drums by Prompt Gamma Neutron Activation Analysis (PGNAA) at CEA Cadarache, France. Detectability in the waste drums is then assessed using the MCNPX model of MEDINA (Multi Element Detection based on Instrumental Neutron Activation), another PGNAA cell dedicated to 200 L drums at FZJ, Germany. For the bituminized waste drum, performances are severely hampered by the high gamma background due to 137Cs, which requires the use of collimator and shield to avoid electronics saturation, these elements being very penalizing for the detection of the weak delayed gamma signal. However, for lower activity concrete drums, detection limits range from 10 to 290 g of 235U or 239Pu, depending on the delayed gamma rays of interest. These detection limits have been determined by using MCNPX to calculate the delayed gamma useful signal, and by measuring the experimental gamma background in MEDINA with a 200 L concrete drum mock-up. The performances could be significantly improved by using a higher interrogating neutron emission and an optimized experimental setup, which would allow characterizing nuclear materials in a wide range of low and medium activity waste packages.

The Development of WARP - A Framework for Continuous Energy Monte Carlo Neutron Transport in General 3D Geometries on GPUs

NASA Astrophysics Data System (ADS)

Bergmann, Ryan

Graphics processing units, or GPUs, have gradually increased in computational power from the small, job-specific boards of the early 1990s to the programmable powerhouses of today. Compared to more common central processing units, or CPUs, GPUs have a higher aggregate memory bandwidth, much higher floating-point operations per second (FLOPS), and lower energy consumption per FLOP. Because one of the main obstacles in exascale computing is power consumption, many new supercomputing platforms are gaining much of their computational capacity by incorporating GPUs into their compute nodes. Since CPU-optimized parallel algorithms are not directly portable to GPU architectures (or at least not without losing substantial performance), transport codes need to be rewritten to execute efficiently on GPUs. Unless this is done, reactor simulations cannot take full advantage of these new supercomputers. WARP, which can stand for ``Weaving All the Random Particles,'' is a three-dimensional (3D) continuous energy Monte Carlo neutron transport code developed in this work as to efficiently implement a continuous energy Monte Carlo neutron transport algorithm on a GPU. WARP accelerates Monte Carlo simulations while preserving the benefits of using the Monte Carlo Method, namely, very few physical and geometrical simplifications. WARP is able to calculate multiplication factors, flux tallies, and fission source distributions for time-independent problems, and can run in both criticality or fixed source modes. WARP can transport neutrons in unrestricted arrangements of parallelepipeds, hexagonal prisms, cylinders, and spheres. WARP uses an event-based algorithm, but with some important differences. Moving data is expensive, so WARP uses a remapping vector of pointer/index pairs to direct GPU threads to the data they need to access. The remapping vector is sorted by reaction type after every transport iteration using a high-efficiency parallel radix sort, which serves to keep the reaction types as contiguous as possible and removes completed histories from the transport cycle. The sort reduces the amount of divergence in GPU ``thread blocks,'' keeps the SIMD units as full as possible, and eliminates using memory bandwidth to check if a neutron in the batch has been terminated or not. Using a remapping vector means the data access pattern is irregular, but this is mitigated by using large batch sizes where the GPU can effectively eliminate the high cost of irregular global memory access. WARP modifies the standard unionized energy grid implementation to reduce memory traffic. Instead of storing a matrix of pointers indexed by reaction type and energy, WARP stores three matrices. The first contains cross section values, the second contains pointers to angular distributions, and a third contains pointers to energy distributions. This linked list type of layout increases memory usage, but lowers the number of data loads that are needed to determine a reaction by eliminating a pointer load to find a cross section value. Optimized, high-performance GPU code libraries are also used by WARP wherever possible. The CUDA performance primitives (CUDPP) library is used to perform the parallel reductions, sorts and sums, the CURAND library is used to seed the linear congruential random number generators, and the OptiX ray tracing framework is used for geometry representation. OptiX is a highly-optimized library developed by NVIDIA that automatically builds hierarchical acceleration structures around user-input geometry so only surfaces along a ray line need to be queried in ray tracing. WARP also performs material and cell number queries with OptiX by using a point-in-polygon like algorithm. WARP has shown that GPUs are an effective platform for performing Monte Carlo neutron transport with continuous energy cross sections. Currently, WARP is the most detailed and feature-rich program in existence for performing continuous energy Monte Carlo neutron transport in general 3D geometries on GPUs, but compared to production codes like Serpent and MCNP, WARP has limited capabilities. Despite WARP's lack of features, its novel algorithm implementations show that high performance can be achieved on a GPU despite the inherently divergent program flow and sparse data access patterns. WARP is not ready for everyday nuclear reactor calculations, but is a good platform for further development of GPU-accelerated Monte Carlo neutron transport. In it's current state, it may be a useful tool for multiplication factor searches, i.e. determining reactivity coefficients by perturbing material densities or temperatures, since these types of calculations typically do not require many flux tallies. (Abstract shortened by UMI.)

Methodes d'optimisation des parametres 2D du reflecteur dans un reacteur a eau pressurisee

NASA Astrophysics Data System (ADS)

Clerc, Thomas

With a third of the reactors in activity, the Pressurized Water Reactor (PWR) is today the most used reactor design in the world. This technology equips all the 19 EDF power plants. PWRs fit into the category of thermal reactors, because it is mainly the thermal neutrons that contribute to the fission reaction. The pressurized light water is both used as the moderator of the reaction and as the coolant. The active part of the core is composed of uranium, slightly enriched in uranium 235. The reflector is a region surrounding the active core, and containing mostly water and stainless steel. The purpose of the reflector is to protect the vessel from radiations, and also to slow down the neutrons and reflect them into the core. Given that the neutrons participate to the reaction of fission, the study of their behavior within the core is capital to understand the general functioning of how the reactor works. The neutrons behavior is ruled by the transport equation, which is very complex to solve numerically, and requires very long calculation. This is the reason why the core codes that will be used in this study solve simplified equations to approach the neutrons behavior in the core, in an acceptable calculation time. In particular, we will focus our study on the diffusion equation and approximated transport equations, such as SPN or S N equations. The physical properties of the reflector are radically different from those of the fissile core, and this structural change causes important tilt in the neutron flux at the core/reflector interface. This is why it is very important to accurately design the reflector, in order to precisely recover the neutrons behavior over the whole core. Existing reflector calculation techniques are based on the Lefebvre-Lebigot method. This method is only valid if the energy continuum of the neutrons is discretized in two energy groups, and if the diffusion equation is used. The method leads to the calculation of a homogeneous reflector. The aim of this study is to create a computational scheme able to compute the parameters of heterogeneous, multi-group reflectors, with both diffusion and SPN/SN operators. For this purpose, two computational schemes are designed to perform such a reflector calculation. The strategy used in both schemes is to minimize the discrepancies between a power distribution computed with a core code and a reference distribution, which will be obtained with an APOLLO2 calculation based on the method Method Of Characteristics (MOC). In both computational schemes, the optimization parameters, also called control variables, are the diffusion coefficients in each zone of the reflector, for diffusion calculations, and the P-1 corrected macroscopic total cross-sections in each zone of the reflector, for SPN/SN calculations (or correction factors on these parameters). After a first validation of our computational schemes, the results are computed, always by optimizing the fast diffusion coefficient for each zone of the reflector. All the tools of the data assimilation have been used to reflect the different behavior of the solvers in the different parts of the core. Moreover, the reflector is refined in six separated zones, corresponding to the physical structure of the reflector. There will be then six control variables for the optimization algorithms. [special characters omitted]. Our computational schemes are then able to compute heterogeneous, 2-group or multi-group reflectors, using diffusion or SPN/SN operators. The optimization performed reduces the discrepancies distribution between the power computed with the core codes and the reference power. However, there are two main limitations to this study: first the homogeneous modeling of the reflector assemblies doesn't allow to properly describe its physical structure near the core/reflector interface. Moreover, the fissile assemblies are modeled in infinite medium, and this model reaches its limit at the core/reflector interface. These two problems should be tackled in future studies. (Abstract shortened by UMI.).

Optimization of the photoneutron target geometry for e-accelerator based BNCT.

PubMed

Chegeni, Nahid; Pur, Saleh Boveiry; Razmjoo, Sasan; Hoseini, Seydeh Khadijed

2017-06-01

Today, electron accelerators are taken into consideration as photoneutron sources. Therefore, for maximum production of epithermal neutron flux, designing a photoneutron target is of significant importance. In this paper, the effect of thickness and geometric shape of a photoneutron target on neutron output were investigated. In this study, a pencil photon source with 13, 15, 18, 20 and 25 MeV energies and a diameter of 2 mm was investigated using Monte Carlo simulation method using MCNP code. To optimize the design of the photoneutron target, the tungsten target with various geometries and thicknesses was investigated. The maximum neutron flux produced for all target geometries and thicknesses occurred at neutron energy peak of around 0.46 MeV. As the thickness increased to 2 cm, neutron flux increased and then a decreasing trend was observed. For various geometrical shapes, the determining factor in photoneutron output was the effective target thickness in the photon interaction path that increased by the increase in the area of interaction. Another factor was the angle of the photon's incidence with the target surface that resulted in a significant decrease in photoneutron output in cone-shaped targets. Three factors including the total neutron flux, neutrons energy spectrum, and convergence of neutrons plays an important role in the selection of geometry and shape of the target that should be investigated considering beam shaping assembly (BSA) shape.

Optimization of the Efficiency of a Neutron Detector to Measure (α, n) Reaction Cross-Section

NASA Astrophysics Data System (ADS)

Perello, Jesus; Montes, Fernando; Ahn, Tony; Meisel, Zach; Joint InstituteNuclear Astrophysics Team

2015-04-01

Nucleosynthesis, the origin of elements, is one of the greatest mysteries in physics. A recent particular nucleosynthesis process of interest is the charge-particle process (cpp). In the cpp, elements form by nuclear fusion reactions during supernovae. This process of nuclear fusion, (α,n), will be studied by colliding beam elements produced and accelerated at the National Superconducting Cyclotron Laboratory (NSCL) to a helium-filled cell target. The elements will fuse with α (helium nuclei) and emit neutrons during the reaction. The neutrons will be detected for a count of fused-elements, thus providing us the probability of such reactions. The neutrons will be detected using the Neutron Emission Ratio Observer (NERO). Currently, NERO's efficiency varies for neutrons at the expected energy range (0-12 MeV). To study (α,n), NERO's efficiency must be near-constant at these energies. Monte-Carlo N-Particle Transport Code (MCNP6), a software package that simulates nuclear processes, was used to optimize NERO configuration for the experiment. MCNP6 was used to simulate neutron interaction with different NERO configurations at the expected neutron energies. By adding additional 3He detectors and polyethylene, a near-constant efficiency at these energies was obtained in the simulations. With the new NERO configuration, study of the (α,n) reactions can begin, which may explain how elements are formed in the cpp. SROP MSU, NSF, JINA, McNair Society.

Compact D-D Neutron Source-Driven Subcritical Multiplier and Beam-Shaping Assembly for Boron Neutron Capture Therapy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francesco Ganda; Jasmina Vujic; Ehud Greenspan

2010-12-01

This work assesses the feasibility of using a small, safe, and inexpensive keff 0.98 subcritical fission assembly [subcritical neutron multiplier (SCM)] to amplify the treatment neutron beam intensity attainable from a compact deuterium-deuterium (D-D) fusion neutron source delivering [approximately]1012 n/s. The objective is to reduce the treatment time for deep-seated brain tumors to [approximately]1 h. The paper describes the optimal SCM design and two optimal beam-shaping assemblies (BSAs) - one designed to maximize the dose rate and the other designed to maximize the total dose that can be delivered to a deep-seated tumor. The neutron beam intensity amplification achieved withmore » the optimized SCM and BSA results in an increase in the treatment dose rate by a factor of 18: from 0.56 Gy/h without the SCM to 10.1 Gy/h. The entire SCM is encased in an aluminum structure. The total amount of 20% enriched uranium required for the SCM is 8.5 kg, and the cost (not including fabrication) is estimated to be less than $60,000. The SCM power level is estimated at 400 W when driven by a 1012 n/s D-D neutron source. This translates into consumption of only [approximately]0.6% of the initially loaded 235U atoms during 50 years of continuous operation and implies that the SCM could operate continuously for the entire lifetime of the facility without refueling. Cooling the SCM does not pose a challenge; it may be accomplished by natural circulation as the maximum heat flux is only 0.034 W/cm2.« less

Application and Development of Microstructured Solid-State Neutron Detectors

NASA Astrophysics Data System (ADS)

Weltz, Adam D.

Neutron detectors are useful for a number of applications, including the identification of nuclear weapons, radiation dosimetry, and nuclear reactor monitoring, among others. Microstructured solid-state neutron detectors (SSNDs) developed at RPI have the potential to reinvent a variety of neutron detection systems due to their compact size, zero bias requirement, competitive thermal neutron detection efficiency (up to 29%), low gamma sensitivity (below the PNNL recommendation of 10-6 corresponding to a 10 mR/hr gamma exposure), and scalability to large surface areas with a single preamplifier (<20% loss in relative efficiency from 1 to 16 cm2). These microstructured SSNDs have semiconducting substrate etched with a repeated, three-dimensional microstructure of high aspect ratio holes filled with 10B. MCNP simulations optimized the dimensions of each microstructure geometry for each detector application, improving the overall performance. This thesis outlines the development of multiple, novel neutron detection applications using microstructured SSNDs developed at RPI. The Directional and Spectral Neutron Detection System (DSNDS) is a modular and portable system that uses rings of microstructured SSNDs embedded in polyethylene in order to gather real-time information about the directionality and spectrum of an unidentified neutron source. This system can be used to identify the presence of diverted special nuclear material (SNM), determine its position, and gather spectral information in real-time. The compact and scalable zero-bias SSNDs allow for customization and modularity of the detector array, which provides design flexibility and enhanced portability. Additionally, a real-time personal neutron dosimeter is a wearable device that uses a combination of fast and thermal microstructured SSNDs in order to determine an individual's neutron dose rate. This system demonstrates that neutron detection systems utilizing microstructured SSNDs are applicable for personal neutron dosimetry. The development of these systems using the compact, zero-bias microstructured SSNDs lays the groundwork for a new generation of neutron detection tools, outlines the challenges and design considerations associated with the implementation of these devices, and demonstrates the value that these detectors bring to the future of neutron detection systems.

Expected total counts for the Self-Interrogation Neutron Resonance Densitometry measurements of spent nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rossa, Riccardo; Universite Libre de Bruxelles; Borella, Alessandro

2015-07-01

The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive neutron technique that aims at a direct quantification of {sup 239}Pu in spent fuel assemblies by measuring the attenuation of the neutron flux in the energy region close to the 0.3 eV resonance of {sup 239}Pu. The {sup 239}Pu mass is estimated by calculating the SINRD signature, that is the ratio between the neutron counts in the fast energy region and around the 0.3 eV resonance region. The SINRD measurement approach in this study consisted in introducing a small neutron detector in the central guide tube of a PWR 17x17 fuelmore » assembly. In order to measure the neutron flux in the energy regions defined in the SINRD signature, different detector types were used. The response of a bare {sup 238}U fission chamber is considered for the determination of the fast neutron flux, while other thermal-epithermal detectors wrapped in neutron absorbers are envisaged to measure the neutron flux around the resonance region. This paper provides an estimation of the total neutron counts that can be achieved with the detector types proposed for the SINRD measurement. In the first section a set of detectors are evaluated in terms of total neutron counts and sensitivity to the {sup 239}Pu content, in order to identify the optimal measurement configuration for each detector type. Then a study is performed to increase the total neutron counts by increasing the detector size. The study shows that the highest total neutron counts are achieved by using either {sup 3}He or {sup 10}B proportional counters because of the high neutron efficiency of these detectors. However, the calculations indicate that the biggest contribution to the measurement uncertainty is due to the measurement of the fast neutron flux. Finally, similar sensitivity to the {sup 239}Pu content is obtained by using the different detector types for the measurement of the neutron flux close to the resonance region. Therefore, the total neutron counts associated to each detector type will play a major role in the selection of the detector types used for the SINRD measurement. (authors)« less

Optimization of small long-life PWR based on thorium fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Subkhi, Moh Nurul, E-mail: nsubkhi@students.itb.ac.id; Physics Dept., Faculty of Science and Technology, State Islamic University of Sunan Gunung Djati Bandung Jalan A.H Nasution 105 Bandung; Suud, Zaki, E-mail: szaki@fi.itb.ac.id

2015-09-30

A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% {sup 233}U & 2.8% {sup 231}Pa, 6% {sup 233}U & 2.8% {sup 231}Pa and 7% {sup 233}U & 6% {supmore » 231}Pa give low excess reactivity.« less

TREAT Neutronics Analysis and Design Support, Part II: Multi-SERTTA-CAL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Woolstenhulme, Nicolas E.; Hill, Connie M.

2016-08-01

Experiment vehicle design is necessary in preparation for Transient Reactor Test (TREAT) facility restart and the resumption of transient testing to support Accident Tolerant Fuel (ATF) characterization and other future fuels testing requirements. Currently the most mature vehicle design is the Multi-SERTTA (Static Environments Rodlet Transient Test Apparatuses), which can accommodate up to four concurrent rodlet-sized specimens under separate environmental conditions. Robust test vehicle design requires neutronics analyses to support design development, optimization of the power coupling factor (PCF) to efficiently maximize energy generation in the test fuel rodlets, and experiment safety analyses. In integral aspect of prior TREAT transientmore » testing was the incorporation of calibration experiments to experimentally evaluate and validate test conditions in preparation of the actual fuel testing. The calibration experiment package established the test parameter conditions to support fine-tuning of the computational models to deliver the required energy deposition to the fuel samples. The calibration vehicle was designed to be as near neutronically equivalent to the experiment vehicle as possible to minimize errors between the calibration and final tests. The Multi-SERTTA-CAL vehicle was designed to serve as the calibration vehicle supporting Multi-SERTTA experimentation. Models of the Multi-SERTTA-CAL vehicle containing typical PWR-fuel rodlets were prepared and neutronics calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries; these results were then compared against those performed for Multi-SERTTA to determine the similarity and possible design modification necessary prior to construction of these experiment vehicles. The estimated reactivity insertion worth into the TREAT core is very similar between the two vehicle designs, with the primary physical difference being a hollow Inconel tube running down the length of the calibration vehicle. Calculations of PCF indicate that on average there is a reduction of approximately 6.3 and 12.6%, respectively, for PWR fuel rodlets irradiated under wet and dry conditions. Changes to the primary or secondary vessel structure in the calibration vehicle can be performed to offset this discrepancy and maintain neutronic equivalency. Current possible modifications to the calibration vehicle include reduction of the primary vessel wall thickness, swapping Zircaloy-4 for stainless steel 316 in the secondary containment, or slight modification to the temperature and pressure of the water environment within the primary vessel. Removal of some of the instrumentation within the calibration vehicle can also serve to slightly increase the PCF. Future efforts include further modification and optimization of the Multi-SERTTA and Multi-SERTTA-CAL designs in preparation of actual TREAT transient testing. Experimental results from both test vehicles will be compared against calculational results and methods to provide validation and support additional neutronics analyses.« less

Mitigating the Effects of the Space Radiation Environment: A Novel Approach of Using Graded-Z Materials

NASA Technical Reports Server (NTRS)

Atwell, William; Rojdev, Kristina; Aghara, Sukesh; Sriprisan, Sirikul

2013-01-01

In this paper we present a novel space radiation shielding approach using various material lay-ups, called "Graded-Z" shielding, which could optimize cost, weight, and safety while mitigating the radiation exposures from the trapped radiation and solar proton environments, as well as the galactic cosmic radiation (GCR) environment, to humans and electronics. In addition, a validation and verification (V&V) was performed using two different high energy particle transport/dose codes (MCNPX & HZETRN). Inherently, we know that materials having high-hydrogen content are very good space radiation shielding materials. Graded-Z material lay-ups are very good trapped electron mitigators for medium earth orbit (MEO) and geostationary earth orbit (GEO). In addition, secondary particles, namely neutrons, are produced as the primary particles penetrate a spacecraft, which can have deleterious effects to both humans and electronics. The use of "dopants," such as beryllium, boron, and lithium, impregnated in other shielding materials provides a means of absorbing the secondary neutrons. Several examples of optimized Graded-Z shielding layups that include the use of composite materials are presented and discussed in detail. This parametric shielding study is an extension of some earlier pioneering work we (William Atwell and Kristina Rojdev) performed in 20041 and 20092.

Development of positron annihilation spectroscopy for investigating deuterium decorated voids in neutron-irradiated tungsten

NASA Astrophysics Data System (ADS)

Taylor, C. N.; Shimada, M.; Merrill, B. J.; Akers, D. W.; Hatano, Y.

2015-08-01

The present work is a continuation of a recent research to develop and optimize positron annihilation spectroscopy (PAS) for characterizing neutron-irradiated tungsten. Tungsten samples were exposed to neutrons in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory and damaged to 0.025 and 0.3 dpa. Subsequently, they were exposed to deuterium plasmas in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory. The implanted deuterium was desorbed through sample heating to 900 °C, and Doppler broadening (DB)-PAS was performed both before and after heating. Results show that deuterium impregnated tungsten is identified as having a smaller S-parameter. The S-parameter increases after deuterium desorption. Microstructural changes also occur during sample heating. These effects can be isolated from deuterium desorption by comparing the S-parameters from the deuterium-free back face with the deuterium-implanted front face. The application of using DB-PAS to examine deuterium retention in tungsten is examined.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feder, Russell; Youssef, Mahamoud; Klabacha, Jonathan

USITER is one of seven partner domestic agencies (DA) contributing components to the ITER project. Four diagnostic port plug packages (two equatorial ports and two upper ports) will be engineered and fabricated by Princeton Plasma Physics Lab (PPPL). Diagnostic port plugs as illustrated in Fig. 1 are large primarily stainless steel structures that serve several roles on ITER. The port plugs are the primary vacuum seal and tritium confinement barriers for the vessel. The port plugs also house several plasma diagnostic systems and other machine service equipment. Finally, each port plug must shield high energy neutrons and gamma photons frommore » escaping and creating radiological problems in maintenance areas behind the port plugs. The optimization of the balance between adequate shielding and the need for high performance, high throughput diagnostics systems is the focus of this paper. Neutronics calculations are also needed for assessing nuclear heating and nuclear damage in the port plug and diagnostic components. Attila, the commercially available discrete-ordinates software package, is used for all diagnostic port plug neutronics analysis studies at PPPL.« less

Forward Helion Scattering and Neutron Polarization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buttimore, N. H.

The elastic scattering of spin half helium-3 nuclei at small angles can show a sufficiently large analyzing power to enable the level of helion polarization to be evaluated. As the helion to a large extent inherits the polarization of its unpaired neutron the asymmetry observed in helion collisions can be transformed into a measurement of the polarization of its constituent neutron. Neutron polarimetry therefore relies upon understanding the spin dependence of the electromagnetic and hadronic interactions in the region of interference where there is an optimal analyzing power.

Simulation and optimization of a new focusing polarizing bender for the diffuse neutrons scattering spectrometer DNS at MLZ

NASA Astrophysics Data System (ADS)

Nemkovski, K.; Ioffe, A.; Su, Y.; Babcock, E.; Schweika, W.; Brückel, Th

2017-06-01

We present the concept and the results of the simulations of a new polarizer for the diffuse neutron scattering spectrometer DNS at MLZ. The concept of the polarizer is based on the idea of a bender made from the stack of the silicon wafers with a double-side supermirror polarizing coating and absorbing spacers in between. Owing to its compact design, such a system provides more free space for the arrangement of other instrument components. To reduce activation of the polarizer in the high intensity neutron beam of the DNS spectrometer we plan to use the Fe/Si supermirrors instead of currently used FeCoV/Ti:N ones. Using the VITESS simulation package we have performed simulations for horizontally focusing polarizing benders with different geometries in the combination with the double-focusing crystal monochromator of DNS. Neutron transmission and polarization efficiency as well as the effects of the focusing for convergent conventional C-benders and S-benders have been analyzed both for wedge-like and plane-parallel convergent geometries of the channels. The results of these simulations and the advantages/disadvantages of the various configurations are discussed.

Optimization study of normal conductor tokamak for commercial neutron source

NASA Astrophysics Data System (ADS)

Fujita, T.; Sakai, R.; Okamoto, A.

2017-05-01

The optimum conceptual design of tokamak with normal conductor coils was studied for minimizing the cost for producing a given neutron flux by using a system code, PEC. It is assumed that the fusion neutrons are used for burning transuranics from the fission reactor spent fuel in the blanket and a fraction of the generated electric power is circulated to opearate the tokamak with moderate plasma fusion gain. The plasma performance was assumed to be moderate ones; {β\\text{N}}~∼ ~3{--}4 in the aspect ratio A~=~2{--}3 and {{H}98y2}~=~1 . The circulating power is an important factor affecting the cost. Though decreasing the aspect ratio is useful to raise the plasma beta and decrease the toroidal field, the maximum field in the coil starts to rise in the very low aspect ratio range and then the circulating power increases with decrease in the plasma aspect ratio A below A~∼ ~2 , while the construction cost increases with A . As a result, the cost per neutron has its minimum around A~∼ ~2.2 , namely, between ST and the conventional tokamak. The average circulating power fraction is expected to be ~51%.

Expected count rate for the Self- Interrogation Neutron Resonance Densitometry measurements of spent nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rossa, Riccardo; Universite libre de Bruxelles, Ecole polytechnique de Bruxelles - Service de Metrologie Nucleaire, CP 165/84, Avenue F.D. Roosevelt, 50 - B1050 Brussels; Borella, Alessandro

The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive neutron technique that aims at a direct quantification of {sup 239}Pu in the fuel assemblies by measuring the attenuation of the neutron flux in the energy region close to the 0.3 eV resonance of {sup 239}Pu. The {sup 239}Pu mass is estimated by calculating the SINRD signature, that is the ratio between the neutron flux integrated over the fast energy region and around the 0.3 eV resonance region. The SINRD measurement approach considered in this study consists in introducing a small neutron detector in the central guide tube of a PWRmore » 17x17 fuel assembly. In order to measure the neutron flux in the energy regions defined in the SINRD signature, different detector types are used. The response of a bare {sup 238}U fission chamber is considered for the determination of the fast neutron flux, while other thermal-epithermal detectors wrapped in neutron absorbers are envisaged to measure the neutron flux around the resonance region. This paper provides an estimation of the count rate that can be achieved with the detector types proposed for the SINRD measurement. In the first section a set of detectors are evaluated in terms of count rate and sensitivity to the {sup 239}Pu content, in order to identify the optimal measurement configuration for each detector type. Then a study is performed to increase the count rate by increasing the detector size. The study shows that the highest count rate is achieved by using either {sup 3}He or {sup 10}B proportional counters because of the high neutron efficiency of these detectors. However, the calculations indicate that the biggest contribution to the measurement uncertainty is due to the measurement of the fast neutron flux. Finally, similar sensitivity to the {sup 239}Pu content is obtained by using the different detector types for the measurement of the neutron flux close to the resonance region. Therefore, the count rate associated to each detector type will play a major role in the selection of the detector types used for the SINRD measurement. (authors)« less

Conceptual design study of small long-life PWR based on thorium cycle fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Subkhi, M. Nurul; Su'ud, Zaki; Waris, Abdul

2014-09-30

A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higher conversion ratio in thermal region compared to uranium cycle produce some significant of {sup 233}U during burn up time. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.3, while the multi-energy-group diffusion calculations were optimized in whole core cylindrical two-dimension R-Z geometry by SRAC-CITATION. this study would be introduced thorium nitride fuel system which ZIRLO is the cladding material. The optimization of 350 MWt small long life PWRmore » result small excess reactivity and reduced power peaking during its operation.« less

A gamma and neutron phoswich read out with SiPM for SPRD

NASA Astrophysics Data System (ADS)

Huang, Tuchen; Fu, Qibin; Yuan, Cenxi; Lin, Shaopeng

2018-02-01

A gamma and neutron phoswich was developed for spectroscopic personal radiation detectors (SPRDs). It consisted of a Φ25 × 25 mm NaI(Tl) crystal for gamma detection and a Φ25 × 3 mm LiI(Eu) crystal for neutron detection. The phoswich was read out by 8 × 8 ch SiPM array (24 × 24 mm2). Radiations in NaI(Tl) and LiI(Eu) were discriminated by pulse shape, while gammas and neutrons in LiI(Eu) were separated by pulse amplitude. For the LiI(Eu), the gamma equivalent energy for thermal neutrons was measured as 3.6 ± 0.1 MeV, providing satisfactory gamma rejection. For NaI(Tl), the response of SiPM array was well linear in the energy range up to 1408 keV, at which a deviation less than 2% was measured. Digital pulse shape discrimination (PSD) was implemented with an 8-bit digitizer running at 50 MSPS sampling rate and offline analysis. The signal pulses from NaI(Tl) and LiI(Eu) showed significant difference in falling edge allowing effective PSD. The best figure of merit (FOM) was measured as 4.4 ± 0.2 with optimized parameters, providing excellent PSD performance. The energy resolutions for 661.6 keV gamma rays in NaI(Tl) and thermal neutrons in LiI(Eu) were measured as 7.0 ± 0.2% and 11.2 ± 0.2% respectively, with selected PSD threshold.

Validation of MCNP NPP Activation Simulations for Decommissioning Studies by Analysis of NPP Neutron Activation Foil Measurement Campaigns

NASA Astrophysics Data System (ADS)

Volmert, Ben; Pantelias, Manuel; Mutnuru, R. K.; Neukaeter, Erwin; Bitterli, Beat

2016-02-01

In this paper, an overview of the Swiss Nuclear Power Plant (NPP) activation methodology is presented and the work towards its validation by in-situ NPP foil irradiation campaigns is outlined. Nuclear Research and consultancy Group (NRG) in The Netherlands has been given the task of performing the corresponding neutron metrology. For this purpose, small Aluminium boxes containing a set of circular-shaped neutron activation foils have been prepared. After being irradiated for one complete reactor cycle, the sets have been successfully retrieved, followed by gamma-spectrometric measurements of the individual foils at NRG. Along with the individual activities of the foils, the reaction rates and thermal, intermediate and fast neutron fluence rates at the foil locations have been determined. These determinations include appropriate corrections for gamma self-absorption and neutron self-shielding as well as corresponding measurement uncertainties. The comparison of the NPP Monte Carlo calculations with the results of the foil measurements is done by using an individual generic MCNP model functioning as an interface and allowing the simulation of individual foil activation by predetermined neutron spectra. To summarize, the comparison between calculation and measurement serve as a sound validation of the Swiss NPP activation methodology by demonstrating a satisfying agreement between measurement and calculation. Finally, the validation offers a chance for further improvements of the existing NPP models by ensuing calibration and/or modelling optimizations for key components and structures.

Development of Low-Cost Method for Fabrication of Metal Neutron Guides

DOE Office of Scientific and Technical Information (OSTI.GOV)

Engelhaupt, Darell; Khaykovich, Boris; Romaine, Suzanne

Neutron scattering is one of the most useful methods of studying the structure and dynamics of matter. US DOE neutron scattering research facilities at Oak Ridge National Laboratory are among the World’s most advanced, providing researchers with unmatched capabilities for probing the structure and properties of materials, including engineering and biological systems. This task is to develop a lower cost process to optimize and produce the required neutron guides capable of efficiently delivering neutron beams for tens of meters between neutron moderators and instruments. Therefore, our effort is to improve the performance and lower the production cost of neutron guides.more » Our approach aims at improving guide quality while controlling their rising costs by adopting a novel electroforming replication approach to their fabrication. These guides will be especially advantageous when used near the neutron source since the radiation resistance of nickel is superior to glass. Additionally, we are depositing low-stress nickel from an extremely low impurity solution completely free of stress-reducing agents, which nominally contain and impart sulfur, carbon and other elements that potentially activate in the neutron environment. This is achieved by using a pulsed periodically reversed current methodology. The best guides quote waviness of 0.1 mrad. It is reasonable to prepare just one mandrel of about 0.5 m long, for production of tens of guide segments, saving both the cost and supply time of guides to neutron facilities. We estimate that we can fabricate a single mandrel for the current cost of an individual one-meter guide, but from this, we can produce tens of meters of guide very inexpensively without mandrel refurbishment. While a multilayer coating will add to the overall cost, we expect this will be less than that of commercially available guides today. Therefore, we will produce higher quality guides, which are less susceptible to radiation damage, at the lower cost than those available today.« less

Development of Low-Cost Method for Fabrication of Metal Neutron Guides

DOE Office of Scientific and Technical Information (OSTI.GOV)

Engelhaupt, Darell; Khaykovich, Boris; Romaine, Suzanne

2017-12-19

Neutron scattering is one of the most useful methods of studying the structure and dynamics of matter. US DOE neutron scattering research facilities at Oak Ridge National Laboratory are among the World’s most advanced, providing researchers with unmatched capabilities for probing the structure and properties of materials, including engineering and biological systems. This task is to develop a lower cost process to optimize and produce the required neutron guides capable of efficiently delivering neutron beams for tens of meters between neutron moderators and instruments. Therefore, our effort is to improve the performance and lower the production cost of neutron guides.more » Our approach aims at improving guide quality while controlling their rising costs by adopting a novel electroforming replication approach to their fabrication. These guides will be especially advantageous when used near the neutron source since the radiation resistance of nickel is superior to glass. Additionally, we are depositing low-stress nickel from an extremely low impurity solution completely free of stress-reducing agents, which nominally contain and impart sulfur, carbon and other elements that potentially activate in the neutron environment. This is achieved by using a pulsed periodically reversed current methodology. The best guides quote waviness of 0.1 mrad. It is reasonable to prepare just one mandrel of about 0.5 m long, for production of tens of guide segments, saving both the cost and supply time of guides to neutron facilities. We estimate that we can fabricate a single mandrel for the current cost of an individual one-meter guide, but from this, we can produce tens of meters of guide very inexpensively without mandrel refurbishment. While a multilayer coating will add to the overall cost, we expect this will be less than that of commercially available guides today. Therefore, we will produce higher quality guides, which are less susceptible to radiation damage, at the lower cost than those available today.« less

Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.; Schneider, A.; Hohl, F.

1976-01-01

The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

Optimization of a Light Collection System for use in the Neutron Lifetime Project

NASA Astrophysics Data System (ADS)

Taylor, C.; O'Shaughnessy, C.; Mumm, P.; Thompson, A.; Huffman, P.

2007-10-01

The Ultracold Neutron (UCN) Lifetime Project is an ongoing experiment with the objective of improving the average measurement of the neutron beta-decay lifetime. A more accurate measurement may increase our understanding of the electroweak interaction and improve astrophysical/cosmological theories on Big Bang nucleosynthesis. The current apparatus uses 0.89 nm cold neutrons to produce UCN through inelastic collisions with superfluid 4He in the superthermal process. The lifetime of the UCN is measured by detection of scintillation light from superfluid 4He created by electrons produced in neutron decay. Competing criteria of high detection efficiency outside of the apparatus and minimum heating of the experimental cell has led to the design of an acrylic light collection system. Initial designs were based on previous generations of the apparatus. ANSYS was used to optimize the cooling system for the light guide by checking simulated end conditions based on width of contact area, number of contact points, and location on the guide itself. SolidWorks and AutoCAD were used for design. The current system is in the production process.

Evaluation of a digital data acquisition system and optimization of n-γ discrimination for a compact neutron spectrometer.

PubMed

Giacomelli, L; Zimbal, A; Reginatto, M; Tittelmeier, K

2011-01-01

A compact NE213 liquid scintillation neutron spectrometer with a new digital data acquisition (DAQ) system is now in operation at the Physikalisch-Technische Bundesanstalt (PTB). With the DAQ system, developed by ENEA Frascati, neutron spectrometry with high count rates in the order of 5×10(5) s(-1) is possible, roughly an order of magnitude higher than with an analog acquisition system. To validate the DAQ system, a new data analysis code was developed and tests were done using measurements with 14-MeV neutrons made at the PTB accelerator. Additional analysis was carried out to optimize the two-gate method used for neutron and gamma (n-γ) discrimination. The best results were obtained with gates of 35 ns and 80 ns. This indicates that the fast and medium decay time components of the NE213 light emission are the ones that are relevant for n-γ discrimination with the digital acquisition system. This differs from what is normally implemented in the analog pulse shape discrimination modules, namely, the fast and long decay emissions of the scintillating light.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fu, Wenkai; Ghosh, Priyarshini; Harrison, Mark

The performance of traditional Hornyak buttons and two proposed variants for fast-neutron hodoscope applications was evaluated using Geant4. The Hornyak button is a ZnS(Ag)-based device previously deployed at the Idaho National Laboratory's TRansient REActor Test Facility (better known as TREAT) for monitoring fast neutrons emitted during pulsing of fissile fuel samples. Past use of these devices relied on pulse-shape discrimination to reduce the significant levels of background Cherenkov radiation. Proposed are two simple designs that reduce the overall light guide mass (here, polymethyl methacrylate or PMMA), employ silicon photomultipliers (SiPMs), and can be operated using pulse-height discrimination alone to eliminatemore » background noise to acceptable levels. Geant4 was first used to model a traditional Hornyak button, and for assumed, hodoscope-like conditions, an intrinsic efficiency of 0.35% for mono-directional fission neutrons was predicted. The predicted efficiency is in reasonably good agreement with experimental data from the literature and, hence, served to validate the physics models and approximations employed. Geant4 models were then developed to optimize the materials and geometries of two alternatives to the Hornyak button, one based on a homogeneous mixture of ZnS(Ag) and PMMA, and one based on alternating layers of ZnS(Ag) and PMMA oriented perpendicular to the incident neutron beam. For the same radiation environment, optimized, 5-cm long (along the beam path) devices of the homogeneous and layered designs were predicted to have efficiencies of approximately 1.3% and 3.3%, respectively. For longer devices, i.e., lengths larger than 25 cm, these efficiencies were shown to peak at approximately 2.2% and 5.9%, respectively. Furthermore, both designs were shown to discriminate Cherenkov noise intrinsically by using an appropriate pulse-height discriminator level, i.e., pulse-shape discrimination is not needed for these devices.« less

Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Indah Rosidah, M., E-mail: indah.maymunah@gmail.com; Suud, Zaki, E-mail: szaki@fi.itb.ac.id; Yazid, Putranto Ilham

The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With themore » tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature and high pressure condition for more than ten years.« less

Medical Isotope Production Analyses In KIPT Neutron Source Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Talamo, Alberto; Gohar, Yousry

Medical isotope production analyses in Kharkov Institute of Physics and Technology (KIPT) neutron source facility were performed to include the details of the irradiation cassette and the self-shielding effect. An updated detailed model of the facility was used for the analyses. The facility consists of an accelerator-driven system (ADS), which has a subcritical assembly using low-enriched uranium fuel elements with a beryllium-graphite reflector. The beryllium assemblies of the reflector have the same outer geometry as the fuel elements, which permits loading the subcritical assembly with different number of fuel elements without impacting the reflector performance. The subcritical assembly is drivenmore » by an external neutron source generated from the interaction of 100-kW electron beam with a tungsten target. The facility construction was completed at the end of 2015, and it is planned to start the operation during the year of 2016. It is the first ADS in the world, which has a coolant system for removing the generated fission power. Argonne National Laboratory has developed the design concept and performed extensive design analyses for the facility including its utilization for the production of different radioactive medical isotopes. 99Mo is the parent isotope of 99mTc, which is the most commonly used medical radioactive isotope. Detailed analyses were performed to define the optimal sample irradiation location and the generated activity, for several radioactive medical isotopes, as a function of the irradiation time.« less

Optimization of the photoneutron target geometry for e-accelerator based BNCT

PubMed Central

Chegeni, Nahid; Pur, Saleh Boveiry; Razmjoo, Sasan; Hoseini, Seydeh Khadijed

2017-01-01

Background and aim Today, electron accelerators are taken into consideration as photoneutron sources. Therefore, for maximum production of epithermal neutron flux, designing a photoneutron target is of significant importance. In this paper, the effect of thickness and geometric shape of a photoneutron target on neutron output were investigated. Methods In this study, a pencil photon source with 13, 15, 18, 20 and 25 MeV energies and a diameter of 2 mm was investigated using Monte Carlo simulation method using MCNP code. To optimize the design of the photoneutron target, the tungsten target with various geometries and thicknesses was investigated. Results The maximum neutron flux produced for all target geometries and thicknesses occurred at neutron energy peak of around 0.46 MeV. As the thickness increased to 2 cm, neutron flux increased and then a decreasing trend was observed. For various geometrical shapes, the determining factor in photoneutron output was the effective target thickness in the photon interaction path that increased by the increase in the area of interaction. Another factor was the angle of the photon’s incidence with the target surface that resulted in a significant decrease in photoneutron output in cone-shaped targets Conclusion Three factors including the total neutron flux, neutrons energy spectrum, and convergence of neutrons plays an important role in the selection of geometry and shape of the target that should be investigated considering beam shaping assembly (BSA) shape. PMID:28848635

A 13C(d,n)-based epithermal neutron source for Boron Neutron Capture Therapy.

PubMed

Capoulat, M E; Kreiner, A J

2017-01-01

Boron Neutron Capture Therapy (BNCT) requires neutron sources suitable for in-hospital siting. Low-energy particle accelerators working in conjunction with a neutron producing reaction are the most appropriate choice for this purpose. One of the possible nuclear reactions is 13 C(d,n) 14 N. The aim of this work is to evaluate the therapeutic capabilities of the neutron beam produced by this reaction, through a 30mA beam of deuterons of 1.45MeV. A Beam Shaping Assembly design was computationally optimized. Depth dose profiles in a Snyder head phantom were simulated with the MCNP code for a number of BSA configurations. In order to optimize the treatment capabilities, the BSA configuration was determined as the one that allows maximizing both the tumor dose and the penetration depth while keeping doses to healthy tissues under the tolerance limits. Significant doses to tumor tissues were achieved up to ∼6cm in depth. Peak doses up to 57Gy-Eq can be delivered in a fractionated scheme of 2 irradiations of approximately 1h each. In a single 1h irradiation, lower but still acceptable doses to tumor are also feasible. Treatment capabilities obtained here are comparable to those achieved with other accelerator-based neutron sources, making of the 13 C(d,n) 14 N reaction a realistic option for producing therapeutic neutron beams through a low-energy particle accelerator. Copyright © 2016 Associazione Italiana di Fisica Medica. Published by Elsevier Ltd. All rights reserved.

Sensitivity Analysis of Cf-252 (sf) Neutron and Gamma Observables in CGMF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carter, Austin Lewis; Talou, Patrick; Stetcu, Ionel

CGMF is a Monte Carlo code that simulates the decay of primary fission fragments by emission of neutrons and gamma rays, according to the Hauser-Feshbach equations. As the CGMF code was recently integrated into the MCNP6.2 transport code, great emphasis has been placed on providing optimal parameters to CGMF such that many different observables are accurately represented. Of these observables, the prompt neutron spectrum, prompt neutron multiplicity, prompt gamma spectrum, and prompt gamma multiplicity are crucial for accurate transport simulations of criticality and nonproliferation applications. This contribution to the ongoing efforts to improve CGMF presents a study of the sensitivitymore » of various neutron and gamma observables to several input parameters for Californium-252 spontaneous fission. Among the most influential parameters are those that affect the input yield distributions in fragment mass and total kinetic energy (TKE). A new scheme for representing Y(A,TKE) was implemented in CGMF using three fission modes, S1, S2 and SL. The sensitivity profiles were calculated for 17 total parameters, which show that the neutron multiplicity distribution is strongly affected by the TKE distribution of the fragments. The total excitation energy (TXE) of the fragments is shared according to a parameter RT, which is defined as the ratio of the light to heavy initial temperatures. The sensitivity profile of the neutron multiplicity shows a second order effect of RT on the mean neutron multiplicity. A final sensitivity profile was produced for the parameter alpha, which affects the spin of the fragments. Higher values of alpha lead to higher fragment spins, which inhibit the emission of neutrons. Understanding the sensitivity of the prompt neutron and gamma observables to the many CGMF input parameters provides a platform for the optimization of these parameters.« less

Characterization of neutron emission from mega-ampere deuterium gas puff Z-pinch at microsecond implosion times

NASA Astrophysics Data System (ADS)

Klir, D.; Shishlov, A. V.; Kokshenev, V. A.; Kubes, P.; Labetsky, A. Yu; Rezac, K.; Cikhardt, J.; Fursov, F. I.; Kovalchuk, B. M.; Kravarik, J.; Kurmaev, N. E.; Ratakhin, N. A.; Sila, O.; Stodulka, J.

2013-08-01

Experiments with deuterium (D2) triple shell gas puffs were carried out on the GIT-12 generator at a 3 MA current level and microsecond implosion times. The outer, middle and inner nozzle diameters were 160 mm, 80 mm and 30 mm, respectively. The influence of the mass of deuterium shells on neutron emission times, neutron yields and neutron energy spectra was studied. The injected linear mass of deuterium varied between 50 and 255 µg cm-1. Gas puffs imploded onto the axis before the peak of generator current at 700-1100 ns. Most of the neutrons were emitted during the second neutron pulse after the development of instabilities. Despite higher currents, heavier gas puffs produced lower neutron yields. Optimal mass and a short time delay between the valve opening and the generator triggering were more important than the better coincidence of stagnation with peak current. The peak neutron yield from D(d, n)3He reactions reached 3 × 1011 at 2.8 MA current, 90 µg cm-1 injected linear mass and 37 mm anode-cathode gap. In the case of lower mass shots, a large number of 10 MeV neutrons were produced either by secondary DT reactions or by DD reactions of deuterons with energies above 7 MeV. The average neutron yield ratio Y>10 MeV/Y2.5 MeV reached (6 ± 3) × 10-4. Such a result can be explained by a power law distribution for deuterons as \\rmd N_d/\\rmd E_d\\propto E_d^{-3} . The optimization of a D2 gas puff Z-pinch and similarities to a plasma focus and its drive parameter are described.

Performance of an RPM based on Gd-lined plastic scintillator for neutron and gamma detection [ANIMMA--2015-IO-372

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fanchini, Erica

A Radiation Portal Monitor (RPM) was developed by the Istituto Nazionale di Fisica Nucleare (INFN) and Ansaldo Nucleare (ANN) within the FP7 SCINTILLA European project. The system was designed to detect both gamma and neutron radiation with a single technology. It is conceived to monitor vehicle and cargo containers in transits across borders or ports, to find radioactive elements and to avoid illegal trafficking of strategic nuclear materials. The system is based on a {sup 3}He-free neutron detection technology using plastic scintillators coupled to Gadolinium to detect and discriminate gamma from neutron signals. During the 3 years of the SCINTILLAmore » project the construction and test of the first two prototypes drove the definition of the final layout of a full RPM system consisting of two twin pillars as a portal for vehicle and cargo container scan. A custom System Control Software (SCS) manages the electronics of the RPM, the ancillary devices and the data analysis. The combination of the detector layout and of the software functionalities enables both to distinguish neutrons and gammas and to identify the energy range of a detected gamma source. The system was initially characterized via static tests with gamma and neutron sources in the INFN laboratory. These measurements were used to calibrate the detector, evaluate the response of the single pillars as well as of the full system, and optimize the RPM configuration and discrimination algorithm. During this phase, specific tests were performed to study the stability over time of the system, monitoring the measured the neutron and gamma count rates over periods of several weeks. The results allow us to demonstrate the reliability and robustness of the RPM. In a second time the RPM performance was studied via dynamic tests performed during the SCINTILLA test and benchmark campaigns. These measurements took place in the JRC ITRAP+10 facility at Ispra (Varese-Italy). The laboratory is equipped with an experimental set-up for dynamic tests of multiple systems according to international standards. The performed measurements utilized radioactive sources with activities selected according to ANSI and IEC standards to test the detector alarm performances in terms of gamma and neutron response, sensitivity to high gamma fields, sensitivity to moderated neutron sources as well as false alarm rates (FAR). In addition, the RPM was tested in challenging configurations exceeding the requirements set by international standards to determine the real limits of the system. The results obtained during these campaigns demonstrated that the system detection efficiency is not only compliant to international standards for its category, but often exceeds them, demonstrating the validity of the chosen technology and of the implemented layout. The positive performance also showed the effectiveness of the SCS and of its functionalities. To further demonstrate the system capabilities, a test in a real-life environment of the RPM is planned to happen in a near future by installing the detectors in a seaport. In this presentation I will give an overview of the RPM characteristics, of its performances as determined in the test campaign mentioned above and of future plans, to demonstrate how this technology can be an effective choice for the realization of {sup 3}He-free RPM detectors. (authors)« less

Lithium Alkaline Halides—Next Generation of Dual Mode Scintillators

NASA Astrophysics Data System (ADS)

Soundara-Pandian, L.; Hawrami, R.; Glodo, J.; Ariesanti, E.; van Loef, E. V.; Shah, K.

2016-04-01

We report on a new family of scintillators - Lithium alkaline halides, developed based on the alkaline halides by introducing lithium for dual mode gamma-neutron detection. Many different compositions were grown, among which LiSr2I5 (LSI), LiCa2I5 (LCI), LiSr2Br5 (LSB) activated with divalent Europium show good gamma and neutron detection properties. LSI shows the main emission at 497 nm under X-ray excitation. It also shows good proportionality, which in combination with the light yield as high as 60000 photons/MeV, results in an energy resolution of 3.5% at 662 keV. The electron or gamma equivalent energy (GEE) of the thermal neutron peak due to the 6Li neutron capture is 4.1 MeV, which amounts to a very high neutron light yield of 245000 photons. The decay times for neutrons are faster compared to that for gamma-rays, hence we achieved good pulse shape discrimination (PSD) between gamma and neutron events. Our initial studies on the effects of Eu concentration on the properties of LSI show that 3%-4% Eu concentration is optimal for the best performance in terms of gamma and neutron light yields and pulse shape discrimination. LCI shows the main emission at 475 nm under X-ray excitation and a very high gamma light yield of 90000 photons/MeV. The measured energy resolution is 6% at 662 keV. The electron equivalent energy for neutron detection has been measured to be around 3 MeV, which gives a neutron light yield of 270 000 photons. The measured decay times for neutrons are faster compared to gamma decays and the PSD between the gamma-rays and neutrons is not as good as LSI. LSB shows two emissions at 410 and 475 nm under X-ray excitation. The measured light yield is 32000 ph/MeV gamma-ray with an energy resolution of 6% at 662 keV. The electron equivalent energy of the 6Li capture peak was measured to be 3.3 MeV.

Characterization of the neutron irradiation system for use in the Low-Dose-Rate Irradiation Facility at Sandia National Laboratories.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Franco, Manuel

The objective of this work was to characterize the neutron irradiation system consisting of americium-241 beryllium (241AmBe) neutron sources placed in a polyethylene shielding for use at Sandia National Laboratories (SNL) Low Dose Rate Irradiation Facility (LDRIF). With a total activity of 0.3 TBq (9 Ci), the source consisted of three recycled 241AmBe sources of different activities that had been combined into a single source. The source in its polyethylene shielding will be used in neutron irradiation testing of components. The characterization of the source-shielding system was necessary to evaluate the radiation environment for future experiments. Characterization of the sourcemore » was also necessary because the documentation for the three component sources and their relative alignment within the Special Form Capsule (SFC) was inadequate. The system consisting of the source and shielding was modeled using Monte Carlo N-Particle transport code (MCNP). The model was validated by benchmarking it against measurements using multiple techniques. To characterize the radiation fields over the full spatial geometry of the irradiation system, it was necessary to use a number of instruments of varying sensitivities. First, the computed photon radiography assisted in determining orientation of the component sources. With the capsule properly oriented inside the shielding, the neutron spectra were measured using a variety of techniques. A N-probe Microspec and a neutron Bubble Dosimeter Spectrometer (BDS) set were used to characterize the neutron spectra/field in several locations. In the third technique, neutron foil activation was used to ascertain the neutron spectra. A high purity germanium (HPGe) detector was used to characterize the photon spectrum. The experimentally measured spectra and the MCNP results compared well. Once the MCNP model was validated to an adequate level of confidence, parametric analyses was performed on the model to optimize for potential experimental configurations and neutron spectra for component irradiation. The final product of this work is a MCNP model validated by measurements, an overall understanding of neutron irradiation system including photon/neutron transport and effective dose rates throughout the system, and possible experimental configurations for future irradiation of components.« less

Targeted Modification of Neutron Energy Spectra for National Security Applications

NASA Astrophysics Data System (ADS)

Bevins, James Edward

At its core, research represents an attempt to break from the "this is the way we have always done it" paradigm. This idea is evidenced from the start in this research effort by the problem formulation to develop a new way to generate synthetic debris that mimics the samples that would be collected for forensics purposes following a nuclear weapon attack on the U.S. or its allies. The philosophy is also demonstrated by the design methodology used to solve the synthetic debris problem, using methods not commonly applied to nuclear engineering problems. Through this research, the bounds of what is deemed possible in neutron spectral shaping are moved ever so slightly. A capability for the production of synthetic debris and fission products was developed for the National Ignition Facility (NIF). Synthetic debris has historically been made in a limited fashion using sample doping techniques since the cessation of nuclear weapons testing, but a more robust alternative approach using neutron spectral shaping was proposed and developed by the University of California-Berkeley and Lawrence Livermore National Laboratory (LLNL). Using NIF as a starting source spectrum, the energy tuning assembly (ETA) developed in this work can irradiate samples with a combined thermonuclear and prompt fission neutron spectrum (TN+PFNS). When used with fissile foils, this irradiation will produce a synthetic fission product distribution that is realistic across all mass chains. To design the ETA, traditional parametric point design approaches were discarded in favor of formal optimization techniques. Finding a lack of suitable algorithms in the literature, a metaheuristic-based optimization algorithm, Gnowee, was developed for rapid convergence to nearly globally optimum solutions for complex, constrained engineering problems with mixed-integer and combinatorial design vectors and high-cost, noisy, discontinuous, black box objective function evaluations. Comparisons between Gnowee and several well-established metaheuristic algorithms are made for a set of continuous, mixed-integer, and combinatorial benchmarks. These results demonstrated Gnoweee to have superior flexibility and convergence characteristics over a wide range of design spaces. The Gnowee algorithm was implemented in Coeus, a new piece of software, to perform optimization of design problems requiring radiation transport for the evaluation of their objective functions. Currently, Coeus solves ETA optimization problems using hybrid radiation transport (ADVANTG and MCNP) to assess design permutations developed by Gnowee. Future enhancements of Coeus will look to expand the geometries and objective functions considered to those beyond ETA design. Coeus was used to generate an ETA design for the TN+PFNS application on NIF. The design achieved a reasonable match with the objective TN+PFNS and associated fission product distributions within the size and weight constraints imposed by the NIF facility. The ETA design was built by American Elements, and initial validation tests were conducted at the Lawrence Berkeley National Laboratory's 88-Inch Cyclotron. These experiments used foil activation and pulse height spectroscopy to measure the ETA-modified spectrum. Additionally, pulse height spectroscopy measurements were taken as the ETA was built-up component-by-component to measure the impact of nuclear data on the ability to model the ETA performance. Some initial analysis of these results is included here. Finally, an integral validation experiment on NIF was proposed using the Coeus generated ETA design. A scoping study conducted by LLNL determined the proposed experiment and ETA design are within NIF facility limitations and current radio-chemistry capabilities. The study found that the proposed ETA experiment was "low risk," has "no show stoppers," and has a "reasonable cost." All that is needed is a sponsor to close the last funding gap and bring the experiment to fruition. This research broke with the current sample doping approach and applied neutron spectral shaping to design an ETA that can create realistic synthetic fission and activation products and improve technical nuclear forensics outcomes. However, the ETA presented in this research represents more than a stand alone point design with a limited scope and application. It is proof of a concept and the product of a unique capability that has a wide range of potential applications. This research demonstrates that the concept of neutron spectral shaping can be used to engineer complex neutron spectra within the confines of physics. There are many possible applications that could benefit from the ability to generate custom energy neutron spectra that fall outside of current sources and methods. The ETA is the product of a general-purpose optimization algorithm, Gnowee, and design framework, Coeus, which enables the use of Gnowee for complex nuclear design problems. Through Gnowee and Coeus, new ETA neutronics designs can be generated in days, not months or years, with a drastic reduction in the research effort required to do so. (Abstract shortened by ProQuest.).

An optimization methodology for heterogeneous minor actinides transmutation

NASA Astrophysics Data System (ADS)

Kooyman, Timothée; Buiron, Laurent; Rimpault, Gérald

2018-04-01

In the case of a closed fuel cycle, minor actinides transmutation can lead to a strong reduction in spent fuel radiotoxicity and decay heat. In the heterogeneous approach, minor actinides are loaded in dedicated targets located at the core periphery so that long-lived minor actinides undergo fission and are turned in shorter-lived fission products. However, such targets require a specific design process due to high helium production in the fuel, high flux gradient at the core periphery and low power production. Additionally, the targets are generally manufactured with a high content in minor actinides in order to compensate for the low flux level at the core periphery. This leads to negative impacts on the fuel cycle in terms of neutron source and decay heat of the irradiated targets, which penalize their handling and reprocessing. In this paper, a simplified methodology for the design of targets is coupled with a method for the optimization of transmutation which takes into account both transmutation performances and fuel cycle impacts. The uncertainties and performances of this methodology are evaluated and shown to be sufficient to carry out scoping studies. An illustration is then made by considering the use of moderating material in the targets, which has a positive impact on the minor actinides consumption but a negative impact both on fuel cycle constraints (higher decay heat and neutron) and on assembly design (higher helium production and lower fuel volume fraction). It is shown that the use of moderating material is an optimal solution of the transmutation problem with regards to consumption and fuel cycle impacts, even when taking geometrical design considerations into account.

Detection Of Special Nuclear Materials Tagged Neutrons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Deyglun, Clement; Perot, Bertrand; Carasco, Cedric

In order to detect Special Nuclear Materials (SNM) in unattended luggage or cargo containers in the field of homeland security, fissions are induced by 14 MeV neutrons produced by an associated particle DT neutron generator, and prompt fission particles correlated with tagged neutron are detected by plastic scintillators. SMN produce high multiplicity events due to induced fissions, whereas nonnuclear materials produce low multiplicity events due to cross-talk, (n,2n) or (n,n'γ) reactions. The data acquisition electronics is made of compact FPGA boards. The coincidence window is triggered by the alpha particle detection, allowing to tag the emission date and direction ofmore » the 14 MeV interrogating neutron. The first part of the paper presents experiment vs. calculation comparisons to validate MCNP-PoliMi simulations and the post-processing tools developed with the data analysis framework ROOT. Measurements have been performed using different targets (iron, lead, graphite), first with small plastic scintillators (10 x 10 x 10 cm{sup 3}) and then with large detectors (10 x 10 x 100 cm{sup 3}) to demonstrate that nuclear materials can be differentiated from nonnuclear dense materials (iron, lead) in iron and wood matrixes. A special attention is paid on SNM detection in abandoned luggage. In the second part of the paper, the performances of a cargo container inspection system are studied by numerical simulation, following previous work reported in. Detectors dimensions and shielding against the neutron generator background are optimized for container inspection. Events not correlated to an alpha particle (uncorrelated background), counting statistics, time and energy resolutions of the data acquisition system are all taken into account in a realistic numerical model. The impact of the container matrix (iron, ceramic, wood) has been investigated by studying the system capability to detect a few kilograms of SNM in different positions in the cargo container, within 10 min acquisitions. (authors)« less

The neutron flux monitors from 20 keV to 1 MeV for BNCT: performance study using accelerator-based neutron sources

NASA Astrophysics Data System (ADS)

Guan, X. C.; Gong, Y.; Murata, I.; Wang, T. S.

2018-05-01

The performance of the neutron flux monitors from 20 keV to 1 MeV developed for boron neutron capture therapy (BNCT) is studied by Monte Carlo simulations using accelerator-based neutron sources (ABNSs). The results show that the performance of the neutron flux monitors is very satisfactory and they can be efficiently used in practical applications to measure the neutron fluxes from 20 keV to 1 MeV of ABNSs for BNCT to high accuracy.

Shielding implications for secondary neutrons and photons produced within the patient during IMPT.

PubMed

DeMarco, J; Kupelian, P; Santhanam, A; Low, D

2013-07-01

Intensity modulated proton therapy (IMPT) uses a combination of computer controlled spot scanning and spot-weight optimized planning to irradiate the tumor volume uniformly. In contrast to passive scattering systems, secondary neutrons and photons produced from inelastic proton interactions within the patient represent the major source of emitted radiation during IMPT delivery. Various published studies evaluated the shielding considerations for passive scattering systems but did not directly address secondary neutron production from IMPT and the ambient dose equivalent on surrounding occupational and nonoccupational work areas. Thus, the purpose of this study was to utilize Monte Carlo simulations to evaluate the energy and angular distributions of secondary neutrons and photons following inelastic proton interactions within a tissue-equivalent phantom for incident proton spot energies between 70 and 250 MeV. Monte Carlo simulation methods were used to calculate the ambient dose equivalent of secondary neutrons and photons produced from inelastic proton interactions in a tissue-equivalent phantom. The angular distribution of emitted neutrons and photons were scored as a function of incident proton energy throughout a spherical annulus at 1, 2, 3, 4, and 5 m from the phantom center. Appropriate dose equivalent conversion factors were applied to estimate the total ambient dose equivalent from secondary neutrons and photons. A reference distance of 1 m from the center of the patient was used to evaluate the mean energy distribution of secondary neutrons and photons and the resulting ambient dose equivalent. For an incident proton spot energy of 250 MeV, the total ambient dose equivalent (3.6 × 10(-3) mSv per proton Gy) was greatest along the direction of the incident proton spot (0°-10°) with a mean secondary neutron energy of 71.3 MeV. The dose equivalent decreased by a factor of 5 in the backward direction (170°-180°) with a mean energy of 4.4 MeV. An 8 × 8 × 8 cm(3) volumetric spot distribution (5 mm FWHM spot size, 4 mm spot spacing) optimized to produce a uniform dose distribution results in an ambient dose equivalent of 4.5 × 10(-2) mSv per proton Gy in the forward direction. This work evaluated the secondary neutron and photon emission due to monoenergetic proton spots between 70 and 250 MeV, incident on a tissue equivalent phantom. Example calculations were performed to estimate concrete shield thickness based upon appropriate workload and shielding design assumptions. Although lower than traditional passive scattered proton therapy systems, the ambient dose equivalent from secondary neutrons produced by the patient during IMPT can be significant relative to occupational and nonoccupational workers in the vicinity of the treatment vault. This work demonstrates that Monte Carlo simulations are useful as an initial planning tool for studying the impact of the treatment room and maze design on surrounding occupational and nonoccupational work areas.

Sensitivity Studies and Experimental Evaluation for Optimizing Transcurium Isotope Production

DOE PAGES

Hogle, Susan L.; Alexander, Charles W.; Burns, Jonathan D.; ...

2017-03-01

This work applies to recent initiatives at the Radiochemical Engineering Development Center at Oak Ridge National Laboratory to optimize the production of transcurium isotopes in the High Flux Isotope Reactor in such a way as to prolong the use of high quality heavy curium feedstock. By studying the sensitivity of fission and transmutation reaction rates to the neutron flux spectrum a means of increasing the fraction of (n,γ) reactions per neutron absorption is explored. Filter materials which preferentially absorb neutrons at energies considered detrimental to optimal transcurium production are identified and transmutation rates are examined with high energy resolution. Experimentalmore » capsules are irradiated employing filter materials and the resulting fission and activation products studied to validate the filtering methodology. Improvement is seen in the production efficiency of heavier curium isotopes in 244Cm and 245Cm targets, and potentially in production of 252Cf from mixed californium targets. Finally, further analysis is recommended to evaluate longer duration irradiations more representative of typical transcurium production.« less

SU-E-T-567: Neutron Dose Equivalent Evaluation for Pencil Beam Scanning Proton Therapy with Apertures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Geng, C; Nanjing University of Aeronautics and Astronautics, Nanjing; Schuemann, J

Purpose: To determine the neutron contamination from the aperture in pencil beam scanning during proton therapy. Methods: A Monte Carlo based proton therapy research platform TOPAS and the UF-series hybrid pediatric phantoms were used to perform this study. First, pencil beam scanning (PBS) treatment pediatric plans with average spot size of 10 mm at iso-center were created and optimized for three patients with and without apertures. Then, the plans were imported into TOPAS. A scripting method was developed to automatically replace the patient CT with a whole body phantom positioned according to the original plan iso-center. The neutron dose equivalentmore » was calculated using organ specific quality factors for two phantoms resembling a 4- and 14-years old patient. Results: The neutron dose equivalent generated by the apertures in PBS is 4–10% of the total neutron dose equivalent for organs near the target, while roughly 40% for organs far from the target. Compared to the neutron dose equivalent caused by PBS without aperture, the results show that the neutron dose equivalent with aperture is reduced in the organs near the target, and moderately increased for those organs located further from the target. This is due to the reduction of the proton dose around the edge of the CTV, which causes fewer neutrons generated in the patient. Conclusion: Clinically, for pediatric patients, one might consider adding an aperture to get a more conformal treatment plan if the spot size is too large. This work shows the somewhat surprising fact that adding an aperture for beam scanning for facilities with large spot sizes reduces instead of increases a potential neutron background in regions near target. Changran Geng is supported by the Chinese Scholarship Council (CSC) and the National Natural Science Foundation of China (Grant No. 11475087)« less

Advancing Materials Science using Neutrons at Oak Ridge National Laboratory

ScienceCinema

Carpenter, John

2018-02-14

Jack Carpenter, pioneer of accelerator-based pulsed spallation neutron sources, talks about neutron science at Oak Ridge National Laboratory (ORNL) and a need for a second target station at the Spallation Neutron Source (SNS). ORNL is the Department of Energy's largest multiprogram science and energy laboratory, and is home to two scientific user facilities serving the neutron science research community: the High Flux Isotope Reactor (HFIR) and SNS. HFIR and SNS provide researchers with unmatched capabilities for understanding the structure and properties of materials, macromolecular and biological systems, and the fundamental physics of the neutron. Neutrons provide a window through which to view materials at a microscopic level that allow researchers to develop better materials and better products. Neutrons enable us to understand materials we use in everyday life. Carpenter explains the need for another station to produce long wavelength neutrons, or cold neutrons, to answer questions that are addressed only with cold neutrons. The second target station is optimized for that purpose. Modern technology depends more and more upon intimate atomic knowledge of materials, and neutrons are an ideal probe.

NEUTRON STAR MASS–RADIUS CONSTRAINTS USING EVOLUTIONARY OPTIMIZATION

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stevens, A. L.; Morsink, S. M.; Fiege, J. D.

The equation of state of cold supra-nuclear-density matter, such as in neutron stars, is an open question in astrophysics. A promising method for constraining the neutron star equation of state is modeling pulse profiles of thermonuclear X-ray burst oscillations from hot spots on accreting neutron stars. The pulse profiles, constructed using spherical and oblate neutron star models, are comparable to what would be observed by a next-generation X-ray timing instrument like ASTROSAT , NICER , or a mission similar to LOFT . In this paper, we showcase the use of an evolutionary optimization algorithm to fit pulse profiles to determinemore » the best-fit masses and radii. By fitting synthetic data, we assess how well the optimization algorithm can recover the input parameters. Multiple Poisson realizations of the synthetic pulse profiles, constructed with 1.6 million counts and no background, were fitted with the Ferret algorithm to analyze both statistical and degeneracy-related uncertainty and to explore how the goodness of fit depends on the input parameters. For the regions of parameter space sampled by our tests, the best-determined parameter is the projected velocity of the spot along the observer’s line of sight, with an accuracy of ≤3% compared to the true value and with ≤5% statistical uncertainty. The next best determined are the mass and radius; for a neutron star with a spin frequency of 600 Hz, the best-fit mass and radius are accurate to ≤5%, with respective uncertainties of ≤7% and ≤10%. The accuracy and precision depend on the observer inclination and spot colatitude, with values of ∼1% achievable in mass and radius if both the inclination and colatitude are ≳60°.« less

Towards radiation hard converter material for SiC-based fast neutron detectors

NASA Astrophysics Data System (ADS)

Tripathi, S.; Upadhyay, C.; Nagaraj, C. P.; Venkatesan, A.; Devan, K.

2018-05-01

In the present work, Geant4 Monte-Carlo simulations have been carried out to study the neutron detection efficiency of the various neutron to other charge particle (recoil proton) converter materials. The converter material is placed over Silicon Carbide (SiC) in Fast Neutron detectors (FNDs) to achieve higher neutron detection efficiency as compared to bare SiC FNDs. Hydrogenous converter material such as High-Density Polyethylene (HDPE) is preferred over other converter materials due to the virtue of its high elastic scattering reaction cross-section for fast neutron detection at room temperature. Upon interaction with fast neutrons, hydrogenous converter material generates recoil protons which liberate e-hole pairs in the active region of SiC detector to provide a detector signal. The neutron detection efficiency offered by HDPE converter is compared with several other hydrogenous materials viz., 1) Lithium Hydride (LiH), 2) Perylene, 3) PTCDA . It is found that, HDPE, though providing highest efficiency among various studied materials, cannot withstand high temperature and harsh radiation environment. On the other hand, perylene and PTCDA can sustain harsh environments, but yields low efficiency. The analysis carried out reveals that LiH is a better material for neutron to other charge particle conversion with competent efficiency and desired radiation hardness. Further, the thickness of LiH has also been optimized for various mono-energetic neutron beams and Am-Be neutron source generating a neutron fluence of 109 neutrons/cm2. The optimized thickness of LiH converter for fast neutron detection is found to be ~ 500 μm. However, the estimated efficiency for fast neutron detection is only 0.1%, which is deemed to be inadequate for reliable detection of neutrons. A sensitivity study has also been done investigating the gamma background effect on the neutron detection efficiency for various energy threshold of Low-Level Discriminator (LLD). The detection efficiency of a stacked structure concept has been explored by juxtaposing several converter-detector layers to improve the efficiency of LiH-SiC-based FNDs . It is observed that approximately tenfold efficiency improvement has been achieved—0.93% for ten layers stacked configuration vis-à-vis 0.1% of single converter-detector layer configuration. Finally, stacked detectors have also been simulated for different converter thicknesses to attain the efficiency as high as ~ 3.25% with the help of 50 stacked layers.

Calculations to support JET neutron yield calibration: Modelling of neutron emission from a compact DT neutron generator

NASA Astrophysics Data System (ADS)

Čufar, Aljaž; Batistoni, Paola; Conroy, Sean; Ghani, Zamir; Lengar, Igor; Milocco, Alberto; Packer, Lee; Pillon, Mario; Popovichev, Sergey; Snoj, Luka; JET Contributors

2017-03-01

At the Joint European Torus (JET) the ex-vessel fission chambers and in-vessel activation detectors are used as the neutron production rate and neutron yield monitors respectively. In order to ensure that these detectors produce accurate measurements they need to be experimentally calibrated. A new calibration of neutron detectors to 14 MeV neutrons, resulting from deuterium-tritium (DT) plasmas, is planned at JET using a compact accelerator based neutron generator (NG) in which a D/T beam impinges on a solid target containing T/D, producing neutrons by DT fusion reactions. This paper presents the analysis that was performed to model the neutron source characteristics in terms of energy spectrum, angle-energy distribution and the effect of the neutron generator geometry. Different codes capable of simulating the accelerator based DT neutron sources are compared and sensitivities to uncertainties in the generator's internal structure analysed. The analysis was performed to support preparation to the experimental measurements performed to characterize the NG as a calibration source. Further extensive neutronics analyses, performed with this model of the NG, will be needed to support the neutron calibration experiments and take into account various differences between the calibration experiment and experiments using the plasma as a source of neutrons.

PENTrack - a versatile Monte Carlo tool for ultracold neutron sources and experiments

NASA Astrophysics Data System (ADS)

Picker, Ruediger; Chahal, Sanmeet; Christopher, Nicolas; Losekamm, Martin; Marcellin, James; Paul, Stephan; Schreyer, Wolfgang; Yapa, Pramodh

2016-09-01

Ultracold neutrons have energies in the hundred nano eV region. They can be stored in traps for hundreds of seconds. This makes them the ideal tool to study the neutron itself. Measurements of neutron decay correlations, lifetime or electric dipole moment are ideally suited for ultracold neutrons, as well as experiments probing the neutron's gravitational levels in the earth's field. We have developed a Monte Carlo simulation tool that can serve to design and optimize these experiments, and possibly correct results: PENTrack is a C++ based simulation code that tracks neutrons, protons and electrons or atoms, as well as their spins, in gravitational and electromagnetic fields. In addition wall interactions of neutrons due to strong interaction are modeled with a Fermi-potential formalism and take surface roughness into account. The presentation will introduce the physics behind the simulation and provide examples of its application.

Performance verification of an epithermal neutron flux monitor using accelerator-based BNCT neutron sources

NASA Astrophysics Data System (ADS)

Guan, X.; Murata, I.; Wang, T.

2017-09-01

The performance of an epithermal neutron flux monitor developed for boron neutron capture therapy (BNCT) is verified by Monte Carlo simulations using accelerator-based neutron sources (ABNSs). The results indicate that the developed epithermal neutron flux monitor works well and it can be efficiently used in practical applications to measure the epithermal neutron fluxes of ABNSs in a high accuracy.

Extracting equation of state parameters from black hole-neutron star mergers: Aligned-spin black holes and a preliminary waveform model

NASA Astrophysics Data System (ADS)

Lackey, Benjamin D.; Kyutoku, Koutarou; Shibata, Masaru; Brady, Patrick R.; Friedman, John L.

2014-02-01

Information about the neutron-star equation of state is encoded in the waveform of a black hole-neutron star system through tidal interactions and the possible tidal disruption of the neutron star. During the inspiral this information depends on the tidal deformability Λ of the neutron star, and we find that the best-measured parameter during the merger and ringdown is consistent with Λ as well. We performed 134 simulations where we systematically varied the equation of state as well as the mass ratio, neutron star mass, and aligned spin of the black hole. Using these simulations we develop an analytic representation of the full inspiral-merger-ringdown waveform calibrated to these numerical waveforms; we use this analytic waveform and a Fisher matrix analysis to estimate the accuracy to which Λ can be measured with gravitational-wave detectors. We find that although the inspiral tidal signal is small, coherently combining this signal with the merger-ringdown matter effect improves the measurability of Λ by a factor of ˜3 over using just the merger-ringdown matter effect alone. However, incorporating correlations between all the waveform parameters then decreases the measurability of Λ by a factor of ˜3. The uncertainty in Λ increases with the mass ratio, but decreases as the black hole spin increases. Overall, a single Advanced LIGO detector can only marginally measure Λ for mass ratios Q =2-5, black hole spins JBH/MBH2=-0.5-0.75, and neutron star masses MNS=1.2M⊙-1.45M⊙ at an optimally oriented distance of 100 Mpc. For the proposed Einstein Telescope, however, the uncertainty in Λ is an order of magnitude smaller.

Development of the dense plasma focus for short-pulse applications

NASA Astrophysics Data System (ADS)

Bennett, N.; Blasco, M.; Breeding, K.; Constantino, D.; DeYoung, A.; DiPuccio, V.; Friedman, J.; Gall, B.; Gardner, S.; Gatling, J.; Hagen, E. C.; Luttman, A.; Meehan, B. T.; Misch, M.; Molnar, S.; Morgan, G.; O'Brien, R.; Robbins, L.; Rundberg, R.; Sipe, N.; Welch, D. R.; Yuan, V.

2017-01-01

The dense plasma focus (DPF) has long been considered a compact source for pulsed neutrons and has traditionally been optimized for the total neutron yield. In this paper, we describe the efforts to optimize the DPF for short-pulse applications by introducing a reentrant cathode at the end of the coaxial plasma gun. The resulting neutron pulse widths are reduced by an average of 21 ±9 % from the traditional long-drift DPF design. Pulse widths and yields achieved from deuterium-tritium fusion at 2 MA are 61.8 ±30.7 ns FWHM and 1.84 ±0.49 ×1012 neutrons per shot. Simulations were conducted concurrently to elucidate the DPF operation and confirm the role of the reentrant cathode. A hybrid fluid-kinetic particle-in-cell modeling capability demonstrates correct sheath velocities, plasma instabilities, and fusion yield rates. Consistent with previous findings that the DPF is dominated by beam-target fusion from superthermal ions, we estimate that the thermonuclear contribution is at the 1% level.

Treatment planning capability assessment of a beam shaping assembly for accelerator-based BNCT.

PubMed

Herrera, M S; González, S J; Burlon, A A; Minsky, D M; Kreiner, A J

2011-12-01

Within the frame of an ongoing project to develop a folded Tandem-Electrostatic-Quadrupole accelerator facility for Accelerator-Based Boron Neutron Capture Therapy (AB-BNCT) a theoretical study was performed to assess the treatment planning capability of different configurations of an optimized beam shaping assembly for such a facility. In particular this study aims at evaluating treatment plans for a clinical case of Glioblastoma. Copyright © 2011 Elsevier Ltd. All rights reserved.

Processing, Fabrication, Characterization and Device Demonstration of High Temperature Superconducting Ceramics

DTIC Science & Technology

1994-07-30

optimize processes for grain alignment in bulk and tape samples; and (4) provide a technology base for utilization of flux-trap magnets . SUXIARY a) A...of the vortices in the bulk material. Neutron scattering experiments can be performed in a magnetic field range of -0.05 T up to several teslas, a...uncorrected for demagnetization ) was then taken as the field associated with the first nonzero value of magnetization -difference, AM; 3ee Figs.5&6

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2015-10-01

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the powermore » coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.« less

Neutron capture measurement on {sup 173}Lu at LANSCE with DANCE detector

DOE Office of Scientific and Technical Information (OSTI.GOV)

Theroine, C.; Ebran, A.; Meot, V.

2013-06-10

The (n,{gamma}) cross section on the unstable {sup 173}Lu(t{sub 1/2} = 1.37y) has been measured from thermal energy up to 200 eV at Los Alamos Neutron Science Center (LANSCE) with The Detector for Advanced Neutron Capture Experiements (DANCE). The main aim of this study is to validate and optimize reaction models for unstable nucleus. A preliminary capture yield will be presented in this paper.

Thermal Neutron Radiography using a High-flux Compact Neutron Generator

NASA Astrophysics Data System (ADS)

Taylor, Michael; Sengbusch, Evan; Seyfert, Chris; Moll, Eli; Radel, Ross

A novel neutron imaging system has been designed and constructed by Phoenix Nuclear Labs to investigate specimens when conventional X-ray imaging will not suffice. A first-generation electronic neutron generator is actively being used by the United States Army and is coupled with activation films for neutron radiography to inspect munitions and other critical defence and aerospace components. A second-generation system has been designed to increase the total neutron output from an upgraded gaseous deuterium target to 5×1011 DD n/s, generating higher neutron flux at the imaging plane and dramatically reducing interrogation time, while maintaining high spatial resolution and low geometric unsharpness. A description of the neutron generator and imaging system, including the beamline, target and detector platform, is given in this paper. State of the art neutron moderators, collimators and imaging detector components are also discussed in the context of increasing specimen throughput and optimizing image quality. Neutron radiographs captured with the neutron radiography system will be further compared against simulated images using the MCNP nuclear simulation code.

Residual stress analysis on tensile MMC specimens after loading/unloading tests in several conditions

NASA Astrophysics Data System (ADS)

Giuliani, Alessandra; Albertini, Gianni; Manescu, Adrian

2004-07-01

Residual stresses have been investigated in samples made of AA6061+22% Al2O3 in order to correlate microstructural characteristics with mechanical performances. In particular, the possible occurrence of a brittle fracture induced by an excessive load transfer from the matrix to the reinforcement was investigated. To this end, macrostresses and microstresses were analysed. A neutron diffraction test on 12 specimens submitted to several loading/unloading conditions at different temperatures was performed. These measurements aimed to establish the optimal temperature for the initial extruded billet in pre-heating stage, before forging the final wheel hub.

Development of a lithium fluoride zinc sulfide based neutron multiplicity counter

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cowles, Christian; Behling, Spencer; Baldez, Phoenix

Here, the feasibility of a full-scale lithium fluoride zinc sulfide (LiF/ZnS) based neutron multiplicity counter has been demonstrated. The counter was constructed of modular neutron detecting stacks that each contain five sheets of LiF/ZnS interleaved between six sheets of wavelength shifting plastic with a photomultiplier tube on each end. Twelve such detector stacks were placed around a sample chamber in a square arrangement with lithiated high-density polyethylene blocks in the corners to reflect high-energy neutrons and capture low-energy neutrons. The final system design was optimized via modeling and small-scale test. Measuring neutrons from a 252Cf source, the counter achieved amore » 36% neutron detection efficiency (ϵϵ) and an View the MathML source11.7μs neutron die-away time (ττ) for a doubles figure-of-merit (ϵ 2/τ) of 109. This is the highest doubles figure-of-merit measured to-date for a 3He-free neutron multiplicity counter.« less

Development of a lithium fluoride zinc sulfide based neutron multiplicity counter

NASA Astrophysics Data System (ADS)

Cowles, Christian; Behling, Spencer; Baldez, Phoenix; Folsom, Micah; Kouzes, Richard; Kukharev, Vladislav; Lintereur, Azaree; Robinson, Sean; Siciliano, Edward; Stave, Sean; Valdez, Patrick

2018-04-01

The feasibility of a full-scale lithium fluoride zinc sulfide (LiF/ZnS) based neutron multiplicity counter has been demonstrated. The counter was constructed of modular neutron detecting stacks that each contain five sheets of LiF/ZnS interleaved between six sheets of wavelength shifting plastic with a photomultiplier tube on each end. Twelve such detector stacks were placed around a sample chamber in a square arrangement with lithiated high-density polyethylene blocks in the corners to reflect high-energy neutrons and capture low-energy neutrons. The final system design was optimized via modeling and small-scale test. Measuring neutrons from a 252Cf source, the counter achieved a 36% neutron detection efficiency (ɛ) and an 11 . 7 μs neutron die-away time (τ) for a doubles figure-of-merit (ɛ2 / τ) of 109. This is the highest doubles figure-of-merit measured to-date for a 3He-free neutron multiplicity counter.

Development of a lithium fluoride zinc sulfide based neutron multiplicity counter

DOE PAGES

Cowles, Christian; Behling, Spencer; Baldez, Phoenix; ...

2018-01-12

Here, the feasibility of a full-scale lithium fluoride zinc sulfide (LiF/ZnS) based neutron multiplicity counter has been demonstrated. The counter was constructed of modular neutron detecting stacks that each contain five sheets of LiF/ZnS interleaved between six sheets of wavelength shifting plastic with a photomultiplier tube on each end. Twelve such detector stacks were placed around a sample chamber in a square arrangement with lithiated high-density polyethylene blocks in the corners to reflect high-energy neutrons and capture low-energy neutrons. The final system design was optimized via modeling and small-scale test. Measuring neutrons from a 252Cf source, the counter achieved amore » 36% neutron detection efficiency (ϵϵ) and an View the MathML source11.7μs neutron die-away time (ττ) for a doubles figure-of-merit (ϵ 2/τ) of 109. This is the highest doubles figure-of-merit measured to-date for a 3He-free neutron multiplicity counter.« less

Possible Experiment for the Demonstration of Neutron Waves Interaction with Spatially Oscillating Potential

NASA Astrophysics Data System (ADS)

Miloi, Mădălina Mihaela; Goryunov, Semyon; Kulin, German

2018-04-01

A wide range of problems in neutron optics is well described by a theory based on application of the effective potential model. It was assumed that the concept of the effective potential in neutron optics have a limited region of validity and ceases to be correct in the case of the giant acceleration of a matter. To test this hypothesis a new Ultra Cold neutron experiment for the observation neutron interaction with potential structure oscillating in space was proposed. The report is focused on the model calculations of the topography of sample surface that oscillate in space. These calculations are necessary to find an optimal parameters and geometry of the planned experiment.

Examination of total cross section resonance structure of niobium and silicon in neutron transmission experiments

NASA Astrophysics Data System (ADS)

Andrianova, Olga; Lomakov, Gleb; Manturov, Gennady

2017-09-01

The neutron transmission experiments are one of the main sources of information about the neutron cross section resonance structure and effect in the self-shielding. Such kind of data for niobium and silicon nuclides in energy range 7 keV to 3 MeV can be obtained from low-resolution transmission measurements performed earlier in Russia (with samples of 0.027 to 0.871 atom/barn for niobium and 0.076 to 1.803 atom/barn for silicon). A significant calculation-to-experiment discrepancy in energy range 100 to 600 keV and 300 to 800 keV for niobium and silicon, respectively, obtained using the evaluated nuclear data library ROSFOND, were found. The EVPAR code was used for estimation the average resonance parameters in energy range 7 to 600 keV for niobium. For silicon a stochastic optimization method was used to modify the resolved resonance parameters in energy range 300 to 800 keV. The improved ROSFOND evaluated nuclear data files were tested in calculation of ICSBEP integral benchmark experiments.

Designing new guides and instruments using McStas

NASA Astrophysics Data System (ADS)

Farhi, E.; Hansen, T.; Wildes, A.; Ghosh, R.; Lefmann, K.

With the increasing complexity of modern neutron-scattering instruments, the need for powerful tools to optimize their geometry and physical performances (flux, resolution, divergence, etc.) has become essential. As the usual analytical methods reach their limit of validity in the description of fine effects, the use of Monte Carlo simulations, which can handle these latter, has become widespread. The McStas program was developed at Riso National Laboratory in order to provide neutron scattering instrument scientists with an efficient and flexible tool for building Monte Carlo simulations of guides, neutron optics and instruments [1]. To date, the McStas package has been extensively used at the Institut Laue-Langevin, Grenoble, France, for various studies including cold and thermal guides with ballistic geometry, diffractometers, triple-axis, backscattering and time-of-flight spectrometers [2]. In this paper, we present some simulation results concerning different guide geometries that may be used in the future at the Institut Laue-Langevin. Gain factors ranging from two to five may be obtained for the integrated intensities, depending on the exact geometry, the guide coatings and the source.

Characterization of multilayer Thick-GEM geometries as 10B converters aiming thermal neutron detection

NASA Astrophysics Data System (ADS)

Natal da Luz, H.; Souza, F. A.; Moralles, M.; Carlin, N.; Oliveira, R. A. N.; Bregant, M.; Suaide, A. A. P.; Chubaci, J. F. D.; Matsuoka, M.; Silva, T. F.; Moro, M. V.; Rodrigues, C. L.; Munhoz, M. G.

2018-02-01

Boron-based thermal neutron detectors have recently regained some attention from the instrumentation community as a strong alternative to helium-3 detectors. From the existing concepts exploiting boron layers in position sensitive detectors, the Cascade [1] is the one that takes full advantage of the 2D capabilities of gaseous detectors, with the position resolution not limited by the architecture of the detector. In this work, a proposal for the Cascade detector, based on Thick-GEMs is presented, together with some preliminary studies of the suitable pitch that optimizes the neutron conversion efficiency, while keeping the collection efficiency intact. The characterization of Thick-GEM prototypes produced in Brazil with hole pitch from 0.75 to 3 mm shows that these devices already present a stable performance at low gains, also resulting in fair energy resolution, when cascaded with a standard KaptonTM 50 µm GEM. Results of the first attempts of boron film depositions with Ion Beam Assisted Deposition and characterization by Ion Beam Analysis are also presented.

Neutron Detection in the A2 Collaboration Experiment on Neutral Pion Photo-production on Neutron

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bulychjov, S. A.; Kudryavtsev, A. E.; Kulikov, V. V.

Neutron detection is of crucial importance for the neutral pion photo-production study on a neutron target that now is in progress at MAMI. Two electro-magnetic calorimeters, based on NaI and BaF 2 crystals, are used in the A2 experiment. While these calorimeters are optimized for pion decay photon detection, they have a reason able efficiency for neutron detection also. The paper describes the method, which has been used to measure this efficiency using the same data taken for pion photo-production study on deuterium target with tagged photon been of 800 MeV maximal energy. As a result, the detection efficiency ismore » a rising function of neutron momentum that reaches 40% near 1 GeV/c.« less

Study of low energy neutron beam formation based on GEANT4 simulations

NASA Astrophysics Data System (ADS)

Avagyan, R.; Avetisyan, R.; Ivanyan, V.; Kerobyan, I.

2017-07-01

The possibility of obtaining thermal/epithermal energy neutron beams using external protons from cyclotron C18/18 is studied based on GEANT4 simulations. This study will be the basis of the Beam Shaped Assembly (BSA) development for future Boron Neutron Capture Therapy (BNCT). Proton induced reactions on 9Be target are considered as a neutron source, and dependence of neutron yield on target thickness is investigated. The problem of reducing the ratio of gamma to neutron yields by inserting a lead sheet after the beryllium target is studied as well. By GEANT4 modeling the optimal thicknesses of 9Be target and lead absorber are determined and the design characteristics of beam shaping assembly, including the materials and thicknesses of reflector and moderator are considered.

Neutron Detection in the A2 Collaboration Experiment on Neutral Pion Photo-production on Neutron

DOE PAGES

Bulychjov, S. A.; Kudryavtsev, A. E.; Kulikov, V. V.; ...

2018-04-09

Neutron detection is of crucial importance for the neutral pion photo-production study on a neutron target that now is in progress at MAMI. Two electro-magnetic calorimeters, based on NaI and BaF 2 crystals, are used in the A2 experiment. While these calorimeters are optimized for pion decay photon detection, they have a reason able efficiency for neutron detection also. The paper describes the method, which has been used to measure this efficiency using the same data taken for pion photo-production study on deuterium target with tagged photon been of 800 MeV maximal energy. As a result, the detection efficiency ismore » a rising function of neutron momentum that reaches 40% near 1 GeV/c.« less

Detecting Shielded Special Nuclear Materials Using Multi-Dimensional Neutron Source and Detector Geometries

NASA Astrophysics Data System (ADS)

Santarius, John; Navarro, Marcos; Michalak, Matthew; Fancher, Aaron; Kulcinski, Gerald; Bonomo, Richard

2016-10-01

A newly initiated research project will be described that investigates methods for detecting shielded special nuclear materials by combining multi-dimensional neutron sources, forward/adjoint calculations modeling neutron and gamma transport, and sparse data analysis of detector signals. The key tasks for this project are: (1) developing a radiation transport capability for use in optimizing adaptive-geometry, inertial-electrostatic confinement (IEC) neutron source/detector configurations for neutron pulses distributed in space and/or phased in time; (2) creating distributed-geometry, gas-target, IEC fusion neutron sources; (3) applying sparse data and noise reduction algorithms, such as principal component analysis (PCA) and wavelet transform analysis, to enhance detection fidelity; and (4) educating graduate and undergraduate students. Funded by DHS DNDO Project 2015-DN-077-ARI095.

Characterization of γ-ray background at IMAT beamline of ISIS Spallation Neutron Source

NASA Astrophysics Data System (ADS)

Festa, G.; Andreani, C.; Arcidiacono, L.; Burca, G.; Kockelmann, W.; Minniti, T.; Senesi, R.

2017-08-01

The environmental γ -ray background on the IMAT beamline at ISIS Spallation Neutron Source, Target Station 2, is characterized via γ spectroscopy. The measurements include gamma exposure at the imaging detector position, along with the gamma background inside the beamline. Present results are discussed and compared with previous measurements recorded at INES and VESUVIO beamlines operating at Target Station 1. They provide new outcome for expanding and optimizing the PGAA experimental capability at the ISIS neutron source for the investigation of materials, engineering components and cultural heritage objects at the ISIS neutron source.

Replacing a 252Cf source with a neutron generator in a shuffler - a conceptual design performed with MCNPX

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schear, Melissa A; Tobin, Stephen J

2009-01-01

The {sup 252}Cf shuffler has been widely used in nuclear safeguards and radioactive waste management to assay fissile isotopes, such as {sup 235}U or {sup 239}Pu, present in a variety of samples, ranging from small cans of uranium waste to metal samples weighing several kilograms. Like other non-destructive assay instruments, the shuffler uses an interrogating neutron source to induce fissions in the sample. Although shufflers with {sup 252}Cf sources have been reliably used for several decades, replacing this isotopic source with a neutron generator presents some distinct advantages. Neutron generators can be run in a continuous or pulsed mode, andmore » may be turned off, eliminating the need for shielding and a shuffling mechanism in the shuffler. There is also essentially no dose to personnel during installation, and no reliance on the availability of {sup 252}Cf. Despite these advantages, the more energetic neutrons emitted from the neutron generator (141 MeV for D-T generators) present some challenges for certain material types. For example when the enrichment of a uranium sample is unknown, the fission of {sup 238}U is generally undesirable. Since measuring uranium is one of the main uses of a shuffler, reducing the delayed neutron contribution from {sup 238}U is desirable. Hence, the shuffler hardware must be modified to accommodate a moderator configuration near the source to tailor the interrogating spectrum in a manner which promotes sub-threshold fissions (below 1 MeV) but avoids the over-moderation of the interrogating neutrons so as to avoid self-shielding. In this study, where there are many material and geometry combinations, the Monte Carlo N-Particle eXtended (MCNPX) transport code was used to model, design, and optimize the moderator configuration within the shuffler geometry. The code is then used to evaluate and compare the assay performances of both the modified shuffler and the current {sup 252}Cf shuffler designs for different test samples. The matrix effect and the non-uniformity of the interrogating flux are investigated and quantified in each case. The modified geometry proposed by this study can serve s a guide in retrofitting shufflers that are already in use.« less

Moderator Demonstration Facility Design and Optimization

DOE Office of Scientific and Technical Information (OSTI.GOV)

McClanahan, Tucker C.; Gallmeier, Franz X.; Iverson, Erik B.

2017-02-01

The Spallation Neutron Source (SNS) facility at Oak Ridge National Laboratory (ORNL) is implementing a Moderator Demonstration Facility (MDF) to demonstrate the performance characteristics of advanced moderators central to the Second Target Station (STS) for SNS. The MDF will use the "spare" front-end installation within the SNS accelerator support complex – an ion source, radio-frequency quadrupole (RFQ) accelerator, and medium-energy beam transport (MEBT) chopper - to provide a 2.5 MeV proton beam of peak current 50 mA and maximum pulse length of less than 10 s at a repetition rate of no more than 60 Hz to a suitable neutron-producingmore » target to demonstrate those aspects of moderator performance necessary to meet the goals of the STS design e ort. The accelerator beam parameters are not open to variation beyond that described above - they are fixed by the nature of the spare front-end installation (the Integrated Test Stand Facility; ITSF). Accordingly, there are some neutronic challenges in developing prototypic moderator illumination from a very non-prototypic primary neutron source; the spallation source we are attempting to mimic has an extended neutron source volume approximately 40 cm long (in the direction of the proton beam), approximately 10 cm wide (horizontally transverse to the proton beam) and approximately 5 cm high (vertically transverse to the proton beam), and an isotropic evaporation energy spectrum with mean energy above 1 MeV. In contrast, the primary neutron source available from the 7Li(p,n) reaction (the most prolific at 2.5 MeV proton energy by more than an order of magnitude) is strongly anisotropic, with an energy spectrum that is both strongly dependent on emission angle and kinematically limited to less than 700 keV, and the interaction zone between the incident protons and any target material (neutron-producing or not) is intrinsically limited to a few tens of microns. The MDF will be unique and innovative amongst the world’s moderator test facilities in that it will be intended to be very prototypic in terms of "moderator illumination" - the spatial variation of the neutron flux entering the moderator itself - as well as capable of testing so-called high-brightness moderators in a wing configuration.« less

Prompt fission neutron spectra from fission induced by 1 to 8 MeV neutrons on U235 and Pu239 using the double time-of-flight technique

NASA Astrophysics Data System (ADS)

Noda, S.; Haight, R. C.; Nelson, R. O.; Devlin, M.; O'Donnell, J. M.; Chatillon, A.; Granier, T.; Bélier, G.; Taieb, J.; Kawano, T.; Talou, P.

2011-03-01

Prompt fission neutron spectra from U235 and Pu239 were measured for incident neutron energies from 1 to 200 MeV at the Weapons Neutron Research facility (WNR) of the Los Alamos Neutron Science Center, and the experimental data were analyzed with the Los Alamos model for the incident neutron energies of 1-8 MeV. A CEA multiple-foil fission chamber containing deposits of 100 mg U235 and 90 mg Pu239 detected fission events. Outgoing neutrons were detected by the Fast Neutron-Induced γ-Ray Observer array of 20 liquid organic scintillators. A double time-of-flight technique was used to deduce the neutron incident energies from the spallation target and the outgoing energies from the fission chamber. These data were used for testing the Los Alamos model, and the total kinetic energy parameters were optimized to obtain a best fit to the data. The prompt fission neutron spectra were also compared with the Evaluated Nuclear Data File (ENDF/B-VII.0). We calculate average energies from both experimental and calculated fission neutron spectra.

A compact DD neutron generator-based NAA system to quantify manganese (Mn) in bone in vivo.

PubMed

Liu, Yingzi; Byrne, Patrick; Wang, Haoyu; Koltick, David; Zheng, Wei; Nie, Linda H

2014-09-01

A deuterium-deuterium (DD) neutron generator-based neutron activation analysis (NAA) system has been developed to quantify metals, including manganese (Mn), in bone in vivo. A DD neutron generator with a flux of up to 3*10(9) neutrons s(-1) was set up in our lab for this purpose. Optimized settings, including moderator, reflector, and shielding material and thickness, were selected based on Monte Carlo (MC) simulations conducted in our previous work. Hand phantoms doped with different Mn concentrations were irradiated using the optimized DD neutron generator irradiation system. The Mn characteristic γ-rays were collected by an HPGe detector system with 100% relative efficiency. The calibration line of the Mn/calcium (Ca) count ratio versus bone Mn concentration was obtained (R(2) = 0.99) using the hand phantoms. The detection limit (DL) was calculated to be about 1.05 μg g(-1) dry bone (ppm) with an equivalent dose of 85.4 mSv to the hand. The DL can be reduced to 0.74 ppm by using two 100% HPGe detectors. The whole body effective dose delivered to the irradiated subject was calculated to be about 17 μSv. Given the average normal bone Mn concentration of 1 ppm in the general population, this system is promising for in vivo bone Mn quantification in humans.

Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi

2013-11-29

This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implementmore » a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.« less

Plasma focus sources: Supplement to the Neutron Resonance Radiography Workshop proceedings

NASA Astrophysics Data System (ADS)

Nardi, Vittorio; Brzosko, Jan

1989-01-01

Since their discovery, plasma focus discharges have been recognized as very intense pulsed sources of deuterium-deuterium (D-D) or deuterium-tritium (D-T) fusion-reaction neutrons, with outstanding capabilities. Specifically, the total neutron emission/shot, Y (sub n), and the rate of neutron emission, Y (sub n), of an optimized plasma focus (PF) are higher than the corresponding quantities observed in any other type of pinched discharge at the same level of powering energy W (sub 0). Recent developments have led to the concept and experimental demonstration of an Advanced Plasma Focus System (APF) that consists of a Mather-geometry plasma focus in which field distortion elements (FDEs) are inserted in the inter-electrode gap for increasing the neutron yield/shot, Y (sub n). The FDE-induced redistribution of the plasma current increases Y (sub n) by a factor approximate to or greater than 5 to 10 above the value obtained without FDEs under otherwise identical conditions of operation of the plasma focus. For example, an APF that is fed by a fast capacitor bank with an energy, W (sub 0) = 6 kJ, and voltage, V (sub 0) = 16.5 kV provides Y (sub n) congruent to 4 x 10 to the 9th D-D neutrons/shot (pure D2 filling) and Y (sub n) = 4 x 10 to the 11th D-T neutrons/shot (filling is 50 pct deuterium and 50 pct tritium). The FDE-induced increase of Y (sub n) for fixed values of (W sub 0, V sub 0), the observed scaling law Y (sub n) proportional to W (sub 0) squared for optimized plasma focus systems, and our experience with neutron scattering in bulk objects lead us to the conclusion that we can use an APF as a source of high-intensity neutron pulses (10 to the 14th n/pulse) in the field off neutron radiography (surface and bulk) with a nanosecond or millisecond time resolution.

SU-G-IeP4-04: DD-Neutron Source Collimation for Neutron Stimulated Emission Computed Tomography: A Monte Carlo Simulation Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fong, G; Kapadia, A

Purpose: To optimize collimation and shielding for a deuterium-deuterium (DD) neutron generator for an inexpensive and compact clinical neutron imaging system. The envisioned application is cancer diagnosis through Neutron Stimulated Emission Computed Tomography (NSECT). Methods: Collimator designs were tested with an isotropic 2.5 MeV neutron source through GEANT4 simulations. The collimator is a 52×52×52 cm{sup 3} polyethylene block coupled with a 1 cm lead sheet in sequence. Composite opening was modeled into the collimator to permit passage of neutrons. The opening varied in shape (cylindrical vs. tapered), size (1–5 cm source-side and target-side openings) and aperture placements (13–39 cm frommore » source-side). Spatial and energy distribution of neutrons and gammas were tracked from each collimator design. Parameters analyzed were primary beam width (FWHM), divergence, and efficiency (percent transmission) for different configurations of the collimator. Select resultant outputs were then used for simulated NSECT imaging of a virtual breast phantom containing a 2.5 cm diameter tumor to assess the effect of the collimator on spatial resolution, noise, and scan time. Finally, composite shielding enclosure made of polyethylene and lead was designed and evaluated to block 99.99% of neutron and gamma radiation generated in the system. Results: Analysis of primary beam indicated the beam-width is linear to the aperture size. Increasing source-side opening allowed at least 20% more neutron throughput for all designs relative to the cylindrical openings. Maximum throughput for all designs was 364% relative to cylindrical openings. Conclusion: The work indicates potential for collimating and shielding a DD neutron generator for use in a clinical NSECT system. The proposed collimator designs produced a well-defined collimated neutron beam that can be used to image samples of interest with millimeter resolution. Balance in output efficiency, noise reduction, and scan time should be considered to determine the optimal design for specific NSECT applications.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oktamuliani, Sri, E-mail: srioktamuliani@ymail.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id

A preliminary study designs SPINNOR (Small Power Reactor, Indonesia, No On-Site Refueling) liquid metal Pb-Bi cooled fast reactors, fuel (U, Pu)N, 150 MWth have been performed. Neutronic calculation uses SRAC which is designed cylindrical core 2D (R-Z) 90 × 135 cm, on the core fuel composed of heterogeneous with percentage difference of PuN 10, 12, 13% and the result of calculation is effective neutron multiplication 1.0488. Power density distribution of the output SRAC is generated for thermal hydraulic calculation using Delphi based on Pascal language that have been developed. The research designed a reactor that is capable of natural circulation atmore » inlet temperature 300 °C with variation of total mass flow rate. Total mass flow rate affect pressure drop and temperature outlet of the reactor core. The greater the total mass flow rate, the smaller the outlet temperature, but increase the pressure drop so that the chimney needed more higher to achieve natural circulation or condition of the system does not require a pump. Optimization of the total mass flow rate produces optimal reactor design on the total mass flow rate of 5000 kg/s with outlet temperature 524,843 °C but require a chimney of 6,69 meters.« less

Optimization of high count rate event counting detector with Microchannel Plates and quad Timepix readout

NASA Astrophysics Data System (ADS)

Tremsin, A. S.; Vallerga, J. V.; McPhate, J. B.; Siegmund, O. H. W.

2015-07-01

Many high resolution event counting devices process one event at a time and cannot register simultaneous events. In this article a frame-based readout event counting detector consisting of a pair of Microchannel Plates and a quad Timepix readout is described. More than 104 simultaneous events can be detected with a spatial resolution of 55 μm, while >103 simultaneous events can be detected with <10 μm spatial resolution when event centroiding is implemented. The fast readout electronics is capable of processing >1200 frames/sec, while the global count rate of the detector can exceed 5×108 particles/s when no timing information on every particle is required. For the first generation Timepix readout, the timing resolution is limited by the Timepix clock to 10-20 ns. Optimization of the MCP gain, rear field voltage and Timepix threshold levels are crucial for the device performance and that is the main subject of this article. These devices can be very attractive for applications where the photon/electron/ion/neutron counting with high spatial and temporal resolution is required, such as energy resolved neutron imaging, Time of Flight experiments in lidar applications, experiments on photoelectron spectroscopy and many others.

The 14 MeV Neutron Irradiation Facility in MARIA Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prokopowicz, R.; Pytel, K.; Dorosz, M.

2015-07-01

The MARIA reactor with thermal neutron flux density up to 3x10{sup 14} cm{sup -2} s{sup -1} and a number of vertical channels is well suited to material testing by thermal neutron treatment. Beside of that some fast neutron irradiation facilities are operated in MARIA reactor as well. One of them is thermal to 14 MeV neutron converter launched in 2014. It is especially devoted to fusion devices material testing irradiation. The ITER and DEMO research thermonuclear facilities are to be run using the deuterium - tritium fusion reaction. Fast neutrons (of energy approximately 14 MeV) resulting from the reaction aremore » essential to carry away the released thermonuclear energy and to breed tritium. However, constructional materials of which thermonuclear reactors are to be built must be specially selected to survive intense fluxes of fast neutrons. Strong sources of 14 MeV neutrons are needed if research on resistance of candidate materials to such fluxes is to be carried out effectively. Nuclear reactor-based converter capable to convert thermal neutrons into 14 MeV fast neutrons may be used to that purpose. The converter based on two stage nuclear reaction on lithium-6 and deuterium compounds leading to 14 MeV neutron production. The reaction chain is begun by thermal neutron capture by lithium-6 nucleus resulted in triton release. The neutron and triton transport calculations have been therefore carried-out to estimate the thermal to 14 MeV neutron conversion efficiency and optimize converter construction. The usable irradiation space of ca. 60 cm{sup 3} has been obtained. The released energy have been calculated. Heat transport has been asses to ensure proper device cooling. A set of thermocouples has been installed in converter to monitor its temperature distribution on-line. Influence of converter on reactor operation has been studied. Safety analyses of steady states and transients have been done. Performed calculations and analyses allow designing the converter and formulate its operation limits and conditions. During first tested operation of the converter the 14 MeV neutron flux density was estimated to 10{sup 9} cm{sup -2} s{sup -1}, whereas fast fission neutrons inside converter achieved 10{sup 12} cm{sup -2} s{sup -1}, and thermal neutrons were reduced down to 109 cm-2 s-1. Taking into account the feasibility of almost incessant converter operation for a number of months, its arisen as one of the most powerful (in terms of fluence), currently available 14 MeV neutron source. Such a converter currently under operation in the MARIA reactor core will be presented. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marques, J.G.; Ramos, A.R.; Fernandes, A.C.

The behavior of electronic components and circuits under radiation is a concern shared by the nuclear industry, the space community and the high-energy physics community. Standard commercial components are used as much as possible instead of radiation hard components, since they are easier to obtain and allow a significant reduction of costs. However, these standard components need to be tested in order to determine their radiation tolerance. The Portuguese Research Reactor (RPI) is a 1 MW pool-type reactor, operating since 1961. The irradiation of electronic components and circuits is one area where a 1 MW reactor can be competitive, sincemore » the fast neutron fluences required for testing are in most cases well below 10{sup 16} n/cm{sup 2}. A program was started in 1999 to test electronics components and circuits for the LHC facility at CERN, initially using a dedicated in-pool irradiation device and later a beam line with tailored neutron and gamma filters. Neutron filters are essential to reduce the intensity of the thermal neutron flux, which does not produce significant defects in electronic components but produces unwanted radiation from activation of contacts and packages of integrated circuits and also of the printed circuit boards. In irradiations performed within the line-of-sight of the core of a fission reactor there is simultaneous gamma radiation which complicates testing in some cases. Filters can be used to reduce its importance and separate testing with a pure gamma radiation source can contribute to clarify some irradiation results. Practice has shown the need to introduce several improvements to the procedures and facilities over the years. We will review improvements done in the following areas: - Optimization of neutron and gamma filters; - Dosimetry procedures in mixed neutron / gamma fields; - Determination of hardness parameter and 1 MeV-equivalent neutron fluence; - Temperature measurement and control during irradiation; - Follow-up of reactor power operational fluctuations; - Study of gamma radiation effects only. The fission neutron spectrum can be limitative for some of the tests, as most neutrons are in the 1-2 MeV energy range. Significant progress has been made lately in compact neutron generators using D-D and D-T fusion reactions, achieving higher neutron fluxes and longer lifetime than previously available. The advantages of using compact neutron generators for testing of electronic components and circuits will be also discussed. (authors)« less

Three-dimensional neutronics optimization of helium-cooled blanket for multi-functional experimental fusion-fission hybrid reactor (FDS-MFX)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, J.; Yuan, B.; Jin, M.

2012-07-01

Three-dimensional neutronics optimization calculations were performed to analyse the parameters of Tritium Breeding Ratio (TBR) and maximum average Power Density (PDmax) in a helium-cooled multi-functional experimental fusion-fission hybrid reactor named FDS (Fusion-Driven hybrid System)-MFX (Multi-Functional experimental) blanket. Three-stage tests will be carried out successively, in which the tritium breeding blanket, uranium-fueled blanket and spent-fuel-fueled blanket will be utilized respectively. In this contribution, the most significant and main goal of the FDS-MFX blanket is to achieve the PDmax of about 100 MW/m3 with self-sustaining tritium (TBR {>=} 1.05) based on the second-stage test with uranium-fueled blanket to check and validate themore » demonstrator reactor blanket relevant technologies based on the viable fusion and fission technologies. Four different enriched uranium materials were taken into account to evaluate PDmax in subcritical blanket: (i) natural uranium, (ii) 3.2% enriched uranium, (iii) 19.75% enriched uranium, and (iv) 64.4% enriched uranium carbide. These calculations and analyses were performed using a home-developed code VisualBUS and Hybrid Evaluated Nuclear Data Library (HENDL). The results showed that the performance of the blanket loaded with 64.4% enriched uranium was the most attractive and it could be promising to effectively obtain tritium self-sufficiency (TBR-1.05) and a high maximum average power density ({approx}100 MW/m{sup 3}) when the blanket was loaded with the mass of {sup 235}U about 1 ton. (authors)« less

Neutron measurement at the thermal column of the Malaysian Triga Mark II reactor using gold foil activation method and TLD

NASA Astrophysics Data System (ADS)

Shalbi, Safwan; Salleh, Wan Norhayati Wan; Mohamad Idris, Faridah; Aliff Ashraff Rosdi, Muhammad; Syahir Sarkawi, Muhammad; Liyana Jamsari, Nur; Nasir, Nur Aishah Mohd

2018-01-01

In order to design facilities for boron neutron capture therapy (BNCT), the neutron measurement must be considered to obtain the optimal design of BNCT facility such as collimator and shielding. The previous feasibility study showed that the thermal column could generate higher thermal neutrons yield for BNCT application at the TRIGA MARK II reactor. Currently, the facility for BNCT are planned to be developed at thermal column. Thus, the main objective was focused on the thermal neutron and epithermal neutron flux measurement at the thermal column. In this measurement, pure gold and cadmium were used as a filter to obtain the thermal and epithermal neutron fluxes from inside and outside of the thermal column door of the 200kW reactor power using a gold foil activation method. The results were compared with neutron fluxes using TLD 600 and TLD 700. The outcome of this work will become the benchmark for the design of BNCT collimator and the shielding

2.5 MeV CW 4-vane RFQ accelerator design for BNCT applications

NASA Astrophysics Data System (ADS)

Zhu, Xiaowen; Wang, Hu; Lu, Yuanrong; Wang, Zhi; Zhu, Kun; Zou, Yubin; Guo, Zhiyu

2018-03-01

Boron Neutron Capture Therapy (BNCT) promises a bright future in cancer therapy for its highly selective destruction of cancer cells, using the 10B +n→7Li +4 He reaction. It offers a more satisfactory therapeutic effect than traditional methods for the treatment of malignant brain tumors, head and neck cancer, melanoma, liver cancer and so on. A CW 4-vane RFQ, operating at 162.5 MHz, provides acceleration of a 20 mA proton beam to 2.5 MeV, bombarding a liquid lithium target for neutron production with a soft neutron energy spectrum. The fast neutron yield is about 1.73×1013 n/s. We preliminarily develop and optimize a beam shaping assembly design for the 7Li(p, n)7Be reaction with a 2.5 MeV proton beam. The epithermal neutron flux simulated at the beam port will reach up to 1 . 575 ×109 n/s/cm2. The beam dynamics design, simulation and benchmark for 2.5 MeV BNCT RFQ have been performed with both ParmteqM (V3.05) and Toutatis, with a transmission efficiency higher than 99.6% at 20 mA. To ease the thermal management in the CW RFQ operation, we adopt a modest inter-vane voltage design (U = 65 kV), though this does increase the accelerator length (reaching 5.2 m). Using the well-developed 3D electromagnetic codes, CST MWS and ANSYS HFSS, we are able to deal with the complexity of the BNCT RFQ, taking the contribution of each component in the RF volume into consideration. This allows us to optimize the longitudinal field distribution in a full-length model. Also, the parametric modeling technique is of great benefit to extensive modifications and simulations. In addition, the resonant frequency tuning of this RFQ is studied, giving the tuning sensitivities of vane channel and wall channel as -16.3 kHz/°C and 12.4 kHz/°C, respectively. Finally, both the multipacting level of this RFQ and multipacting suppressing in the coaxial coupler are investigated.

Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactors

NASA Astrophysics Data System (ADS)

Abdullah, Ade Gafar; Su'ud, Zaki; Kurniadi, Rizal; Kurniasih, Neny; Yulianti, Yanti

2010-12-01

Natural circulation level optimization and the effect during loss of flow accident in the 250 MWt MOX fuelled small Pb-Bi Cooled non-refueling nuclear reactors (SPINNOR) have been performed. The simulation was performed using FI-ITB safety code which has been developed in ITB. The simulation begins with steady state calculation of neutron flux, power distribution and temperature distribution across the core, hot pool and cool pool, and also steam generator. When the accident is started due to the loss of pumping power the power distribution and the temperature distribution of core, hot pool and cool pool, and steam generator change. Then the feedback reactivity calculation is conducted, followed by kinetic calculation. The process is repeated until the optimum power distribution is achieved. The results show that the SPINNOR reactor has inherent safety capability against this accident.

Optimization of a Fast Neutron Scintillator for Real-Time Pulse Shape Discrimination in the Transient Reactor Test Facility (TREAT) Hodoscope

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, James T.; Thompson, Scott J.; Watson, Scott M.

We present a multi-channel, fast neutron/gamma ray detector array system that utilizes ZnS(Ag) scintillator detectors. The system employs field programmable gate arrays (FPGAs) to do real-time all digital neutron/gamma ray discrimination with pulse height and time histograms to allow count rates in excess of 1,000,000 pulses per second per channel. The system detector number is scalable in blocks of 16 channels.

From Neutron Star Observables to the Equation of State. I. An Optimal Parametrization

NASA Astrophysics Data System (ADS)

Raithel, Carolyn A.; Özel, Feryal; Psaltis, Dimitrios

2016-11-01

The increasing number and precision of measurements of neutron star masses, radii, and, in the near future, moments of inertia offer the possibility of precisely determining the neutron star equation of state (EOS). One way to facilitate the mapping of observables to the EOS is through a parametrization of the latter. We present here a generic method for optimizing the parametrization of any physically allowed EOS. We use mock EOS that incorporate physically diverse and extreme behavior to test how well our parametrization reproduces the global properties of the stars, by minimizing the errors in the observables of mass, radius, and the moment of inertia. We find that using piecewise polytropes and sampling the EOS with five fiducial densities between ˜1-8 times the nuclear saturation density results in optimal errors for the smallest number of parameters. Specifically, it recreates the radii of the assumed EOS to within less than 0.5 km for the extreme mock EOS and to within less than 0.12 km for 95% of a sample of 42 proposed, physically motivated EOS. Such a parametrization is also able to reproduce the maximum mass to within 0.04 {M}⊙ and the moment of inertia of a 1.338 {M}⊙ neutron star to within less than 10% for 95% of the proposed sample of EOS.

Development of the dense plasma focus for short-pulse applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bennett, N.; Blasco, M.; Breeding, K.

The dense plasma focus (DPF) has long been considered a compact source for pulsed neutrons and has traditionally been optimized for the total neutron yield. Here, we describe the efforts to optimize the DPF for short-pulse applications by introducing a reentrant cathode at the end of the coaxial plasma gun. We reduced the resulting neutron pulse widths by an average of 21±921±9% from the traditional long-drift DPF design. Pulse widths and yields achieved from deuterium-tritium fusion at 2 MA are 61.8±30.761.8±30.7 ns FWHM and 1.84±0.49×10121.84±0.49×10 12 neutrons per shot. Simulations were conducted concurrently to elucidate the DPF operation and confirmmore » the role of the reentrant cathode. Furthermore, a hybrid fluid-kinetic particle-in-cell modeling capability demonstrates correct sheath velocities, plasma instabilities, and fusion yield rates. Consistent with previous findings that the DPF is dominated by beam-target fusion from superthermal ions, we estimate that the thermonuclear contribution is at the 1% level.« less

Development of the dense plasma focus for short-pulse applications

DOE PAGES

Bennett, N.; Blasco, M.; Breeding, K.; ...

2017-01-05

The dense plasma focus (DPF) has long been considered a compact source for pulsed neutrons and has traditionally been optimized for the total neutron yield. Here, we describe the efforts to optimize the DPF for short-pulse applications by introducing a reentrant cathode at the end of the coaxial plasma gun. We reduced the resulting neutron pulse widths by an average of 21±921±9% from the traditional long-drift DPF design. Pulse widths and yields achieved from deuterium-tritium fusion at 2 MA are 61.8±30.761.8±30.7 ns FWHM and 1.84±0.49×10121.84±0.49×10 12 neutrons per shot. Simulations were conducted concurrently to elucidate the DPF operation and confirmmore » the role of the reentrant cathode. Furthermore, a hybrid fluid-kinetic particle-in-cell modeling capability demonstrates correct sheath velocities, plasma instabilities, and fusion yield rates. Consistent with previous findings that the DPF is dominated by beam-target fusion from superthermal ions, we estimate that the thermonuclear contribution is at the 1% level.« less

Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zamani, M.; End of North Kargar st, Atomic Energy Organization of Iran, P.O. Box: 14155-1339, Tehran; Kasesaz, Y.

2015-07-01

In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials withmore » different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)« less

High-field neutral beam injection for improving the Q of a gas dynamic trap-based fusion neutron source

NASA Astrophysics Data System (ADS)

Zeng, Qiusun; Chen, Dehong; Wang, Minghuang

2017-12-01

In order to improve the fusion energy gain (Q) of a gas dynamic trap (GDT)-based fusion neutron source, a method in which the neutral beam is obliquely injected at a higher magnetic field position rather than at the mid-plane of the GDT is proposed. This method is beneficial for confining a higher density of fast ions at the turning point in the zone with a higher magnetic field, as well as obtaining a higher mirror ratio by reducing the mid-plane field rather than increasing the mirror field. In this situation, collision scattering loss of fast ions with higher density will occur and change the confinement time, power balance and particle balance. Using an updated calculation model with high-field neutral beam injection for a GDT-based fusion neutron source conceptual design, we got four optimal design schemes for a GDT-based fusion neutron source in which Q was improved to two- to three-fold compared with a conventional design scheme and considering the limitation for avoiding plasma instabilities, especially the fire-hose instability. The distribution of fast ions could be optimized by building a proper magnetic field configuration with enough space for neutron shielding and by multi-beam neutral particle injection at different axial points.

Characterization of a 6Li-loaded liquid organic scintillator for fast neutron spectrometry and thermal neutron detection.

PubMed

Bass, C D; Beise, E J; Breuer, H; Heimbach, C R; Langford, T J; Nico, J S

2013-07-01

The characterization of a liquid scintillator incorporating an aqueous solution of enriched lithium chloride to produce a scintillator with 0.40% (6)Li is presented, including the performance of the scintillator in terms of its optical properties and neutron response. The scintillator was incorporated into a fast neutron spectrometer, and the light output spectra from 2.5 MeV, 14.1 MeV, and (252)Cf neutrons were measured using capture-gated coincidence techniques. The spectrometer was operated without coincidence to perform thermal neutron measurements. Possible improvements in spectrometer performance are discussed. Published by Elsevier Ltd.

Direct nn-Scattering Measurement With the Pulsed Reactor YAGUAR.

PubMed

Mitchell, G E; Furman, W I; Lychagin, E V; Muzichka, A Yu; Nekhaev, G V; Strelkov, A V; Sharapov, E I; Shvetsov, V N; Chernuhin, Yu I; Levakov, B G; Litvin, V I; Lyzhin, A E; Magda, E P; Crawford, B E; Stephenson, S L; Howell, C R; Tornow, W

2005-01-01

Although crucial for resolving the issue of charge symmetry in the nuclear force, direct measurement of nn-scattering by colliding free neutrons has never been performed. At present the Russian pulsed reactor YAGUAR is the best neutron source for performing such a measurement. It has a through channel where the neutron moderator is installed. The neutrons are counted by a neutron detector located 12 m from the reactor. In preliminary experiments an instantaneous value of 1.1 × 10(18)/cm(2)s was obtained for the thermal neutron flux density. The experiment will be performed by the DIANNA Collaboration as International Science & Technology Center (ISTC) project No. 2286.

Direct nn-Scattering Measurement With the Pulsed Reactor YAGUAR

PubMed Central

Mitchell, G. E.; Furman, W. I.; Lychagin, E. V.; Muzichka, A. Yu.; Nekhaev, G. V.; Strelkov, A. V.; Sharapov, E. I.; Shvetsov, V. N.; Chernuhin, Yu. I.; Levakov, B. G.; Litvin, V. I.; Lyzhin, A. E.; Magda, E. P.; Crawford, B. E.; Stephenson, S. L.; Howell, C. R.; Tornow, W

2005-01-01

Although crucial for resolving the issue of charge symmetry in the nuclear force, direct measurement of nn-scattering by colliding free neutrons has never been performed. At present the Russian pulsed reactor YAGUAR is the best neutron source for performing such a measurement. It has a through channel where the neutron moderator is installed. The neutrons are counted by a neutron detector located 12 m from the reactor. In preliminary experiments an instantaneous value of 1.1 × 1018/cm2s was obtained for the thermal neutron flux density. The experiment will be performed by the DIANNA Collaboration as International Science & Technology Center (ISTC) project No. 2286. PMID:27308126

A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kamal, I., E-mail: imtiaz-kamal26@yahoo.com; Yunus, S. M., E-mail: yunussm11@yahoo.com; Datta, T. K., E-mail: tk-datta4@yahoo.com

2016-07-12

A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with amore » large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6 mm thickness each. The monochromator design was optimized to provide maximum flux on 3 mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30° (2θ) at each step and covers 120° in 4 steps. When the detector is placed at 1.6 m it subtends 20° at each step and covers 120° in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art technology for diffraction study in our country.« less

A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

NASA Astrophysics Data System (ADS)

Kamal, I.; Yunus, S. M.; Datta, T. K.; Zakaria, A. K. M.; Das, A. K.; Aktar, S.; Hossain, S.; Berliner, R.; Yelon, W. B.

2016-07-01

A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with a large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6mm thickness each. The monochromator design was optimized to provide maximum flux on 3mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30˚ (2θ) at each step and covers 120˚ in 4 steps. When the detector is placed at 1.6 m it subtends 20˚ at each step and covers 120˚ in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art technology for diffraction study in our country.

A Geant4 evaluation of the Hornyak button and two candidate detectors for the TREAT hodoscope

NASA Astrophysics Data System (ADS)

Fu, Wenkai; Ghosh, Priyarshini; Harrison, Mark J.; McGregor, Douglas S.; Roberts, Jeremy A.

2018-05-01

The performance of traditional Hornyak buttons and two proposed variants for fast-neutron hodoscope applications was evaluated using Geant4. The Hornyak button is a ZnS(Ag)-based device previously deployed at the Idaho National Laboratory's TRansient REActor Test Facility (better known as TREAT) for monitoring fast neutrons emitted during pulsing of fissile fuel samples. Past use of these devices relied on pulse-shape discrimination to reduce the significant levels of background Cherenkov radiation. Proposed are two simple designs that reduce the overall light guide mass (here, polymethyl methacrylate or PMMA), employ silicon photomultipliers (SiPMs), and can be operated using pulse-height discrimination alone to eliminate background noise to acceptable levels. Geant4 was first used to model a traditional Hornyak button, and for assumed, hodoscope-like conditions, an intrinsic efficiency of 0.35% for mono-directional fission neutrons was predicted. The predicted efficiency is in reasonably good agreement with experimental data from the literature and, hence, served to validate the physics models and approximations employed. Geant4 models were then developed to optimize the materials and geometries of two alternatives to the Hornyak button, one based on a homogeneous mixture of ZnS(Ag) and PMMA, and one based on alternating layers of ZnS(Ag) and PMMA oriented perpendicular to the incident neutron beam. For the same radiation environment, optimized, 5-cm long (along the beam path) devices of the homogeneous and layered designs were predicted to have efficiencies of approximately 1.3% and 3.3%, respectively. For longer devices, i.e., lengths larger than 25 cm, these efficiencies were shown to peak at approximately 2.2% and 5.9%, respectively. Moreover, both designs were shown to discriminate Cherenkov noise intrinsically by using an appropriate pulse-height discriminator level, i.e., pulse-shape discrimination is not needed for these devices.

A Geant4 evaluation of the Hornyak button and two candidate detectors for the TREAT hodoscope

DOE PAGES

Fu, Wenkai; Ghosh, Priyarshini; Harrison, Mark; ...

2018-02-05

The performance of traditional Hornyak buttons and two proposed variants for fast-neutron hodoscope applications was evaluated using Geant4. The Hornyak button is a ZnS(Ag)-based device previously deployed at the Idaho National Laboratory's TRansient REActor Test Facility (better known as TREAT) for monitoring fast neutrons emitted during pulsing of fissile fuel samples. Past use of these devices relied on pulse-shape discrimination to reduce the significant levels of background Cherenkov radiation. Proposed are two simple designs that reduce the overall light guide mass (here, polymethyl methacrylate or PMMA), employ silicon photomultipliers (SiPMs), and can be operated using pulse-height discrimination alone to eliminatemore » background noise to acceptable levels. Geant4 was first used to model a traditional Hornyak button, and for assumed, hodoscope-like conditions, an intrinsic efficiency of 0.35% for mono-directional fission neutrons was predicted. The predicted efficiency is in reasonably good agreement with experimental data from the literature and, hence, served to validate the physics models and approximations employed. Geant4 models were then developed to optimize the materials and geometries of two alternatives to the Hornyak button, one based on a homogeneous mixture of ZnS(Ag) and PMMA, and one based on alternating layers of ZnS(Ag) and PMMA oriented perpendicular to the incident neutron beam. For the same radiation environment, optimized, 5-cm long (along the beam path) devices of the homogeneous and layered designs were predicted to have efficiencies of approximately 1.3% and 3.3%, respectively. For longer devices, i.e., lengths larger than 25 cm, these efficiencies were shown to peak at approximately 2.2% and 5.9%, respectively. Furthermore, both designs were shown to discriminate Cherenkov noise intrinsically by using an appropriate pulse-height discriminator level, i.e., pulse-shape discrimination is not needed for these devices.« less

A Monte Carlo model for photoneutron generation by a medical LINAC

NASA Astrophysics Data System (ADS)

Sumini, M.; Isolan, L.; Cucchi, G.; Sghedoni, R.; Iori, M.

2017-11-01

For an optimal tuning of the radiation protection planning, a Monte Carlo model using the MCNPX code has been built, allowing an accurate estimate of the spectrometric and geometrical characteristics of photoneutrons generated by a Varian TrueBeam Stx© medical linear accelerator. We considered in our study a device working at the reference energy for clinical applications of 15 MV, stemmed from a Varian Clinac©2100 modeled starting from data collected thanks to several papers available in the literature. The model results were compared with neutron and photon dose measurements inside and outside the bunker hosting the accelerator obtaining a complete dose map. Normalized neutron fluences were tallied in different positions at the patient plane and at different depths. A sensitivity analysis with respect to the flattening filter material were performed to enlighten aspects that could influence the photoneutron production.

Design of a Neutron Temporal Diagnostic for measuring DD or DT burn histories at the NIF

NASA Astrophysics Data System (ADS)

Lahmann, B.; Frenje, J. A.; Sio, H.; Petrasso, R. D.; Bradley, D. K.; Le Pape, S.; MacKinnon, A. J.; Isumi, N.; Macphee, A.; Zayas, C.; Spears, B. K.; Hermann, H.; Hilsabeck, T. J.; Kilkenny, J. D.

2015-11-01

The DD or DT burn history in Inertial Confinement Fusion (ICF) implosions provides essential information about implosion performance and helps to constrain numerical modeling. The capability of measuring this burn history is thus important for the NIF in its pursuit of ignition. Currently, the Gamma Reaction History (GRH) diagnostic is the only system capable of measuring the burn history for DT implosions with yields greater than ~ 1e14. To complement GRH, a new NIF Neutron Temporal Diagnostic (NTD) is being designed for measuring the DD or DT burn history with yields greater than ~ 1e10. A traditional scintillator-based design and a pulse-dilation-based design are being considered. Using MCNPX simulations, both designs have been optimized, validated and contrasted for various types of implosions at the NIF. This work was supported in part by the U.S. DOE, LLNL and LLE.

On the use of the generalized SPRT method in the equivalent hard sphere approximation for nuclear data evaluation

NASA Astrophysics Data System (ADS)

Noguere, Gilles; Archier, Pascal; Bouland, Olivier; Capote, Roberto; Jean, Cyrille De Saint; Kopecky, Stefan; Schillebeeckx, Peter; Sirakov, Ivan; Tamagno, Pierre

2017-09-01

A consistent description of the neutron cross sections from thermal energy up to the MeV region is challenging. One of the first steps consists in optimizing the optical model parameters using average resonance parameters, such as the neutron strength functions. They can be derived from a statistical analysis of the resolved resonance parameters, or calculated with the generalized form of the SPRT method by using scattering matrix elements provided by optical model calculations. One of the difficulties is to establish the contributions of the direct and compound nucleus reactions. This problem was solved by using a slightly modified average R-Matrix formula with an equivalent hard sphere radius deduced from the phase shift originating from the potential. The performances of the proposed formalism are illustrated with results obtained for the 238U+n nuclear systems.

Optimizing the triple-axis spectrometer PANDA at the MLZ for small samples and complex sample environment conditions

NASA Astrophysics Data System (ADS)

Utschick, C.; Skoulatos, M.; Schneidewind, A.; Böni, P.

2016-11-01

The cold-neutron triple-axis spectrometer PANDA at the neutron source FRM II has been serving an international user community studying condensed matter physics problems. We report on a new setup, improving the signal-to-noise ratio for small samples and pressure cell setups. Analytical and numerical Monte Carlo methods are used for the optimization of elliptic and parabolic focusing guides. They are placed between the monochromator and sample positions, and the flux at the sample is compared to the one achieved by standard monochromator focusing techniques. A 25 times smaller spot size is achieved, associated with a factor of 2 increased intensity, within the same divergence limits, ± 2 ° . This optional neutron focusing guide shall establish a top-class spectrometer for studying novel exotic properties of matter in combination with more stringent sample environment conditions such as extreme pressures associated with small sample sizes.

Neutron spectrum determination in a sub-critical assembly using the multi-disc neutron activation technique

NASA Astrophysics Data System (ADS)

Koseoglou, P.; Vagena, E.; Stoulos, S.; Manolopoulou, M.

2016-09-01

Neutron spectrum of the sub-critical nuclear assembly-reactor of Aristotle University of Thessaloniki was measured at three radial distances from the reactor core. The neutron activation technique was applied irradiating 15 thick foils - disc of various elements at each position. The data of 38 (n, γ), (n, p) and (n, α) reactions were analyzed for specific activity determination. Discs instead of foils were used due to the relevant low neutron flux, so the gamma self-absorption as well as the neutron self-shielding factors has been calculated using GEANT simulations in order to determine the activity induced. The specific activities calculated for all isotopes studied were the input to the SANDII code, which was built specifically for the neutron spectrum de-convolution when the neutron activation technique is used. For the optimization of the results a technique was applied in order to minimize the influence of the initial-"guessed" spectrum shape SANDII uses. The neutron spectrum estimated presents a peak in the regions of (i) thermal neutrons ranged between 0.001 and 1 eV peaking at neutron energy ∼0.1 eV and (ii) fast neutrons ranged between 0.1 and 20 MeV peaking at neutron energy ∼1.2 MeV. The reduction of thermal neutrons is higher than the fast one as the distance from the reactor core increases since thermal neutrons capture by natural U-fuel has higher cross section than the fast neutrons.

PREFACE: International Workshop on Neutron Optics and Detectors (NOP&D 2013)

NASA Astrophysics Data System (ADS)

2014-07-01

Every two-three years scientists involved in developments of neutron optics gather together for the International Workshop on Neutron Optics (NOP). Neutron optics has always been considered very important for the development of new neutron instrumentation. The limited brilliance of existing or future neutron sources requires the more effective usage of emitted neutrons. Indeed, improvements of the neutron optical system or an optimization of the neutron-optical tracts of instruments can result in a significant enhancement of their performance. This is especially important at present when the neutron scattering community is strongly engaged in developments of new instrumentation around the spallation neutron sources - SNS, ESS, J-PARC and Second Target Station at ISIS. In 2013 the workshop was organized by the Jülich Centre for Neutron Science of the Forschungszentrum Jülich GmbH and was held at the Conference Centre in Ismaning next to Munich on July 2-7, 2013 on the eve of the ICNS-2013 in Edinburg. It carried on the series of Neutron Optics workshops held in Villigen (1999, 2007), Tokyo (2004) and Alpe d'Huez (2010). This time it is also aimed to compliment the International Conference on Neutron Scattering in Edinburgh (ICNS-2013) by providing a platform for detailed discussions on the latest developments in the field of neutron optics. The scope of the workshop was extended to the neutron detectors (in a way similar to the NOP-2004 held in Tokyo) and was labelled as the International Workshop on Neutron Optics and Detectors, NOP&D-2013. However, in contrast to the Tokyo workshop, the focus of discussions was not the detector technologies (which are the subject of many dedicated meetings), rather than the use of detectors for the purpose of the design of modern instrumentation aiming to inform detector developers about real detectors requirements for new advanced instrumental concepts. The three-full-days workshop gathered a record number of participants, more than 120, that well exceeded usual numbers (around 70) and even exceeded the limit capacity of the conference hall booked for the workshop! It even forced the organizers to stop the registration before the deadline and to establish a kind of waiting list. Such attendance of scientists representing Australia, Japan, India, Saudi Arabia, Russia, USA, Canada and 10 European countries, actually raised the level of this event from Workshop to Conference. Discussions at the workshop were devoted but not limited to new concepts in neutron instrumentation, focusing optics, neutron detection in the light of requirements imposed by neutron instruments, neutron polarization and polarization analysis and simulation packages. The current proceedings are representative of about a half of oral and poster presentations made at the workshop and provide a reader with possibility to outlook the current status and perspectives in the field of neutron optics and related detector developments. Alexander Ioffe Editor and Chairman of NOP&D-2013 The additional financial support by ESS Scandinavia, SwissNeutronics, Mirotron, S-DH Heidelberg and ASTRIUM is gratefully acknowledged.

Durability and shielding performance of borated Ceramicrete coatings in beta and gamma radiation fields

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagh, Arun S.; Sayenko, S. Yu.; Dovbnya, A. N.

2015-07-01

Ceramicrete™, a chemically bonded phosphate ceramic, was developed for nuclear waste immobilization and nuclear radiation shielding. Ceramicrete products are fabricated by an acid–base reaction between magnesium oxide and mono potassium phosphate. Fillers are used to impart desired properties to the product. Ceramicrete’s tailored compositions have resulted in several commercial structural products, including corrosion- and fire-protection coatings. Their borated version, called Borobond™, has been studied for its neutron shielding capabilities and is being used in structures built for storage of nuclear materials. This investigation assesses the durability and shielding performance of borated Ceramicrete coatings when exposed to gamma and beta radiationsmore » to predict the composition needed for optimal shielding performance in a realistic nuclear radiation field. Investigations were conducted using experimental data coupled with predictive Monte Carlo computer model. The results show that it is possible to produce products for simultaneous shielding of all three types of nuclear radiations, viz., neutrons, gamma-, and beta-rays. Additionally, because sprayable Ceramicrete coatings exhibit excellent corrosionand fire-protection characteristics on steel, this research also establishes an opportunity to produce thick coatings to enhance the shielding performance of corrosion and fire protection coatings for use in high radiation environment in nuclear industry.« less

Thermal neutron radiative capture cross-section of 186W(n, γ)187W reaction

NASA Astrophysics Data System (ADS)

Tan, V. H.; Son, P. N.

2016-06-01

The thermal neutron radiative capture cross section for 186W(n, γ)187W reaction was measured by the activation method using the filtered neutron beam at the Dalat research reactor. An optimal composition of Si and Bi, in single crystal form, has been used as neutron filters to create the high-purity filtered neutron beam with Cadmium ratio of Rcd = 420 and peak energy En = 0.025 eV. The induced activities in the irradiated samples were measured by a high resolution HPGe digital gamma-ray spectrometer. The present result of cross section has been determined relatively to the reference value of the standard reaction 197Au(n, γ)198Au. The necessary correction factors for gamma-ray true coincidence summing, and thermal neutron self-shielding effects were taken into account in this experiment by Monte Carlo simulations.

Effect of Different Structural Materials on Neutronic Performance of a Hybrid Reactor

NASA Astrophysics Data System (ADS)

Übeyli, Mustafa; Tel, Eyyüp

2003-06-01

Selection of structural material for a fusion-fission (hybrid) reactor is very important by taking into account of neutronic performance of the blanket. Refractory metals and alloys have much higher operating temperatures and neutron wall load (NWL) capabilities than low activation materials (ferritic/martensitic steels, vanadium alloys and SiC/SiC composites) and austenitic stainless steels. In this study, effect of primary candidate refractory alloys, namely, W-5Re, T111, TZM and Nb-1Zr on neutronic performance of the hybrid reactor was investigated. Neutron transport calculations were conducted with the help of SCALE 4.3 System by solving the Boltzmann transport equation with code XSDRNPM. Among the investigated structural materials, tantalum had the worst performance due to the fact that it has higher neutron absorption cross section than others. And W-5Re and TZM having similar results showed the best performance.

The Impact of Gate Width Setting and Gate Utilization Factors on Plutonium Assay in Passive Correlated Neutron Counting

DOE PAGES

Henzlova, Daniela; Menlove, Howard Olsen; Croft, Stephen; ...

2015-06-15

In the field of nuclear safeguards, passive neutron multiplicity counting (PNMC) is a method typically employed in non-destructive assay (NDA) of special nuclear material (SNM) for nonproliferation, verification and accountability purposes. PNMC is generally performed using a well-type thermal neutron counter and relies on the detection of correlated pairs or higher order multiplets of neutrons emitted by an assayed item. To assay SNM, a set of parameters for a given well-counter is required to link the measured multiplicity rates to the assayed item properties. Detection efficiency, die-away time, gate utilization factors (tightly connected to die-away time) as well as optimummore » gate width setting are among the key parameters. These parameters along with the underlying model assumptions directly affect the accuracy of the SNM assay. In this paper we examine the role of gate utilization factors and the single exponential die-away time assumption and their impact on the measurements for a range of plutonium materials. In addition, we examine the importance of item-optimized coincidence gate width setting as opposed to using a universal gate width value. Finally, the traditional PNMC based on multiplicity shift register electronics is extended to Feynman-type analysis and application of this approach to Pu mass assay is demonstrated.« less

Simultaneous parameter optimization of x-ray and neutron reflectivity data using genetic algorithms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Singh, Surendra, E-mail: surendra@barc.gov.in; Basu, Saibal

2016-05-23

X-ray and neutron reflectivity are two non destructive techniques which provide a wealth of information on thickness, structure and interracial properties in nanometer length scale. Combination of X-ray and neutron reflectivity is well suited for obtaining physical parameters of nanostructured thin films and superlattices. Neutrons provide a different contrast between the elements than X-rays and are also sensitive to the magnetization depth profile in thin films and superlattices. The real space information is extracted by fitting a model for the structure of the thin film sample in reflectometry experiments. We have applied a Genetic Algorithms technique to extract depth dependentmore » structure and magnetic in thin film and multilayer systems by simultaneously fitting X-ray and neutron reflectivity data.« less

Separation of gamma-ray and neutron events with CsI(Tl) pulse shape analysis

NASA Astrophysics Data System (ADS)

Ashida, Y.; Nagata, H.; Koshio, Y.; Nakaya, T.; Wendell, R.

2018-04-01

Fast neutrons are a large background to measurements of gamma-rays emitted from excited nuclei, such that detectors that can efficiently distinguish between the two are essential. In this paper we describe the separation of gamma-rays from neutrons with the pulse shape information of the CsI(Tl) scintillator, using a fast neutron beam and several gamma-ray sources. We find that a figure of merit optimized for this separation takes on large and stable values (nearly 4) between 5 and 10 MeV of electron equivalent deposited energy, the region of most interest to the study of nuclear de-excitation gamma-rays. Accordingly, this work demonstrates the ability of CsI(Tl) scintillators to reject neutron backgrounds to gamma-ray measurements at these energies.

Investigating the response of Micromegas detector to low-energy neutrons using Monte Carlo simulation

NASA Astrophysics Data System (ADS)

Khezripour, S.; Negarestani, A.; Rezaie, M. R.

2017-08-01

Micromegas detector has recently been used for high-energy neutron (HEN) detection, but the aim of this research is to investigate the response of the Micromegas detector to low-energy neutron (LEN). For this purpose, a Micromegas detector (with air, P10, BF3, 3He and Ar/BF3 mixture) was optimized for the detection of 60 keV neutrons using the MCNP (Monte Carlo N Particle) code. The simulation results show that the optimum thickness of the cathode is 1 mm and the optimum of microgrid location is 100 μm above the anode. The output current of this detector for Ar (3%) + BF3 (97%) mixture is greater than the other ones. This mixture is considered as the appropriate gas for the Micromegas neutron detector providing the output current for 60 keV neutrons at the level of 97.8 nA per neutron. Consecuently, this detector can be introduced as LEN detector.

Theoretical neutron damage calculations in industrial robotic manipulators used for non-destructive imaging applications

DOE PAGES

Hashem, Joseph; Schneider, Erich; Pryor, Mitch; ...

2017-01-01

Our paper describes how to use MCNP to evaluate the rate of material damage in a robot incurred by exposure to a neutron flux. The example used in this work is that of a robotic manipulator installed in a high intensity, fast, and collimated neutron radiography beam port at the University of Texas at Austin's TRIGA Mark II research reactor. Our effort includes taking robotic technologies and using them to automate non-destructive imaging tasks in nuclear facilities where the robotic manipulator acts as the motion control system for neutron imaging tasks. Simulated radiation tests are used to analyze the radiationmore » damage to the robot. Once the neutron damage is calculated using MCNP, several possible shielding materials are analyzed to determine the most effective way of minimizing the neutron damage. Furthermore, neutron damage predictions provide users the means to simulate geometrical and material changes, thus saving time, money, and energy in determining the optimal setup for a robotic system installed in a radiation environment.« less

Theoretical neutron damage calculations in industrial robotic manipulators used for non-destructive imaging applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hashem, Joseph; Schneider, Erich; Pryor, Mitch

Our paper describes how to use MCNP to evaluate the rate of material damage in a robot incurred by exposure to a neutron flux. The example used in this work is that of a robotic manipulator installed in a high intensity, fast, and collimated neutron radiography beam port at the University of Texas at Austin's TRIGA Mark II research reactor. Our effort includes taking robotic technologies and using them to automate non-destructive imaging tasks in nuclear facilities where the robotic manipulator acts as the motion control system for neutron imaging tasks. Simulated radiation tests are used to analyze the radiationmore » damage to the robot. Once the neutron damage is calculated using MCNP, several possible shielding materials are analyzed to determine the most effective way of minimizing the neutron damage. Furthermore, neutron damage predictions provide users the means to simulate geometrical and material changes, thus saving time, money, and energy in determining the optimal setup for a robotic system installed in a radiation environment.« less

Design of thermal neutron beam based on an electron linear accelerator for BNCT.

PubMed

Zolfaghari, Mona; Sedaghatizadeh, Mahmood

2016-12-01

An electron linear accelerator (Linac) can be used for boron neutron capture therapy (BNCT) by producing thermal neutron flux. In this study, we used a Varian 2300 C/D Linac and MCNPX.2.6.0 code to simulate an electron-photoneutron source for use in BNCT. In order to decelerate the produced fast neutrons from the photoneutron source, which optimize the thermal neutron flux, a beam-shaping assembly (BSA) was simulated. After simulations, a thermal neutron flux with sharp peak at the beam exit was obtained in the order of 3.09×10 8 n/cm 2 s and 6.19×10 8 n/cm 2 s for uranium and enriched uranium (10%) as electron-photoneutron sources respectively. Also, in-phantom dose analysis indicates that the simulated thermal neutron beam can be used for treatment of shallow skin melanoma in time of about 85.4 and 43.6min for uranium and enriched uranium (10%) respectively. Copyright © 2016. Published by Elsevier Ltd.

Correlated Production and Analog Transport of Fission Neutrons and Photons using Fission Models FREYA, FIFRELIN and the Monte Carlo Code TRIPOLI-4® .

NASA Astrophysics Data System (ADS)

Verbeke, Jérôme M.; Petit, Odile; Chebboubi, Abdelhazize; Litaize, Olivier

2018-01-01

Fission modeling in general-purpose Monte Carlo transport codes often relies on average nuclear data provided by international evaluation libraries. As such, only average fission multiplicities are available and correlations between fission neutrons and photons are missing. Whereas uncorrelated fission physics is usually sufficient for standard reactor core and radiation shielding calculations, correlated fission secondaries are required for specialized nuclear instrumentation and detector modeling. For coincidence counting detector optimization for instance, precise simulation of fission neutrons and photons that remain correlated in time from birth to detection is essential. New developments were recently integrated into the Monte Carlo transport code TRIPOLI-4 to model fission physics more precisely, the purpose being to access event-by-event fission events from two different fission models: FREYA and FIFRELIN. TRIPOLI-4 simulations can now be performed, either by connecting via an API to the LLNL fission library including FREYA, or by reading external fission event data files produced by FIFRELIN beforehand. These new capabilities enable us to easily compare results from Monte Carlo transport calculations using the two fission models in a nuclear instrumentation application. In the first part of this paper, broad underlying principles of the two fission models are recalled. We then present experimental measurements of neutron angular correlations for 252Cf(sf) and 240Pu(sf). The correlations were measured for several neutron kinetic energy thresholds. In the latter part of the paper, simulation results are compared to experimental data. Spontaneous fissions in 252Cf and 240Pu are modeled by FREYA or FIFRELIN. Emitted neutrons and photons are subsequently transported to an array of scintillators by TRIPOLI-4 in analog mode to preserve their correlations. Angular correlations between fission neutrons obtained independently from these TRIPOLI-4 simulations, using either FREYA or FIFRELIN, are compared to experimental results. For 240Pu(sf), the measured correlations were used to tune the model parameters.

Validation of the MCNP computational model for neutron flux distribution with the neutron activation analysis measurement

NASA Astrophysics Data System (ADS)

Tiyapun, K.; Chimtin, M.; Munsorn, S.; Somchit, S.

2015-05-01

The objective of this work is to demonstrate the method for validating the predication of the calculation methods for neutron flux distribution in the irradiation tubes of TRIGA research reactor (TRR-1/M1) using the MCNP computer code model. The reaction rate using in the experiment includes 27Al(n, α)24Na and 197Au(n, γ)198Au reactions. Aluminium (99.9 wt%) and gold (0.1 wt%) foils and the gold foils covered with cadmium were irradiated in 9 locations in the core referred to as CT, C8, C12, F3, F12, F22, F29, G5, and G33. The experimental results were compared to the calculations performed using MCNP which consisted of the detailed geometrical model of the reactor core. The results from the experimental and calculated normalized reaction rates in the reactor core are in good agreement for both reactions showing that the material and geometrical properties of the reactor core are modelled very well. The results indicated that the difference between the experimental measurements and the calculation of the reactor core using the MCNP geometrical model was below 10%. In conclusion the MCNP computational model which was used to calculate the neutron flux and reaction rate distribution in the reactor core can be used for others reactor core parameters including neutron spectra calculation, dose rate calculation, power peaking factors calculation and optimization of research reactor utilization in the future with the confidence in the accuracy and reliability of the calculation.

Beam shaping assembly design of 7Li(p,n)7Be neutron source for boron neutron capture therapy of deep-seated tumor.

PubMed

Zaidi, L; Belgaid, M; Taskaev, S; Khelifi, R

2018-05-31

The development of a medical facility for boron neutron capture therapy at Budker Institute of Nuclear Physics is under way. The neutron source is based on a tandem accelerator with vacuum insulation and lithium target. The proposed accelerator is conceived to deliver a proton beam around 10 mA at 2.3 MeV proton beam. To deliver a therapeutic beam for treatment of deep-seated tumors a typical Beam Shaping Assembly (BSA) based on the source specifications has been explored. In this article, an optimized BSA based on the 7 Li(p,n) 7 Be neutron production reaction is proposed. To evaluate the performance of the designed beam in a phantom, the parameters and the dose profiles in tissues due to the irradiation have been considered. In the simulations, we considered a proton energy of 2.3 MeV, a current of 10 mA, and boron concentrations in tumor, healthy tissues and skin of 52.5 ppm, 15 ppm and 22.5 ppm, respectively. It is found that, for a maximum punctual healthy tissue dose seated to 11 RBE-Gy, a mean dose of 56.5 RBE Gy with a minimum of 52.2 RBE Gy can be delivered to a tumor in 40 min, where the therapeutic ratio is estimated to 5.38. All of these calculations were carried out using the Monte Carlo MCNP code. Copyright © 2018 Elsevier Ltd. All rights reserved.

Feasibility of the Precise Energy Calibration for Fast Neutron Spectrometers

NASA Astrophysics Data System (ADS)

Gaganov, V. V.; Usenko, P. L.; Kryzhanovskaja, M. A.

2017-12-01

Computational studies aimed at improving the accuracy of measurements performed using neutron generators with a tritium target were performed. A measurement design yielding an extremely narrow peak in the energy spectrum of DT neutrons was found. The presence of such a peak establishes the conditions for precise energy calibration of fast-neutron spectrometers.

Inelastic neutron scattering spectrum of cyclotrimethylenetrinitramine: a comparison with solid-state electronic structure calculations.

PubMed

Ciezak, Jennifer A; Trevino, S F

2006-04-20

Solid-state geometry optimizations and corresponding normal-mode analysis of the widely used energetic material cyclotrimethylenetrinitramine (RDX) were performed using density functional theory with both the generalized gradient approximation (BLYP and BP functionals) and the local density approximation (PWC and VWN functionals). The structural results were found to be in good agreement with experimental neutron diffraction data and previously reported calculations based on the isolated-molecule approximation. The vibrational inelastic neutron scattering (INS) spectrum of polycrystalline RDX was measured and compared with simulated INS constructed from the solid-state calculations. The vibrational frequencies calculated from the solid-state methods had average deviations of 10 cm(-1) or less, whereas previously published frequencies based on an isolated-molecule approximation had deviations of 65 cm(-1) or less, illustrating the importance of including crystalline forces. On the basis of the calculations and analysis, it was possible to assign the normal modes and symmetries, which agree well with previous assignments. Four possible "doorway modes" were found in the energy range defined by the lattice modes, which were all found to contain fundamental contributions from rotation of the nitro groups.

HPGe detector shielding optimization with MCNPX for the MEDINA PGNAA cell

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nicol, T.; Perot, B.; Carasco, C.

2015-07-01

Radioactive waste repositories must guarantee the non-toxicity of the waste in the long term, not only regarding radioactivity but also regarding other environmental contamination such as toxic chemicals. Analytical methods already exist for chemical characterization (ICP-MS, ICP-AES...) but they are based on test sampling. A possible alternative, for waste packages with an appropriate gamma radiation level, is to use Prompt Gamma Neutron Activation Analysis (PGNAA), a non-destructive measurement technique sensitive to several toxic chemicals. In view of the characterization of radioactive wastes in Germany and France, collaboration between the CEA Cadarache (France) and the Forschungszentrum Juelich (Germany) was initiated amore » few years ago. FZJ holds a PGNAA graphite cell called MEDINA (Multi Element Detection based on Instrumental Neutron Activation), allowing the characterization of 225 L drums. Fast neutrons are emitted from a D-T pulsed 14 MeV neutron generator and thermalized in graphite to induced radiative captures in the waste materials. Prompt capture gamma rays are detected using a 104% relative efficiency n-type HPGe. However, HPGe crystal is sensitive to fast neutron damage and to thermal neutron activation. A thermal neutron shield made of lithium fluorine and lithium carbonate is already used around the detector. In order to further decrease the current of fast and thermal neutrons coming into the crystal without penalizing MEDINA sensitivity (by decreasing the thermal neutron flux and neutron die away time of the cell, the gamma detection efficiency, or increasing the gamma background), some configurations based on easy-to-implement modifications of MEDINA have been simulated with MCNPX with a model of the cell already validated by experiments. Results show that fast and thermal neutron incoming current in the HPGe could easily be reduced by about a factor of 2 by additional quantities of graphite and by replacing lithium carbonate by lithium fluorine with a higher {sup 6}Li concentration. In addition, these modifications slightly increase the thermal neutron flux in the cell without deteriorating the neutron die away time, and reduce the gamma background about a factor of 2 during the neutron pulse but 5 times less after it. More important changes have also been tested, such as the addition of polyethylene and lead between the neutron generator and the HPGe detector, which is more effective regarding neutron shielding but decreases the neutron die away time, partly compensated by a larger initial thermal neutron flux. Concerning gamma background, hydrogen capture gamma ray (2.23 MeV) is increased due to the presence of polyethylene but lead around the HPGe decreases the total gamma background. In conclusion, simple modifications are possible to improve detector shielding and life time before thermal annealing of the crystal, without reducing MEDINA cell performances. Some of these modifications will be tested in the coming months. (authors)« less

Characterizing a fast-response, low-afterglow liquid scintillator for neutron time-of-flight diagnostics in fast ignition experiments.

PubMed

Abe, Y; Hosoda, H; Arikawa, Y; Nagai, T; Kojima, S; Sakata, S; Inoue, H; Iwasa, Y; Iwano, K; Yamanoi, K; Fujioka, S; Nakai, M; Sarukura, N; Shiraga, H; Norimatsu, T; Azechi, H

2014-11-01

The characteristics of oxygen-enriched liquid scintillators with very low afterglow are investigated and optimized for application to a single-hit neutron spectrometer for fast ignition experiments. It is found that 1,2,4-trimethylbenzene has better characteristics as a liquid scintillator solvent than the conventional solvent, p-xylene. In addition, a benzophenon-doped BBQ liquid scintillator is shown to demonstrate very rapid time response, and therefore has potential for further use in neutron diagnostics with fast time resolution.

Preparation of iridium targets by electrodeposition for neutron capture cross section measurements

DOE PAGES

Bond, Evelyn M.; Moody, W. Allen; Arnold, Charles; ...

2016-03-01

Here, the preparation of 191Ir and 193Ir electrodeposits for neutron capture cross-section measurements at the detector for advanced neutron capture experiments located at the at Los Alamos Neutron Science Center is described. The electrodeposition of iridium in the desired thickness of 0.4–1 mg/cm 2 is challenging. Better yields and thicknesses were obtained using electrodeposition from isopropyl alcohol solutions than from ammonium sulfate solutions. 191Ir and 193Ir targets were initially prepared using the standard single-sided electrodeposition cell. Iridium electrodepositions using a double-sided electrodeposition cell were developed and were optimized, resulting in thick, uniform iridium deposits. LA UR 15-22475.

Development of a moderator system for the High Brilliance Neutron Source project

NASA Astrophysics Data System (ADS)

Dabruck, J. P.; Cronert, T.; Rücker, U.; Bessler, Y.; Klaus, M.; Lange, C.; Butzek, M.; Hansen, W.; Nabbi, R.; Brückel, T.

2016-11-01

The project for an accelerator based high brilliance neutron source HBS driven by Forschungszentrum Jülich forsees the use of the nuclear Be(p,n) or Be(d,n) reaction with accelerated particles in the lower MeV energy range. The lower neutron production compared to spallation has to be compensated by improving the neutron extraction process and optimizing the brilliance. Design and optimiziation of the moderator system are conducted with MCNP and will be validated with measurements at the AKR-2 training reactor by means of a prototype assembly where, e.g., the effect of different liquid H2 ortho/para ratios will be investigated and controlled in realtime via online heat capacity measurements.

Experimental setup for the determination of the correction factors of the neutron doseratemeters in fast neutron fields

DOE Office of Scientific and Technical Information (OSTI.GOV)

Iliescu, Elena; Bercea, Sorin; Dudu, Dorin

2013-12-16

The use of the U-120 Cyclotron of the IFIN-HH allowed to perform a testing bench with fast neutrons in order to determine the correction factors of the doseratemeters dedicated to neutron measurement. This paper deals with researchers performed in order to develop the irradiation facility testing the fast neutrons flux generated at the Cyclotron. This facility is presented, together with the results obtain in determining the correction factor for a doseratemeter dedicated to the neutron dose equivalent rate measurement.

Minimization of spurious strains by using a Si bent-perfect-crystal monochromator: neutron surface strain scanning of a shot-peened sample

NASA Astrophysics Data System (ADS)

Rebelo Kornmeier, Joana; Gibmeier, Jens; Hofmann, Michael

2011-06-01

Neutron strain measurements are critical at the surface. When scanning close to a sample surface, aberration peak shifts arise due to geometrical and divergence effects. These aberration peak shifts can be of the same order as the peak shifts related to residual strains. In this study it will be demonstrated that by optimizing the horizontal bending radius of a Si (4 0 0) monochromator, the aberration peak shifts from surface effects can be strongly reduced. A stress-free sample of fine-grained construction steel, S690QL, was used to find the optimal instrumental conditions to minimize aberration peak shifts. The optimized Si (4 0 0) monochromator and instrument settings were then applied to measure the residual stress depth gradient of a shot-peened SAE 4140 steel sample to validate the effectiveness of the approach. The residual stress depth profile is in good agreement with results obtained by x-ray diffraction measurements from an international round robin test (BRITE-EURAM-project ENSPED). The results open very promising possibilities to bridge the gap between x-ray diffraction and conventional neutron diffraction for non-destructive residual stress analysis close to surfaces.

Investigating a cyclotron HM-30 based neutron source for BNCT of deep-seated tumors by using shifting method

NASA Astrophysics Data System (ADS)

Suharyana; Riyatun; Octaviana, E. F.

2016-11-01

We have successfully proposed a simulation of a neutron beam-shaping assembly using MCNPX Code. This simulation study deals with designing a compact, optimized, and geometrically simple beam shaping assembly for a neutron source based on a proton cyclotron for BNCT purpose. Shifting method was applied in order to lower the fast neutron energy to the epithermal energy range by choosing appropriate materials. Based on a set of MCNPX simulations, it has been found that the best materials for beam shaping assembly are 3 cm Ni layered with 7 cm Pb as the reflector and 13 cm AlF3 the moderator. Our proposed beam shaping assembly configuration satisfies 2 of 5 of the IAEA criteria, namely the epithermal neutron flux 1.25 × 109 n.cm-2 s-1 and the gamma dose over the epithermal neutron flux is 0.18×10 -13 Gy.cm 2 n -1. However, the ratio of the fast neutron dose rate over neutron epithermal flux is still too high. We recommended that the shifting method must be accompanied by the filter method to reduce the fast neutron flux.

Spectral Performance of a Composite Single-Crystal Filtered Thermal Neutron Beam for BNCT Research at the University of Missouri

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Brockman; D. W. Nigg; M. F. Hawthorne

2009-07-01

Parameter studies, design calculations and initial neutronic performance measurements have been completed for a new thermal neutron beamline to be used for neutron capture therapy cell and small-animal radiobiology studies at the University of Missouri Research Reactor. The beamline features the use of single-crystal silicon and bismuth sections for neutron filtering and for reduction of incident gamma radiation. The calculated and measured thermal neutron fluxes produced at the irradiation location are 9.6x108 and 8.8x108 neutrons/cm2-s, respectively. Calculated and measured cadmium ratios (Au foils) are 217 and 132. These results indicate a well-thermalized neutron spectrum with sufficient thermal neutron flux formore » a variety of small animal BNCT studies.« less

Measurement of D-7Li Neutron Production in Neutron Generators Using the Threshold Activation Foil Technique

NASA Astrophysics Data System (ADS)

Coventry, M. D.; Krites, A. M.

Measurements to determine the absolute D-D and D-7Li neutron production rates with a neutron generator running at 100-200 kV acceleration potential were performed using the threshold activation foil technique. This technique provides a clear measure of fast neutron flux and with a suitable model, the neutron output. This approach requires little specialized equipment and is used to calibrate real-time neutron detectors and to verify neutron output. We discuss the activation foil measurement technique and describe its use in determining the relative contributions of D-D and D-7Li reactions to the total neutron yield and real-time detector response and compare to model predictions. The D-7Li reaction produces neutrons with a continuum of energies and a sharp peak around 13.5 MeV for measurement techniques outside of what D-D generators can perform. The ability to perform measurements with D-D neutrons alone, then add D-7Li neutrons for inelastic gamma production presents additional measurement modalities with the same neutron source without the use of tritium. Typically, D-T generators are employed for inelastic scattering applications but have a high regulatory burden from a radiological aspect (tritium inventory, liability concerns) and are export-controlled. D-D and D-7Li generators avoid these issues completely.

New opportunities in quasi elastic neutron scattering spectroscopy

NASA Astrophysics Data System (ADS)

Mezei, F.; Russina, M.

2001-07-01

The high energy resolution usually required in quasi elastic neutron scattering (QENS) spectroscopy is commonly achieved by the use of cold neutrons. This is one of the important research areas where the majority of current work is done on instruments on continuous reactor sources. One particular reason for this is the capability of continuous source time-of-flight spectrometers to use instrumental parameters optimally adapted for best data collection efficiency in each experiment. These parameters include the pulse repetition rate and the length of the pulses to achieve optimal balance between resolution and intensity. In addition, the disc chopper systems used provide perfect symmetrical line shapes with no tails and low background. Recent development of a set of novel techniques enhance the efficiency of cold neutron spectroscopy on existing and future spallation sources in a dramatic fashion. These techniques involve the use of extended pulse length, high intensity coupled moderators, disc chopper systems and advanced neutron optical beam delivery, and they will enable Lujan center at Los Alamos to surpass the best existing reactor instruments in time-of-flight QENS work by more than on order of magnitude in terms of beam flux on the sample. Other applications of the same techniques will allow us to combine advantages of backscattering spectroscopy on continuous and pulsed sources in order to deliver μeV resolution in a very broad energy transfer range.

Monte Carlo simulation of random, porous (foam) structures for neutron detection

NASA Astrophysics Data System (ADS)

Reichenberger, Michael A.; Fronk, Ryan G.; Shultis, J. Kenneth; Roberts, Jeremy A.; Edwards, Nathaniel S.; Stevenson, Sarah R.; Tiner, Christopher N.; McGregor, Douglas S.

2017-01-01

Porous media incorporating highly neutron-sensitive materials are of interest for use in the development of neutron detectors. Previous studies have shown experimentally the feasibility of 6LiF-saturated, multi-layered detectors; however, the random geometry of porous materials has limited the effectiveness of simulation efforts. The results of scatterless neutron transport and subsequent charged reaction product ion energy deposition are reported here using a novel Monte Carlo method and compared to results obtained by MCNP6. This new Dynamic Path Generation (DPG) Monte Carlo method was developed in order to overcome the complexities of modeling a random porous geometry in MCNP6. The DPG method is then applied to determine the optimal coating thickness for 10B4C-coated reticulated vitreous-carbon (RVC) foams. The optimal coating thickness for 4.1275 cm-thick 10B4C-coated reticulated vitreous carbon foams with porosities of 5, 10, 20, 30, 45, and 80 pores per inch (PPI) were determined for ionizing gas pressures of 1.0 and 2.8 atm. A simulated, maximum, intrinsic thermal-neutron detection efficiency of 62.8±0.25% was predicted for an 80 PPI RVC foam with a 0.2 μm thick coating of 10B4C, for a lower level discriminator setting of 75 keV and an argon pressure of 2.8 atm.

Experimental Test of Entropic Noise-Disturbance Uncertainty Relations for Spin-1/2 Measurements.

PubMed

Sulyok, Georg; Sponar, Stephan; Demirel, Bülent; Buscemi, Francesco; Hall, Michael J W; Ozawa, Masanao; Hasegawa, Yuji

2015-07-17

Information-theoretic definitions for noise and disturbance in quantum measurements were given in [Phys. Rev. Lett. 112, 050401 (2014)] and a state-independent noise-disturbance uncertainty relation was obtained. Here, we derive a tight noise-disturbance uncertainty relation for complementary qubit observables and carry out an experimental test. Successive projective measurements on the neutron's spin-1/2 system, together with a correction procedure which reduces the disturbance, are performed. Our experimental results saturate the tight noise-disturbance uncertainty relation for qubits when an optimal correction procedure is applied.

Efficiency Measurement of VANDLE Modules

NASA Astrophysics Data System (ADS)

Peters, William; Matei, C.; Cizewski, J. A.; O'Malley, P. D.; Spassova, I.; Bardayan, D.; Blackmon, J. C.; Brune, C.; Massey, T.; Grzywacz, R. K.; Madurga, M.; Sarazin, F.; Raiola, F.

2010-02-01

The Versatile Array of Neutron Detectors at Low Energy (VANDLE) is a new array of plastic scintillator bars being developed at the Holifield Radioactive Ion Beam Facility (HRIBF) at Oak Ridge National Laboratory (ORNL). The modular design enables optimization of different configurations for particular experiments, such as (d,n) and beta-delayed neutron-decay experiments, with rare ion beams. Two prototype modules were moved to the Edwards Accelerator Laboratory at Ohio University to measure their efficiency using a calibrated ^27Al(d,n) reaction as a neutron source. Results show that one bar with a cross section of 3x3 cm^2 is over 25% efficient to neutrons around 1 MeV with sensitivity down to 100 keV neutrons. Other design features such as wrapping and coupling will be presented, as well as results from resolution tests. )

Neutron beams implemented at nuclear research reactors for BNCT

NASA Astrophysics Data System (ADS)

Bavarnegin, E.; Kasesaz, Y.; Wagner, F. M.

2017-05-01

This paper presents a survey of neutron beams which were or are in use at 56 Nuclear Research Reactors (NRRs) in order to be used for BNCT, either for treatment or research purposes in aspects of various combinations of materials that were used in their Beam Shaping Assembly (BSA) design, use of fission converters and optimized beam parameters. All our knowledge about BNCT is indebted to researches that have been done in NRRs. The results of about 60 years research in BNCT and also the successes of this method in medical treatment of tumors show that, for the development of BNCT as a routine cancer therapy method, hospital-based neutron sources are needed. Achieving a physical data collection on BNCT neutron beams based on NRRs will be helpful for beam designers in developing a non-reactor based neutron beam.

Use of borated polyethylene to improve low energy response of a prompt gamma based neutron dosimeter

NASA Astrophysics Data System (ADS)

Priyada, P.; Ashwini, U.; Sarkar, P. K.

2016-05-01

The feasibility of using a combined sample of borated polyethylene and normal polyethylene to estimate neutron ambient dose equivalent from measured prompt gamma emissions is investigated theoretically to demonstrate improvements in low energy neutron dose response compared to only polyethylene. Monte Carlo simulations have been carried out using the FLUKA code to calculate the response of boron, hydrogen and carbon prompt gamma emissions to mono energetic neutrons. The weighted least square method is employed to arrive at the best linear combination of these responses that approximates the ICRP fluence to dose conversion coefficients well in the energy range of 10-8 MeV to 14 MeV. The configuration of the combined system is optimized through FLUKA simulations. The proposed method is validated theoretically with five different workplace neutron spectra with satisfactory outcome.

Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra.

PubMed

Ďuran, I; Bolshakova, I; Viererbl, L; Sentkerestiová, J; Holyaka, R; Lahodová, Z; Bém, P

2010-10-01

We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10(16) cm(-2) was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

NSDann2BS, a neutron spectrum unfolding code based on neural networks technology and two bonner spheres

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ortiz-Rodriguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.

In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called ''Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres'', (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the ''Robust design of artificial neural networks methodology'' and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored atmore » synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of {sup 252}Cf, {sup 241}AmBe and {sup 239}PuBe neutron sources measured with a Bonner spheres system.« less

NSDann2BS, a neutron spectrum unfolding code based on neural networks technology and two bonner spheres

NASA Astrophysics Data System (ADS)

Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

2013-07-01

In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called "Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres", (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the "Robust design of artificial neural networks methodology" and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored at synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of 252Cf, 241AmBe and 239PuBe neutron sources measured with a Bonner spheres system.

Study of neutron shielding collimators for curved beamlines at the European Spallation Source

NASA Astrophysics Data System (ADS)

Santoro, V.; DiJulio, D. D.; Ansell, S.; Cherkashyna, N.; Muhrer, G.; Bentley, P. M.

2018-06-01

The European Spallation Source is being constructed in Lund, Sweden and is planned to be the world’s brightest pulsed spallation neutron source for cold and thermal neutron beams (≤ 1 eV). The facility uses a 2 GeV proton beam to produce neutrons from a tungsten target. The neutrons are then moderated in a moderator assembly consisting of both liquid hydrogen and water compartments. Surrounding the moderator are 22 beamports, which view the moderator’s outside surfaces. The beamports are connected to long neutron guides that transport the moderated neutrons to the sample position via reflections. As well as the desired moderated neutrons, fast neutrons coming directly from the target can find their way down the beamlines. These can create unwanted sources of background for the instruments. To mitigate such a kind of background, several instruments will use curved guides to lose direct line-of-sight (LoS) to the moderator and the target. In addition instruments can also use shielding collimators to reduce the amount of fast neutrons further traveling down the guide due to albedo reflections or streaming. Several different materials have been proposed for this purpose. We present the results of a study of different options for collimators and identify the optimal choices that balance cost, background and activation levels.

Experiment Design and Analysis Guide - Neutronics & Physics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Misti A Lillo

2014-06-01

The purpose of this guide is to provide a consistent, standardized approach to performing neutronics/physics analysis for experiments inserted into the Advanced Test Reactor (ATR). This document provides neutronics/physics analysis guidance to support experiment design and analysis needs for experiments irradiated in the ATR. This guide addresses neutronics/physics analysis in support of experiment design, experiment safety, and experiment program objectives and goals. The intent of this guide is to provide a standardized approach for performing typical neutronics/physics analyses. Deviation from this guide is allowed provided that neutronics/physics analysis details are properly documented in an analysis report.

Reflector and Protections in a Sodium-cooled Fast Reactor: Modelling and Optimization

NASA Astrophysics Data System (ADS)

Blanchet, David; Fontaine, Bruno

2017-09-01

The ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration) is a Generation IV nuclear reactor concept under development in France [1]. In this frame, studies are underway to optimize radial reflectors and protections. Considering radial protections made in natural boron carbide, this study is conducted to assess the neutronic performances of the MgO as the reference choice for reflector material, in comparison with other possible materials including a more conventional stainless steel. The analysis is based upon a simplified 1-D and 2-D deterministic modelling of the reactor, providing simplified interfaces between core, reflector and protections. Such models allow examining detailed reaction rate distributions; they also provide physical insights into local spectral effects occurring at the Core-Reflector and at the Reflector-Protection interfaces.

Optimizing ZnS/6LiF scintillators for wavelength-shifting-fiber neutron detectors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crow, Lowell; Funk, Loren L; Hannan, Bruce W

2016-01-01

In this paper we compare the performance of grooved and flat ZnS/6LiF scintillators in a wavelength shifting-fiber (WLSF) detector. Flat ZnS/6LiF scintillators with the thickness L=0.2-0.8 mm were characterized using photon counting and pulse-height analysis and compared to a grooved scintillator of approximately 0.8 mm thick. While a grooved scintillator considerably increases the apparent thickness of the scintillator to neutrons for a given coating thickness, we find that the flat scintillators perform better than the grooved scintillators in terms of both light yield and neutron detection efficiency. The flat 0.8-mm-thick scintillator has the highest light output, and it is 52%more » higher compared with a grooved scintillator of same thickness. The lower light output of the grooved scintillator as compared to the flat scintillator is consistent with the greater scintillator-WLSF separation and the much larger average emission angle of the grooved scintillator. We also find that the average light cone width, or photon travel-length as measured using time-of-flight powder diffraction of diamond and vanadium, decreases with increasing L in the range of L=0.6-0.8 mm. This result contrasts with the traditional Swank diffusion model for micro-composite scintillators, and could be explained by a decrease in photon diffusion-coefficient or an increase in micro-particle content in the flat scintillator matrix for the thicker scintillators.« less

Intense fusion neutron sources

NASA Astrophysics Data System (ADS)

Kuteev, B. V.; Goncharov, P. R.; Sergeev, V. Yu.; Khripunov, V. I.

2010-04-01

The review describes physical principles underlying efficient production of free neutrons, up-to-date possibilities and prospects of creating fission and fusion neutron sources with intensities of 1015-1021 neutrons/s, and schemes of production and application of neutrons in fusion-fission hybrid systems. The physical processes and parameters of high-temperature plasmas are considered at which optimal conditions for producing the largest number of fusion neutrons in systems with magnetic and inertial plasma confinement are achieved. The proposed plasma methods for neutron production are compared with other methods based on fusion reactions in nonplasma media, fission reactions, spallation, and muon catalysis. At present, intense neutron fluxes are mainly used in nanotechnology, biotechnology, material science, and military and fundamental research. In the near future (10-20 years), it will be possible to apply high-power neutron sources in fusion-fission hybrid systems for producing hydrogen, electric power, and technological heat, as well as for manufacturing synthetic nuclear fuel and closing the nuclear fuel cycle. Neutron sources with intensities approaching 1020 neutrons/s may radically change the structure of power industry and considerably influence the fundamental and applied science and innovation technologies. Along with utilizing the energy produced in fusion reactions, the achievement of such high neutron intensities may stimulate wide application of subcritical fast nuclear reactors controlled by neutron sources. Superpower neutron sources will allow one to solve many problems of neutron diagnostics, monitor nano-and biological objects, and carry out radiation testing and modification of volumetric properties of materials at the industrial level. Such sources will considerably (up to 100 times) improve the accuracy of neutron physics experiments and will provide a better understanding of the structure of matter, including that of the neutron itself.

Neutron streaming studies along JET shielding penetrations

NASA Astrophysics Data System (ADS)

Stamatelatos, Ion E.; Vasilopoulou, Theodora; Batistoni, Paola; Obryk, Barbara; Popovichev, Sergey; Naish, Jonathan

2017-09-01

Neutronic benchmark experiments are carried out at JET aiming to assess the neutronic codes and data used in ITER analysis. Among other activities, experiments are performed in order to validate neutron streaming simulations along long penetrations in the JET shielding configuration. In this work, neutron streaming calculations along the JET personnel entrance maze are presented. Simulations were performed using the MCNP code for Deuterium-Deuterium and Deuterium- Tritium plasma sources. The results of the simulations were compared against experimental data obtained using thermoluminescence detectors and activation foils.

Thermal neutron radiative capture on cadmium as a counting technique at the INES beam line at ISIS: A preliminary investigation of detector cross-talk.

PubMed

Festa, G; Grazzi, F; Pietropaolo, A; Scherillo, A; Schooneveld, E M

2017-12-01

Experimental tests are presented that assess the cross-talk level among three scintillation detectors used as neutron counters exploiting the thermal neutron radiative capture on Cd. The measurements were done at the INES diffractometer operating at the ISIS spallation neutron source (Rutherford Appleton Laboratory, UK). These tests follow a preliminary set of measurements performed on the same instrument to study the effectiveness of this thermal neutron counting strategy in neutron diffraction measurements, typically performed on INES using squashed 3 He filled gas tubes. The experimental data were collected in two different geometrical configurations of the detectors and compared to results of Monte Carlo simulations, performed using the MCNP code. Copyright © 2017 Elsevier Ltd. All rights reserved.

Neutron assay in mixed radiation fields with a 6Li-loaded plastic scintillator

NASA Astrophysics Data System (ADS)

Balmer, M. J. I.; Gamage, K. A. A.; Taylor, G. C.

2015-08-01

A novel technique for assay of thermal and fast neutrons in a 6Li-loaded plastic scintillator is presented. Existing capture-gated thermal neutron detection techniques were evaluated with the 6Li-loaded plastic scintillator studied in this work. Using simulations and experimental work, shortcomings in its performance were highlighted. As a result, it was proposed that by separating the combined fast and thermal neutron events from gamma events, using established pulse shape discrimination techniques, the thermal neutron events could then be assayed. Experiments were conducted at the National Physical Laboratory, Teddington, performing neutron assays with seven different neutron fields using the proposed technique. For each field, thermal and fast neutron content was estimated and were shown to corroborate with the seven synthesised fields.

A study of gamma-ray and neutron radiation in the interaction of a 2 MeV proton beam with various materials.

PubMed

Kasatov, D; Makarov, A; Shchudlo, I; Taskaev, S

2015-12-01

Epithermal neutron source based on a tandem accelerator with vacuum insulation and lithium target has been proposed, developed and operated in Budker Institute of Nuclear Physics. The source is regarded as a prototype of a future compact device suitable for carrying out BNCT in oncology centers. In this work the measurements of gamma-ray and neutron radiation are presented for the interaction of a 2 MeV proton beam with various materials (Li, C, F, Al, V, Ti, Cu, Mo, stainless steel, and Ta). The obtained results enabled the optimization of the neutron-generating target and the high energy beam transportation path. Copyright © 2015 Elsevier Ltd. All rights reserved.

Prototype of the novel CAMEA concept—A backend for neutron spectrometers

NASA Astrophysics Data System (ADS)

Markó, Márton; Groitl, Felix; Birk, Jonas Okkels; Freeman, Paul Gregory; Lefmann, Kim; Christensen, Niels Bech; Niedermayer, Christof; Jurányi, Fanni; Lass, Jakob; Hansen, Allan; Rønnow, Henrik M.

2018-01-01

The continuous angle multiple energy analysis concept is a backend for both time-of-flight and analyzer-based neutron spectrometers optimized for neutron spectroscopy with highly efficient mapping in the horizontal scattering plane. The design employs a series of several upward scattering analyzer arcs placed behind each other, which are set to different final energies allowing a wide angular coverage with multiple energies recorded simultaneously. For validation of the concept and the model calculations, a prototype was installed at the Swiss neutron source SINQ, Paul Scherrer Institut. The design of the prototype, alignment and calibration procedures, experimental results of background measurements, and proof-of-concept inelastic measurements on LiHoF4 and h-YMnO3 are presented here.

Estimation of Orbital Neutron Detector Spatial Resolution by Systematic Shifting of Differential Topographic Masks

NASA Technical Reports Server (NTRS)

McClanahan, T. P.; Mitrofanov, I. G.; Boynton, W. V.; Chin, G.; Livengood, T.; Starr, R. D.; Evans, L. G.; Mazarico, E.; Smith, D. E.

2012-01-01

We present a method and preliminary results related to determining the spatial resolution of orbital neutron detectors using epithermal maps and differential topographic masks. Our technique is similar to coded aperture imaging methods for optimizing photonic signals in telescopes [I]. In that approach photon masks with known spatial patterns in a telescope aperature are used to systematically restrict incoming photons which minimizes interference and enhances photon signal to noise. Three orbital neutron detector systems with different stated spatial resolutions are evaluated. The differing spatial resolutions arise due different orbital altitudes and the use of neutron collimation techniques. 1) The uncollimated Lunar Prospector Neutron Spectrometer (LPNS) system has spatial resolution of 45km FWHM from approx. 30km altitude mission phase [2]. The Lunar Rennaissance Orbiter (LRO) Lunar Exploration Neutron Detector (LEND) with two detectors at 50km altitude evaluated here: 2) the collimated 10km FWHM spatial resolution detector CSETN and 3) LEND's collimated Sensor for Epithermal Neutrons (SETN). Thus providing two orbital altitudes to study factors of: uncollimated vs collimated and two average altitudes for their effect on fields-of-view.

Status and Perspectives of Neutron Imaging Facilities

NASA Astrophysics Data System (ADS)

Lehmann, E.; Trtik, P.; Ridikas, D.

The methodology and the application range of neutron imaging techniques have been significantly improved at numerous facilities worldwide in the last decades. This progress has been achieved by new detector systems, the setup of dedicated, optimized and flexible beam lines and the much better understanding of the complete imaging process thanks to complementary simulations. Furthermore, new applications and research topics were found and implemented. However, since the quality and the number of neutron imaging facilities depend much on the access to suitable beam ports, there is still an enormous potential to implement state-of-the-art neutron imaging techniques at many more facilities. On the one hand, there are prominent and powerful sources which do not intend/accept the implementation of neutron imaging techniques due to the priorities set for neutron scattering and irradiation techniques exclusively. On the other hand, there are modern and useful devices which remain under-utilized and have either not the capacity or not the know-how to develop attractive user programs and/or industrial partnerships. In this overview of the international status of neutron imaging facilities, we will specify details about the current situation.

Device for Detection of Explosives, Nuclear and Other Hazardous Materials in Luggage and Cargo Containers

NASA Astrophysics Data System (ADS)

Kuznetsov, Andrey; Evsenin, Alexey; Gorshkov, Igor; Osetrov, Oleg; Vakhtin, Dmitry

2009-12-01

Device for detection of explosives, radioactive and heavily shielded nuclear materials in luggage and cargo containers based on Nanosecond Neutron Analysis/Associated Particles Technique (NNA/APT) is under construction. Detection module consists of a small neutron generator with built-in position-sensitive detector of associated alpha-particles, and several scintillator-based gamma-ray detectors. Explosives and other hazardous chemicals are detected by analyzing secondary high-energy gamma-rays from reactions of fast neutrons with materials inside a container. The same gamma-ray detectors are used to detect unshielded radioactive and nuclear materials. An array of several neutron detectors is used to detect fast neutrons from induced fission of nuclear materials. Coincidence and timing analysis allows one to discriminate between fission neutrons and scattered probing neutrons. Mathematical modeling by MCNP5 and MCNP-PoliMi codes was used to estimate the sensitivity of the device and its optimal configuration. Comparison of the features of three gamma detector types—based on BGO, NaI and LaBr3 crystals is presented.

Design of photon converter and photoneutron target for High power electron accelerator based BNCT.

PubMed

Rahmani, Faezeh; Seifi, Samaneh; Anbaran, Hossein Tavakoli; Ghasemi, Farshad

2015-12-01

An electron accelerator, ILU-14, with current of 10 mA and 100 kW in power has been considered as one of the options for neutron source in Boron Neutron Capture Therapy (BNCT). The final design of neutron target has been obtained using MCNPX to optimize the neutron production. Tungsten in strip shape and D2O in cylindrical form have been proposed as the photon converter and the photoneutron target, respectively. In addition calculation of heat deposition in the photon target design has been considered to ensure mechanical stability of target. The results show that about 8.37×10(12) photoneutron/s with average energy of 615 keV can be produced by this neutron source design. In addition, using an appropriate beam shaping assembly an epithermal neutron flux of the order of 1.24×10(8) cm(-2) s(-1) can be obtained for BNCT applications. Copyright © 2015 Elsevier Ltd. All rights reserved.

Production of neutron-rich nuclei approaching r-process by gamma-induced fission of 238U at ELI-NP

NASA Astrophysics Data System (ADS)

Mei, Bo; Balabanski, Dimiter; Constantin, Paul; Anh Le, Tuan; Viet Cuong, Phan

2018-05-01

The investigation of neutron-rich exotic nuclei is crucial not only for nuclear physics but also for nuclear astrophysics. Experimentally, only few neutron-rich nuclei near the stability have been studied, however, most neutron-rich nuclei have not been measured due to their small production cross sections as well as short half-lives. At ELI-NP, gamma beams with high intensities will open new opportunities to investigate very neutron-rich fragments produced by photofission of 238U targets in a gas cell. Based on some simulations, a novel gas cell has been designed to produce, stop and extract 238U photofission fragments. The extraction time and efficiency of photofission fragments have been optimized by using SIMION simulations. According to these simulations, a high extraction efficiency and a short extraction time can be achieved for 238U photofission fragments in the gas cell, which will allow one to measure very neutron-rich fragments with short half-lives by using the IGISOL facility proposed at ELI-NP.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marleau, Peter; Reyna, David

In this work we investigate a method that confirms the operability of neutron detectors requiring neither radiological sources nor radiation-generating devices. This is desirable when radiological sources are not available, but confidence in the functionality of the instrument is required. The “source”, based on the production of neutrons in high-Z materials by muons, provides a tagged, low-background and consistent rate of neutrons that can be used to check the functionality of or calibrate a detector. Using a Monte Carlo guided optimization, an experimental apparatus was designed and built to evaluate the feasibility of this technique. Through a series of trialmore » measurements in a variety of locations we show that gated muon-induced neutrons appear to provide a consistent source of neutrons (35.9 ± 2.3 measured neutrons/10,000 muons in the instrument) under normal environmental variability (less than one statistical standard deviation for 10,000 muons) with a combined environmental + statistical uncertainty of ~18% for 10,000 muons. This is achieved in a single 21-22 minute measurement at sea level.« less

Parallel and Portable Monte Carlo Particle Transport

NASA Astrophysics Data System (ADS)

Lee, S. R.; Cummings, J. C.; Nolen, S. D.; Keen, N. D.

1997-08-01

We have developed a multi-group, Monte Carlo neutron transport code in C++ using object-oriented methods and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and α eigenvalues of the neutron transport equation on a rectilinear computational mesh. It is portable to and runs in parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities are discussed, along with physics and performance results for several test problems on a variety of hardware, including all three Accelerated Strategic Computing Initiative (ASCI) platforms. Current parallel performance indicates the ability to compute α-eigenvalues in seconds or minutes rather than days or weeks. Current and future work on the implementation of a general transport physics framework (TPF) is also described. This TPF employs modern C++ programming techniques to provide simplified user interfaces, generic STL-style programming, and compile-time performance optimization. Physics capabilities of the TPF will be extended to include continuous energy treatments, implicit Monte Carlo algorithms, and a variety of convergence acceleration techniques such as importance combing.

Monte Carlo simulation of explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator.

PubMed

Bergaoui, K; Reguigui, N; Gary, C K; Brown, C; Cremer, J T; Vainionpaa, J H; Piestrup, M A

2014-12-01

An explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator has been simulated using the Monte Carlo N-Particle Transport Code (MCNP5). Nuclear-based explosive detection methods can detect explosives by identifying their elemental components, especially nitrogen. Thermal neutron capture reactions have been used for detecting prompt gamma emission (10.82MeV) following radiative neutron capture by (14)N nuclei. The explosive detection system was built based on a fully high-voltage-shielded, axial D-D neutron generator with a radio frequency (RF) driven ion source and nominal yield of about 10(10) fast neutrons per second (E=2.5MeV). Polyethylene and paraffin were used as moderators with borated polyethylene and lead as neutron and gamma ray shielding, respectively. The shape and the thickness of the moderators and shields are optimized to produce the highest thermal neutron flux at the position of the explosive and the minimum total dose at the outer surfaces of the explosive detection system walls. In addition, simulation of the response functions of NaI, BGO, and LaBr3-based γ-ray detectors to different explosives is described. Copyright © 2014 Elsevier Ltd. All rights reserved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bond, Evelyn M.; Moody, W. Allen; Arnold, Charles

Here, the preparation of 191Ir and 193Ir electrodeposits for neutron capture cross-section measurements at the detector for advanced neutron capture experiments located at the at Los Alamos Neutron Science Center is described. The electrodeposition of iridium in the desired thickness of 0.4–1 mg/cm 2 is challenging. Better yields and thicknesses were obtained using electrodeposition from isopropyl alcohol solutions than from ammonium sulfate solutions. 191Ir and 193Ir targets were initially prepared using the standard single-sided electrodeposition cell. Iridium electrodepositions using a double-sided electrodeposition cell were developed and were optimized, resulting in thick, uniform iridium deposits. LA UR 15-22475.

Data acquisition of neutron crystallography on tetragonal and triclinic forms of hen-egg-white (HEW) lysozyme with an elastically bent Si monochromator

NASA Astrophysics Data System (ADS)

Tanaka, I.; Minezaki, Y.; Harada, K.; Niimura, N.

An elastically bent silicon (EBSi) as a monochromator has been optimized for neutron diffractometers of biocrystallography. It was found that several stacked thin Si plates were easier to be bent much for the near focusing point and they increased neutron reflectivity by aligning the plates. Currently, an EBSi(1 1 1) monochromator system was equipped at a diffractometer (BIX-I). It took 50 days to collect about 12 000 reflections of hen-egg-white lysozyme. The minimum d-spacing was 2.1 Å.

Fast neutron irradiation tests of flash memories used in space environment at the ISIS spallation neutron source

NASA Astrophysics Data System (ADS)

Andreani, C.; Senesi, R.; Paccagnella, A.; Bagatin, M.; Gerardin, S.; Cazzaniga, C.; Frost, C. D.; Picozza, P.; Gorini, G.; Mancini, R.; Sarno, M.

2018-02-01

This paper presents a neutron accelerated study of soft errors in advanced electronic devices used in space missions, i.e. Flash memories performed at the ChipIr and VESUVIO beam lines at the ISIS spallation neutron source. The two neutron beam lines are set up to mimic the space environment spectra and allow neutron irradiation tests on Flash memories in the neutron energy range above 10 MeV and up to 800 MeV. The ISIS neutron energy spectrum is similar to the one occurring in the atmospheric as well as in space and planetary environments, with intensity enhancements varying in the range 108- 10 9 and 106- 10 7 respectively. Such conditions are suitable for the characterization of the atmospheric, space and planetary neutron radiation environments, and are directly applicable for accelerated tests of electronic components as demonstrated here in benchmark measurements performed on flash memories.

Progress Toward Ignition on the National Ignition Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kauffman, R L

2011-10-17

The principal approach to ignition on the National Ignition Facility (NIF) is indirect drive. A schematic of an ignition target is shown in Figure 1. The laser beams are focused through laser entrance holes at each end of a high-Z cylindrical case, or hohlraum. The lasers irradiate the hohlraum walls producing x-rays that ablate and compress the fuel capsule in the center of the hohlraum. The hohlraum is made of Au, U, or other high-Z material. For ignition targets, the hohlraum is {approx}0.5 cm diameter by {approx}1 cm in length. The hohlraum absorbs the incident laser energy producing x-rays formore » symmetrically imploding the capsule. The fuel capsule is a {approx}2-mm-diameter spherical shell of CH, Be, or C filled with DT fuel. The DT fuel is in the form of a cryogenic layer on the inside of the capsule. X-rays ablate the outside of the capsule, producing a spherical implosion. The imploding shell stagnates in the center, igniting the DT fuel. NIC has overseen installation of all of the hardware for performing ignition experiments, including commissioning of approximately 50 diagnostic systems in NIF. The diagnostics measure scattered optical light, x-rays from the hohlraum over the energy range from 100 eV to 500 keV, and x-rays, neutrons, and charged particles from the implosion. An example of a diagnostic is the Magnetic Recoil Spectrometer (MRS) built by a collaboration of scientists from MIT, UR-LLE, and LLNL shown in Figure 2. MRS measures the neutron spectrum from the implosion, providing information on the neutron yield and areal density that are metrics of the quality of the implosion. Experiments on NIF extend ICF research to unexplored regimes in target physics. NIF can produce more than 50 times the laser energy and more than 20 times the power of any previous ICF facility. Ignition scale hohlraum targets are three to four times larger than targets used at smaller facilities, and the ignition drive pulses are two to five times longer. The larger targets and longer pulse lengths produce unique plasma conditions for laser-plasma instabilities that could reduce hohlraum coupling efficiency. Initial experiments have demonstrated efficient coupling of laser energy to x-rays. X-ray drive greater than 300 eV has been measured in gas-filled ignition hohlraum and shows the expected scaling with laser energy and hohlraum scale size. Experiments are now optimizing capsule implosions for ignition. Ignition conditions require assembling the fuel with sufficient density and temperature for thermonuclear burn. X-rays ablate the outside of the capsule, accelerating and spherically compressing the capsule for assembling the fuel. The implosion stagnates, heating the central core and producing a hot spot that ignites and burns the surrounding fuel. The four main characteristics of the implosion are shell velocity, central hot spot shape, fuel adiabat, and mix. Experiments studying these four characteristics of implosions are used to optimize the implosion. Integrated experiments using cryogenic fuel layer experiments demonstrate the quality of the implosion as the optimization experiments progress. The final compressed fuel conditions are diagnosed by measuring the x-ray emission from the hot core and the neutrons and charged particles produced in the fusion reactions. Metrics of the quality of the implosion are the neutron yield and the shell areal density, as well as the size and shape of the core. The yield depends on the amount of fuel in the hot core and its temperature and is a gauge of the energy coupling to the fuel. The areal density, the density of the fuel times its thickness, diagnoses the fuel assembly, which is measured using the fraction of neutrons that are down scattered passing through the dense shell. The yield and fraction of down scattered neutrons, or shell rho-r, from the cryogenic layered implosions are shown in Figure 3. The different sets of data represent results after a series of implosion optimization experiments. Both yield and areal density show significant increases as a result of the optimization. The experimental Ignition Threshold Factor (ITFX) is a measure of the progress toward ignition. ITFX is analogous to the Lawson Criterion in Magnetic Fusion. Implosions have improved by over a factor of 50 since the first cryogenic layered experiments were done in September 2010. This increase is a measure of the progress made toward the ignition goal in the past year. Optimization experiments are planned in the coming year for continued improvement in implosion performance to achieve the ignition goal. In summary, NIF has made significant progress toward ignition in the 30 months since project completion. Diagnostics and all of the supporting equipment are in place for ignition experiments. The Ignition Campaign is under way as a national collaborative effort of all the National Nuclear Security Administration (NNSA) science laboratories as well as international partners.« less

Linear models to perform treaty verification tasks for enhanced information security

DOE PAGES

MacGahan, Christopher J.; Kupinski, Matthew A.; Brubaker, Erik M.; ...

2016-11-12

Linear mathematical models were applied to binary-discrimination tasks relevant to arms control verification measurements in which a host party wishes to convince a monitoring party that an item is or is not treaty accountable. These models process data in list-mode format and can compensate for the presence of variability in the source, such as uncertain object orientation and location. The Hotelling observer applies an optimal set of weights to binned detector data, yielding a test statistic that is thresholded to make a decision. The channelized Hotelling observer applies a channelizing matrix to the vectorized data, resulting in a lower dimensionalmore » vector available to the monitor to make decisions. We demonstrate how incorporating additional terms in this channelizing-matrix optimization offers benefits for treaty verification. We present two methods to increase shared information and trust between the host and monitor. The first method penalizes individual channel performance in order to maximize the information available to the monitor while maintaining optimal performance. Second, we present a method that penalizes predefined sensitive information while maintaining the capability to discriminate between binary choices. Data used in this study was generated using Monte Carlo simulations for fission neutrons, accomplished with the GEANT4 toolkit. Custom models for plutonium inspection objects were measured in simulation by a radiation imaging system. Model performance was evaluated and presented using the area under the receiver operating characteristic curve.« less

Linear models to perform treaty verification tasks for enhanced information security

DOE Office of Scientific and Technical Information (OSTI.GOV)

MacGahan, Christopher J.; Kupinski, Matthew A.; Brubaker, Erik M.

Linear mathematical models were applied to binary-discrimination tasks relevant to arms control verification measurements in which a host party wishes to convince a monitoring party that an item is or is not treaty accountable. These models process data in list-mode format and can compensate for the presence of variability in the source, such as uncertain object orientation and location. The Hotelling observer applies an optimal set of weights to binned detector data, yielding a test statistic that is thresholded to make a decision. The channelized Hotelling observer applies a channelizing matrix to the vectorized data, resulting in a lower dimensionalmore » vector available to the monitor to make decisions. We demonstrate how incorporating additional terms in this channelizing-matrix optimization offers benefits for treaty verification. We present two methods to increase shared information and trust between the host and monitor. The first method penalizes individual channel performance in order to maximize the information available to the monitor while maintaining optimal performance. Second, we present a method that penalizes predefined sensitive information while maintaining the capability to discriminate between binary choices. Data used in this study was generated using Monte Carlo simulations for fission neutrons, accomplished with the GEANT4 toolkit. Custom models for plutonium inspection objects were measured in simulation by a radiation imaging system. Model performance was evaluated and presented using the area under the receiver operating characteristic curve.« less

Linear models to perform treaty verification tasks for enhanced information security

NASA Astrophysics Data System (ADS)

MacGahan, Christopher J.; Kupinski, Matthew A.; Brubaker, Erik M.; Hilton, Nathan R.; Marleau, Peter A.

2017-02-01

Linear mathematical models were applied to binary-discrimination tasks relevant to arms control verification measurements in which a host party wishes to convince a monitoring party that an item is or is not treaty accountable. These models process data in list-mode format and can compensate for the presence of variability in the source, such as uncertain object orientation and location. The Hotelling observer applies an optimal set of weights to binned detector data, yielding a test statistic that is thresholded to make a decision. The channelized Hotelling observer applies a channelizing matrix to the vectorized data, resulting in a lower dimensional vector available to the monitor to make decisions. We demonstrate how incorporating additional terms in this channelizing-matrix optimization offers benefits for treaty verification. We present two methods to increase shared information and trust between the host and monitor. The first method penalizes individual channel performance in order to maximize the information available to the monitor while maintaining optimal performance. Second, we present a method that penalizes predefined sensitive information while maintaining the capability to discriminate between binary choices. Data used in this study was generated using Monte Carlo simulations for fission neutrons, accomplished with the GEANT4 toolkit. Custom models for plutonium inspection objects were measured in simulation by a radiation imaging system. Model performance was evaluated and presented using the area under the receiver operating characteristic curve.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fairchild, R.G.; Kalef-Ezra, J.; Saraf, S.K.

Various calculations indicate that an optimized epithermal neutron beam can be produced by moderating fission neutrons either with a combination of Al and D{sub 2}O, or with Al{sub 2}O{sub 3}. We have designed, installed and tested an Al{sub 2}O{sub 3} moderated epithermal neutron beam at the Brookhaven Medical Research Reactor (BMRR). The epithermal neutron fluence rate of 1.8 {times} 10{sup 9} n/cm{sup 2}-sec produces a peak thermal neutron fluence rate of 1.9 to 2.8 {times} 10{sup 9} n/cm{sup 2}-sec in a tissue equivalent (TE) phantom head, depending on the configuration. Thus a single therapy treatment of 5 {times} 10{sup 12}more » n/cm{sup 2} can be delivered in 30--45 minutes. All irradiation times are given for a BMRR power of 3 MW, which is the highest power which can be delivered continuously. 18 refs., 8 figs., 4 tabs.« less

Capillary optics for radiation focusing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peurrung, A.J.; Reeder, P.L.; Bliss, M.

Capillary lens technology may ultimately bring benefits to neutron and x-ray-based science like conventional lenses with visible light. Although the technology is not yet 10 years old, these lenses have already had a significant impact in engineering, science, and medicine. Capillary lenses are advantageous when it is desirable to increase the radiation flux at a location without regard to its angular divergence. PNNL has worked to improve the technology in several ways. A single, optimally tapered capillary was manufactured, which allows intensity gains of a factor of 270 for an initially parallel, incident x-ray beam. Feasibility of constructing neutron lensesmore » using {sup 58}Ni (particularly effective at reflecting neutrons) has been explored. Three applications for capillary optics have been identified and studied: neutron telescope, Gandolphi x-ray diffractometry, and neutron radiotherapy. A brief guide is given for determining which potential applications are likely to be helped by capillary optics.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youinou, Gilles Jean-Michel

Neutron cross-sections characterize the way neutrons interact with matter. They are essential to most nuclear engineering projects and, even though theoretical progress has been made as far as the predictability of neutron cross-section models, measurements are still indispensable to meet tight design requirements for reduced uncertainties. Within the field of fission reactor technology, one can identify the following specializations that rely on the availability of accurate neutron cross-sections: (1) fission reactor design, (2) nuclear fuel cycles, (3) nuclear safety, (4) nuclear safeguards, (5) reactor monitoring and neutron fluence determination and (6) waste disposal and transmutation. In particular, the assessment ofmore » advanced fuel cycles requires an extensive knowledge of transuranics cross sections. Plutonium isotopes, but also americium, curium and up to californium isotope data are required with a small uncertainty in order to optimize significant features of the fuel cycle that have an impact on feasibility studies (e.g. neutron doses at fuel fabrication, decay heat in a repository, etc.). Different techniques are available to determine neutron cross sections experimentally, with the common denominator that a source of neutrons is necessary. It can either come from an accelerator that produces neutrons as a result of interactions between charged particles and a target, or it can come from a nuclear reactor. When the measurements are performed with an accelerator, they are referred to as differential since the analysis of the data provides the cross-sections for different discrete energies, i.e. σ(E i), and for the diffusion cross sections for different discrete angles. Another approach is to irradiate a very pure sample in a test reactor such as the Advanced Test Reactor (ATR) at INL and, after a given time, determine the amount of the different transmutation products. The precise characterization of the nuclide densities before and after neutron irradiation allows to infer energy-integrated neutron cross sections, i.e. ∫₀ ∞σ(E)φ(E)dE, where φ(E) is the neutron flux “seen” by the sample. This approach, which is usually defined and led by reactor physicists, is referred to as integral and is the object of this report. These two sources of information, i.e. differential and integral, are complementary and are used by the nuclear physicists in charge of producing the evaluated nuclear data files used by the nuclear community (ENDF, JEFF…). The generation of accurate nuclear data files requires an iterative process involving reactor physicists and nuclear data evaluators. This experimental program has been funded by the ATR National Scientific User Facility (ATR-NSUF) and by the DOE Office of Science in the framework of the Recovery Act. It has been given the name MANTRA for Measurement of Actinides Neutron TRAnsmutation.« less

A reassessment study of multi-material-shell gas puff z-pinches as a pulsed neutron source on the sandia ZR

NASA Astrophysics Data System (ADS)

Chong, Y. K.; Velikovich, A. L.; Thornhil, J. W.; Giuliani, J. L.; Knapp, P.; Jennings, C.

2013-10-01

Over the last few years, numerous 1D and 2D MHD simulation studies of deuterium (D) based double-shell gas-puff Z-pinch implosions driven by the Sandia ZR accelerator have been carried out to assess the Z-pinch as a pulsed thermal fusion neutron source. In these studies, an ad-hoc time-dependent shunt impedance model was used within the external driving circuit model in order to account for the unresolved current loss in the MITL and the load. In this study, we incorporate an improved ZR circuit model recently formulated based on the recent Sandia argon gas-puff experiment circuit data into the multi-material version of the Mach +DDTCRE RMHD code. We reinvestigate the effects of multidimensional structure and nonuniform gradients as well as the outer- and inner-shell material interaction on the implosion physics and dynamics of both D-on-D and argon-on-D Z-pinch loads using the model. Then, we characterize the neutron production performance of the Z-pinch loads as a function of total mass, mass ratio and/or radius toward their optimization as a pulsed thernonuclear neutron source. Work supported by DOE/NNSA. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. DOE's NNSA under contract DE-AC04-94AL85000.

An electro-amalgamation approach to isolate no-carrier-added 177Lu from neutron irradiated Yb for biomedical applications.

PubMed

Chakravarty, Rubel; Das, Tapas; Dash, Ashutosh; Venkatesh, Meera

2010-10-01

A novel two-step separation process for the production of no-carrier-added (NCA) (177)Lu from neutron irradiated Yb target through an electrochemical pathway employing mercury-pool cathode has been developed. A two-cycle electrolysis procedure was adopted for separation of (177)Lu from (177)Lu/Yb mixture in lithium citrate medium. The influence of different experimental parameters on the separation process was investigated and optimized for the quantitative deposition of Yb in presence of (177)Lu. The first electrolysis was performed for 50 min in the (177)Lu/Yb feed solution at pH 6 applying a potential of 8 V using platinum electrode as anode and mercury as the cathode. The second electrolysis was performed under the same conditions using fresh electrodes. The radionuclidic and chemical purity of (177)Lu was determined by using gamma ray spectrometry and atomic absorption spectrometry. The suitability of (177)Lu for biomedical applications was ascertained by labeling 1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid D-Phe(1)-Tyr(3)-octreotate(DOTA-TATE) with (177)Lu. This process could provide NCA (177)Lu with >99.99% radionuclidic purity and an overall separation yield of ∼99% was achieved within 3-4 h. The Hg content in the product was determined to be <1 ppm. Radiolabeling yield of >98% was obtained with DOTA-TATE under the optimized reaction conditions. An efficient strategy for the separation of NCA (177)Lu, suitable for biomedical applications, has been developed. Copyright © 2010 Elsevier Inc. All rights reserved.

Exploring Boron Neutron Capture Therapy for non-small cell lung cancer.

PubMed

Farías, Rubén O; Bortolussi, Silva; Menéndez, Pablo R; González, Sara J

2014-12-01

Boron Neutron Capture Therapy (BNCT) is a radiotherapy that combines biological targeting and high LET radiation. It consists in the enrichment of tumour with (10)B and in the successive irradiation of the target with low energy neutrons producing charged particles that mainly cause non-repairable damages to the cells. The feasibility to treat Non Small Cells Lung Cancer (NSCLC) with BNCT was explored. This paper proposes a new approach to determine treatment plans, introducing the possibility to choose the irradiation start and duration to maximize the tumour dose. A Tumour Control Probability (TCP) suited for lung BNCT as well as other high dose radiotherapy schemes was also introduced. Treatment plans were evaluated in localized and disseminated lung tumours. Semi-ideal and real energy spectra beams were employed to assess the best energy range and the performance of non-tailored neutron sources for lung tumour treatments. The optimal neutron energy is within [500 eV-3 keV], lower than the 10 keV suggested for the treatment of deep-seated tumours in the brain. TCPs higher than 0.6 and up to 0.95 are obtained for all cases. Conclusions drawn from [Suzuki et al., Int Canc Conf J 1 (4) (2012) 235-238] supporting the feasibility of BNCT for shallow lung tumours are confirmed, however discussions favouring the treatment of deeper lesions and disseminated disease are also opened. Since BNCT gives the possibility to deliver a safe and potentially effective treatment for NSCLC, it can be considered a suitable alternative for patients with few or no treatment options. Copyright © 2014 Associazione Italiana di Fisica Medica. Published by Elsevier Ltd. All rights reserved.

Analysis of sewage sludge using an experimental prompt gamma neutron activation analysis (pgnaa) set-up with an am-be source

NASA Astrophysics Data System (ADS)

Idiri, Z.; Redjem, F.; Beloudah, N.

2016-09-01

An experimental PGNAA set-up using a 1 Ci Am-Be source has been developed and used for analysis of bulk sewage sludge samples issued from a wastewater treatment plant situated in an industrial area of Algiers. The sample dimensions were optimized using thermal neutron flux calculations carried out with the MCNP5 Monte Carlo Code. A methodology is then proposed to perform quantitative analysis using the absolute method. For this, average thermal neutron flux inside the sludge samples is deduced using average thermal neutron flux in reference water samples and thermal flux measurements with the aid of a 3He neutron detector. The average absolute gamma detection efficiency is determined using the prompt gammas emitted by chlorine dissolved in a water sample. The gamma detection efficiency is normalized for sludge samples using gamma attenuation factors calculated with the MCNP5 code for water and sludge. Wet and dehydrated sludge samples were analyzed. Nutritive elements (Ca, N, P, K) and heavy metals elements like Cr and Mn were determined. For some elements, the PGNAA values were compared to those obtained using Atomic Absorption Spectroscopy (AAS) and Inductively Coupled Plasma (ICP) methods. Good agreement is observed between the different values. Heavy element concentrations are very high compared to normal values; this is related to the fact that the wastewater treatment plant is treating not only domestic but also industrial wastewater that is probably rejected by industries without removal of pollutant elements. The detection limits for almost all elements of interest are sufficiently low for the method to be well suited for such analysis.

Updated neutronics analyses of a water cooled ceramic breeder blanket for the CFETR

NASA Astrophysics Data System (ADS)

Xiaokang, ZHANG; Songlin, LIU; Xia, LI; Qingjun, ZHU; Jia, LI

2017-11-01

The water cooled ceramic breeder (WCCB) blanket employing pressurized water as a coolant is one of the breeding blanket candidates for the China Fusion Engineering Test Reactor (CFETR). Some updating of neutronics analyses was needed, because there were changes in the neutronics performance of the blanket as several significant modifications and improvements have been adopted for the WCCB blanket, including the optimization of radial build-up and customized structure for each blanket module. A 22.5 degree toroidal symmetrical torus sector 3D neutronics model containing the updated design of the WCCB blanket modules was developed for the neutronics analyses. The tritium breeding capability, nuclear heating power, radiation damage, and decay heat were calculated by the MCNP and FISPACT code. The results show that the packing factor and 6Li enrichment of the breeder should both be no less than 0.8 to ensure tritium self-sufficiency. The nuclear heating power of the blanket under 200 MW fusion power reaches 201.23 MW. The displacement per atom per full power year (FPY) of the plasma-facing component and first wall reach 0.90 and 2.60, respectively. The peak H production rate reaches 150.79 appm/FPY and the peak He production reaches 29.09 appm/FPY in blanket module #3. The total decay heat of the blanket modules is 2.64 MW at 1 s after shutdown and the average decay heat density can reach 11.09 kW m-3 at that time. The decay heat density of the blanket modules slowly decreases to lower than 10 W m-3 in more than ten years.

Lithium indium diselenide: A new scintillator for neutron imaging

DOE PAGES

Lukosi, Eric; Herrera, Elan; Hamm, Daniel; ...

2016-05-20

Lithium indium diselenide, 6LiInSe 2 or LISe, is a newly developed neutron detection material that shows both semiconducting and scintillating properties. The 24% atomic density of 6Li yields a thermal neutron mean free path of only 920 μm. This paper reports on the performance of LISe crystals in scintillation mode for its potential use as a converter screen for thermal/cold neutron imaging. The spatial resolution of LISe, determined using a 10% value of the Modulation Transfer Function (MTF), was found to not scale linearly with thickness. Crystals having a thickness of 450 μm or larger resulted in an average spatialmore » resolution of 67 μm, and the thinner crystals exhibited an increase in spatial resolution down to the Nyquist frequency of the CCD. The highest measured spatial resolution of 198 μm thick LISe (27 μm) outperforms a commercial 50 μm thick ZnS(Cu): 6LiF scintillation screen (100 μm) by more than a factor of three. For the thicknesses considered in this study, it has been found that the light yield of LISe did not scale with its thickness, suggesting the need for optimizing the synthesis to enhance the scintillation mechanism. Absorption measurements indicate that the 6Li concentration is uniform throughout the samples and its absorption efficiency as a function of thickness follows general nuclear theory, indicating that the variation in apparent brightness is likely due to a combination of particle escape, light transport, and activation of the scintillation mechanisms. As a result, the presence of 115In and its long-lived 116In activation product did not result in ghosting (memory of past neutron exposure), demonstrating potential for using LISe for imaging transient systems.« less

Fast collimated neutron flux measurement using stilbene scintillator and flashy analog-to-digital converter in JT-60U

NASA Astrophysics Data System (ADS)

Ishikawa, M.; Itoga, T.; Okuji, T.; Nakhostin, M.; Shinohara, K.; Hayashi, T.; Sukegawa, A.; Baba, M.; Nishitani, T.

2006-10-01

A line-integrated neutron emission profile is routinely measured using the radial neutron collimator system in JT-60U tokamak. Stilbene neuron detectors (SNDs), which combine a stilbene organic crystal scintillation detector (SD) with an analog neutron-gamma pulse shape discrimination (PSD) circuit, have been used to measure collimated neutron flux. Although the SND has many advantages as a neutron detector, the maximum count rate is limited up to ˜1×105counts/s due to the analog PSD circuit. To overcome this issue, a digital signal processing system (DSPS) using a flash analog-to-digital converter (Acqiris DC252, 8GHz, 10bits) has been developed at Cyclotron and Radioisotope Center in Tohoku University. In this system anode signals from photomultiplier of the SD are directory stored and digitized. Then, the PSD between neutrons and gamma rays is performed using software. The DSPS has been installed in the vertical neutron collimator system in JT-60U and applied to deuterium experiments. It is confirmed that the PSD is sufficiently performed and collimated neutron flux is successfully measured with count rate up to ˜5×105counts/s without the effect of pileup of detected pulses. The performance of the DSPS as a neutron detector, which supersedes the SND, is demonstrated.

Assessment of beryllium and molybdenum nuclear data files with the RPI neutron scattering system in the energy region from 0.5 to 20 MeV

NASA Astrophysics Data System (ADS)

Daskalakis, Adam; Blain, Ezekiel; Leinweber, Gregory; Rapp, Michael; Barry, Devin; Block, Robert; Danon, Yaron

2017-09-01

A series of neutron scattering benchmark measurements were performed on beryllium and molybdenum with the Rensselaer Polytechnic Institute's Neutron Scattering System. The pulsed neutron source was produced by the Rensselaer Polytechnic Institute's Linear Accelerator and a well collimated neutron beam was incident onto the samples located at a distance of 30.07 m. Neutrons that scattered from the sample were measured using the time-of-flight by eight EJ-301 liquid scintillator detectors positioned 0.5 m from the sample of interest. A total of eight experiments were performed with two sample thicknesses each, measured by detectors placed at two sets of angles. All data were processed using pulse shape analysis that separated the neutron and gamma ray events and included a gamma misclassification correction to account for erroneously identified gamma rays. A detailed model of the neutron scattering system simulated each experiment with several current evaluated nuclear data libraries and their predecessors. Results for each evaluation were compared to the experimental data using a figure-of-merit. The neutron scattering system has been used as a means to quantify a library's performance.

Neutron reflecting supermirror structure

DOEpatents

Wood, J.L.

1992-12-01

An improved neutron reflecting supermirror structure comprising a plurality of stacked sets of bilayers of neutron reflecting materials. The improved neutron reflecting supermirror structure is adapted to provide extremely good performance at high incidence angles, i.e. up to four time the critical angle of standard neutron mirror structures. The reflection of neutrons striking the supermirror structure at a high critical angle provides enhanced neutron throughput, and hence more efficient and economical use of neutron sources. 2 figs.

Neutron reflecting supermirror structure

DOEpatents

Wood, James L.

1992-01-01

An improved neutron reflecting supermirror structure comprising a plurality of stacked sets of bilayers of neutron reflecting materials. The improved neutron reflecting supermirror structure is adapted to provide extremely good performance at high incidence angles, i.e. up to four time the critical angle of standard neutron mirror structures. The reflection of neutrons striking the supermirror structure at a high critical angle provides enhanced neutron throughput, and hence more efficient and economical use of neutron sources.

In-Situ Characterization of Underwater Radioactive Sludge

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simpson, A.P.; Clapham, M.J.; Swinson, B.

2008-07-01

A fundamental requirement underpinning safe clean-up technologies for legacy spent nuclear fuel (SNF) ponds, pools and wet silos is the ability to characterize the radioactive waste form prior to retrieval. The corrosion products resulting from the long term underwater storage of spent nuclear fuel, reactor components and reprocessing debris present a major hazard to facility decontamination and decommissioning in terms of their radioactive content and physical / chemical reactivity. The ability to perform in-situ underwater non-destructive characterization of sludge and debris in a safe and cost-effective manner offers significant benefits over traditional destructive sampling methods. Several techniques are available formore » underwater measurements including (i) Gross gamma counting, (ii) Low-, Medium- and High- Resolution Gamma Spectroscopy, (iii) Passive neutron counting and (iv) Active Neutron Interrogation. The optimum technique depends on (i) the radioactive inventory (ii) mechanical access restrictions for deployment of the detection equipment, interrogation sources etc. (iii) the integrity of plant records and (iv) the extent to which Acceptable Knowledge which may be used for 'fingerprinting' the radioactive contents to a marker nuclide. Prior deployments of underwater SNF characterization equipment around the world have been reviewed with respect to recent developments in gamma and neutron detection technologies, digital electronics advancements, data transfer techniques, remote operation capabilities and improved field ruggedization. Modeling and experimental work has been performed to determine the capabilities, performance envelope and operational limitations of the future generation of non-destructive underwater sludge characterization techniques. Recommendations are given on the optimal design of systems and procedures to provide an acceptable level of confidence in the characterization of residual sludge content of legacy wet storage facilities such that retrieval and repackaging of SNF sludges may proceed safely and efficiently with support of the regulators and the public. (author)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

McMillan, Kyle; Marleau, Peter; Brubaker, Erik

In coded aperture imaging, one of the most important factors determining the quality of reconstructed images is the choice of mask/aperture pattern. In many applications, uniformly redundant arrays (URAs) are widely accepted as the optimal mask pattern. Under ideal conditions, thin and highly opaque masks, URA patterns are mathematically constructed to provide artifact-free reconstruction however, the number of URAs for a chosen number of mask elements is limited and when highly penetrating particles such as fast neutrons and high-energy gamma-rays are being imaged, the optimum is seldom achieved. In this case more robust mask patterns that provide better reconstructed imagemore » quality may exist. Through the use of heuristic optimization methods and maximum likelihood expectation maximization (MLEM) image reconstruction, we show that for both point and extended neutron sources a random mask pattern can be optimized to provide better image quality than that of a URA.« less

Spectroscopy of neutron rich nuclei using cold neutron induced fission of actinide targets at the ILL: the EXILL campaign

NASA Astrophysics Data System (ADS)

de France, G.; Blanc, A.; Drouet, F.; Jentschel, M.; Köster, U.; Mutti, P.; Régis, J. M.; Simpson, G.; Soldner, T.; Stezowski, O.; Ur, C. A.; Urban, W.; Vancrayenest, A.

2014-03-01

A combination of germanium detectors has been installed at the PF1B neutron guide of the ILL to perform the prompt spectroscopy of neutron-rich nuclei produced in the neutron-capture induced-fission of 235U and 241Pu. In addition LaBr3 detectors from the FATIMA collaboration have been installed in complement with the EXOGAM clovers to measure lifetimes of low-lying excited states. The measured characteristics and online spectra indicate very good performances of the overall setup.

Determination of neutron flux distribution in an Am-Be irradiator using the MCNP.

PubMed

Shtejer-Diaz, K; Zamboni, C B; Zahn, G S; Zevallos-Chávez, J Y

2003-10-01

A neutron irradiator has been assembled at IPEN facilities to perform qualitative-quantitative analysis of many materials using thermal and fast neutrons outside the nuclear reactor premises. To establish the prototype specifications, the neutron flux distribution and the absorbed dose rates were calculated using the MCNP computer code. These theoretical predictions then allow one to discuss the optimum irradiator design and its performance.

The Roles and Developments needed for Diagnostics in the ITER Fusion Device

DOE Office of Scientific and Technical Information (OSTI.GOV)

Walsh, Michael

2015-07-01

Harnessing the power from Fusion on earth is an important and challenging task. Excellent work has been carried out in this area over the years with several demonstrations of the ability to produce power. Now, a new large device is being constructed in the south of France. This is called ITER. ITER is a large-scale scientific experiment that aims to demonstrate a possibility to produce commercial energy from fusion. This project is now well underway with the many teams working on the construction and completing various aspects of the design. This device will carry up to 15 MA of plasmamore » current and produce about 500 MW of power, 400 MW approximately in high energy neutrons. The typical temperatures of the electrons inside this device are in the region of a few hundred million Kelvin. It is maintained using a magnetic field. This device is pushing several boundaries from those currently existing. As a result of this, several technologies need to be developed or extended. This is especially true for the systems or diagnostics that measure the performance and provide the control signals for this device. A diagnostic set will be installed on the ITER machine to provide the measurements necessary to control, evaluate and optimize plasma performance in ITER and to further the understanding of plasma physics. These include amongst others, measurements of the plasma shape, temperature, density, impurity concentration, and particle and energy confinement times. The system will comprise about 45 individual measuring systems drawn from the full range of modern plasma diagnostic techniques, including magnetics, lasers, X-rays, neutron cameras, impurity monitors, particle spectrometers, radiation bolometers, pressure and gas analysis, and optical fibres. These devices will have to be made to work in the new and challenging environment inside the vacuum vessel. These systems will have to cope with a range of phenomena that extend the current knowledge in the Fusion field. One amongst them is the parasitic effect of the neutrons on the while all the performing with great accuracy and precision. The levels of neutral particle flux, neutron flux and neutron fluence will be respectively about 5, 10 and 10,000 times higher than the harshest experienced in today's machines. The pulse length of the fusion reaction-or the amount of time the reaction is sustained-will be about 100 times longer. (authors)« less

Increasing the High Voltage Capabilities and Exploring Parameter Space of an Inertial Electrostatic Confinement Fusion Neutron Source for the Detection of Chemical Explosives

NASA Astrophysics Data System (ADS)

Michalak, Matthew K.

The objectives of the work presented here include understanding key operating principles and providing precise data sets that can be used to test inertial electrostatic confinement (IEC) fusion theory and optimize IEC device operation. The underlying physical behavior was separated from superficial trends observed in an IEC device at the University of Wisconsin-Madison (UW). The effects of changing voltage (30-170 kV) and current (30-100 mA) were thoroughly explored, pressure effects (0.15-1.25 mTorr) were mapped, and the effect of impurities in the system was quantified. The most challenging part of this work was designing a high voltage feedthrough that could reliably operate at higher voltages for far longer times than previously attained. A system to detect conventional explosives using fusion neutrons was also designed, constructed, and tested. Precise data sets were created by taking into account and minimizing the effects of short and long term trends in the experiment. Detailed meter current scans were taken that showed a linear relationship of the neutron production rate with current. Cathode voltage scans were slightly greater than linear in the neutron rate from 30 to 170 kV, but the rate increase diminished to near linear as 170 kV was approached. A new high voltage feedthrough was designed that surpassed the performance of past UW IEC lab feedthroughs and shows promise for long duration operation at still higher voltages. Limitations of other equipment in the IEC lab prevented testing the feedthrough to voltages above 175 kV. A more robust construction of the feedthrough and reducing the consequences of a feedthrough failure were also important design criteria that were met. A detector array was made to detect explosives via the 10.8 MeV neutron capture prompt gamma from nitrogen. Signals from four separate detectors were combined to make the individual detectors act similar to one large detector. The detector signals were both summed and combined to compare the performance of the two methods. An overwhelming background radiation signal and insufficient time resolution were two factors that led to the combined signal not performing as well as the summed signal.

A portable neutron spectroscope (NSPECT) for detection, imaging and identification of nuclear material

NASA Astrophysics Data System (ADS)

Ryan, James M.; Bancroft, Christopher; Bloser, Peter; Bravar, Ulisse; Fourguette, Dominique; Frost, Colin; Larocque, Liane; McConnell, Mark L.; Legere, Jason; Pavlich, Jane; Ritter, Greg; Wassick, Greg; Wood, Joshua; Woolf, Richard

2010-08-01

We have developed, fabricated and tested a prototype imaging neutron spectrometer designed for real-time neutron source location and identification. Real-time detection and identification is important for locating materials. These materials, specifically uranium and transuranics, emit neutrons via spontaneous or induced fission. Unlike other forms of radiation (e.g. gamma rays), penetrating neutron emission is very uncommon. The instrument detects these neutrons, constructs images of the emission pattern, and reports the neutron spectrum. The device will be useful for security and proliferation deterrence, as well as for nuclear waste characterization and monitoring. The instrument is optimized for imaging and spectroscopy in the 1-20 MeV range. The detection principle is based upon multiple elastic neutron-proton scatters in organic scintillator. Two detector panel layers are utilized. By measuring the recoil proton and scattered neutron locations and energies, the direction and energy spectrum of the incident neutrons can be determined and discrete and extended sources identified. Event reconstruction yields an image of the source and its location. The hardware is low power, low mass, and rugged. Its modular design allows the user to combine multiple units for increased sensitivity. We will report the results of laboratory testing of the instrument, including exposure to a calibrated Cf-252 source. Instrument parameters include energy and angular resolution, gamma rejection, minimum source identification distances and times, and projected effective area for a fully populated instrument.

Varying Radii of On-Axis Anode Hollows For kJ-Class Dense Plasma Focus

NASA Astrophysics Data System (ADS)

Shaw, Brian; Chapman, Steven; Falabella, Steven; Pankin, Alexei; Liu, Jason; Link, Anthony; Schmidt, Andréa

2017-10-01

A dense plasma focus (DPF) is a compact plasma gun that produces high energy ion beams, up to several MeV, through strong potential gradients. Motivated by particle-in-cell simulations, we have tried a series of hollow anodes on our kJ-class DPF. Each anode has varying hollow sizes, and has been studied to optimize ion beam production in Helium, reduce anode sputter, and increase neutron yields in deuterium. We diagnose the rate at which electrode material is ablated and deposited onto nearby surfaces. This is of interest in the case of solid targets, which perform poorly in the presence of sputter. We have found that the larger the hollow radius produces more energetic ion beams, higher neutron yield, and sputter less than a flat top anode. A complete comparison is presented. This work was prepared by LLNL under Contract DE-AC52-07NA27344 and supported by Office of Defense Nuclear Nonproliferation Research and Development within U.S. Department of Energy's National Nuclear Security Administration.

Modeling the Nab Experiment Electronics in SPICE

NASA Astrophysics Data System (ADS)

Blose, Alexander; Crawford, Christopher; Sprow, Aaron; Nab Collaboration

2017-09-01

The goal of the Nab experiment is to measure the neutron decay coefficients a, the electron-neutrino correlation, as well as b, the Fierz interference term to precisely test the Standard Model, as well as probe for Beyond the Standard Model physics. In this experiment, protons from the beta decay of the neutron are guided through a magnetic field into a Silicon detector. Event reconstruction will be achieved via time-of-flight measurement for the proton and direct measurement of the coincident electron energy in highly segmented silicon detectors, so the amplification circuitry needs to preserve fast timing, provide good amplitude resolution, and be packaged in a high-density format. We have designed a SPICE simulation to model the full electronics chain for the Nab experiment in order to understand the contributions of each stage and optimize them for performance. Additionally, analytic solutions to each of the components have been determined where available. We will present a comparison of the output from the SPICE model, analytic solution, and empirically determined data.

Hohlraum-Driven Ignition-Like Double-Shell Implosion Experiments on Omega: Analysis and Interpretation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amendt, P; Robey, H F; Park, H-S

2003-08-22

An experimental campaign to study hohlraum-driven ignition-like double-shell target performance using the Omega laser facility has begun. These targets are intended to incorporate as many ignition-like properties of the proposed National Ignition Facility (NIF) double-shell ignition design [1,2] as possible, given the energy constraints of the Omega laser. In particular, this latest generation of Omega double-shells is nominally predicted to produce over 99% of the (clean) DD neutron yield from the compressional or stagnation phase of the implosion as required in the NIF ignition design. By contrast, previous double-shell experience on Omega [3] was restricted to cases where a significantmore » fraction of the observed neutron yield was produced during the earlier shock convergence phase where the effects of mix are deemed negligibly small. These new targets are specifically designed to have optimized fall-line behavior for mitigating the effects of pusher-fuel mix after deceleration onset and, thereby, providing maximum neutron yield from the stagnation phase. Experimental results from this recent Omega ignition-like double-shell implosion campaign show favorable agreement with two-dimensional integrated hohlraum simulation studies when enhanced (gold) hohlraum M-band (2-5 keV) radiation is included at a level consistent with observations.« less

Transmutation of uranium and thorium in the particle field of the Quinta sub-critical assembly

NASA Astrophysics Data System (ADS)

Hashemi-Nezhad, S. R.; Asquith, N. L.; Voronko, V. A.; Sotnikov, V. V.; Zhadan, Alina; Zhuk, I. V.; Potapenko, A.; Husak, Krystsina; Chilap, V.; Adam, J.; Baldin, A.; Berlev, A.; Furman, W.; Kadykov, M.; Khushvaktov, J.; Kudashkin, I.; Mar'in, I.; Paraipan, M.; Pronskih, V.; Solnyshkin, A.; Tyutyunnikov, S.

2018-03-01

The fission rates of natural uranium and thorium were measured in the particle field of Quinta, a 512 kg natural uranium target-blanket sub-critical assembly. The Quinta assembly was irradiated with deuterons of energy 4 GeV from the Nuclotron accelerator of the Joint Institute for Nuclear Research (JINR), Dubna, Russia. Fission rates of uranium and thorium were measured using Gamma spectroscopy and fission track techniques. The production rate of 239Np was also measured. The obtained experimental results were compared with Monte Carlo predictions using the MCNPX 2.7 code employing the physics and fission-evaporation models of INCL4-ABLA, CEM03.03 and LAQGSM03.03. Some of the neutronic characteristics of the Quinta are compared with the "Energy plus Transmutation (EpT)" subcritical assembly, which is composed of a lead target and natU blanket. This comparison clearly demonstrates the importance of target material, neutron moderator and reflector types on the performance of a spallation neutron driven subcritical system. As the dimensions of the Quinta are very close to those of an optimal multi-rod-uranium target, the experimental and Monte Carlo calculation results presented in this paper provide insights on the particle field within a uranium target as well as in Accelerator Driven Systems in general.

A Permanent-Magnet Microwave Ion Source For A Compact High-Yield Neutron Generator

NASA Astrophysics Data System (ADS)

Waldmann, O.; Ludewigt, B.

2011-06-01

We present recent work on the development of a microwave ion source that will be used in a high-yield compact neutron generator for active interrogation applications. The sealed tube generator will be capable of producing high neutron yields, 5×1011 n/s for D-T and ˜1×1010 n/s for D-D reactions, while remaining transportable. We constructed a microwave ion source (2.45 GHz) with permanent magnets to provide the magnetic field strength of 87.5 mT necessary for satisfying the electron cyclotron resonance (ECR) condition. Microwave ion sources can produce high extracted beam currents at the low gas pressures required for sealed tube operation and at lower power levels than previously used RF-driven ion sources. A 100 mA deuterium/tritium beam will be extracted through a large slit (60×6 mm2) to spread the beam power over a larger target area. This paper describes the design of the permanent-magnet microwave ion source and discusses the impact of the magnetic field design on the source performance. The required equivalent proton beam current density of 40 mA/cm2 was extracted at a moderate microwave power of 400 W with an optimized magnetic field.

Optimization of a Boiling Water Reactor Loading Pattern Using an Improved Genetic Algorithm

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kobayashi, Yoko; Aiyoshi, Eitaro

2003-08-15

A search method based on genetic algorithms (GA) using deterministic operators has been developed to generate optimized boiling water reactor (BWR) loading patterns (LPs). The search method uses an Improved GA operator, that is, crossover, mutation, and selection. The handling of the encoding technique and constraint conditions is designed so that the GA reflects the peculiar characteristics of the BWR. In addition, some strategies such as elitism and self-reproduction are effectively used to improve the search speed. LP evaluations were performed with a three-dimensional diffusion code that coupled neutronic and thermal-hydraulic models. Strong axial heterogeneities and three-dimensional-dependent constraints have alwaysmore » necessitated the use of three-dimensional core simulators for BWRs, so that an optimization method is required for computational efficiency. The proposed algorithm is demonstrated by successfully generating LPs for an actual BWR plant applying the Haling technique. In test calculations, candidates that shuffled fresh and burned fuel assemblies within a reasonable computation time were obtained.« less

A Feasibility Study on Reactor Based Fission Neutron Radiography of 200-l Waste Packages

NASA Astrophysics Data System (ADS)

Bücherl, T.; Kalthoff, O.; von Gostomski, Ch. Lierse

This feasibility study investigates the applicability of fission neutrons for the non-destructive characterization of radioactive waste packages by means of neutron radiography. Based on a number of mock-up drums of different non-radioactive matrices, but being typical for radioactive waste generated in Europe, radiography measurements at the NECTAR and the ITS facility using fission neutrons and 60Co-gamma-rays, respectively, are performed. The resulting radiographs are compared and qualitatively assessed. In addition, a first approach for the stitching of the fission neutron radiographs to visualize the complete area of 200-l waste drums is performed. While the feasibility of fission neutrons is demonstrated successfully, fields for further improvements are identified.

Conception of a New Recoil Proton Telescope for Real-Time Neutron Spectrometry in Proton-Therapy

NASA Astrophysics Data System (ADS)

Combe, Rodolphe; Arbor, Nicolas; el Bitar, Ziad; Higueret, Stéphane; Husson, Daniel

2018-01-01

Neutrons are the main type of secondary particles emitted in proton-therapy. Because of the risk of secondary cancer and other late occurring effects, the neutron dose should be included in the out-of-field dose calculations. A neutron spectrometer has to be used to take into account the energy dependence of the neutron radiological weighting factor. Due to its high dependence on various parameters of the irradiation (beam, accelerator, patient), the neutron spectrum should be measured independently for each treatment. The current reference method for the measurement of the neutron energy, the Bonner Sphere System, consists of several homogeneous polyethylene spheres with increasing diameters equipped with a proportional counter. It provides a highresolution reconstruction of the neutron spectrum but requires a time-consuming work of signal deconvolution. New neutron spectrometers are being developed, but the main experimental limitation remains the high neutron flux in proton therapy treatment rooms. A new model of a real-time neutron spectrometer, based on a Recoil Proton Telescope technology, has been developed at the IPHC. It enables a real-time high-rate reconstruction of the neutron spectrum from the measurement of the recoil proton trajectory and energy. A new fast-readout microelectronic integrated sensor, called FastPixN, has been developed for this specific purpose. A first prototype, able to detect neutrons between 5 and 20 MeV, has already been validated for metrology with the AMANDE facility at Cadarache. The geometry of the new Recoil Proton Telescope has been optimized via extensive Geant4 Monte Carlo simulations. Uncertainty sources have been carefully studied in order to improve simultaneously efficiency and energy resolution, and solutions have been found to suppress the various expected backgrounds. We are currently upgrading the prototype for secondary neutron detection in proton therapy applications.

Measurement and simulation for a complementary imaging with the neutron and X-ray beams

NASA Astrophysics Data System (ADS)

Hara, Kaoru Y.; Sato, Hirotaka; Kamiyama, Takashi; Shinohara, Takenao

2017-09-01

By using a composite source system, we measured radiographs of the thermal neutron and keV X-ray in the 45-MeV electron linear accelerator facility at Hokkaido University. The source system provides the alternative beam of neutron and X-ray by switching the production target onto the electron beam axis. In the measurement to demonstrate a complementary imaging, the detector based on a vacuum-tube type neutron color image intensifier was applied to the both beams for dual-purpose. On the other hand, for reducing background in a neutron transmission spectrum, test measurements using a gadolinium-type neutron grid were performed with a cold neutron source at Hokkaido University. In addition, the simulations of the neutron and X-ray transmissions for various substances were performed using the PHITS code. A data analysis procedure for estimating the substance of sample was investigated through the simulations.

Imaging of Rabbit VX-2 Hepatic Cancer by Cold and Thermal Neutron Radiography

NASA Astrophysics Data System (ADS)

Tsuchiya, Yoshinori; Matsubayashi, Masahito; Takeda, Tohoru; Lwin, Thet Thet; Wu, Jin; Yoneyama, Akio; Matsumura, Akira; Hori, Tomiei; Itai, Yuji

2003-11-01

Neutron radiography is based on differences in neutron mass attenuation coefficients among the elements and is a non-destructive imaging method. To investigate biomedical applications of neutron radiography, imaging of rabbit VX-2 liver cancer was performed using thermal and cold neutron radiography with a neutron imaging plate. Hepatic vessels and VX-2 tumor were clearly observed by neutron radiography, especially by cold neutron imaging. The image contrast of this modality was better than that of absorption-contrast X-ray radiography.

Neutron Deep Penetration Calculations in Light Water with Monte Carlo TRIPOLI-4® Variance Reduction Techniques

NASA Astrophysics Data System (ADS)

Lee, Yi-Kang

2017-09-01

Nuclear decommissioning takes place in several stages due to the radioactivity in the reactor structure materials. A good estimation of the neutron activation products distributed in the reactor structure materials impacts obviously on the decommissioning planning and the low-level radioactive waste management. Continuous energy Monte-Carlo radiation transport code TRIPOLI-4 has been applied on radiation protection and shielding analyses. To enhance the TRIPOLI-4 application in nuclear decommissioning activities, both experimental and computational benchmarks are being performed. To calculate the neutron activation of the shielding and structure materials of nuclear facilities, the knowledge of 3D neutron flux map and energy spectra must be first investigated. To perform this type of neutron deep penetration calculations with the Monte Carlo transport code, variance reduction techniques are necessary in order to reduce the uncertainty of the neutron activation estimation. In this study, variance reduction options of the TRIPOLI-4 code were used on the NAIADE 1 light water shielding benchmark. This benchmark document is available from the OECD/NEA SINBAD shielding benchmark database. From this benchmark database, a simplified NAIADE 1 water shielding model was first proposed in this work in order to make the code validation easier. Determination of the fission neutron transport was performed in light water for penetration up to 50 cm for fast neutrons and up to about 180 cm for thermal neutrons. Measurement and calculation results were benchmarked. Variance reduction options and their performance were discussed and compared.

Final design of the Energy-Resolved Neutron Imaging System “RADEN” at J-PARC

NASA Astrophysics Data System (ADS)

Shinohara, T.; Kai, T.; Oikawa, K.; Segawa, M.; Harada, M.; Nakatani, T.; Ooi, M.; Aizawa, K.; Sato, H.; Kamiyama, T.; Yokota, H.; Sera, T.; Mochiki, K.; Kiyanagi, Y.

2016-09-01

A new pulsed-neutron instrument, named the Energy-Resolved Neutron Imaging System “RADEN”, has been constructed at the beam line of BL22 in the Materials and Life Science Experimental Facility (MLF) of J-PARC. The primary purpose of this instrument is to perform energy-resolved neutron imaging experiments through the effective utilization of the pulsed nature of the neutron beam, making this the world's first instrument dedicated to pulsed neutron imaging experiments. RADEN was designed to cover a broad energy range: from cold neutrons with energy down to 1.05 meV (or wavelength up to 8.8 Å) with a good wavelength resolution of 0.20% to high-energy neutrons with energy of several tens keV (or wavelength of 10-3 Å). In addition, this instrument is intended to perform state-of-the-art neutron radiography and tomography experiments in Japan. Hence, a maximum beam size of 300 mm square and a high L/D value of up to 7500 are provided.

NEUTRON CHARACTERIZATION OF ENSA-DPT TYPE SPENT FUEL CASK AT TRILLO NUCLEAR POWER PLANT.

PubMed

Méndez-Villafañe, Roberto; Campo-Blanco, Xandra; Embid, Miguel; Yéboles, César A; Morales, Ramón; Novo, Manuel; Sanz, Javier

2018-04-23

The Neutron Standards Laboratory of CIEMAT has conducted the characterization of the independent spent fuel storage installation at the Trillo Nuclear Power Plant. At this facility, the spent fuel assemblies are stored in ENSA-DPT type dual purpose casks. Neutron characterization was performed by dosimetry measurements with a neutron survey meter (LB6411) inside the facility, around an individual cask and between stored casks, and outside the facility. Spectra measurements were also performed with a Bonner sphere system in order to determine the integral quantities and validate the use of the neutron monitor at the different positions. Inside the facility, measured neutron spectra and neutron ambient dose equivalent rate are consistent with the casks spatial distribution and neutron emission rates, and measurements with both instruments are consistent with each other. Outside the facility, measured neutron ambient dose equivalent rates are well below the 0.5 μSv/h limit established by the nuclear regulatory authority.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelley, R.P., E-mail: rpkelley@ufl.edu; Ray, H.; Jordan, K.A.

An empirical investigation of the scintillation mechanism in a pressurized {sup 4}He gas fast neutron detector was conducted using pulse shape fitting. Scintillation signals from neutron interactions were measured and averaged to produce a single generic neutron pulse shape from both a {sup 252}Cf spontaneous fission source and a (d,d) neutron generator. An expression for light output over time was then developed by treating the decay of helium excited states in the same manner as the decay of radioactive isotopes. This pulse shape expression was fitted to the measured neutron pulse shape using a least-squares optimization algorithm, allowing an empiricalmore » analysis of the mechanism of scintillation inside the {sup 4}He detector. A further understanding of this mechanism in the {sup 4}He detector will advance the use of this system as a neutron spectrometer. For {sup 252}Cf neutrons, the triplet and singlet time constants were found to be 970 ns and 686 ns, respectively. For neutrons from the (d,d) generator, the time constants were found to be 884 ns and 636 ns. Differences were noted in the magnitude of these parameters compared to previously published data, however the general relationships were noted to be the same and checked with expected trends from theory. Of the excited helium states produced from a {sup 252}Cf neutron interaction, 76% were found to be born as triplet states, similar to the result from the neutron generator of 71%. The two sources yielded similar pulse shapes despite having very different neutron energy spectra, validating the robustness of the fits across various neutron energies.« less

EXCALIBUR-at-CALIBAN: a neutron transmission experiment for {sup 238}U(n,n'{sub continuum}γ) nuclear data validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bernard, David; Leconte, Pierre; Destouches, Christophe

2015-07-01

Two recent papers justified a new experimental program to give a new basis for the validation of {sup 238}U nuclear data, namely neutron induced inelastic scattering and transport codes at neutron fission energies. The general idea is to perform a neutron transmission experiment through natural uranium material. As shown by Hans Bethe, neutron transmissions measured by dosimetric responses are linked to inelastic cross sections. This paper describes the principle and the results of such an experience called EXCALIBUR performed recently (January and October 2014) at the CALIBAN reactor facility. (authors)

Conceptual Design of a Clinical BNCT Beam in an Adjacent Dry Cell of the Jozef Stefan Institute TRIGA Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maucec, Marko

2000-11-15

The MCNP4B Monte Carlo transport code is used in a feasibility study of the epithermal neutron boron neutron capture therapy facility in the thermalizing column of the 250-kW TRIGA Mark II reactor at the Jozef Stefan Institute (JSI). To boost the epithermal neutron flux at the reference irradiation point, the efficiency of a fission plate with almost 1.5 kg of 20% enriched uranium and 2.3 kW of thermal power is investigated. With the same purpose in mind, the TRIGA reactor core setup is optimized, and standard fresh fuel elements are concentrated partly in the outermost ring of the core. Further,more » a detailed parametric study of the materials and dimensions for all the relevant parts of the irradiation facility is carried out. Some of the standard epithermal neutron filter/moderator materials, as well as 'pressed-only' low-density Al{sub 2}O{sub 3} and AlF{sub 3}, are considered. The proposed version of the BNCT facility, with PbF{sub 2} as the epithermal neutron filter/moderator, provides an epithermal neutron flux of {approx}1.1 x 10{sup 9} n/cm{sup 2}.s, thus enabling patient irradiation times of <60 min. With reasonably low fast neutron and photon contamination ([overdot]D{sub nfast}/{phi}{sub epi} < 5 x 10{sup -13} Gy.cm{sup 2}/n and [overdot]D{sub {gamma}} /{phi}{sub epi} < 3 x 10{sup -13} Gy.cm{sup 2}/n), the in-air performances of the proposed beam are comparable to all existing epithermal BNCT facilities. The design presents an equally efficient alternative to the BNCT beams in TRIGA reactor thermal columns that are more commonly applied. The cavity of the dry cell, a former JSI TRIGA reactor spent-fuel storage facility, adjacent to the thermalizing column, could rather easily be rearranged into a suitable patient treatment room, which would substantially decrease the overall developmental costs.« less

Neutron detection with LiInSe2

NASA Astrophysics Data System (ADS)

Bell, Zane W.; Burger, A.; Matei, Liviu; Groza, Michael; Stowe, Ashley; Tower, Joshua; Kargar, Alireza; Hong, Huicong

2015-08-01

The detection of thermal neutrons has traditionally been accomplished with 3He-tubes, but with the recent shortage of 3He, much research has gone into finding suitable replacements. Both relatively inefficient 10B- and 6LiF-coated silicon diodes and HgI2 have been known for many years, and engineered structures in Si that have been filled with 10B and 6LiF have shown promise. These devices are intended to realize an optimal juxtaposition of neutron-sensitive material and semiconductor and thereby simulate a semiconductor containing B or Li. Such material has been realized for the first time in the form of 6LiInSe2 in which collectable charge from the 6Li(n,t) reaction indicates a neutron event. In this paper we report neutron and gamma responses of 6LiInSe2, we show pulse height spectra from pure gamma sources and from a thermal neutron source, and we derive the μτ product from the position of spectral features as a function of bias voltage. In addition, we demonstrate the observation of the beta decay of 116mIn in samples exposed to thermal neutrons. This feature of the response serves as an additional confirmation of exposure to neutrons.

Controlling neutron orbital angular momentum

NASA Astrophysics Data System (ADS)

Clark, Charles W.; Barankov, Roman; Huber, Michael G.; Arif, Muhammad; Cory, David G.; Pushin, Dmitry A.

2015-09-01

The quantized orbital angular momentum (OAM) of photons offers an additional degree of freedom and topological protection from noise. Photonic OAM states have therefore been exploited in various applications ranging from studies of quantum entanglement and quantum information science to imaging. The OAM states of electron beams have been shown to be similarly useful, for example in rotating nanoparticles and determining the chirality of crystals. However, although neutrons--as massive, penetrating and neutral particles--are important in materials characterization, quantum information and studies of the foundations of quantum mechanics, OAM control of neutrons has yet to be achieved. Here, we demonstrate OAM control of neutrons using macroscopic spiral phase plates that apply a `twist' to an input neutron beam. The twisted neutron beams are analysed with neutron interferometry. Our techniques, applied to spatially incoherent beams, demonstrate both the addition of quantum angular momenta along the direction of propagation, effected by multiple spiral phase plates, and the conservation of topological charge with respect to uniform phase fluctuations. Neutron-based studies of quantum information science, the foundations of quantum mechanics, and scattering and imaging of magnetic, superconducting and chiral materials have until now been limited to three degrees of freedom: spin, path and energy. The optimization of OAM control, leading to well defined values of OAM, would provide an additional quantized degree of freedom for such studies.

Detection of explosives, shielded nuclear materials and other hazardous substances in cargo containers

NASA Astrophysics Data System (ADS)

Kuznetsov, Andrey; Evsenin, Alexey; Vakhtin, Dmitry; Gorshkov, Igor; Osetrov, Oleg; Kalinin, Valery

2006-05-01

Nanosecond Neutron Analysis / Associated Particles Technique (NNA/APT) has been used to create devices for detection of explosives, radioactive and heavily shielded nuclear materials in cargo containers. Explosives and other hazardous materials are detected by analyzing secondary high-energy gamma-rays form reactions of fast neutrons with the materials inside the container. Depending on the dimensions of the inspected containers, the detecting system consists of one or several detection modules, each of which contains a small neutron generator with built-in position sensitive detector of associated alpha-particles and several scintillator-based gamma-ray detectors. The same gamma-ray detectors are used to detect unshielded radioactive and nuclear materials. Array of several detectors of fast neutrons is used to detect neutrons from spontaneous and induced fission of nuclear materials. These neutrons can penetrate thick layers of lead shielding, which can be used to conceal gamma-radioactivity from nuclear materials. Coincidence and timing analysis allows one to discriminate between fission neutrons and scattered probing neutrons. Mathematical modeling by MCNP5 code was used to estimate the sensitivity of the device and its optimal configuration. Capability of the device to detect 1 kg of explosive imitator inside container filled with suitcases and other baggage items has been confirmed experimentally. First experiments with heavily shielded nuclear materials have been carried out.

Advances in Neutron Radiography: Application to Additive Manufacturing Inconel 718

DOE PAGES

Bilheux, Hassina Z; Song, Gian; An, Ke; ...

2016-01-01

Reactor-based neutron radiography is a non-destructive, non-invasive characterization technique that has been extensively used for engineering materials such as inspection of components, evaluation of porosity, and in-operando observations of engineering parts. Neutron radiography has flourished at reactor facilities for more than four decades and is relatively new to accelerator-based neutron sources. Recent advances in neutron source and detector technologies, such as the Spallation Neutron Source (SNS) at the Oak Ridge National Laboratory (ORNL) in Oak Ridge, TN, and the microchannel plate (MCP) detector, respectively, enable new contrast mechanisms using the neutron scattering Bragg features for crystalline information such as averagemore » lattice strain, crystalline plane orientation, and identification of phases in a neutron radiograph. Additive manufacturing (AM) processes or 3D printing have recently become very popular and have a significant potential to revolutionize the manufacturing of materials by enabling new designs with complex geometries that are not feasible using conventional manufacturing processes. However, the technique lacks standards for process optimization and control compared to conventional processes. Residual stresses are a common occurrence in materials that are machined, rolled, heat treated, welded, etc., and have a significant impact on a component s mechanical behavior and durability. They may also arise during the 3D printing process, and defects such as internal cracks can propagate over time as the component relaxes after being removed from its build plate (the base plate utilized to print materials on). Moreover, since access to the AM material is possible only after the component has been fully manufactured, it is difficult to characterize the material for defects a priori to minimize expensive re-runs. Currently, validation of the AM process and materials is mainly through expensive trial-and-error experiments at the component level, whereas in conventional processes the level of confidence in predictive computational modeling is high enough to allow process and materials optimization through computational approaches. Thus, there is a clear need for non-destructive characterization techniques and for the establishment of processing- microstructure databases that can be used for developing and validating predictive modeling tools for AM.« less

Performance of the upgraded ultracold neutron source at Los Alamos National Laboratory and its implication for a possible neutron electric dipole moment experiment

NASA Astrophysics Data System (ADS)

Ito, T. M.; Adamek, E. R.; Callahan, N. B.; Choi, J. H.; Clayton, S. M.; Cude-Woods, C.; Currie, S.; Ding, X.; Fellers, D. E.; Geltenbort, P.; Lamoreaux, S. K.; Liu, C.-Y.; MacDonald, S.; Makela, M.; Morris, C. L.; Pattie, R. W.; Ramsey, J. C.; Salvat, D. J.; Saunders, A.; Sharapov, E. I.; Sjue, S.; Sprow, A. P.; Tang, Z.; Weaver, H. L.; Wei, W.; Young, A. R.

2018-01-01

The ultracold neutron (UCN) source at Los Alamos National Laboratory (LANL), which uses solid deuterium as the UCN converter and is driven by accelerator spallation neutrons, has been successfully operated for over 10 years, providing UCN to various experiments, as the first production UCN source based on the superthermal process. It has recently undergone a major upgrade. This paper describes the design and performance of the upgraded LANL UCN source. Measurements of the cold neutron spectrum and UCN density are presented and compared to Monte Carlo predictions. The source is shown to perform as modeled. The UCN density measured at the exit of the biological shield was 184 (32 ) UCN /cm3 , a fourfold increase from the highest previously reported. The polarized UCN density stored in an external chamber was measured to be 39 (7 ) UCN /cm3 , which is sufficient to perform an experiment to search for the nonzero neutron electric dipole moment with a one-standard-deviation sensitivity of σ (dn) =3 ×10-27e cm .

Development of a detector in order to investigate (n,γ)-cross sections by ToF method with a very short flight path

NASA Astrophysics Data System (ADS)

Wolf, C.; Glorius, J.; Reifarth, R.; Weigand, M.

2018-01-01

The determination of neutron capture cross sections of some radioactive isotopes like 85Kr is very important to improve the knowledge about the s process. Based on its own radioactive decay these isotopes can only be used in small samples inside a TOF facility, which is why the neutron flux of these facilities has to be very high. Unfortunately the neutron flux of the FRANZ setup at Goethe University Frankfurt, which will offer the highest neutron flux in astrophysical energy regions (keV region) [1], is still to low to investigate isotopes like 85Kr. Therefore a new setup called NAUTILUS is under development, which will reduce the flight path from 80 cm to a few centimeter to enhance the angular coverage of the sample and therefore increase the neutron flux by a factor of nearly 100. This implies a higher intensity of the γ-flash energy inside the detector and the neutron induced background. Hence the geometry, the scintillator material and the moderator were optimized by GEANT3 simulations.

Biomolecular Deuteration for Neutron Structural Biology and Dynamics.

PubMed

Haertlein, Michael; Moulin, Martine; Devos, Juliette M; Laux, Valerie; Dunne, Orla; Forsyth, V Trevor

2016-01-01

Neutron scattering studies provide important information in structural biology that is not accessible using other approaches. The uniqueness of the technique, and its complementarity with X-ray scattering, is greatest when full use is made of deuterium labeling. The ability to produce tailor-made deuterium-labeled biological macromolecules allows neutron studies involving solution scattering, crystallography, reflection, and dynamics to be optimized in a manner that has major impact on the scope, quality, and throughput of work in these areas. Deuteration facilities have now been developed at many neutron centres throughout the world; these are having a crucial effect on neutron studies in the life sciences and on biologically related studies in soft matter. This chapter describes methods that have been developed for the efficient production of deuterium-labeled samples for a wide range of neutron scattering applications. Examples are given that illustrate the use of these samples for each of the main techniques. Perspectives for biological deuterium labeling are discussed in relation to developments at current facilities and those that are planned in the future. © 2016 Elsevier Inc. All rights reserved.

A real-time neutron-gamma discriminator based on the support vector machine method for the time-of-flight neutron spectrometer

NASA Astrophysics Data System (ADS)

Wei, ZHANG; Tongyu, WU; Bowen, ZHENG; Shiping, LI; Yipo, ZHANG; Zejie, YIN

2018-04-01

A new neutron-gamma discriminator based on the support vector machine (SVM) method is proposed to improve the performance of the time-of-flight neutron spectrometer. The neutron detector is an EJ-299-33 plastic scintillator with pulse-shape discrimination (PSD) property. The SVM algorithm is implemented in field programmable gate array (FPGA) to carry out the real-time sifting of neutrons in neutron-gamma mixed radiation fields. This study compares the ability of the pulse gradient analysis method and the SVM method. The results show that this SVM discriminator can provide a better discrimination accuracy of 99.1%. The accuracy and performance of the SVM discriminator based on FPGA have been evaluated in the experiments. It can get a figure of merit of 1.30.

Precise calculations in simulations of the interaction of low energy neutrons with nano-dispersed media

NASA Astrophysics Data System (ADS)

Artem'ev, V. A.; Nezvanov, A. Yu.; Nesvizhevsky, V. V.

2016-01-01

We discuss properties of the interaction of slow neutrons with nano-dispersed media and their application for neutron reflectors. In order to increase the accuracy of model simulation of the interaction of neutrons with nanopowders, we perform precise quantum mechanical calculation of potential scattering of neutrons on single nanoparticles using the method of phase functions. We compare results of precise calculations with those performed within first Born approximation for nanodiamonds with the radius of 2-5 nm and for neutron energies 3 × 10-7-10-3 eV. Born approximation overestimates the probability of scattering to large angles, while the accuracy of evaluation of integral characteristics (cross sections, albedo) is acceptable. Using Monte-Carlo method, we calculate albedo of neutrons from different layers of piled up diamond nanopowder.

Neutron residual stress measurements on rail sections for different production conditions

DOT National Transportation Integrated Search

2004-11-13

Rail sectioning with subsequent neutron diffraction experiments has been used to assess residual stresses in the rails. In this study we present the results of neutron stress : measurements performed at the NIST Center for Neutron Research (NCNR) on ...

Albedo Neutron Dosimetry in a Deep Geological Disposal Repository for High-Level Nuclear Waste.

PubMed

Pang, Bo; Becker, Frank

2017-04-28

Albedo neutron dosemeter is the German official personal neutron dosemeter in mixed radiation fields where neutrons contribute to personal dose. In deep geological repositories for high-level nuclear waste, where neutrons can dominate the radiation field, it is of interest to investigate the performance of albedo neutron dosemeter in such facilities. In this study, the deep geological repository is represented by a shielding cask loaded with spent nuclear fuel placed inside a rock salt emplacement drift. Due to the backscattering of neutrons in the drift, issues concerning calibration of the dosemeter arise. Field-specific calibration of the albedo neutron dosemeter was hence performed with Monte Carlo simulations. In order to assess the applicability of the albedo neutron dosemeter in a deep geological repository over a long time scale, spent nuclear fuel with different ages of 50, 100 and 500 years were investigated. It was found out, that the neutron radiation field in a deep geological repository can be assigned to the application area 'N1' of the albedo neutron dosemeter, which is typical in reactors and accelerators with heavy shielding. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Neutron detection using a water Cherenkov detector with pure water and a single PMT

NASA Astrophysics Data System (ADS)

Sidelnik, Iván; Asorey, Hernán; Blostein, Juan Jerónimo; Gómez Berisso, Mariano

2017-12-01

We present the performance of a novel neutron detector based on a water Cherenkov detector (WCD) employing pure water and a single photomultiplier tube (PMT). The experiments presented in this work were performed using 241AmBe and 252Cf neutron sources in different neutron moderator and shielding configurations. We show that fast neutrons from the 241AmBe and 241Cf sources, as well as thermal neutrons from a neutron moderator, despite having different spectral characteristics, produce essentially the same pulse histogram shape. This characteristic pulse-height histogram shapes are recorded as a clear signature of neutrons with energies lower than ≃ 11 MeV . This is verified in different experimental conditions. Our estimation of the neutron detection efficiency is at the level of (15±5)%, for fast neutrons. Since water is the material employed as active volume, the results of this study are of interest for the construction of low cost and large active volume neutron detectors for various applications. Of special importance are those related with space weather phenomena monitoring as well as those for the detection of fissile special nuclear material, including uranium or plutonium.

Neutronic analysis of the 1D and 1E banks reflux detection system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blanchard, A.

1999-12-21

Two H Canyon neutron monitoring systems for early detection of postulated abnormal reflux conditions in the Second Uranium Cycle 1E and 1D Mixer-Settle Banks have been designed and built. Monte Carlo neutron transport simulations using the general purpose, general geometry, n-particle MCNP code have been performed to model expected response of the monitoring systems to varying conditions.The confirmatory studies documented herein conclude that the 1E and 1D neutron monitoring systems are able to achieve adequate neutron count rates for various neutron source and detector configurations, thereby eliminating excessive integration count time. Neutron count rate sensitivity studies are also performed. Conversely,more » the transport studies concluded that the neutron count rates are statistically insensitive to nitric acid content in the aqueous region and to the transition region length. These studies conclude that the 1E and 1D neutron monitoring systems are able to predict the postulated reflux conditions for all examined perturbations in the neutron source and detector configurations. In the cases examined, the relative change in the neutron count rates due to postulated transitions from normal {sup 235}U concentration levels to reflux levels remain satisfactory detectable.« less

Etude des performances de solveurs deterministes sur un coeur rapide a caloporteur sodium

NASA Astrophysics Data System (ADS)

Bay, Charlotte

The reactors of next generation, in particular SFR model, represent a true challenge for current codes and solvers, used mainly for thermic cores. There is no guarantee that their competences could be straight adapted to fast neutron spectrum, or to major design differences. Thus it is necessary to assess the validity of solvers and their potential shortfall in the case of fast neutron reactors. As part of an internship with CEA (France), and at the instigation of EPM Nuclear Institute, this study concerns the following codes : DRAGON/DONJON, ERANOS, PARIS and APOLLO3. The precision assessment has been performed using Monte Carlo code TRIPOLI4. Only core calculation was of interest, namely numerical methods competences in precision and rapidity. Lattice code was not part of the study, that is to say nuclear data, self-shielding, or isotopic compositions. Nor was tackled burnup or time evolution effects. The study consists in two main steps : first evaluating the sensitivity of each solver to calculation parameters, and obtain its optimal calculation set ; then compare their competences in terms of precision and rapidity, by collecting usual quantities (effective multiplication factor, reaction rates map), but also more specific quantities which are crucial to the SFR design, namely control rod worth and sodium void effect. The calculation time is also a key factor. Whatever conclusion or recommendation that could be drawn from this study, they must first of all be applied within similar frameworks, that is to say small fast neutron cores with hexagonal geometry. Eventual adjustments for big cores will have to be demonstrated in developments of this study.

Distinguishing the importance of fullerene phase separation from polymer ordering in the performance of low band gap polymer: Bis-fullerene heterojunctions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, Huipeng; Hsiao, Yu -Che; Chen, Jihua

2014-09-16

It is known, one way to improve power conversion efficiency (PCE) of polymer based bulk-heterojunction (BHJ) photovoltaic cells is to increase the open circuit voltage (V oc). Replacing PCBM with bis-adduct fullerenes significantly improves V oc and the PCE in devices based on the conjugated polymer poly(3-hexyl thiophene) (P3HT). However, for the most promising low band-gap polymer (LBP) system, replacing PCBM with ICBA results in poor short-circuit current (J sc) and PCE although V oc is significantly improved. The optimization of the morphology of as-cast LBP/bis-fullerene BHJ photovoltaics is attempted by adding a co-solvent to the polymer/fullerene solution prior tomore » film deposition. Varying the solubility of polymer and fullerene in the co-solvent, bulk heterojunctions are fabricated with no change of polymer ordering, but with changes in fullerene phase separation. The morphologies of the as-cast samples are characterized by small angle neutron scattering and neutron reflectometry. A homogenous dispersion of ICBA in LBP is found in the samples where the co-solvent is selective to the polymer, giving poor device performance. Aggregates of ICBA are formed in samples where the co-solvent is selective to ICBA. Furthermore, the resultant morphology improves PCE by up to 246%. Finally, a quantitative analysis of the neutron data shows that the interfacial area between ICBA aggregates and its surrounding matrix is improved, facilitating charge transport and improving the PCE.« less

Development of an Efficient Approach to Perform Neutronics Simulations for Plutonium-238 Production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chandler, David; Ellis, Ronald James

Conversion of 238Pu decay heat into usable electricity is imperative to power National Aeronautics and Space Administration (NASA) deep space exploration missions; however, the current stockpile of 238Pu is diminishing and the quality is less than ideal. In response, the US Department of Energy and NASA have undertaken a program to reestablish a domestic 238Pu production program and a technology demonstration sub-project has been initiated. Neutronics simulations for 238Pu production play a vital role in this project because the results guide reactor safety-basis, target design and optimization, and post-irradiation examination activities. A new, efficient neutronics simulation tool written in Pythonmore » was developed to evaluate, with the highest fidelity possible with approved tools, the time-dependent nuclide evolution and heat deposition rates in 238Pu production targets irradiated in the High Flux Isotope Reactor (HFIR). The Python Activation and Heat Deposition Script (PAHDS) was developed specifically for experiment analysis in HFIR and couples the MCNP5 and SCALE 6.1.3 software quality assured tools to take advantage of an existing high-fidelity MCNP HFIR model, the most up-to-date ORIGEN code, and the most up-to-date nuclear data. Three cycle simulations were performed with PAHDS implementing ENDF/B-VII.0, ENDF/B-VII.1, and the Hybrid Library GPD-Rev0 cross-section libraries. The 238Pu production results were benchmarked against VESTA-obtained results and the impact of various cross-section libraries on the calculated metrics were assessed.« less

ENERGY RESPONSE OF FLUORESCENT NUCLEAR TRACK DETECTORS OF VARIOUS COLORATIONS TO MONOENERGETIC NEUTRONS.

PubMed

Fomenko, V; Moreno, B; Million, M; Harrison, J; Akselrod, M

2017-10-25

The neutron-energy dependence of the track-counting sensitivity of fluorescent nuclear track detectors (FNTDs) at two ranges of Mg doping, resulting in different crystal colorations, was investigated. The performance of FNTDs was studied with the following converters: Li-glass for thermal to intermediate-energy neutrons, polyethylene for fast neutrons, and polytetrafluoroethylene (Teflon™) for photon- and radon-background subtraction. The irradiations with monoenergetic neutrons were performed at the National Physics Laboratory (NPL), UK. The energy range was varied from 144 keV to 16.5 MeV in the personal dose equivalent range from 1 to 3 mSv. Monte Carlo simulations were performed to model the response of FNTDs to monoenergetic neutrons. A good agreement with the experimental data was observed suggesting the development of a basic model for future MC studies. Further work will focus on increasing FNTD sensitivity to low-energy neutrons and developing a faster imaging technique for scanning larger areas to improve counting statistics. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

How to polarise all neutrons in one beam: a high performance polariser and neutron transport system

NASA Astrophysics Data System (ADS)

Rodriguez, D. Martin; Bentley, P. M.; Pappas, C.

2016-09-01

Polarised neutron beams are used in disciplines as diverse as magnetism,soft matter or biology. However, most of these applications often suffer from low flux also because the existing neutron polarising methods imply the filtering of one of the spin states, with a transmission of 50% at maximum. With the purpose of using all neutrons that are usually discarded, we propose a system that splits them according to their polarisation, flips them to match the spin direction, and then focuses them at the sample. Monte Carlo (MC) simulations show that this is achievable over a wide wavelength range and with an outstanding performance at the price of a more divergent neutron beam at the sample position.

Perforated semiconductor neutron detectors for battery operated portable modules

NASA Astrophysics Data System (ADS)

McGregor, Douglas S.; Bellinger, Steven L.; Bruno, David; McNeil, Walter J.; Patterson, Eric; Shultis, J. Kenneth; Solomon, C. J.; Unruh, Troy

2007-09-01

Perforated semiconductor diode detectors have been under development for several years at Kansas State University for a variety of neutron detection applications. The fundamental device configuration is a pin diode detector fabricated from high-purity float zone refined Si wafers. Perforations are etched into the diode surface with inductively-coupled plasma (ICP) reactive ion etching (RIE) and backfilled with 6LiF neutron reactive material. The perforation shapes and depths can be optimized to yield a flat response to neutrons over a wide variation of angles. The prototype devices delivered over 3.8% thermal neutron detection efficiency while operating on only 15 volts. The highest efficiency devices thus far have delivered over 12% thermal neutron detection efficiency. The miniature devices are 5.6 mm in diameter and require minimal power to operate, ranging from 3.3 volts to 15 volts, depending upon the amplifying electronics. The battery operated devices have been incorporated into compact modules with a digital readout. Further, the new modules have incorporated wireless readout technology and can be monitored remotely. The neutron detection modules can be used for neutron dosimetry and neutron monitoring. When coupled with high-density polyethylene, the detectors can be used to measure fission neutrons from spontaneous fission sources. Monto Carlo analysis indicates that the devices can be used in cargo containers as a passive search tool for spontaneous fission sources, such as 240Pu. Measurements with a 252Cf source are being conducted for verification.

New developments in the McStas neutron instrument simulation package

NASA Astrophysics Data System (ADS)

Willendrup, P. K.; Knudsen, E. B.; Klinkby, E.; Nielsen, T.; Farhi, E.; Filges, U.; Lefmann, K.

2014-07-01

The McStas neutron ray-tracing software package is a versatile tool for building accurate simulators of neutron scattering instruments at reactors, short- and long-pulsed spallation sources such as the European Spallation Source. McStas is extensively used for design and optimization of instruments, virtual experiments, data analysis and user training. McStas was founded as a scientific, open-source collaborative code in 1997. This contribution presents the project at its current state and gives an overview of the main new developments in McStas 2.0 (December 2012) and McStas 2.1 (expected fall 2013), including many new components, component parameter uniformisation, partial loss of backward compatibility, updated source brilliance descriptions, developments toward new tools and user interfaces, web interfaces and a new method for estimating beam losses and background from neutron optics.

Materials for Low-Energy Neutron Radiation Shielding

NASA Technical Reports Server (NTRS)

Singleterry, Robert C., Jr.; Thibeault, Sheila A.

2000-01-01

Various candidate aircraft and spacecraft materials were analyzed and compared in a low-energy neutron environment using the Monte Carlo N-Particle (MCNP) transport code with an energy range up to 20 MeV. Some candidate materials have been tested in particle beams, and others seemed reasonable to analyze in this manner before deciding to test them. The two metal alloys analyzed are actual materials being designed into or used in aircraft and spacecraft today. This analysis shows that hydrogen-bearing materials have the best shielding characteristics over the metal alloys. It also shows that neutrons above 1 MeV are reflected out of the face of the slab better by larger quantities of carbon in the material. If a low-energy absorber is added to the material, fewer neutrons are transmitted through the material. Future analyses should focus on combinations of scatterers and absorbers to optimize these reaction channels and on the higher energy neutron component (above 50 MeV).

Stabilization of Particle Discrimination Efficiencies for Neutron Spectrum Unfolding With Organic Scintillators

NASA Astrophysics Data System (ADS)

Lawrence, Chris C.; Polack, J. K.; Febbraro, Michael; Kolata, J. J.; Flaska, Marek; Pozzi, S. A.; Becchetti, F. D.

2017-02-01

The literature discussing pulse-shape discrimination (PSD) in organic scintillators dates back several decades. However, little has been written about PSD techniques that are optimized for neutron spectrum unfolding. Variation in n-γ misclassification rates and in γ/n ratio of incident fields can distort the neutron pulse-height response of scintillators and these distortions can in turn cause large errors in unfolded spectra. New applications in arms-control verification call for detection of lower-energy neutrons, for which PSD is particularly problematic. In this article, we propose techniques for removing distortions on pulse-height response that result from the merging of PSD distributions in the low-pulse-height region. These techniques take advantage of the repeatable shapes of PSD distributions that are governed by the counting statistics of scintillation-photon populations. We validate the proposed techniques using accelerator-based time-of-flight measurements and then demonstrate them by unfolding the Watt spectrum from measurement with a 252Cf neutron source.

γ-decay of {}_{8}^{16}{{\\rm{O}}}_{8}\\,{and}\\,{}_{7}^{16}{{\\rm{N}}}_{9} in proton-neutron Tamm-Dancoff and random phase approximations with optimized surface δ interaction

NASA Astrophysics Data System (ADS)

Pahlavani, M. R.; Firoozi, B.

2016-09-01

γ-ray transitions from excited states of {}16{{N}} and {}16{{O}} isomers that appear in the γ spectrum of the {}616{{{C}}}10\\to {}716{{{N}}}9\\to {}816{{{O}}}8 beta decay chain are investigated. The theoretical approach used in this research starts with a mean-field potential consisting of a phenomenological Woods-Saxon potential including spin-orbit and Coulomb terms (for protons) in order to obtain single-particle energies and wave functions for nucleons in a nucleus. A schematic residual surface delta interaction is then employed on the top of the mean field and is treated within the proton-neutron Tamm-Dancoff approximation (pnTDA) and the proton-neutron random phase approximation. The goal is to use an optimized surface delta interaction interaction, as a residual interaction, to improve the results. We have used artificial intelligence algorithms to establish a good agreement between theoretical and experimental energy spectra. The final results of the ‘optimized’ calculations are reasonable via this approach.

Assessing and minimizing contamination in time of flight based validation data

NASA Astrophysics Data System (ADS)

Lennox, Kristin P.; Rosenfield, Paul; Blair, Brenton; Kaplan, Alan; Ruz, Jaime; Glenn, Andrew; Wurtz, Ronald

2017-10-01

Time of flight experiments are the gold standard method for generating labeled training and testing data for the neutron/gamma pulse shape discrimination problem. As the popularity of supervised classification methods increases in this field, there will also be increasing reliance on time of flight data for algorithm development and evaluation. However, time of flight experiments are subject to various sources of contamination that lead to neutron and gamma pulses being mislabeled. Such labeling errors have a detrimental effect on classification algorithm training and testing, and should therefore be minimized. This paper presents a method for identifying minimally contaminated data sets from time of flight experiments and estimating the residual contamination rate. This method leverages statistical models describing neutron and gamma travel time distributions and is easily implemented using existing statistical software. The method produces a set of optimal intervals that balance the trade-off between interval size and nuisance particle contamination, and its use is demonstrated on a time of flight data set for Cf-252. The particular properties of the optimal intervals for the demonstration data are explored in detail.

Effect of Symmetry on Performance of Imploding Capsules using the Big Foot Design

NASA Astrophysics Data System (ADS)

Khan, Shahab; Casey, Daniel; Baker, Kevin; Thomas, Cliff; Nora, Ryan; Spears, Brian; Benedetti, Laura; Izumi, Nobuhiko; Ma, Tammy; Nagel, Sabrina; Pak, Arthur; National Ignition Facility Collaboration

2017-10-01

At the National Ignition Facility, several simultaneous designs are investigated for optimizing Inertial Confinement Fusion (ICF) energy gain of indirectly driven imploding fuel capsules. Relatively high neutron yield has been achieved while exhibiting a non-symmetric central core and/or shell. While developing the ``Big Foot'' design, several tuning steps were undertaken to minimize the asymmetry of both the central hot core as well as the shell. Surrogate capsules (symcaps) were utilized in the 2-D Radiography platform to assess both the shell and central core symmetry. The results of the tuning experiments are presented. In addition, a comparison of performance and shape metrics demonstrates that improving symmetry of the implosion can yield better performance. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344. LLNL-ABS-683471.

Neutron density profile in the lunar subsurface produced by galactic cosmic rays

NASA Astrophysics Data System (ADS)

Ota, Shuya; Sihver, Lembit; Kobayashi, Shingo; Hasebe, Nobuyuki

Neutron production by galactic cosmic rays (GCR) in the lunar subsurface is very important when performing lunar and planetary nuclear spectroscopy and space dosimetry. Further im-provements to estimate the production with increased accuracy is therefore required. GCR, which is a main contributor to the neutron production in the lunar subsurface, consists of not only protons but also of heavy components such as He, C, N, O, and Fe. Because of that, it is important to precisely estimate the neutron production from such components for the lunar spectroscopy and space dosimetry. Therefore, the neutron production from GCR particles in-cluding heavy components in the lunar subsurface was simulated with the Particle and Heavy ion Transport code System (PHITS), using several heavy ion interaction models. This work presents PHITS simulations of the neutron density as a function of depth (neutron density profile) in the lunar subsurface and the results are compared with experimental data obtained by Apollo 17 Lunar Neutron Probe Experiment (LNPE). From our previous study, it has been found that the accuracy of the proton-induced neutron production models is the most influen-tial factor when performing precise calculations of neutron production in the lunar subsurface. Therefore, a benchmarking of proton-induced neutron production models against experimental data was performed to estimate and improve the precision of the calculations. It was found that the calculated neutron production using the best model of Cugnon Old (E < 3 GeV) and JAM (E > 3 GeV) gave up to 30% higher values than experimental results. Therefore, a high energy nuclear data file (JENDL-HE) was used instead of the Cugnon Old model at the energies below 3 GeV. Then, the calculated neutron density profile successfully reproduced the experimental data from LNPE within experimental errors of 15% (measurement) + 30% (systematic). In this presentation, we summarize and discuss our calculated results of neutron production in the lunar subsurface.

Development of a Geant4 application to characterise a prototype neutron detector based on three orthogonal 3He tubes inside an HDPE sphere.

PubMed

Gracanin, V; Guatelli, S; Prokopovich, D; Rosenfeld, A B; Berry, A

2017-01-01

The Bonner Sphere Spectrometer (BSS) system is a well-established technique for neutron dosimetry that involves detection of thermal neutrons within a range of hydrogenous moderators. BSS detectors are often used to perform neutron field surveys in order to determine the ambient dose equivalent H*(10) and estimate health risk to personnel. There is a potential limitation of existing neutron survey techniques, since some detectors do not consider the direction of the neutron field, which can result in overly conservative estimates of dose in neutron fields. This paper shows the development of a Geant4 simulation application to characterise a prototype neutron detector based on three orthogonal 3 He tubes inside a single HDPE sphere built at the Australian Nuclear Science and Technology Organisation (ANSTO). The Geant4 simulation has been validated with respect to experimental measurements performed with an Am-Be source. Crown Copyright © 2016. Published by Elsevier Ltd. All rights reserved.

Fusion neutron source blanket: requirements for calculation accuracy and benchmark experiment precision

NASA Astrophysics Data System (ADS)

Zhirkin, A. V.; Alekseev, P. N.; Batyaev, V. F.; Gurevich, M. I.; Dudnikov, A. A.; Kuteev, B. V.; Pavlov, K. V.; Titarenko, Yu. E.; Titarenko, A. Yu.

2017-06-01

In this report the calculation accuracy requirements of the main parameters of the fusion neutron source, and the thermonuclear blankets with a DT fusion power of more than 10 MW, are formulated. To conduct the benchmark experiments the technical documentation and calculation models were developed for two blanket micro-models: the molten salt and the heavy water solid-state blankets. The calculations of the neutron spectra, and 37 dosimetric reaction rates that are widely used for the registration of thermal, resonance and threshold (0.25-13.45 MeV) neutrons, were performed for each blanket micro-model. The MCNP code and the neutron data library ENDF/B-VII were used for the calculations. All the calculations were performed for two kinds of neutron source: source I is the fusion source, source II is the source of neutrons generated by the 7Li target irradiated by protons with energy 24.6 MeV. The spectral indexes ratios were calculated to describe the spectrum variations from different neutron sources. The obtained results demonstrate the advantage of using the fusion neutron source in future experiments.

Detailed characterisation of the incident neutron beam on the TOSCA spectrometer

NASA Astrophysics Data System (ADS)

Pinna, Roberto S.; Rudić, Svemir; Capstick, Matthew J.; McPhail, David J.; Pooley, Daniel E.; Howells, Gareth D.; Gorini, Giuseppe; Fernandez-Alonso, Felix

2017-10-01

We report a detailed characterisation of the incident neutron beam on the TOSCA spectrometer. A bespoke time-of-flight neutron monitor has been designed, constructed and used to perform extensive spatially resolved measurements of the absolute neutron flux and its underlying time structure at the instrument sample position. The obtained data give a quantitative understanding of the current instrument beyond neutronic simulations and provide a baseline in order to assess the performance of the upgraded instrument. At an average proton current-on-target of 153 μA (ISIS Target Station 1; at the time of measurements) we have found that the wavelength-integrated neutron flux (from 0.28 Å to 4.65 Å) at the position of the TOSCA instrument sample (spatially averaged across the 3 × 3cm2 surface centred around (0,0) position) is approximately 1 . 2 × 106 neutrons cm-2s-1, while the whole beam has a homogeneous distribution across the 3 . 0 × 3 . 5cm2 sample surface. The spectra reproduced the well-known shape of the neutrons moderated by the room temperature water moderator and exhibit a neutron flux of 7 . 3 × 105 neutrons cm-2s-1Å-1 at 1 Å.

[Shielding design and detection of neutrons from medical and industrial electron accelerators--simple method of design calculation for neutron shielding].

PubMed

Nakamura, T; Uwamino, Y

1986-02-01

The neutron leakage from medical and industrial electron accelerators has become an important problem and its detection and shielding is being performed in their facilities. This study provides a new simple method of design calculation for neutron shielding of those electron accelerator facilities by dividing into the following five categories; neutron dose distribution in the accelerator room, neutron attenuation through the wall and the door in the accelerator room, neutron and secondary photon dose distributions in the maze, neutron and secondary photon attenuation through the door at the end of the maze, neutron leakage outside the facility-skyshine.

Enhancing the performance of a tensioned metastable fluid detector based active interrogation system for the detection of SNM in <1 m3 containers using a D-D neutron interrogation source in moderated/reflected geometries

NASA Astrophysics Data System (ADS)

Grimes, T. F.; Hagen, A. R.; Archambault, B. C.; Taleyarkhan, R. P.

2018-03-01

This paper describes the development of a SNM detection system for interrogating 1m3 cargos via the combination of a D-D neutron interrogation source (with and without reflectors) and tensioned metastable fluid detectors (TMFDs). TMFDs have been previously shown (Taleyarkhan et al., 2008; Grimes et al., 2015; Grimes and Taleyarkhan, 2016; Archambault et al., 2017; Hagen et al., 2016) to be capable of using Threshold Energy Neutron Analysis (TENA) techniques to reject the ∼2.45 MeV D-D interrogating neutrons while still remaining sensitive to >2.45 MeV neutrons resulting from fission in the target (HEU) material. In order to enhance the performance, a paraffin reflector was included around the accelerator head. This reflector was used to direct neutrons into the package to increase the fission signal, lower the energy of the interrogating neutrons to increase the fission cross-section with HEU, and, also to direct interrogating neutrons away from the detectors in order to enhance the required discrimination between interrogating and fission neutrons. Experiments performed with a 239 Pu-Be neutron source and MnO2 indicated that impressive performance gains could be made by placing a parabolic paraffin moderator between the interrogation source and an air-filled cargo container with HEU placed at the center. However, experiments with other cargo fillers (as specified in the well-known ANSI N42.41-2007 report), and with HEU placed in locations other than the center of the package indicated that other reflector geometries might be superior due to over-"focusing" and the increased solid angle effects due to the accommodation of the moderator geometry. The best performance for the worst case of source location and box fill was obtained by placing the reflector only behind the D-D neutron source rather than in front of it. Finally, it was shown that there could be significant gains in the ability to detect concealed SNM by operating the system in multiple geometric configurations. Worst case scenarios were created by filling the box with hydrogenous material and placing the HEU as far away as possible from the neutron source. The performance of the system in the worst-case scenarios were greatly improved by exchanging the location of the accelerator and the opposite TMFD panel half way through interrogation. Using this operation, scenarios with positions of the concealed SNM that were once the most challenging to successfully detect became readily detectable.

Optimization of Boiling Water Reactor Loading Pattern Using Two-Stage Genetic Algorithm

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kobayashi, Yoko; Aiyoshi, Eitaro

2002-10-15

A new two-stage optimization method based on genetic algorithms (GAs) using an if-then heuristic rule was developed to generate optimized boiling water reactor (BWR) loading patterns (LPs). In the first stage, the LP is optimized using an improved GA operator. In the second stage, an exposure-dependent control rod pattern (CRP) is sought using GA with an if-then heuristic rule. The procedure of the improved GA is based on deterministic operators that consist of crossover, mutation, and selection. The handling of the encoding technique and constraint conditions by that GA reflects the peculiar characteristics of the BWR. In addition, strategies suchmore » as elitism and self-reproduction are effectively used in order to improve the search speed. The LP evaluations were performed with a three-dimensional diffusion code that coupled neutronic and thermal-hydraulic models. Strong axial heterogeneities and constraints dependent on three dimensions have always necessitated the use of three-dimensional core simulators for BWRs, so that optimization of computational efficiency is required. The proposed algorithm is demonstrated by successfully generating LPs for an actual BWR plant in two phases. One phase is only LP optimization applying the Haling technique. The other phase is an LP optimization that considers the CRP during reactor operation. In test calculations, candidates that shuffled fresh and burned fuel assemblies within a reasonable computation time were obtained.« less

Tomographic Neutron Imaging using SIRT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gregor, Jens; FINNEY, Charles E A; Toops, Todd J

2013-01-01

Neutron imaging is complementary to x-ray imaging in that materials such as water and plastic are highly attenuating while material such as metal is nearly transparent. We showcase tomographic imaging of a diesel particulate filter. Reconstruction is done using a modified version of SIRT called PSIRT. We expand on previous work and introduce Tikhonov regularization. We show that near-optimal relaxation can still be achieved. The algorithmic ideas apply to cone beam x-ray CT and other inverse problems.

High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Test Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Kapernick, Richard

2007-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer. and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics and assess potential design improvements at relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design is developed: this is followed by engineering design, fabrication, and testing to validate the overall design process. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

Signal and background considerations for the MRSt on the National Ignition Facility (NIF).

PubMed

Wink, C W; Frenje, J A; Hilsabeck, T J; Bionta, R; Khater, H Y; Gatu Johnson, M; Kilkenny, J D; Li, C K; Séguin, F H; Petrasso, R D

2016-11-01

A Magnetic Recoil Spectrometer (MRSt) has been conceptually designed for time-resolved measurements of the neutron spectrum at the National Ignition Facility. Using the MRSt, the goals are to measure the time-evolution of the spectrum with a time resolution of ∼20-ps and absolute accuracy better than 5%. To meet these goals, a detailed understanding and optimization of the signal and background characteristics are required. Through ion-optics, MCNP simulations, and detector-response calculations, it is demonstrated that the goals and a signal-to background >5-10 for the down-scattered neutron measurement are met if the background, consisting of ambient neutrons and gammas, at the MRSt is reduced 50-100 times.

The monochromatic imaging mode of a RITA-type neutron spectrometer

NASA Astrophysics Data System (ADS)

Bahl, C. R. H.; Andersen, P.; Klausen, S. N.; Lefmann, K.

2004-12-01

The imaging monochromatic mode of a neutron spectrometer with a multi-bladed RITA analyser system is so far unexplored. We present analytical calculations that define the mode. It is shown that the mode can be realised for PG (0 0 2) analyser crystals, from incident energies of about 3.2 meV and up, allowing the important cases of 3.7, 5.0 and 13.7 meV. Due to beam divergence, the neutron rays from neighbouring analyser blades are found to overlap slightly. Hence, the optimal use of the monochromatic imaging mode would be found by employing an adjustable radial collimator to limit the spread of the ray from each analyser blade.

Validating the performance of correlated fission multiplicity implementation in radiation transport codes with subcritical neutron multiplication benchmark experiments

DOE PAGES

Arthur, Jennifer; Bahran, Rian; Hutchinson, Jesson; ...

2018-06-14

Historically, radiation transport codes have uncorrelated fission emissions. In reality, the particles emitted by both spontaneous and induced fissions are correlated in time, energy, angle, and multiplicity. This work validates the performance of various current Monte Carlo codes that take into account the underlying correlated physics of fission neutrons, specifically neutron multiplicity distributions. The performance of 4 Monte Carlo codes - MCNP®6.2, MCNP®6.2/FREYA, MCNP®6.2/CGMF, and PoliMi - was assessed using neutron multiplicity benchmark experiments. In addition, MCNP®6.2 simulations were run using JEFF-3.2 and JENDL-4.0, rather than ENDF/B-VII.1, data for 239Pu and 240Pu. The sensitive benchmark parameters that in this workmore » represent the performance of each correlated fission multiplicity Monte Carlo code include the singles rate, the doubles rate, leakage multiplication, and Feynman histograms. Although it is difficult to determine which radiation transport code shows the best overall performance in simulating subcritical neutron multiplication inference benchmark measurements, it is clear that correlations exist between the underlying nuclear data utilized by (or generated by) the various codes, and the correlated neutron observables of interest. This could prove useful in nuclear data validation and evaluation applications, in which a particular moment of the neutron multiplicity distribution is of more interest than the other moments. It is also quite clear that, because transport is handled by MCNP®6.2 in 3 of the 4 codes, with the 4th code (PoliMi) being based on an older version of MCNP®, the differences in correlated neutron observables of interest are most likely due to the treatment of fission event generation in each of the different codes, as opposed to the radiation transport.« less

Validating the performance of correlated fission multiplicity implementation in radiation transport codes with subcritical neutron multiplication benchmark experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arthur, Jennifer; Bahran, Rian; Hutchinson, Jesson

Historically, radiation transport codes have uncorrelated fission emissions. In reality, the particles emitted by both spontaneous and induced fissions are correlated in time, energy, angle, and multiplicity. This work validates the performance of various current Monte Carlo codes that take into account the underlying correlated physics of fission neutrons, specifically neutron multiplicity distributions. The performance of 4 Monte Carlo codes - MCNP®6.2, MCNP®6.2/FREYA, MCNP®6.2/CGMF, and PoliMi - was assessed using neutron multiplicity benchmark experiments. In addition, MCNP®6.2 simulations were run using JEFF-3.2 and JENDL-4.0, rather than ENDF/B-VII.1, data for 239Pu and 240Pu. The sensitive benchmark parameters that in this workmore » represent the performance of each correlated fission multiplicity Monte Carlo code include the singles rate, the doubles rate, leakage multiplication, and Feynman histograms. Although it is difficult to determine which radiation transport code shows the best overall performance in simulating subcritical neutron multiplication inference benchmark measurements, it is clear that correlations exist between the underlying nuclear data utilized by (or generated by) the various codes, and the correlated neutron observables of interest. This could prove useful in nuclear data validation and evaluation applications, in which a particular moment of the neutron multiplicity distribution is of more interest than the other moments. It is also quite clear that, because transport is handled by MCNP®6.2 in 3 of the 4 codes, with the 4th code (PoliMi) being based on an older version of MCNP®, the differences in correlated neutron observables of interest are most likely due to the treatment of fission event generation in each of the different codes, as opposed to the radiation transport.« less

Performance of the upgraded ultracold neutron source at Los Alamos National Laboratory and its implication for a possible neutron electric dipole moment experiment

DOE PAGES

Ito, Takeyasu M.; Adamek, E. R.; Callahan, N. B.; ...

2018-01-29

We report the ultracold neutron (UCN) source at Los Alamos National Laboratory (LANL), which uses solid deuterium as the UCN converter and is driven by accelerator spallation neutrons, has been successfully operated for over 10 years, providing UCN to various experiments, as the first production UCN source based on the superthermal process. It has recently undergone a major upgrade. This paper describes the design and performance of the upgraded LANL UCN source. Measurements of the cold neutron spectrum and UCN density are presented and compared to Monte Carlo predictions. The source is shown to perform as modeled. The UCN densitymore » measured at the exit of the biological shield was 184(32) UCN / cm 3, a fourfold increase from the highest previously reported. Finally, the polarized UCN density stored in an external chamber was measured to be 39(7) UCN / cm 3, which is sufficient to perform an experiment to search for the nonzero neutron electric dipole moment with a one-standard-deviation sensitivity of σ(d n) = 3 × 10 -27 e cm.« less

Performance of the upgraded ultracold neutron source at Los Alamos National Laboratory and its implication for a possible neutron electric dipole moment experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ito, Takeyasu M.; Adamek, E. R.; Callahan, N. B.

We report the ultracold neutron (UCN) source at Los Alamos National Laboratory (LANL), which uses solid deuterium as the UCN converter and is driven by accelerator spallation neutrons, has been successfully operated for over 10 years, providing UCN to various experiments, as the first production UCN source based on the superthermal process. It has recently undergone a major upgrade. This paper describes the design and performance of the upgraded LANL UCN source. Measurements of the cold neutron spectrum and UCN density are presented and compared to Monte Carlo predictions. The source is shown to perform as modeled. The UCN densitymore » measured at the exit of the biological shield was 184(32) UCN / cm 3, a fourfold increase from the highest previously reported. Finally, the polarized UCN density stored in an external chamber was measured to be 39(7) UCN / cm 3, which is sufficient to perform an experiment to search for the nonzero neutron electric dipole moment with a one-standard-deviation sensitivity of σ(d n) = 3 × 10 -27 e cm.« less

Optimization of coupled multiphysics methodology for safety analysis of pebble bed modular reactor

NASA Astrophysics Data System (ADS)

Mkhabela, Peter Tshepo

The research conducted within the framework of this PhD thesis is devoted to the high-fidelity multi-physics (based on neutronics/thermal-hydraulics coupling) analysis of Pebble Bed Modular Reactor (PBMR), which is a High Temperature Reactor (HTR). The Next Generation Nuclear Plant (NGNP) will be a HTR design. The core design and safety analysis methods are considerably less developed and mature for HTR analysis than those currently used for Light Water Reactors (LWRs). Compared to LWRs, the HTR transient analysis is more demanding since it requires proper treatment of both slower and much longer transients (of time scale in hours and days) and fast and short transients (of time scale in minutes and seconds). There is limited operation and experimental data available for HTRs for validation of coupled multi-physics methodologies. This PhD work developed and verified reliable high fidelity coupled multi-physics models subsequently implemented in robust, efficient, and accurate computational tools to analyse the neutronics and thermal-hydraulic behaviour for design optimization and safety evaluation of PBMR concept The study provided a contribution to a greater accuracy of neutronics calculations by including the feedback from thermal hydraulics driven temperature calculation and various multi-physics effects that can influence it. Consideration of the feedback due to the influence of leakage was taken into account by development and implementation of improved buckling feedback models. Modifications were made in the calculation procedure to ensure that the xenon depletion models were accurate for proper interpolation from cross section tables. To achieve this, the NEM/THERMIX coupled code system was developed to create the system that is efficient and stable over the duration of transient calculations that last over several tens of hours. Another achievement of the PhD thesis was development and demonstration of full-physics, three-dimensional safety analysis methodology for the PBMR to provide reference solutions. Investigation of different aspects of the coupled methodology and development of efficient kinetics treatment for the PBMR were carried out, which accounts for all feedback phenomena in an efficient manner. The OECD/NEA PBMR-400 coupled code benchmark was used as a test matrix for the proposed investigations. The integrated thermal-hydraulics and neutronics (multi-physics) methods were extended to enable modeling of a wider range of transients pertinent to the PBMR. First, the effect of the spatial mapping schemes (spatial coupling) was studied and quantified for different types of transients, which resulted in implementation of improved mapping methodology based on user defined criteria. The second aspect that was studied and optimized is the temporal coupling and meshing schemes between the neutronics and thermal-hydraulics time step selection algorithms. The coupled code convergence was achieved supplemented by application of methods to accelerate it. Finally, the modeling of all feedback phenomena in PBMRs was investigated and a novel treatment of cross-section dependencies was introduced for improving the representation of cross-section variations. The added benefit was that in the process of studying and improving the coupled multi-physics methodology more insight was gained into the physics and dynamics of PBMR, which will help also to optimize the PBMR design and improve its safety. One unique contribution of the PhD research is the investigation of the importance of the correct representation of the three-dimensional (3-D) effects in the PBMR analysis. The performed studies demonstrated that explicit 3-D modeling of control rod movement is superior and removes the errors associated with the grey curtain (2-D homogenized) approximation.

Development of the new trigger for VANDLE neutron detector

NASA Astrophysics Data System (ADS)

Hasse, Adam; Taylor, Steven; Daugherty, Hadyn; Grzywacz, Robert

2014-09-01

Beta-delayed neutron emission (βn) is the dominant decay channel for the majority of very neutron-rich nuclei. In order to study these decays a new detector system called the Versatile Array of Neutron Detectors at Low Energy (VANDLE) was constructed. A critical part of this neutron time of flight detector is a trigger unit. This trigger is sensitive to electron from beta decay down to very low energies, insensitive to gamma rays and have a good timing performance, better than 1 ns. In order to satisfy these condition, we have developed a new system, which utilizes plastic scintillator but uses recently developed light readout technique, based on the so called Silicon Photomultiplier, manufactured by Sensl. New system has been developed and performance tested using digital data acquisition system at the University of Tennessee and will be utilized in future experiments involving VANDLE. Beta-delayed neutron emission (βn) is the dominant decay channel for the majority of very neutron-rich nuclei. In order to study these decays a new detector system called the Versatile Array of Neutron Detectors at Low Energy (VANDLE) was constructed. A critical part of this neutron time of flight detector is a trigger unit. This trigger is sensitive to electron from beta decay down to very low energies, insensitive to gamma rays and have a good timing performance, better than 1 ns. In order to satisfy these condition, we have developed a new system, which utilizes plastic scintillator but uses recently developed light readout technique, based on the so called Silicon Photomultiplier, manufactured by Sensl. New system has been developed and performance tested using digital data acquisition system at the University of Tennessee and will be utilized in future experiments involving VANDLE. Department of Physics and Astronomy, University of Tennessee, Knoxville, USA.

Cherenkov neutron detector for fusion reaction and runaway electron diagnostics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheon, MunSeong, E-mail: munseong@nfri.re.kr; Kim, Junghee

2015-08-15

A Cherenkov-type neutron detector was newly developed and neutron measurement experiments were performed at Korea Superconducting Tokamak Advanced Research. It was shown that the Cherenkov neutron detector can monitor the time-resolved neutron flux from deuterium-fueled fusion plasmas. Owing to the high temporal resolution of the detector, fast behaviors of runaway electrons, such as the neutron spikes, could be observed clearly. It is expected that the Cherenkov neutron detector could be utilized to provide useful information on runaway electrons as well as fusion reaction rate in fusion plasmas.

Perspectives for online analysis of raw material by pulsed neutron irradiation

NASA Astrophysics Data System (ADS)

Bach, Pierre; Le Tourneur, P.; Poumarede, B.

1997-02-01

On-line analysis by pulsed neutron irradiation is an example of an advanced technology application of nuclear techniques, concerning real problems in the cement, mineral and coal industries. The most significant of these nuclear techniques is their capability of continuous measurement without contact and without sampling, which can lead to improved control of processes and resultant large financial savings. Compared to Californium neutron sources, the use of electrical pulsed neutron generators allows to obtain a higher signal/noise ratio for a more sensitive measurement, and allows to overcome a number of safety problems concerning transportation, installation and maintenance. An experiment related to a possible new on-line raw material analyzer is described, using a pulsed neutron generator. The key factors contributing to an accurate measurement are related to a suitable generator, to a high count rate gamma ray spectroscopy electronics, and to computational tools. Calculation and results for the optimization of the neutron irradiation time diagram are reported. One of the operational characteristics of such an equipment is related to neutron flux available: it is possible to adjust it to the requested accuracy, i.e. for a high accuracy during a few hours/day and for a lower accuracy the rest of the time. This feature allows to operate the neutron tube during a longer time, and then to reduce the cost of analysis.

Control of morphology and function of low band gap polymer–bis-fullerene mixed heterojunctions in organic photovoltaics with selective solvent vapor annealing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, Huipeng; Hsiao, Yu-Che; Hu, Bin

2014-05-07

We reported how by replacing PCBM with a bis-adduct fullerene (i.e. ICBA) we significantly improve the open circuit voltage (VOC) and power conversion efficiency (PCE) in P3HT bulk heterojunctions. But, for the most promising low band-gap polymer (LBP) systems, replacing PCBM with ICBA results in very poor shortcircuit current (JSC) and PCE although the VOC is significantly improved. Therefore, in this work, we have completed small angle neutron scattering and neutron reflectometry experiments to study the impact of post-deposition solvent annealing (SA) with control of solvent quality on the morphology and performance of LBP bis-fullerene BHJ photovoltaics. Our results showmore » that SA in a solvent that is selective for the LBP results in a depletion of bis-fullerene near the air surface, which limits device performance. SA in a solvent vapor which has similar solubility for polymer and bis-fullerene results in a higher degree of polymer ordering, bis-fullerene phase separation, and segregation of the bis-fullerene to the air surface, which facilitates charge transport and increases power conversion efficiency (PCE) by 100%. The highest degree of polymer ordering combined with significant bis-fullerene phase separation and segregation of bis-fullerene to the air surface is obtained by SA in a solvent vapor that is selective for the bis-fullerene. The resultant morphology increases PCE by 190%. These results indicate that solvent annealing with judicious solvent choice provides a unique tool to tune the morphology of LBP bisfullerene BHJ system, providing sufficient polymer ordering, formation of a bis-fullerene pure phase, and segregation of bis-fullerene to the air surface to optimize the morphology of the active layer. Furthermore, this process is broadly applicable to improving current disappointing LBP bis-fullerene systems to optimize their morphology and OPV performance post-deposition, including higher VOC and power conversion efficiency.« less

Opportunities afforded by the intense nanosecond neutron pulses from a plasma focus source for neutron capture therapy and the preliminary simulation results

NASA Astrophysics Data System (ADS)

Giannini, G.; Gribkov, V.; Longo, F.; Ramos Aruca, M.; Tuniz, C.

2012-11-01

The use of short and powerful neutron pulses for boron neutron capture therapy (BNCT) can potentially increase selectivity and reduce the total dose absorbed by the patient. The biological effects of radiation depend on the dose, the dose power and the spatial distribution of the microscopic energy deposition. A dense plasma focus (DPF) device emits very short (in the nanosecond range) and extremely intense pulses of fast neutrons (2.5 or 14 MeV neutrons—from D-D or D-T nuclear reactions) and x-rays. Optimal spectra of neutrons formed for use in BNCT must contain an epithermal part to ensure a reasonable penetration depth into tissues at high enough cross-section on boron. So the powerful nanosecond pulses of fast neutrons generated by DPF must be moderated. After this moderation, the pulse duration must be shorter compared with the duration of the reaction with free radicals, that is, ⩾1 μs. In this work we focus on the development of a detailed simulation of interaction of short-pulse radiation from a DPF with the device's materials and with different types of moderators to estimate the dose power at the cells for this dynamic case. The simulation was carried out by means of the Geant4 toolkit in two main steps: the modeling of the pulsed neutron source device itself; the study of the interaction of fast mono-energetic neutrons with a moderator specific for BNCT.

Polarized neutron scattering on HYSPEC: the HYbrid SPECtrometer at SNS

DOE PAGES

Zaliznyak, Igor A.; Savici, Andrei T.; Ovidiu Garlea, V.; ...

2017-06-20

Here, we describe some of the first polarized neutron scattering measurements performed at HYSPEC [1-4] spectrometer at the Spallation Neutron Source, Oak Ridge National Laboratory. Furthermore, we discuss details of the instrument setup and the experimental procedures in the mode with full polarization analysis. Examples of polarized neutron diffraction and polarized inelastic neutron data obtained on single crystal samples are presented.

Thermal Neutron Die-Way-Time Studies for P and DGNAA of Radioactive Waste Drums at the MEDINA Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mildenberger, Frank; Mauerhofer, Eric

2015-07-01

In Germany, radioactive waste with negligible heat production has to pass through a process of quality checking in order to check its conformance with national regulations prior to its transport, intermediate storage and final disposal. Additionally to its radioactive components, the waste may contain non-radioactive chemically toxic substances that can adversely affect human health and pollute the environment, especially the ground water. After an adequate decay time, the waste radioactivity will become harmless but the non-radioactive substances will persist over time. In principle, these hazardous substances may be quantified from traceability and quality controls performed during the production of themore » waste packages. As a consequence, a research and development program was initiated in 2007 with the aim to develop a nondestructive analytical technique for radioactive waste packages based on prompt and delayed gamma neutron activation analysis (P and DGNAA) employing a DT-neutron generator in pulsed mode. In a preliminary study it was experimentally demonstrated that P and DGNAA is suitable to determine the chemical composition of large samples. In 2010 a facility called MEDINA (Multi Element Detection based on Instrumental Neutron Activation) was developed for the qualitative and quantitative determination of nonradioactive, toxic elements and substances in 200-l steel drums. The determination of hazardous substances and elements is generally achieved measuring the prompt gamma-rays induced by thermal neutrons. Additional information about the composition of the waste matrix could be derived measuring the delayed gamma-rays from short life activation products. However a sensitive detection of these delayed gamma-rays requires that thermal neutrons have almost vanished. Therefore, the thermal neutron die-away-time has to be known in order to achieve an optimal discrimination between prompt and delayed gamma-ray spectra acquisition. Measurements Thermal neutron die-away times have been determined for the following cases: a) the empty chamber, b ) an empty 200-l steel drum, for a 200-l steel drum filled c) with concrete d) with polyethylene and e) with a mixture of polyethylene and concrete by measuring the prompt-gamma ray count rate of relevant isotopes like of {sup 1}H, {sup 10}B, {sup 12}C, {sup 28}Si, {sup 35}Cl, {sup 40}Ca and {sup 56}Fe which are emitted from different parts of the facility and the sample. Additionally, the average die-away-time was determined from the total detector count rate. The neutron generator was operated with a neutron emission of 8x10{sup 7} n.s{sup -1}, a neutron pulse with a length of 250 μs and a repetition time of 5 ms. The spectra were acquired between the neutron pulses over t{sub c}=500 μs after a pre-defined waiting time t{sub D} (multiple of 500 μs). The thermal neutron die-away time was ranging between 0.9 ms and 5 ms according to the sample composition. As an example the measured thermal neutron die-away-time Λ [μs] of a drum filled with concrete is presented. Detailed results of this study will be presented and discussed. (authors)« less

Slotted rotatable target assembly and systematic error analysis for a search for long range spin dependent interactions from exotic vector boson exchange using neutron spin rotation

NASA Astrophysics Data System (ADS)

Haddock, C.; Crawford, B.; Fox, W.; Francis, I.; Holley, A.; Magers, S.; Sarsour, M.; Snow, W. M.; Vanderwerp, J.

2018-03-01

We discuss the design and construction of a novel target array of nonmagnetic test masses used in a neutron polarimetry measurement made in search for new possible exotic spin dependent neutron-atominteractions of Nature at sub-mm length scales. This target was designed to accept and efficiently transmit a transversely polarized slow neutron beam through a series of long open parallel slots bounded by flat rectangular plates. These openings possessed equal atom density gradients normal to the slots from the flat test masses with dimensions optimized to achieve maximum sensitivity to an exotic spin-dependent interaction from vector boson exchanges with ranges in the mm - μm regime. The parallel slots were oriented differently in four quadrants that can be rotated about the neutron beam axis in discrete 90°increments using a Geneva drive. The spin rotation signals from the 4 quadrants were measured using a segmented neutron ion chamber to suppress possible systematic errors from stray magnetic fields in the target region. We discuss the per-neutron sensitivity of the target to the exotic interaction, the design constraints, the potential sources of systematic errors which could be present in this design, and our estimate of the achievable sensitivity using this method.

Experimental characterization of HOTNES: A new thermal neutron facility with large homogeneity area

NASA Astrophysics Data System (ADS)

Bedogni, R.; Sperduti, A.; Pietropaolo, A.; Pillon, M.; Pola, A.; Gómez-Ros, J. M.

2017-01-01

A new thermal neutron irradiation facility, called HOTNES (HOmogeneous Thermal NEutron Source), was established in the framework of a collaboration between INFN-LNF and ENEA-Frascati. HOTNES is a polyethylene assembly, with about 70 cmx70 cm square section and 100 cm height, including a large, cylindrical cavity with diameter 30 cm and height 70 cm. The facility is supplied by a 241Am-B source located at the bottom of this cavity. The facility was designed in such a way that the iso-thermal-fluence surfaces, characterizing the irradiation volume, coincide with planes parallel to the cavity bottom. The thermal fluence rate across a given isofluence plane is as uniform as 1% on a disk with 30 cm diameter. Thermal fluence rate values from about 700 cm-2 s-1 to 1000 cm-2 s-1 can be achieved. The facility design, previously optimized by Monte Carlo simulation, was experimentally verified. The following techniques were used: gold activation foils to assess the thermal fluence rate, semiconductor-based active detector for mapping the irradiation volume, and Bonner Sphere Spectrometer to determine the complete neutron spectrum. HOTNES is expected to be attractive for the scientific community involved in neutron metrology, neutron dosimetry and neutron detector testing.

Geant4 beam model for boron neutron capture therapy: investigation of neutron dose components.

PubMed

Moghaddasi, Leyla; Bezak, Eva

2018-03-01

Boron neutron capture therapy (BNCT) is a biochemically-targeted type of radiotherapy, selectively delivering localized dose to tumour cells diffused in normal tissue, while minimizing normal tissue toxicity. BNCT is based on thermal neutron capture by stable [Formula: see text]B nuclei resulting in emission of short-ranged alpha particles and recoil [Formula: see text]Li nuclei. The purpose of the current work was to develop and validate a Monte Carlo BNCT beam model and to investigate contribution of individual dose components resulting of neutron interactions. A neutron beam model was developed in Geant4 and validated against published data. The neutron beam spectrum, obtained from literature for a cyclotron-produced beam, was irradiated to a water phantom with boron concentrations of 100 μg/g. The calculated percentage depth dose curves (PDDs) in the phantom were compared with published data to validate the beam model in terms of total and boron depth dose deposition. Subsequently, two sensitivity studies were conducted to quantify the impact of: (1) neutron beam spectrum, and (2) various boron concentrations on the boron dose component. Good agreement was achieved between the calculated and measured neutron beam PDDs (within 1%). The resulting boron depth dose deposition was also in agreement with measured data. The sensitivity study of several boron concentrations showed that the calculated boron dose gradually converged beyond 100 μg/g boron concentration. This results suggest that 100μg/g tumour boron concentration may be optimal and above this value limited increase in boron dose is expected for a given neutron flux.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Singh, Arvind, E-mail: anita@barc.gov.in; Topkar, Anita

In order to improve the gamma discrimination capability for thermal neutron measurements using silicon PIN detectors, a novel approach of use of thin epitaxial silicon PIN detectors was investigated. Thin epitaxial silicon detectors with thickness of 15 µm were developed and their performance was tested with thermal neutrons using {sup 10}B converter. The performance of this detector was compared with the performance of a 300 µm silicon detector. The results of experiments presented in this paper indicate that thin epitaxial silicon detectors can significantly improve γ discrimination for thermal neutron measurements.

Neutron and gamma irradiation effects on power semiconductor switches

NASA Technical Reports Server (NTRS)

Schwarze, G. E.; Frasca, A. J.

1990-01-01

The performance characteristics of high-power semiconductor switches subjected to high levels of neutron fluence and gamma dose must be known by the designer of the power conditioning, control and transmission subsystem of space nuclear power systems. Location and the allowable shielding mass budget will determine the level of radiation tolerance required by the switches to meet performance and reliability requirements. Neutron and gamma ray interactions with semiconductor materials and how these interactions affect the electrical and switching characteristics of solid state power switches is discussed. The experimental measurement system and radiation facilities are described. Experimental data showing the effects of neutron and gamma irradiation on the performance characteristics are given for power-type NPN Bipolar Junction Transistors (BJTs), and Metal-Oxide-Semiconductor Field Effect Transistors (MOSFETs). BJTs show a rapid decrease in gain, blocking voltage, and storage time for neutron irradiation, and MOSFETs show a rapid decrease in the gate threshold voltage for gamma irradiation.

Neutron and gamma irradiation effects on power semiconductor switches

NASA Technical Reports Server (NTRS)

Schwarze, G. E.; Frasca, A. J.

1990-01-01

The performance characteristics of high power semiconductor switches subjected to high levels of neutron fluence and gamma dose must be known by the designer of the power conditioning, control and transmission subsystem of space nuclear power systems. Location and the allowable shielding mass budget will determine the level of radiation tolerance required by the switches to meet performance and reliability requirements. Neutron and gamma ray interactions with semiconductor materials and how these interactions affect the electrical and switching characteristics of solid state power switches is discussed. The experimental measurement system and radiation facilities are described. Experimental data showing the effects of neutron and gamma irradiation on the performance characteristics are given for power-type NPN Bipolar Junction Transistors (BJTs), and Metal-Oxide-Semiconductor Field Effect Transistors (MOSFETs). BJTs show a rapid decrease in gain, blocking voltage, and storage time for neutron irradiation, and MOSFETs show a rapid decrease in the gate threshold voltage for gamma irradiation.

Thermal neutron filter design for the neutron radiography facility at the LVR-15 reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soltes, Jaroslav; Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague,; Viererbl, Ladislav

2015-07-01

In 2011 a decision was made to build a neutron radiography facility at one of the unused horizontal channels of the LVR-15 research reactor in Rez, Czech Republic. One of the key conditions for operating an effective radiography facility is the delivery of a high intensity, homogeneous and collimated thermal neutron beam at the sample location. Additionally the intensity of fast neutrons has to be kept as low as possible as the fast neutrons may damage the detectors used for neutron imaging. As the spectrum in the empty horizontal channel roughly copies the spectrum in the reactor core, which hasmore » a high ratio of fast neutrons, neutron filter components have to be installed inside the channel in order to achieve desired beam parameters. As the channel design does not allow the instalment of complex filters and collimators, an optimal solution represent neutron filters made of large single-crystal ingots of proper material composition. Single-crystal silicon was chosen as a favorable filter material for its wide availability in sufficient dimensions. Besides its ability to reasonably lower the ratio of fast neutrons while still keeping high intensities of thermal neutrons, due to its large dimensions, it suits as a shielding against gamma radiation from the reactor core. For designing the necessary filter dimensions the Monte-Carlo MCNP transport code was used. As the code does not provide neutron cross-section libraries for thermal neutron transport through single-crystalline silicon, these had to be created by approximating the theory of thermal neutron scattering and modifying the original cross-section data which are provided with the code. Carrying out a series of calculations the filter thickness of 1 m proved good for gaining a beam with desired parameters and a low gamma background. After mounting the filter inside the channel several measurements of the neutron field were realized at the beam exit. The results have justified the expected calculated values. After the successful filter installing and a series of measurements, first test neutron radiography attempts with test samples could been carried out. (authors)« less

Neutron detection by scintillation of noble-gas excimers

NASA Astrophysics Data System (ADS)

McComb, Jacob Collin

Neutron detection is a technique essential to homeland security, nuclear reactor instrumentation, neutron diffraction science, oil-well logging, particle physics and radiation safety. The current shortage of helium-3, the neutron absorber used in most gas-filled proportional counters, has created a strong incentive to develop alternate methods of neutron detection. Excimer-based neutron detection (END) provides an alternative with many attractive properties. Like proportional counters, END relies on the conversion of a neutron into energetic charged particles, through an exothermic capture reaction with a neutron absorbing nucleus (10B, 6Li, 3He). As charged particles from these reactions lose energy in a surrounding gas, they cause electron excitation and ionization. Whereas most gas-filled detectors collect ionized charge to form a signal, END depends on the formation of diatomic noble-gas excimers (Ar*2, Kr*2,Xe* 2) . Upon decaying, excimers emit far-ultraviolet (FUV) photons, which may be collected by a photomultiplier tube or other photon detector. This phenomenon provides a means of neutron detection with a number of advantages over traditional methods. This thesis investigates excimer scintillation yield from the heavy noble gases following the boron-neutron capture reaction in 10B thin-film targets. Additionally, the thesis examines noble-gas excimer lifetimes with relationship to gas type and gas pressure. Experimental data were collected both at the National Institute of Standards and Technology (NIST) Center for Neutron Research, and on a newly developed neutron beamline at the Maryland University Training Reactor. The components of the experiment were calibrated at NIST and the University of Maryland, using FUV synchrotron radiation, neutron imaging, and foil activation techniques, among others. Computer modeling was employed to simulate charged-particle transport and excimer photon emission within the experimental apparatus. The observed excimer scintillation yields from the 10B( n, alpha)7Li reaction are comparable to the yields of many liquid and solid neutron scintillators. Additionally, the observed slow triplet-state decay of neutron-capture-induced excimers may be used in a practical detector to discriminate neutron interactions from gamma-ray interactions. The results of these measurements and simulations will contribute to the development and optimization of a deployable neutron detector based on noble-gas excimer scintillation.

Tritium target manufacturing for use in accelerators

NASA Astrophysics Data System (ADS)

Bach, P.; Monnin, C.; Van Rompay, M.; Ballanger, A.

2001-07-01

As a neutron tube manufacturer, SODERN is now in charge of manufacturing tritium targets for accelerators, in cooperation with CEA/DAM/DTMN in Valduc. Specific deuterium and tritium targets are manufactured on request, according to the requirements of the users, starting from titanium target on copper substrate, and going to more sophisticated devices. A wide range of possible uses is covered, including thin targets for neutron calibration, thick targets with controlled loading of deuterium and tritium, rotating targets for higher lifetimes, or large size rotating targets for accelerators used in boron neutron therapy. Activity of targets lies in the 1 to 1000 Curie, diameter of targets being up to 30 cm. Special targets are also considered, including surface layer targets for lowering tritium desorption under irradiation, or those made from different kinds of occluders such as titanium, zirconium, erbium, scandium, with different substrates. It is then possible to optimize either neutron output, or lifetime and stability, or thermal behavior.

The Los ALamos Neutron Science Center Hydrogen Moderator System

NASA Astrophysics Data System (ADS)

Jarmer, J. J.; Knudson, J. N.

2006-04-01

At the Los Alamos Neutron Science Center (LANSCE), spallation neutrons are produced by an 800-MeV proton beam interacting with tungsten targets. Gun-barrel-type penetrations through the heavy concrete and steel shielding that surround the targets collimate neutrons to form neutron beams used for scattering experiments. Two liquid hydrogen moderators of one-liter volume each are positioned adjacent to the neutron-production targets. Some of the neutrons that pass through a moderator interact with or scatter from protons in the hydrogen. The neutron-proton interaction reduces the energy or moderates neutrons to lower energies. Lower energy "moderated" neutrons are the most useful for some neutron scattering experiments. We provide a description of the LANSCE hydrogen-moderator system and its cryogenic performance with proton beams of up to 125 micro-amp average current.

Multiple source associated particle imaging for simultaneous capture of multiple projections

DOEpatents

Bingham, Philip R; Hausladen, Paul A; McConchi, Seth M; Mihalczo, John T; Mullens, James A

2013-11-19

Disclosed herein are representative embodiments of methods, apparatus, and systems for performing neutron radiography. For example, in one exemplary method, an object is interrogated with a plurality of neutrons. The plurality of neutrons includes a first portion of neutrons generated from a first neutron source and a second portion of neutrons generated from a second neutron source. Further, at least some of the first portion and the second portion are generated during a same time period. In the exemplary method, one or more neutrons from the first portion and one or more neutrons from the second portion are detected, and an image of the object is generated based at least in part on the detected neutrons from the first portion and the detected neutrons from the second portion.

Gadolinium for neutron detection in current nuclear instrumentation research: A review

NASA Astrophysics Data System (ADS)

Dumazert, J.; Coulon, R.; Lecomte, Q.; Bertrand, G. H. V.; Hamel, M.

2018-02-01

Natural gadolinium displays a number of remarkable physical properties: it is a rare earth element, composed of seven stable or quasi-stable isotopes, with an exceptionally high magnetization and a Curie point near room temperature. Its use in the field of nuclear instrumentation historically relates to its efficiency as a neutron poison in power reactors. Gadolinium is indeed the naturally occurring element with the highest interaction probability with neutrons at thermal energy, shared between Gd-157 (15.65%, 254000 b cross section) and Gd-155 (14.8%, 60900 b) isotopes. Considering that neutron capture results in an isotopic change, followed by a radiative rearrangement of nuclear and atomic structures, Gd may be embodied not merely as a neutron poison but as a neutron converter into a prompt photon and an electron source term. Depending on the nature and energy of the reaction products (from a few-keV Auger electrons up to 8 MeV gamma rays) that the detector aims at isolating as an indirect neutron signature, a variety of sensor media and counting methods have been introduced during the last decades. This review first draws a theoretical description of the radiative cascade following Gd(n , γ) capture. The cascade may be subdivided into regions of interest, each corresponding to dedicated detection designs and optimizations whose current status is detailed. This inventory has allowed the authors to extract and benchmark key figures of merit for the definition of a detection scheme: neutron attenuation, neutron sensitivity (cps/nv), gamma rejection, neutron detection limit in a mixed field, intrinsic or extrinsic moderation, and transportability. On this basis, the authors have identified promising paths for Gd-based neutron detection in contemporary instrumentation.

Drug and tobacco detection using neutron transmission/attenuation

NASA Astrophysics Data System (ADS)

Miller, Thomas G.

1994-10-01

A neutron transmission/attenuation spectrometer has been used to obtain the neutron attenuation signature of cocaine, heroin, hashish, methamphetamine, pipe tobacco and chewing tobacco. A pulsed `white neutron' source was created by bombarding a thick beryllium target with a 5 MeV pulsed deuteron beam. The neutron intensity was measured from about 0.75 MeV to about 4 MeV with the suitcase in and out of the neutron beam to determine the neutron attenuation. Experiments were performed for drugs and tobacco alone and when imbedded in an `average suitcase'. The experimentally determined neutron attenuation curves were used to determine the atomic ratios C/O, N/O, and H/C through the samples using measured neutron cross sections.

High efficiency proportional neutron detector with solid liner internal structures

DOEpatents

Kisner, Roger Allen; Holcomb, David Eugene; Brown, Gilbert M.

2014-08-05

A tube-style neutron detector, a panel-style neutron detector incorporating a plurality of tube-style neutron detectors, and a panel-style neutron detector including a plurality of anode wires are provided. A plurality of channels is provided in a neutron detector such that each channel has an inner surface of a coating layer including a neutron-absorbing material. A wire anode is provided at end of each channel so that electrons generated by a charged daughter particle generated by a neutron are collected to detect a neutron-matter interaction. Moderator units can be incorporated into a neutron detector to provide improved detection efficiencies and/or to determine neutron energy spectrum. Gas-based proportional response from the neutron detectors can be employed for special nuclear material (SNM) detection. This neutron detector can provide similar performance to .sup.3He-based detectors without requiring .sup.3He and without containing toxic, flammable, or high-pressure materials.

Advanced Neutronics Tools for BWR Design Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Santamarina, A.; Hfaiedh, N.; Letellier, R.

2006-07-01

This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method Of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007more » BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy. (authors)« less

The National Spallation Neutron Source Target Station.

NASA Astrophysics Data System (ADS)

Gabriel, T. A.

1997-05-01

The technologies that are being utilized to design and build a state-of-the-art high powered (>= 1 MW), short pulsed (<= 1 μsec), and reliable spallation neutron source target station are discussed. The protons which directly and indirectly produce the neutrons will be obtained from a 1 GeV proton accelerator composed of an ion gun, rfq, linac, and storage ring. Many scientific and technical disciplines are required to produce a successful target station. These disciplines include engineering, remote handling, neutronics, materials, thermal hydraulics, shock analysis, etc. In the areas of engineering and remote handling special emphasis is being given to rapid and efficient assembly and disassembly of critical parts of the target station. In the neutronics area, emphasis is being given to neutron yield and pulse optimization from the moderators, and heating and activation rates throughout the station. Development of structural materials to withstand aggressive radiation environments and that are compatible with other materials is also an important area. Thermal hydraulics and shock analysis are being closely studied since large amounts of energy are being deposited in small volumes in relatively short time periods (< 1 μsec). These areas will be expanded upon in the paper.

Cross correlation calculations and neutron scattering analysis for a portable solid state neutron detection system

NASA Astrophysics Data System (ADS)

Saltos, Andrea

In efforts to perform accurate dosimetry, Oakes et al. [Nucl. Intrum. Mehods. (2013)] introduced a new portable solid state neutron rem meter based on an adaptation of the Bonner sphere and the position sensitive long counter. The system utilizes high thermal efficiency neutron detectors to generate a linear combination of measurement signals that are used to estimate the incident neutron spectra. The inversion problem associated to deduce dose from the counts in individual detector elements is addressed by applying a cross-correlation method which allows estimation of dose with average errors less than 15%. In this work, an evaluation of the performance of this system was extended to take into account new correlation techniques and neutron scattering contribution. To test the effectiveness of correlations, the Distance correlation, Pearson Product-Moment correlation, and their weighted versions were performed between measured spatial detector responses obtained from nine different test spectra, and the spatial response of Library functions generated by MCNPX. Results indicate that there is no advantage of using the Distance Correlation over the Pearson Correlation, and that weighted versions of these correlations do not increase their performance in evaluating dose. Both correlations were proven to work well even at low integrated doses measured for short periods of time. To evaluate the contribution produced by room-return neutrons on the dosimeter response, MCNPX was used to simulate dosimeter responses for five isotropic neutron sources placed inside different sizes of rectangular concrete rooms. Results show that the contribution of scattered neutrons to the response of the dosimeter can be significant, so that for most cases the dose is over predicted with errors as large as 500%. A possible method to correct for the contribution of room-return neutrons is also assessed and can be used as a good initial estimate on how to approach the problem.

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

NASA Astrophysics Data System (ADS)

Koshelev, A. S.; Kovshov, K. N.; Ovchinnikov, M. A.; Pikulina, G. N.; Sokolov, A. B.

2016-12-01

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Kovshov, K. N.; Ovchinnikov, M. A.

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

Amorphous Silicon Based Neutron Detector

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, Liwei

2004-12-12

Various large-scale neutron sources already build or to be constructed, are important for materials research and life science research. For all these neutron sources, neutron detectors are very important aspect. However, there is a lack of a high-performance and low-cost neutron beam monitor that provides time and temporal resolution. The objective of this SBIR Phase I research, collaboratively performed by Midwest Optoelectronics, LLC (MWOE), the University of Toledo (UT) and Oak Ridge National Laboratory (ORNL), is to demonstrate the feasibility for amorphous silicon based neutron beam monitors that are pixilated, reliable, durable, fully packaged, and fabricated with high yield usingmore » low-cost method. During the Phase I effort, work as been focused in the following areas: 1) Deposition of high quality, low-defect-density, low-stress a-Si films using very high frequency plasma enhanced chemical vapor deposition (VHF PECVD) at high deposition rate and with low device shunting; 2) Fabrication of Si/SiO2/metal/p/i/n/metal/n/i/p/metal/SiO2/ device for the detection of alpha particles which are daughter particles of neutrons through appropriate nuclear reactions; and 3) Testing of various devices fabricated for alpha and neutron detection; As the main results: · High quality, low-defect-density, low-stress a-Si films have been successfully deposited using VHF PECVD on various low-cost substrates; · Various single-junction and double junction detector devices have been fabricated; · The detector devices fabricated have been systematically tested and analyzed. · Some of the fabricated devices are found to successfully detect alpha particles. Further research is required to bring this Phase I work beyond the feasibility demonstration toward the final prototype devices. The success of this project will lead to a high-performance, low-cost, X-Y pixilated neutron beam monitor that could be used in all of the neutron facilities worldwide. In addition, the technologies developed here could be used to develop X-ray and neutron monitors that could be used in the future for security checks at the airports and other critical facilities. The project would lead to devices that could significantly enhance the performance of multi-billion dollar neutron source facilities in the US and bring our nation to the forefront of neutron beam sciences and technologies which have enormous impact to materials, life science and military research and applications.« less

Application of an ultraminiature thermal neutron monitor for irradiation field study of accelerator-based neutron capture therapy

PubMed Central

Ishikawa, Masayori; Tanaka, Kenichi; Endo, Satrou; Hoshi, Masaharu

2015-01-01

Abstract Phantom experiments to evaluate thermal neutron flux distribution were performed using the Scintillator with Optical Fiber (SOF) detector, which was developed as a thermal neutron monitor during boron neutron capture therapy (BNCT) irradiation. Compared with the gold wire activation method and Monte Carlo N-particle (MCNP) calculations, it was confirmed that the SOF detector is capable of measuring thermal neutron flux as low as 105 n/cm2/s with sufficient accuracy. The SOF detector will be useful for phantom experiments with BNCT neutron fields from low-current accelerator-based neutron sources. PMID:25589504

Neutron reflecting supermirror structure

DOEpatents

Wood, James L.

1992-01-01

An improved neutron reflecting supermirror structure comprising a plurality of stacked sets of bilayers of neutron reflecting materials. The improved neutron reflecting supermirror structure is adapted to provide extremely good performance at high incidence angles, i.e. up to four time the critical angle of standard neutron mirror structures. The reflection of neutrons striking the supermirror structure at a high critical angle provides enhanced neutron throughput, and hence more efficient and economical use of neutron sources. One layer of each set of bilayers consist of titanium, and the second layer of each set of bilayers consist of an alloy of nickel with carbon interstitially present in the nickel alloy.

Conceptual design of the early implementation of the NEutron Detector Array (NEDA) with AGATA

NASA Astrophysics Data System (ADS)

Hüyük, Tayfun; Di Nitto, Antonio; Jaworski, Grzegorz; Gadea, Andrés; Javier Valiente-Dobón, José; Nyberg, Johan; Palacz, Marcin; Söderström, Pär-Anders; Jose Aliaga-Varea, Ramon; de Angelis, Giacomo; Ataç, Ayşe; Collado, Javier; Domingo-Pardo, Cesar; Egea, Francisco Javier; Erduran, Nizamettin; Ertürk, Sefa; de France, Gilles; Gadea, Rafael; González, Vicente; Herrero-Bosch, Vicente; Kaşkaş, Ayşe; Modamio, Victor; Moszynski, Marek; Sanchis, Enrique; Triossi, Andrea; Wadsworth, Robert

2016-03-01

The NEutron Detector Array (NEDA) project aims at the construction of a new high-efficiency compact neutron detector array to be coupled with large γ-ray arrays such as AGATA. The application of NEDA ranges from its use as selective neutron multiplicity filter for fusion-evaporation reaction to a large solid angle neutron tagging device. In the present work, possible configurations for the NEDA coupled with the Neutron Wall for the early implementation with AGATA has been simulated, using Monte Carlo techniques, in order to evaluate their performance figures. The goal of this early NEDA implementation is to improve, with respect to previous instruments, efficiency and capability to select multiplicity for fusion-evaporation reaction channels in which 1, 2 or 3 neutrons are emitted. Each NEDA detector unit has the shape of a regular hexagonal prism with a volume of about 3.23l and it is filled with the EJ301 liquid scintillator, that presents good neutron- γ discrimination properties. The simulations have been performed using a fusion-evaporation event generator that has been validated with a set of experimental data obtained in the 58Ni + 56Fe reaction measured with the Neutron Wall detector array.

Measurement and simulation of thermal neutron flux distribution in the RTP core

NASA Astrophysics Data System (ADS)

Rabir, Mohamad Hairie B.; Jalal Bayar, Abi Muttaqin B.; Hamzah, Na'im Syauqi B.; Mustafa, Muhammad Khairul Ariff B.; Karim, Julia Bt. Abdul; Zin, Muhammad Rawi B. Mohamed; Ismail, Yahya B.; Hussain, Mohd Huzair B.; Mat Husin, Mat Zin B.; Dan, Roslan B. Md; Ismail, Ahmad Razali B.; Husain, Nurfazila Bt.; Jalil Khan, Zareen Khan B. Abdul; Yakin, Shaiful Rizaide B. Mohd; Saad, Mohamad Fauzi B.; Masood, Zarina Bt.

2018-01-01

The in-core thermal neutron flux distribution was determined using measurement and simulation methods for the Malaysian’s PUSPATI TRIGA Reactor (RTP). In this work, online thermal neutron flux measurement using Self Powered Neutron Detector (SPND) has been performed to verify and validate the computational methods for neutron flux calculation in RTP calculations. The experimental results were used as a validation to the calculations performed with Monte Carlo code MCNP. The detail in-core neutron flux distributions were estimated using MCNP mesh tally method. The neutron flux mapping obtained revealed the heterogeneous configuration of the core. Based on the measurement and simulation, the thermal flux profile peaked at the centre of the core and gradually decreased towards the outer side of the core. The results show a good agreement (relatively) between calculation and measurement where both show the same radial thermal flux profile inside the core: MCNP model over estimation with maximum discrepancy around 20% higher compared to SPND measurement. As our model also predicts well the neutron flux distribution in the core it can be used for the characterization of the full core, that is neutron flux and spectra calculation, dose rate calculations, reaction rate calculations, etc.

Neutron capture reactions at DANCE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bredeweg, T. A.

2008-05-12

The Detector for Advanced Neutron Capture Experiments (DANCE) is a 4{pi} BaF{sub 2} array consisting of 160 active detector elements. The primary purpose of the array is to perform neutron capture cross section measurements on small (> or approx.100 {mu}g) and/or radioactive (< or approx. 100 mCi) species. The measurements made possible with this array will be useful in answering outstanding questions in the areas of national security, threat reduction, nuclear astrophysics, advanced reactor design and accelerator transmutation of waste. Since the commissioning of DANCE we have performed neutron capture cross section measurements on a wide array of medium tomore » heavy mass nuclides. Measurements to date include neutron capture cross sections on {sup 241,243}Am, neutron capture and neutron-induced fission cross sections and capture-to-fission ratio ({alpha} = {sigma}{sub {gamma}}/{sigma}{sub f}) for {sup 235}U using a new fission-tagging detector as well as neutron capture cross sections for several astrophysics branch-point nuclei. Results from several of these measurements will be presented along with a discussion of additional physics information that can be extracted from the DANCE data.« less

The EURITRACK project: development of a tagged neutron inspection system for cargo containers

NASA Astrophysics Data System (ADS)

Perot, Bertrand; Perret, Gregory; Mariani, Alain; Ma, Jean-Luc; Szabo, Jean-Louis; Mercier, Emmanuel; Sannie, Guillaume; Viesti, Giuseppe; Nebbia, Giancarlo; Pesente, Silvia; Lunardon, Marcello; Formisano, Paola; Moretto, Sandra; Fabris, Daniela; Zenoni, Aldo; Bonomi, Germano; Donzella, Antonietta; Fontana, Andrea; Boghen, Gaia; Valkovic, Vladivoj; Sudac, Darovin; Moszynski, Marek; Batsch, Tadeusz; Gierlik, Michal; Wolski, Dariusz; Klamra, Wlodzimierz; Isaksson, Patrick; Le Tourneur, Philippe; Lhuissier, Miguel; Colonna, Annamaria; Tintori, Carlo; Peerani, Paolo; Sequeira, Vitor; Salvato, Martino

2006-05-01

The EURopean Illicit TRAfficing Countermeasures Kit project is part of the 6th European Union Framework Program, and aims at developing a neutron inspection system for detecting threat materials (explosives, drugs, etc.) in cargo containers. Neutron interaction in the container produces specific gamma-rays used to determine the chemical composition of the inspected material. An associated particle sealed tube neutron generator is developed to allow precise location of the interaction point by direction and time-of-flight measurements of the neutrons tagged by alpha-particles. The EURITRACK project consists in developing: a transportable deuterium-tritium neutron generator including a position sensitive alpha detector (8×8 matrix of YAP:Ce crystals coupled to a multi-anode photomultiplier), fast neutron and gamma-ray detectors, front-end electronics to perform coincidence and spectroscopic measurements, and an integrated software which manages neutron generator and detectors positioning, data acquisition and analysis. Hardware components have been developed and tested by the consortium partners. Current status of this work and provisional performances of the system assessed by Monte Carlo calculations are presented.

Note: Fast neutron efficiency in CR-39 nuclear track detectors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cavallaro, S.

2015-03-15

CR-39 samples are commonly employed for fast neutron detection in fusion reactors and in inertial confinement fusion experiments. The literature reported efficiencies are strongly depending on experimental conditions and, in some cases, highly dispersed. The present note analyses the dependence of efficiency as a function of various parameters and experimental conditions in both the radiator-assisted and the stand-alone CR-39 configurations. Comparisons of literature experimental data with Monte Carlo calculations and optimized efficiency values are shown and discussed.

Boron neutron capture therapy for oral precancer: proof of principle in an experimental animal model

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. Monti Hughes; ECC Pozzi; S. Thorp

Field-cancerized tissue can give rise to second primary tumours, causing therapeutic failure. Boron neutron capture therapy (BNCT) is based on biological targeting and would serve to treat undetectable foci of malignant transformation. The aim of this study was to optimize BNCT for the integral treatment for oral cancer, with particular emphasis on the inhibitory effect on tumour development originating in precancerous conditions, and radiotoxicity of different BNCT protocols in a hamster cheek pouch oral precancer model.

A neutron activation spectrometer and neutronic experimental platform for the National Ignition Facility (invited)

NASA Astrophysics Data System (ADS)

Yeamans, C. B.; Gharibyan, N.

2016-11-01

At the National Ignition Facility, the diagnostic instrument manipulator-based neutron activation spectrometer is used as a diagnostic of implosion performance for inertial confinement fusion experiments. Additionally, it serves as a platform for independent neutronic experiments and may be connected to fast recording systems for neutron effect tests on active electronics. As an implosion diagnostic, the neutron activation spectrometers are used to quantify fluence of primary DT neutrons, downscattered neutrons, and neutrons above the primary DT neutron energy created by reactions of upscattered D and T in flight. At a primary neutron yield of 1015 and a downscattered fraction of neutrons in the 10-12 MeV energy range of 0.04, the downscattered neutron fraction can be measured to a relative uncertainty of 8%. Significant asymmetries in downscattered neutrons have been observed. Spectrometers have been designed and fielded to measure the tritium-tritium and deuterium-tritium neutron outputs simultaneously in experiments using DT/TT fusion ratio as a direct measure of mix of ablator into the gas.

A neutron activation spectrometer and neutronic experimental platform for the National Ignition Facility (invited).

PubMed

Yeamans, C B; Gharibyan, N

2016-11-01

At the National Ignition Facility, the diagnostic instrument manipulator-based neutron activation spectrometer is used as a diagnostic of implosion performance for inertial confinement fusion experiments. Additionally, it serves as a platform for independent neutronic experiments and may be connected to fast recording systems for neutron effect tests on active electronics. As an implosion diagnostic, the neutron activation spectrometers are used to quantify fluence of primary DT neutrons, downscattered neutrons, and neutrons above the primary DT neutron energy created by reactions of upscattered D and T in flight. At a primary neutron yield of 10 15 and a downscattered fraction of neutrons in the 10-12 MeV energy range of 0.04, the downscattered neutron fraction can be measured to a relative uncertainty of 8%. Significant asymmetries in downscattered neutrons have been observed. Spectrometers have been designed and fielded to measure the tritium-tritium and deuterium-tritium neutron outputs simultaneously in experiments using DT/TT fusion ratio as a direct measure of mix of ablator into the gas.

Re-appraisal and extension of the Gratton-Vargas two-dimensional analytical snowplow model of plasma focus. III. Scaling theory for high pressure operation and its implications

NASA Astrophysics Data System (ADS)

Auluck, S. K. H.

2016-12-01

Recent work on the revised Gratton-Vargas model (Auluck, Phys. Plasmas 20, 112501 (2013); 22, 112509 (2015) and references therein) has demonstrated that there are some aspects of Dense Plasma Focus (DPF), which are not sensitive to details of plasma dynamics and are well captured in an oversimplified model assumption, which contains very little plasma physics. A hyperbolic conservation law formulation of DPF physics reveals the existence of a velocity threshold related to specific energy of dissociation and ionization, above which, the work done during shock propagation is adequate to ensure dissociation and ionization of the gas being ingested. These developments are utilized to formulate an algorithmic definition of DPF optimization that is valid in a wide range of applications, not limited to neutron emission. This involves determination of a set of DPF parameters, without performing iterative model calculations, that lead to transfer of all the energy from the capacitor bank to the plasma at the time of current derivative singularity and conversion of a preset fraction of this energy into magnetic energy, while ensuring that electromagnetic work done during propagation of the plasma remains adequate for dissociation and ionization of neutral gas being ingested. Such a universal optimization criterion is expected to facilitate progress in new areas of DPF research that include production of short lived radioisotopes of possible use in medical diagnostics, generation of fusion energy from aneutronic fuels, and applications in nanotechnology, radiation biology, and materials science. These phenomena are expected to be optimized for fill gases of different kinds and in different ranges of mass density compared to the devices constructed for neutron production using empirical thumb rules. A universal scaling theory of DPF design optimization is proposed and illustrated for designing devices working at one or two orders higher pressure of deuterium than the current practice of designs optimized at pressures less than 10 mbar of deuterium. These examples show that the upper limit for operating pressure is of technological (and not physical) origin.

14 MeV Neutron Irradiation Effect on Superconducting Magnet Materials for Fusion Device

NASA Astrophysics Data System (ADS)

Nishimura, A.; Hishinuma, Y.; Seo, K.; Tanaka, T.; Muroga, T.; Nishijima, S.; Katagiri, K.; Takeuchi, T.; Shindo, Y.; Ochiai, K.; Nishitani, T.; Okuno, K.

2006-03-01

As a large-scale plasma experimental device is planned and designed, the importance of investigations on irradiation effect of 14 MeV neutron increases and an experimental database is desired to be piled up. Recently, intense streaming of fast neutron from ports are reported and degradation of superconducting magnet performance is anticipated. To investigate the pure neutron effect on superconducting magnet materials, a cryogenic target system was newly developed and installed at Fusion Neutronics Source in Japan Atomic Energy Research Institute. Although production rate of 14 MeV neutron is not large, only 14 MeV neutron can be supplied to irradiation test without gamma ray. Copper wires, superconducting wires, glass fiber reinforced composites are irradiated and the irradiation effects are characterized. At the same time, sensors for measuring temperature and magnetic field are irradiated and their performance was investigated after irradiation. This paper presents outline of the cryogenic target system and some irradiation test results.

Neutron resonance spin echo with longitudinal DC fields

NASA Astrophysics Data System (ADS)

Krautloher, Maximilian; Kindervater, Jonas; Keller, Thomas; Häußler, Wolfgang

2016-12-01

We report on the design, construction, and performance of a neutron resonance spin echo (NRSE) instrument employing radio frequency (RF) spin flippers combining RF fields with DC fields, the latter oriented parallel (longitudinal) to the neutron propagation direction (longitudinal NRSE (LNRSE)). The advantage of the longitudinal configuration is the inherent homogeneity of the effective magnetic path integrals. In the center of the RF coils, the sign of the spin precession phase is inverted by a π flip of the neutron spins, such that non-uniform spin precession at the boundaries of the RF flippers is canceled. The residual inhomogeneity can be reduced by Fresnel- or Pythagoras-coils as in the case of conventional spin echo instruments (neutron spin echo (NSE)). Due to the good intrinsic homogeneity of the B0 coils, the current densities required for the correction coils are at least a factor of three less than in conventional NSE. As the precision and the current density of the correction coils are the limiting factors for the resolution of both NSE and LNRSE, the latter has the intrinsic potential to surpass the energy resolution of present NSE instruments. Our prototype LNRSE spectrometer described here was implemented at the resonance spin echo for diverse applications (RESEDA) beamline at the MLZ in Garching, Germany. The DC fields are generated by B0 coils, based on resistive split-pair solenoids with an active shielding for low stray fields along the beam path. One pair of RF flippers at a distance of 2 m generates a field integral of ˜0.5 Tm. The LNRSE technique is a future alternative for high-resolution spectroscopy of quasi-elastic excitations. In addition, it also incorporates the MIEZE technique, which allows to achieve spin echo resolution for spin depolarizing samples and sample environments. Here we present the results of numerical optimization of the coil geometry and first data from the prototype instrument.

Preliminary On-Orbit Neutron Dose Equivalent and Energy Spectrum Results from the ISS-RAD Fast Neutron Detector (FND)

NASA Technical Reports Server (NTRS)

Semones, Edward; Leitgab, Martin

2016-01-01

The ISS-RAD instrument was activated on ISS on February 1st, 2016. Integrated in ISS-RAD, the Fast Neutron Detector (FND) performs, for the first time on ISS, routine and precise direct neutron measurements between 0.5 and 8 MeV. Preliminary results for neutron dose equivalent and neutron flux energy distributions from online/on-board algorithms and offline ground analyses will be shown, along with comparisons to simulated data and previously measured neutron spectral data. On-orbit data quality and pre-launch analysis validation results will be discussed as well.

Bruyères-le-Châtel Neutron Evaluations of Actinides with the TALYS Code: The Fission Channel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Romain, P., E-mail: pascal.romain@cea.fr; Morillon, B.; Duarte, H.

For several years, various neutron evaluations of plutonium and uranium isotopes have been performed at Bruyères-le-Châtel (BRC), from 1 keV up to 30 MeV. Since only nuclear reaction models have been used to produce these evaluations, our approach was named the “Full Model” approach. Total, shape elastic and direct inelastic cross sections were obtained from the coupled channels model using a dispersive optical potential developed for actinides, with a large enough coupling scheme including the lowest octupolar band. All other cross sections were calculated using the Hauser-Feshbach theory (TALYS code) with a pre-equilibrium component above 8–10 MeV. In this paper,more » we focus our attention on the fission channel. More precisely, we will present the BRC contribution to fission modeling and the philosophy adopted in our “Full Model” approach. Performing evaluations with the “Full Model” approach implies the optimization of a large number of model parameters. With increasing neutron incident energy, many residual nuclei produced by nucleon emission also lead to fission. All available experimental data assigned to various fission mechanisms of the same nucleus were used to determine fission barrier parameters. For uranium isotopes, triple-humped fission barriers were required in order to reproduce accurately variations of the experimental fission cross sections. Our BRC fission modeling has shown that the effects of the class II or class III states located in the wells of the fission barrier sometimes provide an anti-resonant transmission rather than a resonant one. Consistent evaluations were produced for a large series of U and Pu isotopes. Resulting files were tested against integral data.« less

Study of a nTHGEM-based thermal neutron detector

NASA Astrophysics Data System (ADS)

Li, Ke; Zhou, Jian-Rong; Wang, Xiao-Dong; Xiong, Tao; Zhang, Ying; Xie, Yu-Guang; Zhou, Liang; Xu, Hong; Yang, Gui-An; Wang, Yan-Feng; Wang, Yan; Wu, Jin-Jie; Sun, Zhi-Jia; Hu, Bi-Tao

2016-07-01

With new generation neutron sources, traditional neutron detectors cannot satisfy the demands of the applications, especially under high flux. Furthermore, facing the global crisis in 3He gas supply, research on new types of neutron detector as an alternative to 3He is a research hotspot in the field of particle detection. GEM (Gaseous Electron Multiplier) neutron detectors have high counting rate, good spatial and time resolution, and could be one future direction of the development of neutron detectors. In this paper, the physical process of neutron detection is simulated with Geant4 code, studying the relations between thermal conversion efficiency, boron thickness and number of boron layers. Due to the special characteristics of neutron detection, we have developed a novel type of special ceramic nTHGEM (neutron THick GEM) for neutron detection. The performance of the nTHGEM working in different Ar/CO2 mixtures is presented, including measurements of the gain and the count rate plateau using a copper target X-ray source. A detector with a single nTHGEM has been tested for 2-D imaging using a 252Cf neutron source. The key parameters of the performance of the nTHGEM detector have been obtained, providing necessary experimental data as a reference for further research on this detector. Supported by National Natural Science Foundation of China (11127508, 11175199, 11205253, 11405191), Key Laboratory of Neutron Physics, CAEP (2013DB06, 2013BB04) and CAS (YZ201512)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nekoogar, F; Dowla, F; Wang, T

Recent advancements in the ultra-wide band Radio Frequency Identification (RFID) technology and solid state pillar type neutron detectors have enabled us to move forward in combining both technologies for advanced neutron monitoring. The LLNL RFID tag is totally passive and will operate indefinitely without the need for batteries. The tag is compact, can be directly mounted on metal, and has high performance in dense and cluttered environments. The LLNL coin-sized pillar solid state neutron detector has achieved a thermal neutron detection efficiency of 20% and neutron/gamma discrimination of 1E5. These performance values are comparable to a fieldable {sup 3}He basedmore » detector. In this paper we will discuss features about the two technologies and some potential applications for the advanced safeguarding of nuclear materials.« less

Probing the effect of neutron excess on the dynamics of Hf compound system

NASA Astrophysics Data System (ADS)

Sharma, Ishita; Kumar, Raj; Sharma, Manoj K.

2018-05-01

The reaction dynamics of 170Hf* and 174Hf* compound systems formed in 46,50Ti +124Sn reactions have been analyzed within the framework of Dynamical Cluster-decay Model (DCM) over an energy range of Ec.m.=115-156 MeV. The experimental data of fusion evaporation cross sections for 170,174Hf* systems is successfully addressed by optimizing the value of neck length ΔR. It is to be noted that calculations are performed by taking quadrupole deformations β2 with the optimum orientation of decaying fragments. Here, the effect of neutron excess has been examined on the structural properties of decaying fragments via fragmentation potential and preformation probability. The observation depicts that the isotopic effect prevails at fission region and as we move from lighter nucleus i.e. 170Hf* to heavier nucleus i.e. 174Hf*, the heavy mass fragments (HMF) start contributing along with fission fragments which are equally evident for 170Hf* as well. The ΔR values so obtained are compared at center of mass-energies which consequently provides information regarding the relative time scale of decaying fragments.

Microstructural and micromechanical characterization of IN718 theta shaped specimens built with electron beam melting

DOE PAGES

Cakmak, Ercan; Kirka, Michael M.; Watkins, Thomas R.; ...

2016-02-23

Theta-shaped specimens were additively manufactured out of Inconel 718 powders using an electron beam melting technique, as a model complex load bearing structure. We employed two different build strategies; producing two sets of specimens. Microstructural and micro-mechanical characterizations were performed using electron back-scatter, synchrotron x-ray and in-situ neutron diffraction techniques. In particular, the cross-members of the specimens were the focus of the synchrotron x-ray and in-situ neutron diffraction measurements. The build strategies employed resulted in the formation of distinct microstructures and crystallographic textures, signifying the importance of build-parameter manipulation for microstructural optimization. Large strain anisotropy of the different lattice planesmore » was observed during in-situ loading. Texture was concluded to have a distinct effect upon both the axial and transverse strain responses of the cross-members. In particular, the (200), (220) and (420) transverse lattice strains all showed unexpected overlapping trends in both builds. This was related to the strong {200} textures along the build/loading direction, providing agreement between the experimental and calculated results.« less

Study of the dynamics of poly(ethylene oxide) by combining molecular dynamic simulations and neutron scattering experiments

NASA Astrophysics Data System (ADS)

Brodeck, M.; Alvarez, F.; Arbe, A.; Juranyi, F.; Unruh, T.; Holderer, O.; Colmenero, J.; Richter, D.

2009-03-01

We performed quasielastic neutron scattering experiments and atomistic molecular dynamics simulations on a poly(ethylene oxide) (PEO) homopolymer system above the melting point. The excellent agreement found between both sets of data, together with a successful comparison with literature diffraction results, validates the condensed-phase optimized molecular potentials for atomistic simulation studies (COMPASS) force field used to produce our dynamic runs and gives support to their further analysis. This provided direct information on magnitudes which are not accessible from experiments such as the radial probability distribution functions of specific atoms at different times and their moments. The results of our simulations on the H-motions and different experiments indicate that in the high-temperature range investigated the dynamics is Rouse-like for Q-values below ≈0.6 Å-1. We then addressed the single chain dynamic structure factor with the simulations. A mode analysis, not possible directly experimentally, reveals the limits of applicability of the Rouse model to PEO. We discuss the possible origins for the observed deviations.

Study of the dynamics of poly(ethylene oxide) by combining molecular dynamic simulations and neutron scattering experiments.

PubMed

Brodeck, M; Alvarez, F; Arbe, A; Juranyi, F; Unruh, T; Holderer, O; Colmenero, J; Richter, D

2009-03-07

We performed quasielastic neutron scattering experiments and atomistic molecular dynamics simulations on a poly(ethylene oxide) (PEO) homopolymer system above the melting point. The excellent agreement found between both sets of data, together with a successful comparison with literature diffraction results, validates the condensed-phase optimized molecular potentials for atomistic simulation studies (COMPASS) force field used to produce our dynamic runs and gives support to their further analysis. This provided direct information on magnitudes which are not accessible from experiments such as the radial probability distribution functions of specific atoms at different times and their moments. The results of our simulations on the H-motions and different experiments indicate that in the high-temperature range investigated the dynamics is Rouse-like for Q-values below approximately 0.6 A(-1). We then addressed the single chain dynamic structure factor with the simulations. A mode analysis, not possible directly experimentally, reveals the limits of applicability of the Rouse model to PEO. We discuss the possible origins for the observed deviations.

rf improvements for Spallation Neutron Source H- ion sourcea)

NASA Astrophysics Data System (ADS)

Kang, Y. W.; Fuja, R.; Goulding, R. H.; Hardek, T.; Lee, S.-W.; McCarthy, M. P.; Piller, M. C.; Shin, K.; Stockli, M. P.; Welton, R. F.

2010-02-01

The Spallation Neutron Source at Oak Ridge National Laboratory is ramping up the accelerated proton beam power to 1.4 MW and just reached 1 MW. The rf-driven multicusp ion source that originates from the Lawrence Berkeley National Laboratory has been delivering ˜38 mA H- beam in the linac at 60 Hz, 0.9 ms. To improve availability, a rf-driven external antenna multicusp ion source with a water-cooled ceramic aluminum nitride (AlN) plasma chamber is developed. Computer modeling and simulations have been made to analyze and optimize the rf performance of the new ion source. Operational statistics and test runs with up to 56 mA medium energy beam transport beam current identify the 2 MHz rf system as a limiting factor in the system availability and beam production. Plasma ignition system is under development by using a separate 13 MHz system. To improve the availability of the rf power system with easier maintenance, we tested a 70 kV isolation transformer for the 80 kW, 6% duty cycle 2 MHz amplifier to power the ion source from a grounded solid-state amplifier.

rf improvements for Spallation Neutron Source H- ion source.

PubMed

Kang, Y W; Fuja, R; Goulding, R H; Hardek, T; Lee, S-W; McCarthy, M P; Piller, M C; Shin, K; Stockli, M P; Welton, R F

2010-02-01

The Spallation Neutron Source at Oak Ridge National Laboratory is ramping up the accelerated proton beam power to 1.4 MW and just reached 1 MW. The rf-driven multicusp ion source that originates from the Lawrence Berkeley National Laboratory has been delivering approximately 38 mA H(-) beam in the linac at 60 Hz, 0.9 ms. To improve availability, a rf-driven external antenna multicusp ion source with a water-cooled ceramic aluminum nitride (AlN) plasma chamber is developed. Computer modeling and simulations have been made to analyze and optimize the rf performance of the new ion source. Operational statistics and test runs with up to 56 mA medium energy beam transport beam current identify the 2 MHz rf system as a limiting factor in the system availability and beam production. Plasma ignition system is under development by using a separate 13 MHz system. To improve the availability of the rf power system with easier maintenance, we tested a 70 kV isolation transformer for the 80 kW, 6% duty cycle 2 MHz amplifier to power the ion source from a grounded solid-state amplifier.

Simulating variable source problems via post processing of individual particle tallies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bleuel, D.L.; Donahue, R.J.; Ludewigt, B.A.

2000-10-20

Monte Carlo is an extremely powerful method of simulating complex, three dimensional environments without excessive problem simplification. However, it is often time consuming to simulate models in which the source can be highly varied. Similarly difficult are optimization studies involving sources in which many input parameters are variable, such as particle energy, angle, and spatial distribution. Such studies are often approached using brute force methods or intelligent guesswork. One field in which these problems are often encountered is accelerator-driven Boron Neutron Capture Therapy (BNCT) for the treatment of cancers. Solving the reverse problem of determining the best neutron source formore » optimal BNCT treatment can be accomplished by separating the time-consuming particle-tracking process of a full Monte Carlo simulation from the calculation of the source weighting factors which is typically performed at the beginning of a Monte Carlo simulation. By post-processing these weighting factors on a recorded file of individual particle tally information, the effect of changing source variables can be realized in a matter of seconds, instead of requiring hours or days for additional complete simulations. By intelligent source biasing, any number of different source distributions can be calculated quickly from a single Monte Carlo simulation. The source description can be treated as variable and the effect of changing multiple interdependent source variables on the problem's solution can be determined. Though the focus of this study is on BNCT applications, this procedure may be applicable to any problem that involves a variable source.« less

Multi-particle inspection using associated particle sources

DOEpatents

Bingham, Philip R.; Mihalczo, John T.; Mullens, James A.; McConchie, Seth M.; Hausladen, Paul A.

2016-02-16

Disclosed herein are representative embodiments of methods, apparatus, and systems for performing combined neutron and gamma ray radiography. For example, one exemplary system comprises: a neutron source; a set of alpha particle detectors configured to detect alpha particles associated with neutrons generated by the neutron source; neutron detectors positioned to detect at least some of the neutrons generated by the neutron source; a gamma ray source; a set of verification gamma ray detectors configured to detect verification gamma rays associated with gamma rays generated by the gamma ray source; a set of gamma ray detectors configured to detect gamma rays generated by the gamma ray source; and an interrogation region located between the neutron source, the gamma ray source, the neutron detectors, and the gamma ray detectors.

Neutron irradiation effects on plasma facing materials

NASA Astrophysics Data System (ADS)

Barabash, V.; Federici, G.; Rödig, M.; Snead, L. L.; Wu, C. H.

2000-12-01

This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed.

Performance and accuracy of criticality calculations performed using WARP – A framework for continuous energy Monte Carlo neutron transport in general 3D geometries on GPUs

DOE PAGES

Bergmann, Ryan M.; Rowland, Kelly L.; Radnović, Nikola; ...

2017-05-01

In this companion paper to "Algorithmic Choices in WARP - A Framework for Continuous Energy Monte Carlo Neutron Transport in General 3D Geometries on GPUs" (doi:10.1016/j.anucene.2014.10.039), the WARP Monte Carlo neutron transport framework for graphics processing units (GPUs) is benchmarked against production-level central processing unit (CPU) Monte Carlo neutron transport codes for both performance and accuracy. We compare neutron flux spectra, multiplication factors, runtimes, speedup factors, and costs of various GPU and CPU platforms running either WARP, Serpent 2.1.24, or MCNP 6.1. WARP compares well with the results of the production-level codes, and it is shown that on the newestmore » hardware considered, GPU platforms running WARP are between 0.8 to 7.6 times as fast as CPU platforms running production codes. Also, the GPU platforms running WARP were between 15% and 50% as expensive to purchase and between 80% to 90% as expensive to operate as equivalent CPU platforms performing at an equal simulation rate.« less

Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

NASA Astrophysics Data System (ADS)

Acır, Adem; Altunok, Taner

2010-10-01

HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li20Sn80 coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.

Performance and accuracy of criticality calculations performed using WARP – A framework for continuous energy Monte Carlo neutron transport in general 3D geometries on GPUs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bergmann, Ryan M.; Rowland, Kelly L.; Radnović, Nikola

In this companion paper to "Algorithmic Choices in WARP - A Framework for Continuous Energy Monte Carlo Neutron Transport in General 3D Geometries on GPUs" (doi:10.1016/j.anucene.2014.10.039), the WARP Monte Carlo neutron transport framework for graphics processing units (GPUs) is benchmarked against production-level central processing unit (CPU) Monte Carlo neutron transport codes for both performance and accuracy. We compare neutron flux spectra, multiplication factors, runtimes, speedup factors, and costs of various GPU and CPU platforms running either WARP, Serpent 2.1.24, or MCNP 6.1. WARP compares well with the results of the production-level codes, and it is shown that on the newestmore » hardware considered, GPU platforms running WARP are between 0.8 to 7.6 times as fast as CPU platforms running production codes. Also, the GPU platforms running WARP were between 15% and 50% as expensive to purchase and between 80% to 90% as expensive to operate as equivalent CPU platforms performing at an equal simulation rate.« less

Development of Diagnostics for the Livermore DPF Devices

NASA Astrophysics Data System (ADS)

Mitrani, James; Prasad, Rahul R.; Podpaly, Yuri A.; Cooper, Christopher M.; Chapman, Steven F.; Shaw, Brian H.; Povilus, Alexander P.; Schmidt, Andrea

2017-10-01

LLNL is commissioning several new diagnostics to understand and optimize ion and neutron production in their dense plasma focus (DPF) systems. Gas fills used in DPF devices at LLNL are Deuterium (D2) and He accelerated onto a Be target, for production of neutrons. Neutron yields are currently measured with Helium-3 tubes, and development of yttrium-based activation detectors is currently underway. Neutron time-of-flight (nTOF) signals from prompt neutrons will be measured with gadolinium-doped liquid scintillators. An ion energy analyzer will be used to diagnose energy distribution of D + and He +2 ions. Additionally, a fast frame ICCD camera has been applied to image the plasma sheath during the rundown and pinch phases. Sheath velocity will be measured with an array of discrete photodiodes with ns time responses. A discussion of our results will be presented. Prepared by LLNL under Contract DE-AC52-07NA27344, and supported by the Laboratory Directed Research and Development Program (15-ERD-034) at LLNL and the Office of Defense Nuclear Nonproliferation Research and Development within U.S. Department of Energy.

Development of a digital method for neutron/gamma-ray discrimination based on matched filtering

NASA Astrophysics Data System (ADS)

Korolczuk, S.; Linczuk, M.; Romaniuk, R.; Zychor, I.

2016-09-01

Neutron/gamma-ray discrimination is crucial for measurements with detectors sensitive to both neutron and gamma-ray radiation. Different techniques to discriminate between neutrons and gamma-rays based on pulse shape analysis are widely used in many applications, e.g., homeland security, radiation dosimetry, environmental monitoring, fusion experiments, nuclear spectroscopy. A common requirement is to improve a radiation detection level with a high detection reliability. Modern electronic components, such as high speed analog to digital converters and powerful programmable digital circuits for signal processing, allow us to develop a fully digital measurement system. With this solution it is possible to optimize digital signal processing algorithms without changing any electronic components in an acquisition signal path. We report on results obtained with a digital acquisition system DNG@NCBJ designed at the National Centre for Nuclear Research. A 2'' × 2'' EJ309 liquid scintillator was used to register mixed neutron and gamma-ray radiation from PuBe sources. A dedicated algorithm for pulse shape discrimination, based on real-time filtering, was developed and implemented in hardware.

Neutron Characterization for Additive Manufacturing

NASA Technical Reports Server (NTRS)

Watkins, Thomas; Bilheux, Hassina; An, Ke; Payzant, Andrew; DeHoff, Ryan; Duty, Chad; Peter, William; Blue, Craig; Brice, Craig A.

2013-01-01

Oak Ridge National Laboratory (ORNL) is leveraging decades of experience in neutron characterization of advanced materials together with resources such as the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR) shown in Fig. 1 to solve challenging problems in additive manufacturing (AM). Additive manufacturing, or three-dimensional (3-D) printing, is a rapidly maturing technology wherein components are built by selectively adding feedstock material at locations specified by a computer model. The majority of these technologies use thermally driven phase change mechanisms to convert the feedstock into functioning material. As the molten material cools and solidifies, the component is subjected to significant thermal gradients, generating significant internal stresses throughout the part (Fig. 2). As layers are added, inherent residual stresses cause warping and distortions that lead to geometrical differences between the final part and the original computer generated design. This effect also limits geometries that can be fabricated using AM, such as thin-walled, high-aspect- ratio, and overhanging structures. Distortion may be minimized by intelligent toolpath planning or strategic placement of support structures, but these approaches are not well understood and often "Edisonian" in nature. Residual stresses can also impact component performance during operation. For example, in a thermally cycled environment such as a high-pressure turbine engine, residual stresses can cause components to distort unpredictably. Different thermal treatments on as-fabricated AM components have been used to minimize residual stress, but components still retain a nonhomogeneous stress state and/or demonstrate a relaxation-derived geometric distortion. Industry, federal laboratory, and university collaboration is needed to address these challenges and enable the U.S. to compete in the global market. Work is currently being conducted on AM technologies at the ORNL Manufacturing Demonstration Facility (MDF) sponsored by the DOE's Advanced Manufacturing Office. The MDF is focusing on R&D of both metal and polymer AM pertaining to in-situ process monitoring and closed-loop controls; implementation of advanced materials in AM technologies; and demonstration, characterization, and optimization of next-generation technologies. ORNL is working directly with industry partners to leverage world-leading facilities in fields such as high performance computing, advanced materials characterization, and neutron sciences to solve fundamental challenges in advanced manufacturing. Specifically, MDF is leveraging two of the world's most advanced neutron facilities, the HFIR and SNS, to characterize additive manufactured components.

Data-optimized source modeling with the Backwards Liouville Test–Kinetic method

DOE PAGES

Woodroffe, J. R.; Brito, T. V.; Jordanova, V. K.; ...

2017-09-14

In the standard practice of neutron multiplicity counting , the first three sampled factorial moments of the event triggered neutron count distribution were used to quantify the three main neutron source terms: the spontaneous fissile material effective mass, the relative (α,n) production and the induced fission source responsible for multiplication. Our study compares three methods to quantify the statistical uncertainty of the estimated mass: the bootstrap method, propagation of variance through moments, and statistical analysis of cycle data method. Each of the three methods was implemented on a set of four different NMC measurements, held at the JRC-laboratory in Ispra,more » Italy, sampling four different Pu samples in a standard Plutonium Scrap Multiplicity Counter (PSMC) well counter.« less

Signal and background considerations for the MRSt on the National Ignition Facility (NIF)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wink, C. W., E-mail: cwink@mit.edu; Frenje, J. A.; Gatu Johnson, M.

2016-11-15

A Magnetic Recoil Spectrometer (MRSt) has been conceptually designed for time-resolved measurements of the neutron spectrum at the National Ignition Facility. Using the MRSt, the goals are to measure the time-evolution of the spectrum with a time resolution of ∼20-ps and absolute accuracy better than 5%. To meet these goals, a detailed understanding and optimization of the signal and background characteristics are required. Through ion-optics, MCNP simulations, and detector-response calculations, it is demonstrated that the goals and a signal-to background >5–10 for the down-scattered neutron measurement are met if the background, consisting of ambient neutrons and gammas, at the MRStmore » is reduced 50–100 times.« less

Signal and background considerations for the MRSt on the National Ignition Facility (NIF)

DOE PAGES

Wink, C. W.; Frenje, J. A.; Hilsabeck, T. J.; ...

2016-08-03

A Magnetic Recoil Spectrometer (MRSt) has been conceptually designed for time-resolved measurements of the neutron spectrum at the National Ignition Facility. Using the MRSt, the goals are to measure the time-evolution of the spectrum with a time resolution of ~20-ps and absolute accuracy better than 5%. To meet these goals, a detailed understanding and optimization of the signal and background characteristics are required. Through ion-optics, MCNP simulations, and detector-response calculations, we demonstrate that the goals and a signal-to background >5-10 for the down-scattered neutron measurement are met if the background, consisting of ambient neutrons and gammas, at the MRSt ismore » reduced 50-100 times.« less

Experimental parameters optimization of instrumental neutron activation analysis in order to determine selected elements in some industrial soils in Turkey

NASA Astrophysics Data System (ADS)

Haciyakupoglu, Sevilay; Nur Esen, Ayse; Erenturk, Sema

2014-08-01

The purpose of this study is optimization of the experimental parameters for analysis of soil matrix by instrumental neutron activation analysis and quantitative determination of barium, cerium, lanthanum, rubidium, scandium and thorium in soil samples collected from industrialized urban areas near Istanbul. Samples were irradiated in TRIGA MARK II Research Reactor of Istanbul Technical University. Two types of reference materials were used to check the accuracy of the applied method. The achieved results were found to be in compliance with certified values of the reference materials. The calculated En numbers for mentioned elements were found to be less than 1. The presented data of element concentrations in soil samples will help to trace the pollution as an impact of urbanization and industrialization, as well as providing database for future studies.

EXOGAM at the ILL: the EXILL campaign

NASA Astrophysics Data System (ADS)

de France, G.; Blanc, A.; Drouet, F.; Jentschel, M.; Köster, U.; Mutti, P.; Régis, J. M.; Simpson, G.; Soldner, T.; Stezowski, O.; Ur, C. A.; Urban, W.; Vancrayenest, A.

2018-02-01

A combination of germanium detectors has been installed at the PF1B neutron guide of the ILL to perform the prompt spectroscopy of neutron-rich nuclei produced in the neutron-capture induced-fission of 235U and 241Pu. Radiative capture reactions on rare targets have also been performed. LaBr3 detectors from the FATIMA collaboration have also been installed in complement with the EXOGAM clovers to measure lifetimes of low-lying excited states. The measured characteristics indicate very good performances of the overall setup. Some recent results will be discussed.

A measurement of the absolute neutron beam polarization produced by an optically pumped 3He neutron spin filter

NASA Astrophysics Data System (ADS)

Rich, D. R.; Bowman, J. D.; Crawford, B. E.; Delheij, P. P. J.; Espy, M. A.; Haseyama, T.; Jones, G.; Keith, C. D.; Knudson, J.; Leuschner, M. B.; Masaike, A.; Masuda, Y.; Matsuda, Y.; Penttilä, S. I.; Pomeroy, V. R.; Smith, D. A.; Snow, W. M.; Szymanski, J. J.; Stephenson, S. L.; Thompson, A. K.; Yuan, V.

2002-04-01

The capability of performing accurate absolute measurements of neutron beam polarization opens a number of exciting opportunities in fundamental neutron physics and in neutron scattering. At the LANSCE pulsed neutron source we have measured the neutron beam polarization with an absolute accuracy of 0.3% in the neutron energy range from 40 meV to 10 eV using an optically pumped polarized 3He spin filter and a relative transmission measurement technique. 3He was polarized using the Rb spin-exchange method. We describe the measurement technique, present our results, and discuss some of the systematic effects associated with the method.

Characterization of plastic and boron carbide additive manufactured neutron collimators

NASA Astrophysics Data System (ADS)

Stone, M. B.; Siddel, D. H.; Elliott, A. M.; Anderson, D.; Abernathy, D. L.

2017-12-01

Additive manufacturing techniques allow for the production of materials with complicated geometries with reduced costs and production time over traditional methods. We have applied this technique to the production of neutron collimators for use in thermal and cold neutron scattering instrumentation directly out of boron carbide. We discuss the design and generation of these collimators. We also provide measurements at neutron scattering beamlines which serve to characterize the performance of these collimators. Additive manufacturing of parts using neutron absorbing material may also find applications in radiography and neutron moderation.

Transmutation approximations for the application of hybrid Monte Carlo/deterministic neutron transport to shutdown dose rate analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Biondo, Elliott D.; Wilson, Paul P. H.

In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using deterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation ofmore » an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed here, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR problem, GT-CADIS provides speedups of 200 100 relative to global variance reduction with the Forward-Weighted (FW)-CADIS method and 9 ± 5 • 104 relative to analog. As a result, this work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR analysis.« less

Transmutation approximations for the application of hybrid Monte Carlo/deterministic neutron transport to shutdown dose rate analysis

DOE PAGES

Biondo, Elliott D.; Wilson, Paul P. H.

2017-05-08

In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using deterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation ofmore » an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed here, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR problem, GT-CADIS provides speedups of 200 100 relative to global variance reduction with the Forward-Weighted (FW)-CADIS method and 9 ± 5 • 104 relative to analog. As a result, this work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR analysis.« less

Neutron Scattering Studies on Large Length Scale Sample Structures

NASA Astrophysics Data System (ADS)

Feng, Hao

Neutron scattering can be used to study structures of matter. Depending on the interested sample properties, different scattering techniques can be chosen. Neutron reflectivity is more often used to detect in-depth profile of layered structures and the interfacial roughness while transmission is more sensitive to sample bulk properties. Neutron Reflectometry (NR) technique, one technique in neutron reflectivity, is first discussed in this thesis. Both specular reflectivity and the first order Bragg intensity were measured in the NR experiment with a diffraction grating in order to study the in-depth and the lateral structure of a sample (polymer) deposited on the grating. However, the first order Bragg intensity solely is sometimes inadequate to determine the lateral structure and high order Bragg intensities are difficult to measure using traditional neutron scattering techniques due to the low brightness of the current neutron sources. Spin Echo Small Angle Neutron Scattering (SESANS) technique overcomes this resolution problem by measuring the Fourier transforms of all the Bragg intensities, resulting in measuring the real-space density correlations of samples and allowing the accessible length scale from few-tens of nanometers to several microns. SESANS can be implemented by using two pairs of magnetic Wollaston prims (WP) and the accessible length scale is proportional to the magnetic field intensity in WPs. To increase the magnetic field and thus increase the accessible length scale, an apparatus named Superconducting Wollaston Prisms (SWP) which has a series of strong, well-defined shaped magnetic fields created by superconducting coils was developed in Indiana University in 2016. Since then, various kinds of optimization have been implemented, which are addressed in this thesis. Finally, applications of SWPs in other neutron scattering techniques like Neutron Larmor Diffraction (NLD) are discussed.

Progress towards developing neutron tolerant magnetostrictive and piezoelectric transducers

NASA Astrophysics Data System (ADS)

Reinhardt, Brian; Tittmann, Bernhard; Rempe, Joy; Daw, Joshua; Kohse, Gordon; Carpenter, David; Ames, Michael; Ostrovsky, Yakov; Ramuhalli, Pradeep; Montgomery, Robert; Chien, Hualte; Wernsman, Bernard

2015-03-01

Current generation light water reactors (LWRs), sodium cooled fast reactors (SFRs), small modular reactors (SMRs), and next generation nuclear plants (NGNPs) produce harsh environments in and near the reactor core that can severely tax material performance and limit component operational life. To address this issue, several Department of Energy Office of Nuclear Energy (DOE-NE) research programs are evaluating the long duration irradiation performance of fuel and structural materials used in existing and new reactors. In order to maximize the amount of information obtained from Material Testing Reactor (MTR) irradiations, DOE is also funding development of enhanced instrumentation that will be able to obtain in-situ, real-time data on key material characteristics and properties, with unprecedented accuracy and resolution. Such data are required to validate new multi-scale, multi-physics modeling tools under development as part of a science-based, engineering driven approach to reactor development. It is not feasible to obtain high resolution/microscale data with the current state of instrumentation technology. However, ultrasound-based sensors offer the ability to obtain such data if it is demonstrated that these sensors and their associated transducers are resistant to high neutron flux, high gamma radiation, and high temperature. To address this need, the Advanced Test Reactor National Scientific User Facility (ATR-NSUF) is funding an irradiation, led by PSU, at the Massachusetts Institute of Technology Research Reactor to test the survivability of ultrasound transducers. As part of this effort, PSU and collaborators have designed, fabricated, and provided piezoelectric and magnetostrictive transducers that are optimized to perform in harsh, high flux, environments. Four piezoelectric transducers were fabricated with either aluminum nitride, zinc oxide, or bismuth titanate as the active element that were coupled to either Kovar or aluminum waveguides and two magnetostrictive transducers were fabricated with Remendur or Galfenol as the active elements. Pulse-echo ultrasonic measurements of these transducers are made in-situ. This paper will present an overview of the test design including selection criteria for candidate materials and optimization of test assembly parameters, data obtained from both out-of-pile and in-pile testing at elevated temperatures, and an assessment based on initial data of the expected performance of ultrasonic devices in irradiation conditions.

A GDT-based fusion neutron source for academic and industrial applications

NASA Astrophysics Data System (ADS)

Anderson, J. K.; Forest, C. B.; Mirnov, V. V.; Peterson, E. E.; Waleffe, R.; Wallace, J.; Harvey, R. W.

2017-10-01

The design of a fusion neutron source based on the gas dynamic trap (GDT) configuration is underway. The motivation is both the ends and the means. There are immediate applications for neutrons including medical isotope production and actinide burners. Taking the next step in the magnetic mirror path will leverage advances in high-temperature superconducting magnets and additive manufacturing in confining a fusion plasma, and both the technological and physics bases exist. Recent breakthrough results at the GDT facility in Russia demonstrate stable confinement of a beta 60% mirror plasma at high Te ( 1 keV). These scale readily to a fusion neutron source with an increase in magnetic field, mirror ratio, and ion energy. Studies of a next-step compact device focus on calculations of MHD equilibrium and stability, and Fokker-Planck modeling to optimize the heating scenario. The conceptualized device uses off-the-shelf MRI magnets for a 1 T central field, REBCO superconducting mirror coils (which can currently produce fields in excess of 30T), and existing 75 keV NBI and 140 GHz ECRH. High harmonic fast wave injection is damped on beam ions, dramatically increasing the fusion reactivity for an incremental bump in input power. MHD stability is achieved with the vortex confinement scheme, where a biasing profile imposes optimal ExB rotation of the plasma. Liquid metal divertors are being considered in the end cells. Work supported by the Wisconsin Alumni Research Foundation.

First measurement of the VESUVIO neutron spectrum in the 30-80 MeV energy range using a Proton Recoil Telescope technique

NASA Astrophysics Data System (ADS)

Cazzaniga, C.; Tardocchi, M.; Croci, G.; Frost, C.; Giacomelli, L.; Grosso, G.; Hjalmarsson, A.; Rebai, M.; Rhodes, N. J.; Schooneveld, E. M.; Gorini, G.

2013-11-01

Measurements of the fast neutron energy spectrum at the ISIS spallation source are reported. The measurements were performed with a Proton Recoil Telescope consisting of a thin plastic foil placed in the neutron beam and two scintillator detectors. Results in the neutron energy range 30 MeV < En < 80 MeV are in good agreement with Monte Carlo simulations of the neutron spectrum.

Development of Fast Measurement System of Neutron Emission Profile Using a Digital Signal Processing Technique in JT-60U

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ishikawa, M.; Shinohara, K.; Itoga, T.

2008-03-12

Neutron emission profiles are routinely measured in JT-60U Tokamak. Stinbene neuron detectors (SNDs), which combine a Stilbene organic crystal scintillation detector (Stilbene detector) with an analog neutron-gamma pulse shape discrimination (PSD) circuit, have been used to measure neutron flux efficiently. Although the SND has many advantages as a neutron detector, the maximum count rate is limited up to {approx}1x 10{sup 5} counts/s due to the dead time of the analog PSD circuit. To overcome this issue, a digital signal processing (DSP) system using a Flash-ADC has been developed. In this system, anode signals from the photomultiplier of the Stilbene detectormore » are fed to the Flash ADC and digitized. Then, the PSD between neutrons and gamma-rays are performed using software. The photomultiplier tube is also modified to suppress and correct gain fluctuation of the photomultiplier. The DSP system has been installed in the center channel of the vertical neutron collimator system in JT-60U and applied to measurements of neutron flux in JT-60U experiments. Neutron flux are successfully measured with count rate up to {approx}1x 10{sup 6} counts/s without the effect of pile up of detected pulses. The performance of the DSP system as a neutron detector is demonstrated.« less

Assay of the Martian Regolith with Neutrons

NASA Technical Reports Server (NTRS)

Drake, Darrell M.

1997-01-01

The purpose of the research is to combine experiments and Monte Carlo transport of neutrons through volume of soil in an attempt to model neutron leakage from planetary surfaces. Emphasis is given to the change of neutron spectra as a function of water content and location. During the first stage of effort, two experiments were conducted in which leakage of neutrons from a Pu-Be source through about 30 g/cm(exp 2) of soil were measured with several counters. A Monte Carlo code, MCNP, has been used to model many of the 100 individual runs of the experiment. Hydrogen is the element that has the most dramatic effect on the neutron spectrum and its effect on the neutron spectrum is almost the same whether it is in the form of water or polyethylene. In order to simulate various water configurations, sheets of polyethylene have been used between layers of soil as well as water in several concentrations up to 18%. Comparison of experimental results to theoretical predictions made with the MCNP code were disappointing for low concentrations of water. We have made extensive calculations to see if room return could be the cause of the discrepancies. Water concentrations of the 'dry' soil were measured by two different laboratories and differed only by 0.5%. We have made calculations to optimize the next experiment and are investigating other methods of determining the water content of 'dry' soil.

Time-resolved Neutron-gamma-ray Data Acquisition for in Situ Subsurface Planetary Geochemistry

NASA Technical Reports Server (NTRS)

Bodnarik, Julie G.; Burger, Dan Michael; Burger, A.; Evans, L. G.; Parsons, A. M.; Schweitzer, J. S.; Starr R. D.; Stassun, K. G.

2013-01-01

The current gamma-ray/neutron instrumentation development effort at NASA Goddard Space Flight Center aims to extend the use of active pulsed neutron interrogation techniques to probe the subsurface elemental composition of planetary bodies in situ. Previous NASA planetary science missions, that used neutron and/or gamma-ray spectroscopy instruments, have relied on neutrons produced from galactic cosmic rays. One of the distinguishing features of this effort is the inclusion of a high intensity 14.1 MeV pulsed neutron generator synchronized with a custom data acquisition system to time each event relative to the pulse. With usually only one opportunity to collect data, it is difficult to set a priori time-gating windows to obtain the best possible results. Acquiring time-tagged, event-by-event data from nuclear induced reactions provides raw data sets containing channel/energy, and event time for each gamma ray or neutron detected. The resulting data set can be plotted as a function of time or energy using optimized analysis windows after the data are acquired. Time windows can now be chosen to produce energy spectra that yield the most statistically significant and accurate elemental composition results that can be derived from the complete data set. The advantages of post-processing gamma-ray time-tagged event-by-event data in experimental tests using our prototype instrument will be demonstrated.

Monte Carlo simulations and benchmark measurements on the response of TE(TE) and Mg(Ar) ionization chambers in photon, electron and neutron beams

NASA Astrophysics Data System (ADS)

Lin, Yi-Chun; Huang, Tseng-Te; Liu, Yuan-Hao; Chen, Wei-Lin; Chen, Yen-Fu; Wu, Shu-Wei; Nievaart, Sander; Jiang, Shiang-Huei

2015-06-01

The paired ionization chambers (ICs) technique is commonly employed to determine neutron and photon doses in radiology or radiotherapy neutron beams, where neutron dose shows very strong dependence on the accuracy of accompanying high energy photon dose. During the dose derivation, it is an important issue to evaluate the photon and electron response functions of two commercially available ionization chambers, denoted as TE(TE) and Mg(Ar), used in our reactor based epithermal neutron beam. Nowadays, most perturbation corrections for accurate dose determination and many treatment planning systems are based on the Monte Carlo technique. We used general purposed Monte Carlo codes, MCNP5, EGSnrc, FLUKA or GEANT4 for benchmark verifications among them and carefully measured values for a precise estimation of chamber current from absorbed dose rate of cavity gas. Also, energy dependent response functions of two chambers were calculated in a parallel beam with mono-energies from 20 keV to 20 MeV photons and electrons by using the optimal simple spherical and detailed IC models. The measurements were performed in the well-defined (a) four primary M-80, M-100, M120 and M150 X-ray calibration fields, (b) primary 60Co calibration beam, (c) 6 MV and 10 MV photon, (d) 6 MeV and 18 MeV electron LINACs in hospital and (e) BNCT clinical trials neutron beam. For the TE(TE) chamber, all codes were almost identical over the whole photon energy range. In the Mg(Ar) chamber, MCNP5 showed lower response than other codes for photon energy region below 0.1 MeV and presented similar response above 0.2 MeV (agreed within 5% in the simple spherical model). With the increase of electron energy, the response difference between MCNP5 and other codes became larger in both chambers. Compared with the measured currents, MCNP5 had the difference from the measurement data within 5% for the 60Co, 6 MV, 10 MV, 6 MeV and 18 MeV LINACs beams. But for the Mg(Ar) chamber, the derivations reached 7.8-16.5% below 120 kVp X-ray beams. In this study, we were especially interested in BNCT doses where low energy photon contribution is less to ignore, MCNP model is recognized as the most suitable to simulate wide photon-electron and neutron energy distributed responses of the paired ICs. Also, MCNP provides the best prediction of BNCT source adjustment by the detector's neutron and photon responses.

Epithermal neutron beams from the 7 Li(p,n) reaction near the threshold for neutron capture therapy

NASA Astrophysics Data System (ADS)

Porras, I.; Praena, J.; Arias de Saavedra, F.; Pedrosa, M.; Esquinas, P.; L. Jiménez-Bonilla, P.

2016-11-01

Two applications for neutron capture therapy of epithermal neutron beams calculated from the 7Li ( p , n reaction are discussed. In particular, i) for a proton beam of 1920 keV of a 30 mA, a neutron beam of adequate features for BNCT is found at an angle of 80° from the forward direction; and ii) for a proton beam of 1910 keV, a neutron beam is obtained at the forward direction suitable for performing radiobiology experiments for the determination of the biological weighting factors of the fast dose component in neutron capture therapy.

A study on the optimum fast neutron flux for boron neutron capture therapy of deep-seated tumors.

PubMed

Rasouli, Fatemeh S; Masoudi, S Farhad

2015-02-01

High-energy neutrons, named fast neutrons which have a number of undesirable biological effects on tissue, are a challenging problem in beam designing for Boron Neutron Capture Therapy, BNCT. In spite of this fact, there is not a widely accepted criterion to guide the beam designer to determine the appropriate contribution of fast neutrons in the spectrum. Although a number of researchers have proposed a target value for the ratio of fast neutron flux to epithermal neutron flux, it can be shown that this criterion may not provide the optimum treatment condition. This simulation study deals with the determination of the optimum contribution of fast neutron flux in the beam for BNCT of deep-seated tumors. Since the dose due to these high-energy neutrons damages shallow tissues, delivered dose to skin is considered as a measure for determining the acceptability of the designed beam. To serve this purpose, various beam shaping assemblies that result in different contribution of fast neutron flux are designed. The performances of the neutron beams corresponding to such configurations are assessed in a simulated head phantom. It is shown that the previously used criterion, which suggests a limit value for the contribution of fast neutrons in beam, does not necessarily provide the optimum condition. Accordingly, it is important to specify other complementary limits considering the energy of fast neutrons. By analyzing various neutron spectra, two limits on fast neutron flux are proposed and their validity is investigated. The results show that considering these limits together with the widely accepted IAEA criteria makes it possible to have a more realistic assessment of sufficiency of the designed beam. Satisfying these criteria not only leads to reduction of delivered dose to skin, but also increases the advantage depth in tissue and delivered dose to tumor during the treatment time. The Monte Carlo Code, MCNP-X, is used to perform these simulations. Copyright © 2014 Elsevier Ltd. All rights reserved.

The cold neutron chopper spectrometer at the Spallation Neutron Source—A review of the first 8 years of operation

DOE PAGES

Ehlers, G.; Podlesnyak, A. A.; Kolesnikov, A. I.

2016-09-13

The first eight years of operation of the Cold Neutron Chopper Spectrometer (CNCS) at the Spallation Neutron Source in Oak Ridge is being reviewed. The instrument has been part of the facility user program since 2009, and more than 250 individual user experiments have been performed to date. CNCS is an extremely powerful and versatile instrument and offers leading edge performance in terms of beam intensity, energy resolution, and flexibility to trade one for another. In addition, experiments are being routinely performed with the sample at extreme conditions: T ≲ 0.05 K, p ≳ 2 GPa, and B = 8more » T can be achieved individually or in combination. In particular, CNCS is in a position to advance the state of the art with inelastic neutron scattering under pressure, and some of the recent accomplishments in this area will be presented in more detail.« less