NEUTRON DENSITY CONTROL IN A NEUTRONIC REACTOR

DOEpatents

Young, G.J.

1959-06-30

The method and means for controlling the neutron density in a nuclear reactor is described. It describes the method and means for flattening the neutron density distribution curve across the reactor by spacing the absorbing control members to varying depths in the central region closer to the center than to the periphery of the active portion of the reactor to provide a smaller neutron reproduction ratio in the region wherein the members are inserted, than in the remainder of the reactor thereby increasing the over-all potential power output.

Optimization of radiation shielding material aiming at compactness, lightweight, and low activation for a vehicle-mounted accelerator-driven D-T neutron source.

PubMed

Cai, Yao; Hu, Huasi; Lu, Shuangying; Jia, Qinggang

2018-05-01

To minimize the size and weight of a vehicle-mounted accelerator-driven D-T neutron source and protect workers from unnecessary irradiation after the equipment shutdown, a method to optimize radiation shielding material aiming at compactness, lightweight, and low activation for the fast neutrons was developed. The method employed genetic algorithm, combining MCNP and ORIGEN codes. A series of composite shielding material samples were obtained by the method step by step. The volume and weight needed to build a shield (assumed as a coaxial tapered cylinder) were adopted to compare the performance of the materials visually and conveniently. The results showed that the optimized materials have excellent performance in comparison with the conventional materials. The "MCNP6-ACT" method and the "rigorous two steps" (R2S) method were used to verify the activation grade of the shield irradiated by D-T neutrons. The types of radionuclide, the energy spectrum of corresponding decay gamma source, and the variation in decay gamma dose rate were also computed. Copyright © 2018 Elsevier Ltd. All rights reserved.

Determination of rhenium in molybdenite by neutron-activation analysis.

PubMed

Terada, K; Yoshimura, Y; Osaki, S; Kiba, T

1967-01-01

A neutron-activation method is described for the determination of rhenium in molybdenite. Radiochemical separation by a carrier technique was carried out very rapidly by means of successive liquid-liquid extraction processes. The recovery of rhenium, which was determined by a spectrophotometric method, was about 93%. About 10 samples could be analysed within 6 hr in parallel runs.

Monte Carlo simulation of thermal neutron flux of americium-beryllium source used in neutron activation analysis

NASA Astrophysics Data System (ADS)

Didi, Abdessamad; Dadouch, Ahmed; Bencheikh, Mohamed; Jai, Otman

2017-09-01

The neutron activation analysis is a method of exclusively elemental analysis. Its implementation of irradiates the sample which can be analyzed by a high neutron flux, this method is widely used in developed countries with nuclear reactors or accelerators of particle. The purpose of this study is to develop a prototype to increase the neutron flux such as americium-beryllium and have the opportunity to produce radioisotopes. Americium-beryllium is a mobile source of neutron activity of 20 curie, and gives a thermal neutron flux of (1.8 ± 0.0007) × 106 n/cm2 s when using water as moderator, when using the paraffin, the thermal neutron flux increases to (2.2 ± 0.0008) × 106 n/cm2 s, in the case of adding two solid beryllium barriers, the distance between them is 24 cm, parallel and symmetrical about the source, the thermal flux is increased to (2.5 ± 0.0008) × 106 n/cm2 s and in the case of multi-source (6 sources), with-out barriers, increases to (1.17 ± 0.0008) × 107 n/cm2 s with a rate of increase equal to 4.3 and with the both barriers flux increased to (1.37 ± 0.0008) × 107 n/cm2 s.

Accidental neutron dosimetry with human hair

NASA Astrophysics Data System (ADS)

Ekendahl, Daniela; Bečková, Věra; Zdychová, Vlasta; Bulánek, Boris; Prouza, Zdeněk; Štefánik, Milan

2014-11-01

Human hair contains sulfur, which can be activated by fast neutrons. The 32S(n,p)32P reaction with a threshold of 2.5 MeV was used for fast neutron dose estimation. It is a very important parameter for individual dose reconstruction with regards to the heterogeneity of the neutron transfer to the human body. Samples of human hair were irradiated in a radial channel of a training reactor VR-1. 32P activity in hair was measured both, directly by means of a proportional counter, and as ash dispersed in a liquid scintillator. Based on neutron spectrum estimation, a relationship between the neutron dose and induced activity was derived. The experiment verified the practical feasibility of this dosimetry method in cases of criticality accidents or malevolent acts with nuclear materials.

Non-destructive method for determining neutron exposure and constituent concentrations of a body

DOEpatents

Gold, Raymond; McElroy, William N.

1986-01-01

A non-destructive method for determination of neutron exposure and constituent concentrations in an object, such as reactor pressure vessel, is based on the observation of characteristic gamma-rays emitted by activation products in the object by using a unique continuous gamma-ray spectrometer. The spectrometer views the object through appropriate collimators to determine the absolute emission rate of these characteristic gamma-rays, thereby ascertaining the absolute activity of given activation products in the object. These data can then be used to deduce the spatial and angular dependence of neutron exposure or the spatial constituent concentration at regions of interest within the object.

Non-destructive method for determining neutron exposure and constituent concentrations of a body

DOEpatents

Gold, R.; McElroy, W.N.

1984-02-22

A non-destructive method for determination of neutron exposure and constituent concentrations in an object, such as a reactor pressure vessel, is based on the observation of characteristic gamma-rays emitted by activation products in the object by using a unique continuous gamma-ray spectrometer. The spectrometer views the object through appropriate collimators to determine the absolute emission rate of these characteristic gamma-rays, thereby ascertaining the absolute activity of given activation products in the object. These data can then be used to deduce the spatial and angular dependence of neutron exposure or the spatial constituent concentrations at regions of interest within the object.

Mock-up experiment at Birmingham University for BNCT project of Osaka University--Neutron flux measurement with gold foil.

PubMed

Tamaki, S; Sakai, M; Yoshihashi, S; Manabe, M; Zushi, N; Murata, I; Hoashi, E; Kato, I; Kuri, S; Oshiro, S; Nagasaki, M; Horiike, H

2015-12-01

Mock-up experiment for development of accelerator based neutron source for Osaka University BNCT project was carried out at Birmingham University, UK. In this paper, spatial distribution of neutron flux intensity was evaluated by foil activation method. Validity of the design code system was confirmed by comparing measured gold foil activities with calculations. As a result, it was found that the epi-thermal neutron beam was well collimated by our neutron moderator assembly. Also, the design accuracy was evaluated to have less than 20% error. Copyright © 2015 Elsevier Ltd. All rights reserved.

Accuracy of neutron self-activation method with iodine-containing scintillators for quantifying 128I generation using decay-fitting technique

NASA Astrophysics Data System (ADS)

Nohtomi, Akihiro; Wakabayashi, Genichiro

2015-11-01

We evaluated the accuracy of a self-activation method with iodine-containing scintillators in quantifying 128I generation in an activation detector; the self-activation method was recently proposed for photo-neutron on-line measurements around X-ray radiotherapy machines. Here, we consider the accuracy of determining the initial count rate R0, observed just after termination of neutron irradiation of the activation detector. The value R0 is directly related to the amount of activity generated by incident neutrons; the detection efficiency of radiation emitted from the activity should be taken into account for such an evaluation. Decay curves of 128I activity were numerically simulated by a computer program for various conditions including different initial count rates (R0) and background rates (RB), as well as counting statistical fluctuations. The data points sampled at minute intervals and integrated over the same period were fit by a non-linear least-squares fitting routine to obtain the value R0 as a fitting parameter with an associated uncertainty. The corresponding background rate RB was simultaneously calculated in the same fitting routine. Identical data sets were also evaluated by a well-known integration algorithm used for conventional activation methods and the results were compared with those of the proposed fitting method. When we fixed RB = 500 cpm, the relative uncertainty σR0 /R0 ≤ 0.02 was achieved for R0/RB ≥ 20 with 20 data points from 1 min to 20 min following the termination of neutron irradiation used in the fitting; σR0 /R0 ≤ 0.01 was achieved for R0/RB ≥ 50 with the same data points. Reasonable relative uncertainties to evaluate initial count rates were reached by the decay-fitting method using practically realistic sampling numbers. These results clarified the theoretical limits of the fitting method. The integration method was found to be potentially vulnerable to short-term variations in background levels, especially instantaneous contaminations by spike-like noise. The fitting method easily detects and removes such spike-like noise.

Determination of neutron capture cross sections of 232Th at 14.1 MeV and 14.8 MeV using the neutron activation method

NASA Astrophysics Data System (ADS)

Lan, Chang-Lin; Zhang, Yi; Lv, Tao; Xie, Bao-Lin; Peng, Meng; Yao, Ze-En; Chen, Jin-Gen; Kong, Xiang-Zhong

2017-04-01

The 232Th(n, γ)233Th neutron capture reaction cross sections were measured at average neutron energies of 14.1 MeV and 14.8 MeV using the activation method. The neutron flux was determined using the monitor reaction 27Al(n,α)24Na. The induced gamma-ray activities were measured using a low background gamma ray spectrometer equipped with a high resolution HPGe detector. The experimentally determined cross sections were compared with the data in the literature, and the evaluated data of ENDF/B-VII.1, JENDL-4.0u+, and CENDL-3.1. The excitation functions of the 232Th(n,γ)233Th reaction were also calculated theoretically using the TALYS1.6 computer code. Supported by Chinese TMSR Strategic Pioneer Science and Technology Project-The Th-U Fuel Physics Term (XDA02010100) and National Natural Science Foundation of China (11205076, 21327801)

Physical principles of neutron-gamma materials monitoring

NASA Astrophysics Data System (ADS)

Pekarskii, G. Sh.

1986-03-01

The physical principles of secondary radiation methods in nondestructive testing are discussed. Among the techniques considered are: neutron activation analysis (NAA); the induced-radiation method; and quasialbedo recording of secondary gamma-radiation. Emphasis is given to the neutron-gamma method which consists of exposing test material to a neutron flux and recording the secondary gamma-radiation by means of a spectrometer. The limitations of the method in detecting local inhomogeneous defects (filled pores cracks, and inclusions) in metal layers and multicomponents materials are described, and some advantages of the method over NAA are discussed. Formulas are derived for estimating the optimum density of the gamma-ray flux which is received by the detector.

Physical principles of neutron-gamma materials monitoring

NASA Astrophysics Data System (ADS)

Pekarskii, G. Sh.

1985-07-01

The physical principles of secondary radiation methods in nondestructive testing are discussed. Among the techniques considered are: neutron activation analysis (NAA); the induced-radiation method; and quasialbedo recording of secondary gamma-radiation. Emphasis is given to the neutron-gamma method which consists of exposing test material to a neutron flux and recording the secondary gamma-radiation by means of a spectrometer. The limitations of the method in detecting local inhomogeneous defects (filled pores cracks, and inclusions) in metal layers and multicomponents materials are described, and some advantages of the method over NAA are discussed. Formulas are derived for estimating the optimum density of the gamma-ray flux which is received by the detector.

Transportable, Low-Dose Active Fast-Neutron Imaging

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

Mihalczo, John T.; Wright, Michael C.; McConchie, Seth M.

2017-08-01

This document contains a description of the method of transportable, low-dose active fast-neutron imaging as developed by ORNL. The discussion begins with the technique and instrumentation and continues with the image reconstruction and analysis. The analysis discussion includes an example of how a gap smaller than the neutron production spot size and detector size can be detected and characterized depending upon the measurement time.

Determination of the composition of HgCdTe oxide films by neutron activation analysis

NASA Astrophysics Data System (ADS)

Gnade, B.; Simmons, A.; Little, D.; Strong, R.

1987-04-01

The composition of HgCdTe oxides grown by anodic oxidation in a standard KOH/ethylene glycol solution has been determined by neutron activation analysis (NAA). This technique is not hindered by the difficulties normally associated with methods using ion beams or electron beams. Neutron activation analysis has the advantage of being quantitative, and also NAA is not affected by the chemical composition of the matrix. The analysis of the KOH/ethylene glycol oxide film by neutron activation yields Hg:Cd:Te ratios of 0.534:0.19:1, in close agreement with Rutherford backscattering spectroscopy analysis (R.L. Strong et al., J. Vac. Sci. Technol. A4 (4) (1986) 1992).

Sulfur activation at the Little Boy-Comet Critical Assembly: a replica of the Hiroshima bomb

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

Kerr, G.D.; Emery, J.F.; Pace, J.V. III

1985-04-01

Studies have been completed on the activation of sulfur by fast neutrons from the Little Boy-Comet Critical Assembly which replicates the general features of the Hiroshima bomb. The complex effects of the bomb's design and construction on leakage of sulfur-activation neutrons were investigated both experimentally and theoretically. Our sulfur activation studies were performed as part of a larger program to provide benchmark data for testing of methods used in recent source-term calculations for the Hiroshima bomb. Source neutrons capable of activating sulfur play an important role in determining neutron doses in Hiroshima at a kilometer or more from the pointmore » of explosion. 37 refs., 5 figs., 6 tabs.« less

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

Evaluation of radioisotope tracer and activation analysis techniques for contamination monitoring in space environment simulation chambers

NASA Technical Reports Server (NTRS)

Smathers, J. B.; Kuykendall, W. E., Jr.; Wright, R. E., Jr.; Marshall, J. R.

1973-01-01

Radioisotope measurement techniques and neutron activation analysis are evaluated for use in identifying and locating contamination sources in space environment simulation chambers. The alpha range method allows the determination of total contaminant concentration in vapor state and condensate state. A Cf-252 neutron activation analysis system for detecting oils and greases tagged with stable elements is described. While neutron activation analysis of tagged contaminants offers specificity, an on-site system is extremely costly to implement and provides only marginal detection sensitivity under even the most favorable conditions.

A nuclear method to authenticate Buddha images

NASA Astrophysics Data System (ADS)

Khaweerat, S.; Ratanatongchai, W.; Channuie, J.; Wonglee, S.; Picha, R.; Promping, J.; Silva, K.; Liamsuwan, T.

2015-05-01

The value of Buddha images in Thailand varies dramatically depending on authentication and provenance. In general, people use their individual skills to make the justification which frequently leads to obscurity, deception and illegal activities. Here, we propose two non-destructive techniques of neutron radiography (NR) and neutron activation autoradiography (NAAR) to reveal respectively structural and elemental profiles of small Buddha images. For NR, a thermal neutron flux of 105 n cm-2s-1 was applied. NAAR needed a higher neutron flux of 1012 n cm-2 s-1 to activate the samples. Results from NR and NAAR revealed unique characteristic of the samples. Similarity of the profile played a key role in the classification of the samples. The results provided visual evidence to enhance the reliability of authenticity approval. The method can be further developed for routine practice which impact thousands of customers in Thailand.

Trace elements study of high purity nanocrystalline silicon carbide (3C-SiC) using k0-INAA method

NASA Astrophysics Data System (ADS)

Huseynov, Elchin; Jazbec, Anze

2017-07-01

Silicon carbide (3C-SiC) nanoparticles have been irradiated by neutron flux (2×1013 n·cm-2·s-1) at TRIGA Mark II type research reactor. After neutron irradiation, the radioisotopes of trace elements in the nanocrystalline 3C-SiC were studied as time functions. The identification of isotopes which significantly increased the activity of the samples as a result of neutron radiation was carried out. Nanocrystalline 3C-SiC are synthesized by standard laser technique and the purity of samples was determined by the k0-based Instrumental Neutron Activation Analysis (k0-INAA) method. Trace elements concentration in the 3C-SiC nanoparticles were determined by the radionuclides of appropriate elements. The trace element isotopes concentration have been calculated in percentage according to k0-INAA method.

Absorption of Thermal Neutrons in Uranium

DOE R&D Accomplishments Database

Creutz, E. C.; Wilson, R. R.; Wigner, E. P.

1941-09-26

A knowledge of the absorption processes for neutrons in uranium is important for planning a chain reaction experiment. The absorption of thermal neutrons in uranium and uranium oxide has been studied. Neutrons from the cyclotron were slowed down by passage through a graphite block. A uranium or uranium oxide sphere was placed at various positions in the block. The neutron intensity at different points in the sphere and in the graphite was measured by observing the activity induced in detectors or uranium oxide or manganese. It was found that both the fission activity in the uranium oxide and the activity induced in manganese was affected by non-thermal neutrons. An experimental correction for such effects was made by making measurements with the detectors surrounded by cadmium. After such corrections the results from three methods of procedure with the uranium oxide detectors and from the manganese detectors were consistent to within a few per cent.

Using the nuclear activation AMS method for determining chlorine in solids at ppb-levels and below

NASA Astrophysics Data System (ADS)

Winkler, Stephan R.; Eigl, Rosmarie; Forstner, Oliver; Martschini, Martin; Steier, Peter; Sterba, Johannes H.; Golser, Robin

2015-10-01

Neutron activation analysis using decay counting of the activated element is a well-established method in elemental analysis. However, for chlorine there is a better alternative to measuring decay of the short-lived activation product chlorine-38 (t1/2 = 37.24 min) - accelerator mass spectrometry (AMS) of 36Cl: the relatively high neutron capture cross section of chlorine-35 for thermal neutrons (43.7 b) and combined the AMS technique for chlorine-36 (t1/2 = 301 ka) allow for determination of chlorine down to ppb-levels using practical sample sizes and common exposure durations. The combination of neutron activation and AMS can be employed for a few other elements (nitrogen, thorium, and uranium) as well. For bulk solid samples an advantage of the method is that lab contamination can be rendered irrelevant. The chlorine-35 in the sample is activated to chlorine-36, and surface chlorine can be removed after the irradiation. Subsequent laboratory contamination, however, will not carry a prominent chlorine-36 signature. After sample dissolution and addition of sufficient amounts of stable chlorine carrier the produced chlorine-36 and thus the original chlorine-35 of the sample can be determined using AMS. We have developed and applied the method for analysis of chlorine in steel samples. The chlorine content of steel is of interest to nuclear industry, precisely because of above mentioned high neutron capture cross section for chlorine-35, which leads to accumulation of chlorine-36 as long-term nuclear waste. The samples were irradiated at the TRIGA Mark II reactor of the Atominstitut in Vienna and the 36Cl-AMS setup at the Vienna Environmental Research Accelerator (VERA) was used for 36Cl/Cl analysis.

Development of a technique using MCNPX code for determination of nitrogen content of explosive materials using prompt gamma neutron activation analysis method

NASA Astrophysics Data System (ADS)

Nasrabadi, M. N.; Bakhshi, F.; Jalali, M.; Mohammadi, A.

2011-12-01

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 10.8 MeV following radioactive neutron capture by 14N nuclei. We aimed to study the feasibility of using field-portable prompt gamma neutron activation analysis (PGNAA) along with improved nuclear equipment to detect and identify explosives, illicit substances or landmines. A 252Cf radio-isotopic source was embedded in a cylinder made of high-density polyethylene (HDPE) and the cylinder was then placed in another cylindrical container filled with water. Measurements were performed on high nitrogen content compounds such as melamine (C3H6N6). Melamine powder in a HDPE bottle was placed underneath the vessel containing water and the neutron source. Gamma rays were detected using two NaI(Tl) crystals. The results were simulated with MCNP4c code calculations. The theoretical calculations and experimental measurements were in good agreement indicating that this method can be used for detection of explosives and illicit drugs.

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

Application of activation methods on the Dubna experimental transmutation set-ups.

PubMed

Stoulos, S; Fragopoulou, M; Adloff, J C; Debeauvais, M; Brandt, R; Westmeier, W; Krivopustov, M; Sosnin, A; Papastefanou, C; Zamani, M; Manolopoulou, M

2003-02-01

High spallation neutron fluxes were produced by irradiating massive heavy targets with proton beams in the GeV range. The experiments were performed at the Dubna High Energy Laboratory using the nuclotron accelerator. Two different experimental set-ups were used to produce neutron spectra convenient for transmutation of radioactive waste by (n,x) reactions. By a theoretical analysis neutron spectra can be reproduced from activation measurements. Thermal-epithermal and fast-super-fast neutron fluxes were estimated using the 197Au, 238U (n,gamma) and (n,2n) reactions, respectively. Depleted uranium transmutation rates were also studied in both experiments.

Total body nitrogen analysis. [neutron activation analysis

NASA Technical Reports Server (NTRS)

Palmer, H. E.

1975-01-01

Studies of two potential in vivo neutron activation methods for determining total and partial body nitrogen in animals and humans are described. A method using the CO-11 in the expired air as a measure of nitrogen content was found to be adequate for small animals such as rats, but inadequate for human measurements due to a slow excretion rate. Studies on the method of measuring the induced N-13 in the body show that with further development, this method should be adequate for measuring muscle mass changes occurring in animals or humans during space flight.

Normalization of a collimated 14.7 MeV neutron source in a neutron spectrometry system for benchmark experiments

NASA Astrophysics Data System (ADS)

Ofek, R.; Tsechanski, A.; Shani, G.

1988-05-01

In the present study a method used to normalize a collimated 14.7 MeV neutron beam is introduced. It combined a measurement of the fast neutron scalar flux passing through the collimator, using a copper foil activation, with a neutron transport calculation of the foil activation per unit source neutron, carried out by the discrete-ordinates transport code DOT 4.2. The geometry of the collimated neutron beam is composed of a D-T neutron source positioned 30 cm in front of a 6 cm diameter collimator, through a 120 cm thick paraffin wall. The neutron flux emitted from the D-T source was counted by an NE-213 scintillator, simultaneously with the irradiation of the copper foil. Thus, the determination of the normalization factor of the D-T source is used for an absolute flux calibration of the NE-213 scintillator. The major contributions to the uncertainty in the determination of the normalization factor, and their origins, are discussed.

Ex-vessel neutron dosimetry analysis for westinghouse 4-loop XL pressurized water reactor plant using the RadTrack{sup TM} Code System with the 3D parallel discrete ordinates code RAPTOR-M3G

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

Chen, J.; Alpan, F. A.; Fischer, G.A.

2011-07-01

Traditional two-dimensional (2D)/one-dimensional (1D) SYNTHESIS methodology has been widely used to calculate fast neutron (>1.0 MeV) fluence exposure to reactor pressure vessel in the belt-line region. However, it is expected that this methodology cannot provide accurate fast neutron fluence calculation at elevations far above or below the active core region. A three-dimensional (3D) parallel discrete ordinates calculation for ex-vessel neutron dosimetry on a Westinghouse 4-Loop XL Pressurized Water Reactor has been done. It shows good agreement between the calculated results and measured results. Furthermore, the results show very different fast neutron flux values at some of the former plate locationsmore » and elevations above and below an active core than those calculated by a 2D/1D SYNTHESIS method. This indicates that for certain irregular reactor internal structures, where the fast neutron flux has a very strong local effect, it is required to use a 3D transport method to calculate accurate fast neutron exposure. (authors)« less

Nuclear instrumentation in VENUS-F

NASA Astrophysics Data System (ADS)

Wagemans, J.; Borms, L.; Kochetkov, A.; Krása, A.; Van Grieken, C.; Vittiglio, G.

2018-01-01

VENUS-F is a fast zero power reactor with 30 wt% U fuel and Pb/Bi as a coolant simulator. Depending on the experimental configuration, various neutron spectra (fast, epithermal, and thermal islands) are present. This paper gives a review of the nuclear instrumentation that is applied for reactor control and in a large variety of physics experiments. Activation foils and fission chambers are used to measure spatial neutron flux profiles, spectrum indices, reactivity effects (with positive period and compensation method or the MSM method) and kinetic parameters (with the Rossi-alpha method). Fission chamber calibrations are performed in the standard irradiation fields of the BR1 reactor (prompt fission neutron spectrum and Maxwellian thermal neutron spectrum).

A new irradiation method with a neutron filter for silicon neutron transmutation doping at the Japan research reactor no. 3 (JRR-3).

PubMed

Komeda, Masao; Kawasaki, Kozo; Obara, Toru

2013-04-01

We studied a new silicon irradiation holder with a neutron filter designed to make the vertical neutron flux profile uniform. Since an irradiation holder has to be made of a low activation material, we applied aluminum blended with B4C as the holder material. Irradiation methods to achieve uniform flux with a filter are discussed using Monte-Carlo calculation code MVP. Validation of the use of the MVP code for the holder's analyses is also discussed via characteristic experiments. Copyright © 2013 Elsevier Ltd. All rights reserved.

Progress in obtaining an absolute calibration of a total deuterium-tritium neutron yield diagnostic based on copper activation.

PubMed

Ruiz, C L; Chandler, G A; Cooper, G W; Fehl, D L; Hahn, K D; Leeper, R J; McWatters, B R; Nelson, A J; Smelser, R M; Snow, C S; Torres, J A

2012-10-01

The 350-keV Cockroft-Walton accelerator at Sandia National laboratory's Ion Beam facility is being used to calibrate absolutely a total DT neutron yield diagnostic based on the (63)Cu(n,2n)(62)Cu(β+) reaction. These investigations have led to first-order uncertainties approaching 5% or better. The experiments employ the associated-particle technique. Deuterons at 175 keV impinge a 2.6 μm thick erbium tritide target producing 14.1 MeV neutrons from the T(d,n)(4)He reaction. The alpha particles emitted are measured at two angles relative to the beam direction and used to infer the neutron flux on a copper sample. The induced (62)Cu activity is then measured and related to the neutron flux. This method is known as the F-factor technique. Description of the associated-particle method, copper sample geometries employed, and the present estimates of the uncertainties to the F-factor obtained are given.

Neutron flux characterization of californium-252 Neutron Research Facility at the University of Texas - Pan American by nuclear analytical technique

NASA Astrophysics Data System (ADS)

Wahid, Kareem; Sanchez, Patrick; Hannan, Mohammad

2014-03-01

In the field of nuclear science, neutron flux is an intrinsic property of nuclear reaction facilities that is the basis for experimental irradiation calculations and analysis. In the Rio Grande Valley (Texas), the UTPA Neutron Research Facility (NRF) is currently the only neutron facility available for experimental research purposes. The facility is comprised of a 20-microgram californium-252 neutron source surrounded by a shielding cascade containing different irradiation cavities. Thermal and fast neutron flux values for the UTPA NRF have yet to be fully investigated and may be of particular interest to biomedical studies in low neutron dose applications. Though a variety of techniques exist for the characterization of neutron flux, neutron activation analysis (NAA) of metal and nonmetal foils is a commonly utilized experimental method because of its detection sensitivity and availability. The aim of our current investigation is to employ foil activation in the determination of neutron flux values for the UTPA NSRF for further research purposes. Neutron spectrum unfolding of the acquired experimental data via specialized software and subsequent comparison for consistency with computational models lends confidence to the results.

Measurement of Activation Cross Sections Producing Short-Lived Nuclei with Pulsed Neutron Beam

NASA Astrophysics Data System (ADS)

Shimizu, Toshiaki; Arakita, Kazumasa; Miyazaki, Itaru; Shibata, Michihiro; Kawade, Kiyoshi; Hori, Jun-ichi; Ochiai, Kentaro; Nishitani, Takeo

2005-05-01

Activation cross sections for the (n, n') reaction were measured by means of the activation method at the neutron energies of 3.1 and 2.54 MeV by using a pulsed neutron beam. Target nuclei were 79Br, 90Zr, 197Au, and 207Pb, whose half-lives were between 0.8 and 8 s. The cross section for the 90Zr (n, n') 90mZr reaction was obtained for the first time in this energy range. The d-D neutrons were generated by bombarding a deuterated titanium target with a 350-keV d+ beam at the 80-degree beam line of the Fusion Neutronics Source at the Japan Atomic Energy Research Institute. In order to obtain reliable activation cross sections, careful attention was paid to correct the efficiency for a volume source, and the self-absorption of gamma rays in an irradiated sample. The systematics of the (n, n') reaction at the neutron energy of 3.1 MeV, which could be predicted within an accuracy of 50%, was proposed on the basis of our data.

Neutron activation analysis: trends in developments and applications

NASA Astrophysics Data System (ADS)

de Goeij, J. J.; Bode, P.

1995-03-01

New developments in instrumentation for, and methodology of, Instrumental Neutron Activation Analysis (INAA) may lead to new niches for this method of elemental analysis. This paper describes the possibilities of advanced detectors, automated irradiation and counting stations, and very large sample analysis. An overview is given of some typical new fields of application.

Application of the Monte Carlo method to estimate doses due to neutron activation of different materials in a nuclear reactor

NASA Astrophysics Data System (ADS)

Ródenas, José

2017-11-01

All materials exposed to some neutron flux can be activated independently of the kind of the neutron source. In this study, a nuclear reactor has been considered as neutron source. In particular, the activation of control rods in a BWR is studied to obtain the doses produced around the storage pool for irradiated fuel of the plant when control rods are withdrawn from the reactor and installed into this pool. It is very important to calculate these doses because they can affect to plant workers in the area. The MCNP code based on the Monte Carlo method has been applied to simulate activation reactions produced in the control rods inserted into the reactor. Obtained activities are introduced as input into another MC model to estimate doses produced by them. The comparison of simulation results with experimental measurements allows the validation of developed models. The developed MC models have been also applied to simulate the activation of other materials, such as components of a stainless steel sample introduced into a training reactors. These models, once validated, can be applied to other situations and materials where a neutron flux can be found, not only nuclear reactors. For instance, activation analysis with an Am-Be source, neutrography techniques in both medical applications and non-destructive analysis of materials, civil engineering applications using a Troxler, analysis of materials in decommissioning of nuclear power plants, etc.

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

Bushuev, A. V.; Kozhin, A. F., E-mail: alexfkozhin@yandex.ru; Aleeva, T. B.

An active neutron method for measuring the residual mass of {sup 235}U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual {sup 235}U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of {sup 238}U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be usedmore » in experiments where the studied object needs to be irradiated with a uniform fluence.« less

Anisotropy of the neutron fluence from a plasma focus.

NASA Technical Reports Server (NTRS)

Lee, J. H.; Shomo, L. P.; Kim, K. H.

1972-01-01

The fluence of neutrons from a plasma focus was measured by gamma spectrometry of an activated silver target. This method results in a significant increase in accuracy over the beta-counting method. Multiple detectors were used in order to measure the anisotropy of the fluence of neutrons. The fluence was found to be concentrated in a cone with a half-angle of 30 deg about the axis, and to drop off rapidly outside of this cone; the anisotropy was found to depend upon the total yield of neutrons. This dependence was strongest on the axis. Neither the axial concentration of the fluence of neutrons nor its dependence on the total yield of neutrons is explained by any of the currently proposed models. Some other explanations, including the possibility of an axially distributed source, are considered.

Improvement of non-destructive fissile mass assays in α low-level waste drums: A matrix correction method based on neutron capture gamma-rays and a neutron generator

NASA Astrophysics Data System (ADS)

Jallu, F.; Loche, F.

2008-08-01

Within the framework of radioactive waste control, non-destructive assay (NDA) methods may be employed. The active neutron interrogation (ANI) method is now well-known and effective in quantifying low α-activity fissile masses (mainly 235U, 239Pu, 241Pu) with low densities, i.e. less than about 0.4, in radioactive waste drums of volumes up to 200 l. The PROMpt Epithermal and THErmal interrogation Experiment (PROMETHEE [F. Jallu, A. Mariani, C. Passard, A.-C. Raoux, H. Toubon, Alpha low level waste control: improvement of the PROMETHEE 6 assay system performances. Nucl. Technol. 153 (January) (2006); C. Passard, A. Mariani, F. Jallu, J. Romeyer-Dherber, H. Recroix, M. Rodriguez, J. Loridon, C. Denis, PROMETHEE: an alpha low level waste assay system using passive and active neutron measurement methods. Nucl. Technol. 140 (December) (2002) 303-314]) based on ANI has been under development since 1996 to reach the incinerating α low level waste (LLW) criterion of about 50 Bq[α] per gram of crude waste (≈50 μg Pu) in 118 l drums on the date the drums are conditioned. Difficulties arise when dealing with matrices containing neutron energy moderators such as H and neutron absorbents such as Cl. These components may have a great influence on the fissile mass deduced from the neutron signal measured by ANI. For example, the calibration coefficient measured in a 118 l drum containing a cellulose matrix (density d = 0.144 g cm -3) may be 50 times higher than that obtained in a poly-vinyl-chloride matrix ( d = 0.253 g cm -3). Without any information on the matrix, the fissile mass is often overestimated due to safety procedures and by considering the most disadvantageous calibration coefficient corresponding to the most absorbing and moderating calibration matrix. The work discussed in this paper was performed at the CEA Nuclear Measurement Laboratory in France. It concerns the development of a matrix effect correction method, which consists in identifying and quantifying the matrix components by using prompt gamma-rays following neutron capture. The method aims to refine the value of the adequate calibration coefficient used for ANI analysis. This paper presents the final results obtained for 118 l waste drums with low α-activity and low density. This paper discusses the experimental and modelling studies and describes the development of correction abacuses based on gamma-ray spectrometry signals.

Neutron detector using sol-gel absorber

DOEpatents

Hiller, John M.; Wallace, Steven A.; Dai, Sheng

1999-01-01

An neutron detector composed of fissionable material having ions of lithium, uranium, thorium, plutonium, or neptunium, contained within a glass film fabricated using a sol-gel method combined with a particle detector is disclosed. When the glass film is bombarded with neutrons, the fissionable material emits fission particles and electrons. Prompt emitting activated elements yielding a high energy electron contained within a sol-gel glass film in combination with a particle detector is also disclosed. The emissions resulting from neutron bombardment can then be detected using standard UV and particle detection methods well known in the art, such as microchannel plates, channeltrons, and silicon avalanche photodiodes.

Early clinical experience utilizing scintillator with optical fiber (SOF) detector in clinical boron neutron capture therapy: its issues and solutions.

PubMed

Ishikawa, Masayori; Yamamoto, Tetsuya; Matsumura, Akira; Hiratsuka, Junichi; Miyatake, Shin-Ichi; Kato, Itsuro; Sakurai, Yoshinori; Kumada, Hiroaki; Shrestha, Shubhechha J; Ono, Koji

2016-08-09

Real-time measurement of thermal neutrons in the tumor region is essential for proper evaluation of the absorbed dose in boron neutron capture therapy (BNCT) treatment. The gold wire activation method has been routinely used to measure the neutron flux distribution in BNCT irradiation, but a real-time measurement using gold wire is not possible. To overcome this issue, the scintillator with optical fiber (SOF) detector has been developed. The purpose of this study is to demonstrate the feasibility of the SOF detector as a real-time thermal neutron monitor in clinical BNCT treatment and also to report issues in the use of SOF detectors in clinical practice and their solutions. Clinical measurements using the SOF detector were carried out in 16 BNCT clinical trial patients from December 2002 until end of 2006 at the Japanese Atomic Energy Agency (JAEA) and Kyoto University Research Reactor Institute (KURRI). The SOF detector worked effectively as a real-time thermal neutron monitor. The neutron fluence obtained by the gold wire activation method was found to differ from that obtained by the SOF detector. The neutron fluence obtained by the SOF detector was in better agreement with the expected fluence than with gold wire activation. The estimation error for the SOF detector was small in comparison to the gold wire measurement. In addition, real-time monitoring suggested that the neutron flux distribution and intensity at the region of interest (ROI) may vary due to the reactor condition, patient motion and dislocation of the SOF detector. Clinical measurements using the SOF detector to measure thermal neutron flux during BNCT confirmed that SOF detectors are effective as a real-time thermal neutron monitor. To minimize the estimation error due to the displacement of the SOF probe during treatment, a loop-type SOF probe was developed.

Comparative analysis of non-destructive methods to control fissile materials in large-size containers

NASA Astrophysics Data System (ADS)

Batyaev, V. F.; Sklyarov, S. V.

2017-09-01

The analysis of various non-destructive methods to control fissile materials (FM) in large-size containers filled with radioactive waste (RAW) has been carried out. The difficulty of applying passive gamma-neutron monitoring FM in large containers filled with concreted RAW is shown. Selection of an active non-destructive assay technique depends on the container contents; and in case of a concrete or iron matrix with very low activity and low activity RAW the neutron radiation method appears to be more preferable as compared with the photonuclear one. Note to the reader: the pdf file has been changed on September 22, 2017.

Improved Fission Neutron Data Base for Active Interrogation of Actinides

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

Pozzi, Sara; Czirr, J. Bart; Haight, Robert

2013-11-06

This project will develop an innovative neutron detection system for active interrogation measurements. Many active interrogation methods to detect fissionable material are based on the detection of neutrons from fission induced by fast neutrons or high-energy gamma rays. The energy spectrum of the fission neutrons provides data to identify the fissionable isotopes and materials such as shielding between the fissionable material and the detector. The proposed path for the project is as follows. First, the team will develop new neutron detection systems and algorithms by Monte Carlo simulations and bench-top experiments. Next, They will characterize and calibrate detection systems bothmore » with monoenergetic and white neutron sources. Finally, high-fidelity measurements of neutron emission from fissions induced by fast neutrons will be performed. Several existing fission chambers containing U-235, Pu-239, U-238, or Th-232 will be used to measure the neutron-induced fission neutron emission spectra. The challenge for making confident measurements is the detection of neutrons in the energy ranges of 0.01 – 1 MeV and above 8 MeV, regions where the basic data on the neutron energy spectrum emitted from fission is least well known. In addition, improvements in the specificity of neutron detectors are required throughout the complete energy range: they must be able to clearly distinguish neutrons from other radiations, in particular gamma rays and cosmic rays. The team believes that all of these challenges can be addressed successfully with emerging technologies under development by this collaboration. In particular, the collaboration will address the area of fission neutron emission spectra for isotopes of interest in the advanced fuel cycle initiative (AFCI).« less

Active-Interrogation Measurements of Induced-Fission Neutrons from Low-Enriched Uranium

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

J. L. Dolan; M. J. Marcath; M. Flaska

2012-07-01

Protection and control of nuclear fuels is paramount for nuclear security and safeguards; therefore, it is important to develop fast and robust controlling mechanisms to ensure the safety of nuclear fuels. Through both passive- and active-interrogation methods we can use fast-neutron detection to perform real-time measurements of fission neutrons for process monitoring. Active interrogation allows us to use different ranges of incident neutron energy to probe for different isotopes of uranium. With fast-neutron detectors, such as organic liquid scintillation detectors, we can detect the induced-fission neutrons and photons and work towards quantifying a sample’s mass and enrichment. Using MCNPX-PoliMi, amore » system was designed to measure induced-fission neutrons from U-235 and U-238. Measurements were then performed in the summer of 2010 at the Joint Research Centre in Ispra, Italy. Fissions were induced with an associated particle D-T generator and an isotopic Am-Li source. The fission neutrons, as well as neutrons from (n, 2n) and (n, 3n) reactions, were measured with five 5” by 5” EJ-309 organic liquid scintillators. The D-T neutron generator was available as part of a measurement campaign in place by Padova University. The measurement and data-acquisition systems were developed at the University of Michigan utilizing a CAEN V1720 digitizer and pulse-shape discrimination algorithms to differentiate neutron and photon detections. Low-enriched uranium samples of varying mass and enrichment were interrogated. Acquired time-of-flight curves and cross-correlation curves are currently analyzed to draw relationships between detected neutrons and sample mass and enrichment. In the full paper, the promise of active-interrogation measurements and fast-neutron detection will be assessed through the example of this proof-of-concept measurement campaign. Additionally, MCNPX-PoliMi simulation results will be compared to the measured data to validate the MCNPX-PoliMi code when used for active-interrogation simulations.« less

Neutron generator flux enhancement techniques and construction of a more efficient neutron detector for borehole-logging gamma-ray spectroscopy

NASA Astrophysics Data System (ADS)

Ghias, Asghar

1999-11-01

Neutron activation methods and bore-hole gamma-ray spectrometry have been versatile techniques for real time field evaluation in mineral exploration. The most common neutron generators producing 14 MeV and 2.5 MeV neutrons accelerate deuterium ions into a tritium or deuterium target via the 3H( 2H,n)4He or the 2H(2H,n) 3H reactions. The development and design of bore-hole 2.5 MeV high flux neutron generator coupled with an efficient gamma-ray detector is the primary focus of this work, which is needed by the coal and petroleum industries. A 2.5 MeV neutron generator, which used the D(D,n)T reaction, was constructed similar to a conventional Zetatron 14 MeV generator. The performance of the low energy neutron generator was studied under various operating conditions. In order to enhance the neutron flux of the generator, an r.f. field was applied to the ion source which increased the neutron yield per pulse by about thirty percent. A theoretical study of the r.f enhancement has been made to explain the operation of the r.f. added Zetatron tube. An alternative, method of neutron flux enhancement by use of laser-excitation is discussed and explained theoretically. The laser technique although not experimentally verified, is based on the recent development of vibronic lasers, the neutron flux can be enhanced several orders of magnitude by precise tuning of the wavelength within vibronic band. Activation experiments using a large coal sample (about I ton) were conducted, and studies were made on inter and intrapulse counting, detector gated spectra, and comparison of the spectra using different neutron sources. Preliminary results on coal analysis reveal that lower energy (2.5 MeV) is superior to high energy (14 MeV) neutrons. During the course of this work it became necessary to measure fast neutrons, efficiently and in real time. A new type of detector was consequently developed using SnO2 as sheath material around a BGO detector to measure the capture gamma-rays of oxygen. Using neutron activation studies of coal, the feasibility of applying the technique to aid medical diagnostics is also discussed in this dissertation.

A Comparison of Neutron-Based Non-Destructive Assessment Methods for Chemical Warfare Materiel and High Explosives

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

E.H. Seabury; D.L. Chichester; C.J. Wharton

2008-08-01

Prompt Gamma Neutron Activation Analysis (PGNAA) systems employ neutrons as a probe to interrogate items, e.g. chemical warfare materiel-filled munitions. The choice of a neutron source in field-portable systems is determined by its ability to excite nuclei of interest, operational concerns such as radiological safety and ease-of-use, and cost. Idaho National Laboratory’s PINS Chemical Assay System has traditionally used a Cf-252 isotopic neutron source, but recently a Deuterium-Tritium (DT) Electronic Neutron Generator (ENG) has been tested as an alternate neutron source. This paper presents the results of using both of these neutron sources to interrogate chemical warfare materiel (CWM) andmore » high explosive (HE) filled munitions.« less

A Comparison of Neutron-Based Non-Destructive Assessment Methods for Chemical Warfare Material and High Explosives

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

Seabury, E. H.; Chichester, D. L.; Wharton, C. J.

2009-03-10

Prompt Gamma Neutron Activation Analysis (PGNAA) systems employ neutrons as a probe to interrogate items, e.g. chemical warfare materiel-filled munitions. The choice of a neutron source in field-portable systems is determined by its ability to excite nuclei of interest, operational concerns such as radiological safety and ease-of-use, and cost. Idaho National Laboratory's PINS Chemical Assay System has traditionally used a {sup 252}Cf isotopic neutron source, but recently a deuterium-tritium (DT) electronic neutron generator (ENG) has been tested as an alternate neutron source. This paper presents the results of using both of these neutron sources to interrogate chemical warfare materiel (CWM)more » and high explosive (HE) filled munitions.« 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.

Nuclear accident dosimetry intercomparison studies

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

Sims, C.S.

1989-09-01

Twenty-two nuclear accident dosimetry intercomparison studies utilizing the fast-pulse Health Physics Research Reactor at the Oak Ridge National Laboratory have been conducted since 1965. These studies have provided a total of 62 different organizations a forum for discussion of criticality accident dosimetry, an opportunity to test their neutron and gamma-ray dosimetry systems under a variety of simulated criticality accident conditions, and the experience of comparing results with reference dose values as well as with the measured results obtained by others making measurements under identical conditions. Sixty-nine nuclear accidents (27 with unmoderated neutron energy spectra and 42 with eight different shieldedmore » spectra) have been simulated in the studies. Neutron doses were in the 0.2-8.5 Gy range and gamma doses in the 0.1-2.0 Gy range. A total of 2,289 dose measurements (1,311 neutron, 978 gamma) were made during the intercomparisons. The primary methods of neutron dosimetry were activation foils, thermoluminescent dosimeters, and blood sodium activation. The main methods of gamma dose measurement were thermoluminescent dosimeters, radiophotoluminescent glass, and film. About 68% of the neutron measurements met the accuracy guidelines (+/- 25%) and about 52% of the gamma measurements met the accuracy criterion (+/- 20%) for accident dosimetry.« less

Gold nanoparticles production using reactor and cyclotron based methods in assessment of (196,198)Au production yields by (197)Au neutron absorption for therapeutic purposes.

PubMed

Khorshidi, Abdollah

2016-11-01

Medical nano-gold radioisotopes is produced regularly using high-flux nuclear reactors, and an accelerator-driven neutron activator can turn out higher yield of (197)Au(n,γ)(196,198)Au reactions. Here, nano-gold production via radiative/neutron capture was investigated using irradiated Tehran Research Reactor flux and also simulated proton beam of Karaj cyclotron in Iran. (197)Au nano-solution, including 20nm shaped spherical gold and water, was irradiated under Tehran reactor flux at 2.5E+13n/cm(2)/s for (196,198)Au activity and production yield estimations. Meanwhile, the yield was examined using 30MeV proton beam of Karaj cyclotron via simulated new neutron activator containing beryllium target, bismuth moderator around the target, and also PbF2 reflector enclosed the moderator region. Transmutation in (197)Au nano-solution samples were explored at 15 and 25cm distances from the target. The neutron flux behavior inside the water and bismuth moderators was investigated for nano-gold particles transmutation. The transport of fast neutrons inside bismuth material as heavy nuclei with a lesser lethargy can be contributed in enhanced nano-gold transmutation with long duration time than the water moderator in reactor-based method. Cyclotron-driven production of βeta-emitting radioisotopes for brachytherapy applications can complete the nano-gold production technology as a safer approach as compared to the reactor-based method. Copyright © 2016 Elsevier B.V. All rights reserved.

The application of selected radionuclides for monitoring of the D-D reactions produced by dense plasma-focus device.

PubMed

Jednorog, S; Szydlowski, A; Bienkowska, B; Prokopowicz, R

The dense plasma focus (DPF) device-DPF-1000U which is operated at the Institute of Plasma Physics and Laser Microfusion is the largest that type plasma experiment in the world. The plasma that is formed in large plasma experiments is characterized by vast numbers of parameters. All of them need to be monitored. A neutron activation method occupies a high position among others plasma diagnostic methods. The above method is off-line, remote, and an integrated one. The plasma which has enough temperature to bring about nuclear fusion reactions is always a strong source of neutrons that leave the reactions area and take along energy and important information on plasma parameters and properties as well. Silver as activated material is used as an effective way of neutrons measurement, especially when they are emitted in the form of short pulses like as it happens from the plasma produced in Dense Plasma-Focus devices. Other elements such as beryllium and yttrium are newly introduced and currently tested at the Institute of Plasma Physics and Laser Microfusion to use them in suitable activation neutron detectors. Some specially designed massive indium samples have been recently adopted for angular neutrons distribution measurements (vertical and horizontal) and have been used in the recent plasma experiment conducted on the DPF-1000U device. This choice was substantiated by relatively long half-lives of the neutron induced isotopes and the threshold character of the 115 In(n,n') 115m In nuclear reaction.

Spallation yield of neutrons produced in thick lead target bombarded with 250 MeV protons

NASA Astrophysics Data System (ADS)

Chen, L.; Ma, F.; Zhanga, X. Y.; Ju, Y. Q.; Zhang, H. B.; Ge, H. L.; Wang, J. G.; Zhou, B.; Li, Y. Y.; Xu, X. W.; Luo, P.; Yang, L.; Zhang, Y. B.; Li, J. Y.; Xu, J. K.; Liang, T. J.; Wang, S. L.; Yang, Y. W.; Gu, L.

2015-01-01

The neutron yield from thick target of Pb irradiated with 250 MeV protons has been studied experimentally. The neutron production was measured with the water-bath gold method. The thermal neutron distributions in the water were determined according to the measured activities of Au foils. Corresponding results calculated with the Monte Carlo code MCNPX were compared with the experimental data. It was found out that the Au foils with cadmium cover significantly changed the spacial distribution of the thermal neutron field. The corrected neutron yield was deduced to be 2.23 ± 0.19 n/proton by considering the influence of the Cd cover on the thermal neutron flux.

Determination of selected trace elements in foodstuffs and biological materials by destructive neutron activation analysis.

PubMed

Bayat, I; Etehadiyan, M; Ansar, M

1995-01-01

Concentration of trace elements in Nescafé, Fariman sugar, and Sadaf turmeric and mercury content in cancerous blood were determined by radiochemical, neutron activation analysis. By this separation method levels of 110mAg, 198Au, 203Hg, 76Se, 51Cr, 24Na, 42K, 99Mo, 122Sb, 82Br, 59Fe, 60Co were measured without interference in the gamma spectroscopy. A nondestructive method has also been used for the analysis of sodium, potassium, and bromine.

Monitoring of the Irradiated Neutron Fluence in the Neutron Transmutation Doping Process of Hanaro

NASA Astrophysics Data System (ADS)

Kim, Myong-Seop; Park, Sang-Jun

2009-08-01

Neutron transmutation doping (NTD) for silicon is a process of the creation of phosphorus impurities in intrinsic or extrinsic silicon by neutron irradiation to obtain silicon semiconductors with extremely uniform dopant distribution. HANARO has two vertical holes for the NTD, and the irradiation for 5 and 6 inch silicon ingots has been going on at one hole. In order to achieve the accurate neutron fluence corresponding to the target resistivity, the real time neutron flux is monitored by self-powered neutron detectors. After irradiation, the total irradiation fluence is confirmed by measuring the absolute activity of activation detectors. In this work, a neutron fluence monitoring method using zirconium foils with the mass of 10 ~ 50 mg was applied to the NTD process of HANARO. We determined the proportional constant of the relationship between the resistivity of the irradiated silicon and the neutron fluence determined by using zirconium foils. The determined constant for the initially n-type silicon was 3.126 × 1019 n·Ω/cm. It was confirmed that the difference between this empirical value and the theoretical one was only 0.5%. Conclusively, the practical methodology to perform the neutron transmutation doping of silicon was established.

Measured neutron and gamma spectra from californium-252 in a tissue-equivalent medium.

PubMed

Elson, H R; Stupar, T A; Shapiro, A; Kereiakes, J G

1979-01-01

A method of experimentally obtaining both neutron and gamma-ray spectra in a scattering medium is described. The method utilizes a liquid-organic scintillator (NE-213) coupled with a pulse-shape discrimination circuit. This allows the separation of the neutron-induced pulse-height data from the gamma-ray pulse-height data. Using mathematical unfolding techniques, the two sets of pulse-height data were transformed to obtain the neutron and gamma-ray energy spectra. A small spherical detector was designed and constructed to reduce the errors incurred by attempting spectral measurements in a scattering medium. Demonstration of the utility of the system to obtain the neutron and gamma-ray spectra in a scattering medium was performed by characterizing the neutron and gamma-ray spectra at various sites about a 3.7-microgram (1.5 cm active length) californium-252 source in a tissue-equivalent medium.

Neutron absorption detector

DOEpatents

Bell, Zane William [Oak Ridge, TN; Boatner, Lynn Allen [Oak Ridge, TN

2011-05-31

A method of detecting an activator, the method including impinging a receptor material that is not predominately water and lacks a photoluminescent material with an activator and generating Cherenkov effect light due to the activator impinging the receptor material. The method further including identifying a characteristic of the activator based on the light.

SU-F-T-154: An Evaluation and Quantification of Secondary Neutron Radiation Dose Due to Double Scatter and Pencil Beam Scanning Proton Therapy

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

Glick, A; Diffenderfer, E

2016-06-15

Proton radiation therapy can deliver high radiation doses to tumors while sparing normal tissue. However, protons yield secondary neutron and gamma radiation that is difficult to detect, small in comparison to the prescribed dose, and not accounted for in most treatment planning systems. The risk for secondary malignancies after proton therapy may be dependent on the quality of this dose. Consequently, there is interest in characterizing the secondary radiation. Previously, we used the dual ionization chamber method to measure the separate absorbed dose from gamma-rays and neutrons secondary to the proton beam1, relying on characterization of ionization chamber response inmore » the unknown neutron spectrum from Monte Carlo simulation. We developed a procedure to use Shieldwerx activation foils, with neutron activation energies ranging from 0.025 eV to 13.5 MeV, to measure the neutron energy spectrum from double scattering (DS) and pencil beam scanning (PBS) protons outside of the treatment volume in a water tank. The activated foils are transferred to a NaI well chamber for gamma-ray spectroscopy and activity measurement. Since PBS treats in layers, the switching time between layers is used to correct for the decay of the activated foils and the relative dose per layer is assumed to be proportional to the neutron fluence per layer. MATLAB code was developed to incorporate the layer delivery and switching time into a calculation of foil activity, which is then used to determine the neutron energy fluence from tabulated foil activation energy thresholds.1. Diffenderfer et. al., Med. Phys., 38(11) 2011.« less

Determination of elements in hospital waste with neutron activation analysis method

NASA Astrophysics Data System (ADS)

Dwijananti, P.; Astuti, B.; Alwiyah; Fianti

2018-03-01

The producer of the biggest B3 waste is hospital. The waste is from medical and laboratory activities. The purpose of this study is to determine the elements contained in the liquid waste from hospital and calculate the levels of these elements. This research was done by analysis of the neutron activation conducted at BATAN Yogyakarta. The neutron activation analysis is divided into two stages: activation of the samples using neutron sources of reactor Kartini, then chopping by using a set of tools, gamma spectrometer with HPGe detector. Qualitative and quantitative analysis were done by matching the gamma spectrum peak to the Neutron Activation Table. The sample was taken from four points of the liquid waste treatment plant (WWTP) Bhakti Wira Tamtama Semarang hospital. The results showed that the samples containing elements of Cr, Zn, Fe, Co, and Na, with the levels of each element is Cr (0.033 - 0.075) mg/L, Zn (0.090 - 1.048) mg/L, Fe (2.937-37.743) mg/L, Co (0.005-0.023) mg/L, and Na (61.088-116.330) mg/L. Comparing to the standard value, the liquid is safe to the environment.

Multi-element analysis of emeralds and associated rocks by k(o) neutron activation analysis

PubMed

Acharya; Mondal; Burte; Nair; Reddy; Reddy; Reddy; Manohar

2000-12-01

Multi-element analysis was carried out in natural emeralds, their associated rocks and one sample of beryl obtained from Rajasthan, India. The concentrations of 21 elements were assayed by Instrumental Neutron Activation Analysis using the k0 method (k0 INAA method) and high-resolution gamma ray spectrometry. The data reveal the segregation of some elements from associated (trapped and host) rocks to the mineral beryl forming the gemstones. A reference rock standard of the US Geological Survey (USGS BCR-1) was also analysed as a control of the method.

Determination of carrier yields for neutron activation analysis using energy dispersive X-ray spectrometry

USGS Publications Warehouse

Johnson, R.G.; Wandless, G.A.

1984-01-01

A new method is described for determining carrier yield in the radiochemical neutron activation analysis of rare-earth elements in silicate rocks by group separation. The method involves the determination of the rare-earth elements present in the carrier by means of energy-dispersive X-ray fluorescence analysis, eliminating the need to re-irradiate samples in a nuclear reactor after the gamma ray analysis is complete. Results from the analysis of USGS standards AGV-1 and BCR-1 compare favorably with those obtained using the conventional method. ?? 1984 Akade??miai Kiado??.

Neutron production by stopping 55 MeV deuterons in carbon and heavy water

NASA Astrophysics Data System (ADS)

Lhersonneau, G.; Malkiewicz, T.; Jones, P.; Ketelhut, S.; Trzaska, W. H.

2012-09-01

Neutron production by stopping 55 MeV deuterons in thick carbon and heavy-water targets has been measured by the activation method. The geometry was close to the one defined for the SPIRAL2 uranium-carbide target in the initial phase. A comparative method for obtaining the neutron flux has been used and is presented in detail. The neutron flux generated by 55 MeV deuterons on carbon is 2.3 times the flux at the deuteron energy of 40 MeV. The flux further increases by a factor 1.4 when using a heavy-water target. These results are discussed in the context of an energy upgrade of the SPIRAL2 driver accelerator.

Photonuclear activation of pure isotopic mediums.

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

Grohman, Mark A.; Lukosi, Eric Daniel

2010-06-01

This work simulated the response of idealized isotopic U-235, U-238, Th-232, and Pu-239 mediums to photonuclear activation with various photon energies. These simulations were conducted using MCNPX version 2.6.0. It was found that photon energies between 14-16 MeV produce the highest response with respect to neutron production rates from all photonuclear reactions. In all cases, Pu-239 responds the highest, followed by U-238. Th-232 produces more overall neutrons at lower photon energies then U-235 when material thickness is above 3.943 centimeters. The time it takes each isotopic material to reach stable neutron production rates in time is directly proportional to themore » material thickness and stopping power of the medium, where thicker mediums take longer to reach stable neutron production rates and thinner media display a neutron production plateau effect, due to the lack of significant attenuation of the activating photons in the isotopic mediums. At this time, no neutron sensor system has time resolutions capable of verifying these simulations, but various indirect methods are possible and should be explored for verification of these results.« less

Characterization of the CALIBAN Critical Assembly Neutron Spectra using Several Adjustment Methods Based on Activation Foils Measurement

NASA Astrophysics Data System (ADS)

Casoli, Pierre; Grégoire, Gilles; Rousseau, Guillaume; Jacquet, Xavier; Authier, Nicolas

2016-02-01

CALIBAN is a metallic critical assembly managed by the Criticality, Neutron Science and Measurement Department located on the French CEA Center of Valduc. The reactor is extensively used for benchmark experiments dedicated to the evaluation of nuclear data, for electronic hardening or to study the effect of the neutrons on various materials. Therefore CALIBAN irradiation characteristics and especially its central cavity neutron spectrum have to be very accurately evaluated. In order to strengthen our knowledge of this spectrum, several adjustment methods based on activation foils measurements are being studied for a few years in the laboratory. Firstly two codes included in the UMG package have been tested and compared: MAXED and GRAVEL. More recently, the CALIBAN cavity spectrum has been studied using CALMAR, a new adjustment tool currently under development at the CEA Center of Cadarache. The article will discuss and compare the results and the quality of spectrum rebuilding obtained with the UMG codes and with the CALMAR software, from a set of activation measurements carried out in the CALIBAN irradiation cavity.

Proton Neutron Gamma-X Detection (PNGXD): An introduction to contrast agent detection during proton therapy via prompt gamma neutron activation

NASA Astrophysics Data System (ADS)

Gräfe, James L.

2017-09-01

Proton therapy is an alternative external beam cancer treatment modality to the conventional linear accelerator-based X-ray radiotherapy. An inherent by-product of proton-nuclear interactions is the production of secondary neutrons. These neutrons have long been thought of as a secondary contaminant, nuisance, and source of secondary cancer risk. In this paper, a method is proposed to use these neutrons to identify and localize the presence of the tumor through neutron capture reactions with the gadolinium-based MRI contrast agent. This could provide better confidence in tumor targeting by acting as an additional quality assurance tool of tumor position during treatment. This effectively results in a neutron induced nuclear medicine scan. Gadolinium (Gd), is an ideal candidate for this novel nuclear contrast imaging procedure due to its unique nuclear properties and its widespread use as a contrast agent in MRI. Gd has one of the largest thermal neutron capture cross sections of all the stable nuclides, and the gadolinium-based contrast agents localize in leaky tissues and tumors. Initial characteristics of this novel concept were explored using the Monte Carlo code MCNP6. The number of neutron capture reactions per Gy of proton dose was found to be approximately 50,000 neutron captures/Gy, for a 8 cm3 tumor containing 300 ppm Gd at 8 cm depth with a simple simulation designed to represent the active delivery method. Using the passive method it is estimated that this number can be up to an order of magnitude higher. The thermal neutron distribution was found to not be localized within the spread out Bragg peak (SOBP) for this geometrical configuration and therefore would not allow for the identification of a geometric miss of the tumor by the proton SOBP. However, this potential method combined with nuclear medicine imaging and fused with online CBCT and prior MRI or CT imaging could help to identify tumor position during treatment. More computational and experimental work are required to determine the feasibility of this new technique termed Proton Neutron Gamma-X Detection (PNGXD). The initial concept of this procedure is presented in this paper as well as future research directions.

Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors

DOEpatents

Caldwell, J.T.; Kunz, W.E.; Atencio, J.D.

1982-03-31

A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify /sup 233/U, /sup 235/U and /sup 239/Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as /sup 240/Pu, /sup 244/Cm and /sup 252/Cf, and the spontaneous alpha particle emitter /sup 241/Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether permanent low-level burial is appropriate for the waste sample.

Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors

DOEpatents

Caldwell, John T.; Kunz, Walter E.; Atencio, James D.

1984-01-01

A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify .sup.233 U, .sup.235 U and .sup.239 Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as .sup.240 Pu, .sup.244 Cm and .sup.252 Cf, and the spontaneous alpha particle emitter .sup.241 Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether "permanent" low-level burial is appropriate for the waste sample.

Stagnancy of the pygmy dipole resonance

NASA Astrophysics Data System (ADS)

Sun, Xu-Wei; Chen, Jing; Lu, Ding-Hui

2018-01-01

The pygmy dipole resonance (PDR) of nickel isotopes is studied using the deformed random phase approximation method. The isoscalar character of the pygmy resonance is confirmed, and the correlation between the pygmy resonance and neutron skin thickness is discussed. Our investigation shows a linear correlation between PDR integral cross section and neutron skin thickness when the excess neutrons lie in pf orbits, with a correlation rate of about 0.27 fm-1. However, in more neutron-rich nickel isotopes, the growth of the pygmy dipole resonance is stagnant. Although the neutron skin thickness increases, the whole skin is not active. There is an inertial part in the nuclei 70-78Ni which does not participate in the pygmy resonance actively and as a result, contributes little to the photo-absorption cross section. Supported by National Science Foundation of China

In field application of differential Die-Away time technique for detecting gram quantities of fissile materials

NASA Astrophysics Data System (ADS)

Remetti, Romolo; Gandolfo, Giada; Lepore, Luigi; Cherubini, Nadia

2017-10-01

In the frame of Chemical, Biological, Radiological, and Nuclear defense European activities, the ENEA, the Italian National Agency for New Technologies, Energy and Sustainable Economic Development, is proposing the Neutron Active Interrogation system (NAI), a device designed to find transuranic-based Radioactive Dispersal Devices hidden inside suspected packages. It is based on Differential Die-Away time Analysis, an active neutron technique targeted in revealing the presence of fissile material through detection of induced fission neutrons. Several Monte Carlo simulations, carried out by MCNPX code, and the development of ad-hoc design methods, have led to the realization of a first prototype based on a 14 MeV d-t neutron generator coupled with a tailored moderating structure, and an array of helium-3 neutron detectors. The complete system is characterized by easy transportability, light weight, and real-time response. First results have shown device's capability to detect gram quantities of fissile materials.

Identification of heavy metals on vegetables at the banks of Kaligarang river using neutron analysis activation method

NASA Astrophysics Data System (ADS)

Yulianti, D.; Marwoto, P.; Fianti

2018-03-01

This research aims to determine the type, concentration, and distribution of heavy metals in vegetables on the banks river Kaligarang using Neutron Analysis Activation (NAA) Method. The result is then compared to its predefined threshold. Vegetable samples included papaya leaf, cassava leaf, spinach, and water spinach. This research was conducted by taking a snippet of sediment and vegetation from 4 locations of Kaligarang river. These snippets are then prepared for further irradiated in the reactor for radioactive samples emiting γ-ray. The level of γ-ray energy determines the contained elements of sample that would be matched to Neutron Activation Table. The results showed that vegetablesat Kaligarang are containing Cr-50, Co-59, Zn-64, Fe-58, and Mn-25, and well distributed at all research locations. Furthermore, the level of the detected metal elements is less than the predefined threshold.

Neutron-induced 63Ni activity and microscopic observation of copper samples exposed to the Hiroshima atomic bomb

NASA Astrophysics Data System (ADS)

Shizuma, Kiyoshi; Endo, Satoru; Shinozaki, Kenji; Fukushima, Hiroshi

2013-05-01

Fast neutron activation data for 63Ni in copper samples exposed to the Hiroshima atomic bomb are important in evaluating neutron doses to the survivors. Up to until now, accelerator mass spectrometry and liquid scintillation counting methods have been applied in 63Ni measurements and data were accumulated within 1500 m from the hypocenter. The slope of the activation curve versus distance shows reasonable agreement with the calculation result, however, data near the hypocenter are scarce. In the present work, two copper samples obtained from the Atomic bomb dome (155 m from the hypocenter) and the Bank of Japan building (392 m) were utilized in 63Ni beta-ray measurement with a Si surface barrier detector. Additionally, microscopic observation of the metal surfaces was performed for the first time. Only upper limit of 63Ni production was obtained for copper sample of the Atomic bomb dome. The result of the 63Ni measurement for Bank of Japan building show reasonable agreement with the AMS measurement and to fast neutron activation calculations based on the Dosimetry System 2002 (DS02) neutrons.

The D-D Neutron Generator as an Alternative to Am(Li) Isotopic Neutron Source in the Active Well Coincidence Counter

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

McElroy, Robert Dennis; Cleveland, Steven L.

The 235U mass assay of bulk uranium items, such as oxide canisters, fuel pellets, and fuel assemblies, is not achievable by traditional gamma-ray assay techniques due to the limited penetration of the item by the characteristic 235U gamma rays. Instead, fast neutron interrogation methods such as active neutron coincidence counting must be used. For international safeguards applications, the most commonly used active neutron systems, the Active Well Coincidence Counter (AWCC), Uranium Neutron Collar (UNCL) and 252Cf Shuffler, rely on fast neutron interrogation using an isotopic neutron source [i.e., 252Cf or Am(Li)] to achieve better measurement accuracies than are possible usingmore » gamma-ray techniques for high-mass, high-density items. However, the Am(Li) sources required for the AWCC and UNCL systems are no longer manufactured, and newly produced systems rely on limited supplies of sources salvaged from disused instruments. The 252Cf shuffler systems rely on the use of high-output 252Cf sources, which while still available have become extremely costly for use in routine operations and require replacement every five to seven years. Lack of a suitable alternative neutron interrogation source would leave a potentially significant gap in the safeguarding of uranium processing facilities. In this work, we made use of Oak Ridge National Laboratory’s (ORNL’s) Large Volume Active Well Coincidence Counter (LV-AWCC) and a commercially available deuterium-deuterium (D-D) neutron generator to examine the potential of the D-D neutron generator as an alternative to the isotopic sources. We present the performance of the LV-AWCC with D-D generator for the assay of 235U based on the results of Monte Carlo N-Particle (MCNP) simulations and measurements of depleted uranium (DU), low enriched uranium (LEU), and highly enriched uranium (HEU) items.« less

The sciences and applications of the Electron LINAC-driven neutron source in Argentina

NASA Astrophysics Data System (ADS)

Granada, J. R.; Mayer, R. E.; Dawidowski, J.; Santisteban, J. R.; Cantargi, F.; Blostein, J. J.; Rodríguez Palomino, L. A.; Tartaglione, A.

2016-06-01

The Neutron Physics group at Centro Atómico Bariloche (CNEA, Argentina) has evolved for more than forty five years around a small 25MeV linear electron accelerator. It constitutes our compact accelerator-driven neutron source (CANS), which is dedicated to the use and development of neutronic methods to tackle problems of basic sciences and technological applications. Its historical first commitment has been the determination of the total cross sections of materials as a function of neutron energy by means of transmission experiments for thermal and sub-thermal neutrons. This also allowed testing theoretical models for the generation of scattering kernels and cross sections. Through the years, our interests moved from classic pulsed neutron diffraction, which included the development of high-precision methods for the determination of very low hydrogen content in metals, towards deep inelastic neutron scattering (DINS), a powerful tool for the determination of atomic momentum distribution in condensed matter. More recently non-intrusive techniques aimed at the scanning of large cargo containers have started to be developed with our CANS, testing the capacity and limitations to detect special nuclear material and dangerous substances. Also, the ever-present "bremsstrahlung" radiation has been recognized and tested as a useful complement to instrumental neutron activation, as it permits to detect other nuclear species through high-energy photon activation. The facility is also used for graduate and undergraduate students' experimental work within the frame of Instituto Balseiro Physics and Nuclear Engineering courses of study, and also MSc and PhD theses work.

Neutron activation determination of iridium, gold, platinum, and silver in geologic samples

USGS Publications Warehouse

Millard, H.T.

1987-01-01

Low-level methods for the determination of iridium and other noble metals have become increasingly important in recent years due to interest in locating abundance anomalies associated with the Cretaceous and Tertiary (K-T) boundary. Typical iridium anomalies are in the range of 1 to 100 ??g/kg (ppb). Thus methods with detection limits near 0.1 ??g/kg should be adequate to detect K-T boundary anomalies. Radiochemical neutron activation analysis methods continue to be required although instrumental neutron activation analysis techniques employing elaborate gamma-counters are under development. In the procedure developed in this study samples irradiated in the epithermal neutron facility of the U. S. Geological Survey TRIGA Reactor (Denver, Colorado) are treated with a mini-fire assay technique. The iridium, gold, and silver are collected in a 1-gram metallic lead button. Primary contaminants at this stage are arsenic and antimony. These can be removed by heating the button with a mixture of sodium perioxide and sodium hydroxide. The resulting 0.2-gram lead bead is counted in a Compton suppression spectrometer. Carrier yields are determined by reirradiation of the lead beads. This procedure has been applied to the U.S.G.S. Standard Rock PCC-1 and samples from K-T boundary sites in the Western Interior of North America. ?? 1987 Akade??miai Kiado??.

Study of dietary supplements compositions by neutron activation analysis at the VR-1 training reactor

NASA Astrophysics Data System (ADS)

Stefanik, Milan; Rataj, Jan; Huml, Ondrej; Sklenka, Lubomir

2017-11-01

The VR-1 training reactor operated by the Czech Technical University in Prague is utilized mainly for education of students and training of various reactor staff; however, R&D is also carried out at the reactor. The experimental instrumentation of the reactor can be used for the irradiation experiments and neutron activation analysis. In this paper, the neutron activation analysis (NAA) is used for a study of dietary supplements containing the zinc (one of the essential trace elements for the human body). This analysis includes the dietary supplement pills of different brands; each brand is represented by several different batches of pills. All pills were irradiated together with the standard activation etalons in the vertical channel of the VR-1 reactor at the nominal power (80 W). Activated samples were investigated by the nuclear gamma-ray spectrometry technique employing the semiconductor HPGe detector. From resulting saturated activities, the amount of mineral element (Zn) in the pills was determined using the comparative NAA method. The results show clearly that the VR-1 training reactor is utilizable for neutron activation analysis experiments.

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

Huml, O.

The objective of this work was to determine the neutron flux density distribution in various places of the training reactor VR-1 Sparrow. This experiment was performed on the new core design C1, composed of the new low-enriched uranium fuel cells IRT-4M (19.7 %). This fuel replaced the old high-enriched uranium fuel IRT-3M (36 %) within the framework of the RERTR Program in September 2005. The measurement used the neutron activation analysis method with gold wires. The principle of this method consists in neutron capture in a nucleus of the material forming the activation detector. This capture can change the nucleusmore » in a radioisotope, whose activity can be measured. The absorption cross-section values were evaluated by MCNP computer code. The gold wires were irradiated in seven different positions in the core C1. All irradiations were performed at reactor power level 1E8 (1 kW{sub therm}). The activity of segments of irradiated wires was measured by special automatic device called 'Drat' (Wire in English). (author)« less

Earthquake effects in thermal neutron variations at the high-altitude station of Northern

NASA Astrophysics Data System (ADS)

Antonova, Valentina; Chubenko, Alexandr; Kryukov, Sergey; Lutsenko, Vadim

2016-04-01

Results of study of thermal neutron variations under various space and geophysical conditions on the basis of measurements on stationary installations with high statistical accuracy are presented. Installations are located close to the fault of the earth's crust at the high-altitude station of cosmic rays (3340 m above sea level, 43.02 N, 76.56 E, 20 km from Almaty) in the mountains of Northern Tien-Shan. Responses of the most effective gelio- and geophysical events (variations of atmospheric pressure, coronal mass ejections, earthquakes) has consistently considered in the variations of the thermal neutron flux and compared with variations of high-energy neutrons (standard monitor 18NM64) of galactic origin during these periods. Coefficients of correlation were calculated between data of thermal neutron detectors and data of the neutron monitor, recording the intensity of high-energy particles. High correlation coefficients and similarity of responses to changes of space and geophysical conditions are obtained, that confirms the conclusion of the genetic connection of thermal neutrons with high-energy neutrons of galactic origin and suggests same sources of disturbances in the absence of seismic activity. Observations and analysis of experimental data during the activation of seismic activity in the vicinity of Almaty showed the frequent breakdown of the correlation between the intensity of thermal and high-energy neutrons and the absence of similarity between variations during these periods. We suppose that the additional thermal neutron flux of the lithospheric origin appears under these conditions. Method of separating of thermal neutron flux variations of the lithospheric origin from neutrons variations generated in the atmosphere by subtracting the normalized data is proposed, taking into account the conclusion that variations caused with the atmospheric and interplanetary origins in thermal neutron detectors are similar to variations of high-energy neutrons, and the probability of detecting by 18NM64 monitor of thermal neutrons is extremely low (less than 0, 01). We used it for analysis variations of thermal neutrons during earthquakes 2006-2015. The catalog of earthquakes in the vicinity of Almaty with intensity ≥ 3b, including 25 events, is composed on the basis of observations of the Kazakhstan National Data center. Experimental data of registration of thermal and high-energy neutrons (≥ 200 MeV) with duration not less than 14 days are prepared for an each event. The main statistical characteristics of experimental data are calculated and the normalization is carried out. The increase of thermal neutrons flux of the lithospheric origin during of seismic processes activation is observed for ~ 60% of events. However, before the earthquake the increase of thermal neutron flux is observed only for ~ 30-35% of events. It is shown that the amplitude of the additional thermal neutron flux from the Earth's crust is equal to 5-7% of the background level. Sometimes it reaches values of 10-12%. We propose to employ method of allocating the thermal neutron flux of the lithospheric origin for short-term prediction of earthquakes in seismoactive regions.

Gamma neutron assay method and apparatus

DOEpatents

Cole, J.D.; Aryaeinejad, R.; Greenwood, R.C.

1995-01-03

The gamma neutron assay technique is an alternative method to standard safeguards techniques for the identification and assaying of special nuclear materials in a field or laboratory environment, as a tool for dismantlement and destruction of nuclear weapons, and to determine the isotopic ratios for a blend-down program on uranium. It is capable of determining the isotopic ratios of fissionable material from the spontaneous or induced fission of a sample to within approximately 0.5%. This is based upon the prompt coincidence relationships that occur in the fission process and the proton conservation and quasi-conservation of nuclear mass (A) that exists between the two fission fragments. The system is used in both passive (without an external neutron source) and active (with an external neutron source) mode. The apparatus consists of an array of neutron and gamma-ray detectors electronically connected to determine coincident events. The method can also be used to assay radioactive waste which contains fissile material, even in the presence of a high background radiation field. 7 figures.

Gamma neutron assay method and apparatus

DOEpatents

Cole, Jerald D.; Aryaeinejad, Rahmat; Greenwood, Reginald C.

1995-01-01

The gamma neutron assay technique is an alternative method to standard safeguards techniques for the identification and assaying of special nuclear materials in a field or laboratory environment, as a tool for dismantlement and destruction of nuclear weapons, and to determine the isotopic ratios for a blend-down program on uranium. It is capable of determining the isotopic ratios of fissionable material from the spontaneous or induced fission of a sample to within approximately 0.5%. This is based upon the prompt coincidence relationships that occur in the fission process and the proton conservation and quasi-conservation of nuclear mass (A) that exists between the two fission fragments. The system is used in both passive (without an external neutron source and active (with an external neutron source) mode. The apparatus consists of an array of neutron and gamma-ray detectors electronically connected to determine coincident events. The method can also be used to assay radioactive waste which contains fissile material, even in the presence of a high background radiation field.

Improved lithium iodide neutron scintillator with Eu 2+ activation: The elimination of Suzuki-Phase precipitates

DOE PAGES

Boatner, Lynn A.; Comer, Eleanor P.; Wright, Gomez W.; ...

2017-02-21

Monovalent alkali halides such as NaI, CsI, and LiI are widely used as inorganic scintillators for radiation detection due to their light yield, the capability for the growth of large single crystals, relatively low cost, and other favorable characteristics. These materials are frequently activated through the addition of small amounts (e.g., a few hundred ppm) of elements such as thallium - or sodium in the case of CsI. The monovalent alkali halide scintillators can also be activated with low concentrations of Eu 2+, however Eu activation has previously not been widely employed due to the non-uniform segregation of the divalentmore » Eu dopant that leads to the formation of unwanted phases during Bridgman or other solidification crystal-growth methods. Specifically, for Eu concentrations near and above ~0.5%, Suzuki Phase precipitates form in the course of the melt-growth process, and these Suzuki Phase particles scatter the scintillation light. This adversely affects the scintillator performance via reduction in the optical transmission of the material, and depending on the crystal thickness and precipitated-particle concentration, this reduction can occur up to the point of opacity. Here we describe a post-growth process for the removal of Suzuki Phase precipitates from single crystals of the neutron scintillator LiI activated with Eu 2+ at concentrations up to and in excess of 3 wt.%, and we correlate the resulting neutron-detection performance with the thermal processing methods used to remove the Suzuki Phase particles. Furthermore, the resulting improved scintillator properties using increased Eu activator levels are applicable to neutron imaging and active interrogation systems, and pulse-height gamma-ray spectroscopy rather than pulse-shape discrimination can be used to discriminate between gamma ray and neutron interaction events.« less

Improved lithium iodide neutron scintillator with Eu2+ activation: The elimination of Suzuki-Phase precipitates

NASA Astrophysics Data System (ADS)

Boatner, L. A.; Comer, E. P.; Wright, G. W.; Ramey, J. O.; Riedel, R. A.; Jellison, G. E.; Kolopus, J. A.

2017-05-01

Monovalent alkali halides such as NaI, CsI, and LiI are widely used as inorganic scintillators for radiation detection due to their light yield, the capability for the growth of large single crystals, relatively low cost, and other favorable characteristics. These materials are frequently activated through the addition of small amounts (e.g., a few hundred ppm) of elements such as thallium - or sodium in the case of CsI. The monovalent alkali halide scintillators can also be activated with low concentrations of Eu2+, however Eu activation has previously not been widely employed due to the non-uniform segregation of the divalent Eu dopant that leads to the formation of unwanted phases during Bridgman or other solidification crystal-growth methods. Specifically, for Eu concentrations near and above 0.5%, Suzuki Phase precipitates form in the course of the melt-growth process, and these Suzuki Phase particles scatter the scintillation light. This adversely affects the scintillator performance via reduction in the optical transmission of the material, and depending on the crystal thickness and precipitated-particle concentration, this reduction can occur up to the point of opacity. Here we describe a post-growth process for the removal of Suzuki Phase precipitates from single crystals of the neutron scintillator LiI activated with Eu2+ at concentrations up to and in excess of 3 wt%, and we correlate the resulting neutron-detection performance with the thermal processing methods used to remove the Suzuki Phase particles. The resulting improved scintillator properties using increased Eu activator levels are applicable to neutron imaging and active interrogation systems, and pulse-height gamma-ray spectroscopy rather than pulse-shape discrimination can be used to discriminate between gamma ray and neutron interaction events.

Neutron-activatable radionuclide cancer therapy using graphene oxide nanoplatelets.

PubMed

Kim, Junghyun; Jay, Michael

2017-09-01

Neutron-activation is a promising method of generating radiotherapeutics with minimal handling of radioactive materials. Graphene oxide nanoplatelets (GONs) were examined as a carrier for neutron-activatable holmium with the purpose of exploiting inherent characteristics for theranostic application. GONs were hypothesized to be an ideal candidate for this application owing to their desirable characteristics such as a rigid structure, high metal loading capacity, low density, heat resistance, and the ability to withstand harsh environments associated with the neutron-activation process. Non-covalently PEGylated GONs (GONs-PEG) offered enhanced dispersibility and biocompatibility, and also exhibited increased holmium loading capacity nearly two-fold greater than GONs. Holmium leaching was investigated over a wide pH range, including conditions that mimic the tumor microenvironment, following neutron irradiation. The in vitro cell-based cytotoxicity analysis of GONs-based formulations with non-radioactive holmium confirmed their safety profile within cells. The results demonstrate the potential of GONs as a carrier of neutron-activatable radiotherapeutic agents. Copyright © 2017 Elsevier Inc. All rights reserved.

Neutron radiative capture methods for surface elemental analysis

USGS Publications Warehouse

Trombka, J.I.; Senftle, F.; Schmadebeck, R.

1970-01-01

Both an accelerator and a 252Cf neutron source have been used to induce characteristic gamma radiation from extended soil samples. To demonstrate the method, measurements of the neutron-induced radiative capture and activation gamma rays have been made with both Ge(Li) and NaI(Tl) detectors, Because of the possible application to space flight geochemical analysis, it is believed that NaI(Tl) detectors must be used. Analytical procedures have been developed to obtain both qualitative and semiquantitative results from an interpretation of the measured NaI(Tl) pulse-height spectrum. Experiment results and the analytic procedure are presented. ?? 1970.

"Chiron": A Proposed Remote Sensing Prompt Gamma Ray Activation Analysis Instrument for a Nuclear Powered Prometheus Mission

NASA Technical Reports Server (NTRS)

Floyd, Samuel R.; Keller, John W.; Dworkin, Jason P.; Mildner, David F. R.

2004-01-01

Prompt Gamma Ray Activation Analysis (PGAA) from neutron capture is an important experimental method that yields information on the elemental abundance of target materials. Gamma ray analysis has been used in planetary exploration missions by taking advantage of the production of neutrons as a result of Galactic Cosmic Ray interaction within the planetary surfaces. The .gamma ray signal that can be obtained from the GCR production of neutrons is very low, so we seek a superior neutron source. NASA s Project Prometheus and the Dept. of Energy aim to develop a nuclear power system for planetary exploration. This provides us with a tremendous opportunity to harness the reactor as a source of neutrons that can be used for PGAA. We envision a narrow stream of neutrons from the reactor directed toward the surface of an asteroid or comet producing the prompt gamma ray signal for analysis. Under ideal conditions of neutron flux and spacecraft orbit, both the signal strength and the spatial resolution will improved by several orders of magnitude over previously missions.

NEUTRON MEASURING METHOD AND APPARATUS

DOEpatents

Seaborg, G.T.; Friedlander, G.; Gofman, J.W.

1958-07-29

A fast neutron fission detecting apparatus is described consisting of a source of fast neutrons, an ion chamber containing air, two electrodes within the ion chamber in confronting spaced relationship, a high voltage potential placed across the electrodes, a shield placed about the source, and a suitable pulse annplifier and recording system in the electrode circuit to record the impulse due to fissions in a sannple material. The sample material is coated onto the active surface of the disc electrode and shielding means of a material having high neutron capture capabilities for thermal neutrons are provided in the vicinity of the electrodes and about the ion chamber so as to absorb slow neutrons of thermal energy to effectively prevent their diffusing back to the sample and causing an error in the measurement of fast neutron fissions.

Neutron production for 250 MeV protons bombarding on thick grain-made tungsten target

NASA Astrophysics Data System (ADS)

Zhang, Xueying; Zhang, Yanbin; Ma, Fei; Ju, Yongqin; Chen, Liang; Zhang, Hongbin; Li, Yanyan; Wan, Bo; Wang, Jianguo; Ge, Honglin

2015-08-01

Neutron yield for 250 MeV protons incident on a tungsten target has been measured using the water bath method. The target was made of many randomly placed tungsten grains. Through analyzing the activity of Au foils, the neutron flux distribution in water was obtained. The neutrons slowing down process shows that the neutrons from tungsten have an average energy lower than neutrons from the lead target. The neutron yield was experimentally determined to be 2.02 ± 0.15 neutron/proton. Detailed simulation was also performed with the Geant4 toolkit. Comparison has been made with the experimentally derived neutron yield. It was found that, around 250 MeV, experimental results were described satisfactorily with a combination of high-energy spallation, low-energy neutron reaction and scattering. It was shown that the grain-packed target does not affect much the main neutronic properties, which are of crucial importance for the design of the spallation target.

Improving the neutron-to-photon discrimination capability of detectors used for neutron dosimetry in high energy photon beam radiotherapy.

PubMed

Irazola, L; Terrón, J A; Bedogni, R; Pola, A; Lorenzoli, M; Sánchez-Nieto, B; Gómez, F; Sánchez-Doblado, F

2016-09-01

The increasing interest of the medical community to radioinduced second malignancies due to photoneutrons in patients undergoing high-energy radiotherapy, has stimulated in recent years the study of peripheral doses, including the development of some dedicated active detectors. Although these devices are designed to respond to neutrons only, their parasitic photon response is usually not identically zero and anisotropic. The impact of these facts on measurement accuracy can be important, especially in points close to the photon field-edge. A simple method to estimate the photon contribution to detector readings is to cover it with a thermal neutron absorber with reduced secondary photon emission, such as a borated rubber. This technique was applied to the TNRD (Thermal Neutron Rate Detector), recently validated for thermal neutron measurements in high-energy photon radiotherapy. The positive results, together with the accessibility of the method, encourage its application to other detectors and different clinical scenarios. Copyright © 2016 Elsevier Ltd. All rights reserved.

Evaluation of neutron flux parameters in irradiation sites of research reactor using the Westcott-formalism for the k0 neutron activation analysis method

NASA Astrophysics Data System (ADS)

Kasban, H.; Hamid, Ashraf

2015-12-01

Instrumental Neutron Activation Analysis using k0 (k0-INAA) method has been used to determine a number of elements in sediment samples collected from El-Manzala Lake in Egypt. k0-INAA according to Westcott's formalism has been implemented using the complete irradiation kit of the fast pneumatic rabbit and some selected manually loaded irradiation sites for short and long irradiation at Egypt Second Research Reactor (ETRR-2). Zr-Au and Co sets as neutron flux monitors are used to determine the neutron flux parameters (f and α) in each irradiation sites. Two reference materials IAEA Soil-7 samples have been inserted and implemented for data validation and an internal monostandard multi monitor used (k0 based IM-NAA). It was given a good agreement between the experimental analyzed values and that obtained of the certified values. The major and trace elements in the sediment samples have been evaluated with the use of Co as an internal and Au as an external monostandard comparators. The concentrations of the elements (Cr, Mn and Zn) in the sediment samples of the present work are discussed regarding to those obtained from other sites.

Application of neutron transmutation doping method to initially p-type silicon material.

PubMed

Kim, Myong-Seop; Kang, Ki-Doo; Park, Sang-Jun

2009-01-01

The neutron transmutation doping (NTD) method was applied to the initially p-type silicon in order to extend the NTD applications at HANARO. The relationship between the irradiation neutron fluence and the final resistivity of the initially p-type silicon material was investigated. The proportional constant between the neutron fluence and the resistivity was determined to be 2.3473x10(19)nOmegacm(-1). The deviation of the final resistivity from the target for almost all the irradiation results of the initially p-type silicon ingots was at a range from -5% to 2%. In addition, the burn-up effect of the boron impurities, the residual (32)P activity and the effect of the compensation characteristics for the initially p-type silicon were studied. Conclusively, the practical methodology to perform the neutron transmutation doping of the initially p-type silicon ingot was established.

Ir and Rare Earth's Elements determination by Neutron Activation Analysis and ICP - MS in soil samples

NASA Astrophysics Data System (ADS)

Salvini, A.; Cattadori, C.; Broggini, C.; Cagnazzo, M.; Ori, Gian Gabriele; Nisi, S.; Borio, A.; Manera, S.

2006-05-01

The platinum metals depleted in the earth's crust are relative to their cosmic abundance; concentration of these elements in sediments may thus indicate influxes of extraterrestrial material. Analysis of these parameters are done easily by Neutron Activation Analysis (NAA) and comparative results with ICP-MS technique show a good match. Results, adjust parameters and limits of this method will be displayed in tables.

Neutrons Flux Distributions of the Pu-Be Source and its Simulation by the MCNP-4B Code

NASA Astrophysics Data System (ADS)

Faghihi, F.; Mehdizadeh, S.; Hadad, K.

Neutron Fluence rate of a low intense Pu-Be source is measured by Neutron Activation Analysis (NAA) of 197Au foils. Also, the neutron fluence rate distribution versus energy is calculated using the MCNP-4B code based on ENDF/B-V library. Theoretical simulation as well as our experimental performance are a new experience for Iranians to make reliability with the code for further researches. In our theoretical investigation, an isotropic Pu-Be source with cylindrical volume distribution is simulated and relative neutron fluence rate versus energy is calculated using MCNP-4B code. Variation of the fast and also thermal neutrons fluence rate, which are measured by NAA method and MCNP code, are compared.

SNL-SAND-IV v. 0.9 (beta)

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

Griffin, Patrick J.

2016-10-05

The code is used to provide an unfolded/adjusted energy-dependent fission reactor neutron spectrum based upon an input trial spectrum and a set of measured activities. This is part of a neutron environment characterization that supports doing testing in a given reactor environment. An iterative perturbation method is used to obtain a "best fit" neutron flux spectrum for a given input set of infinitely dilute foil activities. The calculational procedure consists of the selection of a trial flux spectrum to serve as the initial approximation to the solution, and subsequent iteration to a form acceptable as an appropriate solution. The solutionmore » is specified either as time-integrated flux (fluence) for a pulsed environment or as a flux for a steady-state neutron environment.« less

Certification of biological candidates reference materials by neutron activation analysis

NASA Astrophysics Data System (ADS)

Kabanov, Denis V.; Nesterova, Yulia V.; Merkulov, Viktor G.

2018-03-01

The paper gives the results of interlaboratory certification of new biological candidate reference materials by neutron activation analysis recommended by the Institute of Nuclear Chemistry and Technology (Warsaw, Poland). The correctness and accuracy of the applied method was statistically estimated for the determination of trace elements in candidate reference materials. The procedure of irradiation in the reactor thermal fuel assembly without formation of fast neutrons was carried out. It excluded formation of interfering isotopes leading to false results. The concentration of more than 20 elements (e.g., Ba, Br, Ca, Co, Ce, Cr, Cs, Eu, Fe, Hf, La, Lu, Rb, Sb, Sc, Ta, Th, Tb, Yb, U, Zn) in candidate references of tobacco leaves and bottom sediment compared to certified reference materials were determined. It was shown that the average error of the applied method did not exceed 10%.

Method and apparatus for measuring reactivity of fissile material

DOEpatents

Lee, David M.; Lindquist, Lloyd O.

1985-01-01

Given are a method and apparatus for measuring nondestructively and non-invasively (i.e., using no internal probing) the burnup, reactivity, or fissile content of any material which emits neutrons and which has fissionable components. No external neutron-emitting interrogation source or fissile material is used and no scanning is required, although if a profile is desired scanning can be used. As in active assays, here both reactivity and content of fissionable material can be measured. The assay is accomplished by altering the return flux of neutrons into the fuel assembly. The return flux is altered by changing the reflecting material. The existing passive neutron emissions in the material being assayed are used as the source of interrogating neutrons. Two measurements of either emitted neutron or emitted gamma-ray count rates are made and are then correlated to either reactivity, burnup, or fissionable content of the material being assayed, thus providing a measurement of either reactivity, burnup, or fissionable content of the material being assayed. Spent fuel which has been freshly discharged from a reactor can be assayed using this method and apparatus. Precisions of 1000 MWd/tU appear to be feasible.

Thermal Neutron Tomography for Cultural Heritage at INR

NASA Astrophysics Data System (ADS)

Dinca, Marin; Mandescu, Dragos

The neutron and gamma imaging facility placed at the tangential channel of the TRIGA-ACPR from INR was used for tomography investigations on a test object with good results and shortly followed its involvement for tomography investigations on prehistoric statues of clay from the Arges County Museum. This activity was performed in connection with a research contract with IAEA with title ;The neutron and gamma imaging method combined with neutron-based analytical methods for cultural heritage research;, in the frame of a current CRP, that helps curators to reveal the internal structure and composition of the objects. The detector system has been developed based on two interchangeable scintillators, one for thermal neutrons and the other one for gamma radiations, a mirror of float glass coated with aluminum and two interchangeable CCD cameras. Experiments of tomography imaging for two prehistoric statues of clay with CCD STARLIGHT XPRESS SXV-H9 camera with XD-4 type image intensifier are presented in this paper. The tomography reconstructions with Octopus software have shown the potential of good results even for 100 projections/1800. This was a good opportunity for the dissemination of the investigation methods based on neutrons for cultural heritage and beyond this area.

THE DETERMINATION OF TRACES OF IRON IN SAMPLES OF PLATINUM BY NE TRON- ACTIVATION ANALYSIS

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

Morris, D.F.C.; Killick, R.A.

1963-11-01

A neutron-activation analysis method for the determination of traces of iron in samples of purified platinum is described. The nuclear reactor BEPO at Harwell was used as the neutron source. A rapid radiochemical separation procedure using carriers was employed to decontaminate the iron activity from most other induced activities. The analysis is completed by discriminated gamma scintillation counting. Results of analyses of seven samples of platinum are quoted. The method of analysis has the advantage that it obviates difficulties caused by reagent blanks or by contamination from traces of inactive iron after irradiation. Interference resulting from nuclear reactions of elementsmore » other than iron in the samples appears to be of no consequence. (auth)« less

Determination of the neutron activation profile of core drill samples by gamma-ray spectrometry.

PubMed

Gurau, D; Boden, S; Sima, O; Stanga, D

2018-04-01

This paper provides guidance for determining the neutron activation profile of core drill samples taken from the biological shield of nuclear reactors using gamma spectrometry measurements. Thus, it provides guidance for selecting a model of the right form to fit data and using least squares methods for model fitting. The activity profiles of two core samples taken from the biological shield of a nuclear reactor were determined. The effective activation depth and the total activity of core samples along with their uncertainties were computed by Monte Carlo simulation. Copyright © 2017 Elsevier Ltd. All rights reserved.

Development of Techniques for Spent Fuel Assay – Differential Dieaway Final Report

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

Swinhoe, Martyn Thomas; Goodsell, Alison; Ianakiev, Kiril Dimitrov

This report summarizes the work done under a DNDO R&D funded project on the development of the differential dieaway method to measure plutonium in spent fuel. There are large amounts of plutonium that are contained in spent fuel assemblies, and currently there is no way to make quantitative non-destructive assay. This has led NA24 under the Next Generation Safeguards Initiative (NGSI) to establish a multi-year program to investigate, develop and implement measurement techniques for spent fuel. The techniques which are being experimentally tested by the existing NGSI project do not include any pulsed neutron active techniques. The present work coversmore » the active neutron differential dieaway technique and has advanced the state of knowledge of this technique as well as produced a design for a practical active neutron interrogation instrument for spent fuel. Monte Carlo results from the NGSI effort show that much higher accuracy (1-2%) for the Pu content in spent fuel assemblies can be obtained with active neutron interrogation techniques than passive techniques, and this would allow their use for nuclear material accountancy independently of any information from the operator. The main purpose of this work was to develop an active neutron interrogation technique for spent nuclear fuel.« less

A Spatial and Temporal Characterization of the Background Neutron Environment at the Navy and Marine Corps Memorial Stadium

DTIC Science & Technology

2017-04-01

developing informed survey protocols. Experimental Method The neutron detector used in this research was the Large Neutron Sensor (LNS), containing...are useful in planning, conducting, and assessing the utility and limitations of radiation surveys using current state-of-the-art portable or...34,000. In a security environment, where large public venues may be a target for terrorist activity, the ability to survey venues for radio- logical

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.

Absolute determination of copper and silver in ancient coins using 14 MeV neutrons

NASA Astrophysics Data System (ADS)

Chalouhi, Ch.; Hourani, E.; Loos, R.; Melki, S.

1982-09-01

A method for absolute determination of copper and silver in ancient coins is described. Activation analysis by 14 MeV neutrons is performed. In the experimental procedure emphasis is placed on corrections for neutrons and gamma attenuation. In the analytical procedure, a multi linear-regression calculation is used to separate different contributions to the 511 keV gamma peak. The precision in the absolute determination of Cu and Ag is better than 2% in recent coins of definite shapes, whereas it is a somewhat lower in ancient coins of irregular shapes. The method was applied to ancient coins provided by the Museum of the American University of Beirut. Overall consistency and suitability of the method were obtained.

Determination of tungsten in geochemical reference material basalt Columbia River 2 by radiochemical neutron activation analysis and inductively coupled plasma mass spectrometry

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

Morrison, Samuel S.; Beck, Chelsie L.; Bowen, James M.

Environmental tungsten (W) analyses are inhibited by a lack of reference materials and practical methods to remove isobaric and radiometric interferences. We present a method that evaluates the potential use of commercially available sediment, Basalt Columbia River-2 (BCR-2), as a reference material using neutron activation analysis (NAA) and mass spectrometry. Tungsten concentrations using both methods are in statistical agreement at the 95% confidence interval (92 ± 4 ng/g for NAA and 100 ±7 ng/g for mass spectrometry) with recoveries greater than 95%. These results indicate that BCR-2 may be suitable as a reference material for future studies.

A large 2D PSD for thermal neutron detection

NASA Astrophysics Data System (ADS)

Knott, R. B.; Smith, G. C.; Watt, G.; Boldeman, J. W.

1997-02-01

A 2D PSD based on a MWPC has been constructed for a small angle neutron scattering instrument. The active area of the detector was 640 × 640 mm 2. To meet the specifications for neutron detection efficiency and spatial resolution, and to minimise parallax, the gas mixture was 190 kPa 3He plus 100 kPa CF 4, and the active volume had a thickness of 30 mm. The design maximum neutron count rate of the detector was 10 5 events per secod. The (calculated) neutron detection efficiency was 60% for 2 Å neutrons and the (measured) neutron energy resolution on the anode grid was typically 20% (fwhm). The location of a neutron detection event within the active area was determined using the wire-by-wire method: the spatial resolution (5 × 5 mm 2) was thereby defined by the wire geometry. A 16-channel charge-sensitive preamplifier/amplifier/comparator module has been developed with a channel sensitivity of 0.1 V/fC, noise line width of 0.4 fC (fwhm) and channel-to-channel cross-talk of less than 5%. The Proportional Counter Operating System (PCOS III) (LeCroy Corp, USA) was used for event encoding. The ECL signals produced by the 16 channel modules were latched in PCOS III by a trigger pulse from the anode and the fast encoders produce a position and width for each event. The information was transferred to a UNIX workstation for accumulation and online display.

A comparative study of neutron activation analysis and proton-induced X-ray emission analysis for the determination of heavy metals in estuarine sediments

NASA Astrophysics Data System (ADS)

Randle, K.; Al-Jundi, J.; Mamas, C. J. V.; Sokhi, R. S.; Earwaker, L. G.

1993-06-01

Our work on heavy metals in the estuarine environment has involved the use of two multielement techniques: neutron activation analysis (NAA) and proton-induced X-ray emission (PIXE) analysis. As PIXE is essentially a surface analytical technique problems may arise due to sample inhomogeneity and surface roughness. In order to assess the contribution of these effects we have compared the results from PIXE analysis with those from a technique which analyzes a larger bulk sample rather than just the surface. An obvious method was NAA. A series of sediment samples containing particles of variable diameter were compared. Pellets containing a few mg of sediment were prepared from each sample and analyzed by the PIXE technique using both an absolute and a comparitive method. For INAA the rest of the sample was then irradiated with thermal neutrons and element concentrations determined from analyses of the subsequent gamma-ray spectrum. Results from the two methods are discussed.

Determination of the Secondary Neutron Flux at the Massive Natural Uranium Spallation Target

NASA Astrophysics Data System (ADS)

Zeman, M.; Adam, J.; Baldin, A. A.; Furman, W. I.; Gustov, S. A.; Katovsky, K.; Khushvaktov, J.; Mar`in, I. I.; Novotny, F.; Solnyshkin, A. A.; Tichy, P.; Tsoupko-Sitnikov, V. M.; Tyutyunnikov, S. I.; Vespalec, R.; Vrzalova, J.; Wagner, V.; Zavorka, L.

The flux of secondary neutrons generated in collisions of the 660 MeV proton beam with the massive natural uranium spallation target was investigated using a set of monoisotopic threshold activation detectors. Sandwiches made of thin high-purity Al, Co, Au, and Bi metal foils were installed in different positions across the whole spallation target. The gamma-ray activity of products of (n,xn) and other studied reactions was measured offline with germanium semiconductor detectors. Reaction yields of radionuclides with half-life exceeding 100 min and with effective neutron energy thresholds between 3.6 MeV and 186 MeV provided us with information about the spectrum of spallation neutrons in this energy region and beyond. The experimental neutron flux was determined using the measured reaction yields and cross-sections calculated with the TALYS 1.8 nuclear reaction program and INCL4-ABLA event generator of MCNP6. Neutron spectra in the region of activation sandwiches were also modeled with the radiation transport code MCNPX 2.7. Neutron flux based on excitation functions from TALYS provides a reasonable description of the neutron spectrum inside the spallation target and is in good agreement with Monte-Carlo predictions. The experimental flux that uses INCL4 cross-sections rather underestimates the modeled spectrum in the whole region of interest, but the agreement within few standard deviations was reached as well. The paper summarizes basic principles of the method for determining the spectrum of high-energy neutrons without employing the spectral adjustment routines and points out to the need for model improvements and precise cross-section measurements.

Update on reactors and reactor instruments in Asia

NASA Astrophysics Data System (ADS)

Rao, K. R.

1991-10-01

The 1980s have seen the commissioning of several medium flux (∼10 14 neutrons/cm 2s) research reactors in Asia. The reactors are based on indigenous design and development in India and China. At Dhruva reactor (India), a variety of neutron spectrometers have been established that have provided useful data related to the structure of high- Tc materials, phonon density of states, magnetic moment distributions and micellar aggregation during the last couple of years. Polarised neutron analysis, neutron interferometry and neutron spin echo methods are some of the new techniques under development. The spectrometers and associated automaton, detectors and neutron guides have all been indigenously developed. This paper summarises the developments and on-going activities in Bangladesh, China, India, Indonesia, Korea, Malaysia, the Philippines and Thailand.

66.7-keV γ -line intensity of 171Tm determined via neutron activation

NASA Astrophysics Data System (ADS)

Weigand, M.; Heftrich, T.; Düllmann, Ch. E.; Eberhardt, K.; Fiebiger, S.; Glorius, J.; Göbel, K.; Haas, R.; Langer, C.; Lohse, S.; Reifarth, R.; Renisch, D.; Wolf, C.

2018-03-01

Background: About 50% of the heavy elements are produced in stars during the slow neutron capture process. The analysis of branching points allows to set constraints on the temperature and the neutron density in the interior of stars. The temperature dependence of the branch point 171Tm is weak. Hence, the 171Tm neutron capture cross section can be used to constrain the neutron density during the main component of the s process in thermally pulsing asymptotic giant branch stars. Purpose: In order to perform neutron capture experiments on 171Tm, sample material has to be produced and characterized. The characterization is done by γ spectroscopy, relying on the intensities of the involved γ lines. Only the 66.7-keV γ line can be observed whose intensity was uncertain so far. Method: An enriched 170Er sample was activated with thermal neutrons at the TRIGA (Training, Research, Isotopes, General Atomics) research reactor at the Johannes Gutenberg-Universität Mainz. The activation resulted in an easily quantifiable number of 171Er nuclei that subsequently decayed to 171Tm. Result: The intensity of the 66.7-keV γ line of the 171Tm decay was measured to Iγ=(0.144 ±0.010 )% . Conclusions: Our result is in good agreement with the value found in the literature.

Measurements of 89Y(n,2n)88Y and 89Y(n,3n)87Y, 87mY cross sections for fast neutrons at KIRAMS

NASA Astrophysics Data System (ADS)

In, Eun Jin; Bak, Sang-In; Ham, Cheolmin; Kim, Do Yoon; Myung, Hyunjeong; Shim, Chungbo; Shin, Jae Won; Min, Kyung Joo; Zhou, Yujie; Park, Tae-Sun; Hong, Seung-Woo; Bhoraskar, V. N.

2017-09-01

A proton cyclotron MC-50 in Korea Institute of Radiological & Medical Science (KIRAMS) is used to carry out neutron activation experiments with Y2O3 targets irradiated with neutron beams of a continuous spectrum produced by proton beams on a thick beryllium target. Neutrons are generated by 9Be (p, n) reaction with an incident proton intensity of 20 μA. The neutron spectra generated by proton beams of 30, 35, and 40 MeV are calculated by GEANT4 simulations. Nb powders are used for neutron flux monitoring by measuring the activities of 92mNb through the reaction 93Nb (n, 2n). By using a subtraction method, the average cross section of 89Y(n,2n) and 89Y(n,3n) reactions at the neutron energies of 29.8 ± 1.8 MeV and 34.8 ± 1.8 MeV are extracted and are found to be close to the existing cross sections from the EXFOR data and the evaluated nuclear data libraries such as TENDL-2015 or EAF-2010.

Determination of specific activity of cobalt (60Co/Co) in steel samples exposed to the atomic bomb in Hiroshima.

PubMed

Kimura, T; Takano, N; Iba, T; Fujita, S; Watanabe, T; Maruyama, T; Hamada, T

1990-06-01

Specific activity 60Co/Co in two steel samples taken at 687m S and 1295m NNW from the hypocenter was measured by gamma-ray spectrometry and neutron activation analysis. The results were, respectively, (2.64 +/- 0.38) x 10(1) and (3.09 +/- 0.48) x 10(-1) dpm/mg Co at the time of bombing, which are consistent with previous data by Hashizume et al. for steel rings on the surface of roofs of buildings. The present data are expected to serve as verification of the bomb neutron transport calculations. Content of nickel and copper in the samples, determined by colorimetric and neutron activation methods, respectively, was too small to account for any significant 60Co production by the (n,p) and (n, alpha) reactions.

Nondestructive Inspection System for Special Nuclear Material Using Inertial Electrostatic Confinement Fusion Neutrons and Laser Compton Scattering Gamma-Rays

NASA Astrophysics Data System (ADS)

Ohgaki, H.; Daito, I.; Zen, H.; Kii, T.; Masuda, K.; Misawa, T.; Hajima, R.; Hayakawa, T.; Shizuma, T.; Kando, M.; Fujimoto, S.

2017-07-01

A Neutron/Gamma-ray combined inspection system for hidden special nuclear materials (SNMs) in cargo containers has been developed under a program of Japan Science and Technology Agency in Japan. This inspection system consists of an active neutron-detection system for fast screening and a laser Compton backscattering gamma-ray source in coupling with nuclear resonance fluorescence (NRF) method for precise inspection. The inertial electrostatic confinement fusion device has been adopted as a neutron source and two neutron-detection methods, delayed neutron noise analysis method and high-energy neutron-detection method, have been developed to realize the fast screening system. The prototype system has been constructed and tested in the Reactor Research Institute, Kyoto University. For the generation of the laser Compton backscattering gamma-ray beam, a race track microtron accelerator has been used to reduce the size of the system. For the NRF measurement, an array of LaBr3(Ce) scintillation detectors has been adopted to realize a low-cost detection system. The prototype of the gamma-ray system has been demonstrated in the Kansai Photon Science Institute, National Institutes for Quantum and Radiological Science and Technology. By using numerical simulations based on the data taken from these prototype systems and the inspection-flow, the system designed by this program can detect 1 kg of highly enriched 235U (HEU) hidden in an empty 20-ft container within several minutes.

Gadolinium-loaded Plastic Scintillators for Thermal Neutron Detection using Compensation

NASA Astrophysics Data System (ADS)

Dumazert, Jonathan; Coulon, Romain; Hamel, Matthieu; Carrel, Frédérick; Sguerra, Fabien; Normand, Stéphane; Méchin, Laurence; Bertrand, Guillaume H. V.

2016-06-01

Plastic scintillator loading with gadolinium-rich organometallic complexes shows a high potential for the deployment of efficient and cost-effective neutron detectors. Due to the low-energy photon and electron signature of thermal neutron capture by Gd-155 and Gd-157, alternative treatment to pulse-shape discrimination has to be proposed in order to display a count rate. This paper discloses the principle of a compensation method applied to a two-scintillator system: a detection scintillator interacts with photon and fast neutron radiation and is loaded with gadolinium organometallic compound to become a thermal neutron absorber, while a not-gadolinium loaded compensation scintillator solely interacts with the fast neutron and photon part of incident radiation. After the nonlinear smoothing of the counting signals, a hypothesis test determines whether the resulting count rate post-background response compensation falls into statistical fluctuations or provides a robust indication of neutron activity. Laboratory samples are tested under both photon and neutron irradiations, allowing the authors to investigate the performance of the overall detection system in terms of sensitivity and detection limits, especially with regards to a similar-active volume He-3 based commercial counter. The study reveals satisfactory figures of merit in terms of sensitivity and directs future investigation toward promising paths.

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

Heyse, J.; Becker, B.; Kopecky, S.

Neutrons can be used as a tool to study properties of materials and objects. An evolving activity in this field focusses on neutron induced reaction cross sections. The probability that a neutron interacts with nuclei strongly depends on the energy of the neutron. The cross sections reveal the presence of resonance structures, the energy and width of which are isotope specific. As such, these resonance structures can be used as fingerprints to determine the elemental and isotopic composition of materials and objects. They are the basis of two analytical methods which have been developed at Institute for Reference Materials andmore » Measurements of the European Commission's Joint Research Centre (EC-JRC-IRMM): Neutron Resonance Capture Analysis (NRCA) and Neutron Resonance Transmission Analysis (NRTA). The first technique is based on the detection of gamma rays emitted during a neutron capture reaction in the sample being studied; the latter determines the fraction of neutrons transmitted through a sample positioned in a neutron beam. In the past both techniques have been applied to determine the composition of archaeological objects and to characterize nuclear reference materials. More recently a combination of NRTA and NRCA is being studied as a non-destructive method to determine the heavy metal content of particle-like debris of melted fuel that is formed in severe nuclear accidents such as the one which occurred at the Fukushima Daiichi nuclear power plant in Japan. This study is part of a collaboration between the Japan Atomic Energy Agency (JAEA) and ECJRC- IRMM and is a spin-off from the core activity of IRMM, i.e. the production of nuclear data for nuclear technology applications. This contribution focusses on a newly developed NRTA measurement station that has been set up recently at one of the flight paths of the neutron time-of-flight facility GELINA at the EC-JRC-IRMM. The basic principles of NRTA and first results of measurements at the new set up will be discussed. (authors)« less

Determination of the thermal and epithermal neutron sensitivities of an LBO chamber.

PubMed

Endo, Satoru; Sato, Hitoshi; Shimazaki, Takuto; Nakajima, Erika; Kotani, Kei; Suda, Mitsuru; Hamano, Tsuyoshi; Kajimoto, Tsuyoshi; Tanaka, Kenichi; Hoshi, Masaharu

2017-08-01

An LBO (Li 2 B 4 O 7 ) walled ionization chamber was designed to monitor the epithermal neutron fluence in boron neutron capture therapy clinical irradiation. The thermal and epithermal neutron sensitivities of the device were evaluated using accelerator neutrons from the 9 Be(d, n) reaction at a deuteron energy of 4 MeV (4 MeV d-Be neutrons). The response of the chamber in terms of the electric charge induced in the LBO chamber was compared with the thermal and epithermal neutron fluences measured using the gold-foil activation method. The thermal and epithermal neutron sensitivities obtained were expressed in units of pC cm 2 , i.e., from the chamber response divided by neutron fluence (cm -2 ). The measured LBO chamber sensitivities were 2.23 × 10 -7  ± 0.34 × 10 -7 (pC cm 2 ) for thermal neutrons and 2.00 × 10 -5  ± 0.12 × 10 -5 (pC cm 2 ) for epithermal neutrons. This shows that the LBO chamber is sufficiently sensitive to epithermal neutrons to be useful for epithermal neutron monitoring in BNCT irradiation.

Research on the self-absorption corrections for PGNAA of large samples

NASA Astrophysics Data System (ADS)

Yang, Jian-Bo; Liu, Zhi; Chang, Kang; Li, Rui

2017-02-01

When a large sample is analysed with the prompt gamma neutron activation analysis (PGNAA) neutron self-shielding and gamma self-absorption affect the accuracy, the correction method for the detection efficiency of the relative H of each element in a large sample is described. The influences of the thickness and density of the cement samples on the H detection efficiency, as well as the impurities Fe2O3 and SiO2 on the prompt γ ray yield for each element in the cement samples, were studied. The phase functions for Ca, Fe, and Si on H with changes in sample thickness and density were provided to avoid complicated procedures for preparing the corresponding density or thickness scale for measuring samples under each density or thickness value and to present a simplified method for the measurement efficiency scale for prompt-gamma neutron activation analysis.

Background and Source Term Identification in Active Neutron Interrogation Methods

DTIC Science & Technology

2011-03-24

interactions occurred to observe gamma ray peaks and not unduly increase simulation time. Not knowing the uranium enrichment modeled by Gozani, pure U...neutron interactions can occur. The uranium targets, though, should have increased neutron fluencies as the energy levels become below 2 MeV. This is...Assessment Monitor Site (TEAMS) at Kirtland AFB, NM. Iron (Fe-56), lead (Pb-207), polyethylene (C2H4 –– > C-12 & H-1), and uranium (U-235 and U-238) were

Application of Neutron Tomography in Culture Heritage research.

PubMed

Mongy, T

2014-02-01

Neutron Tomography (NT) investigation of Culture Heritages (CH) is an efficient tool for understanding the culture of ancient civilizations. Neutron imaging (NI) is a-state-of-the-art non-destructive tool in the area of CH and plays an important role in the modern archeology. The NI technology can be widely utilized in the field of elemental analysis. At Egypt Second Research Reactor (ETRR-2), a collimated Neutron Radiography (NR) beam is employed for neutron imaging purposes. A digital CCD camera is utilized for recording the beam attenuation in the sample. This helps for the detection of hidden objects and characterization of material properties. Research activity can be extended to use computer software for quantitative neutron measurement. Development of image processing algorithms can be used to obtain high quality images. In this work, full description of ETRR-2 was introduced with up to date neutron imaging system as well. Tomographic investigation of a clay forged artifact represents CH object was studied by neutron imaging methods in order to obtain some hidden information and highlight some attractive quantitative measurements. Computer software was used for imaging processing and enhancement. Also the Astra Image 3.0 Pro software was employed for high precise measurements and imaging enhancement using advanced algorithms. This work increased the effective utilization of the ETRR-2 Neutron Radiography/Tomography (NR/T) technique in Culture Heritages activities. © 2013 Elsevier Ltd. All rights reserved.

Neutron-capture cross-section measurements of Xe136 between 0.4 and 14.8 MeV

NASA Astrophysics Data System (ADS)

Bhike, Megha; Tornow, W.

2014-03-01

Fast-neutron-capture cross-section data on Xe136 have been measured with the activation method between 0.4 and 14.8 MeV. The cross section was found to be of the order of 1 mb at the eleven energies investigated. This result is important to interpret potential neutron-induced backgrounds in the enriched xenon observatory and KamLAND-Zen neutrinoless double-β decay searches that use xenon as both source and detector. A high-pressure sphere filled with Xe136 was irradiated with monoenergetic neutrons produced by the reactions 3H(p ,n)3He, 2H(d ,n)3He, and 3H(d ,n)4He. Indium and gold monitor foils were irradiated simultaneously with the Xe136 to determine the incident neutron flux. The activities of the reaction products were measured with high-resolution γ-ray spectroscopy. The present results are compared to predictions from ENDF/B-VII.1 and TENDL-2012.

Neutron resonance spectroscopy for the characterization of materials and objects

NASA Astrophysics Data System (ADS)

Schillebeeckx, P.; Borella, A.; Emiliani, F.; Gorini, G.; Kockelmann, W.; Kopecky, S.; Lampoudis, C.; Moxon, M.; Perelli Cippo, E.; Postma, H.; Rhodes, N. J.; Schooneveld, E. M.; Van Beveren, C.

2012-03-01

The resonance structure in neutron induced reaction cross sections can be used to determine the elemental compositions of materials or objects. The occurrence of resonances is the basis of neutron resonance capture analysis (NRCA) and neutron resonance transmission analysis (NRTA). NRCA and NRTA are fully non-destructive methods to determine the bulk elemental composition without the need of any sample preparation and resulting in a negligible residual activity. They have been applied to determine the elemental composition of archaeological objects and to characterize reference materials used for cross section measurements. For imaging applications a position sensitive neutron detector has been developed within the ANCIENT CHARM project. The detector is based on a 10 × 10 array of 6Li-glass scintillators mounted on a pitch of 2.5 mm, resulting in a 25 × 25 mm2 active area. The detector has been tested at the time-of-flight facility GELINA and used at the ISIS spallation source to study cultural heritage objects.

Dual neutral particle induced transmutation in CINDER2008

NASA Astrophysics Data System (ADS)

Martin, W. J.; de Oliveira, C. R. E.; Hecht, A. A.

2014-12-01

Although nuclear transmutation methods for fission have existed for decades, the focus has been on neutron-induced reactions. Recent novel concepts have sought to use both neutrons and photons for purposes such as active interrogation of cargo to detect the smuggling of highly enriched uranium, a concept that would require modeling the transmutation caused by both incident particles. As photonuclear transmutation has yet to be modeled alongside neutron-induced transmutation in a production code, new methods need to be developed. The CINDER2008 nuclear transmutation code from Los Alamos National Laboratory is extended from neutron applications to dual neutral particle applications, allowing both neutron- and photon-induced reactions for this modeling with a focus on fission. Following standard reaction modeling, the induced fission reaction is understood as a two-part reaction, with an entrance channel to the excited compound nucleus, and an exit channel from the excited compound nucleus to the fission fragmentation. Because photofission yield data-the exit channel from the compound nucleus-are sparse, neutron fission yield data are used in this work. With a different compound nucleus and excitation, the translation to the excited compound state is modified, as appropriate. A verification and validation of these methods and data has been performed. This has shown that the translation of neutron-induced fission product yield sets, and their use in photonuclear applications, is appropriate, and that the code has been extended correctly.

Nuclear reactor with internal thimble-type delayed neutron detection system

DOEpatents

Gross, Kenny C.; Poloncsik, John; Lambert, John D. B.

1990-01-01

This invention teaches improved apparatus for the method of detecting a breach in cladded fuel used in a nuclear reactor. The detector apparatus is located in the primary heat exchanger which conveys part of the reactor coolant past at least three separate delayed-neutron detectors mounted in this heat exchanger. The detectors are spaced apart such that the coolant flow time from the core to each detector is different, and these differences are known. The delayed-neutron activity at the detectors is a function of the delay time after the reaction in the fuel until the coolant carrying the delayed-neutron emitter passes the respective detector. This time delay is broken down into separate components including an isotopic holdup time required for the emitter to move through the fuel from the reaction to the coolant at the breach, and two transit times required for the emitter now in the coolant to flow from the breach to the detector loop and then via the loop to the detector. At least two of these time components are determined during calibrated operation of the reactor. Thereafter during normal reactor operation, repeated comparisons are made by the method of regression approximation of the third time component for the best-fit line correlating measured delayed-neutron activity against activity that is approximated according to specific equations. The equations use these time-delay components and known parameter values of the fuel and of the part and emitting daughter isotopes.

The impact of heavy metals from environmental tobacco smoke on indoor air quality as determined by Compton suppression neutron activation analysis.

PubMed

Landsberger, S; Wu, D

1995-12-01

The method of instrumental neutron activation analysis (NAA) has been improved for air filter samples in the determination of low level heavy metals in indoor air. By using the techniques of epithermal neutron irradiation in conjunction with Compton suppression, the detection limits of cadmium, arsenic and antimony measurements have been dramatically reduced to 2 ng for Cd, 0.2 ng for As, and 0.03 ng for Sb. The determination of these heavy metals in particulate material generated from cigarette smoking in indoor environments has been conducted. Other elements, Br, Cl, Na, K, Zn were also found at elevated levels.

Determination of the origin of the medieval glass bracelets discovered in Dubna, Moscow region, Russia using the neutron activation analysis

NASA Astrophysics Data System (ADS)

Dmitrieva, S. O.; Frontasyeva, M. V.; Dmitriev, A. A.; Dmitriev, A. Yu.

2017-01-01

The work is dedicated to the determination of the origin of archaeological finds from medieval glass using the method of neutron activation analysis (NAA). Among such objects we can discover not only things produced in ancient Russian glassmaking workshops but also imported from Byzantium. The authors substantiate the ancient Russian origin of the medieval glass bracelets of pre-Mongol period, found on the ancient Dubna settlement. The conclusions are based on data about the glass chemical composition obtained as a result of NAA of 10 fragments of bracelets at the IBR-2 reactor (Frank Laboratory of Neutron Physics, Joint Institute for Nuclear Research).

Measurement of the Neutron Flux of the Reaction C$sup 12$(d,n)N$sup 13$ Based on the Activity of N$sup 13$. Report No. 28; MEDICION DEL FLUJO NEUTRONICO DE LA REACCION C$sup 12$(d,n)N$sup 13$ A BASE DE LA ACTIVIDAD DEL N$sup 13$. INFORME NO 28

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

Keller, A.

1960-08-01

The principal problem in the measurement of neutron flux from the reaction C/sup 12/(d,n)N/sup 13/, based on the activity of N/sup 13/, is the possibility of nitrogen escape. A method for the measurement of possibie N/sup 13/ losses is described, and measurements with TiC targets irradiated with 600- kev deuterons are reported. The formulas for the calculation of the neutron flux are indicated. (tr-auth)

Advanced analysis techniques for uranium assay

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

Geist, W. H.; Ensslin, Norbert; Carrillo, L. A.

2001-01-01

Uranium has a negligible passive neutron emission rate making its assay practicable only with an active interrogation method. The active interrogation uses external neutron sources to induce fission events in the uranium in order to determine the mass. This technique requires careful calibration with standards that are representative of the items to be assayed. The samples to be measured are not always well represented by the available standards which often leads to large biases. A technique of active multiplicity counting is being developed to reduce some of these assay difficulties. Active multiplicity counting uses the measured doubles and triples countmore » rates to determine the neutron multiplication (f4) and the product of the source-sample coupling ( C ) and the 235U mass (m). Since the 35U mass always appears in the multiplicity equations as the product of Cm, the coupling needs to be determined before the mass can be known. A relationship has been developed that relates the coupling to the neutron multiplication. The relationship is based on both an analytical derivation and also on empirical observations. To determine a scaling constant present in this relationship, known standards must be used. Evaluation of experimental data revealed an improvement over the traditional calibration curve analysis method of fitting the doubles count rate to the 235Um ass. Active multiplicity assay appears to relax the requirement that the calibration standards and unknown items have the same chemical form and geometry.« less

1983 international intercomparison of nuclear accident dosimetry systems at Oak Ridge National Laboratory

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

Swaja, R.E.; Greene, R.T.; Sims, C.S.

1985-04-01

An international intercomparison of nuclear accident dosimetry systems was conducted during September 12-16, 1983, at Oak Ridge National Laboratory (ORNL) using the Health Physics Research Reactor operated in the pulse mode to simulate criticality accidents. This study marked the twentieth in a series of annual accident dosimetry intercomparisons conducted at ORNL. Participants from ten organizations attended this intercomparison and measured neutron and gamma doses at area monitoring stations and on phantoms for three different shield conditions. Results of this study indicate that foil activation techniques are the most popular and accurate method of determining accident-level neutron doses at area monitoringmore » stations. For personnel monitoring, foil activation, blood sodium activation, and thermoluminescent (TL) methods are all capable of providing accurate dose estimates in a variety of radiation fields. All participants in this study used TLD's to determine gamma doses with very good results on the average. Chemical dosemeters were also shown to be capable of yielding accurate estimates of total neutron plus gamma doses in a variety of radiation fields. While 83% of all neutron measurements satisfied regulatory standards relative to reference values, only 39% of all gamma results satisfied corresponding guidelines for gamma measurements. These results indicate that continued improvement in accident dosimetry evaluation and measurement techniques is needed.« less

Progress in obtaining an absolute calibration of a total deuterium-tritium neutron yield diagnostic based on copper activationa)

NASA Astrophysics Data System (ADS)

Ruiz, C. L.; Chandler, G. A.; Cooper, G. W.; Fehl, D. L.; Hahn, K. D.; Leeper, R. J.; McWatters, B. R.; Nelson, A. J.; Smelser, R. M.; Snow, C. S.; Torres, J. A.

2012-10-01

The 350-keV Cockroft-Walton accelerator at Sandia National laboratory's Ion Beam facility is being used to calibrate absolutely a total DT neutron yield diagnostic based on the 63Cu(n,2n)62Cu(β+) reaction. These investigations have led to first-order uncertainties approaching 5% or better. The experiments employ the associated-particle technique. Deuterons at 175 keV impinge a 2.6 μm thick erbium tritide target producing 14.1 MeV neutrons from the T(d,n)4He reaction. The alpha particles emitted are measured at two angles relative to the beam direction and used to infer the neutron flux on a copper sample. The induced 62Cu activity is then measured and related to the neutron flux. This method is known as the F-factor technique. Description of the associated-particle method, copper sample geometries employed, and the present estimates of the uncertainties to the F-factor obtained are given.

Compact DD generator based in vivo neutron activation analysis (IVNAA) system to determine sodium concentrations in human bone.

PubMed

Coyne, Mychaela Dawn; Neumann, Colby R; Zhang, Xinxin; Byrne, Patrick; Liu, Yingzi; Weaver, Connie M; Nie, Linda Huiling

2018-04-16

This study presents the development of a non-invasive method for monitoring Na in human bone. Many diseases, such as hypertension and osteoporosis, are closely associated with sodium (Na) retention in the human body. Na retention is generally evaluated by calculating the difference between dietary intake and excretion. There is currently no method to directly quantify Na retained in the body. Bone is a storage for many elements, including Na, which renders bone Na an ideal biomarker to study Na metabolism and retention. Approach: A customized compact deuterium-deuterium (DD) neutron generator was used to produce neutrons for in vivo neutron activation analysis (IVNAA), with a moderator/ reflector/ shielding assembly optimized for human hand irradiation in order to maximize the thermal neutron flux inside the irradiation cave and to limit radiation exposure to the hand and the whole body. Main Results: The experimental results show that the system is able to detect sodium levels in the bone as low as 12 g Na/g dry bone with an effective dose to the body of about 27 μSv. The simulation results agree with the numbers estimated from the experiment. Significance: This is expected to be a feasible method for measuring the change of Na in bone. The low detection limit indicates this will be a useful system to study the association between Na retention and related diseases. © 2018 Institute of Physics and Engineering in Medicine.

Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code.

PubMed

Khattab, K; Sulieman, I

2009-04-01

The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

Neutronics calculation of RTP core

NASA Astrophysics Data System (ADS)

Rabir, Mohamad Hairie B.; Zin, Muhammad Rawi B. Mohamed; Karim, Julia Bt. Abdul; Bayar, Abi Muttaqin B. Jalal; Usang, Mark Dennis Anak; Mustafa, Muhammad Khairul Ariff B.; Hamzah, Na'im Syauqi B.; Said, Norfarizan Bt. Mohd; Jalil, Muhammad Husamuddin B.

2017-01-01

Reactor calculation and simulation are significantly important to ensure safety and better utilization of a research reactor. The Malaysian's PUSPATI TRIGA Reactor (RTP) achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. Since early 90s, neutronics modelling were used as part of its routine in-core fuel management activities. The are several computer codes have been used in RTP since then, based on 1D neutron diffusion, 2D neutron diffusion and 3D Monte Carlo neutron transport method. This paper describes current progress and overview on neutronics modelling development in RTP. Several important parameters were analysed such as keff, reactivity, neutron flux, power distribution and fission product build-up for the latest core configuration. The developed core neutronics model was validated by means of comparison with experimental and measurement data. Along with the RTP core model, the calculation procedure also developed to establish better prediction capability of RTP's behaviour.

Use and imaging performance of CMOS flat panel imager with LiF/ZnS(Ag) and Gadox scintillation screens for neutron radiography

NASA Astrophysics Data System (ADS)

Cha, B. K.; kim, J. Y.; Kim, T. J.; Sim, C.; Cho, G.; Lee, D. H.; Seo, C.-W.; Jeon, S.; Huh, Y.

2011-01-01

In digital neutron radiography system, a thermal neutron imaging detector based on neutron-sensitive scintillating screens with CMOS(complementary metal oxide semiconductor) flat panel imager is introduced for non-destructive testing (NDT) application. Recently, large area CMOS APS (active-pixel sensor) in conjunction with scintillation films has been widely used in many digital X-ray imaging applications. Instead of typical imaging detectors such as image plates, cooled-CCD cameras and amorphous silicon flat panel detectors in combination with scintillation screens, we tried to apply a scintillator-based CMOS APS to neutron imaging detection systems for high resolution neutron radiography. In this work, two major Gd2O2S:Tb and 6LiF/ZnS:Ag scintillation screens with various thickness were fabricated by a screen printing method. These neutron converter screens consist of a dispersion of Gd2O2S:Tb and 6LiF/ZnS:Ag scintillating particles in acrylic binder. These scintillating screens coupled-CMOS flat panel imager with 25x50mm2 active area and 48μm pixel pitch was used for neutron radiography. Thermal neutron flux with 6x106n/cm2/s was utilized at the NRF facility of HANARO in KAERI. The neutron imaging characterization of the used detector was investigated in terms of relative light output, linearity and spatial resolution in detail. The experimental results of scintillating screen-based CMOS flat panel detectors demonstrate possibility of high sensitive and high spatial resolution imaging in neutron radiography system.

Apparatus for and method of monitoring for breached fuel elements

DOEpatents

Gross, Kenny C.; Strain, Robert V.

1983-01-01

This invention teaches improved apparatus for the method of detecting a breach in cladded fuel used in a nuclear reactor. The detector apparatus uses a separate bypass loop for conveying part of the reactor coolant away from the core, and at least three separate delayed-neutron detectors mounted proximate this detector loop. The detectors are spaced apart so that the coolant flow time from the core to each detector is different, and these differences are known. The delayed-neutron activity at the detectors is a function of the dealy time after the reaction in the fuel until the coolant carrying the delayed-neutron emitter passes the respective detector. This time delay is broken down into separate components including an isotopic holdup time required for the emitter to move through the fuel from the reaction to the coolant at the breach, and two transit times required for the emitter now in the coolant to flow from the breach to the detector loop and then via the loop to the detector. At least two of these time components are determined during calibrated operation of the reactor. Thereafter during normal reactor operation, repeated comparisons are made by the method of regression approximation of the third time component for the best-fit line correlating measured delayed-neutron activity against activity that is approximated according to specific equations. The equations use these time-delay components and known parameter values of the fuel and of the part and emitting daughter isotopes.

A large 2D PSD for thermal neutron detection

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

Knott, R.B.; Watt, G.; Boldeman, J.W.

1996-12-31

A 2D PSD based on a MWPC has been constructed for a small angle neutron scattering instrument. The active area of the detector was 640 x 640 mm{sup 2}. To meet the specifications for neutron detection efficiency and spatial resolution, and to minimize parallax, the gas mixture was 190 kPa {sup 3}He plus 100 kPa CF{sub 4} and the active volume had a thickness of 30 mm. The design maximum neutron count-rate of the detector was 10{sup 5} events per second. The (calculated) neutron detection efficiency was 60% for 2{angstrom} neutrons and the (measured) neutron energy resolution on the anodemore » grid was typically 20% (fwhm). The location of a neutron detection event within the active area was determined using the wire-by-wire method: the spatial resolution (5 x 5 mm{sup 2}) was thereby defined by the wire geometry. A 16 channel charge-sensitive preamplifier/amplifier/comparator module has been developed with a channel sensitivity of 0.1 V/fC, noise linewidth of 0.4 fC (fwhm) and channel-to-channel cross-talk of less than 5%. The Proportional Counter Operating System (PCOS III) (LeCroy Corp USA) was used for event encoding. The ECL signals produced by the 16 channel modules were latched in PCOS III by a trigger pulse from the anode and the fast encoders produce a position and width for each event. The information was transferred to a UNIX workstation for accumulation and online display.« less

Thermal neutron capture and resonance integral cross sections of 45Sc

NASA Astrophysics Data System (ADS)

Van Do, Nguyen; Duc Khue, Pham; Tien Thanh, Kim; Thi Hien, Nguyen; Kim, Guinyun; Kim, Kwangsoo; Shin, Sung-Gyun; Cho, Moo-Hyun; Lee, Manwoo

2015-11-01

The thermal neutron cross section (σ0) and resonance integral (I0) of the 45Sc(n,γ)46Sc reaction have been measured relative to that of the 197Au(n,γ)198Au reaction by means of the activation method. High-purity natural scandium and gold foils without and with a cadmium cover of 0.5 mm thickness were irradiated with moderated pulsed neutrons produced from the Pohang Neutron Facility (PNF). The induced activities in the activated foils were measured with a high purity germanium (HPGe) detector. In order to improve the accuracy of the experimental results the counting losses caused by the thermal (Gth) and resonance (Gepi) neutron self-shielding, the γ-ray attenuation (Fg) and the true γ-ray coincidence summing effects were made. In addition, the effect of non-ideal epithermal spectrum was also taken into account by determining the neutron spectrum shape factor (α). The thermal neutron cross-section and resonance integral of the 45Sc(n,γ)46Sc reaction have been determined relative to the reference values of the 197Au(n,γ)198Au reaction, with σo,Au = 98.65 ± 0.09 barn and Io,Au = 1550 ± 28 barn. The present thermal neutron cross section has been determined to be σo,Sc = 27.5 ± 0.8 barn. According to the definition of cadmium cut-off energy at 0.55 eV, the present resonance integral cross section has been determined to be Io,Sc = 12.4 ± 0.7 barn. The present results are compared with literature values and discussed.

Verification of threshold activation detection (TAD) technique in prompt fission neutron detection using scintillators containing 19F

NASA Astrophysics Data System (ADS)

Sibczynski, P.; Kownacki, J.; Moszyński, M.; Iwanowska-Hanke, J.; Syntfeld-Każuch, A.; Gójska, A.; Gierlik, M.; Kaźmierczak, Ł.; Jakubowska, E.; Kędzierski, G.; Kujawiński, Ł.; Wojnarowicz, J.; Carrel, F.; Ledieu, M.; Lainé, F.

2015-09-01

In the present study ⌀ 5''× 3'' and ⌀ 2''× 2'' EJ-313 liquid fluorocarbon as well as ⌀ 2'' × 3'' BaF2 scintillators were exposed to neutrons from a 252Cf neutron source and a Sodern Genie 16GT deuterium-tritium (D+T) neutron generator. The scintillators responses to β- particles with maximum endpoint energy of 10.4 MeV from the n+19F reactions were studied. Response of a ⌀ 5'' × 3'' BC-408 plastic scintillator was also studied as a reference. The β- particles are the products of interaction of fast neutrons with 19F which is a component of the EJ-313 and BaF2 scintillators. The method of fast neutron detection via fluorine activation is already known as Threshold Activation Detection (TAD) and was proposed for photofission prompt neutron detection from fissionable and Special Nuclear Materials (SNM) in the field of Homeland Security and Border Monitoring. Measurements of the number of counts between 6.0 and 10.5 MeV with a 252Cf source showed that the relative neutron detection efficiency ratio, defined as epsilonBaF2 / epsilonEJ-313-5'', is 32.0% ± 2.3% and 44.6% ± 3.4% for front-on and side-on orientation of the BaF2, respectively. Moreover, the ⌀ 5'' EJ-313 and side-on oriented BaF2 were also exposed to neutrons from the D+T neutron generator, and the relative efficiency epsilonBaF2 / epsilonEJ-313-5'' was estimated to be 39.3%. Measurements of prompt photofission neutrons with the BaF2 detector by means of data acquisition after irradiation (out-of-beam) of nuclear material and between the beam pulses (beam-off) techniques were also conducted on the 9 MeV LINAC of the SAPHIR facility.

Design and construction of pulsed neutron diagnostic system for plasma focus device (SBUPF1).

PubMed

Moghadam, Sahar Rajabi; Davani, Fereydoon Abbasi

2010-07-01

In this paper, two designs of pulsed neutron counter structure are introduced. To increase the activation counter efficiency, BC-400 plastic scintillator plates along with silver foils are utilized. Rectangular cubic and cylindrical geometries for activation counter cell are modeled using MCNP4C code. Eventually, an optimum length of 14 cm is calculated for the detector cell and optimum numbers of 20 silver foils for rectangular cubic geometry and ten foils for cylindrical geometry have been acquired. Due to the high cost of cutting, polishing of plastics, and etc., the rectangular cubic design is found to be more economical than the other design. In order to examine the functionality and ensure the detector output and corresponding designing, neutron yield of a 2.48 kJ plasma focus device (SBUPF1) in 8 mbar pressure with removal source method for calibration was measured (3.71+/-0.32)x10(7) neutrons per shot.

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.

Progress on Establishing the Feasibility of Lead Slowing Down Spectroscopy for Direct Measurement of Plutonium in Used Fuel

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

Kulisek, Jonathan A.; Anderson, Kevin K.; Bowyer, Sonya M.

2012-07-19

Developing a method for the accurate, direct, and independent assay of the fissile isotopes in bulk materials (such as used fuel) of next-generation domestic nuclear fuel cycles is a goal of the Office of Nuclear Energy, Fuel Cycle R&D, Material Protection and Control Technology (MPACT) Campaign. To meet this goal, MPACT continues to support a multi-institutional collaboration to address the feasibility of Lead Slowing Down Spectroscopy (LSDS) as an active nondestructive assay method that has the potential to provide independent, direct measurement of Pu and U isotopic masses in used fuel with an uncertainty considerably lower than the approximately 10%more » typical of today’s confirmatory assay methods. An LSDS is comprised of a stack of lead (typically 1-6 m3) in which materials to be measured are placed in the lead and a pulse of neutrons is injected. The neutrons in this pulse lose energy due to inelastic and (subsequently) elastic scattering and the average energy of the neutrons decreases as the time increases by a well-defined relationship. In the interrogation energy region (~0.1-1000 eV) the neutrons have little energy spread (~30%) about the average neutron energy. Due to this characteristic, the energy of the (assay) neutrons can then be determined by measuring the time elapsed since the neutron pulse. By measuring the induced fission neutrons emitted from the used fuel, it is possible to determine isotopic-mass content by unfolding the unique structure of isotopic resonances across the interrogation energy region. This paper will present efforts on the development of time-spectral analysis algorithms, fast neutron detector advances, and validation and testing measurements.« less

Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

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

Root, M. A.; Menlove, H. O.; Lanza, R. C.

The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« less

Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

DOE PAGES

Root, M. A.; Menlove, H. O.; Lanza, R. C.; ...

2018-03-21

The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« less

Production of 92Y via the 92Zr(n,p) reaction using the C(d,n) accelerator neutron source

NASA Astrophysics Data System (ADS)

Kin, Tadahiro; Sanzen, Yukimasa; Kamida, Masaki; Watanabe, Yukinobu; Itoh, Masatoshi

2017-09-01

We have proposed a new method of producing medical radioisotope 92Y as a candidate of alternatives of 111In bioscan prior to 90Y ibritumomab tiuxetan treatment. The 92Y isotope is produced via the 92Zr (n,p) reaction using accelerator neutrons generated by the interaction of deuteron beams with carbon. A feasibility experiment was performed at Cyclotron and Radioisotope Center, Tohoku University. A carbon thick target was irradiated by 20-MeV deuterons to produce accelerator neutrons. The thick target neutron yield (TTNY) was measured by using the multiple foils activation method. The foils were made of Al, Fe, Co, Ni, Zn, Zr, Nb, and Au. The production amount of 92Y and induced impurities were estimated by simulation with the measured TTNY and the JENDL-4.0 nuclear data.

Development and Characterization of a High Sensitivity Segmented Fast Neutron Spectrometer (FaNS-2)

PubMed Central

Langford, T.J.; Beise, E.J.; Breuer, H.; Heimbach, C.R.; Ji, G.; Nico, J.S.

2016-01-01

We present the development of a segmented fast neutron spectrometer (FaNS-2) based upon plastic scintillator and 3He proportional counters. It was designed to measure both the flux and spectrum of fast neutrons in the energy range of few MeV to 1 GeV. FaNS-2 utilizes capture-gated spectroscopy to identify neutron events and reject backgrounds. Neutrons deposit energy in the plastic scintillator before capturing on a 3He nucleus in the proportional counters. Segmentation improves neutron energy reconstruction while the large volume of scintillator increases sensitivity to low neutron fluxes. A main goal of its design is to study comparatively low neutron fluxes, such as cosmogenic neutrons at the Earth's surface, in an underground environment, or from low-activity neutron sources. In this paper, we present details of its design and construction as well as its characterization with a calibrated 252Cf source and monoenergetic neutron fields of 2.5 MeV and 14 MeV. Detected monoenergetic neutron spectra are unfolded using a Singular Value Decomposition method, demonstrating a 5% energy resolution at 14 MeV. Finally, we discuss plans for measuring the surface and underground cosmogenic neutron spectra with FaNS-2. PMID:27226807

Analysis of calibration data for the uranium active neutron coincidence counting collar with attention to errors in the measured neutron coincidence rate

DOE PAGES

Croft, Stephen; Burr, Thomas Lee; Favalli, Andrea; ...

2015-12-10

We report that the declared linear density of 238U and 235U in fresh low enriched uranium light water reactor fuel assemblies can be verified for nuclear safeguards purposes using a neutron coincidence counter collar in passive and active mode, respectively. The active mode calibration of the Uranium Neutron Collar – Light water reactor fuel (UNCL) instrument is normally performed using a non-linear fitting technique. The fitting technique relates the measured neutron coincidence rate (the predictor) to the linear density of 235U (the response) in order to estimate model parameters of the nonlinear Padé equation, which traditionally is used to modelmore » the calibration data. Alternatively, following a simple data transformation, the fitting can also be performed using standard linear fitting methods. This paper compares performance of the nonlinear technique to the linear technique, using a range of possible error variance magnitudes in the measured neutron coincidence rate. We develop the required formalism and then apply the traditional (nonlinear) and alternative approaches (linear) to the same experimental and corresponding simulated representative datasets. Lastly, we find that, in this context, because of the magnitude of the errors in the predictor, it is preferable not to transform to a linear model, and it is preferable not to adjust for the errors in the predictor when inferring the model parameters« less

Infra-red signature neutron detector

DOEpatents

Bell, Zane William [Oak Ridge, TN; Boatner, Lynn Allen [Oak Ridge, TN

2009-10-13

A method of detecting an activator, the method including impinging with an activator a receptor material that includes a photoluminescent material that generates infrared radiation and generation a by-product of a nuclear reaction due to the activator impinging the receptor material. The method further includes generating light from the by-product via the Cherenkov effect, wherein the light activates the photoluminescent material so as to generate the infrared radiation. Identifying a characteristic of the activator based on the infrared radiation.

The Multi-Step CADIS method for shutdown dose rate calculations and uncertainty propagation

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

Ibrahim, Ahmad M.; Peplow, Douglas E.; Grove, Robert E.

2015-12-01

Shutdown dose rate (SDDR) analysis requires (a) a neutron transport calculation to estimate neutron flux fields, (b) an activation calculation to compute radionuclide inventories and associated photon sources, and (c) a photon transport calculation to estimate final SDDR. In some applications, accurate full-scale Monte Carlo (MC) SDDR simulations are needed for very large systems with massive amounts of shielding materials. However, these simulations are impractical because calculation of space- and energy-dependent neutron fluxes throughout the structural materials is needed to estimate distribution of radioisotopes causing the SDDR. Biasing the neutron MC calculation using an importance function is not simple becausemore » it is difficult to explicitly express the response function, which depends on subsequent computational steps. Furthermore, the typical SDDR calculations do not consider how uncertainties in MC neutron calculation impact SDDR uncertainty, even though MC neutron calculation uncertainties usually dominate SDDR uncertainty.« less

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

Dickens, Peter T.; Marcial, Jose; McCloy, John

In this study, LiAlO 2 crystals doped with rare-earth elements and Ti were produced by the CZ method and spectroscopic and neutron detection properties were investigated. Photoluminescence revealed no clear luminescent activation of LiAlO 2 by the rare-earth dopants though some interesting luminescence was observed from secondary phases within the crystal. Gamma-ray pulse height spectra collected using a 137Cs source exhibited only a Compton edge for the crystals. Neutron modeling using Monte Carlo N-Particle Transport Code revealed most neutrons used in the detection setup are thermalized, and while using natural lithium in the crystal growth, which contains 7.6% 6Li, amore » 10 mm Ø by 10 mm sample of LiAlO 2 has a 70.7% intrinsic thermal neutron capture efficiency. Furthermore, the pulse height spectra collected using a 241Am-Be neutron source demonstrated a distinct neutron peak.« less

Monte Carlo source simulation technique for solution of interference reactions in INAA experiments: a preliminary report

NASA Astrophysics Data System (ADS)

Allaf, M. Athari; Shahriari, M.; Sohrabpour, M.

2004-04-01

A new method using Monte Carlo source simulation of interference reactions in neutron activation analysis experiments has been developed. The neutron spectrum at the sample location has been simulated using the Monte Carlo code MCNP and the contributions of different elements to produce a specified gamma line have been determined. The produced response matrix has been used to measure peak areas and the sample masses of the elements of interest. A number of benchmark experiments have been performed and the calculated results verified against known values. The good agreement obtained between the calculated and known values suggests that this technique may be useful for the elimination of interference reactions in neutron activation analysis.

Neutron radiation tolerance of Au-activated silicon

NASA Technical Reports Server (NTRS)

Joyner, W. T.

1987-01-01

Double injection devices prepared by the introduction of deep traps, using the Au activation method have been found to tolerate gamma irradiation into the Gigarad (Si) region without significant degradation of operating characteristics. Silicon double injection devices, using deep levels creacted by Au diffusion, can tolerate fast neutron irradiation up to 10 to the 15th n/sq cm. Significant parameter degradation occurs at 10 to the 16th n/sq cm. However, since the actual doping of the basic material begins to change as a result of the transmutation of silicon into phosphorus for neutron fluences greater than 10 to the 17th/sq cm, the radiation tolerance of these devices is approaching the limit possible for any device based on initially doped silicon.

Measurements of neutron distribution in neutrons-gamma-rays mixed field using imaging plate for neutron capture therapy.

PubMed

Tanaka, Kenichi; Endo, Satoru; Hoshi, Masaharu

2010-01-01

The imaging plate (IP) technique is tried to be used as a handy method to measure the spatial neutron distribution via the (157)Gd(n,gamma)(158)Gd reaction for neutron capture therapy (NCT). For this purpose, IP is set in a water phantom and irradiated in a mixed field of neutrons and gamma-rays. The Hiroshima University Radiobiological Research Accelerator is utilized for this experiment. The neutrons are moderated with 20-cm-thick D(2)O to obtain suitable neutron field for NCT. The signal for IP doped with Gd as a neutron-response enhancer is subtracted with its contribution by gamma-rays, which was estimated using IP without Gd. The gamma-ray response of Gd-doped IP to non-Gd IP is set at 1.34, the value measured for (60)Co gamma-rays, in estimating the gamma-ray contribution to Gd-doped IP signal. Then measured distribution of the (157)Gd(n,gamma)(158)Gd reaction rate agrees within 10% with the calculated value based on the method that has already been validated for its reproducibility of Au activation. However, the evaluated distribution of the (157)Gd(n,gamma)(158)Gd reaction rate is so sensitive to gamma-ray energy, e.g. the discrepancy of the (157)Gd(n,gamma)(158)Gd reaction rate between measurement and calculation becomes 30% for the photon energy change from 33keV to 1.253MeV.

Beam Characterization at the Neutron Radiography Facility

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

Sarah Morgan; Jeffrey King

The quality of a neutron imaging beam directly impacts the quality of radiographic images produced using that beam. Fully characterizing a neutron beam, including determination of the beam’s effective length-to-diameter ratio, neutron flux profile, energy spectrum, image quality, and beam divergence, is vital for producing quality radiographic images. This project characterized the east neutron imaging beamline at the Idaho National Laboratory Neutron Radiography Reactor (NRAD). The experiments which measured the beam’s effective length-to-diameter ratio and image quality are based on American Society for Testing and Materials (ASTM) standards. An analysis of the image produced by a calibrated phantom measured themore » beam divergence. The energy spectrum measurements consist of a series of foil irradiations using a selection of activation foils, compared to the results produced by a Monte Carlo n-Particle (MCNP) model of the beamline. Improvement of the existing NRAD MCNP beamline model includes validation of the model’s energy spectrum and the development of enhanced image simulation methods. The image simulation methods predict the radiographic image of an object based on the foil reaction rate data obtained by placing a model of the object in front of the image plane in an MCNP beamline model.« less

Accurate determination of arsenic in arsenobetaine standard solutions of BCR-626 and NMIJ CRM 7901-a by neutron activation analysis coupled with internal standard method.

PubMed

Miura, Tsutomu; Chiba, Koichi; Kuroiwa, Takayoshi; Narukawa, Tomohiro; Hioki, Akiharu; Matsue, Hideaki

2010-09-15

Neutron activation analysis (NAA) coupled with an internal standard method was applied for the determination of As in the certified reference material (CRM) of arsenobetaine (AB) standard solutions to verify their certified values. Gold was used as an internal standard to compensate for the difference of the neutron exposure in an irradiation capsule and to improve the sample-to-sample repeatability. Application of the internal standard method significantly improved linearity of the calibration curve up to 1 microg of As, too. The analytical reliability of the proposed method was evaluated by k(0)-standardization NAA. The analytical results of As in AB standard solutions of BCR-626 and NMIJ CRM 7901-a were (499+/-55)mgkg(-1) (k=2) and (10.16+/-0.15)mgkg(-1) (k=2), respectively. These values were found to be 15-20% higher than the certified values. The between-bottle variation of BCR-626 was much larger than the expanded uncertainty of the certified value, although that of NMIJ CRM 7901-a was almost negligible. Copyright (c) 2010 Elsevier B.V. All rights reserved.

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.

Spectroscopic and neutron detection properties of rare earth and titanium doped LiAlO 2 single crystals

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

Dickens, Peter T.; Marcial, José; McCloy, John

In this study, LiAlO2 crystals doped with rare-earth elements and Ti were produced by the CZ method and spectroscopic and neutron detection properties were investigated. Photoluminescence revealed no clear luminescent activation of LiAlO2 by the rare-earth dopants though some interesting luminescence was observed from secondary phases within the crystal. Gamma-ray pulse height spectra collected using a 137Cs source exhibited only a Compton edge for the crystals. Neutron modeling using Monte Carlo N-Particle Transport Code revealed most neutrons used in the detection setup are thermalized, and while using natural lithium in the crystal growth, which contains 7.6 % 6Li, a 10more » mm Ø by 10 mm sample of LiAlO2 has a 70.7 % intrinsic thermal neutron capture efficiency. Furthermore, the pulse height spectra collected using a 241Am-Be neutron source demonstrated a distinct neutron peak.« less

Spectroscopic and neutron detection properties of rare earth and titanium doped LiAlO 2 single crystals

DOE PAGES

Dickens, Peter T.; Marcial, Jose; McCloy, John; ...

2017-05-17

In this study, LiAlO 2 crystals doped with rare-earth elements and Ti were produced by the CZ method and spectroscopic and neutron detection properties were investigated. Photoluminescence revealed no clear luminescent activation of LiAlO 2 by the rare-earth dopants though some interesting luminescence was observed from secondary phases within the crystal. Gamma-ray pulse height spectra collected using a 137Cs source exhibited only a Compton edge for the crystals. Neutron modeling using Monte Carlo N-Particle Transport Code revealed most neutrons used in the detection setup are thermalized, and while using natural lithium in the crystal growth, which contains 7.6% 6Li, amore » 10 mm Ø by 10 mm sample of LiAlO 2 has a 70.7% intrinsic thermal neutron capture efficiency. Furthermore, the pulse height spectra collected using a 241Am-Be neutron source demonstrated a distinct neutron peak.« less

Crystallographic changes in lead zirconate titanate due to neutron irradiation

DOE PAGES

Henriques, Alexandra; Graham, Joseph T.; Landsberger, Sheldon; ...

2014-11-17

Piezoelectric and ferroelectric materials are useful as the active element in non-destructive monitoring devices for high-radiation areas. Here, crystallographic structural refinement (i.e., the Rietveld method) is used to quantify the type and extent of structural changes in PbZr 0 .5Ti 0 .5O 3 after exposure to a 1 MeV equivalent neutron fluence of 1.7 × 10 15 neutrons/cm 2. The results show a measurable decrease in the occupancy of Pb and O due to irradiation, with O vacancies in the tetragonal phase being created preferentially on one of the two O sites. The results demonstrate a method by which themore » effects of radiation on crystallographic structure may be investigated.« less

Fission and activation of uranium by fusion-plasma neutrons

NASA Technical Reports Server (NTRS)

Lee, J. H.; Hohl, F.; Mcfarland, D. R.

1978-01-01

Fusion-fission hybrid reactors are discussed in terms of two main purposes: to breed fissile materials (Pu 233 and Th 233 from U 238 or Th 232) for use in low-reactivity breeders, and to produce tritium from lithium to refuel fusion plasma cores. Neutron flux generation is critical for both processes. Various methods for generating the flux are described, with attention to new geometries for multiple plasma focus arrays, e.g., hypocycloidal pinch and staged plasma focus devices. These methods are evaluated with reference to their applicability to D-D fusion reactors, which will ensure a virtually unlimited energy supply. Accurate observations of the neutron flux from such schemes are obtained by using different target materials in the plasma focus.

Neutron Productions from thin Be target irradiated by 50 MeV/u 238U beam

NASA Astrophysics Data System (ADS)

Lee, Hee-Seock; Oh, Joo-Hee; Jung, Nam-Suk; Oranj, Leila Mokhtari; Nakao, Noriaki; Uwamino, Yoshitomo

2017-09-01

Neutrons generated from thin beryllium target by 50 MeV/u 238U beam were measured using activation analysis at 15, 30, 45, and 90 degrees from the beam direction. A 0.085 mm-thick Be stripper of RIBF was used as the neutron generating target. Activation detectors of bismuth, cobalt, and aluminum were placed out of the stripper chamber. The threshold reactions of 209Bi(n, xn)210-xBi(x=4 8), 59Co(n, xn)60-xCO(x=2 5), 59Co(n, 2nα)54Mn, 27Al(n, α)24Na, and 27Al(n,2nα)22Na were applied to measure the production rates of radionuclides. The neutron spectra were obtained using an unfolding method with the SAND-II code. All of production rates and neutron spectra were compared with the calculated results using Monte Carlo codes, the PHITS and the FLUKA. The FLUKA results showed better agreement with the measurements than the PHITS. The discrepancy between the measurements and the calculations were discussed.

Completely non-destructive elemental analysis of bulky samples by PGAA

NASA Astrophysics Data System (ADS)

Oura, Y.; Nakahara, H.; Sueki, K.; Sato, W.; Saito, A.; Tomizawa, T.; Nishikawa, T.

1999-01-01

NBAA (neutron beam activation analysis), which is a combination of PGAA and INAA by a single neutron irradiation, using an internal monostandard method is proposed as a very unique and promising method for the elemental analysis of voluminous and invaluable archaeological samples which do not allow even a scrape of the surface. It was applied to chinawares, Sueki ware, and bronze mirrors, and proved to be a very effective method for nondestructive analysis of not only major elements but also some minor elements such as boron that help solve archaeological problems of ears and sites of their production.

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

Boatner, Lynn A.; Comer, Eleanor P.; Wright, Gomez W.

Monovalent alkali halides such as NaI, CsI, and LiI are widely used as inorganic scintillators for radiation detection due to their light yield, the capability for the growth of large single crystals, relatively low cost, and other favorable characteristics. These materials are frequently activated through the addition of small amounts (e.g., a few hundred ppm) of elements such as thallium - or sodium in the case of CsI. The monovalent alkali halide scintillators can also be activated with low concentrations of Eu 2+, however Eu activation has previously not been widely employed due to the non-uniform segregation of the divalentmore » Eu dopant that leads to the formation of unwanted phases during Bridgman or other solidification crystal-growth methods. Specifically, for Eu concentrations near and above ~0.5%, Suzuki Phase precipitates form in the course of the melt-growth process, and these Suzuki Phase particles scatter the scintillation light. This adversely affects the scintillator performance via reduction in the optical transmission of the material, and depending on the crystal thickness and precipitated-particle concentration, this reduction can occur up to the point of opacity. Here we describe a post-growth process for the removal of Suzuki Phase precipitates from single crystals of the neutron scintillator LiI activated with Eu 2+ at concentrations up to and in excess of 3 wt.%, and we correlate the resulting neutron-detection performance with the thermal processing methods used to remove the Suzuki Phase particles. Furthermore, the resulting improved scintillator properties using increased Eu activator levels are applicable to neutron imaging and active interrogation systems, and pulse-height gamma-ray spectroscopy rather than pulse-shape discrimination can be used to discriminate between gamma ray and neutron interaction events.« less

The energy spectrum of neutrons from 7Li(d,n)8Be reaction at deuteron energy 2.9 MeV

NASA Astrophysics Data System (ADS)

Mitrofanov, Konstantin V.; Piksaikin, Vladimir M.; Zolotarev, Konstantin I.; Egorov, Andrey S.; Gremyachkin, Dmitrii E.

2017-09-01

The neutron beams generated at the electrostatic accelerators using nuclear reactions T(p,n)3He, D(d,n)3He, 7Li(p,n)7Be, T(d,n)4He, 7Li(d,n)8Be, 9Be(d,n)10B are widely used in neutron physics and in many practical applications. Among these reactions the least studied reactions are 7Li(d,n)8Be and 9Be(d,n)10B. The present work is devoted to the measurement of the neutron spectrum from 7Li(d,n)8Be reaction at 0∘ angle to the deuteron beam axis on the electrostatic accelerator Tandetron (JSC "SSC RF - IPPE") using activation method and a stilbene crystal scintillation detector. The first time ever 7Li(d,n)8Be reaction was measured by activation method. The target was a thick lithium layer on metallic backing. The energy of the incident deuteron was 2.9 MeV. As activation detectors a wide range of nuclear reactions were used: 27Al(n,p)27Mg, 27Al(n,α)24Na, 113In(n,n')113mIn, 115In(n,n')115mIn, 115In(n,γ)116mIn, 58Ni(n,p)58mCo, 58Ni(n,2n)57Ni, 197Au(n,γ)198Au, 197Au(n,2n)196Au, 59Co(n,p)59Fe, 59Co(n,2n)58m+gCo, 59Co (n,g)60Co. Measurement of the induced gamma-activity was carried out using HPGe detector Canberra GX5019 [1]. The up-to-date evaluations of the cross sections for these reactions were used in processing of the data. The program STAYSL was used to unfold the energy spectra. The neutron spectra obtained by activation detectors is consistent with the corresponding data measured by a stilbene crystal scintillation detector within their uncertainties.

Standard-less analysis of Zircaloy clad samples by an instrumental neutron activation method

NASA Astrophysics Data System (ADS)

Acharya, R.; Nair, A. G. C.; Reddy, A. V. R.; Goswami, A.

2004-03-01

A non-destructive method for analysis of irregular shape and size samples of Zircaloy has been developed using the recently standardized k0-based internal mono standard instrumental neutron activation analysis (INAA). The samples of Zircaloy-2 and -4 tubes, used as fuel cladding in Indian boiling water reactors (BWR) and pressurized heavy water reactors (PHWR), respectively, have been analyzed. Samples weighing in the range of a few tens of grams were irradiated in the thermal column of Apsara reactor to minimize neutron flux perturbations and high radiation dose. The method utilizes in situ relative detection efficiency using the γ-rays of selected activation products in the sample for overcoming γ-ray self-attenuation. Since the major and minor constituents (Zr, Sn, Fe, Cr and/or Ni) in these samples were amenable to NAA, the absolute concentrations of all the elements were determined using mass balance instead of using the concentration of the internal mono standard. Concentrations were also determined in a smaller size Zircaloy-4 sample by irradiating in the core position of the reactor to validate the present methodology. The results were compared with literature specifications and were found to be satisfactory. Values of sensitivities and detection limits have been evaluated for the elements analyzed.

Measurement of Insulation Compaction in the Cryogenic Fuel Tanks at Kennedy Space Center by Fast/Thermal Neutron Techniques

NASA Technical Reports Server (NTRS)

Livingston, R. A.; Schweitzer, J. S.; Parsons, Ann M.; Arens, Ellen E.

2010-01-01

The liquid hydrogen and oxygen cryogenic storage tanks at John F. Kennedy Space Center (KSC) use expanded perlite as thermal insulation. Th ere is evidence that some of the perlite has compacted over time, com promising the thermal performance and possibly also structural integr ity of the tanks. Therefore an Non-destructive Testing (NDT) method for measuring the perlite density or void fraction is urgently needed. Methods based on neutrons are good candidates because they can readil y penetrate through the 1.75 cm outer steel shell and through the ent ire 120 cm thickness of the perlite zone. Neutrons interact with the nuclei of materials to produce characteristic gamma rays which are the n detected. The gamma ray signal strength is proportional to the atom ic number density. Consequently, if the perlite is compacted then the count rates in the individual peaks in the gamma ray spectrum will i ncrease. Perlite is a feldspathic volcanic rock made up of the major elements Si, AI, Na, K and 0 along with some water. With commercially available portable neutron generators it is possible to produce simul taneously fluxes of neutrons in two energy ranges: fast (14 MeV) and thermal (25 meV). Fast neutrons produce gamma rays by inelastic scatt ering which is sensitive to Fe and O. Thermal neutrons produce gamma rays by radiative capture in prompt gamma neutron activation (PGNA) and this is sensitive to Si, AI, Na, Kand H. Thus the two energy ranges produce complementary information. The R&D program has three phases: numerical simulations of neutron and gamma ray transport with MCNP s oftware, evaluation of the system in the laboratory on test articles and finally mapping of the perlite density in the cryogenic tanks at KSC. The preliminary MCNP calculations have shown that the fast/therma l neutron NDT method is capable of distinguishing between expanded an d compacted perlite with excellent statistics.

In situ calibration of neutron activation system on the large helical device

NASA Astrophysics Data System (ADS)

Pu, N.; Nishitani, T.; Isobe, M.; Ogawa, K.; Kawase, H.; Tanaka, T.; Li, S. Y.; Yoshihashi, S.; Uritani, A.

2017-11-01

In situ calibration of the neutron activation system on the Large Helical Device (LHD) was performed by using an intense 252Cf neutron source. To simulate a ring-shaped neutron source, we installed a railway inside the LHD vacuum vessel and made a train loaded with the 252Cf source run along a typical magnetic axis position. Three activation capsules loaded with thirty pieces of indium foils stacked with total mass of approximately 18 g were prepared. Each capsule was irradiated over 15 h while the train was circulating. The activation response coefficient (9.4 ± 1.2) × 10-8 of 115In(n, n')115mIn reaction obtained from the experiment is in good agreement with results from three-dimensional neutron transport calculations using the Monte Carlo neutron transport simulation code 6. The activation response coefficients of 2.45 MeV birth neutron and secondary 14.1 MeV neutron from deuterium plasma were evaluated from the activation response coefficient obtained in this calibration experiment with results from three-dimensional neutron calculations using the Monte Carlo neutron transport simulation code 6.

Neutron Spectrum Measurements from Irradiations at NCERC

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

Jackman, Kevin Richard; Mosby, Michelle A.; Bredeweg, Todd Allen

2015-04-15

Several irradiations have been conducted on assemblies (COMET/ZEUS and Flattop) at the National Criticality Experiments Research Center (NCERC) located at the Nevada National Security Site (NNSS). Configurations of the assemblies and irradiated materials changed between experiments. Different metallic foils were analyzed using the radioactivation method by gamma-ray spectrometry to understand/characterize the neutron spectra. Results of MCNP calculations are shown. It was concluded that MCNP simulated spectra agree with experimental measurements, with the caveats that some data are limited by statistics at low-energies and some activation foils have low activities.

Measurements of fusion neutron yields by neutron activation technique: Uncertainty due to the uncertainty on activation cross-sections

NASA Astrophysics Data System (ADS)

Stankunas, Gediminas; Batistoni, Paola; Sjöstrand, Henrik; Conroy, Sean; JET Contributors

2015-07-01

The neutron activation technique is routinely used in fusion experiments to measure the neutron yields. This paper investigates the uncertainty on these measurements as due to the uncertainties on dosimetry and activation reactions. For this purpose, activation cross-sections were taken from the International Reactor Dosimetry and Fusion File (IRDFF-v1.05) in 640 groups ENDF-6 format for several reactions of interest for both 2.5 and 14 MeV neutrons. Activation coefficients (reaction rates) have been calculated using the neutron flux spectra at JET vacuum vessel, both for DD and DT plasmas, calculated by MCNP in the required 640-energy group format. The related uncertainties for the JET neutron spectra are evaluated as well using the covariance data available in the library. These uncertainties are in general small, but not negligible when high accuracy is required in the determination of the fusion neutron yields.

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.

DD fusion neutron production at UW-Madison using IEC devices

NASA Astrophysics Data System (ADS)

Fancher, Aaron; Michalak, Matt; Kulcinski, Gerald; Santarius, John; Bonomo, Richard

2017-10-01

An inertial electrostatic confinement (IEC) device using spherical, gridded electrodes at high voltage accelerates deuterium ions, allowing for neutrons to be produced within the device from DD fusion reactions. The effects of the device cathode voltage (30-170 kV), current (30-100 mA), and pressure (0.15-1.25 mTorr) on the neutron production rate have been measured. New high voltage capabilities have resulted in the achievement of a steady state neutron production rate of 3.3x108 n/s at 175 kV, 100 mA, and 1.0 mTorr of deuterium. Applications of IEC devices include the production of DD neutrons to detect chemical explosives and special nuclear materials using active interrogation methods. Research supported by US Dept. of Homeland Security Grant 2015-DN-077-AR1095 and the Grainger Foundation.

Ion source and beam guiding studies for an API neutron generator

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

Sy, A.; Ji, Q.; Persaud, A.

2013-04-19

Recently developed neutron imaging methods require high neutron yields for fast imaging times and small beam widths for good imaging resolution. For ion sources with low current density to be viable for these types of imaging methods, large extraction apertures and beam focusing must be used. We present recent work on the optimization of a Penning-type ion source for neutron generator applications. Two multi-cusp magnet configurations have been tested and are shown to increase the extracted ion current density over operation without multi-cusp magnetic fields. The use of multi-cusp magnetic confinement and gold electrode surfaces have resulted in increased ionmore » current density, up to 2.2 mA/cm{sup 2}. Passive beam focusing using tapered dielectric capillaries has been explored due to its potential for beam compression without the cost and complexity issues associated with active focusing elements. Initial results from first experiments indicate the possibility of beam compression. Further work is required to evaluate the viability of such focusing methods for associated particle imaging (API) systems.« less

Least-Squares Neutron Spectral Adjustment with STAYSL PNNL

NASA Astrophysics Data System (ADS)

Greenwood, L. R.; Johnson, C. D.

2016-02-01

The STAYSL PNNL computer code, a descendant of the STAY'SL code [1], performs neutron spectral adjustment of a starting neutron spectrum, applying a least squares method to determine adjustments based on saturated activation rates, neutron cross sections from evaluated nuclear data libraries, and all associated covariances. STAYSL PNNL is provided as part of a comprehensive suite of programs [2], where additional tools in the suite are used for assembling a set of nuclear data libraries and determining all required corrections to the measured data to determine saturated activation rates. Neutron cross section and covariance data are taken from the International Reactor Dosimetry File (IRDF-2002) [3], which was sponsored by the International Atomic Energy Agency (IAEA), though work is planned to update to data from the IAEA's International Reactor Dosimetry and Fusion File (IRDFF) [4]. The nuclear data and associated covariances are extracted from IRDF-2002 using the third-party NJOY99 computer code [5]. The NJpp translation code converts the extracted data into a library data array format suitable for use as input to STAYSL PNNL. The software suite also includes three utilities to calculate corrections to measured activation rates. Neutron self-shielding corrections are calculated as a function of neutron energy with the SHIELD code and are applied to the group cross sections prior to spectral adjustment, thus making the corrections independent of the neutron spectrum. The SigPhi Calculator is a Microsoft Excel spreadsheet used for calculating saturated activation rates from raw gamma activities by applying corrections for gamma self-absorption, neutron burn-up, and the irradiation history. Gamma self-absorption and neutron burn-up corrections are calculated (iteratively in the case of the burn-up) within the SigPhi Calculator spreadsheet. The irradiation history corrections are calculated using the BCF computer code and are inserted into the SigPhi Calculator workbook for use in correcting the measured activities. Output from the SigPhi Calculator is automatically produced, and consists of a portion of the STAYSL PNNL input file data that is required to run the spectral adjustment calculations. Within STAYSL PNNL, the least-squares process is performed in one step, without iteration, and provides rapid results on PC platforms. STAYSL PNNL creates multiple output files with tabulated results, data suitable for plotting, and data formatted for use in subsequent radiation damage calculations using the SPECTER computer code (which is not included in the STAYSL PNNL suite). All components of the software suite have undergone extensive testing and validation prior to release and test cases are provided with the package.

A comparison of two methods of in vivo dosimetry for a high energy neutron beam.

PubMed

Blake, S W; Bonnett, D E; Finch, J

1990-06-01

Two methods of in vivo dosimetry have been compared in a high energy neutron beam. These were activation dosimetry and thermoluminescence dosimetry (TLD). Their suitability was determined by comparison with estimates of total dose, obtained using a tissue equivalent ionization chamber. Measurements were made on the central axis and a profile of a 10 x 10 cm square field and also behind a shielding block in order to simulate conditions of clinical use. The TLD system was found to provide the best estimate of total dose.

Dual-fission chamber and neutron beam characterization for fission product yield measurements using monoenergetic neutrons

NASA Astrophysics Data System (ADS)

Bhatia, C.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.; Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rundberg, R. S.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Macri, R.; Ryan, C.; Sheets, S. A.; Stoyer, M. A.; Tonchev, A. P.

2014-09-01

A program has been initiated to measure the energy dependence of selected high-yield fission products used in the analysis of nuclear test data. We present out initial work of neutron activation using a dual-fission chamber with quasi-monoenergetic neutrons and gamma-counting method. Quasi-monoenergetic neutrons of energies from 0.5 to 15 MeV using the TUNL 10 MV FM tandem to provide high-precision and self-consistent measurements of fission product yields (FPY). The final FPY results will be coupled with theoretical analysis to provide a more fundamental understanding of the fission process. To accomplish this goal, we have developed and tested a set of dual-fission ionization chambers to provide an accurate determination of the number of fissions occurring in a thick target located in the middle plane of the chamber assembly. Details of the fission chamber and its performance are presented along with neutron beam production and characterization. Also presented are studies on the background issues associated with room-return and off-energy neutron production. We show that the off-energy neutron contribution can be significant, but correctable, while room-return neutron background levels contribute less than <1% to the fission signal.

Critical Dimensions of Water-tamped Slabs and Spheres of Active Material

DOE R&D Accomplishments Database

Greuling, E.; Argo, H.: Chew, G.; Frankel, M. E.; Konopinski, E.J.; Marvin, C.; Teller, E.

1946-08-06

The magnitude and distribution of the fission rate per unit area produced by three energy groups of moderated neutrons reflected from a water tamper into one side of an infinite slab of active material is calculated approximately in section II. This rate is directly proportional to the current density of fast neutrons from the active material incident on the water tamper. The critical slab thickness is obtained in section III by solving an inhomogeneous transport integral equation for the fast-neutron current density into the tamper. Extensive use is made of the formulae derived in "The Mathematical Development of the End-Point Method" by Frankel and Goldberg. In section IV slight alterations in the theory outlined in sections II and III were made so that one could approximately compute the critical radius of a water-tamper sphere of active material. The derived formulae were applied to calculate the critical dimensions of water-tamped slabs and spheres of solid UF{sub 6} leaving various (25) isotope enrichment fractions. Decl. Dec. 16, 1955.

Microwave acid digestion and preconcentration neutron activation analysis of biological and diet samples for iodine.

PubMed

Rao, R R; Chatt, A

1991-07-01

A simple preconcentration neutron activation analysis (PNAA) method has been developed for the determination of low levels of iodine in biological and nutritional materials. The method involves dissolution of the samples by microwave digestion in the presence of acids in closed Teflon bombs and preconcentration of total iodine, after reduction to iodide with hydrazine sulfate, by coprecipitation with bismuth sulfide. The effects of different factors such as acidity, time for complete precipitation, and concentrations of bismuth, sulfide, and diverse ions on the quantitative recovery of iodide have been studied. The absolute detection limit of the PNAA method is 5 ng of iodine. Precision of measurement, expressed in terms of relative standard deviation, is about 5% at 100 ppb and 10% at 20 ppb levels of iodine. The PNAA method has been applied to several biological reference materials and total diet samples.

Determination of molybdenum in sea and estuarine water with BETA-naphthoin oxime and neutron activation

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

Kuathilake, A.I.; Chatt, A.

1980-05-01

An analytical method has been developed for the determination of submicrogram quantities of molybdenum in sea and esturaine water. The method consists of preconcentration of molybdenum with BETA-naphthoin oxime followed by the determination of the element employing neutron activation analysis. Various factors that can influence yield and selectivity of the preconcentration process have been investigated in detail. A comparison study between ..cap alpha..-benzoin oxime and BETA-naphthoin oxime in preconcentrating molybdenum has been carried out using a standard steel sample. The method has been applied to determine molybdenum content of sea and estuarine water. A detection limit of 0.32 ..mu..g Momore » L/sup -1/ seawater has been acheived. The precision and accuracy of the method have been evaluated using an intercomparison fresh water and a biological standard reference material. 1 figure, 9 tables.« less

Comparison between experimental data and Monte-Carlo simulations of neutron production in spallation reactions of 0.7-1.5 GeV protons on a thick, lead target

NASA Astrophysics Data System (ADS)

Krása, A.; Majerle, M.; Krízek, F.; Wagner, V.; Kugler, A.; Svoboda, O.; Henzl, V.; Henzlová, D.; Adam, J.; Caloun, P.; Kalinnikov, V. G.; Krivopustov, M. I.; Stegailov, V. I.; Tsoupko-Sitnikov, V. M.

2006-05-01

Relativistic protons with energies 0.7-1.5 GeV interacting with a thick, cylindrical, lead target, surrounded by a uranium blanket and a polyethylene moderator, produced spallation neutrons. The spatial and energetic distributions of the produced neutron field were measured by the Activation Analysis Method using Al, Au, Bi, and Co radio-chemical sensors. The experimental yields of isotopes induced in the sensors were compared with Monte-Carlo calculations performed with the MCNPX 2.4.0 code.

Application of a Bonner sphere spectrometer for determination of the energy spectra of neutrons generated by ≈1 MJ plasma focus.

PubMed

Králík, M; Krása, J; Velyhan, A; Scholz, M; Ivanova-Stanik, I M; Bienkowska, B; Miklaszewski, R; Schmidt, H; Řezáč, K; Klír, D; Kravárik, J; Kubeš, P

2010-11-01

The spectra of neutrons outside the plasma focus device PF-1000 with an upper energy limit of ≈1 MJ was measured using a Bonner spheres spectrometer in which the active detector of thermal neutrons was replaced by nine thermoluminescent chips. As an a priori spectrum for the unfolding procedure, the spectrum calculated by means of the Monte Carlo method with a simplified model of the discharge chamber was selected. Differences between unfolded and calculated spectra are discussed with respect to properties of the discharge vessel and the laboratory layout.

Standardization of the neutron probe for the assessment of masonry deterioration

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

Livingston, R.A.

1992-01-01

The repair of the infrastructure will require nondestructive methods to assess the condition of existing buildings and other structures, many of which are constructed of masonry. One possible technology is the neutron probe, a prompt-gamma neutron activation (PGNA) technique that can perform non- destructive elemental analyses in the field. It is based on a very low intensity [sup 252]Cf neutron source and a high-purity germanium detector for the gamma rays emitted by neutron capture within the material. The thermal neutron capture cross sections for hydrogen and chlorine are very large, and in masonry, these elements are found primarily in moisturemore » and chlorides. These are major causes of deterioration in porous materials such as brick masonry. The moisture damages the material through expansive stresses during freeze-thaw cycles. Chlorides also generate expansive stresses through periodic cycles of dissolution and recrystallization in response to relative humidity cycles in the atmosphere. Similar problems also occur in reinforced concrete, where chlorides cause additional damage through corrosion of the reinforcing steel. The sensitivity of the neutron probe to hydrogen and chlorine thus means it can be used to map the distribution of these agents of deterioration. Preliminary field work at Colonial Williamsburg and Venice, Italy, showed that the technique could yield useful qualitative information. However, to be a quantitative method, the neutron probe had to be standardized in the laboratory on materials of known composition and specified moisture and chloride content.« less

Measurement of Fission Product Yields from Fast-Neutron Fission

NASA Astrophysics Data System (ADS)

Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Henderson, R.; Kenneally, J.; Macri, R.; McNabb, D.; Ryan, C.; Sheets, S.; Stoyer, M. A.; Tonchev, A. P.; Bhatia, C.; Bhike, M.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.

2014-09-01

One of the aims of the Stockpile Stewardship Program is a reduction of the uncertainties on fission data used for analyzing nuclear test data [1,2]. Fission products such as 147Nd are convenient for determining fission yields because of their relatively high yield per fission (about 2%) and long half-life (10.98 days). A scientific program for measuring fission product yields from 235U,238U and 239Pu targets as a function of bombarding neutron energy (0.1 to 15 MeV) is currently underway using monoenergetic neutron beams produced at the 10 MV Tandem Accelerator at TUNL. Dual-fission chambers are used to determine the rate of fission in targets during activation. Activated targets are counted in highly shielded HPGe detectors over a period of several weeks to identify decaying fission products. To date, data have been collected at neutron bombarding energies 4.6, 9.0, 14.5 and 14.8 MeV. Experimental methods and data reduction techniques are discussed, and some preliminary results are presented.

Measurements of the thermal neutron cross-section and resonance integral for the 108Pd(n,γ)109Pd reaction

NASA Astrophysics Data System (ADS)

Hien, Nguyen Thi; Kim, Guinyun; Kim, Kwangsoo; Do, Nguyen Van; Khue, Pham Duc; Thanh, Kim Tien; Shin, Sung-Gyun; Cho, Moo-Hyun

2018-06-01

The thermal neutron capture cross-section (σ0) and resonance integral (I0) of the 108Pd(n,γ)109Pd reaction have been measured relative to that of the monitor reaction 197Au(n,γ)198Au. The measurements were carried out using the neutron activation with the cadmium ratio method. Both the samples and monitors were irradiated with and without cadmium cover of 0.5 mm thickness. The induced activities of the reaction products were measured with a well calibrated HPGe γ-ray detector. In order to improve the accuracy of the results, the necessary corrections for the counting losses were made. The thermal neutron capture cross-section and resonance integral of the 108Pd(n,γ)109Pd reaction were determined to be σ0,Pd = 8.68 ± 0.41 barn and I0,Pd = 245.6 ± 24.8 barn, respectively. The obtained results are compared with literature values and discussed.

Studies on separation and purification of fission (99)Mo from neutron activated uranium aluminum alloy.

PubMed

Rao, Ankita; Kumar Sharma, Abhishek; Kumar, Pradeep; Charyulu, M M; Tomar, B S; Ramakumar, K L

2014-07-01

A new method has been developed for separation and purification of fission (99)Mo from neutron activated uranium-aluminum alloy. Alkali dissolution of the irradiated target (100mg) results in aluminum along with (99)Mo and a few fission products passing into solution, while most of the fission products, activation products and uranium remain undissolved. Subsequent purification steps involve precipitation of aluminum as Al(OH)3, iodine as AgI/AgIO3 and molybdenum as Mo-α-benzoin oxime. Ruthenium is separated by volatilization as RuO4 and final purification of (99)Mo was carried out using anion exchange method. The radiochemical yield of fission (99)Mo was found to be >80% and the purity of the product was in conformity with the international pharmacopoeia standards. Copyright © 2014 Elsevier Ltd. All rights reserved.

Neutronic reactor

DOEpatents

Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

1983-01-01

A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

Analysis of low levels of rare earths by radiochemical neutron activation analysis

USGS Publications Warehouse

Wandless, G.A.; Morgan, J.W.

1985-01-01

A procedure for the radiochemical neutron-activation analysis for the rare earth elements (REE) involves the separation of the REE as a group by rapid ion-exchange methods and determination of yields by reactivation or by energy dispersive X-ray fluorescence (EDXRF) spectrometry. The U. S. Geological Survey (USGS) standard rocks, BCR-1 and AGV-1, were analyzed to determine the precision and accuracy of the method. We found that the precision was ??5-10% on the basis of replicate analysis and that, in general the accuracy was within ??5% of accepted values for most REE. Data for USGS standard rocks BIR-1 (Icelandic basalt) and DNC-1 (North Carolina diabase) are also presented. ?? 1985 Akade??miai Kiado??.

Measuring neutron spectra in radiotherapy using the nested neutron spectrometer.

PubMed

Maglieri, Robert; Licea, Angel; Evans, Michael; Seuntjens, Jan; Kildea, John

2015-11-01

Out-of-field neutron doses resulting from photonuclear interactions in the head of a linear accelerator pose an iatrogenic risk to patients and an occupational risk to personnel during radiotherapy. To quantify neutron production, in-room measurements have traditionally been carried out using Bonner sphere systems (BSS) with activation foils and TLDs. In this work, a recently developed active detector, the nested neutron spectrometer (NNS), was tested in radiotherapy bunkers. The NNS is designed for easy handling and is more practical than the traditional BSS. Operated in current-mode, the problem of pulse pileup due to high dose-rates is overcome by measuring current, similar to an ionization chamber. In a bunker housing a Varian Clinac 21EX, the performance of the NNS was evaluated in terms of reproducibility, linearity, and dose-rate effects. Using a custom maximum-likelihood expectation-maximization algorithm, measured neutron spectra at various locations inside the bunker were then compared to Monte Carlo simulations of an identical setup. In terms of dose, neutron ambient dose equivalents were calculated from the measured spectra and compared to bubble detector neutron dose equivalent measurements. The NNS-measured spectra for neutrons at various locations in a treatment room were found to be consistent with expectations for both relative shape and absolute magnitude. Neutron fluence-rate decreased with distance from the source and the shape of the spectrum changed from a dominant fast neutron peak near the Linac head to a dominant thermal neutron peak in the moderating conditions of the maze. Monte Carlo data and NNS-measured spectra agreed within 30% at all locations except in the maze where the deviation was a maximum of 40%. Neutron ambient dose equivalents calculated from the authors' measured spectra were consistent (one standard deviation) with bubble detector measurements in the treatment room. The NNS may be used to reliably measure the neutron spectrum of a radiotherapy beam in less than 1 h, including setup and data unfolding. This work thus represents a new, fast, and practical method for neutron spectral measurements in radiotherapy.

Fusion-neutron-yield, activation measurements at the Z accelerator: design, analysis, and sensitivity.

PubMed

Hahn, K D; Cooper, G W; Ruiz, C L; Fehl, D L; Chandler, G A; Knapp, P F; Leeper, R J; Nelson, A J; Smelser, R M; Torres, J A

2014-04-01

We present a general methodology to determine the diagnostic sensitivity that is directly applicable to neutron-activation diagnostics fielded on a wide variety of neutron-producing experiments, which include inertial-confinement fusion (ICF), dense plasma focus, and ion beam-driven concepts. This approach includes a combination of several effects: (1) non-isotropic neutron emission; (2) the 1/r(2) decrease in neutron fluence in the activation material; (3) the spatially distributed neutron scattering, attenuation, and energy losses due to the fielding environment and activation material itself; and (4) temporally varying neutron emission. As an example, we describe the copper-activation diagnostic used to measure secondary deuterium-tritium fusion-neutron yields on ICF experiments conducted on the pulsed-power Z Accelerator at Sandia National Laboratories. Using this methodology along with results from absolute calibrations and Monte Carlo simulations, we find that for the diagnostic configuration on Z, the diagnostic sensitivity is 0.037% ± 17% counts/neutron per cm(2) and is ∼ 40% less sensitive than it would be in an ideal geometry due to neutron attenuation, scattering, and energy-loss effects.

Neutron Spectroscopy for pulsed beams with frame overlap using a double time-of-flight technique

NASA Astrophysics Data System (ADS)

Harrig, K. P.; Goldblum, B. L.; Brown, J. A.; Bleuel, D. L.; Bernstein, L. A.; Bevins, J.; Harasty, M.; Laplace, T. A.; Matthews, E. F.

2018-01-01

A new double time-of-flight (dTOF) neutron spectroscopy technique has been developed for pulsed broad spectrum sources with a duty cycle that results in frame overlap, where fast neutrons from a given pulse overtake slower neutrons from previous pulses. Using a tunable beam at the 88-Inch Cyclotron at Lawrence Berkeley National Laboratory, neutrons were produced via thick-target breakup of 16 MeV deuterons on a beryllium target in the cyclotron vault. The breakup spectral shape was deduced from a dTOF measurement using an array of EJ-309 organic liquid scintillators. Simulation of the neutron detection efficiency of the scintillator array was performed using both GEANT4 and MCNP6. The efficiency-corrected spectral shape was normalized using a foil activation technique to obtain the energy-dependent flux of the neutron beam at zero degrees with respect to the incoming deuteron beam. The dTOF neutron spectrum was compared to spectra obtained using HEPROW and GRAVEL pulse height spectrum unfolding techniques. While the unfolding and dTOF results exhibit some discrepancies in shape, the integrated flux values agree within two standard deviations. This method obviates neutron time-of-flight spectroscopy challenges posed by pulsed beams with frame overlap and opens new opportunities for pulsed white neutron source facilities.

Shielding analyses of an AB-BNCT facility using Monte Carlo simulations and simplified methods

NASA Astrophysics Data System (ADS)

Lai, Bo-Lun; Sheu, Rong-Jiun

2017-09-01

Accurate Monte Carlo simulations and simplified methods were used to investigate the shielding requirements of a hypothetical accelerator-based boron neutron capture therapy (AB-BNCT) facility that included an accelerator room and a patient treatment room. The epithermal neutron beam for BNCT purpose was generated by coupling a neutron production target with a specially designed beam shaping assembly (BSA), which was embedded in the partition wall between the two rooms. Neutrons were produced from a beryllium target bombarded by 1-mA 30-MeV protons. The MCNP6-generated surface sources around all the exterior surfaces of the BSA were established to facilitate repeated Monte Carlo shielding calculations. In addition, three simplified models based on a point-source line-of-sight approximation were developed and their predictions were compared with the reference Monte Carlo results. The comparison determined which model resulted in better dose estimation, forming the basis of future design activities for the first ABBNCT facility in Taiwan.

Fast neutron imaging device and method

DOEpatents

Popov, Vladimir; Degtiarenko, Pavel; Musatov, Igor V.

2014-02-11

A fast neutron imaging apparatus and method of constructing fast neutron radiography images, the apparatus including a neutron source and a detector that provides event-by-event acquisition of position and energy deposition, and optionally timing and pulse shape for each individual neutron event detected by the detector. The method for constructing fast neutron radiography images utilizes the apparatus of the invention.

Nanostructure Neutron Converter Layer Development

NASA Technical Reports Server (NTRS)

Park, Cheol (Inventor); Lowther, Sharon E. (Inventor); Kang, Jin Ho (Inventor); Thibeault, Sheila A. (Inventor); Sauti, Godfrey (Inventor); Bryant, Robert G. (Inventor)

2016-01-01

Methods for making a neutron converter layer are provided. The various embodiment methods enable the formation of a single layer neutron converter material. The single layer neutron converter material formed according to the various embodiments may have a high neutron absorption cross section, tailored resistivity providing a good electric field penetration with submicron particles, and a high secondary electron emission coefficient. In an embodiment method a neutron converter layer may be formed by sequential supercritical fluid metallization of a porous nanostructure aerogel or polyimide film. In another embodiment method a neutron converter layer may be formed by simultaneous supercritical fluid metallization of a porous nanostructure aerogel or polyimide film. In a further embodiment method a neutron converter layer may be formed by in-situ metalized aerogel nanostructure development.

SU-E-T-557: Measuring Neutron Activation of Cardiac Devices Irradiated During Proton Therapy Using Indium Foils

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

Avery, S; Christodouleas, J; Delaney, K

2014-06-01

Purpose: Measuring Neutron Activation of Cardiac devices Irradiated during Proton Therapy using Indium Foils Methods: The foils had dimensions of 25mm x 25mm x 1mm. After being activated, the foils were placed in a Canberra Industries well chamber utilizing a NaI(Tl) scintillation detector. The resulting gamma spectrum was acquired and analyzed using Genie 2000 spectroscopy software. One activation foil was placed over the upper, left chest of RANDO where a pacemaker would be. The rest of the foils were placed over the midline of the patient at different distances, providing a spatial distribution over the phantom. Using lasers and BBsmore » to align the patient, 200 MU square fields were delivered to various treatment sites: the brain, the pancreas, and the prostate. Each field was shot at least a day apart, giving more than enough time for activity of the foil to decay (t1=2 = 54.12 min). Results: The net counts (minus background) of the three aforementioned peaks were used for our measurements. These counts were adjusted to account for detector efficiency, relative photon yields from decay, and the natural abundance of 115-In. The average neutron flux for the closed multi-leaf collimator irradiation was measured to be 1.62 x 106 - 0.18 x 106 cm2 s-1. An order of magnitude estimate of the flux for neutrons up to 1 keV from Diffenderfer et al. gives 3 x 106 cm2 s-1 which does agree on the order of magnitude. Conclusion: Lower energy neutrons have higher interaction cross-sections and are more likely to damage pacemakers. The thermal/slow neutron component may be enough to estimate the overall risk. The true test of the applicability of activation foils is whether or not measurements are capable of predicting cardiac device malfunction. For that, additional studies are needed to provide clinical evidence one way or the other.« less

Neutron capture cross-section studies of Tellurium isotopes for neutrinoless double beta decay applications

NASA Astrophysics Data System (ADS)

Bhike, Megha; Tornow, Werner

2014-09-01

The CUORE detector at Gran Sasso, aimed at searching for neutrinoless double-beta decay of 130Te, employs an array of TeO2 bolometer modules. To understand and identify the contribution of muon and (α,n) induced neutrons to the CUORE background, fast neutron cature cross-section data of the tellurium isotopes 126Te, 128Te and 130Te have been measured with the activation method at eight different energies in the neutron energy range 0.5-7.5 MeV. Plastic pill boxes of diameter 1.6 cm and width 1 cm containing Te were irradiated with mono-energetic neutrons produced via the 3H(p,n)3He and 2H(d,n)3He reactions. The cross-sections were determined relative to the 197Au(n, γ)198Au and 115In(n,n')115m In standard cross sections. The activities of the products were measured using 60% lead-shielded HPGe detectors at TUNL's low background counting facility. The present results are compared with the evaluated data from TENDL-2012, ENDF/B-VII.1, JEFF-3.2 and JENDL-4.0, as well as with literature data.

DN/DG Screening of Environmental Swipe Samples: FY2016 Report

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

Glasgow, David C.; Croft, Stephen; Venkataraman, Ramkumar

The Delayed Neutron Delayed Gamma (DNDG) technique provides a new analytical capability to the International Atomic Energy Agency (IAEA) for detecting undeclared nuclear activities. IAEA’s Long Term R&D (LTRD) plan has a stated high urgency need to develop elemental and isotopic signatures of nuclear fuel cycle activities and processes (LTRD 2.2). The new DNDG capability is used to co-detect both uranium and plutonium as an extension of a DN only method that is already being utilized by the IAEA for the analysis of swipes to inform on undeclared nuclear activities. Analytical method involving irradiation of swipe samples potentially containing tracemore » quantities of fissile material in a thermal neutron field, followed by the counting of delayed neutrons, is a well-known technique in the field of safeguards and nonproliferation. It is used for detecting the presence of microscopic amounts of fissile material, (typically a linear combination of 233U, 235U, 239Pu, and 241Pu)and quantifying it in terms of the equivalent mass of 235U. The delayed neutron (DN) technique is very sensitive and is been routinely employed at the High Flux Isotope Reactor (HFIR) facility at Oak Ridge National Laboratory (ORNL). Both uranium and plutonium are of high safeguards value. However, the DN technique is not well suited for distinguishing between U and Pu isotopes since the decay curves overlap closely. The delayed gamma (DG) technique will help detect the presence of 239Pu in a mixture of U and Pu. Thus the DNDG approach combines the best of both worlds; the sensitivity of DN counting and the isotopic specificity of DG counting. The present work seeks to build on the delayed neutron and delayed gamma methods that have been developed at ORNL. It is recognized that the distribution profile of heavy fission products remains fairly invariant for the fissile nuclides whereas the distribution of light fission products varies from one isotope to another. That is, the ratio of the yield of a light fission fragment to a heavy fission fragments is isotope specific. Measurement of the ratio of the net full energy peak (FEP) from low/high mass fission products is an elegant way to characterize the fraction of fissile materials present in a mixture. By empirically calibrating the ratio of the net FEP as a function of known concentration of the binary mixture, one can determine the fraction of fissile isotopes in an unknown sample. In the work done in fiscal year (FY) 2016, samples of single fissile material isotopes as well as binary mixtures were irradiated in a well thermalized irradiation field in the HFIR. Delayed neutron counting was performed using the neutron counter at the HFIR Neutron Activation Analysis (NAA) laboratory. Delayed gamma counting was performed using a shielded high purity germanium (HPGe) detector. Delayed neutron decay curve results highlighted the difficulty of distinguishing between U and Pu isotopes, and the need for including the delayed gamma component. Based on delayed gamma spectrometry, twelve ratios of low mass/high fission product gamma ray FEP have been identified as valid candidates. Linearity of the ratios, as a function of 239Pu fraction in 235U+ 239Pu mixtures, was confirmed for the low mass/high mass candidates that were selected. The DNDG method we are spearheading allows not only the presence of total fissile content to be detected, but whether the material is predominantly U or predominantly Pu, or a mixture. This provides additional SG relevant information.« less

Active interrogation of highly enriched uranium

NASA Astrophysics Data System (ADS)

Fairrow, Nannette Lea

Safeguarding special nuclear material (SNM) in the Department of Energy Complex is vital to the national security of the United States. Active and passive nondestructive assays are used to confirm the presence of SNM in various configurations ranging from waste to nuclear weapons. Confirmation measurements for nuclear weapons are more challenging because the design complicates the detection of a distinct signal for highly enriched uranium. The emphasis of this dissertation was to investigate a new nondestructive assay technique that provides an independent and distinct signal to confirm the presence of highly enriched uranium (HEU). Once completed and tested this assay method could be applied to confirmation measurements of nuclear weapons. The new system uses a 14-MeV neutron source for interrogation and records the arrival time of neutrons between the pulses with a high efficiency detection system. The data is then analyzed by the Feynman reduced variance method. The analysis determined the amount of correlation in the data and provided a unique signature of correlated fission neutrons. Measurements of HEU spheres were conducted at Los Alamos with the new system. Then, Monte Carlo calculations were performed to verify hypothesis made about the behavior of the neutrons in the experiment. Comparisons of calculated counting rates by the Monte Carlo N-Particle Transport Code (MCNP) were made with the experimental data to confirm that the measured response reflected the desired behavior of neutron interactions in the highly enriched uranium. In addition, MCNP calculations of the delayed neutron build-up were compared with the measured data. Based on the results obtained from this dissertation, this measurement method has the potential to be expanded to include mass determinations of highly enriched uranium. Although many safeguards techniques exist for measuring special nuclear material, the number of assays that can be used to confirm HEU in shielded systems is limited. These assays also rely on secondary characteristics of the material to be measured. A review of the nondestructive techniques with potential applications for nuclear weapons confirmatory measurements were evaluated with summaries of the pros and cons involved in implementing the methods at production type facilities.

The investigation of fast neutron Threshold Activation Detectors (TAD)

NASA Astrophysics Data System (ADS)

Gozani, T.; King, M. J.; Stevenson, J.

2012-02-01

The detection of fast neutrons is usually done by liquid hydrogenous organic scintillators, where the separation between the ever present gamma rays and neutrons is achieved by the pulse shape discrimination (PSD). In many practical situation the detection of fast neutrons has to be carried out while the intense source (be it neutrons, gamma rays or x-rays) that creates these neutrons, for example by the fission process, is present. This source, or ``flash'', usually blinds the neutron detectors and temporarily incapacitates them. By the time the detectors recover the prompt neutron signature does not exist. Thus to overcome the blinding background, one needs to search for processes whereby the desired signature, such as fission neutrons could in some way be measured long after the fission occurred and when the neutron detector is fully recovered from the overload. A new approach was proposed and demonstrated a good sensitivity for the detection of fast neutrons in adverse overload situations where normally it could not be done. A temporal separation of the fission event from the prompt neutrons detection is achieved via the activation process. The main idea, called Threshold Activation Detection (or detector)-TAD, is to find appropriate substances that can be selectively activated by the fission neutrons and not by the source radiation, and then measure the radioactively decaying activation products (typically beta and γ-rays) well after the source pulse has ended. The activation material should possess certain properties: a suitable half-life; an energy threshold below which the numerous source neutrons will not activate it (e.g. about 3 MeV); easily detectable activation products and has a usable cross section for the selected reaction. Ideally the substance would be part of the scintillator. There are several good candidates for TAD. The first one we have selected is based on fluorine. One of the major advantages of this element is the fact that it is a major constituent of available scintillators (e.g., BaF2, CaF2, hydrogen free liquid fluorocarbon). Thus the activation products of the fast prompt neutrons, in particular, the beta particles, can be measured with a very high efficiency in the detector. Other detectors and substances were investigated, such as 6Li and even common detectors such as NaI. The principles and experimental results obtained with F, NaI and 6Li based TAD are shown. The various contributing activation products are identified. The insensitivity of the fluorine based TAD to (d,D) neutrons is demonstrated. Ways and means to reduce or subtract the various neutron induced activations of NaI detector are elucidated along with its fast neutron detection capabilities. 6Li could also be a useful TAD.

Feasibility study of prompt gamma neutron activation for NDT measurement of moisture in stone and brick

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

Livingston, R. A.; Al-Sheikhly, M.; Grissom, C.

2014-02-18

The conservation of stone and brick architecture or sculpture often involves damage caused by moisture. The feasibility of a NDT method based on prompt gamma neutron activation (PGNA) for measuring the element hydrogen as an indication of water is being evaluated. This includes systematic characterization of the lithology and physical properties of seven building stones and one brick type used in the buildings of the Smithsonian Institution in Washington, D.C. To determine the required dynamic range of the NDT method, moisture-related properties were measured by standard methods. Cold neutron PGNA was also used to determine chemically bound water (CBW) content.more » The CBW does not damage porous masonry, but creates an H background that defines the minimum level of detection of damaging moisture. The CBW was on the order of 0.5% for all the stones. This rules out the measurement of hygric processes in all of the stones and hydric processed for the stones with fine scale pore-size distributions The upper bound of moisture content, set by porosity through water immersion, was on the order of 5%. The dynamic range is about 10–20. The H count rates were roughly 1–3 cps. Taking into account differences in neutron energies and fluxes and sample volume between cold PGNA and a portable PGNA instrument, it appears that it is feasible to apply PGNA in the field.« less

Flux trap effect study in a sub-critical neutron assembly using activation methods

NASA Astrophysics Data System (ADS)

Routsonis, K.; Stoulos, S.; Clouvas, A.; Catsaros, N.; Varvayianni, M.; Manolopoulou, M.

2016-09-01

The neutron flux trap effect was experimentally studied in the subcritical assembly of the Atomic and Nuclear Physics Laboratory of the Aristotle University of Thessaloniki, using delayed gamma neutron activation analysis. Measurements were taken within the natural uranium fuel grid, in vertical levels symmetrical to the Am-Be neutron source, before and after the removal of fuel elements, permitting likewise a basic study of the vertical flux profile. Three identical flux traps of diamond shape were created by removing four fuel rods for each one. Two (n, γ) reactions and one (n, p) threshold reaction were selected for thermal, epithermal and fast flux study. Results of thermal and epithermal flux obtained through the 197Au (n, γ) 198Au and 186W (n, γ) 187W reactions, with and without Cd covers, to differentiate between the two flux regions. The 58Ni (n, p) 58Co reaction was used for the fast flux determination. An interpolation technique based on local procedures was applied to fit the cross sections data and the neutron flux spectrum. End results show a maximum thermal flux increase of 105% at the source level, pointing to a high potential to increase in the available thermal flux for future experiments. The increase in thermal flux is not accompanied by a comparable decrease in epithermal or fast flux, since thermal flux gain is higher than epithermal and fast neutron flux loss. So, the neutron reflection is mainly responsible for the thermal neutron increase, contributing to 89% at the central axial position.

Neutron detector and fabrication method thereof

DOEpatents

Bhandari, Harish B.; Nagarkar, Vivek V.; Ovechkina, Olena E.

2016-08-16

A neutron detector and a method for fabricating a neutron detector. The neutron detector includes a photodetector, and a solid-state scintillator operatively coupled to the photodetector. In one aspect, the method for fabricating a neutron detector includes providing a photodetector, and depositing a solid-state scintillator on the photodetector to form a detector structure.

Quantitative Evaluation of Ion-implanted Arsenic in Silicon by Instrumental Neutron Activation Analysis

NASA Astrophysics Data System (ADS)

Takatsuka, Toshiko; Hirata, Kouichi; Kobayashi, Yoshinori; Kuroiwa, Takayoshi; Miura, Tsutomu; Matsue, Hideaki

2008-11-01

Certified reference materials (CRMs) of shallow arsenic implants in silicon are now under development at the National Metrology Institute of Japan (NMIJ). The amount of ion-implanted arsenic atoms is quantified by Instrumental Neutron Activation Analysis (INAA) using research reactor JRR-3 in Japan Atomic Energy Agency (JAEA). It is found that this method can evaluate arsenic amounts of 1015 atoms/cm2 with small uncertainties, and is adaptable to shallower dopants. The estimated uncertainties can satisfy the industrial demands for reference materials to calibrate the implanted dose of arsenic at shallow junctions.

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

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

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

An extension of the point kinetics model is developed in this paper to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If themore » detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. Finally, the spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.« less

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

DOE PAGES

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

2017-06-13

An extension of the point kinetics model is developed in this paper to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If themore » detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. Finally, the spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.« less

Neutron dose estimation in a zero power nuclear reactor

NASA Astrophysics Data System (ADS)

Triviño, S.; Vedelago, J.; Cantargi, F.; Keil, W.; Figueroa, R.; Mattea, F.; Chautemps, A.; Santibañez, M.; Valente, M.

2016-10-01

This work presents the characterization and contribution of neutron and gamma components to the absorbed dose in a zero power nuclear reactor. A dosimetric method based on Fricke gel was implemented to evaluate the separation between dose components in the mixed field. The validation of this proposed method was performed by means of direct measurements of neutron flux in different positions using Au and Mg-Ni activation foils. Monte Carlo simulations were conversely performed using the MCNP main code with a dedicated subroutine to incorporate the exact complete geometry of the nuclear reactor facility. Once nuclear fuel elements were defined, the simulations computed the different contributions to the absorbed dose in specific positions inside the core. Thermal/epithermal contributions of absorbed dose were assessed by means of Fricke gel dosimetry using different isotopic compositions aimed at modifying the sensitivity of the dosimeter for specific dose components. Clear distinctions between gamma and neutron capture dose were obtained. Both Monte Carlo simulations and experimental results provided reliable estimations about neutron flux rate as well as dose rate during the reactor operation. Simulations and experimental results are in good agreement in every positions measured and simulated in the core.

Neutron activation analysis: A primary method of measurement

NASA Astrophysics Data System (ADS)

Greenberg, Robert R.; Bode, Peter; De Nadai Fernandes, Elisabete A.

2011-03-01

Neutron activation analysis (NAA), based on the comparator method, has the potential to fulfill the requirements of a primary ratio method as defined in 1998 by the Comité Consultatif pour la Quantité de Matière — Métrologie en Chimie (CCQM, Consultative Committee on Amount of Substance — Metrology in Chemistry). This thesis is evidenced in this paper in three chapters by: demonstration that the method is fully physically and chemically understood; that a measurement equation can be written down in which the values of all parameters have dimensions in SI units and thus having the potential for metrological traceability to these units; that all contributions to uncertainty of measurement can be quantitatively evaluated, underpinning the metrological traceability; and that the performance of NAA in CCQM key-comparisons of trace elements in complex matrices between 2000 and 2007 is similar to the performance of Isotope Dilution Mass Spectrometry (IDMS), which had been formerly designated by the CCQM as a primary ratio method.

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.

Coupled multi-group neutron photon transport for the simulation of high-resolution gamma-ray spectroscopy applications

NASA Astrophysics Data System (ADS)

Burns, Kimberly Ann

The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of coupled Monte Carlo-deterministic methods for the simulation of neutron-induced photons for high-resolution gamma-ray spectroscopy applications. RAdiation Detection Scenario Analysis Toolbox (RADSAT), a code which couples deterministic and Monte Carlo transport to perform radiation detection scenario analysis in three dimensions [1], was used as the building block for the methods derived in this work. RADSAT was capable of performing coupled deterministic-Monte Carlo simulations for gamma-only and neutron-only problems. The purpose of this work was to develop the methodology necessary to perform coupled neutron-photon calculations and add this capability to RADSAT. Performing coupled neutron-photon calculations requires four main steps: the deterministic neutron transport calculation, the neutron-induced photon spectrum calculation, the deterministic photon transport calculation, and the Monte Carlo detector response calculation. The necessary requirements for each of these steps were determined. A major challenge in utilizing multigroup deterministic transport methods for neutron-photon problems was maintaining the discrete neutron-induced photon signatures throughout the simulation. Existing coupled neutron-photon cross-section libraries and the methods used to produce neutron-induced photons were unsuitable for high-resolution gamma-ray spectroscopy applications. Central to this work was the development of a method for generating multigroup neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so the neutron-induced photon signatures were preserved. The RADSAT-NG cross-section library was developed as a specialized multigroup neutron-photon cross-section set for the simulation of high-resolution gamma-ray spectroscopy applications. The methodology and cross sections were tested using code-to-code comparison with MCNP5 [2] and NJOY [3]. A simple benchmark geometry was used for all cases compared with MCNP. The geometry consists of a cubical sample with a 252Cf neutron source on one side and a HPGe gamma-ray spectrometer on the opposing side. Different materials were examined in the cubical sample: polyethylene (C2H4), P, N, O, and Fe. The cross sections for each of the materials were compared to cross sections collapsed using NJOY. Comparisons of the volume-averaged neutron flux within the sample, volume-averaged photon flux within the detector, and high-purity gamma-ray spectrometer response (only for polyethylene) were completed using RADSAT and MCNP. The code-to-code comparisons show promising results for the coupled Monte Carlo-deterministic method. The RADSAT-NG cross-section production method showed good agreement with NJOY for all materials considered although some additional work is needed in the resonance region and in the first and last energy bin. Some cross section discrepancies existed in the lowest and highest energy bin, but the overall shape and magnitude of the two methods agreed. For the volume-averaged photon flux within the detector, typically the five most intense lines agree to within approximately 5% of the MCNP calculated flux for all of materials considered. The agreement in the code-to-code comparisons cases demonstrates a proof-of-concept of the method for use in RADSAT for coupled neutron-photon problems in high-resolution gamma-ray spectroscopy applications. One of the primary motivators for using the coupled method over pure Monte Carlo method is the potential for significantly lower computational times. For the code-to-code comparison cases, the run times for RADSAT were approximately 25--500 times shorter than for MCNP, as shown in Table 1. This was assuming a 40 mCi 252Cf neutron source and 600 seconds of "real-world" measurement time. The only variance reduction technique implemented in the MCNP calculation was forward biasing of the source toward the sample target. Improved MCNP runtimes could be achieved with the addition of more advanced variance reduction techniques.

Neutron spectroscopy for pulsed beams with frame overlap using a double time-of-flight technique

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

Harrig, K. P.; Goldblum, B. L.; Brown, J. A.

A new double time-of- ight (dTOF) neutron spectroscopy technique has been developed for pulsed broad spectrum sources with a duty cycle that results in frame overlap, where fast neutrons from a given pulse overtake slower neutrons from previous pulses. Using a tunable beam at the 88-Inch Cyclotron at Lawrence Berkeley National Laboratory, neutrons were produced via thick-target breakup of 16 MeV deuterons on a beryllium target in the cyclotron vault. The breakup spectral shape was deduced from a dTOF measurement using an array of EJ-309 organic liquid scintillators. Simulation of the neutron detection efficiency of the scintillator array was performedmore » using both GEANT4 and MCNP6. The efficiency- corrected spectral shape was normalized using a foil activation technique to obtain the energy-dependent flux of the neutron beam at zero degrees with respect to the incoming deuteron beam. The dTOF neutron spectrum was compared to spectra obtained using HEPROW and GRAVEL pulse height spectrum unfolding techniques. While the unfolding and dTOF results exhibit some discrepancies in shape, the integrated flux values agree within two standard deviations. As a result, this method obviates neutron time-of-flight spectroscopy challenges posed by pulsed beams« less

Neutron spectroscopy for pulsed beams with frame overlap using a double time-of-flight technique

DOE PAGES

Harrig, K. P.; Goldblum, B. L.; Brown, J. A.; ...

2017-10-16

A new double time-of- ight (dTOF) neutron spectroscopy technique has been developed for pulsed broad spectrum sources with a duty cycle that results in frame overlap, where fast neutrons from a given pulse overtake slower neutrons from previous pulses. Using a tunable beam at the 88-Inch Cyclotron at Lawrence Berkeley National Laboratory, neutrons were produced via thick-target breakup of 16 MeV deuterons on a beryllium target in the cyclotron vault. The breakup spectral shape was deduced from a dTOF measurement using an array of EJ-309 organic liquid scintillators. Simulation of the neutron detection efficiency of the scintillator array was performedmore » using both GEANT4 and MCNP6. The efficiency- corrected spectral shape was normalized using a foil activation technique to obtain the energy-dependent flux of the neutron beam at zero degrees with respect to the incoming deuteron beam. The dTOF neutron spectrum was compared to spectra obtained using HEPROW and GRAVEL pulse height spectrum unfolding techniques. While the unfolding and dTOF results exhibit some discrepancies in shape, the integrated flux values agree within two standard deviations. As a result, this method obviates neutron time-of-flight spectroscopy challenges posed by pulsed beams« less

Neutron radiative capture reactions on nuclei of relevance to 0νββ, dark matter and neutrino/antineutrino searches

NASA Astrophysics Data System (ADS)

Tornow, W.; Bhike, Megha

2015-05-01

A program is underway at the Triangle Universities Nuclear Laboratory (TUNL) to measure the neutron capture cross section in the 0.5 to 15 MeV energy range on nuclei whose radioactive daughters could potentially create backgrounds in searches for rare events. Here, we refer to neutrino-less double-beta decay and dark-matter searches, and to detectors built for neutrino and/or antineutrino studies. Neutron capture cross-section data obtained by using the activation method are reported for 40Ar, 74,76Ge, 128,130Te and 136Xe and compared to model calculations and evaluations.

Solvent extraction of oil-sand components for determination of trace elements by neutron activation analysis

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

Jacobs, F.S.; Filby, R.H.

Instrumental neutron activation analysis was used to measure the concentrations of 30 elements in Athabasca oil sands and oil-sand components. The oil sands were separated into solid residue, bitumen, and fines by Soxhlet extraction with toluene-bitumen extract. The mineral content of the extracted bitumen was dependent on the treatment of the oil sand prior to extraction. The geochemically important and organically associated trace element contents of the bitumen (and asphaltenes) were determined by subtracting the mineral contributions from the total measured concentrations. The method allows analysis of the bitumen without the necessity of ultracentrifugation or membrane filtration, which might removemore » geochemically important components of the bitumen. The method permits classification of trace elements into organic and inorganic combinations.« less

A method to measure neutron polarization using P-even asymmetry of {gamma}-quantum emission in the neutron-nuclear interaction

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

Gledenov, Yu. M.; Nesvizhevsky, V. V.; Sedyshev, P. V.

2012-07-15

A new method to measure polarization of cold/thermal neutrons using P-even asymmetry in nuclear reactions induced by polarized neutrons is proposed. A scheme profiting from a large correlation of the neutron spin and the circular {gamma}-quantum polarization in the reaction (n, {gamma}) of polarized neutrons with nuclei is analyzed. This method could be used, for instance, to measure the neutron-beam polarization in experiments with frequently varying configuration. We show that high accuracy and reliability of measurements could be expected.

A neutron activation analysis procedure for the determination of uranium, thorium and potassium in geologic samples

USGS Publications Warehouse

Aruscavage, P. J.; Millard, H.T.

1972-01-01

A neutron activation analysis procedure was developed for the determination of uranium, thorium and potassium in basic and ultrabasic rocks. The three elements are determined in the same 0.5-g sample following a 30-min irradiation in a thermal neutron flux of 2??1012 n??cm-2??sec-1. Following radiochemical separation, the nuclides239U (T=23.5 m),233Th (T=22.2 m) and42K (T=12.36 h) are measured by ??-counting. A computer program is used to resolve the decay curves which are complex owing to contamination and the growth of daughter activities. The method was used to determine uranium, throium and potassium in the U. S. Geological Survey standard rocks DTS-1, PCC-1 and BCR-1. For 0.5-g samples the limits of detection for uranium, throium and potassium are 0.7, 1.0 and 10 ppb, respectively. ?? 1972 Akade??miai Kiado??.

Improvement of analytical capabilities of neutron activation analysis laboratory at the Colombian Geological Survey

NASA Astrophysics Data System (ADS)

Parrado, G.; Cañón, Y.; Peña, M.; Sierra, O.; Porras, A.; Alonso, D.; Herrera, D. C.; Orozco, J.

2016-07-01

The Neutron Activation Analysis (NAA) laboratory at the Colombian Geological Survey has developed a technique for multi-elemental analysis of soil and plant matrices, based on Instrumental Neutron Activation Analysis (INAA) using the comparator method. In order to evaluate the analytical capabilities of the technique, the laboratory has been participating in inter-comparison tests organized by Wepal (Wageningen Evaluating Programs for Analytical Laboratories). In this work, the experimental procedure and results for the multi-elemental analysis of four soil and four plant samples during participation in the first round on 2015 of Wepal proficiency test are presented. Only elements with radioactive isotopes with medium and long half-lives have been evaluated, 15 elements for soils (As, Ce, Co, Cr, Cs, Fe, K, La, Na, Rb, Sb, Sc, Th, U and Zn) and 7 elements for plants (Br, Co, Cr, Fe, K, Na and Zn). The performance assessment by Wepal based on Z-score distributions showed that most results obtained |Z-scores| ≤ 3.

Development of high flux thermal neutron generator for neutron activation analysis

NASA Astrophysics Data System (ADS)

Vainionpaa, Jaakko H.; Chen, Allan X.; Piestrup, Melvin A.; Gary, Charles K.; Jones, Glenn; Pantell, Richard H.

2015-05-01

The new model DD110MB neutron generator from Adelphi Technology produces thermal (<0.5 eV) neutron flux that is normally achieved in a nuclear reactor or larger accelerator based systems. Thermal neutron fluxes of 3-5 · 107 n/cm2/s are measured. This flux is achieved using four ion beams arranged concentrically around a target chamber containing a compact moderator with a central sample cylinder. Fast neutron yield of ∼2 · 1010 n/s is created at the titanium surface of the target chamber. The thickness and material of the moderator is selected to maximize the thermal neutron flux at the center. The 2.5 MeV neutrons are quickly thermalized to energies below 0.5 eV and concentrated at the sample cylinder. The maximum flux of thermal neutrons at the target is achieved when approximately half of the neutrons at the sample area are thermalized. In this paper we present simulation results used to characterize performance of the neutron generator. The neutron flux can be used for neutron activation analysis (NAA) prompt gamma neutron activation analysis (PGNAA) for determining the concentrations of elements in many materials. Another envisioned use of the generator is production of radioactive isotopes. DD110MB is small enough for modest-sized laboratories and universities. Compared to nuclear reactors the DD110MB produces comparable thermal flux but provides reduced administrative and safety requirements and it can be run in pulsed mode, which is beneficial in many neutron activation techniques.

The "neutron channel design"—A method for gaining the desired neutrons

NASA Astrophysics Data System (ADS)

Hu, G.; Hu, H. S.; Wang, S.; Pan, Z. H.; Jia, Q. G.; Yan, M. F.

2016-12-01

The neutrons with desired parameters can be obtained after initial neutrons penetrating various structure and component of the material. A novel method, the "neutron channel design", is proposed in this investigation for gaining the desired neutrons. It is established by employing genetic algorithm (GA) combining with Monte Carlo software. This method is verified by obtaining 0.01eV to 1.0eV neutrons from the Compact Accelerator-driven Neutron Source (CANS). One layer polyethylene (PE) moderator was designed and installed behind the beryllium target in CANS. The simulations and the experiment for detection the neutrons were carried out. The neutron spectrum at 500cm from the PE moderator was simulated by MCNP and PHITS software. The counts of 0.01eV to 1.0eV neutrons were simulated by MCNP and detected by the thermal neutron detector in the experiment. These data were compared and analyzed. Then this method is researched on designing the complex structure of PE and the composite material consisting of PE, lead and zirconium dioxide.

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

Analytical study of comet nucleus samples

NASA Technical Reports Server (NTRS)

Albee, A. L.

1989-01-01

Analytical procedures for studying and handling frozen (130 K) core samples of comet nuclei are discussed. These methods include neutron activation analysis, x ray fluorescent analysis and high resolution mass spectroscopy.

Associated Particle Tagging (APT) in Magnetic Spectrometers

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

Jordan, David V.; Baciak, James E.; Stave, Sean C.

2012-10-16

Summary In Brief The Associated Particle Tagging (APT) project, a collaboration of Pacific Northwest National Laboratory (PNNL), Idaho National Laboratory (INL) and the Idaho State University (ISU)/Idaho Accelerator Center (IAC), has completed an exploratory study to assess the role of magnetic spectrometers as the linchpin technology in next-generation tagged-neutron and tagged-photon active interrogation (AI). The computational study considered two principle concepts: (1) the application of a solenoidal alpha-particle spectrometer to a next-generation, large-emittance neutron generator for use in the associated particle imaging technique, and (2) the application of tagged photon beams to the detection of fissile material via active interrogation.more » In both cases, a magnetic spectrometer momentum-analyzes charged particles (in the neutron case, alpha particles accompanying neutron generation in the D-T reaction; in the tagged photon case, post-bremsstrahlung electrons) to define kinematic properties of the relevant neutral interrogation probe particle (i.e. neutron or photon). The main conclusions of the study can be briefly summarized as follows: Neutron generator: • For the solenoidal spectrometer concept, magnetic field strengths of order 1 Tesla or greater are required to keep the transverse size of the spectrometer smaller than 1 meter. The notional magnetic spectrometer design evaluated in this feasibility study uses a 5-T magnetic field and a borehole radius of 18 cm. • The design shows a potential for 4.5 Sr tagged neutron solid angle, a factor of 4.5 larger than achievable with current API neutron-generator designs. • The potential angular resolution for such a tagged neutron beam can be less than 0.5o for modest Si-detector position resolution (3 mm). Further improvement in angular resolution can be made by using Si-detectors with better position resolution. • The report documents several features of a notional generator design incorporating the alpha-particle spectrometer concept, and outlines challenges involved in the magnetic field design. Tagged photon interrogation: • We investigated a method for discriminating fissile from benign cargo-material response to an energy-tagged photon beam. The method relies upon coincident detection of the tagged photon and a photoneutron or photofission neutron produced in the target material. The method exploits differences in the shape of the neutron production cross section as a function of incident photon energy in order to discriminate photofission yield from photoneutrons emitted by non-fissile materials. Computational tests of the interrogation method as applied to material composition assay of a simple, multi-layer target suggest that the tagged-photon information facilitates precise (order 1% thickness uncertainty) reconstruction of the constituent thicknesses of fissile (uranium) and high-Z (Pb) constituents of the test targets in a few minutes of photon-beam exposure. We assumed an 18-MeV endpoint tagged photon beam for these simulations. • The report addresses several candidate design and data analysis issues for beamline infrastructure required to produce a tagged photon beam in a notional AI-dedicated facility, including the accelerator and tagging spectrometer.« less

Tracing footprints of environmental events in tree ring chemistry using neutron activation analysis

NASA Astrophysics Data System (ADS)

Sahin, Dagistan

The aim of this study is to identify environmental effects on tree-ring chemistry. It is known that industrial pollution, volcanic eruptions, dust storms, acid rain and similar events can cause substantial changes in soil chemistry. Establishing whether a particular group of trees is sensitive to these changes in soil environment and registers them in the elemental chemistry of contemporary growth rings is the over-riding goal of any Dendrochemistry research. In this study, elemental concentrations were measured in tree-ring samples of absolutely dated eleven modern forest trees, grown in the Mediterranean region, Turkey, collected and dated by the Malcolm and Carolyn Wiener Laboratory for Aegean and Near Eastern Dendrochronology laboratory at Cornell University. Correlations between measured elemental concentrations in the tree-ring samples were analyzed using statistical tests to answer two questions. Does the current concentration of a particular element depend on any other element within the tree? And, are there any elements showing correlated abnormal concentration changes across the majority of the trees? Based on the detailed analysis results, the low mobility of sodium and bromine, positive correlations between calcium, zinc and manganese, positive correlations between trace elements lanthanum, samarium, antimony, and gold within tree-rings were recognized. Moreover, zinc, lanthanum, samarium and bromine showed strong, positive correlations among the trees and were identified as possible environmental signature elements. New Dendrochemistry information found in this study would be also useful in explaining tree physiology and elemental chemistry in Pinus nigra species grown in Turkey. Elemental concentrations in tree-ring samples were measured using Neutron Activation Analysis (NAA) at the Pennsylvania State University Radiation Science and Engineering Center (RSEC). Through this study, advanced methodologies for methodological, computational and experimental NAA were developed to ensure an acceptable accuracy and certainty in the elemental concentration measurements in tree-ring samples. Two independent analysis methods of NAA were used; the well known k-zero method and a novel method developed in this study, called the Multi-isotope Iterative Westcott (MIW) method. The MIW method uses reaction rate probabilities for a group of isotopes, which can be calculated by a neutronic simulation or measured by experimentation, and determines the representative values for the neutron flux and neutron flux characterization parameters based on Westcott convention. Elemental concentration calculations for standard reference material and tree-ring samples were then performed using the MIW and k-zero analysis methods of the NAA and the results were cross verified. In the computational part of this study, a detailed burnup coupled neutronic simulation was developed to analyze real-time neutronic changes in a TRIGA Mark III reactor core, in this study, the Penn State Breazeale Reactor (PSBR) core. To the best of the author`s knowledge, this is the first burnup coupled neutronic simulation with realistic time steps and full fuel temperature profile for a TRIGA reactor using Monte Carlo Utility for Reactor Evolutions (MURE) code and Monte Carlo Neutral-Particle Code (MCNP) coupling. High fidelity and flexibility in the simulation was aimed to replicate the real core operation through the day. This approach resulted in an enhanced accuracy in neutronic representation of the PSBR core with respect to previous neutronic simulation models for the PSBR core. An important contribution was made in the NAA experimentation practices employed in Dendrochemistry studies at the RSEC. Automated laboratory control and analysis software for NAA measurements in the RSEC Radionuclide Applications Laboratory was developed. Detailed laboratory procedures were written in this study comprising preparation, handling and measurements of tree-ring samples in the Radionuclide Applications Laboratory.

Prompt-gamma neutron activation analysis system design: Effects of D-T versus D-D neutron generator source selection

USDA-ARS?s Scientific Manuscript database

Prompt-gamma neutron activation (PGNA) analysis is used for the non-invasive measurement of human body composition. Advancements in portable, compact neutron generator design have made those devices attractive as neutron sources. Two distinct generators are available: D-D with 2.5 MeV and D-T with...

Neutron-capture cross-section measurements of 74Ge and 76Ge in the energy region 0.4-14.8 MeV for neutrinoless double β decay applications

NASA Astrophysics Data System (ADS)

Bhike, Megha; Tornow, Werner

2013-10-01

Fast neutron capture cross sections for the reactions 74Ge(n, γ)75Ge and 76Ge(n, γ)77Ge have been measured in the neutron energy region 0.4-14.8 MeV with the activation method. The results are important to identify backgrounds in the neutrinoless double- β decay experiments GERDA and MAJORANA, which use germanium as both source and detector. Isotopically enriched targets which consisted of 86% of 76Ge and 14% of 74Ge were irradiated with mono-energetic neutrons produced via 3H(p,n)3He, 2H(d,n)3He and 3H(d,n)4He reactions. The cross sections were determined relative to 197Au(n, γ)198Au, 115In(n,n')115mIn and 197Au(n,2n)196Au standard cross sections. The activities of the products were measured using high-resolution γ-ray spctroscopy. The present results are compared with the evaluated data from ENDF/B-VII.1 and TALYS.

A probe for neutron activation analysis in a drill hole using 252Cf, and a Ge(Li) detector cooled by a melting cryogen

USGS Publications Warehouse

Tanner, A.B.; Moxham, R.M.; Senftle, F.E.; Baicker, J.A.

1972-01-01

A sonde has been built for high-resolution measurement of natural or neutron-induced gamma rays in boreholes. The sonde is 7.3 cm in diameter and about 2.2 m in length and weighs about 16 kg. The lithium-compensated germanium semiconductor detector is stabilized at -185 to -188??C for as much as ten hours by a cryostatic reservoir containing melting propane. During periods when the sonde is not in use the propane is kept frozen by a gravity-fed trickle of liquid nitrogen from a reservoir temporarily attached to the cryostat section. A 252Cf source, shielded from the detector, may be placed in the bottom section of the sonde for anlysis by measurement of neutron-activation or neutron-capture gamma rays. Stability of the cryostat with changing hydrostatic pressure, absence of vibration, lack of need for power to the cryostat during operation, and freedom of orientation make the method desirable for borehole, undersea, space, and some laboratory applications. ?? 1972.

Active Neutron-Based Interrogation System with D-D Neutron Source for Detection of Special Nuclear Materials

NASA Astrophysics Data System (ADS)

Takahashi, Y.; Misawa, T.; Yagi, T.; Pyeon, C. H.; Kimura, M.; Masuda, K.; Ohgaki, H.

2015-10-01

The detection of special nuclear materials (SNM) is an important issue for nuclear security. The interrogation systems used in a sea port and an airport are developed in the world. The active neutron-based interrogation system is the one of the candidates. We are developing the active neutron-based interrogation system with a D-D fusion neutron source for the nuclear security application. The D-D neutron source is a compact discharge-type fusion neutron source called IEC (Inertial-Electrostatic Confinement fusion) device which provides 2.45 MeV neutrons. The nuclear materials emit the highenergy neutrons by fission reaction. High-energy neutrons with energies over 2.45 MeV amount to 30% of all the fission neutrons. By using the D-D neutron source, the detection of SNMs is considered to be possible with the attention of fast neutrons if there is over 2.45 MeV. Ideally, neutrons at En>2.45 MeV do not exist if there is no nuclear materials. The detection of fission neutrons over 2.45 MeV are hopeful prospect for the detection of SNM with a high S/N ratio. In the future, the experiments combined with nuclear materials and a D-D neutron source will be conducted. Furthermore, the interrogation system will be numerically investigated by using nuclear materials, a D-D neutron source, and a steel container.

Determination of U, Th and K in bricks by gamma-ray spectrometry, X-ray fluorescence analysis and neutron activation analysis

NASA Astrophysics Data System (ADS)

Bártová, H.; Kučera, J.; Musílek, L.; Trojek, T.; Gregorová, E.

2017-11-01

Knowledge of the content of natural radionuclides in bricks can be important in some cases in dosimetry and application of ionizing radiation. Dosimetry of naturally occurring radionuclides in matter (NORM) in general is one of them, the other one, related to radiation protection, is radon exposure evaluation, and finally, it is needed for the thermoluminescence (TL) dating method. The internal dose rate inside bricks is caused mostly by contributions of the natural radionuclides 238U, 232Th, radionuclides of their decay chains, and 40K. The decay chain of 235U is usually much less important. The concentrations of 238U, 232Th and 40K were measured by various methods, namely by gamma-ray spectrometry, X-ray fluorescence analysis (XRF), and neutron activation analysis (NAA) which was used as a reference method. These methods were compared from the point of view of accuracy, limit of detection (LOD), amount of sample needed and sample handling, time demands, and instrument availability.

Determination and distribution of rare earth elements in beach rock samples using instrumental neutron activation analysis (INAA)

NASA Astrophysics Data System (ADS)

Ravisankar, R.; Manikandan, E.; Dheenathayalu, M.; Rao, Brahmaji; Seshadreesan, N. P.; Nair, K. G. M.

2006-10-01

Beach rocks are a peculiar type of formation when compared to other types of rocks. Rare earth element (REE) concentrations in beach rock samples collected from the South East Coast of Tamilnadu, India, have been measured using the instrumental neutron activation analysis (INAA) single comparator K0 method. The irradiations were carried out using a thermal neutron flux of ˜10 11 n cm -2 s -1 at 20 kW power using the Kalpakkam mini reactor (KAMINI), IGCAR, Kalpakkam, Tamilnadu. Accuracy and precision were evaluated by assaying irradiated standard reference material (SRM 1646a estuarine sediment). The results being found to be in good agreement with certified values. REE elements have been determined from 15 samples using high-resolution gamma spectrometry. The geochemical behavior of REE in beach rock, in particular REE (chondrite-normalized) pattern has been studied.

Hydrogen analysis for granite using proton-proton elastic recoil coincidence spectrometry.

PubMed

Komatsubara, T; Sasa, K; Ohshima, H; Kimura, H; Tajima, Y; Takahashi, T; Ishii, S; Yamato, Y; Kurosawa, M

2008-07-01

In an effort to develop DS02, a new radiation dosimetry system for the atomic bomb survivors of Hiroshima and Nagasaki, measurements of neutron-induced activities have provided valuable information to reconstruct the radiation situation at the time of the bombings. In Hiroshima, the depth profile of (152)Eu activity measured in a granite pillar of the Motoyasu Bridge (128 m from the hypocenter) was compared with that calculated using the DS02 methodology. For calculation of the (152)Eu production due to the thermal-neutron activation reaction, (151)Eu(n,gamma)(152)Eu, information on the hydrogen content in granite is important because the transport and slowing-down process of neutrons penetrating into the pillar is strongly affected by collisions with the protons of hydrogen. In this study, proton-proton elastic recoil coincidence spectrometry has been used to deduce the proton density in the Motoyasu pillar granite. Slices of granite samples were irradiated by a 20 MeV proton beam, and the energies of scattered and recoil protons were measured with a coincidence method. The water concentration in the pillar granite was evaluated to be 0.30 +/- 0.07%wt. This result is consistent with earlier data on adsorptive water (II) and bound water obtained by the Karl Fisher method.

Neutron threshold activation detectors (TAD) for the detection of fissions

NASA Astrophysics Data System (ADS)

Gozani, Tsahi; Stevenson, John; King, Michael J.

2011-10-01

Prompt fission neutrons are one of the strongest signatures of the fission process. Depending on the fission inducing radiation, their average number ranges from 2.5 to 4 neutrons per fission. They are more energetic and abundant, by about 2 orders of magnitude, than the delayed neutrons (≈3 vs. ≈0.01) that are commonly used as indicators for the presence of fissionable materials. The detection of fission prompt neutrons, however, has to be done in the presence of extremely intense probing radiation that stimulated them. During irradiation, the fission stimulation radiation, X-rays or neutrons, overwhelms the neutron detectors and temporarily incapacitate them. Consequently, by the time the detectors recover from the source radiation, fission prompt neutrons are no longer emitted. In order to measure the prompt fission signatures under these circumstances, special measures are usually taken with the detectors such as heavy shielding with collimation, use of inefficient geometries, high pulse height bias and gamma-neutron separation via pulse-shape discrimination with an appropriate organic scintillator. These attempts to shield the detector from the flash of radiation result in a major loss of sensitivity. It can lead to a complete inability to detect the fission prompt neutrons. In order to overcome the blinding induced background from the source radiation, the detection of prompt fission neutrons needs to occur long after the fission event and after the detector has fully recovered from the source overload. A new approach to achieve this is to detect the delayed activation induced by the fission neutrons. The approach demonstrates a good sensitivity in adverse overload situations (gamma and neutron "flash") where fission prompt neutrons could normally not be detected. The new approach achieves the required temporal separation between the detection of prompt neutrons and the detector overload by the neutron activation of the detector material. The technique, called Threshold Activation Detection (TAD), is to utilize appropriate substances that can be selectively activated by the fission neutrons and not by the source radiation and then measure the radioactively decaying activation products (typically beta and gamma rays) well after the source pulse. The activation material should possess certain properties: a suitable half-life of the order of seconds; an energy threshold below which the numerous source neutrons will not activate it (e.g., 3 MeV); easily detectable activation products (typically >1 MeV beta and gamma rays) and have a usable cross-section for the selected reaction. Ideally the substance would be a part of the scintillator. There are several good material candidates for the TAD, including fluorine, which is a major constituent of available scintillators such as BaF 2, CaF 2 and hydrogen free liquid fluorocarbon. Thus the fluorine activation products, in particular the beta particles, can be measured with a very high efficiency in the detector. The principles, applications and experimental results obtained with the fluorine based TAD are discussed.

Photonuclear Contributions to SNS Pulse Shapes

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

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

Short-pulsed sources like the Spallation Neutron Source (SNS) and ISIS produce bursts of neutron pulses at rates of 10-60 Hz, with sub-microsecond proton pulses impacting on high-Z target materials. Moderators are grouped around the target to receive the fast neutrons generated from spallation reactions to moderate them effciently to thermal and sub-thermal energies and to feed narrow neutron pulses to neutron scattering instruments. The scattering instruments use the neutrons as a probe for material investigations, and make use of time-of-flight (TOF) methods for resolving the neutron energy. The energy resolution of scattering instruments depends on the narrow time-structure of themore » neutron pulses, while neutrons in the long tail of the emission time distributions can degrade the instrument performance and add undesired background to measurements. The SNS neutronics team is investigating a possible source term impacting the background at short-pulsed spallation sources. The ISIS TS2 project claims to have significantly reduced neutron scattering instrument background levels by the elimination or reduction of iron shielding in the target-moderator-reflector assembly. An alternative hypothesis, also proposed by ISIS, suggests that this apparent reduction arises from moving beamline shielding away from the neutron guide channels, reducing albedo down the beamlines. In both hypotheses, the background neutrons in question are believed to be generated by photonuclear reactions. If the background neutrons are indeed generated via photonuclear channels, then they are generated in a time-dependent fashion, since most of the high-energy photons capable of inducing photonuclear production are gone within a few microseconds following the proton pulse. To evaluate this e ect, we have enabled photonuclear reactions in a series of studies for the SNS first target station (FTS) taking advantage of its Monte Carlo model. Using a mixture of ENDF/B VII.0 and TENDL-2014 photonuclear cross sections available and the CEM03 physics model within MCNPX 2.6.0 in the simulation, we are able to estimate the impact of photoneutron production on both overall neutron production and delayed neutron production. We find that a significant number of photon-induced neutrons are produced a few milliseconds after the proton pulse, following prompt gamma emission through the capture of neutrons in the slowing-down and thermalization processes. We name these "slowing-down delayed neutrons" to distinguish them from either "activation-delayed neutrons" or "beta-delayed neutrons." The beta-delayed and activation-delayed neutrons were not part of this study, and will be addressed elsewhere. While these other delayed neutron channels result in the time-independent (constant) production of fast neutrons outside of the prompt pulse, the slowing-down delayed neutrons also a ect the shape of the pulses. Although numerically insignificant in most cases, we describe a set of scenarios related to T0-chopper operation in which the slowing-down delayed neutrons may be important.« less

Evaluation of equivalent dose from neutrons and activation products from a 15-MV X-ray LINAC

PubMed Central

Israngkul-Na-Ayuthaya, Isra; Suriyapee, Sivalee; Pengvanich, Phongpheath

2015-01-01

A high-energy photon beam that is more than 10 MV can produce neutron contamination. Neutrons are generated by the [γ,n] reactions with a high-Z target material. The equivalent neutron dose and gamma dose from activation products have been estimated in a LINAC equipped with a 15-MV photon beam. A Monte Carlo simulation code was employed for neutron and photon dosimetry due to mixed beam. The neutron dose was also experimentally measured using the Optically Stimulated Luminescence (OSL) under various conditions to compare with the simulation. The activation products were measured by gamma spectrometer system. The average neutron energy was calculated to be 0.25 MeV. The equivalent neutron dose at the isocenter obtained from OSL measurement and MC calculation was 5.39 and 3.44 mSv/Gy, respectively. A gamma dose rate of 4.14 µSv/h was observed as a result of activations by neutron inside the treatment machine. The gamma spectrum analysis showed 28Al, 24Na, 54Mn and 60Co. The results confirm that neutrons and gamma rays are generated, and gamma rays remain inside the treatment room after the termination of X-ray irradiation. The source of neutrons is the product of the [γ,n] reactions in the machine head, whereas gamma rays are produced from the [n,γ] reactions (i.e. neutron activation) with materials inside the treatment room. The most activated nuclide is 28Al, which has a half life of 2.245 min. In practice, it is recommended that staff should wait for a few minutes (several 28Al half-lives) before entering the treatment room after the treatment finishes to minimize the dose received. PMID:26265661

Phase 3 experiments of the JAERI/USDOE collaborative program on fusion blanket neutronics. Volume 1: Experiment

NASA Astrophysics Data System (ADS)

Oyama, Yukio; Konno, Chikara; Ikeda, Yujiro; Maekawa, Fujio; Kosako, Kazuaki; Nakamura, Tomoo; Maekawa, Hiroshi; Youssef, Mahmoud Z.; Kumar, Anil; Abdou, Mohamed A.

1994-02-01

A pseudo-line source has been realized by using an accelerator based D-T point neutron source. The pseudo-line source is obtained by time averaging of continuously moving point source or by superposition of finely distributed point sources. The line source is utilized for fusion blanket neutronics experiments with an annular geometry so as to simulate a part of a tokamak reactor. The source neutron characteristics were measured for two operational modes for the line source, continuous and step-wide modes, with the activation foil and the NE213 detectors, respectively. In order to give a source condition for a successive calculational analysis on the annular blanket experiment, the neutron source characteristics was calculated by a Monte Carlo code. The reliability of the Monte Carlo calculation was confirmed by comparison with the measured source characteristics. The shape of the annular blanket system was a rectangular with an inner cavity. The annular blanket was consist of 15 mm-thick first wall (SS304) and 406 mm-thick breeder zone with Li2O at inside and Li2CO3 at outside. The line source was produced at the center of the inner cavity by moving the annular blanket system in the span of 2 m. Three annular blanket configurations were examined; the reference blanket, the blanket covered with 25 mm thick graphite armor and the armor-blanket with a large opening. The neutronics parameters of tritium production rate, neutron spectrum and activation reaction rate were measured with specially developed techniques such as multi-detector data acquisition system, spectrum weighting function method and ramp controlled high voltage system. The present experiment provides unique data for a higher step of benchmark to test a reliability of neutronics design calculation for a realistic tokamak reactor.

Method and apparatus for determining the content and distribution of a thermal neutron absorbing material in an object

DOEpatents

Crane, Thomas W.

1986-01-01

The disclosure is directed to an apparatus and method for determining the content and distribution of a thermal neutron absorbing material within an object. Neutrons having an energy higher than thermal neutrons are generated and thermalized. The thermal neutrons are detected and counted. The object is placed between the neutron generator and the neutron detector. The reduction in the neutron flux corresponds to the amount of thermal neutron absorbing material in the object. The object is advanced past the neutron generator and neutron detector to obtain neutron flux data for each segment of the object. The object may comprise a space reactor heat pipe and the thermal neutron absorbing material may comprise lithium.

Method and apparatus for determining the content and distribution of a thermal neutron absorbing material in an object

DOEpatents

Crane, T.W.

1983-12-21

The disclosure is directed to an apparatus and method for determining the content and distribution of a thermal neutron absorbing material within an object. Neutrons having an energy higher than thermal neutrons are generated and thermalized. The thermal neutrons are detected and counted. The object is placed between the neutron generator and the neutron detector. The reduction in the neutron flux corresponds to the amount of thermal neutron absorbing material in the object. The object is advanced past the neutron generator and neutron detector to obtain neutron flux data for each segment of the object. The object may comprise a space reactor heat pipe and the thermal neutron absorbing material may comprise lithium.

Detection of special nuclear material from delayed neutron emission induced by a dual-particle monoenergetic source

DOE PAGES

Mayer, Michael F.; Nattress, J.; Jovanovic, I.

2016-06-27

Detection of unique signatures of special nuclear materials is critical for their interdiction in a variety of nuclear security and nonproliferation scenarios. We report on the observation of delayed neutrons from fission of uranium induced in dual-particle active interrogation based on the 11B(d,n γ) 12C nuclear reaction. Majority of the fissions are attributed to fast fission induced by the incident quasi-monoenergetic neutrons. A Li-doped glass–polymer composite scintillation neutron detector, which displays excellent neutron/γ discrimination at low energies, was used in the measurements, along with a recoil-based liquid scintillation detector. Time- dependent buildup and decay of delayed neutron emission from 238Umore » were measured between the interrogating beam pulses and after the interrogating beam was turned off, respectively. Characteristic buildup and decay time profiles were compared to the common parametrization into six delayed neutron groups, finding a good agreement between the measurement and nuclear data. Furthermore, this method is promising for detecting fissile and fissionable materials in cargo scanning applications and can be readily integrated with transmission radiography using low-energy nuclear reaction sources.« less

Compensated gadolinium-loaded plastic scintillators for thermal neutron detection (and counting)

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

Dumazert, Jonathan; Coulon, Romain; Bertrand, Guillaume H. V.

2015-07-01

Plastic scintillator loading with gadolinium-rich organometallic complexes shows a high potential for the deployment of efficient and cost-effective neutron detectors. Due to the low-energy photon and electron signature of thermal neutron capture by gadolinium-155 and gadolinium-157, alternative treatment to Pulse Shape Discrimination has to be proposed in order to display a trustable count rate. This paper discloses the principle of a compensation method applied to a two-scintillator system: a detection scintillator interacts with photon radiation and is loaded with gadolinium organometallic compound to become a thermal neutron absorber, while a non-gadolinium loaded compensation scintillator solely interacts with the photon partmore » of the incident radiation. Posterior to the nonlinear smoothing of the counting signals, a hypothesis test determines whether the resulting count rate after photon response compensation falls into statistical fluctuations or provides a robust image of a neutron activity. A laboratory prototype is tested under both photon and neutron irradiations, allowing us to investigate the performance of the overall compensation system in terms of neutron detection, especially with regards to a commercial helium-3 counter. The study reveals satisfactory results in terms of sensitivity and orientates future investigation toward promising axes. (authors)« less

Neutron radiography in Indian space programme

NASA Astrophysics Data System (ADS)

Viswanathan, K.

1999-11-01

Pyrotechnic devices are indispensable in any space programme to perform such critical operations as ignition, stage separation, solar panel deployment, etc. The nature of design and configuration of different types of pyrotechnic devices, and the type of materials that are put in their construction make the inspection of them with thermal neutrons more favourable than any other non destructive testing methods. Although many types of neutron sources are available for use, generally the radiographic quality/exposure duration and cost of source run in opposite directions even after four decades of research and development. But in the area of space activity, by suitably combining the X-ray and neutron radiographic requirements, the inspection of the components can be made economically viable. This is demonstrated in the Indian space programme by establishing a 15 MeV linear accelerator based neutron generator facility to inspect medium to giant solid propellant boosters by X-ray inspection and all types of critical pyro and some electronic components by neutron radiography. Since the beam contains unacceptable gamma, transfer imaging technique has been evolved and the various parameters have been optimised to get a good quality image.

Method for the melting of metals

DOEpatents

White, Jack C.; Traut, Davis E.

1992-01-01

A method of quantitatively determining the molten pool configuration in melting of metals. The method includes the steps of introducing hafnium metal seeds into a molten metal pool at intervals to form ingots, neutron activating the ingots and determining the hafnium location by radiometric means. Hafnium possesses exactly the proper metallurgical and radiochemical properties for this use.

Thick-foils activation technique for neutron spectrum unfolding with the MINUIT routine-Comparison with GEANT4 simulations

NASA Astrophysics Data System (ADS)

Vagena, E.; Theodorou, K.; Stoulos, S.

2018-04-01

Neutron activation technique has been applied using a proposed set of twelve thick metal foils (Au, As, Cd, In, Ir, Er, Mn, Ni, Se, Sm, W, Zn) for off-site measurements to obtain the neutron spectrum over a wide energy range (from thermal up to a few MeV) in intense neutron-gamma mixed fields such as around medical Linacs. The unfolding procedure takes into account the activation rates measured using thirteen (n , γ) and two (n , p) reactions without imposing a guess solution-spectrum. The MINUIT minimization routine unfolds a neutron spectrum that is dominated by fast neutrons (70%) peaking at 0.3 MeV, while the thermal peak corresponds to the 15% of the total neutron fluence equal to the epithermal-resonances area. The comparison of the unfolded neutron spectrum against the simulated one with the GEANT4 Monte-Carlo code shows a reasonable agreement within the measurement uncertainties. Therefore, the proposed set of activation thick-foils could be a useful tool in order to determine low flux neutrons spectrum in intense mixed field.

Active Interrogation of Sensitive Nuclear Material Using Laser Driven Neutron Beams

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

Favalli, Andrea; Roth, Markus

2015-05-01

An investigation of the viability of a laser-driven neutron source for active interrogation is reported. The need is for a fast, movable, operationally safe neutron source which is energy tunable and has high-intensity, directional neutron production. Reasons for the choice of neutrons and lasers are set forth. Results from the interrogation of an enriched U sample are shown.

Mineral exploration and soil analysis using in situ neutron activation

USGS Publications Warehouse

Senftle, F.E.; Hoyte, A.F.

1966-01-01

A feasibility study has been made to operate by remote control an unshielded portable positive-ion accelerator type neutron source to induce activities in the ground or rock by "in situ" neutron irradiation. Selective activation techniques make it possible to detect some thirty or more elements by irradiating the ground for periods of a few minutes with either 3-MeV or 14-MeV neutrons. The depth of penetration of neutrons, the effect of water content of the soil on neutron moderation, gamma ray attenuation in the soil and other problems are considered. The analysis shows that, when exploring for most elements of economic interest, the reaction 2H(d,n)3He yielding ??? 3-MeV neutrons is most practical to produce a relatively uniform flux of neutrons of less than 1 keV to a depth of 19???-20???. Irradiation with high energy neutrons (??? 14 MeV) can also be used and may be better suited for certain problems. However, due to higher background and lower sensitivity for the heavy minerals, it is not a recommended neutron source for general exploration use. Preliminary experiments have been made which indicate that neutron activation in situ is feasible for a mineral exploration or qualititative soil analysis. ?? 1976.

Quantitative NDA measurements of advanced reprocessing product materials containing uranium, neptunium, plutonium, and americium

NASA Astrophysics Data System (ADS)

Goddard, Braden

The ability of inspection agencies and facility operators to measure powders containing several actinides is increasingly necessary as new reprocessing techniques and fuel forms are being developed. These powders are difficult to measure with nondestructive assay (NDA) techniques because neutrons emitted from induced and spontaneous fission of different nuclides are very similar. A neutron multiplicity technique based on first principle methods was developed to measure these powders by exploiting isotope-specific nuclear properties, such as the energy-dependent fission cross sections and the neutron induced fission neutron multiplicity. This technique was tested through extensive simulations using the Monte Carlo N-Particle eXtended (MCNPX) code and by one measurement campaign using the Active Well Coincidence Counter (AWCC) and two measurement campaigns using the Epithermal Neutron Multiplicity Counter (ENMC) with various (alpha,n) sources and actinide materials. Four potential applications of this first principle technique have been identified: (1) quantitative measurement of uranium, neptunium, plutonium, and americium materials; (2) quantitative measurement of mixed oxide (MOX) materials; (3) quantitative measurement of uranium materials; and (4) weapons verification in arms control agreements. This technique still has several challenges which need to be overcome, the largest of these being the challenge of having high-precision active and passive measurements to produce results with acceptably small uncertainties.

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

A feasibility study of the Tehran research reactor as a neutron source for BNCT.

PubMed

Kasesaz, Yaser; Khalafi, Hossein; Rahmani, Faezeh; Ezati, Arsalan; Keyvani, Mehdi; Hossnirokh, Ashkan; Shamami, Mehrdad Azizi; Monshizadeh, Mahdi

2014-08-01

Investigation on the use of the Tehran Research Reactor (TRR) as a neutron source for Boron Neutron Capture Therapy (BNCT) has been performed by calculating and measuring energy spectrum and the spatial distribution of neutrons in all external irradiation facilities, including six beam tubes, thermal column, and the medical room. Activation methods with multiple foils and a copper wire have been used for the mentioned measurements. The results show that (1) the small diameter and long length beam tubes cannot provide sufficient neutron flux for BNCT; (2) in order to use the medical room, the TRR core should be placed in the open pool position, in this situation the distance between the core and patient position is about 400 cm, so neutron flux cannot be sufficient for BNCT; and (3) the best facility which can be adapted for BNCT application is the thermal column, if all graphite blocks can be removed. The epithermal and fast neutron flux at the beginning of this empty column are 4.12×10(9) and 1.21×10(9) n/cm(2)/s, respectively, which can provide an appropriate neutron beam for BNCT by designing and constructing a proper Beam Shaping Assembly (BSA) structure. Copyright © 2014 Elsevier Ltd. All rights reserved.

MCNP-REN - A Monte Carlo Tool for Neutron Detector Design Without Using the Point Model

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

Abhold, M.E.; Baker, M.C.

1999-07-25

The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo N-Particle code (MCNP) was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP - Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program (TAP) predict neutron detector response without using the pointmore » reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of MOX fresh fuel made using the Underwater Coincidence Counter (UWCC) as well as measurements of HEU reactor fuel using the active neutron Research Reactor Fuel Counter (RRFC) are compared with calculations. The method used in MCNP-REN is demonstrated to be fundamentally sound and shown to eliminate the need to use the point model for detector performance predictions.« less

Methods for Probing Magnetic Films with Neutrons

NASA Astrophysics Data System (ADS)

Kozhevnikov, S. V.; Ott, F.; Radu, F.

2018-03-01

We review various methods in the investigation of magnetic films with neutrons, including those based on the effects of Larmor precession, Zeeman spatial splitting of the beam, neutron spin resonance, and polarized neutron channeling. The underlying principles, examples of the investigated systems, specific features, applications, and perspectives of these methods are discussed.

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

Ryan, M.T.; Butler, H.M.; Gupton, E.D.

The UCC-ND Employee Identification Badge contains an indium foil disc that is intended for use as a dosimetry screening device in the event of a criticality accident. While it is recognized that indium is not a precise mixed neutron-gamma dosimeter, its activation by neutrons provides adequate means for separating potentially exposed persons into three groups. These groups are: (1) personnel exposed below annual dose limits, (2) personnel exposed above annual dose limits but below 25 rem, and (3) personnel exposed above 25 rem. This screening procedure is designed to facilitate dosimeter processing in order to meet regulatory reporting requirements. Amore » quick method of interpreting induced activity measurements is presented and discussed.« less

In situ investigation of deformation mechanisms in magnesium-based metal matrix composites

NASA Astrophysics Data System (ADS)

Farkas, Gergely; Choe, Heeman; Máthis, Kristián; Száraz, Zoltán; Noh, Yoonsook; Trojanová, Zuzanka; Minárik, Peter

2015-07-01

We studied the effect of short fibers on the mechanical properties of a magnesium alloy. In particular, deformation mechanisms in a Mg-Al-Sr alloy reinforced with short alumina fibers were studied in situ using neutron diffraction and acoustic emission methods. The fibers' plane orientation with respect to the loading axis was found to be a key parameter, which influences the acting deformation processes, such as twinning or dislocation slip. Furthermore, the twinning activity was much more significant in samples with parallel fiber plane orientation, which was confirmed by both acoustic emission and electron backscattering diffraction results. Neutron diffraction was also used to assist in analyzing the acoustic emission and electron backscattering diffraction results. The simultaneous application of the two in situ methods, neutron diffraction and acoustic emission, was found to be beneficial for obtaining complementary datasets about the twinning and dislocation slip in the magnesium alloys and composites used in this study.

A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

DOE PAGES

Menlove, Howard Olsen; Belian, Anthony P.; Geist, William H.; ...

2017-10-07

The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd 2O 3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate themore » 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. Here, we have named the system the fast-neutron passive collar (FNPC).« less

A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

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

Menlove, Howard Olsen; Belian, Anthony P.; Geist, William H.

The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd 2O 3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate themore » 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. Here, we have named the system the fast-neutron passive collar (FNPC).« less

A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

NASA Astrophysics Data System (ADS)

Menlove, Howard; Belian, Anthony; Geist, William; Rael, Carlos

2018-01-01

The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd2O3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate the 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. We have named the system the fast-neutron passive collar (FNPC).

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.

Neutron capture studies with a short flight path

NASA Astrophysics Data System (ADS)

Walter, Stephan; Heil, Michael; Käppeler, Franz; Plag, Ralf; Reifarth, René

The time of flight (TOF) method is an important tool for the experimental determination of neu- tron capture cross sections which are needed for s-process nucleosynthesis in general, and for analyses of branchings in the s-process reaction path in particular. So far, sample masses of at least several milligrams are required to compensate limitations in the currently available neutron fluxes. This constraint leads to unacceptable backgrounds for most of the relevant unstable branch point nuclei, due to the decay activity of the sample. A possible solution has been proposed by the NCAP project at the University of Frankfurt. A first step in this direction is reported here, which aims at enhancing the sensitivity of the Karlsruhe TOF array by reducing the neutron flight path to only a few centimeters. Though sample masses in the microgram regime can be used by this approach, the increase in neutron flux has to be paid by a higher background from the prompt flash related to neutron production. Test measurements with Au samples are reported.

Neutron Capture gamma ENDF libraries for modeling and identification of neutron sources

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

Sleaford, B

2007-10-29

There are a number of inaccuracies and data omissions with respect to gammas from neutron capture in the ENDF libraries used as field reference information and by modeling codes used in JTOT. As the use of Active Neutron interrogation methods is expanded, these shortfalls become more acute. A new, more accurate and complete evaluated experimental database of gamma rays (over 35,000 lines for 262 isotopes up to U so far) from thermal neutron capture has recently become available from the IAEA. To my knowledge, none of this new data has been installed in ENDF libraries and disseminated. I propose tomore » upgrade libraries of {sup 184,186}W, {sup 56}Fe, {sup 204,206,207}Pb, {sup 104}Pd, and {sup 19}F the 1st year. This will involve collaboration with Richard Firestone at LBL in evaluating the data and installing it in the libraries. I will test them with the transport code MCNP5.« less

Expert system for surveillance and diagnosis of breach fuel elements

DOEpatents

Gross, K.C.

1988-01-21

An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil area of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor. 2 figs.

Expert system for surveillance and diagnosis of breach fuel elements

DOEpatents

Gross, Kenny C.

1989-01-01

An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil areas of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor.

Neutron spectral measurements in an intense photon field associated with a high-energy x-ray radiotherapy machine.

PubMed

Holeman, G R; Price, K W; Friedman, L F; Nath, R

1977-01-01

High-energy x-ray radiotherapy machines in the supermegavoltage region generate complex neutron energy spectra which make an exact evaluation of neutron shielding difficult. Fast neutrons resulting from photonuclear reactions in the x-ray target and collimators undergo successive collisions in the surrounding materials and are moderated by varying amounts. In order to examine the neutron radiation exposures quantitatively, the neutron energy spectra have been measured inside and outside the treatment room of a Sagittaire medical linear accelerator (25-MV x rays) located at Yale-New Haven Hospital. The measurements were made using a Bonner spectrometer consisting of 2-, 3-, 5-, 8-, 10- and 12-in.-diameter polyethylene spheres with 6Li and 7Li thermoluminescent dosimeter (TLD) chips at the centers, in addition to bare and cadmium-covered chips. The individual TLD chips were calibrated for neutron and photon response. The spectrometer was calibrated using a known PuBe spectrum Spectrometer measurements were made at Yale Electron Accelerator Laboratory and results compared with a neutron time-of-flight spectrometer and an activation technique. The agreement between the results from these independent methods is found to be good, except for the measurements in the direct photon beam. Quality factors have been inferred for the neutron fields inside and outside the treatment room. Values of the inferred quality factors fall primarily between 4 and 8, depending on location.

Benchmark test of neutron transport calculations: indium, nickel, gold, europium, and cobalt activation with and without energy moderated fission neutrons by iron simulating the Hiroshima atomic bomb casing.

PubMed

Iwatani, K; Hoshi, M; Shizuma, K; Hiraoka, M; Hayakawa, N; Oka, T; Hasai, H

1994-10-01

A benchmark test of the Monte Carlo neutron and photon transport code system (MCNP) was performed using a bare- and energy-moderated 252Cf fission neutron source which was obtained by transmission through 10-cm-thick iron. An iron plate was used to simulate the effect of the Hiroshima atomic bomb casing. This test includes the activation of indium and nickel for fast neutrons and gold, europium, and cobalt for thermal and epithermal neutrons, which were inserted in the moderators. The latter two activations are also to validate 152Eu and 60Co activity data obtained from the atomic bomb-exposed specimens collected at Hiroshima and Nagasaki, Japan. The neutron moderators used were Lucite and Nylon 6 and the total thickness of each moderator was 60 cm or 65 cm. Measured activity data (reaction yield) of the neutron-irradiated detectors in these moderators decreased to about 1/1,000th or 1/10,000th, which corresponds to about 1,500 m ground distance from the hypocenter in Hiroshima. For all of the indium, nickel, and gold activity data, the measured and calculated values agreed within 25%, and the corresponding values for europium and cobalt were within 40%. From this study, the MCNP code was found to be accurate enough for the bare- and energy-moderated 252Cf neutron activation calculations of these elements using moderators containing hydrogen, carbon, nitrogen, and oxygen.

Study of (n,2n) reaction on 191,193Ir isotopes and isomeric cross section ratios

NASA Astrophysics Data System (ADS)

Vlastou, R.; Kalamara, A.; Kokkoris, M.; Patronis, N.; Serris, M.; Georgoulakis, M.; Hassapoglou, S.; Kobothanasis, K.; Axiotis, M.; Lagoyannis, A.

2017-09-01

The cross section of 191Ir(n,2n)190Irg+m1 and 191Ir(n,2n)190Irm2 reactions has been measured at 17.1 and 20.9 MeV neutron energies at the 5.5 MV tandem T11/25 Accelerator Laboratory of NCSR "Demokritos", using the activation method. The neutron beams were produced by means of the 3H(d,n)4He reaction at a flux of the order of 2 × 105 n/cm2s. The neutron flux has been deduced implementing the 27Al(n,α) reaction, while the flux variation of the neutron beam was monitored by using a BF3 detector. The 193Ir(n,2n)192Ir reaction cross section has also been determined, taking into account the contribution from the contaminant 191Ir(n,γ)192Ir reaction. The correction method is based on the existing data in ENDF for the contaminant reaction, convoluted with the neutron spectra which have been extensively studied by means of simulations using the NeusDesc and MCNP codes. Statistical model calculations using the code EMPIRE 3.2.2 and taking into account pre-equilibrium emission, have been performed on the data measured in this work as well as on data reported in literature.

System and method for delivery of neutron beams for medical therapy

DOEpatents

Nigg, David W.; Wemple, Charles A.

1999-01-01

A neutron delivery system that provides improved capability for tumor control during medical therapy. The system creates a unique neutron beam that has a bimodal or multi-modal energy spectrum. This unique neutron beam can be used for fast-neutron therapy, boron neutron capture therapy (BNCT), or both. The invention includes both an apparatus and a method for accomplishing the purposes of the invention.

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.

Arsenic activation neutron detector

DOEpatents

Jacobs, E.L.

1980-01-28

A detector of bursts of neutrons from a deuterium-deuteron reaction includes a quantity of arsenic adjacent a gamma detector such as a scintillator and photomultiplier tube. The arsenic is activated by the 2.5-MeV neutrons to release gamma radiation which is detected to give a quantitative representation of detected neutrons.

Arsenic activation neutron detector

DOEpatents

Jacobs, Eddy L.

1981-01-01

A detector of bursts of neutrons from a deuterium-deuteron reaction includes a quantity of arsenic adjacent a gamma detector such as a scintillator and photomultiplier tube. The arsenic is activated by the 2.5 Mev neutrons to release gamma radiation which is detected to give a quantitative representation of detected neutrons.

Differential neutron energy spectra measured on spacecraft low Earth orbit

NASA Technical Reports Server (NTRS)

Benton, E. V.; Frank, A. L.; Dudkin, E. V.; Potapov, Yu. V.; Akopova, A. B.; Melkumyan, L. V.

1995-01-01

Two methods for measuring neutrons in the range from thermal energies to dozens of MeV were used. In the first method, alpha-particles emitted from the (sup 6) Li(n.x)T reaction are detected with the help of plastic nuclear track detectors, yielding results on thermal and resonance neutrons. Also, fission foils are used to detect fast neutrons. In the second method, fast neutrons are recorded by nuclear photographic emulsions (NPE). The results of measurements on board various satellites are presented. The neutron flux density does not appear to correlate clearly with orbital parameters. Up to 50% of neutrons are due to albedo neutrons from the atmosphere while the fluxes inside the satellites are 15-20% higher than those on the outside. Estimates show that the neutron contribution to the total equivalent radiation dose reaches 20-30%.

Applicability of the two-angle differential method to response measurement of neutron-sensitive devices at the RCNP high-energy neutron facility

NASA Astrophysics Data System (ADS)

Masuda, Akihiko; Matsumoto, Tetsuro; Iwamoto, Yosuke; Hagiwara, Masayuki; Satoh, Daiki; Sato, Tatsuhiko; Iwase, Hiroshi; Yashima, Hiroshi; Nakane, Yoshihiro; Nishiyama, Jun; Shima, Tatsushi; Tamii, Atsushi; Hatanaka, Kichiji; Harano, Hideki; Nakamura, Takashi

2017-03-01

Quasi-monoenergetic high-energy neutron fields induced by 7Li(p,n) reactions are used for the response evaluation of neutron-sensitive devices. The quasi-monoenergetic high-energy field consists of high-energy monoenergetic peak neutrons and unwanted continuum neutrons down to the low-energy region. A two-angle differential method has been developed to compensate for the effect of the continuum neutrons in the response measurements. In this study, the two-angle differential method was demonstrated for Bonner sphere detectors, which are typical examples of moderator-based neutron-sensitive detectors, to investigate the method's applicability and its dependence on detector characteristics. Experiments were performed under 96-387 MeV quasi-monoenergetic high-energy neutron fields at the Research Center for Nuclear Physics (RCNP), Osaka University. The measurement results for large high-density polyethylene (HDPE) sphere detectors agreed well with Monte Carlo calculations, which verified the adequacy of the two-angle differential method. By contrast, discrepancies were observed in the results for small HDPE sphere detectors and metal-induced sphere detectors. The former indicated that detectors that are particularly sensitive to low-energy neutrons may be affected by penetrating neutrons owing to the geometrical features of the RCNP facility. The latter discrepancy could be consistently explained by a problem in the evaluated cross-section data for the metals used in the calculation. Through those discussions, the adequacy of the two-angle differential method was experimentally verified, and practical suggestions were made pertaining to this method.

Fission Signatures for Nuclear Material Detection

NASA Astrophysics Data System (ADS)

Gozani, Tsahi

2009-06-01

Detection and interdiction of nuclear materials in all forms of transport is one of the most critical security issues facing the United States and the rest of the civilized world. Naturally emitted gamma rays by these materials, while abundant and detectable when unshielded, are low in energy and readily shielded. X-ray radiography is useful in detecting the possible presence of shielding material. Positive detection of concealed nuclear materials requires methods which unequivocally detect specific attributes of the materials. These methods typically involve active interrogation by penetrating radiation of neutrons, photons or other particles. Fortunately, nuclear materials, probed by various types of radiation, yield very unique and often strong signatures. Paramount among them are the detectable fission signatures, namely prompt neutrons and gamma rays, and delayed neutrons gamma rays. Other useful signatures are the nuclear states excited by neutrons, via inelastic scattering, or photons, via nuclear resonance fluorescence and absorption. The signatures are very different in magnitude, level of specificity, ease of excitation and detection, signal to background ratios, etc. For example, delayed neutrons are very unique to the fission process, but are scarce, have low energy, and hence are easily absorbed. Delayed gamma rays are more abundant but "featureless", and have a higher background from natural sources and more importantly, from activation due to the interrogation sources. The prompt fission signatures need to be measured in the presence of the much higher levels of probing radiation. This requires taking special measures to look for the signatures, sometimes leading to a significant sensitivity loss or a complete inability to detect them. Characteristic gamma rays induced in nuclear materials reflecting their nuclear structure, while rather unique, require very high intensity of interrogation radiation and very high resolution in energy and/or time. The trade off of signatures, their means of stimulation, and methods of detection, will be reviewed.

System and method for delivery of neutron beams for medical therapy

DOEpatents

Nigg, D.W.; Wemple, C.A.

1999-07-06

A neutron delivery system that provides improved capability for tumor control during medical therapy is disclosed. The system creates a unique neutron beam that has a bimodal or multi-modal energy spectrum. This unique neutron beam can be used for fast-neutron therapy, boron neutron capture therapy (BNCT), or both. The invention includes both an apparatus and a method for accomplishing the purposes of the invention. 5 figs.

Method of Operating a Neutronic Reactor

NASA Astrophysics Data System (ADS)

Fermi, Enrico; Szilard, Leo

This Patent is a later,1 almost faithful, copy of Patent No. 2,708,656 (which is then not reported in the present volume). This revised version was probably prepared (by the authors) in order to correct several misprints of the previous version. As emphasized in The New York Times of May 19, 1955, Patent No. 2,708,656, an "historic Patent, covering the first nuclear reactor", is the first one on this topic issued by the U.S. Patent Office, and served as a reference for the subsequent Patents on the same subject. In this long Patent, the theory, exper- imental data and principles of construction and operation of "any" type of nuclear reactor known at that time are discussed in an extremely detailed way. Various possible fission fragments produced by the reactor, several forms of the uranium employed (metal, oxide and so on, grouped in different geometrical forms), various materials adopted as moderators, several cooling systems, different geometries of the reactors, etc. are considered accurately. The theoretical description, centered around the achievement of a self-sustaining chain reaction, is exhaustive, and great attention is devoted to any possible cause of neutron loss, to the resonance capture of neutrons and to the effect of the presence of relevant impurities in the reactor. The chain production of neutrons in the pile is described in great detail, along with the theoretical arguments underlying the exponential experiment. The problem of the variation of the multiplication factor due to the production of radioactive elements, such as xenon, is discussed extensively. In particular it is pointed out that, although the initial production of xenon lowers the multiplication factor K due to its relevant neutron absorption, it subsequently increases again due to the decay of xenon into another isotope which absorbs fewer neutrons. The building up of reactors with solid (graphite) or liquid (heavy water) moderators is discussed, as well as other possible moderators such as light water or beryllium. In particular, the ratio is given of the absorption cross section to the scattering cross section for several moderators. Procedures for the purification of uranium are described as well. Several methods (i.e., the exponential pile or the "shotgun" method; see Patent No. 2.969,307) are reported for testing the purity against neutron absorption of different materials. The effect of the boron and vanadium impurities in the graphite and light water in the heavy water are considered. Different cooling systems for the reactors are considered and compared in the Patent, based on the circulation of a gas (typically, air) or a liquid (light or heavy water, diphenyl, etc.). The principles and practice for the construction, functioning and control of several kinds of reactors are reported in detail. One reactor considered in the present Patent is a low power uranium-graphite one without cooling system, where the active part consists in (small) cylinders of metallic uranium or pseudo-spheres of uranium oxide (or cylinders of U3O8). The control rods are made of steel with boron inserts, while limitation and safety rods are made of cadmium. In addition, an uranium-graphite pile cooled by air or even by water or diphenyl is considered. It is pointed out that dyphenil should usually be preferred with respect to water, due to its lower absorption of neutrons and to its higher boiling temperature, but the disadvantage related to its use is mainly due to the closed pumping system required and to the possible occurrence of polymerization which makes the fluid viscous. Another kind of reactor described in detail is made of uranium (vertical) bars immersed in heavy water. When, during the operation, heavy water is dissociated into D2 and O2, these two gaseous elements are carried by an inert gas (helium) into a recombination device. The control and safety rods are made of cadmium. Hybrid reactors composed of different lattices in the same neutronic reactor, in order to increase the multiplication factor K, are considered as well. A description of the possible uses of nuclear reactors, other than as power supplies, including the production of collimated beams of fast neutrons, the production of plutonium (a fissionable material usable in other reactors) or several other radioactive isotopes (for possible utilization in medicine) is as well given. As it results clear, no published reference article behind the present Patent exists. Some partial results may be found in several papers2 of Volume II of [Fermi (1962)] (see, for example, [Fermi (1952)]), but here very many technical data and some information of historic interest (mainly on the experiments performed in order to obtain the data reported) are given. The most "relevant" change of Patent No. 2,798,847 with respect to the original Patent No. 2,708,656 is the replacement of the 8 claims of the previous one by the following only one claim, which well summarizes the work done: "A method of operating a neutronic reactor including an active portion having a neutron reproduction ratio substantially in excess of unity in the absence of high neutron absorbing bodies, said method comprising the steps of inserting in the active portion a shim member consisting essentially of a high neutron absorbing body in an amount to reduce the neutron reproduction ratio to a value slightly higher than unit to prevent a dangerous reactivity level, controlling the reaction by moving a control member consisting essentially of a second high neutron absorbing body inwardly and outwardly in response to variations in neutron density, to maintain the neutron reproduction ratio substantially at unity, and withdrawing successive portions of the shim member to the extent necessary to enable the reactor to be controlled by movement of the control member after the neutron reproduction value has been lowered to the point where the outward movement of the control member is insufficient to maintain the neutron reproduction ratio at the desired point, and thus to maintain the range of control effected by such movement of the control member substantially constant despite diminution of neutron reproduction ratio caused by operation of the reactor, the active portion being substantially free of high neutron absorber other than the control member and the shim member."

The fluctuating ribosome: thermal molecular dynamics characterized by neutron scattering

NASA Astrophysics Data System (ADS)

Zaccai, Giuseppe; Natali, Francesca; Peters, Judith; Řihová, Martina; Zimmerman, Ella; Ollivier, J.; Combet, J.; Maurel, Marie-Christine; Bashan, Anat; Yonath, Ada

2016-11-01

Conformational changes associated with ribosome function have been identified by X-ray crystallography and cryo-electron microscopy. These methods, however, inform poorly on timescales. Neutron scattering is well adapted for direct measurements of thermal molecular dynamics, the ‘lubricant’ for the conformational fluctuations required for biological activity. The method was applied to compare water dynamics and conformational fluctuations in the 30 S and 50 S ribosomal subunits from Haloarcula marismortui, under high salt, stable conditions. Similar free and hydration water diffusion parameters are found for both subunits. With respect to the 50 S subunit, the 30 S is characterized by a softer force constant and larger mean square displacements (MSD), which would facilitate conformational adjustments required for messenger and transfer RNA binding. It has been shown previously that systems from mesophiles and extremophiles are adapted to have similar MSD under their respective physiological conditions. This suggests that the results presented are not specific to halophiles in high salt but a general property of ribosome dynamics under corresponding, active conditions. The current study opens new perspectives for neutron scattering characterization of component functional molecular dynamics within the ribosome.

[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.

The Prompt Gamma Neutron Activation Analysis Facility at ICN—Pitesti

NASA Astrophysics Data System (ADS)

Bǎrbos, D.; Pǎunoiu, C.; Mladin, M.; Cosma, C.

2008-08-01

PGNAA is a very widely applicable technique for determining the presence and amount of many elements simultaneously in samples ranging in size from micrograms to many grams. PGNAA is characterized by its capability for nondestructive multi-elemental analysis and its ability to analyse elements that cannot be determined by INAA. By means of this PGNAA method we are able to increase the performace of INAA method. A facility has been developed at Institute for Nuclear Research—Piteşti so that the unique features of prompt gamma-ray neutron activation analysis can be used to measure trace and major elements in samples. The facility is linked at the radial neutron beam tube at ACPR-TRIGA reactor. During the PGNAA—facility is in use the ACPR reactor will be operated in steady-state mode at 250 KW maximum power. The facility consists of a radial beam-port, external sample position with shielding, and induced prompt gamma-ray counting system. Thermal neutron flux with energy lower than cadmium cut-off at the sample position was measured using thin gold foil is: φscd = 1.106 n/cm2/s with a cadmium ratio of:80. The gamma-ray detection system consist of an HpGe detector of 16% efficiency (detector model GC1518) with 1.85 keV resolution capability. The HpGe is mounted with its axis at 90° with respect to the incident neutron beam at distance about 200mm from the sample position. To establish the performance capabilities of the facility, irradiation of pure element or sample compound standards were performed to identify the gama-ray energies from each element and their count rates.

Neutron imaging with bubble chambers for inertial confinement fusion

NASA Astrophysics Data System (ADS)

Ghilea, Marian C.

One of the main methods to obtain energy from controlled thermonuclear fusion is inertial confinement fusion (ICF), a process where nuclear fusion reactions are initiated by heating and compressing a fuel target, typically in the form of a pellet that contains deuterium and tritium, relying on the inertia of the fuel mass to provide confinement. In inertial confinement fusion experiments, it is important to distinguish failure mechanisms of the imploding capsule and unambiguously diagnose compression and hot spot formation in the fuel. Neutron imaging provides such a technique and bubble chambers are capable of generating higher resolution images than other types of neutron detectors. This thesis explores the use of a liquid bubble chamber to record high yield 14.1 MeV neutrons resulting from deuterium-tritium fusion reactions on ICF experiments. A design tool to deconvolve and reconstruct penumbral and pinhole neutron images was created, using an original ray tracing concept to simulate the neutron images. The design tool proved that misalignment and aperture fabrication errors can significantly decrease the resolution of the reconstructed neutron image. A theoretical model to describe the mechanism of bubble formation was developed. A bubble chamber for neutron imaging with Freon 115 as active medium was designed and implemented for the OMEGA laser system. High neutron yields resulting from deuterium-tritium capsule implosions were recorded. The bubble density was too low for neutron imaging on OMEGA but agreed with the model of bubble formation. The research done in here shows that bubble detectors are a promising technology for the higher neutron yields expected at National Ignition Facility (NIF).

Neutron imaging integrated circuit and method for detecting neutrons

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

Nagarkar, Vivek V.; More, Mitali J.

The present disclosure provides a neutron imaging detector and a method for detecting neutrons. In one example, a method includes providing a neutron imaging detector including plurality of memory cells and a conversion layer on the memory cells, setting one or more of the memory cells to a first charge state, positioning the neutron imaging detector in a neutron environment for a predetermined time period, and reading a state change at one of the memory cells, and measuring a charge state change at one of the plurality of memory cells from the first charge state to a second charge statemore » less than the first charge state, where the charge state change indicates detection of neutrons at said one of the memory cells.« less

Fission Product Yield Study of 235U, 238U and 239Pu Using Dual-Fission Ionization Chambers

NASA Astrophysics Data System (ADS)

Bhatia, C.; Fallin, B.; Howell, C.; Tornow, W.; Gooden, M.; Kelley, J.; Arnold, C.; Bond, E.; Bredeweg, T.; Fowler, M.; Moody, W.; Rundberg, R.; Rusev, G.; Vieira, D.; Wilhelmy, J.; Becker, J.; Macri, R.; Ryan, C.; Sheets, S.; Stoyer, M.; Tonchev, A.

2014-05-01

To resolve long-standing differences between LANL and LLNL regarding the correct fission basis for analysis of nuclear test data [M.B. Chadwick et al., Nucl. Data Sheets 111, 2891 (2010); H. Selby et al., Nucl. Data Sheets 111, 2891 (2010)], a collaboration between TUNL/LANL/LLNL has been established to perform high-precision measurements of neutron induced fission product yields. The main goal is to make a definitive statement about the energy dependence of the fission yields to an accuracy better than 2-3% between 1 and 15 MeV, where experimental data are very scarce. At TUNL, we have completed the design, fabrication and testing of three dual-fission chambers dedicated to 235U, 238U, and 239Pu. The dual-fission chambers were used to make measurements of the fission product activity relative to the total fission rate, as well as for high-precision absolute fission yield measurements. The activation method was employed, utilizing the mono-energetic neutron beams available at TUNL. Neutrons of 4.6, 9.0, and 14.5 MeV were produced via the 2H(d,n)3He reaction, and for neutrons at 14.8 MeV, the 3H(d,n)4He reaction was used. After activation, the induced γ-ray activity of the fission products was measured for two months using high-resolution HPGe detectors in a low-background environment. Results for the yield of seven fission fragments of 235U, 238U, and 239Pu and a comparison to available data at other energies are reported. For the first time results are available for neutron energies between 2 and 14 MeV.

Detectors for Active Interrogation Applications

NASA Astrophysics Data System (ADS)

Clarke, S. D.; Hamel, M. C.; Bourne, M. M.; Pozzi, S. A.

Active interrogation creates an environment that is particularly challenging from a radiation-detection standpoint: the elevated background levels from the source can mask the desired signatures from the SNM. Neutron based interrogation experiments have shown that nanosecond-level timing is required to discriminate induced-fission neutrons from the scattered source neutrons. Previous experiments using high-energy bremsstrahlung X-rays have demonstrated the ability to induce and detect prompt photofission neutrons from single target materials; however, a real-world application would require spectroscopic capability to discern between photofission neutrons emitted by SNM and neutrons emitted by other reactions in non-SNM. Using digital pulse-shape discrimination, organic liquid scintillators are capable of reliably detecting neutrons in an intense gamma-ray field. Photon misclassification rates as low as 1 in 106 have been achieved, which is approaching the level of gaseous neutron detectors such as 3He without the need for neutron moderation. These scintillators also possess nanosecond-timing resolution, making them candidates for both neutron-and photon-driven active interrogation systems. We have applied an array of liquid and NaI(Tl) scintillators to successfully image 13.7 kg of HEU interrogated by a DT neutron generator; the system was in the direct presence of the accelerator during the experiment.

Applying an analytical method to study neutron behavior for dosimetry

NASA Astrophysics Data System (ADS)

Shirazi, S. A. Mousavi

2016-12-01

In this investigation, a new dosimetry process is studied by applying an analytical method. This novel process is associated with a human liver tissue. The human liver tissue has compositions including water, glycogen and etc. In this study, organic compound materials of liver are decomposed into their constituent elements based upon mass percentage and density of every element. The absorbed doses are computed by analytical method in all constituent elements of liver tissue. This analytical method is introduced applying mathematical equations based on neutron behavior and neutron collision rules. The results show that the absorbed doses are converged for neutron energy below 15MeV. This method can be applied to study the interaction of neutrons in other tissues and estimating the absorbed dose for a wide range of neutron energy.

Effects of geochemical composition on neutron die-away measurements: Implications for Mars Science Laboratory's Dynamic Albedo of Neutrons experiment

NASA Astrophysics Data System (ADS)

Hardgrove, C.; Moersch, J.; Drake, D.

2011-12-01

The Dynamic Albedo of Neutrons (DAN) experiment, part of the scientific payload of the Mars Science Laboratory (MSL) rover mission, will have the ability to assess both the abundance and the burial depth of subsurface hydrogen as the rover traverses the Martian surface. DAN will employ a method of measuring neutron fluxes called “neutron die-away” that has not been used in previous planetary exploration missions. This method requires the use of a pulsed neutron generator that supplements neutrons produced via spallation in the subsurface by the cosmic ray background. It is well established in neutron remote sensing that low-energy (thermal) neutrons are sensitive not only to hydrogen content, but also to the macroscopic absorption cross-section of near-surface materials. To better understand the results that will be forthcoming from DAN, we model the effects of varying abundances of high absorption cross-section elements that are likely to be found on the Martian surface (Cl, Fe) on neutron die-away measurements made from a rover platform. Previously, the Mars Exploration Rovers (MER) Spirit and Opportunity found that elevated abundances of these two elements are commonly associated with locales that have experienced some form of aqueous activity in the past, even though hydrogen-rich materials are not necessarily still present. By modeling a suite of H and Cl compositions, we demonstrate that (for abundance ranges reasonable for Mars) both the elements will significantly affect DAN thermal neutron count rates. Additionally, we show that the timing of thermal neutron arrivals at the detector can be used together with the thermal neutron count rates to independently determine the abundances of hydrogen and high neutron absorption cross-section elements (the most important being Cl). Epithermal neutron die-away curves may also be used to separate these two components. We model neutron scattering in actual Martian compositions that were determined by the MER Alpha Proton X-Ray Spectrometer (APXS), as examples of local geochemical anomalies that DAN would be sensitive to if they were present at the MSL landing site. These MER targets, named “Eileen Dean,” “Jack Russell,” and “Kenosha Comets,” all have unusually high or low Cl or Fe abundances as a result of geochemical interactions involving water. Using these examples we demonstrate that DAN can be used not only to assess the amount of present-day hydrogen in the near-surface but also to identify locations that may preserve a geochemical record of past aqueous processes.

A laser-induced repetitive fast neutron source applied for gold activation analysis

NASA Astrophysics Data System (ADS)

Lee, Sungman; Park, Sangsoon; Lee, Kitae; Cha, Hyungki

2012-12-01

A laser-induced repetitively operated fast neutron source was developed for applications in laser-driven nuclear physics research. The developed neutron source, which has a neutron yield of approximately 4 × 105 n/pulse and can be operated up to a pulse repetition rate of 10 Hz, was applied for a gold activation analysis. Relatively strong delayed gamma spectra of the activated gold were measured at 333 keV and 355 keV, and proved the possibility of the neutron source for activation analyses. In addition, the nuclear reactions responsible for the measured gamma spectra of gold were elucidated by the 14 MeV fast neutrons resulting from the D(t,n)He4 nuclear reaction, for which the required tritium originated from the primary fusion reaction, D(d,p)T3.

A laser-induced repetitive fast neutron source applied for gold activation analysis.

PubMed

Lee, Sungman; Park, Sangsoon; Lee, Kitae; Cha, Hyungki

2012-12-01

A laser-induced repetitively operated fast neutron source was developed for applications in laser-driven nuclear physics research. The developed neutron source, which has a neutron yield of approximately 4 × 10(5) n/pulse and can be operated up to a pulse repetition rate of 10 Hz, was applied for a gold activation analysis. Relatively strong delayed gamma spectra of the activated gold were measured at 333 keV and 355 keV, and proved the possibility of the neutron source for activation analyses. In addition, the nuclear reactions responsible for the measured gamma spectra of gold were elucidated by the 14 MeV fast neutrons resulting from the D(t,n)He(4) nuclear reaction, for which the required tritium originated from the primary fusion reaction, D(d,p)T(3).

Evaluation of equivalent dose from neutrons and activation products from a 15-MV X-ray LINAC.

PubMed

Israngkul-Na-Ayuthaya, Isra; Suriyapee, Sivalee; Pengvanich, Phongpheath

2015-11-01

A high-energy photon beam that is more than 10 MV can produce neutron contamination. Neutrons are generated by the [γ,n] reactions with a high-Z target material. The equivalent neutron dose and gamma dose from activation products have been estimated in a LINAC equipped with a 15-MV photon beam. A Monte Carlo simulation code was employed for neutron and photon dosimetry due to mixed beam. The neutron dose was also experimentally measured using the Optically Stimulated Luminescence (OSL) under various conditions to compare with the simulation. The activation products were measured by gamma spectrometer system. The average neutron energy was calculated to be 0.25 MeV. The equivalent neutron dose at the isocenter obtained from OSL measurement and MC calculation was 5.39 and 3.44 mSv/Gy, respectively. A gamma dose rate of 4.14 µSv/h was observed as a result of activations by neutron inside the treatment machine. The gamma spectrum analysis showed (28)Al, (24)Na, (54)Mn and (60)Co. The results confirm that neutrons and gamma rays are generated, and gamma rays remain inside the treatment room after the termination of X-ray irradiation. The source of neutrons is the product of the [γ,n] reactions in the machine head, whereas gamma rays are produced from the [n,γ] reactions (i.e. neutron activation) with materials inside the treatment room. The most activated nuclide is (28)Al, which has a half life of 2.245 min. In practice, it is recommended that staff should wait for a few minutes (several (28)Al half-lives) before entering the treatment room after the treatment finishes to minimize the dose received. © The Author 2015. Published by Oxford University Press on behalf of The Japan Radiation Research Society and Japanese Society for Radiation Oncology.

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.

Precision determination of absolute neutron flux

DOE PAGES

Yue, A. T.; Anderson, E. S.; Dewey, M. S.; ...

2018-06-08

A technique for establishing the total neutron rate of a highly-collimated monochromatic cold neutron beam was demonstrated using an alpha–gamma counter. The method involves only the counting of measured rates and is independent of neutron cross sections, decay chain branching ratios, and neutron beam energy. For the measurement, a target of 10B-enriched boron carbide totally absorbed the neutrons in a monochromatic beam, and the rate of absorbed neutrons was determined by counting 478 keV gamma rays from neutron capture on 10B with calibrated high-purity germanium detectors. A second measurement based on Bragg diffraction from a perfect silicon crystal was performedmore » to determine the mean de Broglie wavelength of the beam to a precision of 0.024%. With these measurements, the detection efficiency of a neutron monitor based on neutron absorption on 6Li was determined to an overall uncertainty of 0.058%. We discuss the principle of the alpha–gamma method and present details of how the measurement was performed including the systematic effects. We further describe how this method may be used for applications in neutron dosimetry and metrology, fundamental neutron physics, and neutron cross section measurements.« less

Precision determination of absolute neutron flux

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

Yue, A. T.; Anderson, E. S.; Dewey, M. S.

A technique for establishing the total neutron rate of a highly-collimated monochromatic cold neutron beam was demonstrated using an alpha–gamma counter. The method involves only the counting of measured rates and is independent of neutron cross sections, decay chain branching ratios, and neutron beam energy. For the measurement, a target of 10B-enriched boron carbide totally absorbed the neutrons in a monochromatic beam, and the rate of absorbed neutrons was determined by counting 478 keV gamma rays from neutron capture on 10B with calibrated high-purity germanium detectors. A second measurement based on Bragg diffraction from a perfect silicon crystal was performedmore » to determine the mean de Broglie wavelength of the beam to a precision of 0.024%. With these measurements, the detection efficiency of a neutron monitor based on neutron absorption on 6Li was determined to an overall uncertainty of 0.058%. We discuss the principle of the alpha–gamma method and present details of how the measurement was performed including the systematic effects. We further describe how this method may be used for applications in neutron dosimetry and metrology, fundamental neutron physics, and neutron cross section measurements.« less

ATRC Neutron Detector Testing Quick Look Report

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

Troy C. Unruh; Benjamin M. Chase; Joy L. Rempe

2013-08-01

As part of the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program, a joint Idaho State University (ISU) / French Alternative Energies and Atomic Energy Commission (CEA) / Idaho National Laboratory (INL) project was initiated in FY-10 to investigate the feasibility of using neutron sensors to provide online measurements of the neutron flux and fission reaction rate in the ATR Critical Facility (ATRC). A second objective was to provide initial neutron spectrum and flux distribution information for physics modeling and code validation using neutron activation based techniques in ATRC as well as ATR during depressurized operations. Detailed activationmore » spectrometry measurements were made in the flux traps and in selected fuel elements, along with standard fission rate distribution measurements at selected core locations. These measurements provide additional calibration data for the real-time sensors of interest as well as provide benchmark neutronics data that will be useful for the ATR Life Extension Program (LEP) Computational Methods and V&V Upgrade project. As part of this effort, techniques developed by Prof. George Imel will be applied by Idaho State University (ISU) for assessing the performance of various flux detectors to develop detailed procedures for initial and follow-on calibrations of these sensors. In addition to comparing data obtained from each type of detector, calculations will be performed to assess the performance of and reduce uncertainties in flux detection sensors and compare data obtained from these sensors with existing integral methods employed at the ATRC. The neutron detectors required for this project were provided to team participants at no cost. Activation detectors (foils and wires) from an existing, well-characterized INL inventory were employed. Furthermore, as part of an on-going ATR NSUF international cooperation, the CEA sent INL three miniature fission chambers (one for detecting fast flux and two for detecting thermal flux) with associated electronics for assessment. In addition, Prof. Imel, ISU, has access to an inventory of Self-Powered Neutron Detectors (SPNDs) with a range of response times as well as Back-to-Back (BTB) fission chambers from prior research he conducted at the Transient REActor Test Facility (TREAT) facility and Neutron RADiography (NRAD) reactors. Finally, SPNDs from the National Atomic Energy Commission of Argentina (CNEA) were provided in connection with the INL effort to upgrade ATR computational methods and V&V protocols that are underway as part of the ATR LEP. Work during fiscal year 2010 (FY10) focussed on design and construction of Experiment Guide Tubes (EGTs) for positioning the flux detectors in the ATRC N-16 locations as well as obtaining ATRC staff concurrence for the detector evaluations. Initial evaluations with CEA researchers were also started in FY10 but were cut short due to reactor reliability issues. Reactor availability issues caused experimental work to be delayed during FY11/12. In FY13, work resumed; and evaluations were completed. The objective of this "Quick Look" report is to summarize experimental activities performed from April 4, 2013 through May 16, 2013.« less

Sample registration software for process automation in the Neutron Activation Analysis (NAA) Facility in Malaysia nuclear agency

NASA Astrophysics Data System (ADS)

Rahman, Nur Aira Abd; Yussup, Nolida; Salim, Nazaratul Ashifa Bt. Abdullah; Ibrahim, Maslina Bt. Mohd; Mokhtar, Mukhlis B.; Soh@Shaari, Syirrazie Bin Che; Azman, Azraf B.; Ismail, Nadiah Binti

2015-04-01

Neutron Activation Analysis (NAA) had been established in Nuclear Malaysia since 1980s. Most of the procedures established were done manually including sample registration. The samples were recorded manually in a logbook and given ID number. Then all samples, standards, SRM and blank were recorded on the irradiation vial and several forms prior to irradiation. These manual procedures carried out by the NAA laboratory personnel were time consuming and not efficient. Sample registration software is developed as part of IAEA/CRP project on `Development of Process Automation in the Neutron Activation Analysis (NAA) Facility in Malaysia Nuclear Agency (RC17399)'. The objective of the project is to create a pc-based data entry software during sample preparation stage. This is an effective method to replace redundant manual data entries that needs to be completed by laboratory personnel. The software developed will automatically generate sample code for each sample in one batch, create printable registration forms for administration purpose, and store selected parameters that will be passed to sample analysis program. The software is developed by using National Instruments Labview 8.6.

A compact in vivo neutron activation analysis system to quantify manganese in human hand bone

NASA Astrophysics Data System (ADS)

Liu, Yingzi

As an urgent issue of correlating cumulative manganese (Mn) exposure to neurotoxicity, bone has emerged as an attractive biomarker for long-term Mn deposition and storage. A novel Deuterium-Deuterium (DD) neutron generator irradiation system has been simulated and constructed, incorporating moderator, reflector and shielding. This neutron activation analysis (NAA) irradiation assembly presents several desirable features, including high neutron flux, improved detection limit and acceptable neutron & photon dose, which would allow it be ready for clinical measurement. Key steps include simulation modeling and verifying, irradiation system design, detector characterization, and neutron flux and dose assessment. Activation foils were also analyzed to reveal the accurate neutron spectrum in the irradiation cave. The detection limit with this system is 0.428 ppm with 36 mSv equivalent hand dose and 52 microSv whole body effective dose.

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

PubMed

Kurosawa, Masahiko

2005-01-01

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.

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.

Neutron Transport Simulations for NIST Neutron Lifetime Experiment

NASA Astrophysics Data System (ADS)

Li, Fangchen; BL2 Collaboration Collaboration

2016-09-01

Neutrons in stable nuclei can exist forever; a free neutron lasts for about 15 minutes on average before it beta decays to a proton, an electron, and an antineutrino. Precision measurements of the neutron lifetime test the validity of weak interaction theory and provide input into the theory of the evolution of light elements in the early universe. There are two predominant ways of measuring the neutron lifetime: the bottle method and the beam method. The bottle method measures decays of ultracold neutrons that are stored in a bottle. The beam method measures decay protons in a beam of cold neutrons of known flux. An improved beam experiment is being prepared at the National Institute of Science and Technology (Gaithersburg, MD) with the goal of reducing statistical and systematic uncertainties to the level of 1 s. The purpose of my studies was to develop computer simulations of neutron transport to determine the beam collimation and study the neutron distribution's effect on systematic effects for the experiment, such as the solid angle of the neutron flux monitor. The motivation for the experiment and the results of this work will be presented. This work was supported, in part, by a Grant to Gettysburg College from the Howard Hughes Medical Institute through the Precollege and Undergraduate Science Education Program.

Method for measuring dose-equivalent in a neutron flux with an unknown energy spectra and means for carrying out that method

DOEpatents

Distenfeld, Carl H.

1978-01-01

A method for measuring the dose-equivalent for exposure to an unknown and/or time varing neutron flux which comprises simultaneously exposing a plurality of neutron detecting elements of different types to a neutron flux and combining the measured responses of the various detecting elements by means of a function, whose value is an approximate measure of the dose-equivalent, which is substantially independent of the energy spectra of the flux. Also, a personnel neutron dosimeter, which is useful in carrying out the above method, comprising a plurality of various neutron detecting elements in a single housing suitable for personnel to wear while working in a radiation area.

Application of instrumental neutron activation and X-ray fluorescence analysis to the examination of objects of art

NASA Astrophysics Data System (ADS)

Panczyk, E.; Ligeza, M.; Walis, L.

1999-01-01

In the Institute of Nuclear Chemistry and Technology in Warsaw in collaboration with the Department of Preservation and Restoration of Works of Art of the Academy of Fine Arts in Cracow and National Museum in Warsaw systematic studies using nuclear methods, particulary instrumental neutron activation analysis and X-ray fluorescence analysis, have been carried out on the panel paintings from the Krakowska- Nowosadecka School and Silesian School of the period from the XIV-XVII century, Chinese and Thai porcelains and mummies fillings of Egyptian sarcophagi. These studies will provide new data to the existing data base, will permit to compare materials used by various schools and individual artists.

Container Inspection Utilizing 14 MeV Neutrons

NASA Astrophysics Data System (ADS)

Valkovic, Vladivoj; Sudac, Davorin; Nad, Karlo; Obhodas, Jasmina

2016-06-01

A proposal for an autonomous and flexible ship container inspection system is presented. This could be accomplished by the incorporation of an inspection system on various container transportation devices (straddle carriers, yard gentry cranes, automated guided vehicles, trailers). The configuration is terminal specific and it should be defined by the container terminal operator. This enables that no part of the port operational area is used for inspection. The inspection scenario includes container transfer from ship to transportation device with the inspection unit mounted on it. The inspection is performed during actual container movement to the container location. A neutron generator without associated alpha particle detection is used. This allows the use of higher neutron intensities (5 × 109 - 1010 n/s in 4π). The inspected container is stationary in the “inspection position” on the transportation device while the “inspection unit” moves along its side. The following analytical methods will be used simultaneously: neutron radiography, X-ray radiography, neutron activation analysis, (n, γ) and (n,n'γ) reactions, neutron absorption. and scattering, X-ray backscattering. The neutron techniques will utilize “smart collimators” for neutrons and gamma rays, both emitted and detected. The inspected voxel is defined by the intersection of the neutron generator and the detectors solid angles. The container inspection protocol is based on identification of discrepancies between the cargo manifest, elemental “fingerprint” and radiography profiles. In addition, the information on container weight is obtained during the container transport and screening by measuring of density of material in the container.

An iterative method for the localization of a neutron source in a large box (container)

NASA Astrophysics Data System (ADS)

Dubinski, S.; Presler, O.; Alfassi, Z. B.

2007-12-01

The localization of an unknown neutron source in a bulky box was studied. This can be used for the inspection of cargo, to prevent the smuggling of neutron and α emitters. It is important to localize the source from the outside for safety reasons. Source localization is necessary in order to determine its activity. A previous study showed that, by using six detectors, three on each parallel face of the box (460×420×200 mm 3), the location of the source can be found with an average distance of 4.73 cm between the real source position and the calculated one and a maximal distance of about 9 cm. Accuracy was improved in this work by applying an iteration method based on four fixed detectors and the successive iteration of positioning of an external calibrating source. The initial positioning of the calibrating source is the plane of detectors 1 and 2. This method finds the unknown source location with an average distance of 0.78 cm between the real source position and the calculated one and a maximum distance of 3.66 cm for the same box. For larger boxes, localization without iterations requires an increase in the number of detectors, while localization with iterations requires only an increase in the number of iteration steps. In addition to source localization, two methods for determining the activity of the unknown source were also studied.

Safety control circuit for a neutronic reactor

DOEpatents

Ellsworth, Howard C.

2004-04-27

A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, a safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper.

Experimental research of neutron yield and spectrum from deuterium gas-puff z-pinch on the GIT-12 generator at current above 2 MA

NASA Astrophysics Data System (ADS)

Cherdizov, R. K.; Fursov, F. I.; Kokshenev, V. A.; Kurmaev, N. E.; Labetsky, A. Yu; Ratakhin, N. A.; Shishlov, A. V.; Cikhardt, J.; Cikhardtova, B.; Klir, D.; Kravarik, J.; Kubes, P.; Rezac, K.; Dudkin, G. N.; Garapatsky, A. A.; Padalko, V. N.; Varlachev, V. A.

2017-05-01

The Z-pinch experiments with deuterium gas-puff surrounded by an outer plasma shell were carried out on the GIT-12 generator (Tomsk, Russia) at currents of 2 MA. The plasma shell consisting of hydrogen and carbon ions was formed by 48 plasma guns. The deuterium gas-puff was created by a fast electromagnetic valve. This configuration provides an efficient mode of the neutron production in DD reaction, and the neutron yield reaches a value above 1012 neutrons per shot. Neutron diagnostics included scintillation TOF detectors for determination of the neutron energy spectrum, bubble detectors BD-PND, a silver activation detector, and several activation samples for determination of the neutron yield analysed by a Sodium Iodide (NaI) and a high-purity Germanium (HPGe) detectors. Using this neutron diagnostic complex, we measured the total neutron yield and amount of high-energy neutrons.

The intensive DT neutron generator of TU Dresden

NASA Astrophysics Data System (ADS)

Klix, Axel; DÖring, Toralf; Domula, Alexander; Zuber, Kai

2018-01-01

TU Dresden operates an accelerator-based intensive DT neutron generator. Experimental activities comprise investigation into material activation and decay, neutron and photon transport in matter and R&D work on radiation detectors for harsh environments. The intense DT neutron generator is capable to produce a maximum of 1012 n/s. The neutron source is a solid-type water-cooled tritium target based on a titanium matrix on a copper carrier. The neutron yield at a typical deuteron beam current of 1 mA is of the order of 1011 n/s in 4Π. A pneumatic sample transport system is available for short-time irradiations and connected to wo high-purity germanium detector spectrometers for the measurement of induced activities. The overall design of the experimental hall with the neutron generator allows a flexible setup of experiments including the possibility of investigating larger structures and cooled samples or samples at high temperatures.

Detectors for Active Interrogation Applications

DOE PAGES

Clarke, S. D.; Hamel, M. C.; Bourne, M. M.; ...

2017-10-26

Active interrogation creates an environment that is particularly challenging from a radiation-detection standpoint: the elevated background levels from the source can mask the desired signatures from the SNM. Neutron based interrogation experiments have shown that nanosecond-level timing is required to discriminate induced-fission neutrons from the scattered source neutrons. Previous experiments using high-energy bremsstrahlung X-rays have demonstrated the ability to induce and detect prompt photofission neutrons from single target materials; however, a real-world application would require spectroscopic capability to discern between photofission neutrons emitted by SNM and neutrons emitted by other reactions in non-SNM. Using digital pulseshape discrimination, organic liquid scintillatorsmore » are capable of reliably detecting neutrons in an intense gamma-ray field. Photon misclassification rates as low as 1 in 10 6 have been achieved, which is approaching the level of gaseous neutron detectors such as 3He without the need for neutron moderation. These scintillators also possess nanosecond-timing resolution, making them candidates for both neutron-and photon-driven active interrogation systems. Lastly, we have applied an array of liquid and NaI(Tl) scintillators to successfully image 13.7 kg of HEU interrogated by a DT neutron generator; the system was in the direct presence of the accelerator during the experiment.« less

Method of preparing high specific activity platinum-195m

DOEpatents

Mirzadeh, Saed; Du, Miting; Beets, Arnold L.; Knapp, Jr., Furn F.

2004-06-15

A method of preparing high-specific-activity .sup.195m Pt includes the steps of: exposing .sup.193 Ir to a flux of neutrons sufficient to convert a portion of the .sup.193 Ir to .sup.195m Pt to form an irradiated material; dissolving the irradiated material to form an intermediate solution comprising Ir and Pt; and separating the Pt from the Ir by cation exchange chromatography to produce .sup.195m Pt.

Neutron Activation Analysis of Water - A Review

NASA Technical Reports Server (NTRS)

Buchanan, John D.

1971-01-01

Recent developments in this field are emphasized. After a brief review of basic principles, topics discussed include sources of neutrons, pre-irradiation physical and chemical treatment of samples, neutron capture and gamma-ray analysis, and selected applications. Applications of neutron activation analysis of water have increased rapidly within the last few years and may be expected to increase in the future.

High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

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

Guillen, Donna; Greenwood, Lawrence R.; Parry, James

2014-06-22

A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated formore » up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.« less

Neutron absorbers and methods of forming at least a portion of a neutron absorber

DOEpatents

Guillen, Donna P; Porter, Douglas L; Swank, W David; Erickson, Arnold W

2014-12-02

Methods of forming at least a portion of a neutron absorber include combining a first material and a second material to form a compound, reducing the compound into a plurality of particles, mixing the plurality of particles with a third material, and pressing the mixture of the plurality of particles and the third material. One or more components of neutron absorbers may be formed by such methods. Neutron absorbers may include a composite material including an intermetallic compound comprising hafnium aluminide and a matrix material comprising pure aluminum.

Nondispersive neutron focusing method beyond the critical angle of mirrors

DOEpatents

Ice, Gene E.

2008-10-21

This invention extends the Kirkpatrick-Baez (KB) mirror focusing geometry to allow nondispersive focusing of neutrons with a convergence on a sample much larger than is possible with existing KB optical schemes by establishing an array of at least three mirrors and focusing neutrons by appropriate multiple deflections via the array. The method may be utilized with supermirrors, multilayer mirrors, or total external reflection mirrors. Because high-energy x-rays behave like neutrons in their absorption and reflectivity rates, this method may be used with x-rays as well as neutrons.

Experimental studies of the medical radioisotopes production in neutron spectra generated by 660 MeV protons and 1-8 GeV deuterons in massive uranium target

NASA Astrophysics Data System (ADS)

Zhadan, A.; Sotnikov, V.; Adam, J.; Solnyshkin, A.; Tyutyunnikov, S.; Voronko, V.; Zhivkov, P.; Zavorka, L.

2017-06-01

The possibility of medical radionuclide 64,67Cu production in spallation neutron spectrum induced by proton and deuteron beams has been studied. Experiments were performed on a massive natural uranium target at the accelerators Phasotron and Nuclotron JINR, Dubna. The main disadvantage of this method is a high 64Cu/67Cu ratio in the final product at EOB. Significantly reduce 64Cu/67Cu ratio is only possible if you use zinc target enriched with 68Zn or 67Zn. The MCNPX simulation of 67,64Cu production and definition of the theoretical limit of the specific activity of 67,64Cu by irradiation of natural zinc and zinc enriched by the 68 isotope were performed. The neutron flux density shouldnot be less than 5.1013 n/cm2/s if we want to obtain high specific activity (>200 GBq/mg) of 67Cu.

The feasibility of well-logging measurements of arsenic levels using neutron-activation analysis

USGS Publications Warehouse

Oden, C.P.; Schweitzer, J.S.; McDowell, G.M.

2006-01-01

Arsenic is an extremely toxic metal, which poses a significant problem in many mining environments. Arsenic contamination is also a major problem in ground and surface waters. A feasibility study was conducted to determine if neutron-activation analysis is a practical method of measuring in situ arsenic levels. The response of hypothetical well-logging tools to arsenic was simulated using a readily available Monte Carlo simulation code (MCNP). Simulations were made for probes with both hyperpure germanium (HPGe) and bismuth germanate (BGO) detectors using accelerator and isotopic neutron sources. Both sources produce similar results; however, the BGO detector is much more susceptible to spectral interference than the HPGe detector. Spectral interference from copper can preclude low-level arsenic measurements when using the BGO detector. Results show that a borehole probe could be built that would measure arsenic concentrations of 100 ppm by weight to an uncertainty of 50 ppm in about 15 min. ?? 2006 Elsevier Ltd. All rights reserved.

Isotope effect in heavy/light water suspensions of optically active gold nanoparticles

NASA Astrophysics Data System (ADS)

Kutsenko, V. Y.; Artykulnyi, O. P.; Petrenko, V. I.; Avdeev, M. V.; Marchenko, O. A.; Bulavin, L. A.; Snegir, S. V.

2018-04-01

Aqueous suspensions of optically active gold nanoparticles coated with trisodium citrate were synthesized in light (H2O) water and mixture of light and heavy (H2O/D2O) water using the modified Turkevich protocol. The objective of the paper was to verify sensitivity of neutron scattering methods (in particular, neutron reflectometry) to the potential isotope H/D substitution in the stabilizing organic shell around particles in colloidal solutions. First, the isotope effect was studied with respect to the changes in the structural properties of metal particles (size, shape, crystalline morphology) in solutions by electron microscopy including high-resolution transmission electron microscopy from dried systems. The structural factors determining the variation in the adsorption spectra in addition to the change in the optical properties of surrounding medium were discussed. Then, neutron reflectometry was applied to the layered nanoparticles anchored on a silicon wafer via 3-aminopropyltriethoxysilane molecules to reveal the presence of deuterated water molecules in the shell presumably formed by citrate molecules around the metallic core.

IEC-Based Neutron Generator for Security Inspection System

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

Wu, Linchun; Miley, George H.

2002-07-01

Large nuclear reactors are widely employed for electricity power generation, but small nuclear radiation sources can also be used for a variety of industrial/government applications. In this paper we will discuss the use of a small neutron source based on Inertial Electrostatic Confinement (IEC) of accelerated deuterium ions. There is an urgent need of highly effective detection systems for explosives, especially in airports. While current airport inspection systems are strongly based on X-ray technique, neutron activation including Thermal Neutron Analysis (TNA) and Fast Neutron Analysis (FNA) is powerful in detecting certain types of explosives in luggage and in cargoes. Basicmore » elements present in the explosives can be measured through the (n, n'?) reaction initiated by fast neutrons. Combined with a time-of-flight technique, a complete imaging of key elements, hence of the explosive materials, is obtained. Among the various neutron source generators, the IEC is an ideal candidate to meet the neutron activation analysis requirements. Compared with other accelerators and radioisotopes such as {sup 252}Cf, the IEC is simpler, can be switched on or off, and can reliably produce neutrons with minimum maintenance. Theoretical and experimental studies of a spherical IEC have been conducted at the University of Illinois. In a spherical IEC device, 2.54-MeV neutrons of {approx}10{sup 8} n/s via DD reactions over recent years or 14-MeV neutrons of {approx}2x10{sup 10} n/s via DT reactions can be obtained using an ion gun injection technique. The possibility of the cylindrical IEC in pulsed operation mode combining with pulsed FNA method would also be discussed. In this paper we examine the possibility of using an alternative cylindrical IEC configuration. Such a device was studied earlier at the University of Illinois and it provides a very convenient geometry for security inspection. However, to calculate the neutron yield precisely with this configuration, an understanding of the potential wall trapping and acceleration of ions is needed. The theory engaged is an extension of original analytic study by R.L. Hirsh on the potential well structure in a spherical IEC device, i.e. roughly a 'line' source of neutrons from a cylindrical IEC is a 'point' source from the spherical geometry. Thus our present study focuses on the cylindrical IEC for its convenient application in an FNA detecting system. The conceptual design and physics of ion trapping and re-circulation in a cylindrical IEC intended for neutron-based inspection system will be presented. (authors)« less

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.

Activation Inventories after Exposure to DD/DT Neutrons in Safety Analysis of Nuclear Fusion Installations.

PubMed

Stankunas, Gediminas; Cufar, Aljaz; Tidikas, Andrius; Batistoni, Paola

2017-11-23

Irradiations with 14 MeV fusion neutrons are planned at Joint European Torus (JET) in DT operations with the objective to validate the calculation of the activation of structural materials in functional materials expected in ITER and fusion plants. This study describes the activation and dose rate calculations performed for materials irradiated throughout the DT plasma operation during which the samples of real fusion materials are exposed to 14 MeV neutrons inside the JET vacuum vessel. Preparatory activities are in progress during the current DD operations with dosimetry foils to measure the local neutron fluence and spectrum at the sample irradiation position. The materials included those used in the manufacturing of the main in-vessel components, such as ITER-grade W, Be, CuCrZr, 316 L(N) and the functional materials used in diagnostics and heating systems. The neutron-induced activities and dose rates at shutdown were calculated by the FISPACT code, using the neutron fluxes and spectra that were provided by the preceding MCNP neutron transport calculations. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Enhanced NIF neutron activation diagnostics.

PubMed

Yeamans, C B; Bleuel, D L; Bernstein, L A

2012-10-01

The NIF neutron activation diagnostic suite relies on removable activation samples, leading to operational inefficiencies and a fundamental lower limit on the half-life of the activated product that can be observed. A neutron diagnostic system measuring activation of permanently installed samples could remove these limitations and significantly enhance overall neutron diagnostic capabilities. The physics and engineering aspects of two proposed systems are considered: one measuring the (89)Zr/(89 m)Zr isomer ratio in the existing Zr activation medium and the other using potassium zirconate as the activation medium. Both proposed systems could improve the signal-to-noise ratio of the current system by at least a factor of 5 and would allow independent measurement of fusion core velocity and fuel areal density.

Conversion from film to image plates for transfer method neutron radiography of nuclear fuel

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

Craft, Aaron E.; Papaioannou, Glen C.; Chichester, David L.

This paper summarizes efforts to characterize and qualify a computed radiography (CR) system for neutron radiography of irradiated nuclear fuel at Idaho National Laboratory (INL). INL has multiple programs that are actively developing, testing, and evaluating new nuclear fuels. Irradiated fuel experiments are subjected to a number of sequential post-irradiation examination techniques that provide insight into the overall behavior and performance of the fuel. One of the first and most important of these exams is neutron radiography, which provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Results from neutronmore » radiography are often the driver for subsequent examinations of the PIE program. Features of interest that can be evaluated using neutron radiography include irradiation-induced swelling, isotopic and fuel-fragment redistribution, plate deformations, and fuel fracturing. The NRAD currently uses the foil-film transfer technique with film for imaging fuel. INL is pursuing multiple efforts to advance its neutron imaging capabilities for evaluating irradiated fuel and other applications, including conversion from film to CR image plates. Neutron CR is the current state-of-the-art for neutron imaging of highly-radioactive objects. Initial neutron radiographs of various types of nuclear fuel indicate that radiographs can be obtained of comparable image quality currently obtained using film. This paper provides neutron radiographs of representative irradiated fuel pins along with neutron radiographs of standards that informed the qualification of the neutron CR system for routine use. Additionally, this paper includes evaluations of some of the CR scanner parameters and their effects on image quality.« less

CORRELATIONS OF EFFECTIVE TEMPERATURES FOR NEUTRON SPECTRA EMITTED IN U$sup 235$ AND Pu$sup 239$ FISSION BY FAST AND SLOW NEUTRONS (in Russian)

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

Smirenkin, G.M.

1959-12-01

The method of threshold indicators was used for determining the magnitude of dT/sub ef//dE/sub n/. The fission neutrons were produced by fast and slow neutron bombardment. tron bombardment was accomplished in a paraffin block. Hollow metallic U/sup 235/ (90% enriched) and Pu/sup 239/ specimens with 50 mm outside diameters and 10 mm thick contained the threshold activator Ag/sup 107/(n,2n)Ag/sup 106m/. Measurements were made of the neutron flux above the threshold (9.5 Mev) of the Ag/sup 107/(n,2n)Ag/sup 106m/ reaction and the number of fission in the sphere (N/sub gamma /). The final data showed that for U/sup 235/ dT/sub ef//dE/sub n/more » = 0.008 plus or minus 0.004 and for Pu/sup 239/ dT/ sub ef/ /dE/sub n/ = tron fission distributions explain part of the error of the measurement. (R.V.J.)« less

Neutron flux measurements around PLT

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

Zankl, G.; Strachan, J.D.; Lewis, R.

1980-09-01

Using Indium activation foils, the toroidal and poloidal neutron emission patterns were determined for PLT plasmas which include ICRF and neutral beam heating. The activities produced the /sup 115/In (n,n') /sup 115m/In reaction were determined by counting the 336 keV ..gamma.. line of the /sup 115m/In decay. This activation cross section falls just below 2.5 MeV so that the influence of scattered neutrons of degraded energies is reduced. From the magnitude of the activity, the absolute calibration of the PLT fusion neutron emission is obtained with less than or equal to 40% accuracy.

Application of Physics and Chemistry to Archeology: A New Undergraduate Course

ERIC Educational Resources Information Center

Meschel, Susan V.

1976-01-01

Describes a course that covers such topics as the archeological dating processes and methods that enable the identification and authentication of artifacts, including X-ray diffraction, optical emission spectroscopy, infrared spectroscopy, and neutron activation analysis. (MLH)

Direct Discrete Method for Neutronic Calculations

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

Vosoughi, Naser; Akbar Salehi, Ali; Shahriari, Majid

The objective of this paper is to introduce a new direct method for neutronic calculations. This method which is named Direct Discrete Method, is simpler than the neutron Transport equation and also more compatible with physical meaning of problems. This method is based on physic of problem and with meshing of the desired geometry, writing the balance equation for each mesh intervals and with notice to the conjunction between these mesh intervals, produce the final discrete equations series without production of neutron transport differential equation and mandatory passing from differential equation bridge. We have produced neutron discrete equations for amore » cylindrical shape with two boundary conditions in one group energy. The correction of the results from this method are tested with MCNP-4B code execution. (authors)« less

Calculation of the neutron diffusion equation by using Homotopy Perturbation Method

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

Koklu, H., E-mail: koklu@gantep.edu.tr; Ozer, O.; Ersoy, A.

The distribution of the neutrons in a nuclear fuel element in the nuclear reactor core can be calculated by the neutron diffusion theory. It is the basic and the simplest approximation for the neutron flux function in the reactor core. In this study, the neutron flux function is obtained by the Homotopy Perturbation Method (HPM) that is a new and convenient method in recent years. One-group time-independent neutron diffusion equation is examined for the most solved geometrical reactor core of spherical, cubic and cylindrical shapes, in the frame of the HPM. It is observed that the HPM produces excellent resultsmore » consistent with the existing literature.« less

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.

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.

Apparatus for and method of monitoring for breached fuel elements

DOEpatents

Gross, K.C.; Strain, R.V.

1981-04-28

This invention teaches improved apparatus for the method of detecting a breach in cladded fuel used in a nuclear reactor. The detector apparatus uses a separate bypass loop for conveying part of the reactor coolant away from the core, and at least three separate delayed-neutron detectors mounted proximate this detector loop. The detectors are spaced apart so that the coolant flow time from the core to each detector is different, and these differences are known. The delayed-neutron activity at the detectors is a function of the delay time after the reaction in the fuel until the coolant carrying the delayed-neutron emitter passes the respective detector. This time delay is broken down into separate components including an isotopic holdup time required for the emitter to move through the fuel from the reaction to the coolant at the breach, and two transit times required for the emitter now in the coolant to flow from the breach to the detector loop and then via the loop to the detector.

Source Correlated Prompt Neutron Activation Analysis for Material Identification and Localization

NASA Astrophysics Data System (ADS)

Canion, Bonnie; McConchie, Seth; Landsberger, Sheldon

2017-07-01

This paper investigates the energy spectrum of photon signatures from an associated particle imaging deuterium tritium (API-DT) neutron generator interrogating shielded uranium. The goal is to investigate if signatures within the energy spectrum could be used to indirectly characterize shielded uranium when the neutron signature is attenuated. By utilizing the correlated neutron cone associated with each pixel of the API-DT neutron generator, certain materials can be identified and located via source correlated spectrometry of prompt neutron activation gamma rays. An investigation is done to determine if fission neutrons induce a significant enough signature within the prompt neutron-induced gamma-ray energy spectrum in shielding material to be useful for indirect nuclear material characterization. The signature deriving from the induced fission neutrons interacting with the shielding material was slightly elevated in polyethylene-shielding depleted uranium (DU), but was more evident in some characteristic peaks from the aluminum shielding surrounding DU.

A method for in situ absolute DD yield calibration of neutron time-of-flight detectors on OMEGA using CR-39-based proton detectors.

PubMed

Waugh, C J; Rosenberg, M J; Zylstra, A B; Frenje, J A; Séguin, F H; Petrasso, R D; Glebov, V Yu; Sangster, T C; Stoeckl, C

2015-05-01

Neutron time of flight (nTOF) detectors are used routinely to measure the absolute DD neutron yield at OMEGA. To check the DD yield calibration of these detectors, originally calibrated using indium activation systems, which in turn were cross-calibrated to NOVA nTOF detectors in the early 1990s, a direct in situ calibration method using CR-39 range filter proton detectors has been successfully developed. By measuring DD neutron and proton yields from a series of exploding pusher implosions at OMEGA, a yield calibration coefficient of 1.09 ± 0.02 (relative to the previous coefficient) was determined for the 3m nTOF detector. In addition, comparison of these and other shots indicates that significant reduction in charged particle flux anisotropies is achieved when bang time occurs significantly (on the order of 500 ps) after the trailing edge of the laser pulse. This is an important observation as the main source of the yield calibration error is due to particle anisotropies caused by field effects. The results indicate that the CR-39-nTOF in situ calibration method can serve as a valuable technique for calibrating and reducing the uncertainty in the DD absolute yield calibration of nTOF detector systems on OMEGA, the National Ignition Facility, and laser megajoule.

A method for in situ absolute DD yield calibration of neutron time-of-flight detectors on OMEGA using CR-39-based proton detectors

DOE PAGES

Waugh, C. J.; Rosenberg, M. J.; Zylstra, A. B.; ...

2015-05-27

Neutron time of flight (nTOF) detectors are used routinely to measure the absolute DD neutron yield at OMEGA. To check the DD yield calibration of these detectors, originally calibrated using indium activation systems, which in turn were cross-calibrated to NOVA nTOF detectors in the early 1990s, a direct in situ calibration method using CR-39 range filter proton detectors has been successfully developed. By measuring DD neutron and proton yields from a series of exploding pusher implosions at OMEGA, a yield calibration coefficient of 1.09 ± 0.02 (relative to the previous coefficient) was determined for the 3m nTOF detector. In addition,more » comparison of these and other shots indicates that significant reduction in charged particle flux anisotropies is achieved when bang time occurs significantly (on the order of 500 ps) after the trailing edge of the laser pulse. This is an important observation as the main source of the yield calibration error is due to particle anisotropies caused by field effects. The results indicate that the CR-39-nTOF in situ calibration method can serve as a valuable technique for calibrating and reducing the uncertainty in the DD absolute yield calibration of nTOF detector systems on OMEGA, the National Ignition Facility, and laser megajoule.« less

A method for in situ absolute DD yield calibration of neutron time-of-flight detectors on OMEGA using CR-39-based proton detectors

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

Waugh, C. J.; Rosenberg, M. J.; Zylstra, A. B.

Neutron time of flight (nTOF) detectors are used routinely to measure the absolute DD neutron yield at OMEGA. To check the DD yield calibration of these detectors, originally calibrated using indium activation systems, which in turn were cross-calibrated to NOVA nTOF detectors in the early 1990s, a direct in situ calibration method using CR-39 range filter proton detectors has been successfully developed. By measuring DD neutron and proton yields from a series of exploding pusher implosions at OMEGA, a yield calibration coefficient of 1.09 ± 0.02 (relative to the previous coefficient) was determined for the 3m nTOF detector. In addition,more » comparison of these and other shots indicates that significant reduction in charged particle flux anisotropies is achieved when bang time occurs significantly (on the order of 500 ps) after the trailing edge of the laser pulse. This is an important observation as the main source of the yield calibration error is due to particle anisotropies caused by field effects. The results indicate that the CR-39-nTOF in situ calibration method can serve as a valuable technique for calibrating and reducing the uncertainty in the DD absolute yield calibration of nTOF detector systems on OMEGA, the National Ignition Facility, and laser megajoule.« less

A method for in situ absolute DD yield calibration of neutron time-of-flight detectors on OMEGA using CR-39-based proton detectors

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

Waugh, C. J., E-mail: cjwaugh@mit.edu; Zylstra, A. B.; Frenje, J. A.

2015-05-15

Neutron time of flight (nTOF) detectors are used routinely to measure the absolute DD neutron yield at OMEGA. To check the DD yield calibration of these detectors, originally calibrated using indium activation systems, which in turn were cross-calibrated to NOVA nTOF detectors in the early 1990s, a direct in situ calibration method using CR-39 range filter proton detectors has been successfully developed. By measuring DD neutron and proton yields from a series of exploding pusher implosions at OMEGA, a yield calibration coefficient of 1.09 ± 0.02 (relative to the previous coefficient) was determined for the 3m nTOF detector. In addition,more » comparison of these and other shots indicates that significant reduction in charged particle flux anisotropies is achieved when bang time occurs significantly (on the order of 500 ps) after the trailing edge of the laser pulse. This is an important observation as the main source of the yield calibration error is due to particle anisotropies caused by field effects. The results indicate that the CR-39-nTOF in situ calibration method can serve as a valuable technique for calibrating and reducing the uncertainty in the DD absolute yield calibration of nTOF detector systems on OMEGA, the National Ignition Facility, and laser megajoule.« 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.

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

Neutron activation analysis system

DOEpatents

Taylor, M.C.; Rhodes, J.R.

1973-12-25

A neutron activation analysis system for monitoring a generally fluid media, such as slurries, solutions, and fluidized powders, including two separate conduit loops for circulating fluid samples within the range of radiation sources and detectors is described. Associated with the first loop is a neutron source that emits s high flux of slow and thermal neutrons. The second loop employs a fast neutron source, the flux from which is substantially free of thermal neutrons. Adjacent to both loops are gamma counters for spectrographic determination of the fluid constituents. Other gsmma sources and detectors are arranged across a portion of each loop for deterMining the fluid density. (Official Gazette)

Neutron Activation Diagnostics in Deuterium Gas-Puff Experiments on the 3 MA GIT-12 Z-Pinch

NASA Astrophysics Data System (ADS)

Cikhardt, J.; Klir, D.; Rezac, K.; Cikhardtova, B.; Kravarik, J.; Kubes, P.; Sila, O.; Shishlov, A. V.; Cherdizov, R. K.; Fursov, F. I.; Kokshenev, V. A.; Kurmaev, N. E.; Labetsky, A. Yu; Ratakhin, N. A.; Dudkin, G. N.; Garapatsky, A. A.; Padalko, V. N.; Varlachev, V. A.; Turek, K.

2016-10-01

The experiments with a deuterium z-pinch on the GIT-12 generator at IHCE in Tomsk were performed in the frame of the Czech-Russian agreement. A set of neutron diagnostics included scintillation time-of-flight detectors, bubble detectors, and several kinds of threshold nuclear activation detectors in the order to obtain information about the yield, anisotropy, and spectrum of the neutrons produced by a deuterium gas-puff. The average neutron yield in these experiments was of the order of 1012 neutrons per a single shot. The energy spectrum of the produced neutrons was evaluated using neutron time-of-flight detectors and a set of neutron activation detectors. Because the deuterons in the pinch achieve multi-MeV energies, non-DD neutrons are produced by nuclear reactions of deuterons with a stainless steel vacuum chamber and aluminum components of diagnostics inside the chamber. An estimated number of the non-DD was of the order of 1011. GACR (Grant No. 16-07036S), CME (Grant Nos. LD14089, LG13029, and LH13283), MESRF (Grant No. RFMEFI59114X0001), IAEA (Grant No. RC17088), CTU (Grant No. SGS 16/223/OHK3/3T/13).

New experimental research stand SVICKA neutron field analysis using neutron activation detector technique

NASA Astrophysics Data System (ADS)

Varmuza, Jan; Katovsky, Karel; Zeman, Miroslav; Stastny, Ondrej; Haysak, Ivan; Holomb, Robert

2018-04-01

Knowledge of neutron energy spectra is very important because neutrons with various energies have a different material impact or a biological tissue impact. This paper presents basic results of the neutron flux distribution inside the new experimental research stand SVICKA which is located at Brno University of Technology in Brno, Czech Republic. The experiment also focused on the investigation of the sandwich biological shielding quality that protects staff against radiation effects. The set of indium activation detectors was used to the investigation of neutron flux distribution. The results of the measurement provide basic information about the neutron flux distribution inside all irradiation channels and no damage or cracks are present in the experimental research stand biological shielding.

Radiological risks of neutron interrogation of food.

PubMed

Albright, S; Seviour, R

2015-09-01

In recent years there has been growing interest in the use of neutron scanning techniques for security. Neutron techniques with a range of energy spectra including thermal, white and fast neutrons have been shown to work in different scenarios. As international interest in neutron scanning increases the risk of activating cargo, especially foodstuffs must be considered. There has been a limited amount of research into the activation of foods by neutron beams and we have sought to improve the amount of information available. In this paper we show that for three important metrics; activity, ingestion dose and Time to Background there is a strong dependence on the food being irradiated and a weak dependence on the energy of irradiation. Previous studies into activation used results based on irradiation of pharmaceuticals as the basis for research into activation of food. The earlier work reports that (24)Na production is the dominant threat which motivated the search for (24)Na(n,γ)(24)Na in highly salted foods. We show that (42)K can be more significant than (24)Na in low sodium foods such as Bananas and Potatoes.

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.

The Use of Radioactivation Analysis in Biology; APPLICAZIONI DELL'ANALISI PER RADIOATTIVAZIONE ALLA BIOLOGIA

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

Merlini, M.

1962-01-01

The principles of activation analysis and the methods of detection and measurement of radioactivity from neutron irradiated samples are described. The application of this method in different fields is mentioned. An example of the use of activation analysis in biology is given; and the results of a study on the manganese content in different parts of a lamellibranch, Unio mancus elongatus (Pfeiffer) of Lago Maggiore, are presented and discussed. (auth)

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.).

Active Well Counting Using New PSD Plastic Detectors

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

Hausladen, Paul; Newby, Jason; McElroy, Robert Dennis

This report presents results and analysis from a series of proof-of-concept measurements to assess the suitability of segmented detectors constructed from Eljen EJ-299-34 PSD-plastic scintillator with pulse-shape discrimination capability for the purposes of quantifying uranium via active neutron coincidence counting. Present quantification of bulk uranium materials for international safeguards and domestic materials control and accounting relies on active neutron coincidence counting systems, such as the Active Well Coincidence Counter (AWCC) and the Uranium Neutron Coincidence Collar (UNCL), that use moderated He-3 proportional counters along with necessarily low-intensity 241Am(Li) neutron sources. Scintillation-based fast-neutron detectors are a potentially superior technology to themore » existing AWCC and UNCL designs due to their spectroscopic capability and their inherently short neutron coincidence times that largely eliminate random coincidences and enable interrogation by stronger sources. One of the past impediments to the investigation and adoption of scintillation counters for the purpose of quantifying bulk uranium was the commercial availability of scintillators having the necessary neutron-gamma pulse-shape discrimination properties only as flammable liquids. Recently, Eljen EJ-299-34 PSD-plastic scintillator became commercially available. The present work is the first assessment of an array of PSD-plastic detectors for the purposes of quantifying bulk uranium. The detector panel used in the present work was originally built as the focal plane for a fast-neutron imager, but it was repurposed for the present investigation by construction of a stand to support the inner well of an AWCC immediately in front of the detector panel. The detector panel and data acquisition of this system are particularly well suited for performing active-well fast-neutron counting of LEU and HEU samples because the active detector volume is solid, the 241Am(Li) interrogating neutrons are largely below the detector threshold, and the segmented construction of the detector modules allow for separation of true neutron-neutron coincidences from inter-detector scattering using the kinematics of neutron scattering. The results from a series of measurements of a suite of uranium standards are presented, and compared to measurements of the same standards and source configurations using the AWCC. Using these results, the performance of the segmented detectors reconfigured as a well counter is predicted and outperforms the AWCC.« less

Advanced techniques for determining long term compatibility of materials with propellants

NASA Technical Reports Server (NTRS)

Green, R. L.; Stebbins, J. P.; Smith, A. W.; Pullen, K. E.

1973-01-01

A method for the prediction of propellant-material compatibility for periods of time up to ten years is presented. Advanced sensitive measurement techniques used in the prediction method are described. These include: neutron activation analysis, radioactive tracer technique, and atomic absorption spectroscopy with a graphite tube furnace sampler. The results of laboratory tests performed to verify the prediction method are presented.

Neutron detection with a NaI spectrometer using high-energy photons

NASA Astrophysics Data System (ADS)

Holm, Philip; Peräjärvi, Kari; Sihvonen, Ari-Pekka; Siiskonen, Teemu; Toivonen, Harri

2013-01-01

Neutrons can be indirectly detected by high-energy photons. The performance of a 4″×4″×16″ NaI portal monitor was compared to a 3He-based portal monitor with a comparable cross-section of the active volume. Measurements were performed with bare and shielded 252Cf and AmBe sources. With an optimum converter and moderator structure for the NaI detector, the detection efficiencies and minimum detectable activities of the portal monitors were similar. The NaI portal monitor preserved its detection efficiency much better with shielded sources, making the method very interesting for security applications. For heavily shielded sources, the NaI detector was 2-3 times more sensitive than the 3He-based detector.

Container inspection in the port container terminal by using 14 MeV neutrons

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

Valkovic, Vladivoj; Kvinticka 62, Zagreb; Sudac, Davorin

A proposal for an autonomous and flexible ship container inspection system is presented. This could be accomplished by the incorporation of inspection system on the container transportation devices (straddle carriers, yard gentry cranes automated guided vehicles, trailers). This configuration is terminal specific and it will be decided by container terminal operator. In such a way no part of port operational area will be used for inspection. The inspection scenario will include container transfer from ship to transportation device with inspection unit mounted on it, inspection during container movement to the container location. A neutron generator without associated alpha particle detectionmore » will be used. This will allow the use of higher neutron intensity (5x10{sup 9} - 10{sup 10} n/s in 4π). The inspected container will be stationary in the 'inspection position' on the transportation device while the 'inspection unit' will move along its side. Following analytical methods will be used simultaneously: neutron radiography, X-ray radiography, neutron activation analysis, (n,γ) and (n,n'γ) reactions, neutron absorption, and scattering, X-ray backscattering, Neutron techniques will take the advantage of using 'smart collimators' for neutrons and gammas, both emitted and detected. The inspected voxel will be defined by intersections/union of neutron generator and detectors solid angles. The container inspection protocol will be based on identification of discrepancies between its cargo manifest and its elemental 'fingerprint' and radiography profiles. In addition, the information on container weight will be obtained during the container transport and foreseen screening from the measurement of density of material in the container. (authors)« less

Neutron Zeeman beam-splitting for the investigation of magnetic nanostructures

NASA Astrophysics Data System (ADS)

Kozhevnikov, S. V.; Ott, F.; Semenova, E.

2017-03-01

Zeeman spatial splitting of a neutron beam takes place during a neutron spin-flip in magnetically non-collinear systems at grazing incidence geometry. We apply the neutron beam-splitting method for the investigation of magnetically non-collinear clusters of submicron size in a thin film. The experimental results are compared with ones obtained by other methods.

Progress toward a new beam measurement of the neutron lifetime

NASA Astrophysics Data System (ADS)

Hoogerheide, Shannon Fogwell

2016-09-01

Neutron beta decay is the simplest example of nuclear beta decay. A precise value of the neutron lifetime is important for consistency tests of the Standard Model and Big Bang Nucleosysnthesis models. The beam neutron lifetime method requires the absolute counting of the decay protons in a neutron beam of precisely known flux. Recent work has resulted in improvements in both the neutron and proton detection systems that should permit a significant reduction in systematic uncertainties. A new measurement of the neutron lifetime using the beam method will be performed at the National Institute of Standards and Technology Center for Neutron Research. The projected uncertainty of this new measurement is 1 s. An overview of the measurement and the technical improvements will be discussed.

Benchmark test of transport calculations of gold and nickel activation with implications for neutron kerma at Hiroshima.

PubMed

Hoshi, M; Hiraoka, M; Hayakawa, N; Sawada, S; Munaka, M; Kuramoto, A; Oka, T; Iwatani, K; Shizuma, K; Hasai, H

1992-11-01

A benchmark test of the Monte Carlo neutron and photon transport code system (MCNP) was performed using a 252Cf fission neutron source to validate the use of the code for the energy spectrum analyses of Hiroshima atomic bomb neutrons. Nuclear data libraries used in the Monte Carlo neutron and photon transport code calculation were ENDF/B-III, ENDF/B-IV, LASL-SUB, and ENDL-73. The neutron moderators used were granite (the main component of which is SiO2, with a small fraction of hydrogen), Newlight [polyethylene with 3.7% boron (natural)], ammonium chloride (NH4Cl), and water (H2O). Each moderator was 65 cm thick. The neutron detectors were gold and nickel foils, which were used to detect thermal and epithermal neutrons (4.9 eV) and fast neutrons (> 0.5 MeV), respectively. Measured activity data from neutron-irradiated gold and nickel foils in these moderators decreased to about 1/1,000th or 1/10,000th, which correspond to about 1,500 m ground distance from the hypocenter in Hiroshima. For both gold and nickel detectors, the measured activities and the calculated values agreed within 10%. The slopes of the depth-yield relations in each moderator, except granite, were similar for neutrons detected by the gold and nickel foils. From the results of these studies, the Monte Carlo neutron and photon transport code was verified to be accurate enough for use with the elements hydrogen, carbon, nitrogen, oxygen, silicon, chlorine, and cadmium, and for the incident 252Cf fission spectrum neutrons.

Non-streaming high-efficiency perforated semiconductor neutron detectors, methods of making same and measuring wand and detector modules utilizing same

DOEpatents

McGregor, Douglas S.; Shultis, John K.; Rice, Blake B.; McNeil, Walter J.; Solomon, Clell J.; Patterson, Eric L.; Bellinger, Steven L.

2010-12-21

Non-streaming high-efficiency perforated semiconductor neutron detectors, method of making same and measuring wands and detector modules utilizing same are disclosed. The detectors have improved mechanical structure, flattened angular detector responses, and reduced leakage current. A plurality of such detectors can be assembled into imaging arrays, and can be used for neutron radiography, remote neutron sensing, cold neutron imaging, SNM monitoring, and various other applications.

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.

Pulsed Neurton Elemental On-Line Material Analyzer

DOEpatents

Vourvopoulos, George

2002-08-20

An on-line material analyzer which utilizes pulsed neutron generation in order to determine the composition of material flowing through the apparatus. The on-line elemental material analyzer is based on a pulsed neutron generator. The elements in the material interact with the fast and thermal neutrons produced from the pulsed generator. Spectra of gamma-rays produced from fast neutrons interacting with elements of the material are analyzed and stored separately from spectra produced from thermal neutron reactions. Measurements of neutron activation takes place separately from the above reactions and at a distance from the neutron generator. A primary passageway allows the material to flow through at a constant rate of speed and operators to provide data corresponding to fast and thermal neutron reactions. A secondary passageway meters the material to allow for neutron activation analysis. The apparatus also has the capability to determine the density of the flowed material. Finally, the apparatus continually utilizes a neutron detector in order to normalize the yield of the gamma ray detectors and thereby automatically calibrates and adjusts the spectra data for fluctuations in neutron generation.

Multiple-wavelength neutron holography with pulsed neutrons

PubMed Central

Hayashi, Kouichi; Ohoyama, Kenji; Happo, Naohisa; Matsushita, Tomohiro; Hosokawa, Shinya; Harada, Masahide; Inamura, Yasuhiro; Nitani, Hiroaki; Shishido, Toetsu; Yubuta, Kunio

2017-01-01

Local structures around impurities in solids provide important information for understanding the mechanisms of material functions, because most of them are controlled by dopants. For this purpose, the x-ray absorption fine structure method, which provides radial distribution functions around specific elements, is most widely used. However, a similar method using neutron techniques has not yet been developed. If one can establish a method of local structural analysis with neutrons, then a new frontier of materials science can be explored owing to the specific nature of neutron scattering—that is, its high sensitivity to light elements and magnetic moments. Multiple-wavelength neutron holography using the time-of-flight technique with pulsed neutrons has great potential to realize this. We demonstrated multiple-wavelength neutron holography using a Eu-doped CaF2 single crystal and obtained a clear three-dimensional atomic image around trivalent Eu substituted for divalent Ca, revealing an interesting feature of the local structure that allows it to maintain charge neutrality. The new holography technique is expected to provide new information on local structures using the neutron technique. PMID:28835917

Multiple-wavelength neutron holography with pulsed neutrons.

PubMed

Hayashi, Kouichi; Ohoyama, Kenji; Happo, Naohisa; Matsushita, Tomohiro; Hosokawa, Shinya; Harada, Masahide; Inamura, Yasuhiro; Nitani, Hiroaki; Shishido, Toetsu; Yubuta, Kunio

2017-08-01

Local structures around impurities in solids provide important information for understanding the mechanisms of material functions, because most of them are controlled by dopants. For this purpose, the x-ray absorption fine structure method, which provides radial distribution functions around specific elements, is most widely used. However, a similar method using neutron techniques has not yet been developed. If one can establish a method of local structural analysis with neutrons, then a new frontier of materials science can be explored owing to the specific nature of neutron scattering-that is, its high sensitivity to light elements and magnetic moments. Multiple-wavelength neutron holography using the time-of-flight technique with pulsed neutrons has great potential to realize this. We demonstrated multiple-wavelength neutron holography using a Eu-doped CaF 2 single crystal and obtained a clear three-dimensional atomic image around trivalent Eu substituted for divalent Ca, revealing an interesting feature of the local structure that allows it to maintain charge neutrality. The new holography technique is expected to provide new information on local structures using the neutron technique.

Focusing mirrors for enhanced neutron radiography with thermal neutrons and application for irradiated nuclear fuel

NASA Astrophysics Data System (ADS)

Rai, Durgesh K.; Abir, Muhammad; Wu, Huarui; Khaykovich, Boris; Moncton, David E.

2018-01-01

Neutron radiography is a powerful method of probing the structure of materials based on attenuation of neutrons. This method is most suitable for materials containing heavy metals, which are not transparent to X-rays, for example irradiated nuclear fuel and other nuclear materials. Neutron radiography is one of the first non-distractive post-irradiated examination methods, which is applied to gain an overview of the integrity of irradiated nuclear fuel and other nuclear materials. However, very powerful gamma radiation emitted by the samples is damaging to the electronics of digital imaging detectors and has so far precluded the use of modern detectors. Here we describe a design of a neutron microscope based on focusing mirrors suitable for thermal neutrons. As in optical microscopes, the sample is separated from the detector, decreasing the effect of gamma radiation. In addition, the application of mirrors would result in a thirty-fold gain in flux and a resolution of better than 40 μm for a field-of-view of about 2.5 cm. Such a thermal neutron microscope can be useful for other applications of neutron radiography, where thermal neutrons are advantageous.

Calibrating and training of neutron based NSA techniques with less SNM standards

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

Geist, William H; Swinhoe, Martyn T; Bracken, David S

2010-01-01

Accessing special nuclear material (SNM) standards for the calibration of and training on nondestructive assay (NDA) instruments has become increasingly difficult in light of enhanced safeguards and security regulations. Limited or nonexistent access to SNM has affected neutron based NDA techniques more than gamma ray techniques because the effects of multiplication require a range of masses to accurately measure the detector response. Neutron based NDA techniques can also be greatly affected by the matrix and impurity characteristics of the item. The safeguards community has been developing techniques for calibrating instrumentation and training personnel with dwindling numbers of SNM standards. Montemore » Carlo methods have become increasingly important for design and calibration of instrumentation. Monte Carlo techniques have the ability to accurately predict the detector response for passive techniques. The Monte Carlo results are usually benchmarked to neutron source measurements such as californium. For active techniques, the modeling becomes more difficult because of the interaction of the interrogation source with the detector and nuclear material; and the results cannot be simply benchmarked with neutron sources. A Monte Carlo calculated calibration curve for a training course in Indonesia of material test reactor (MTR) fuel elements assayed with an active well coincidence counter (AWCC) will be presented as an example. Performing training activities with reduced amounts of nuclear material makes it difficult to demonstrate how the multiplication and matrix properties of the item affects the detector response and limits the knowledge that can be obtained with hands-on training. A neutron pulse simulator (NPS) has been developed that can produce a pulse stream representative of a real pulse stream output from a detector measuring SNM. The NPS has been used by the International Atomic Energy Agency (IAEA) for detector testing and training applications at the Agency due to the lack of appropriate SNM standards. This paper will address the effect of reduced access to SNM for calibration and training of neutron NDA applications along with the advantages and disadvantages of some solutions that do not use standards, such as the Monte Carlo techniques and the NPS.« less

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

Trahan, Alexis Chanel

The objectives of this presentation are to introduce the basic physics of neutron production, interactions and detection; identify the processes that generate neutrons; explain the most common neutron mechanism, spontaneous and induced fission and (a,n) reactions; describe the properties of neutron from different sources; recognize advantages of neutron measurements techniques; recognize common neutrons interactions; explain neutron cross section measurements; describe the fundamental of 3He detector function and designs; and differentiate between passive and active assay techniques.

New generation non-stationary portable neutron generators for biophysical applications of Neutron Activation Analysis.

PubMed

Marchese, N; Cannuli, A; Caccamo, M T; Pace, C

2017-01-01

Neutron sources are increasingly employed in a wide range of research fields. For some specific purposes an alternative to existing large-scale neutron scattering facilities, can be offered by the new generation of portable neutron devices. This review reports an overview for such recently available neutron generators mainly addressed to biophysics applications with specific reference to portable non-stationary neutron generators applied in Neutron Activation Analysis (NAA). The review reports a description of a typical portable neutron generator set-up addressed to biophysics applications. New generation portable neutron devices, for some specific applications, can constitute an alternative to existing large-scale neutron scattering facilities. Deuterium-Deuterium pulsed neutron sources able to generate 2.5MeV neutrons, with a neutron yield of 1.0×10 6 n/s, a pulse rate of 250Hz to 20kHz and a duty factor varying from 5% to 100%, when combined with solid-state photon detectors, show that this kind of compact devices allow rapid and user-friendly elemental analysis. "This article is part of a Special Issue entitled "Science for Life" Guest Editor: Dr. Austen Angell, Dr. Salvatore Magazù and Dr. Federica Migliardo". Copyright © 2016 Elsevier B.V. All rights reserved.

Improved neutron-gamma discrimination for a 6Li-glass neutron detector using digital signal analysis methods

DOE PAGES

Wang, Cai -Lin; Riedel, Richard A.

2016-01-14

A 6Li-glass scintillator (GS20) based neutron Anger camera was developed for time-of-flight single-crystal diffraction instruments at SNS. Traditional pulse-height analysis (PHA) for neutron-gamma discrimination (NGD) resulted in the neutron-gamma efficiency ratio (defined as NGD ratio) on the order of 10 4. The NGD ratios of Anger cameras need to be improved for broader applications including neutron reflectometers. For this purpose, five digital signal analysis methods of individual waveforms from PMTs were proposed using: i). pulse-amplitude histogram; ii). power spectrum analysis combined with the maximum pulse amplitude; iii). two event parameters (a 1, b 0) obtained from Wiener filter; iv). anmore » effective amplitude (m) obtained from an adaptive least-mean-square (LMS) filter; and v). a cross-correlation (CC) coefficient between an individual waveform and a reference. The NGD ratios can be 1-102 times those from traditional PHA method. A brighter scintillator GS2 has better NGD ratio than GS20, but lower neutron detection efficiency. The ultimate NGD ratio is related to the ambient, high-energy background events. Moreover, our results indicate the NGD capability of neutron Anger cameras can be improved using digital signal analysis methods and brighter neutron scintillators.« less

BOREHOLE NEUTRON ACTIVATION: THE RARE EARTHS.

USGS Publications Warehouse

Mikesell, J.L.; Senftle, F.E.

1987-01-01

Neutron-induced borehole gamma-ray spectroscopy has been widely used as a geophysical exploration technique by the petroleum industry, but its use for mineral exploration is not as common. Nuclear methods can be applied to mineral exploration, for determining stratigraphy and bed correlations, for mapping ore deposits, and for studying mineral concentration gradients. High-resolution detectors are essential for mineral exploration, and by using them an analysis of the major element concentrations in a borehole can usually be made. A number of economically important elements can be detected at typical ore-grade concentrations using this method. Because of the application of the rare-earth elements to high-temperature superconductors, these elements are examined in detail as an example of how nuclear techniques can be applied to mineral exploration.

Sample registration software for process automation in the Neutron Activation Analysis (NAA) Facility in Malaysia nuclear agency

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

Rahman, Nur Aira Abd, E-mail: nur-aira@nuclearmalaysia.gov.my; Yussup, Nolida; Ibrahim, Maslina Bt. Mohd

Neutron Activation Analysis (NAA) had been established in Nuclear Malaysia since 1980s. Most of the procedures established were done manually including sample registration. The samples were recorded manually in a logbook and given ID number. Then all samples, standards, SRM and blank were recorded on the irradiation vial and several forms prior to irradiation. These manual procedures carried out by the NAA laboratory personnel were time consuming and not efficient. Sample registration software is developed as part of IAEA/CRP project on ‘Development of Process Automation in the Neutron Activation Analysis (NAA) Facility in Malaysia Nuclear Agency (RC17399)’. The objective ofmore » the project is to create a pc-based data entry software during sample preparation stage. This is an effective method to replace redundant manual data entries that needs to be completed by laboratory personnel. The software developed will automatically generate sample code for each sample in one batch, create printable registration forms for administration purpose, and store selected parameters that will be passed to sample analysis program. The software is developed by using National Instruments Labview 8.6.« less

Neutron range spectrometer

DOEpatents

Manglos, S.H.

1988-03-10

A neutron range spectrometer and method for determining the neutron energy spectrum of a neutron emitting source are disclosed. Neutrons from the source are colliminated along a collimation axis and a position sensitive neutron counter is disposed in the path of the collimated neutron beam. The counter determines positions along the collimation axis of interactions between the neutrons in the neutron beam and a neutron-absorbing material in the counter. From the interaction positions, a computer analyzes the data and determines the neutron energy spectrum of the neutron beam. The counter is preferably shielded and a suitable neutron-absorbing material is He-3. 1 fig.

Stellar neutron capture cross sections of 41K and 45Sc

NASA Astrophysics Data System (ADS)

Heil, M.; Plag, R.; Uberseder, E.; Bisterzo, S.; Käppeler, F.; Mengoni, A.; Pignatari, M.

2016-05-01

The neutron capture cross sections of light nuclei (A <56 ) are important for s -process scenarios since they act as neutron poisons. We report on measurements of the neutron capture cross sections of 41K and 45Sc, which were performed at the Karlsruhe 3.7 MV Van de Graaff accelerator via the activation method in a quasistellar neutron spectrum corresponding to a thermal energy of k T =25 keV. Systematic effects were controlled by repeated irradiations, resulting in overall uncertainties of less than 3%. The measured spectrum-averaged data have been used to normalize the energy-dependent (n ,γ ) cross sections from the main data libraries JEFF-3.2, JENDL-4.0, and ENDF/B-VII.1, and a set of Maxwellian averaged cross sections was calculated for improving the s -process nucleosynthesis yields in AGB stars and in massive stars. At k T =30 keV, the new Maxwellian averaged cross sections of 41K and 45Sc are 19.2 ±0.6 mb and 61.3 ±1.8 mb, respectively. Both values are 20% lower than previously recommended. The effect of neutron poisons is discussed for nuclei with A <56 in general and for the investigated isotopes in particular.

Feasibility study for wax deposition imaging in oil pipelines by PGNAA technique.

PubMed

Cheng, Can; Jia, Wenbao; Hei, Daqian; Wei, Zhiyong; Wang, Hongtao

2017-10-01

Wax deposition in pipelines is a crucial problem in the oil industry. A method based on the prompt gamma-ray neutron activation analysis technique was applied to reconstruct the image of wax deposition in oil pipelines. The 2.223MeV hydrogen capture gamma rays were used to reconstruct the wax deposition image. To validate the method, both MCNP simulation and experiments were performed for wax deposited with a maximum thickness of 20cm. The performance of the method was simulated using the MCNP code. The experiment was conducted with a 252 Cf neutron source and a LaBr 3 : Ce detector. A good correspondence between the simulations and the experiments was observed. The results obtained indicate that the present approach is efficient for wax deposition imaging in oil pipelines. Copyright © 2017 Elsevier Ltd. All rights reserved.

Performance assessment of imaging plates for the JHR transfer Neutron Imaging System

NASA Astrophysics Data System (ADS)

Simon, E.; Guimbal, P. AB(; )

2018-01-01

The underwater Neutron Imaging System to be installed in the Jules Horowitz Reactor (JHR-NIS) is based on a transfer method using a neutron activated beta-emitter like Dysprosium. The information stored in the converter is to be offline transferred on a specific imaging system, still to be defined. Solutions are currently under investigation for the JHR-NIS in order to anticipate the disappearance of radiographic films commonly used in these applications. We report here the performance assessment of Computed Radiography imagers (Imaging Plates) performed at LLB/Orphée (CEA Saclay). Several imaging plate types are studied, in one hand in the configuration involving an intimate contact with an activated dysprosium foil converter: Fuji BAS-TR, Fuji UR-1 and Carestream Flex XL Blue imaging plates, and in the other hand by using a prototypal imaging plate doped with dysprosium and thus not needing any contact with a separate converter foil. The results for these imaging plates are compared with those obtained with gadolinium doped imaging plate used in direct neutron imaging (Fuji BAS-ND). The detection performances of the different imagers are compared regarding resolution and noise. The many advantages of using imaging plates over radiographic films (high sensitivity, linear response, high dynamic range) could palliate its lower intrinsic resolution.

APPARATUS FOR CONTROLLING NEUTRONIC REACTORS

DOEpatents

Dietrich, J.R.; Harrer, J.M.

1958-09-16

A device is described for rapidly cortrolling the reactivity of an active portion of a reactor. The inveniion consists of coaxially disposed members each having circumferenital sections of material having dlfferent neutron absorbing characteristics and means fur moving the members rotatably and translatably relative to each other within the active portion to vary the neutron flux therein. The angular and translational movements of any member change the neutron flux shadowing effect of that member upon the other member.

Results on the neutron energy distribution measurements at the RECH-1 Chilean nuclear reactor

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

Aguilera, P., E-mail: paguilera87@gmail.com; Romero-Barrientos, J.; Universidad de Chile, Dpto. de Física, Facultad de Ciencias, Las Palmeras 3425, Nuñoa, Santiago

2016-07-07

Neutron activations experiments has been perform at the RECH-1 Chilean Nuclear Reactor to measure its neutron flux energy distribution. Samples of pure elements was activated to obtain the saturation activities for each reaction. Using - ray spectroscopy we identify and measure the activity of the reaction product nuclei, obtaining the saturation activities of 20 reactions. GEANT4 and MCNP was used to compute the self shielding factor to correct the cross section for each element. With the Expectation-Maximization algorithm (EM) we were able to unfold the neutron flux energy distribution at dry tube position, near the RECH-1 core. In this work,more » we present the unfolding results using the EM algorithm.« less

Method and apparatus for detecting neutrons

DOEpatents

Perkins, R.W.; Reeder, P.L.; Wogman, N.A.; Warner, R.A.; Brite, D.W.; Richey, W.C.; Goldman, D.S.

1997-10-21

The instant invention is a method for making and using an apparatus for detecting neutrons. Scintillating optical fibers are fabricated by melting SiO{sub 2} with a thermal neutron capturing substance and a scintillating material in a reducing atmosphere. The melt is then drawn into fibers in an anoxic atmosphere. The fibers may then be coated and used directly in a neutron detection apparatus, or assembled into a geometrical array in a second, hydrogen-rich, scintillating material such as a polymer. Photons generated by interaction with thermal neutrons are trapped within the coated fibers and are directed to photoelectric converters. A measurable electronic signal is generated for each thermal neutron interaction within the fiber. These electronic signals are then manipulated, stored, and interpreted by normal methods to infer the quality and quantity of incident radiation. When the fibers are arranged in an array within a second scintillating material, photons generated by kinetic neutrons interacting with the second scintillating material and photons generated by thermal neutron capture within the fiber can both be directed to photoelectric converters. These electronic signals are then manipulated, stored, and interpreted by normal methods to infer the quality and quantity of incident radiation. 5 figs.

Method and apparatus for detecting neutrons

DOEpatents

Perkins, Richard W.; Reeder, Paul L.; Wogman, Ned A.; Warner, Ray A.; Brite, Daniel W.; Richey, Wayne C.; Goldman, Don S.

1997-01-01

The instant invention is a method for making and using an apparatus for detecting neutrons. Scintillating optical fibers are fabricated by melting SiO.sub.2 with a thermal neutron capturing substance and a scintillating material in a reducing atmosphere. The melt is then drawn into fibers in an anoxic atmosphere. The fibers may then be coated and used directly in a neutron detection apparatus, or assembled into a geometrical array in a second, hydrogen-rich, scintillating material such as a polymer. Photons generated by interaction with thermal neutrons are trapped within the coated fibers and are directed to photoelectric converters. A measurable electronic signal is generated for each thermal neutron interaction within the fiber. These electronic signals are then manipulated, stored, and interpreted by normal methods to infer the quality and quantity of incident radiation. When the fibers are arranged in an array within a second scintillating material, photons generated by kinetic neutrons interacting with the second scintillating material and photons generated by thermal neutron capture within the fiber can both be directed to photoelectric converters. These electronic signals are then manipulated, stored, and interpreted by normal methods to infer the quality and quantity of incident radiation.

Improved neutron-gamma discrimination for a {sup 6}Li-glass neutron detector using digital signal analysis methods

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

Wang, C. L., E-mail: wangc@ornl.gov; Riedel, R. A.

2016-01-15

A {sup 6}Li-glass scintillator (GS20) based neutron Anger camera was developed for time-of-flight single-crystal diffraction instruments at Spallation Neutron Source. Traditional Pulse-Height Analysis (PHA) for Neutron-Gamma Discrimination (NGD) resulted in the neutron-gamma efficiency ratio (defined as NGD ratio) on the order of 10{sup 4}. The NGD ratios of Anger cameras need to be improved for broader applications including neutron reflectometers. For this purpose, six digital signal analysis methods of individual waveforms acquired from photomultiplier tubes were proposed using (i) charge integration, (ii) pulse-amplitude histograms, (iii) power spectrum analysis combined with the maximum pulse-amplitude, (iv) two event parameters (a{sub 1}, b{submore » 0}) obtained from a Wiener filter, (v) an effective amplitude (m) obtained from an adaptive least-mean-square filter, and (vi) a cross-correlation coefficient between individual and reference waveforms. The NGD ratios are about 70 times those from the traditional PHA method. Our results indicate the NGD capabilities of neutron Anger cameras based on GS20 scintillators can be significantly improved with digital signal analysis methods.« less

Active detection of shielded SNM with 60-keV neutrons

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

Hagmann, C; Dietrich, D; Hall, J

2008-07-08

Fissile materials, e.g. {sup 235}U and {sup 239}Pu, can be detected non-invasively by active neutron interrogation. A unique characteristic of fissile material exposed to neutrons is the prompt emission of high-energy (fast) fission neutrons. One promising mode of operation subjects the object to a beam of medium-energy (epithermal) neutrons, generated by a proton beam impinging on a Li target. The emergence of fast secondary neutrons then clearly indicates the presence of fissile material. Our interrogation system comprises a low-dose 60-keV neutron generator (5 x 10{sup 6}/s), and a 1 m{sup 2} array of scintillators for fast neutron detection. Preliminary experimentalmore » results demonstrate the detectability of small quantities (370 g) of HEU shielded by steel (200 g/cm{sup 2}) or plywood (30 g/cm{sup 2}), with a typical measurement time of 1 min.« less

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.

Towards an In-Beam Measurement of the Neutron Lifetime to 1 Second

NASA Astrophysics Data System (ADS)

Mulholland, Jonathan

2014-03-01

A precise value for the neutron lifetime is required for consistency tests of the Standard Model and is an essential parameter in the theory of Big Bang Nucleosynthesis. A new measurement of the neutron lifetime using the in-beam method is planned at the National Institute of Standards and Technology Center for Neutron Research. The systematic effects associated with the in-beam method are markedly different than those found in storage experiments utilizing ultracold neutrons. Experimental improvements, specifically recent advances in the determination of absolute neutron fluence, should permit an overall uncertainty of 1 second on the neutron lifetime. The dependence of the primordial mass fraction on the neutron lifetime, technical improvements of the in-beam technique, and the path toward improving the precision of the new measurement will be discussed.

Progress toward a new beam measurement of the neutron lifetime

NASA Astrophysics Data System (ADS)

Hoogerheide, Shannon Fogwell; BL2 Collaboration

2017-01-01

Neutron beta decay is the simplest example of nuclear beta decay. A precise value of the neutron lifetime is important for consistency tests of the Standard Model and Big Bang Nucleosynthesis models. The beam neutron lifetime method requires the absolute counting of the decay protons in a neutron beam of precisely known flux. Recent work has resulted in improvements in both the neutron and proton detection systems that should permit a significant reduction in systematic uncertainties. A new measurement of the neutron lifetime using the beam method is underway at the National Institute of Standards and Technology Center for Neutron Research. The projected uncertainty of this new measurement is 1 s. An overview of the measurement, its current status, and the technical improvements will be discussed.

Mercury and gold concentrations of highly polluted environmental samples determined using prompt gamma-ray analysis and instrument neutron activation analysis

NASA Astrophysics Data System (ADS)

Osawa, Takahito; Hatsukawa, Yuichi; Appel, Peter W. U.; Matsue, Hideaki

2011-04-01

The authors have established a method of determining mercury and gold in severely polluted environmental samples using prompt gamma-ray analysis (PGA) and instrumental neutron activation analysis (INAA). Since large amounts of mercury are constantly being released into the environment by small-scale gold mining in many developing countries, the mercury concentration in tailings and water has to be determined to mitigate environmental pollution. Cold-vapor atomic absorption analysis, the most pervasive method of mercury analysis, is not suitable because tailings and water around mining facilities have extremely high mercury concentrations. On the other hand, PGA can determine high mercury concentrations in polluted samples as it has an appropriate level of sensitivity. Moreover, gold concentrations can be determined sequentially by using INAA after PGA. In conclusion, the analytical procedure established in this work using PGA and INAA is the best way to evaluate the degree of pollution and the tailing resource value. This method will significantly contribute to mitigating problems in the global environment.

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.

Method and apparatus for determination of temperature, neutron absorption cross section and neutron moderating power

DOEpatents

Vagelatos, Nicholas; Steinman, Donald K.; John, Joseph; Young, Jack C.

1981-01-01

A nuclear method and apparatus determines the temperature of a medium by injecting fast neutrons into the medium and detecting returning slow neutrons in three first energy ranges by producing three respective detection signals. The detection signals are combined to produce three derived indicia each systematically related to the population of slow neutrons returning from the medium in a respective one of three second energy ranges, specifically exclusively epithermal neutrons, exclusively substantially all thermal neutrons and exclusively a portion of the thermal neutron spectrum. The derived indicia are compared with calibration indicia similarly systematically related to the population of slow neutrons in the same three second energy ranges returning from similarly irradiated calibration media for which the relationships temperature, neutron absorption cross section and neutron moderating power to such calibration indicia are known. The comparison indicates the temperature at which the calibration indicia correspond to the derived indicia and consequently the temperature of the medium. The neutron absorption cross section and moderating power of the medium can be identified at the same time.

Improved neutron activation prediction code system development

NASA Technical Reports Server (NTRS)

Saqui, R. M.

1971-01-01

Two integrated neutron activation prediction code systems have been developed by modifying and integrating existing computer programs to perform the necessary computations to determine neutron induced activation gamma ray doses and dose rates in complex geometries. Each of the two systems is comprised of three computational modules. The first program module computes the spatial and energy distribution of the neutron flux from an input source and prepares input data for the second program which performs the reaction rate, decay chain and activation gamma source calculations. A third module then accepts input prepared by the second program to compute the cumulative gamma doses and/or dose rates at specified detector locations in complex, three-dimensional geometries.

Calibration of the JET neutron activation system for DT operation

NASA Astrophysics Data System (ADS)

Bertalot, L.; Roquemore, A. L.; Loughlin, M.; Esposito, B.

1999-01-01

The neutron activation system at JET is a pneumatic transfer system capable of positioning activation samples close to the plasma. Its primary purpose is to provide a calibration for the time-dependent neutron yield monitors (fission chambers and solid state detectors). Various activation reactions with different high energy thresholds were used including 56Fe(n,p) 56Mn, 27Al(n,α) 24Na, 93Nb(n,2n) 92mNb, and 28Si(n,p) 28Al reactions. The silicon reaction, with its short half life (2.25 min), provides a prompt determination of the 14 MeV DT yield. The neutron induced γ-ray activity of the Si samples was measured using three sodium iodide scintillators, while two high purity germanium detectors were used for other foils. It was necessary to use a range of sample masses and different counting geometries in order to cover the wide range of neutron yields (1015-1019 neutrons) while avoiding excessive count rates in the detectors. The absolute full energy peak efficiency calibration of the detectors was measured taking into account the source-detector geometry, the self-attenuation of the samples and cross-talk effects. An error analysis of the neutron yield measurement was performed including uncertainties in efficiency calibration, neutron transport calculations, cross sections, and counting statistics. Cross calibrations between the different irradiation ends were carried out in DD and DT (with 1% and 10% tritium content) discharges. The effect of the plasma vertical displacement was also experimentally studied. An agreement within 10% was found between the 14 MeV neutron yields measured from Si, Fe, Al, Nb samples in DT discharges.

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.

Nodal weighting factor method for ex-core fast neutron fluence evaluation

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

Chiang, R. T.

The nodal weighting factor method is developed for evaluating ex-core fast neutron flux in a nuclear reactor by utilizing adjoint neutron flux, a fictitious unit detector cross section for neutron energy above 1 or 0.1 MeV, the unit fission source, and relative assembly nodal powers. The method determines each nodal weighting factor for ex-core neutron fast flux evaluation by solving the steady-state adjoint neutron transport equation with a fictitious unit detector cross section for neutron energy above 1 or 0.1 MeV as the adjoint source, by integrating the unit fission source with a typical fission spectrum to the solved adjointmore » flux over all energies, all angles and given nodal volume, and by dividing it with the sum of all nodal weighting factors, which is a normalization factor. Then, the fast neutron flux can be obtained by summing the various relative nodal powers times the corresponding nodal weighting factors of the adjacent significantly contributed peripheral assembly nodes and times a proper fast neutron attenuation coefficient over an operating period. A generic set of nodal weighting factors can be used to evaluate neutron fluence at the same location for similar core design and fuel cycles, but the set of nodal weighting factors needs to be re-calibrated for a transition-fuel-cycle. This newly developed nodal weighting factor method should be a useful and simplified tool for evaluating fast neutron fluence at selected locations of interest in ex-core components of contemporary nuclear power reactors. (authors)« less

Methods for absorbing neutrons

DOEpatents

Guillen, Donna P [Idaho Falls, ID; Longhurst, Glen R [Idaho Falls, ID; Porter, Douglas L [Idaho Falls, ID; Parry, James R [Idaho Falls, ID

2012-07-24

A conduction cooled neutron absorber may include a metal matrix composite that comprises a metal having a thermal neutron cross-section of at least about 50 barns and a metal having a thermal conductivity of at least about 1 W/cmK. Apparatus for providing a neutron flux having a high fast-to-thermal neutron ratio may include a source of neutrons that produces fast neutrons and thermal neutrons. A neutron absorber positioned adjacent the neutron source absorbs at least some of the thermal neutrons so that a region adjacent the neutron absorber has a fast-to-thermal neutron ratio of at least about 15. A coolant in thermal contact with the neutron absorber removes heat from the neutron absorber.

Radioactivity in atomic-bomb samples from exposure to environmental neutrons.

PubMed

Endo, S; Shizuma, K; Tanaka, K; Ishikawa, M; Rühm, W; Egbert, S D; Hoshi, M

2007-12-01

For about one decade, activation measurements performed on environmental samples from a distance larger than 1 km from the hypocenter of the atomic-bomb explosion over Hiroshima suggested much higher thermal neutron fluences to the survivors than predicted. This caused concern among the radiation protection community and prompted a complete re-evaluation of all aspects of survivor dosimetry. While it was shown recently that secondary neutrons from cosmic radiation and other sources have probably been the reason for the high measured concentrations of the long-lived radioisotope 36Cl in these samples, the source for high measured concentrations of the short-lived radionuclides 152Eu and 60Co has not yet been investigated in detail. In order to quantify the production of 152Eu and 60Co in environmental samples by secondary neutrons from cosmic radiation, thermal neutron fluxes were measured by means of a He gas proportional counter in various buildings where these samples had been and still are being stored. Because a 252Cf neutron source has been operated occasionally close to one of the sample storage rooms, additional neutron flux measurements were carried out when the neutron source was in operation. The thermal neutron fluxes measured ranged from 0.00017 to 0.00093 n cm(-2) s(-1) and depended on the floor number of the investigated building. Based on the measured neutron fluxes, the specific activities from the reactions 151Eu(n,gamma)152Eu and 59Co(n,gamma)60Co in the atomic-bomb samples were estimated to be 7.9 mBq g(-1) Eu and 0.27 mBq g(-1) Co, respectively, in saturation. These activities are much lower than those recently measured in samples that had been exposed to atomic-bomb neutrons. It is therefore concluded that environmental and moderated 252Cf neutrons are not the source for the high activities that had been measured in these samples.

A two-fold reduction in measurement time for neutron assay: Initial tests of a prototype dual-gated shift register

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

Stewart, J.E.; Bourret, S.C.; Krick, M.S.

1996-09-01

Neutron coincidence counting (NCC) is used routinely around the world for nondestructive mass assay of uranium and plutonium in many forms, including waste. Compared with other methods, NCC is generally the most flexible, economic, and rapid. Many applications of NCC would benefit from a reduction in counting time required for a fixed random error. We have developed and tested the first prototype of a dual- gated, shift-register-based electronics unit that offers the potential of decreased measurement time for all passive and active NCC applications.

A two-fold reduction in measurement time for neutron assay: Initial tests of a prototype dual-gated shift register

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

Stewart, J.E.; Bourret, S.C.; Krick, M.S.

1996-12-31

Neutron coincidence counting (NCC) is used routinely around the world for nondestructive mass assay of uranium and plutonium in many forms, including waste. Compared with other methods, NCC is generally the most flexible, economic, and rapid. Many applications of NCC would benefit from a reduction in counting time required for a fixed random error. The authors have developed and tested the first prototype of a dual-gated, shift-register-based electronics unit that offers the potential of decreased measurement time for all passive and active NCC applications.

Neutron capture therapy with deep tissue penetration using capillary neutron focusing

DOEpatents

Peurrung, Anthony J.

1997-01-01

An improved method for delivering thermal neutrons to a subsurface cancer or tumor which has been first doped with a dopant having a high cross section for neutron capture. The improvement is the use of a guide tube in cooperation with a capillary neutron focusing apparatus, or neutron focusing lens, for directing neutrons to the tumor, and thereby avoiding damage to surrounding tissue.

Problems encountered in the use of neutron methods for elemental analysis on planetary surfaces

USGS Publications Warehouse

Senftle, F.; Philbin, P.; Moxham, R.; Boynton, G.; Trombka, J.

1974-01-01

From experimental studies of gamma rays from fast and thermal neutron reactions in hydrogeneous and non-hydrogeneous, semi-infinite samples and from Monte Carlo calculations on soil of a composition which might typically be encountered on planetary surfaces, it is found that gamma rays from fast or inelastic scattering reactions would dominate the observed spectra. With the exception of gamma rays formed by inelastically scattered neutrons on oxygen, useful spectra would be limited to energies below 3 MeV. Other experiments were performed which show that if a gamma-ray detector were placed within 6 m of an isotopic neutron source in a spacecraft, it would be rendered useless for gamma-ray spectrometry below 3 MeV because of internal activation produced by neutron exposure during space travel. Adequate shielding is not practicable because of the size and weight constraints for planetary missions. Thus, it is required that the source be turned off or removed to a safe distance during non-measurement periods. In view of these results an accelerator or an off-on isotopic source would be desirable for practical gamma-ray spectral analysis on planetary surfaces containing but minor amounts of hydrogen. ?? 1974.

Stability evaluation and correction of a pulsed neutron generator prompt gamma activation analysis system

USDA-ARS?s Scientific Manuscript database

Source output stability is important for accurate measurement in prompt gamma neutron activation. This is especially true when measuring low-concentration elements such as in vivo nitrogen (~2.5% of body weight). We evaluated the stability of the compact DT neutron generator within an in vivo nitrog...

Radiotherapeutic bandage based on electrospun polyacrylonitrile containing holmium-166 iron garnet nanoparticles for the treatment of skin cancer.

PubMed

Munaweera, Imalka; Levesque-Bishop, Daniel; Shi, Yi; Di Pasqua, Anthony J; Balkus, Kenneth J

2014-12-24

Radiation therapy is used as a primary treatment for inoperable tumors and in patients that cannot or will not undergo surgery. Radioactive holmium-166 ((166)Ho) is a viable candidate for use against skin cancer. Nonradioactive holmium-165 ((165)Ho) iron garnet nanoparticles have been incorporated into a bandage, which, after neutron-activation to (166)Ho, can be applied to a tumor lesion. The (165)Ho iron garnet nanoparticles ((165)HoIG) were synthesized and introduced into polyacrylonitrile (PAN) polymer solutions. The polymer solutions were then electrospun to produce flexible nonwoven bandages, which are stable to neutron-activation. The fiber mats were characterized using scanning electron microscopy, transmission electron microscopy, powder X-ray diffraction, Fourier transform infrared spectroscopy, thermogravimetric analysis and inductively coupled plasma mass spectrometry. The bandages are stable after neutron-activation at a thermal neutron-flux of approximately 3.5 × 10(12) neutrons/cm(2)·s for at least 4 h and 100 °C. Different amounts of radioactivity can be produced by changing the amount of the (165)HoIG nanoparticles inside the bandage and the duration of neutron-activation, which is important for different stages of skin cancer. Furthermore, the radioactive bandage can be easily manipulated to irradiate only the tumor site by cutting the bandage into specific shapes and sizes that cover the tumor prior to neutron-activation. Thus, exposure of healthy cells to high energy β-particles can be avoided. Moreover, there is no leakage of radioactive material after neutron activation, which is critical for safe handling by healthcare professionals treating skin cancer patients.

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.

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

Clarke, S. D.; Hamel, M. C.; Bourne, M. M.

Active interrogation creates an environment that is particularly challenging from a radiation-detection standpoint: the elevated background levels from the source can mask the desired signatures from the SNM. Neutron based interrogation experiments have shown that nanosecond-level timing is required to discriminate induced-fission neutrons from the scattered source neutrons. Previous experiments using high-energy bremsstrahlung X-rays have demonstrated the ability to induce and detect prompt photofission neutrons from single target materials; however, a real-world application would require spectroscopic capability to discern between photofission neutrons emitted by SNM and neutrons emitted by other reactions in non-SNM. Using digital pulseshape discrimination, organic liquid scintillatorsmore » are capable of reliably detecting neutrons in an intense gamma-ray field. Photon misclassification rates as low as 1 in 10 6 have been achieved, which is approaching the level of gaseous neutron detectors such as 3He without the need for neutron moderation. These scintillators also possess nanosecond-timing resolution, making them candidates for both neutron-and photon-driven active interrogation systems. Lastly, we have applied an array of liquid and NaI(Tl) scintillators to successfully image 13.7 kg of HEU interrogated by a DT neutron generator; the system was in the direct presence of the accelerator during the experiment.« less

A bismuth activation counter for high sensitivity pulsed 14 MeV neutrons

NASA Astrophysics Data System (ADS)

Burns, E. J. T.; Thacher, P. D.; Hassig, G. J.; Decker, R. D.; Romero, J. A.; Barrett, K. P.

2011-08-01

We have built a fast neutron bismuth activation counter that measures activation counts from pulsed 14-MeV neutron generators for incident neutron fluences between 30 and 300 neutrons/cm2 at 15.2 cm (6 in.). The activation counter consists of a large bismuth germanate (BGO) detector surrounded by a bismuth metal shield in front of and concentric with the cylindrical detector housing. The 14 MeV neutrons activate the 2.6-millisecond (ms) isomer in the shield and the detector by the reaction 209Bi (n,2nγ) 208mBi. The use of millisecond isomers and activation counting times minimizes the background from other activated materials and the environment. In addition to activation, the bismuth metal shields against other outside radiation sources. We have tested the bismuth activation counter, simultaneously, with two data acquisition systems (DASs) and both give similar results. The two-dimensional (2D) DAS and three dimensional (3D) DAS both consist of pulse height analysis (PHA) systems that can be used to discriminate against gamma radiations below 300 keV photon energy, so that the detector can be used strictly as a counter. If the counting time is restricted to less than 25 ms after the neutron pulse, there are less than 10 counts of background for single pulse operation in all our operational environments tested so far. High-fluence neutron generator operations are restricted by large dead times and pulse height saturation. When we operate our 3D DAS PHA system in list mode acquisition (LIST), real-time corrections to dead time or live time can be made on the scale of 1 ms time windows or dwell times. The live time correction is consistent with nonparalyzable models for dead time of 1.0±0.2 μs for our 3D DAS and 1.5±0.3 μs for our 2D DAS dominated by our fixed time width analog to digital converters (ADCs). With the same solid angle, we have shown that the bismuth activation counter has a factor of 4 increase in sensitivity over our lead activation counter, because of higher counts and negligible backgrounds.

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 Activated Samarium-153 Microparticles for Transarterial Radioembolization of Liver Tumour with Post-Procedure Imaging Capabilities

PubMed Central

Hashikin, Nurul Ab. Aziz; Yeong, Chai-Hong; Abdullah, Basri Johan Jeet; Ng, Kwan-Hoong; Chung, Lip-Yong; Dahalan, Rehir; Perkins, Alan Christopher

2015-01-01

Introduction Samarium-153 (153Sm) styrene divinylbenzene microparticles were developed as a surrogate for Yttrium-90 (90Y) microspheres in liver radioembolization therapy. Unlike the pure beta emitter 90Y, 153Sm possess both therapeutic beta and diagnostic gamma radiations, making it possible for post-procedure imaging following therapy. Methods The microparticles were prepared using commercially available cation exchange resin, Amberlite IR-120 H+ (620–830 μm), which were reduced to 20–40 μm via ball mill grinding and sieve separation. The microparticles were labelled with 152Sm via ion exchange process with 152SmCl3, prior to neutron activation to produce radioactive 153Sm through 152Sm(n,γ)153Sm reaction. Therapeutic activity of 3 GBq was referred based on the recommended activity used in 90Y-microspheres therapy. The samples were irradiated in 1.494 x 1012 n.cm-2.s-1 neutron flux for 6 h to achieve the nominal activity of 3.1 GBq.g-1. Physicochemical characterisation of the microparticles, gamma spectrometry, and in vitro radiolabelling studies were carried out to study the performance and stability of the microparticles. Results Fourier Transform Infrared (FTIR) spectroscopy of the Amberlite IR-120 resins showed unaffected functional groups, following size reduction of the beads. However, as shown by the electron microscope, the microparticles were irregular in shape. The radioactivity achieved after 6 h neutron activation was 3.104 ± 0.029 GBq. The specific activity per microparticle was 53.855 ± 0.503 Bq. Gamma spectrometry and elemental analysis showed no radioactive impurities in the samples. Radiolabelling efficiencies of 153Sm-Amberlite in distilled water and blood plasma over 48 h were excellent and higher than 95%. Conclusion The laboratory work revealed that the 153Sm-Amberlite microparticles demonstrated superior characteristics for potential use in hepatic radioembolization. PMID:26382059

Applications of neutron activation analysis in determination of natural and man-made radionuclides, including PA-231

NASA Astrophysics Data System (ADS)

Byrne, A. R.; Benedik, L.

1999-01-01

Neutron activation analysis (NAA), being essentially an isotopic and not an elemental method of analysis, is capable of determining a number of important radionuclides of radioecological interest by transformation into another, more easily quantifiable radionuclide. The nuclear characteristics which favour this technique may be summarized in an advantage factor relative to radiometric analysis of the original radioanalyte. Well known or hardly known examples include235U,238U,232Th,230Th,129I,99Tc,237Np and231Pa; a number of these are discussed and illustrated in analysis of real samples of environmental and biological origin. In particular, determination of231Pa by RNAA was performed using both postirradiation and preseparation methods. Application of INAA to enable the use of238U and232Th as endogenous (internal) radiotracers in alpha spectrometric analyses of uranium and thorium radioisotopes in radioecological studies is described, also allowing independent data sets to be obtained for quality control.

Dual Neutral Particle Beam Interrogation of Intermodal Shipping Containers for Special Nuclear Material

NASA Astrophysics Data System (ADS)

Keith, Rodney Lyman

Intermodal shipping containers entering the United States provide an avenue to smuggle unsecured or stolen special nuclear material (SNM). The only direct method fielded to indicate the presence of SNM is by passive photon/neutron radiation detection. Active interrogation using neutral particle beams to induce fission in SNM is a method under consideration. One by-product of fission is the creation of fragments that undergo radioactive decay over a time period on the order of tens of seconds after the initial event. The "delayed" gamma-rays emitted from these fragments over this period are considered a hallmark for the presence of SNM. A fundamental model is developed using homogenized cargos with a SNM target embedded at the center and computationally interrogated using simultaneous neutron and photon beams. Findings from analysis of the delayed gamma emissions from these experiments are intended to mitigate the effects of poor quality information about the composition and disposition of suspect cargo before examination in an active interrogation portal.

Reactor Dosimetry State of the Art 2008

NASA Astrophysics Data System (ADS)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G. Williams, A. P. Ribaric and T. Schnauber. Agile high-fidelity MCNP model development techniques for rapid mechanical design iteration / J. A. Kulesza.Extension of Raptor-M3G to r-8-z geometry for use in reactor dosimetry applications / M. A. Hunter, G. Longoni and S. L. Anderson. In vessel exposure distributions evaluated with MCNP5 for Atucha II / J. M. Longhino, H. Blaumann and G. Zamonsky. Atucha I nuclear power plant azimutal ex-vessel flux profile evaluation / J. M. Longhino ... [et al.]. UFTR thermal column characterization and redesign for maximized thermal flux / C. Polit and A. Haghighat. Activation counter using liquid light-guide for dosimetry of neutron burst / M. Hayashi ... [et al.]. Control rod reactivity curves for the annular core research reactor / K. R. DePriest ... [et al.]. Specification of irradiation conditions in VVER-440 surveillance positions / V. Kochkin ... [et al.]. Simulations of Mg-Ar ionisation and TE-TE ionisation chambers with MCNPX in a straightforward gamma and beta irradiation field / S. Nievaart ... [et al.]. The change of austenitic stainless steel elements content in the inner parts of VVER-440 reactor during operation / V. Smutný, J. Hep and P. Novosad. Fast neutron environmental spectrometry using disk activation / G. Lövestam ... [et al.]. Optimization of the neutron activation detector location scheme for VVER-lOOO ex-vessel dosimetry / V. N. Bukanov ... [et al.]. Irradiation conditions for surveillance specimens located into plane containers installed in the WWER-lOOO reactor of unit 2 of the South-Ukrainian NPP / O. V. Grytsenko. V. N. Bukanov and S. M. Pugach. Conformity between LRO mock-ups and VVERS NPP RPV neutron flux attenuation / S. Belousov. Kr. Ilieva and D. Kirilova. FLUOLE: a new relevant experiment for PWR pressure vessel surveillance / D. Beretz ... [et al.]. Transport of neutrons and photons through the iron and water layers / M. J. Kost'ál ... [et al.]. Condition evaluation of spent nuclear fuel assemblies from the first-generation nuclear-powered submarines by gamma scanning / A. F. Usatyi. L. A. Serdyukova and B. S. Stepennov -- Oral session 3: Power plant surveillance. Upgraded neutron dosimetry procedure for VVER-440 surveillance specimens / V. Kochkin ... [et al.]. Neutron dosimetry on the full-core first generation VVER-440 aimed to reactor support structure load evaluation / P. Borodkin ... [et al.]. Ex-vessel neutron dosimetry programs for PWRs in Korea / C. S. Yoo. B. C. Kim and C. C. Kim. Comparison of irradiation conditions of VVER-1000 reactor pressure vessel and surveillance specimens for various core loadings / V. N. Bukanov ... [et al.]. Re-evaluation of dosimetry in the new surveillance program for the Loviisa 1 VVER-440 reactor / T. Serén -- Oral session 4: Benchmarks, intercomparisons and adjustment methods. Determination of the neutron parameter's uncertainties using the stochastic methods of uncertainty propagation and analysis / G. Grégoire ... [et al.].Covariance matrices for calculated neutron spectra and measured dosimeter responses / J. G. Williams ... [et al.]. The role of dosimetry at the high flux reactor / S. C. van der Marek ... [et al.]. Calibration of a manganese bath relative to Cf-252 nu-bar / D. M. Gilliam, A. T. Yue and M. Scott Dewey. Major upgrade of the reactor dosimetry interpretation methodology used at the CEA: general principle / C. Destouches ... [et al.] -- Oral session 5: power plant surveillance. The role of ex-vessel neutron dosimetry in reactor vessel surveillance in South Korea / B.-C. Kim ... [et al.]. Spanish RPV surveillance programmes: lessons learned and current activities / A. Ballesteros and X. Jardí. Atucha I nuclear power plant extended dosimetry and assessment / H. Blaumann ... [et al.]. Monitoring of radiation load of pressure vessels of Russian VVER in compliance with license amendments / G. Borodkin ... [et al.] -- Poster session 2: Test reactors, accelerators and advanced systems; cross sections, nuclear data, damage correlations. Two-dimensional mapping of the calculated fission power for the full-size fuel plate experiment irradiated in the advanced test reactor / G. S. Chang and M. A. Lillo. The radiation safety information computational center: a resource for reactor dosimetry software and nuclear data / B. L. Kirk. Irradiated xenon isotopic ratio measurement for failed fuel detection and location in fast reactor / C. Ito, T. Iguchi and H. Harano. Characterization of dosimetry of the BMRR horizontal thimble tubes and broad beam facility / J.-P. Hu, R. N. Reciniello and N. E. Holden. 2007 nuclear data review / N. E. Holden. Further dosimetry studies at the Rhode Island nuclear science / R. N. Reciniello ... [et al.]. Characterization of neutron fields in the experimental fast reactor Joyo MK-III core / S. Maeda ... [et al.]. Measuring [symbol]Li(n, t) and [symbol]B(n, [symbol]) cross sections using the NIST alpha-gamma apparatus / M. S. Dewey ... [et al.]. Improvement of neutron/gamma field evaluation for restart of JMTR / Y. Nagao ... [et al.]. Monitoring of the irradiated neutron fluence in the neutron transmutation doping process of HANARO / M.-S. Kim and S.-J. Park.Training reactor VR-l neutron spectrum determination / M. Vins, A. Kolros and K. Katovsky. Differential cross sections for gamma-ray production by 14 MeV neutrons on iron and bismuth / V. M. Bondar ... [et al.]. The measurements of the differential elastic neutron cross-sections of carbon for energies from 2 to 133 ke V / O. Gritzay ... [et al.]. Determination of neutron spectrum by the dosimetry foil method up to 35 Me V / S. P. Simakov ... [et al.]. Extension of the BGL broad group cross section library / D. Kirilova, S. Belousov and Kr. Ilieva. Measurements of neutron capture cross-section for tantalum at the neutron filtered beams / O. Gritzayand V. Libman. Measurements of microscopic data at GELINA in support of dosimetry / S. Kopecky ... [et al.]. Nuclide guide and international chart of nuclides - 2008 / T. Golashvili -- Oral session 6: Test reactors, accelerators and advanced systems. Neutronic analyses in support of the HFIR beamline modifications and lifetime extension / I. Remec and E. D. Blakeman. Characterization of neutron test facilities at Sandia National Laboratories / D. W. Vehar ... [et al.]. LYRA irradiation experiments: neutron metrology and dosimetry / B. Acosta and L. Debarberis. Calculated neutron and gamma-ray spectra across the prismatic very high temperature reactor core / J. W. Sterbentz. Enhancement of irradiation capability of the experimental fast reactor joyo / S. Maeda ... [et al.]. Neutron spectrum analyses by foil activation method for high-energy proton beams / C. H. Pyeon ... [et al.] -- Oral session 7: Cross sections, nuclear data, damage correlations. Investigation of new reaction cross-section evaluations in order to update and extend the IRDF-2002 reactor dosimetry library / É. M. Zsolnay, H. J. Nolthenius and A. L. Nichols. A novel approach towards DPA calculations / A. Hogenbirk and D. F. Da Cruz. A new ENDFIB-VII.O based multigroup cross-section library for reactor dosimetry / F. A. Alpan and S. L. Anderson. Activities at the NEA for dosimetry applications / H. Henriksson and I. Kodeli. Validation and verification of covariance data from dosimetry reaction cross-section evaluations / S. Badikov. Status of the neutron cross section standards / A. D. Carlson -- Oral session 8: transport calculations. A dosimetry assessment for the core restraint of an advanced gas cooled reactor / D. A. Thornton ... [et al.]. Neutron dosimetry study in the region of the support structure of a VVER-1000 type reactor / G. Borodkin ... [et al.]. SNS moderator poison design and experiment validation of the moderator performance / W. Lu ... [et al.]. Analysis of OSIRIS in-core surveillance dosimetry for GONDOLE steel irradiation program by using TRIPOLI-4 Monte Carlo code / Y. K. Lee and F. Malouch.Reactor dosimetry applications using RAPTOR-M3G: a new parallel 3-D radiation transport code / G. Longoni and S. L. Anderson.

Automated lithology prediction from PGNAA and other geophysical logs.

PubMed

Borsaru, M; Zhou, B; Aizawa, T; Karashima, H; Hashimoto, T

2006-02-01

Different methods of lithology predictions from geophysical data have been developed in the last 15 years. The geophysical logs used for predicting lithology are the conventional logs: sonic, neutron-neutron, gamma (total natural-gamma) and density (backscattered gamma-gamma). The prompt gamma neutron activation analysis (PGNAA) is another established geophysical logging technique for in situ element analysis of rocks in boreholes. The work described in this paper was carried out to investigate the application of PGNAA to the lithology interpretation. The data interpretation was conducted using the automatic interpretation program LogTrans based on statistical analysis. Limited test suggests that PGNAA logging data can be used to predict the lithology. A success rate of 73% for lithology prediction was achieved from PGNAA logging data only. It can also be used in conjunction with the conventional geophysical logs to enhance the lithology prediction.

The MCNP Simulation of a PGNAA System at TRR-1/M1

NASA Astrophysics Data System (ADS)

Sangaroon, S.; Ratanatongchai, W.; Picha, R.; Khaweerat, S.; Channuie, J.

2017-06-01

The prompt-gamma neutron activation analysis system (PGNAA) has been installed at Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1999. The purpose of the system is for elemental and isotopic analyses. The system mainly consists of a series of the moderator and collimator, neutron and gamma-ray shielding and the HPGe detector. In this work, the condition of the system is carried out based on the Monte Carlo method using Monte Carlo N-Particle transport code and the experiment. The flux ratios (Φthermal/Φepithermal and Φthermal/Φfast) and thermal neutron flux have been obtained. The simulated prompt gamma rays of the Portland cement sample have been carried out. The simulation provides significant contribution in upgrading the PGNAA station to be available in various applications.

REACTOR MONITORING

DOEpatents

Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

1959-02-01

A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

Correlation of the neutron yield from the plasma focus upon variations in the magnetic field energy of the discharge circuit

NASA Astrophysics Data System (ADS)

Ablesimov, V. E.; Dolin, Yu. N.; Kalinychev, A. E.; Tsibikov, Z. S.

2017-10-01

The relation between neutron yield Y and magnetic field energy variations Δ W in the discharge circuit has been studied for a Mather-type plasma-focus camera. The activation technique (activation of silver isotopes) has been used to measure the integral yield of DD neutrons from the source. The time dependence of the neutron yield has been recorded by scintillation detectors. For the device used in the investigations, the neutron yield exhibits a linear dependence on variations in the magnetic field energy Δ W in the discharge circuit at the instant of neutron generation. It is also found that this dependence is related to the initial deuteron pressure in the discharge chamber.

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

Method and system for detecting explosives

DOEpatents

Reber, Edward L [Idaho Falls, ID; Jewell, James K [Idaho Falls, ID; Rohde, Kenneth W [Idaho Falls, ID; Seabury, Edward H [Idaho Falls, ID; Blackwood, Larry G [Idaho Falls, ID; Edwards, Andrew J [Idaho Falls, ID; Derr, Kurt W [Idaho Falls, ID

2009-03-10

A method of detecting explosives in a vehicle includes providing a first rack on one side of the vehicle, the rack including a neutron generator and a plurality of gamma ray detectors; providing a second rack on another side of the vehicle, the second rack including a neutron generator and a plurality of gamma ray detectors; providing a control system, remote from the first and second racks, coupled to the neutron generators and gamma ray detectors; using the control system, causing the neutron generators to generate neutrons; and performing gamma ray spectroscopy on spectra read by the gamma ray detectors to look for a signature indicative of presence of an explosive. Various apparatus and other methods are also provided.

Explosives detection system and method

DOEpatents

Reber, Edward L.; Jewell, James K.; Rohde, Kenneth W.; Seabury, Edward H.; Blackwood, Larry G.; Edwards, Andrew J.; Derr, Kurt W.

2007-12-11

A method of detecting explosives in a vehicle includes providing a first rack on one side of the vehicle, the rack including a neutron generator and a plurality of gamma ray detectors; providing a second rack on another side of the vehicle, the second rack including a neutron generator and a plurality of gamma ray detectors; providing a control system, remote from the first and second racks, coupled to the neutron generators and gamma ray detectors; using the control system, causing the neutron generators to generate neutrons; and performing gamma ray spectroscopy on spectra read by the gamma ray detectors to look for a signature indicative of presence of an explosive. Various apparatus and other methods are also provided.

Improved neutron-gamma discrimination for a 3He neutron detector using subspace learning methods

DOE PAGES

Wang, C. L.; Funk, L. L.; Riedel, R. A.; ...

2017-02-10

3He gas based neutron linear-position-sensitive detectors (LPSDs) have been applied for many neutron scattering instruments. Traditional Pulse-Height Analysis (PHA) for Neutron-Gamma Discrimination (NGD) resulted in the neutron-gamma efficiency ratio on the orders of 10 5-10 6. The NGD ratios of 3He detectors need to be improved for even better scientific results from neutron scattering. Digital Signal Processing (DSP) analyses of waveforms were proposed for obtaining better NGD ratios, based on features extracted from rise-time, pulse amplitude, charge integration, a simplified Wiener filter, and the cross-correlation between individual and template waveforms of neutron and gamma events. Fisher linear discriminant analysis (FLDA)more » and three multivariate analyses (MVAs) of the features were performed. The NGD ratios are improved by about 10 2-10 3 times compared with the traditional PHA method. Finally, our results indicate the NGD capabilities of 3He tube detectors can be significantly improved with subspace-learning based methods, which may result in a reduced data-collection time and better data quality for further data reduction.« less

ANITA-IEAF activation code package - updating of the decay and cross section data libraries and validation on the experimental data from the Karlsruhe Isochronous Cyclotron

NASA Astrophysics Data System (ADS)

Frisoni, Manuela

2017-09-01

ANITA-IEAF is an activation package (code and libraries) developed in the past in ENEA-Bologna in order to assess the activation of materials exposed to neutrons with energies greater than 20 MeV. An updated version of the ANITA-IEAF activation code package has been developed. It is suitable to be applied to the study of the irradiation effects on materials in facilities like the International Fusion Materials Irradiation Facility (IFMIF) and the DEMO Oriented Neutron Source (DONES), in which a considerable amount of neutrons with energies above 20 MeV is produced. The present paper summarizes the main characteristics of the updated version of ANITA-IEAF, able to use decay and cross section data based on more recent evaluated nuclear data libraries, i.e. the JEFF-3.1.1 Radioactive Decay Data Library and the EAF-2010 neutron activation cross section library. In this paper the validation effort related to the comparison between the code predictions and the activity measurements obtained from the Karlsruhe Isochronous Cyclotron is presented. In this integral experiment samples of two different steels, SS-316 and F82H, pure vanadium and a vanadium alloy, structural materials of interest in fusion technology, were activated in a neutron spectrum similar to the IFMIF neutron field.

EDITORIAL: Instrumentation and Methods for Neutron Scattering—papers from the 4th European Conference on Neutron Scattering in Lund, Sweden, June 2007

NASA Astrophysics Data System (ADS)

Rennie, Adrian R.

2008-03-01

Neutron scattering is used as a tool to study problems in disciplines that include chemistry, materials science, biology and condensed matter physics as well as problems from neighbouring disciplines such as geology, environmental sciences and archaeology. Equipment for these studies is found at laboratories with research reactors or spallation neutron sources and there are many recent or current developments with new instruments and even entirely new facilities such as the Spallation Neutron Source at Oak Ridge, USA, the OPAL reactor at Lucas Heights, Australia and the second target station at the ISIS facility in the UK. Design and optimization of the instruments at these facilities involves work with many research laboratories and groups in universities. Every four years the European Conference on Neutron Scattering (ECNS) brings together both the specialists in neutron instrumentation and the community of users (in intervening years there are International and American conferences). In June 2007 about 700 delegates came to the 4th ECNS that was held in Lund, Sweden. There were more than 600 presentations as talks and posters. The opportunity to publish papers in Measurement Science and Technology that relate to neutron scattering instrumentation and method development was offered to the participants, and the papers that follow describe some of the recent activity in this field. Accounts of work on condensed matter science and the applications of neutron scattering appear separately in Journal of Physics: Condensed Matter. There are, of course, many features of neutron instrumentation that are specific to this particular field of measurement. However, there are also many elements of apparatus and experiment design that can usefully be shared with a broader community. It is hoped that this issue with papers from ECNS will find a broad community of interest. Apart from descriptions of overall design of diffractometers and spectrometers there are accounts of new components such as detectors, polarizers, focusing devices and sample environments. The organizers and participants are extremely grateful to numerous sponsors that helped to make the conference a great success. An equal debt of gratitude is due to the Institute of Physics and the editorial and publishing staff for agreeing to publish these papers and organizing the refereeing and editorial process efficiently and promptly in a friendly way.

Establishment of gold-quartz standard GQS-1

USGS Publications Warehouse

Millard, Hugh T.; Marinenko, John; McLane, John E.

1969-01-01

A homogeneous gold-quartz standard, GQS-1, was prepared from a heterogeneous gold-bearing quartz by chemical treatment. The concentration of gold in GQS-1 was determined by both instrumental neutron activation analysis and radioisotope dilution analysis to be 2.61?0.10 parts per million. Analysis of 10 samples of the standard by both instrumental neutron activation analysis and radioisotope dilution analysis failed to reveal heterogeneity within the standard. The precision of the analytical methods, expressed as standard error, was approximately 0.1 part per million. The analytical data were also used to estimate the average size of gold particles. The chemical treatment apparently reduced the average diameter of the gold particles by at least an order of magnitude and increased the concentration of gold grains by a factor of at least 4,000.

Neutron capture therapy with deep tissue penetration using capillary neutron focusing

DOEpatents

Peurrung, A.J.

1997-08-19

An improved method is disclosed for delivering thermal neutrons to a subsurface cancer or tumor which has been first doped with a dopant having a high cross section for neutron capture. The improvement is the use of a guide tube in cooperation with a capillary neutron focusing apparatus, or neutron focusing lens, for directing neutrons to the tumor, and thereby avoiding damage to surrounding tissue. 1 fig.

Neutron Detector Signal Processing to Calculate the Effective Neutron Multiplication Factor of Subcritical Assemblies

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

Talamo, Alberto; Gohar, Yousry

2016-06-01

This report describes different methodologies to calculate the effective neutron multiplication factor of subcritical assemblies by processing the neutron detector signals using MATLAB scripts. The subcritical assembly can be driven either by a spontaneous fission neutron source (e.g. californium) or by a neutron source generated from the interactions of accelerated particles with target materials. In the latter case, when the particle accelerator operates in a pulsed mode, the signals are typically stored into two files. One file contains the time when neutron reactions occur and the other contains the times when the neutron pulses start. In both files, the timemore » is given by an integer representing the number of time bins since the start of the counting. These signal files are used to construct the neutron count distribution from a single neutron pulse. The built-in functions of MATLAB are used to calculate the effective neutron multiplication factor through the application of the prompt decay fitting or the area method to the neutron count distribution. If the subcritical assembly is driven by a spontaneous fission neutron source, then the effective multiplication factor can be evaluated either using the prompt neutron decay constant obtained from Rossi or Feynman distributions or the Modified Source Multiplication (MSM) method.« less

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

Couture, Aaron Joseph

This report documents aspects of direct and indirect neutron capture. The importance of neutron capture rates and methods to determine them are presented. The following conclusions are drawn: direct neutron capture measurements remain a backbone of experimental study; work is being done to take increased advantage of indirect methods for neutron capture; both instrumentation and facilities are making new measurements possible; more work is needed on the nuclear theory side to understand what is needed furthest from stability.

Variable control of neutron albedo in toroidal fusion devices

DOEpatents

Jassby, D.L.; Micklich, B.J.

1983-06-01

This invention pertains to methods of controlling in the steady state, neutron albedo in toroidal fusion devices, and in particular, to methods of controlling the flux and energy distribution of collided neutrons which are incident on an outboard wall of a toroidal fusion device.

Organic metal neutron detector

DOEpatents

Butler, M.A.; Ginley, D.S.

1984-11-21

A device for detection of neutrons comprises: as an active neutron sensing element, a conductive organic polymer having an electrical conductivity and a cross-section for said neutrons whereby a detectable change in said conductivity is caused by impingement of said neutrons on the conductive organic polymer which is responsive to a property of said polymer which is altered by impingement of said neutrons on the polymer; and means for associating a change in said alterable property with the presence of neutrons at the location of said device.

A Dosimetry Study of Deuterium-Deuterium Neutron Generator-based In Vivo Neutron Activation Analysis.

PubMed

Sowers, Daniel; Liu, Yingzi; Mostafaei, Farshad; Blake, Scott; Nie, Linda H

2015-12-01

A neutron irradiation cavity for in vivo neutron activation analysis (IVNAA) to detect manganese, aluminum, and other potentially toxic elements in human hand bone has been designed and its dosimetric specifications measured. The neutron source is a customized deuterium-deuterium neutron generator that produces neutrons at 2.45 MeV by the fusion reaction 2H(d, n)3He at a calculated flux of 7 × 10(8) ± 30% s(-1). A moderator/reflector/shielding [5 cm high density polyethylene (HDPE), 5.3 cm graphite and 5.7 cm borated (HDPE)] assembly has been designed and built to maximize the thermal neutron flux inside the hand irradiation cavity and to reduce the extremity dose and effective dose to the human subject. Lead sheets are used to attenuate bremsstrahlung x rays and activation gammas. A Monte Carlo simulation (MCNP6) was used to model the system and calculate extremity dose. The extremity dose was measured with neutron and photon sensitive film badges and Fuji electronic pocket dosimeters (EPD). The neutron ambient dose outside the shielding was measured by Fuji NSN3, and the photon dose was measured by a Bicron MicroREM scintillator. Neutron extremity dose was calculated to be 32.3 mSv using MCNP6 simulations given a 10-min IVNAA measurement of manganese. Measurements by EPD and film badge indicate hand dose to be 31.7 ± 0.8 mSv for neutrons and 4.2 ± 0.2 mSv for photons for 10 min; whole body effective dose was calculated conservatively to be 0.052 mSv. Experimental values closely match values obtained from MCNP6 simulations. These are acceptable doses to apply the technology for a manganese toxicity study in a human population.

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

Klix, A.; Fischer, U.; Raj, P.

Fusion power reactors will rely on the internal production of the fuel tritium from lithium in the tritium breeding blanket. Test Blanket Modules (TBM) will be installed in ITER with the aim to investigate the nuclear performance of different breeding blanket designs. Currently there is no fully qualified nuclear instrumentation available for the measurement of neutron fluxes and tritium production rates which would be able to withstand the harsh environment conditions in the TBM such as high temperature (>400 deg. C) and, depending on the operation scenario, intense radiation levels. As partner of the European Consortium on Nuclear Data andmore » Measurement Techniques in the framework of several F4E specific grants and contracts, KIT and ENEA have jointly studied the possibility to develop and test detectors suitable to operate in ITER-TBMs. Here we present an overview of ongoing work on three types of neutron flux monitors under development for the TBMs with focus on the KIT activities. A neutron activation system (NAS) with pneumatic sample transport could provide absolute neutron flux measurements in selected positions. A test system for investigating activation materials with short half-lives was constructed at the DT neutron generator laboratory of Technical University of Dresden to investigate the neutronics aspects. Several irradiations have been performed with focus on the simultaneous measurement of the extracted activated probes. An engineering assessment of a TBM NAS in the conceptual design phase has been done which considered issues of design requirements and integration. Last but not least, a mechanical test bench is under construction at KIT which will address issues of driving the activation probes, solutions for loading the system etc. experimentally. Self-powered neutron detectors (SPND) are widely applied in fission reactor monitoring, and the commercially available SPNDs are sensitive to thermal neutrons. We are investigating novel materials for SPND which would be sensitive also to the fast neutron flux expected in the TBMs. To this end simulations were done with the European Activation System EASY and neutron flux spectra which were calculated with MCNP for the HCPB TBM. Preliminary tests with commercial SPND in a fast reactor were performed. As a result of these activities, several materials have been found which may be suitable for the measurement of fast neutron fluxes in the TBM. Test detectors are under preparation for testing with DT neutron generators. Within the I{sub S}MART project, funded by KIC InnoEnergy, KIT is developing an online detector based on silicon carbide electronics for the TBMs. The operation of such detectors at TBM relevant temperatures is expected to incur lower accumulated radiation damage to them than at room temperature due to annealing effects. Detectors of several designs have been already irradiated with DT neutrons. Irradiation tests at elevated temperatures have been done and further tests are currently underway. This paper summarizes the status of the work for these three neutron flux monitor systems. (authors)« less

Monte-Carlo gamma response simulation of fast/thermal neutron interactions with soil elements

USDA-ARS?s Scientific Manuscript database

Soil elemental analysis using characteristic gamma rays induced by neutrons is an effective method of in situ soil content determination. The nuclei of soil elements irradiated by neutrons issue characteristic gamma rays due to both inelastic neutron scattering (e.g., Si, C) and thermal neutron capt...

Activation analysis study on Li-ion batteries for nuclear forensic applications

NASA Astrophysics Data System (ADS)

Johnson, Erik B.; Whitney, Chad; Holbert, Keith E.; Zhang, Taipeng; Stannard, Tyler; Christie, Anthony; Harper, Peter; Anderson, Blake; Christian, James F.

2015-06-01

The nuclear materials environment has been increasing significantly in complexity over the past couple of decades. The prevention of attacks from nuclear weapons is becoming more difficult, and nuclear forensics is a deterrent by providing detailed information on any type of nuclear event for proper attribution. One component of the nuclear forensic analysis is a measurement of the neutron spectrum. As an example, the neutron component provides information on the composition of the weapons, whether boosting is involved or the mechanisms used in creating a supercritical state. As 6Li has a large cross-section for thermal neutrons, the lithium battery is a primary candidate for assessing the neutron spectrum after detonation. The absorption process for 6Li yields tritium, which can be measured at a later point after the nuclear event, as long as the battery can be processed in a manner to successfully extract the tritium content. In addition, measuring the activated constituents after exposure provides a means to reconstruct the incident neutron spectrum. The battery consists of a spiral or folded layers of material that have unique, energy dependent interactions associated with the incident neutron flux. A detailed analysis on the batteries included a pre-irradiated mass spectrometry analysis to be used as input for neutron spectrum reconstruction. A set of batteries were exposed to a hard neutron spectrum delivered by the University of Massachusetts, Lowell research reactor Fast Neutron Irradiator (FNI). The gamma spectra were measured from the batteries within a few days and within a week after the exposure to obtain sufficient data on the activated materials in the batteries. The activity was calculated for a number of select isotopes, indicating the number of associated neutron interactions. The results from tritium extraction are marginal. A measurable increase in detected particles (gammas and betas) below 50 keV not self-attenuated by the battery was observed, yet as the spectra are coarse, the gamma information is not separable from tritium spectra. The activation analysis was successful, and the incident neutron spectrum was reconstructed using materials found in lithium batteries.

A prestorage method to measure neutron transmission of ultracold neutron guides

NASA Astrophysics Data System (ADS)

Blau, B.; Daum, M.; Fertl, M.; Geltenbort, P.; Göltl, L.; Henneck, R.; Kirch, K.; Knecht, A.; Lauss, B.; Schmidt-Wellenburg, P.; Zsigmond, G.

2016-01-01

There are worldwide efforts to search for physics beyond the Standard Model of particle physics. Precision experiments using ultracold neutrons (UCN) require very high intensities of UCN. Efficient transport of UCN from the production volume to the experiment is therefore of great importance. We have developed a method using prestored UCN in order to quantify UCN transmission in tubular guides. This method simulates the final installation at the Paul Scherrer Institute's UCN source where neutrons are stored in an intermediate storage vessel serving three experimental ports. This method allowed us to qualify UCN guides for their intended use and compare their properties.

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

WINDOWS: a program for the analysis of spectral data foil activation measurements

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

Stallmann, F.W.; Eastham, J.F.; Kam, F.B.K.

The computer program WINDOWS together with its subroutines is described for the analysis of neutron spectral data foil activation measurements. In particular, the unfolding of the neutron differential spectrum, estimated windows and detector contributions, upper and lower bounds for an integral response, and group fluxes obtained from neutron transport calculations. 116 references. (JFP)

Neutron-Activated Gamma-Emission: Technology Review

DTIC Science & Technology

2012-01-01

valid OMB control number. PLEASE DO NOT RETURN YOUR FORM TO THE ABOVE ADDRESS. 1. REPORT DATE (DD-MM-YYYY) January 2012 2. REPORT TYPE Progress 3... DATES COVERED (From - To) January to March 2010 4. TITLE AND SUBTITLE Neutron-Activated Gamma-Emission: Technology Review 5a. CONTRACT NUMBER...Backscatter Analysis Techniques........................................................................13 3. Sources of Neutrons 15 3.1 Radioisotope

Neutron track length estimator for GATE Monte Carlo dose calculation in radiotherapy.

PubMed

Elazhar, H; Deschler, T; Létang, J M; Nourreddine, A; Arbor, N

2018-06-20

The out-of-field dose in radiation therapy is a growing concern in regards to the late side-effects and secondary cancer induction. In high-energy x-ray therapy, the secondary neutrons generated through photonuclear reactions in the accelerator are part of this secondary dose. The neutron dose is currently not estimated by the treatment planning system while it appears to be preponderant for distances greater than 50 cm from the isocenter. Monte Carlo simulation has become the gold standard for accurately calculating the neutron dose under specific treatment conditions but the method is also known for having a slow statistical convergence, which makes it difficult to be used on a clinical basis. The neutron track length estimator, a neutron variance reduction technique inspired by the track length estimator method has thus been developped for the first time in the Monte Carlo code GATE to allow a fast computation of the neutron dose in radiotherapy. The details of its implementation, as well as the comparison of its performances against the analog MC method, are presented here. A gain of time from 15 to 400 can be obtained by our method, with a mean difference in the dose calculation of about 1% in comparison with the analog MC method.

Double difference method in deep inelastic neutron scattering on the VESUVIO spectrometer

NASA Astrophysics Data System (ADS)

Andreani, C.; Colognesi, D.; Degiorgi, E.; Filabozzi, A.; Nardone, M.; Pace, E.; Pietropaolo, A.; Senesi, R.

2003-02-01

The principles of the Double Difference (DD) method, applied to the neutron spectrometer VESUVIO, are discussed. VESUVIO, an inverse geometry spectrometer operating at the ISIS pulsed neutron source in the eV energy region, has been specifically designed to measure the single particle dynamical properties in condensed matter. The width of the nuclear resonance of the absorbing filter, used for the neutron energy analysis, provides the most important contribution to the energy resolution of the inverse geometry instruments. In this paper, the DD method, which is based on a linear combination of two measurements recorded with filter foils of the same resonance material but of different thickness, is shown to improve significantly the instrumental energy resolution, as compared with the Single Difference (SD) method. The asymptotic response functions, derived through Monte-Carlo simulations for polycrystalline Pb and ZrH 2 samples, are analysed in both DD and SD methods, and compared with the experimental ones for Pb sample. The response functions have been modelled for two distinct experimental configurations of the VESUVIO spectrometer, employing 6Li-glass neutron detectors and NaI γ detectors revealing the γ-ray cascade from the ( n,γ) reaction, respectively. The DD method appears to be an effective experimental procedure for Deep Inelastic Neutron Scattering measurements on VESUVIO spectrometer, since it reduces the experimental resolution of the instrument in both 6Li-glass neutron detector and γ detector configurations.

CHARACTERIZATION OF A THIN SILICON SENSOR FOR ACTIVE NEUTRON PERSONAL DOSEMETERS.

PubMed

Takada, M; Nunomiya, T; Nakamura, T; Matsumoto, T; Masuda, A

2016-09-01

A thin silicon sensor has been developed for active neutron personal dosemeters for use by aircrews and first responders. This thin silicon sensor is not affected by the funneling effect, which causes detection of cosmic protons and over-response to cosmic neutrons. There are several advantages to the thin silicon sensor: a decrease in sensitivity to gamma rays, an improvement of the energy detection limit for neutrons down to 0.8 MeV and an increase in the sensitivity to fast neutrons. Neutron response functions were experimentally obtained using 2.5 and 5 MeV monoenergy neutron beams and a (252)Cf neutron source. Simulation results using the Monte Carlo N-Particle transport code agree quite well with the experimental ones when an energy deposition region shaped like a circular truncated cone is used in place of a cylindrical region. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Technical basis for the use of a correlated neutron source in the uranium neutron coincidence collar

DOE PAGES

Root, Margaret A.; Menlove, Howard Olsen; Lanza, Richard C.; ...

2017-01-16

Active neutron coincidence systems are commonly used by international inspectorates to verify a material balance across the various stages of the nuclear fuel cycle. The Uranium Neutron Coincidence Collar (UNCL) is one such instrument; it is used to measure the linear density of 235U (g 235U/cm of active length in assembly) in fresh light water reactor fuel in nuclear fuel fabrication facilities. The UNCL and other active neutron interrogation detectors have historically relied on americium lithium ( 241AmLi) sources to induce fission within the sample in question. Californium-252 is under consideration as a possible alternative to the traditional 241AmLi source.more » Finally, this work relied upon a combination of experiments and Monte Carlo simulations to demonstrate the technical basis for the replacement of 241AmLi sources with 252Cf sources by evaluating the statistical uncertainty in the measurements incurred by each source and assessing the penetrability of neutrons from each source for the UNCL.« less

Technical basis for the use of a correlated neutron source in the uranium neutron coincidence collar

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

Root, Margaret A.; Menlove, Howard Olsen; Lanza, Richard C.

Active neutron coincidence systems are commonly used by international inspectorates to verify a material balance across the various stages of the nuclear fuel cycle. The Uranium Neutron Coincidence Collar (UNCL) is one such instrument; it is used to measure the linear density of 235U (g 235U/cm of active length in assembly) in fresh light water reactor fuel in nuclear fuel fabrication facilities. The UNCL and other active neutron interrogation detectors have historically relied on americium lithium ( 241AmLi) sources to induce fission within the sample in question. Californium-252 is under consideration as a possible alternative to the traditional 241AmLi source.more » Finally, this work relied upon a combination of experiments and Monte Carlo simulations to demonstrate the technical basis for the replacement of 241AmLi sources with 252Cf sources by evaluating the statistical uncertainty in the measurements incurred by each source and assessing the penetrability of neutrons from each source for the UNCL.« less

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.

TWO NEW SINGLE-EXPOSURE, MULTI-DETECTOR NEUTRON SPECTROMETERS FOR RADIATION PROTECTION APPLICATIONS IN WORKPLACE MONITORING.

PubMed

Gómez-Ros, J M; Bedogni, R; Bortot, D; Domingo, C; Esposito, A; Introini, M V; Lorenzoli, M; Mazzitelli, G; Moraleda, M; Pola, A; Sacco, D

2017-04-01

This communication describes two new instruments, based on multiple active thermal neutron detectors arranged within a single moderator, that permit to unfold the neutron spectrum (from thermal to hundreds of MeV) and to determine the corresponding integral quantities with only one exposure. This makes them especially advantageous for neutron field characterisation and workplace monitoring in neutron-producing facilities. One of the devices has spherical geometry and nearly isotropic response, the other one has cylindrical symmetry and it is only sensitive to neutrons incident along the cylinder axis. In both cases, active detectors have been specifically developed looking for the criteria of miniaturisation, high sensitivity, linear response and good photon rejection. The calculated response matrix has been validated by experimental irradiations in neutron reference fields with a global uncertainty of 3%. The measurements performed in realistic neutron fields permitted to determine the neutron spectra and the integral quantities, in particular H*(10). © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Shape-Independent Model of Monitor Neutron Activation Analysis

NASA Astrophysics Data System (ADS)

Yusuf, Siaka Ojo

The technique of monitor neutron activation analysis has been improved by developing a shape-independent model to solve the problem of the treatment of the epithermal reaction contribution to the reaction rate in reactor neutron activation analysis. It is a form of facility characterization in which differential approximations to neither the neutron flux distribution as a function of energy nor the reaction cross section as a function of energy are necessary. The model predicts a linear relationship when the k-factors (ratios of reaction rates of two nuclides at a given irradiation position) for element x, k _{c} (x), is plotted against the k-factor for the monitor, k_{c} (m). The slope of this line, B(x,c,m) is measured for each element x to provide the calibration of the irradiation facility for monitor activation analysis. In this thesis, scandium was chosen as the comparator and antimony as the epithermal monitor. B(x, Sc, Sb) has been accurately measured for a number of nuclides in three different reactors. The measurement was done by irradiating filter papers containing binary mixture of the elements x and the flux monitor Sc at the various irradiation positions in these three reactors. The experiment was designed in such a way that systematic errors due to mass ratios and efficiency ratios cancel out. Also, rate related errors and backgrounds were kept at negligible values. The results show that B(x,c,m) depends not only on x, c, and m, but also on the type of moderator used for the reactor. We want this new approach to be adopted at all laboratories where routine analysis of multi-element samples are done with the monitor method since the choices of c and m are flexible.

Calibration of the borated ion chamber at NIST reactor thermal column.

PubMed

Wang, Z; Hertel, N E; Lennox, A

2007-01-01

In boron neutron capture therapy and boron neutron capture enhanced fast neutron therapy, the absorbed dose of tissue due to the boron neutron capture reaction is difficult to measure directly. This dose can be computed from the measured thermal neutron fluence rate and the (10)B concentration at the site of interest. A borated tissue-equivalent (TE) ion chamber can be used to directly measure the boron dose in a phantom under irradiation by a neutron beam. Fermilab has two Exradin 0.5 cm(3) Spokas thimble TE ion chambers, one loaded with boron, available for such measurements. At the Fermilab Neutron Therapy Facility, these ion chambers are generally used with air as the filling gas. Since alpha particles and lithium ions from the (10)B(n,alpha)(7)Li reactions have very short ranges in air, the Bragg-Gray principle may not be satisfied for the borated TE ion chamber. A calibration method is described in this paper for the determination of boron capture dose using paired ion chambers. The two TE ion chambers were calibrated in the thermal column of the National Institute of Standards and Technology (NIST) research reactor. The borated TE ion chamber is loaded with 1,000 ppm of natural boron (184 ppm of (10)B). The NIST thermal column has a cadmium ratio of greater than 400 as determined by gold activation. The thermal neutron fluence rate during the calibration was determined using a NIST fission chamber to an accuracy of 5.1%. The chambers were calibrated at two different thermal neutron fluence rates: 5.11 x 10(6) and 4.46 x 10(7)n cm(-2) s(-1). The non-borated ion chamber reading was used to subtract collected charge not due to boron neutron capture reactions. An optically thick lithium slab was used to attenuate the thermal neutrons from the neutron beam port so the responses of the chambers could be corrected for fast neutrons and gamma rays in the beam. The calibration factor of the borated ion chamber was determined to be 1.83 x 10(9) +/- 5.5% (+/- 1sigma) n cm(-2) per nC at standard temperature and pressure condition.

Measuring The Neutron Lifetime to One Second Using in Beam Techniques

NASA Astrophysics Data System (ADS)

Mulholland, Jonathan; NIST In Beam Lifetime Collaboration

2013-10-01

The decay of the free neutron is the simplest nuclear beta decay and is the prototype for charged current semi-leptonic weak interactions. A precise value for the neutron lifetime is required for consistency tests of the Standard Model and is an essential parameter in the theory of Big Bang Nucleosynthesis. A new measurement of the neutron lifetime using the in-beam method is planned at the National Institute of Standards and Technology Center for Neutron Research. The systematic effects associated with the in-beam method are markedly different than those found in storage experiments utilizing ultracold neutrons. Experimental improvements, specifically recent advances in the determination of absolute neutron fluence, should permit an overall uncertainty of 1 second on the neutron lifetime. The technical improvements in the in-beam technique, and the path toward improving the precision of the new measurement will be discussed.

Antiradiation Vaccine: Technology Development Of Prophylaxis, Prevention And Treatment Of Biological Consequences And Complications After Neutron Irradiation.

NASA Astrophysics Data System (ADS)

Popov, Dmitri; Maliev, Slava; Jones, Jeffrey

Introduction: Neutrons irradiation produce a unique biological effectiveness compare to different types of radiation because their ability to create a denser trail of ionized atoms in biological living tissues[Straume 1982; Latif et al.2010; Katz 1978; Bogatyrev 1982]. The efficacy of an Anti-Radiation Vaccine for the prophylaxis, prevention and therapy of acute radiation pathology was studied in a neutron exposure facility. The biological effects of fast neutrons include damage of central nervous system and cardiovascular system with development of Acute Cerebrovascular and Cardiovascular forms of acute radiation pathology. After irradiation by high doses of fast neutron, formation of neurotoxins, hematotoxins,cytotoxins forming from cell's or tissue structures. High doses of Neutron Irradiation generate general and specific toxicity, inflammation reactions. Current Acute Medical Management and Methods of Radiation Protection are not effective against moderate and high doses of neutron irradiation. Our experiments demonstrate that Antiradiation Vaccine is the most effective radioprotectant against high doses of neutron-radiation. Radiation Toxins(biological substances with radio-mimetic properties) isolated from central lymph of gamma-irradiated animals could be working substance with specific antigenic properties for vaccination against neutron irradiation. Methods: Antiradiation Vaccine preparation standard - mixture of a toxoid form of Radiation Toxins - include Cerebrovascular RT Neurotoxin, Cardiovascular RT Neurotoxin, Gastrointestinal RT Neurotoxin, Hematopoietic RT Hematotoxin. Radiation Toxins were isolated from the central lymph of gamma-irradiated animals with different forms of Acute Radiation Syndromes - Cerebrovascular, Cardiovascular, Gastrointestinal, Hematopoietic forms. Devices for Y-radiation were "Panorama","Puma". Neutron exposure was accomplished at the Department of Research Institute of Nuclear Physics, Dubna, Russia. The neutrons irradiation generated in a canal of Research Reactor BBP-M and BBP-M. Mixed neutron beam contained 95% of fast neutron irradiation and 5% of gamma-irradiation. Neutron energy - 1.98 - 2.30 Me V energy. Dose - 10.7 Gy., 0.22 Gy-min. Scheme of experiments: Rabbits from all groups were irradiated in a canal of Research Reactor together. Group A: control-5 rabbits; Group B:placebo-5 rabbits; Group C: radioprotectant Cystamine (50 mg-kg)-5 rabbits, 15 minutes before irradiation Group D:Radio-protectant Mexamine (10 mg-kg)-5 rabbits { 15 minutes before irradiation; Group E: Antiradiation Vaccine: subcutaneus administration or I-M - 2 ml of active substance , 20 days before irradiation. Results: Control Group A - 100% mortality within the next two hours after neutron irradiation with clinical symptoms of acute cerebrovascular syndrome. Group B - 100% mortality less than two hours following irradiation. Group C - 100% mortality within 8-10 hours after irradiation. Group D - 100% mortality within 8-11 hours after irradiation. In Groups A - D the development of extremely severe form of Acute Radiation Cerebrovascular Syndrome produced rapid death. Group E - 100% mortality within 240 hours ( 9|10 days) following neutron irradiation with animals exhibiting cardiovascular, cerebrovascular and gastrointestinal clinical symptoms. Discussion: A pre-irradiation vaccination with Antiradiation Vaccine is effective against mild and even high doses of neutron radiation. Vaccination with antiradiation Vaccine prolonged survival time of rabbits, exposed to a high dose LD100, of neutron radiation: from two hours (control) up to 11 days. We also postulate that radiation toxins,isolated from lymph of gamma-irradiated animals are likely similar to structure of radiation toxins circulated in blood and lymph of neutron irradiated animals. Toxico-kinetics and toxico-dynamics of radiation toxins of after neutron-irradiation were quite unique and distinguished from different types of radiation

Neutron-detecting apparatuses and methods of fabrication

DOEpatents

Dahal, Rajendra P.; Huang, Jacky Kuan-Chih; Lu, James J. Q.; Danon, Yaron; Bhat, Ishwara B.

2015-10-06

Neutron-detecting structures and methods of fabrication are provided which include: a substrate with a plurality of cavities extending into the substrate from a surface; a p-n junction within the substrate and extending, at least in part, in spaced opposing relation to inner cavity walls of the substrate defining the plurality of cavities; and a neutron-responsive material disposed within the plurality of cavities. The neutron-responsive material is responsive to neutrons absorbed for releasing ionization radiation products, and the p-n junction within the substrate spaced in opposing relation to and extending, at least in part, along the inner cavity walls of the substrate reduces leakage current of the neutron-detecting structure.

Neutron Resonance Spin Determination Using Multi-Segmented Detector DANCE

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

Baramsai, B.; Mitchell, G. E.; Chyzh, A.

2011-06-01

A sensitive method to determine the spin of neutron resonances is introduced based on the statistical pattern recognition technique. The new method was used to assign the spins of s-wave resonances in {sup 155}Gd. The experimental neutron capture data for these nuclei were measured with the DANCE (Detector for Advanced Neutron Capture Experiment) calorimeter at the Los Alamos Neutron Science Center. The highly segmented calorimeter provided detailed multiplicity distributions of the capture {gamma}-rays. Using this information, the spins of the neutron capture resonances were determined. With these new spin assignments, level spacings are determined separately for s-wave resonances with J{supmore » {pi}} = 1{sup -} and 2{sup -}.« less

Development of deterministic transport methods for low energy neutrons for shielding in space

NASA Technical Reports Server (NTRS)

Ganapol, Barry

1993-01-01

Transport of low energy neutrons associated with the galactic cosmic ray cascade is analyzed in this dissertation. A benchmark quality analytical algorithm is demonstrated for use with BRYNTRN, a computer program written by the High Energy Physics Division of NASA Langley Research Center, which is used to design and analyze shielding against the radiation created by the cascade. BRYNTRN uses numerical methods to solve the integral transport equations for baryons with the straight-ahead approximation, and numerical and empirical methods to generate the interaction probabilities. The straight-ahead approximation is adequate for charged particles, but not for neutrons. As NASA Langley improves BRYNTRN to include low energy neutrons, a benchmark quality solution is needed for comparison. The neutron transport algorithm demonstrated in this dissertation uses the closed-form Green's function solution to the galactic cosmic ray cascade transport equations to generate a source of neutrons. A basis function expansion for finite heterogeneous and semi-infinite homogeneous slabs with multiple energy groups and isotropic scattering is used to generate neutron fluxes resulting from the cascade. This method, called the FN method, is used to solve the neutral particle linear Boltzmann transport equation. As a demonstration of the algorithm coded in the programs MGSLAB and MGSEMI, neutron and ion fluxes are shown for a beam of fluorine ions at 1000 MeV per nucleon incident on semi-infinite and finite aluminum slabs. Also, to demonstrate that the shielding effectiveness against the radiation from the galactic cosmic ray cascade is not directly proportional to shield thickness, a graph of transmitted total neutron scalar flux versus slab thickness is shown. A simple model based on the nuclear liquid drop assumption is used to generate cross sections for the galactic cosmic ray cascade. The ENDF/B V database is used to generate the total and scattering cross sections for neutrons in aluminum. As an external verification, the results from MGSLAB and MGSEMI were compared to ANISN/PC, a routinely used neutron transport code, showing excellent agreement. In an application to an aluminum shield, the FN method seems to generate reasonable results.

Halogenated sulfidohydroboranes for nuclear medicine and boron neutron capture therapy

DOEpatents

Miura, Michiko; Slatkin, Daniel N.

1997-03-18

A method for performing boron neutron capture therapy for the treatment of tumors is disclosed. The method includes administering to a patient an iodinated sulfidohydroborane, a boron-10-containing compound. The site of the tumor is localized by visualizing the increased concentration of the iodine labelled compound at the tumor. The targeted tumor is then irradiated with a beam of neutrons having an energy distribution effective for neutron capture. Destruction of the tumor occurs due to high LET particle irradiation of the tissue secondary to the incident neutrons being captured by the boron-10 nuclei. Iodinated sulfidohydroboranes are disclosed which are especially suitable for the method of the invention. In a preferred embodiment, a compound having the formula Na.sub.4 B.sub.12 I.sub.11 SSB.sub.12 I.sub.11, or another pharmaceutically acceptable salt of the compound, may be administered to a cancer patient for boron neutron capture therapy.

Halogenated sulfidohydroboranes for nuclear medicine and boron neutron capture therapy

DOEpatents

Miura, Michiko; Slatkin, Daniel N.

1995-10-03

A method for performing boron neutron capture therapy for the treatment of tumors is disclosed. The method includes administering to a patient an iodinated sulfidohydroborane, a boron-10-containing compound. The site of the tumor is localized by visualizing the increased concentration of the iodine labelled compound at the tumor. The targeted tumor is then irradiated with a beam of neutrons having an energy distribution effective for neutron capture. Destruction of the tumor occurs due to high LET particle irradiation of the tissue secondary to the incident neutrons being captured by the boron-10 nuclei. Iodinated sulfidohydroboranes are disclosed which are especially suitable for the method of the invention. In a preferred embodiment, a compound having the formula Na.sub.4 B.sub.12 I.sub.11 SSB.sub.12 I.sub.11, or another pharmaceutically acceptable salt of the compound, may be administered to a cancer patient for boron neutron capture therapy.

Halogenated sulfidohydroboranes for nuclear medicine and boron neutron capture therapy

DOEpatents

Miura, Michiko; Slatkin, Daniel N.

1997-08-05

A method for performing boron neutron capture therapy for the treatment of tumors is disclosed. The method includes administering to a patient an iodinated sulfidohydroborane, a boron-10-containing compound. The site of the tumor is localized. by visualizing the increased concentration of the iodine labelled compound at the tumor. The targeted tumor is then irradiated with a beam of neutrons having an energy distribution effective for neutron capture. Destruction of the tumor occurs due to high LET particle irradiation of the tissue secondary to the incident neutrons being captured by the boron-10 nuclei. Iodinated sulfidohydroboranes are disclosed which are especially suitable for the method of the invention. In a preferred embodiment, a compound having the formula Na.sub.4 B.sub.12 I.sub.11 SSB.sub.12 I.sub.11, or another pharmaceutically acceptable salt of the compound, may be administered to a cancer patient for boron neutron capture therapy.

Halogenated sulfidohydroboranes for nuclear medicine and boron neutron capture therapy

DOEpatents

Miura, M.; Slatkin, D.N.

1995-10-03

A method for performing boron neutron capture therapy for the treatment of tumors is disclosed. The method includes administering to a patient an iodinated sulfidohydroborane, a boron-10-containing compound. The site of the tumor is localized by visualizing the increased concentration of the iodine labelled compound at the tumor. The targeted tumor is then irradiated with a beam of neutrons having an energy distribution effective for neutron capture. Destruction of the tumor occurs due to high LET particle irradiation of the tissue secondary to the incident neutrons being captured by the boron-10 nuclei. Iodinated sulfidohydroboranes are disclosed which are especially suitable for the method of the invention. In a preferred embodiment, a compound having the formula Na{sub 4}B{sub 12}I{sub 11}SSB{sub 12}I{sub 11}, or another pharmaceutically acceptable salt of the compound, may be administered to a cancer patient for boron neutron capture therapy. 1 fig.

Image enhancement using MCNP5 code and MATLAB in neutron radiography.

PubMed

Tharwat, Montaser; Mohamed, Nader; Mongy, T

2014-07-01

This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work. Copyright © 2014 Elsevier Ltd. All rights reserved.

Halogenated sulfidohydroboranes for nuclear medicine and boron neutron capture therapy

DOEpatents

Miura, M.; Slatkin, D.N.

1997-03-18

A method for performing boron neutron capture therapy for the treatment of tumors is disclosed. The method includes administering to a patient an iodinated sulfidohydroborane, a boron-10-containing compound. The site of the tumor is localized by visualizing the increased concentration of the iodine labelled compound at the tumor. The targeted tumor is then irradiated with a beam of neutrons having an energy distribution effective for neutron capture. Destruction of the tumor occurs due to high LET particle irradiation of the tissue secondary to the incident neutrons being captured by the boron-10 nuclei. Iodinated sulfidohydroboranes are disclosed which are especially suitable for the method of the invention. In a preferred embodiment, a compound having the formula Na{sub 4}B{sub 12}I{sub 11}SSB{sub 12}I{sub 11}, or another pharmaceutically acceptable salt of the compound, may be administered to a cancer patient for boron neutron capture therapy. 1 fig.

Halogenated sulfidohydroboranes for nuclear medicine and boron neutron capture therapy

DOEpatents

Miura, M.; Slatkin, D.N.

1997-08-05

A method for performing boron neutron capture therapy for the treatment of tumors is disclosed. The method includes administering to a patient an iodinated sulfidohydroborane, a boron-10-containing compound. The site of the tumor is localized by visualizing the increased concentration of the iodine labelled compound at the tumor. The targeted tumor is then irradiated with a beam of neutrons having an energy distribution effective for neutron capture. Destruction of the tumor occurs due to high LET particle irradiation of the tissue secondary to the incident neutrons being captured by the boron-10 nuclei. Iodinated sulfidohydroboranes are disclosed which are especially suitable for the method of the invention. In a preferred embodiment, a compound having the formula Na{sub 4}B{sub 12}I{sub 11}SSB{sub 12}I{sub 11}, or another pharmaceutically acceptable salt of the compound, may be administered to a cancer patient for boron neutron capture therapy. 1 fig.

Evaluation of Am–Li neutron spectra data for active well type neutron multiplicity measurements of uranium

DOE PAGES

Goddard, Braden; Croft, Stephen; Lousteau, Angela; ...

2016-05-25

Safeguarding nuclear material is an important and challenging task for the international community. One particular safeguards technique commonly used for uranium assay is active neutron correlation counting. This technique involves irradiating unused uranium with ( α,n) neutrons from an Am-Li source and recording the resultant neutron pulse signal which includes induced fission neutrons. Although this non-destructive technique is widely employed in safeguards applications, the neutron energy spectra from an Am-Li sources is not well known. Several measurements over the past few decades have been made to characterize this spectrum; however, little work has been done comparing the measured spectra ofmore » various Am-Li sources to each other. This paper examines fourteen different Am-Li spectra, focusing on how these spectra affect simulated neutron multiplicity results using the code Monte Carlo N-Particle eXtended (MCNPX). Two measurement and simulation campaigns were completed using Active Well Coincidence Counter (AWCC) detectors and uranium standards of varying enrichment. The results of this work indicate that for standard AWCC measurements, the fourteen Am-Li spectra produce similar doubles and triples count rates. Finally, the singles count rates varied by as much as 20% between the different spectra, although they are usually not used in quantitative analysis.« less

Characterization of the graphite pile as a source of thermal neutrons

NASA Astrophysics Data System (ADS)

Vykydal, Zdenek; Králík, Miloslav; Jančář, Aleš; Kopecký, Zdeněk; Dressler, Jan; Veškrna, Martin

2015-11-01

A new graphite pile designed to serve as a standard source of thermal neutrons has been built at the Czech Metrology Institute. Actual dimensions of the pile are 1.95 m (W)×1.95 m (L)×2.0 m (H). At its center, there is a measurement channel whose dimensions are 0.4 m×0.4 m×1.25 m (depth). The channel is equipped with a calibration bench, which allows reproducible placement of the tested/calibrated device. At a distance of 80 cm from the channel axis, six holes are symmetrically located allowing the placement of radionuclide neutron sources of Pu-Be and/or Am-Be type. Spatial distribution of thermal neutron fluence in the cavity was calculated in detail with the MCNP neutron transport code. Experimentally, it was measured with two active detectors: a small 3He proportional detector by the French company LMT, type 0.5 NH 1/1 KF, and a silicon pixel detector Timepix with 10B converter foil. The relative values of thermal neutron fluence rate obtained with active detectors were converted to absolute ones using thermal neutron fluence rates measured by means of gold foil activation. The quality of thermal neutron field was characterized by the cadmium ratio.

Theoretical and experimental physical methods of neutron-capture therapy

NASA Astrophysics Data System (ADS)

Borisov, G. I.

2011-09-01

This review is based to a substantial degree on our priority developments and research at the IR-8 reactor of the Russian Research Centre Kurchatov Institute. New theoretical and experimental methods of neutron-capture therapy are developed and applied in practice; these are: A general analytical and semi-empiric theory of neutron-capture therapy (NCT) based on classical neutron physics and its main sections (elementary theories of moderation, diffuse, reflection, and absorption of neutrons) rather than on methods of mathematical simulation. The theory is, first of all, intended for practical application by physicists, engineers, biologists, and physicians. This theory can be mastered by anyone with a higher education of almost any kind and minimal experience in operating a personal computer.

Resolving the neutron lifetime puzzle

NASA Astrophysics Data System (ADS)

Mumm, Pieter

2018-05-01

Free electrons and protons are stable, but outside atomic nuclei, free neutrons decay into a proton, electron, and antineutrino through the weak interaction, with a lifetime of ∼880 s (see the figure). The most precise measurements have stated uncertainties below 1 s (0.1%), but different techniques, although internally consistent, disagree by 4 standard deviations given the quoted uncertainties. Resolving this “neutron lifetime puzzle” has spawned much experimental effort as well as exotic theoretical mechanisms, thus far without a clear explanation. On page 627 of this issue, Pattie et al. (1) present the most precise measurement of the neutron lifetime to date. A new method of measuring trapped neutrons in situ allows a more detailed exploration of one of the more pernicious systematic effects in neutron traps, neutron phase-space evolution (the changing orbits of neutrons in the trap), than do previous methods. The precision achieved, combined with a very different set of systematic uncertainties, gives hope that experiments such as this one can help resolve the current situation with the neutron lifetime.

Evaluation of radioactivity in the bodies of mice induced by neutron exposure from an epi-thermal neutron source of an accelerator-based boron neutron capture therapy system

PubMed Central

NAKAMURA, Satoshi; IMAMICHI, Shoji; MASUMOTO, Kazuyoshi; ITO, Masashi; WAKITA, Akihisa; OKAMOTO, Hiroyuki; NISHIOKA, Shie; IIJIMA, Kotaro; KOBAYASHI, Kazuma; ABE, Yoshihisa; IGAKI, Hiroshi; KURITA, Kazuyoshi; NISHIO, Teiji; MASUTANI, Mitsuko; ITAMI, Jun

2017-01-01

This study aimed to evaluate the residual radioactivity in mice induced by neutron irradiation with an accelerator-based boron neutron capture therapy (BNCT) system using a solid Li target. The radionuclides and their activities were evaluated using a high-purity germanium (HP-Ge) detector. The saturated radioactivity of the irradiated mouse was estimated to assess the radiation protection needs for using the accelerator-based BNCT system. 24Na, 38Cl, 80mBr, 82Br, 56Mn, and 42K were identified, and their saturated radioactivities were (1.4 ± 0.1) × 102, (2.2 ± 0.1) × 101, (3.4 ± 0.4) × 102, 2.8 ± 0.1, 8.0 ± 0.1, and (3.8 ± 0.1) × 101 Bq/g/mA, respectively. The 24Na activation rate at a given neutron fluence was found to be consistent with the value reported from nuclear-reactor-based BNCT experiments. The induced activity of each nuclide can be estimated by entering the saturated activity of each nuclide, sample mass, irradiation time, and proton current into the derived activation equation in our accelerator-based BNCT system. PMID:29225308

Evaluation of radioactivity in the bodies of mice induced by neutron exposure from an epi-thermal neutron source of an accelerator-based boron neutron capture therapy system.

PubMed

Nakamura, Satoshi; Imamichi, Shoji; Masumoto, Kazuyoshi; Ito, Masashi; Wakita, Akihisa; Okamoto, Hiroyuki; Nishioka, Shie; Iijima, Kotaro; Kobayashi, Kazuma; Abe, Yoshihisa; Igaki, Hiroshi; Kurita, Kazuyoshi; Nishio, Teiji; Masutani, Mitsuko; Itami, Jun

2017-01-01

This study aimed to evaluate the residual radioactivity in mice induced by neutron irradiation with an accelerator-based boron neutron capture therapy (BNCT) system using a solid Li target. The radionuclides and their activities were evaluated using a high-purity germanium (HP-Ge) detector. The saturated radioactivity of the irradiated mouse was estimated to assess the radiation protection needs for using the accelerator-based BNCT system. 24 Na, 38 Cl, 80m Br, 82 Br, 56 Mn, and 42 K were identified, and their saturated radioactivities were (1.4 ± 0.1) × 10 2 , (2.2 ± 0.1) × 10 1 , (3.4 ± 0.4) × 10 2 , 2.8 ± 0.1, 8.0 ± 0.1, and (3.8 ± 0.1) × 10 1 Bq/g/mA, respectively. The 24 Na activation rate at a given neutron fluence was found to be consistent with the value reported from nuclear-reactor-based BNCT experiments. The induced activity of each nuclide can be estimated by entering the saturated activity of each nuclide, sample mass, irradiation time, and proton current into the derived activation equation in our accelerator-based BNCT system.

Neutron-neutron angular correlations in spontaneous fission of 252Cf and 240Pu

NASA Astrophysics Data System (ADS)

Verbeke, J. M.; Nakae, L. F.; Vogt, R.

2018-04-01

Background: Angular anisotropy has been observed between prompt neutrons emitted during the fission process. Such an anisotropy arises because the emitted neutrons are boosted along the direction of the parent fragment. Purpose: To measure the neutron-neutron angular correlations from the spontaneous fission of 252Cf and 240Pu oxide samples using a liquid scintillator array capable of pulse-shape discrimination. To compare these correlations to simulations combining the Monte Carlo radiation transport code MCNPX with the fission event generator FREYA. Method: Two different analysis methods were used to study the neutron-neutron correlations with varying energy thresholds. The first is based on setting a light output threshold while the second imposes a time-of-flight cutoff. The second method has the advantage of being truly detector independent. Results: The neutron-neutron correlation modeled by FREYA depends strongly on the sharing of the excitation energy between the two fragments. The measured asymmetry enabled us to adjust the FREYA parameter x in 240Pu, which controls the energy partition between the fragments and is so far inaccessible in other measurements. The 240Pu data in this analysis was the first available to quantify the energy partition for this isotope. The agreement between data and simulation is overall very good for 252Cf(sf ) and 240Pu(sf ) . Conclusions: The asymmetry in the measured neutron-neutron angular distributions can be predicted by FREYA. The shape of the correlation function depends on how the excitation energy is partitioned between the two fission fragments. Experimental data suggest that the lighter fragment is disproportionately excited.

NAA For Human Serum Analysis: Comparison With Conventional Analyses

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

Oliveira, Laura C.; Zamboni, Cibele B.; Medeiros, Jose A. G.

2010-08-04

Instrumental and Comparator methods of Neutron Activation Analysis (NAA) were applied to determine elements of clinical relevancy in serum samples of adult population (Sao Paulo city, Brazil). A comparison with the conventional analyses, Colorimetric for calcium, Titrymetric for chlorine and Ion Specific Electrode for sodium and potassium determination were also performed permitting a discussion about the performance of NAA methods for clinical chemistry research.

Investigations on landmine detection by neutron-based techniques.

PubMed

Csikai, J; Dóczi, R; Király, B

2004-07-01

Principles and techniques of some neutron-based methods used to identify the antipersonnel landmines (APMs) are discussed. New results have been achieved in the field of neutron reflection, transmission, scattering and reaction techniques. Some conclusions are as follows: The neutron hand-held detector is suitable for the observation of anomaly caused by a DLM2-like sample in different soils with a scanning speed of 1m(2)/1.5 min; the reflection cross section of thermal neutrons rendered the determination of equivalent thickness of different soil components possible; a simple method was developed for the determination of the thermal neutron flux perturbation factor needed for multi-elemental analysis of bulky samples; unfolded spectra of elastically backscattered neutrons using broad-spectrum sources render the identification of APMs possible; the knowledge of leakage spectra of different source neutrons is indispensable for the determination of the differential and integrated reaction rates and through it the dimension of the interrogated volume; the precise determination of the C/O atom fraction requires the investigations on the angular distribution of the 6.13MeV gamma-ray emitted in the (16)O(n,n'gamma) reaction. These results, in addition to the identification of landmines, render the improvement of the non-intrusive neutron methods possible.

A 23-GROUP NEUTRON THERMALIZATION CROSS SECTION LIBRARY

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

Doctor, R.D.; Boling, M.A.

1963-07-15

A set of 23-group neutron cross sections for use in the calculation of neutron thermalization and thermal neutron spectral effects in SNAP reactors is compiled. The sources and methods used to obtain the cross sections are described. (auth)

Practical new method of measuring thermal-neutron fluence

NASA Technical Reports Server (NTRS)

Siebold, J. R.; Warman, E. A.

1967-01-01

Thermoluminescence dosimeter technique measures thermal-neutron fluence by encapsulating lithium flouride phosphor powder and exposing it to a neutron environment. The capsule is heated in a dosimeter reader, which results in light emission proportional to the neutron fluence.

Neutron dosimetry

DOEpatents

Quinby, Thomas C.

1976-07-27

A method of measuring neutron radiation within a nuclear reactor is provided. A sintered oxide wire is disposed within the reactor and exposed to neutron radiation. The induced radioactivity is measured to provide an indication of the neutron energy and flux within the reactor.

X-ray and neutron diffraction studies of crystallinity in hydroxyapatite coatings.

PubMed

Girardin, E; Millet, P; Lodini, A

2000-02-01

To standardize industrial implant production and make comparisons between different experimental results, we have to be able to quantify the crystallinity of hydroxyapatite. Methods of measuring crystallinity ratio were developed for various HA samples before and after plasma spraying. The first series of methods uses X-ray diffraction. The advantage of these methods is that X-ray diffraction equipment is used widely in science and industry. In the second series, a neutron diffraction method is developed and the results recorded are similar to those obtained by the modified X-ray diffraction methods. The advantage of neutron diffraction is the ability to obtain measurements deep inside a component. It is a nondestructive method, owing to the very low absorption of neutrons in most materials. Copyright 2000 John Wiley & Sons, Inc.

Designing a new type of neutron detector for neutron and gamma-ray discrimination via GEANT4.

PubMed

Shan, Qing; Chu, Shengnan; Ling, Yongsheng; Cai, Pingkun; Jia, Wenbao

2016-04-01

Design of a new type of neutron detector, consisting of a fast neutron converter, plastic scintillator, and Cherenkov detector, to discriminate 14-MeV fast neutrons and gamma rays in a pulsed n-γ mixed field and monitor their neutron fluxes is reported in this study. Both neutrons and gamma rays can produce fluorescence in the scintillator when they are incident on the detector. However, only the secondary charged particles of the gamma rays can produce Cherenkov light in the Cherenkov detector. The neutron and gamma-ray fluxes can be calculated by measuring the fluorescence and Cherenkov light. The GEANT4 Monte Carlo simulation toolkit is used to simulate the whole process occurring in the detector, whose optimum parameters are known. Analysis of the simulation results leads to a calculation method of neutron flux. This method is verified by calculating the neutron fluxes using pulsed n-γ mixed fields with different n/γ ratios, and the results show that the relative errors of all calculations are <5%. Copyright © 2016 Elsevier Ltd. All rights reserved.

Status of the BL2 beam measurement of the neutron lifetime

NASA Astrophysics Data System (ADS)

Hoogerheide, Shannon Fogwell; BL2 Collaboration

2017-09-01

Neutron beta decay is the simplest example of nuclear beta decay and a precise value of the neutron lifetime is important for consistency tests of the Standard Model and Big Bang Nucleosynthesis models. A new measurement of the neutron lifetime, utilizing the beam method, is underway at the National Institute of Standards and Technology Center for Neutron Research with a projected uncertainty of 1 s. A review of the beam method and the technical improvements in this experiment will be presented. The status of the experiment, as well as preliminary measurements, beam characteristics, and early data will be discussed.

A Ramsey’s Method With Pulsed Neutrons for a T-Violation Experiment

PubMed Central

Masuda, Y.; Ino, T.; Muto, S.; Skoy, V.

2005-01-01

A Ramsey’s method with pulsed neutrons is discussed for neutron spin manipulation in a time reversal (T) symmetry violation experiment. The neutron spin (sn) is aligned to the direction of a vector product of the nuclear spin (I) and the neutron momentum (kn) for the measurement of a T-odd correlation term, which is represented as sn · (kn × I), during propagation through a polarized nuclear target. The phase control and amplitude modulation of separated oscillatory fields are discussed for the measurement of the T-odd correlation term. PMID:27308171

A Ramsey's Method With Pulsed Neutrons for a T-Violation Experiment.

PubMed

Masuda, Y; Ino, T; Muto, S; Skoy, V

2005-01-01

A Ramsey's method with pulsed neutrons is discussed for neutron spin manipulation in a time reversal (T) symmetry violation experiment. The neutron spin (s n) is aligned to the direction of a vector product of the nuclear spin ( I ) and the neutron momentum ( k n) for the measurement of a T-odd correlation term, which is represented as s n · ( k n × I ), during propagation through a polarized nuclear target. The phase control and amplitude modulation of separated oscillatory fields are discussed for the measurement of the T-odd correlation term.

High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

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

Donna Post Guillen; Larry R. Greenwood; James R. Parry

2014-06-22

A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluencemore » monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.« less

Passive Neutron Non-Destructive Assay for Remediation of Radiological Waste at Hanford Burial Grounds- 13189

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

Simpson, A.; Pitts, M.; Ludowise, J.D.

The Hanford burial grounds contains a broad spectrum of low activity radioactive wastes, transuranic (TRU) wastes, and hazardous wastes including fission products, byproduct material (thorium and uranium), plutonium and laboratory chemicals. A passive neutron non-destructive assay technique has been developed for characterization of shielded concreted drums exhumed from the burial grounds. This method facilitates the separation of low activity radiological waste containers from TRU waste containers exhumed from the burial grounds. Two identical total neutron counting systems have been deployed, each consisting of He-3 detectors surrounded by a polyethylene moderator. The counts are processed through a statistical filter that removesmore » outliers in order to suppress cosmic spallation events and electronic noise. Upon completion of processing, a 'GO / NO GO' signal is provided to the operator based on a threshold level equivalent to 0.5 grams of weapons grade plutonium in the container being evaluated. This approach allows instantaneous decisions to be made on how to proceed with the waste. The counting systems have been set up using initial on-site measurements (neutron emitting standards loaded into surrogate waste containers) combined with Monte Carlo modeling techniques. The benefit of this approach is to allow the systems to extend their measurement ranges, in terms of applicable matrix types and container sizes, with minimal interruption to the operations at the burial grounds. (authors)« less

SNS Sample Activation Calculator Flux Recommendations and Validation

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

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

The Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) uses the Sample Activation Calculator (SAC) to calculate the activation of a sample after the sample has been exposed to the neutron beam in one of the SNS beamlines. The SAC webpage takes user inputs (choice of beamline, the mass, composition and area of the sample, irradiation time, decay time, etc.) and calculates the activation for the sample. In recent years, the SAC has been incorporated into the user proposal and sample handling process, and instrument teams and users have noticed discrepancies in the predicted activation of their samples.more » The Neutronics Analysis Team validated SAC by performing measurements on select beamlines and confirmed the discrepancies seen by the instrument teams and users. The conclusions were that the discrepancies were a result of a combination of faulty neutron flux spectra for the instruments, improper inputs supplied by SAC (1.12), and a mishandling of cross section data in the Sample Activation Program for Easy Use (SAPEU) (1.1.2). This report focuses on the conclusion that the SAPEU (1.1.2) beamline neutron flux spectra have errors and are a significant contributor to the activation discrepancies. The results of the analysis of the SAPEU (1.1.2) flux spectra for all beamlines will be discussed in detail. The recommendations for the implementation of improved neutron flux spectra in SAPEU (1.1.3) are also discussed.« less

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 a generalized perturbation theory method for sensitivity analysis using continuous-energy Monte Carlo methods

DOE PAGES

Perfetti, Christopher M.; Rearden, Bradley T.

2016-03-01

The sensitivity and uncertainty analysis tools of the ORNL SCALE nuclear modeling and simulation code system that have been developed over the last decade have proven indispensable for numerous application and design studies for nuclear criticality safety and reactor physics. SCALE contains tools for analyzing the uncertainty in the eigenvalue of critical systems, but cannot quantify uncertainty in important neutronic parameters such as multigroup cross sections, fuel fission rates, activation rates, and neutron fluence rates with realistic three-dimensional Monte Carlo simulations. A more complete understanding of the sources of uncertainty in these design-limiting parameters could lead to improvements in processmore » optimization, reactor safety, and help inform regulators when setting operational safety margins. A novel approach for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was recently explored as academic research and has been found to accurately and rapidly calculate sensitivity coefficients in criticality safety applications. The work presented here describes a new method, known as the GEAR-MC method, which extends the CLUTCH theory for calculating eigenvalue sensitivity coefficients to enable sensitivity coefficient calculations and uncertainty analysis for a generalized set of neutronic responses using high-fidelity continuous-energy Monte Carlo calculations. Here, several criticality safety systems were examined to demonstrate proof of principle for the GEAR-MC method, and GEAR-MC was seen to produce response sensitivity coefficients that agreed well with reference direct perturbation sensitivity coefficients.« less

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.

Naturally induced secondary radiation in interplanetary space: Preliminary analyses for gamma radiation and radioisotope production from thermal neutron activation

NASA Technical Reports Server (NTRS)

Plaza-Rosado, Heriberto

1991-01-01

Thermal neutron activation analyses were carried out for various space systems components to determine gamma radiation dose rates and food radiation contamination levels. The space systems components selected were those for which previous radiation studies existed. These include manned space vehicle radiation shielding, liquid hydrogen propellant tanks for a Mars mission, and a food supply used as space vehicle radiation shielding. The computational method used is based on the fast neutron distribution generated by the BRYNTRN and HZETRN transport codes for Galactic Cosmic Rays (GCR) at solar minimum conditions and intense solar flares in space systems components. The gamma dose rates for soft tissue are calculated for water and aluminum space vehicle slab shields considering volumetric source self-attenuation and exponential buildup factors. In the case of the lunar habitat with regolith shielding, a completely exposed spherical habitat was assumed for mathematical convenience and conservative calculations. Activation analysis of the food supply used as radiation shielding is presented for four selected nutrients: potassium, calcium, sodium, and phosphorus. Radioactive isotopes that could represent a health hazard if ingested are identified and their concentrations are identified. For nutrients soluble in water, it was found that all induced radioactivity was below the accepted maximum permissible concentrations.

Naturally induced secondary radiation in interplanetary space: Preliminary analyses for gamma radiation and radioisotope production from thermal neutron activation

NASA Astrophysics Data System (ADS)

Plaza-Rosado, Heriberto

1991-09-01

Thermal neutron activation analyses were carried out for various space systems components to determine gamma radiation dose rates and food radiation contamination levels. The space systems components selected were those for which previous radiation studies existed. These include manned space vehicle radiation shielding, liquid hydrogen propellant tanks for a Mars mission, and a food supply used as space vehicle radiation shielding. The computational method used is based on the fast neutron distribution generated by the BRYNTRN and HZETRN transport codes for Galactic Cosmic Rays (GCR) at solar minimum conditions and intense solar flares in space systems components. The gamma dose rates for soft tissue are calculated for water and aluminum space vehicle slab shields considering volumetric source self-attenuation and exponential buildup factors. In the case of the lunar habitat with regolith shielding, a completely exposed spherical habitat was assumed for mathematical convenience and conservative calculations. Activation analysis of the food supply used as radiation shielding is presented for four selected nutrients: potassium, calcium, sodium, and phosphorus. Radioactive isotopes that could represent a health hazard if ingested are identified and their concentrations are identified. For nutrients soluble in water, it was found that all induced radioactivity was below the accepted maximum permissible concentrations.

Nonnegative Tensor Factorization Approach Applied to Fission Chamber’s Output Signals Blind Source Separation

NASA Astrophysics Data System (ADS)

Laassiri, M.; Hamzaoui, E.-M.; Cherkaoui El Moursli, R.

2018-02-01

Inside nuclear reactors, gamma-rays emitted from nuclei together with the neutrons introduce unwanted backgrounds in neutron spectra. For this reason, powerful extraction methods are needed to extract useful neutron signal from recorded mixture and thus to obtain clearer neutron flux spectrum. Actually, several techniques have been developed to discriminate between neutrons and gamma-rays in a mixed radiation field. Most of these techniques, tackle using analogue discrimination methods. Others propose to use some organic scintillators to achieve the discrimination task. Recently, systems based on digital signal processors are commercially available to replace the analog systems. As alternative to these systems, we aim in this work to verify the feasibility of using a Nonnegative Tensor Factorization (NTF) to blind extract neutron component from mixture signals recorded at the output of fission chamber (WL-7657). This last have been simulated through the Geant4 linked to Garfield++ using a 252Cf neutron source. To achieve our objective of obtaining the best possible neutron-gamma discrimination, we have applied the two different NTF algorithms, which have been found to be the best methods that allow us to analyse this kind of nuclear data.

Determination of selenium in the environment and in biological material.

PubMed Central

Bem, E M

1981-01-01

This paper reviews the following problems, sampling, decomposition procedures and most important analytical methods used for selenium determination, e.g., neutron activation analysis, atomic absorption spectrometry, gas-liquid chromatography, spectrophotometry, fluorimetry, and x-ray fluorescence. This review covers the literature mainly from 1975 to 1977. PMID:7007035

An active target for the accelerator-based transmutation system

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

Grebyonkin, K.F.

1995-10-01

Consideration is given to the possibility of radical reduction in power requirements to the proton accelerator of the electronuclear reactor due to neutron multiplication both in the blanket and the target of an active material. The target is supposed to have the fast-neutron spectrum, and the blanket-the thermal one. The blanket and the target are separated by the thermal neutrons absorber, which is responsible for the neutron decoupling of the active target and blanket. Also made are preliminary estimations which illustrate that the realization of the idea under consideration can lead to significant reduction in power requirements to the protonmore » beam and, hence considerably improve economic characteristics of the electronuclear reactor.« less

Relativistic neutrons in active galactic nuclei

NASA Technical Reports Server (NTRS)

Sikora, Marek; Begelman, Mitchell C.; Rudak, Bronislaw

1989-01-01

The acceleration of protons to relativistic energies in active galactic nuclei leads to the creation of relativistic neutrons which escape from the central engine. The neutrons decay at distances of up to 1-100 pc, depositing their energies and momenta in situ. Energy deposition by decaying neutrons may inhibit spherical accretion and drive a wind, which could be responsible for the velocity fields in emission-line regions and the outflow of broad absorption line systems. Enhanced pressure in the neutron decay region may also help to confine emission line clouds. A fraction of the relativistic proton energy is radiated in gamma-rays with energies which may be as large as about 100,000 GeV.

OVERVIEW OF NEUTRON MEASUREMENTS IN JET FUSION DEVICE.

PubMed

Batistoni, P; Villari, R; Obryk, B; Packer, L W; Stamatelatos, I E; Popovichev, S; Colangeli, A; Colling, B; Fonnesu, N; Loreti, S; Klix, A; Klosowski, M; Malik, K; Naish, J; Pillon, M; Vasilopoulou, T; De Felice, P; Pimpinella, M; Quintieri, L

2017-10-05

The design and operation of ITER experimental fusion reactor requires the development of neutron measurement techniques and numerical tools to derive the fusion power and the radiation field in the device and in the surrounding areas. Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case in ITER and power plant studies. The required radiation transport calculations are extremely challenging because of the large physical extent of the reactor plant, the complexity of the geometry, and the combination of deep penetration and streaming paths. This article reports the experimental activities which are carried-out at JET to validate the neutronics measurements methods and numerical tools used in ITER and power plant design. A new deuterium-tritium campaign is proposed in 2019 at JET: the unique 14 MeV neutron yields produced will be exploited as much as possible to validate measurement techniques, codes, procedures and data currently used in ITER design thus reducing the related uncertainties and the associated risks in the machine operation. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

STEAM FORMING NEUTRONIC REACTOR AND METHOD OF OPERATING IT

DOEpatents

Untermyer, S.

1960-05-10

The heterogeneous reactor is liquid moderated and cooled by a steam forming coolant and is designed to produce steam from the coolant directly within the active portion of the reactor while avoiding the formation of bubbles in the liquid moderator. This reactor achieves inherent stability as a result of increased neutron leakage and increased neutron resonance absorption in the U/sup 238/ fuel with the formation of bubbles. The invention produces certain conditions under which the formation of vapor bubbles as a result of a neutron flux excursion from the injection of a reactivity increment into the reactor will operate to nullify the reactivity increment within a sufficiently short period of time to prevent unsafe reactor operating conditions from developing. This is obtained by disposing a plurality of fuel elements within a mass of steam forming coolant in the core with the ratio of the volume of steam forming coolant to the volume of fissionable isotopes being within the range yielding a multiplication factor greater than unity and a negative reactivity to core void coefficient at the boiling temperature of the coolant.

Progress toward the development and testing of source reconstruction methods for NIF neutron imaging.

PubMed

Loomis, E N; Grim, G P; Wilde, C; Wilson, D C; Morgan, G; Wilke, M; Tregillis, I; Merrill, F; Clark, D; Finch, J; Fittinghoff, D; Bower, D

2010-10-01

Development of analysis techniques for neutron imaging at the National Ignition Facility is an important and difficult task for the detailed understanding of high-neutron yield inertial confinement fusion implosions. Once developed, these methods must provide accurate images of the hot and cold fuels so that information about the implosion, such as symmetry and areal density, can be extracted. One method under development involves the numerical inversion of the pinhole image using knowledge of neutron transport through the pinhole aperture from Monte Carlo simulations. In this article we present results of source reconstructions based on simulated images that test the methods effectiveness with regard to pinhole misalignment.

Method of using deuterium-cluster foils for an intense pulsed neutron source

DOEpatents

Miley, George H.; Yang, Xiaoling

2013-09-03

A method is provided for producing neutrons, comprising: providing a converter foil comprising deuterium clusters; focusing a laser on the foil with power and energy sufficient to cause deuteron ions to separate from the foil; and striking a surface of a target with the deuteron ions from the converter foil with energy sufficient to cause neutron production by a reaction selected from the group consisting of D-D fusion, D-T fusion, D-metal nuclear spallation, and p-metal. A further method is provided for assembling a plurality of target assemblies for a target injector to be used in the previously mentioned manner. A further method is provided for producing neutrons, comprising: splitting a laser beam into a first beam and a second beam; striking a first surface of a target with the first beam, and an opposite second surface of the target with the second beam with energy sufficient to cause neutron production.

Shutdown Dose Rate Analysis Using the Multi-Step CADIS Method

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

Ibrahim, Ahmad M.; Peplow, Douglas E.; Peterson, Joshua L.

2015-01-01

The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) hybrid Monte Carlo (MC)/deterministic radiation transport method was proposed to speed up the shutdown dose rate (SDDR) neutron MC calculation using an importance function that represents the neutron importance to the final SDDR. This work applied the MS-CADIS method to the ITER SDDR benchmark problem. The MS-CADIS method was also used to calculate the SDDR uncertainty resulting from uncertainties in the MC neutron calculation and to determine the degree of undersampling in SDDR calculations because of the limited ability of the MC method to tally detailed spatial and energy distributions. The analysismore » that used the ITER benchmark problem compared the efficiency of the MS-CADIS method to the traditional approach of using global MC variance reduction techniques for speeding up SDDR neutron MC calculation. Compared to the standard Forward-Weighted-CADIS (FW-CADIS) method, the MS-CADIS method increased the efficiency of the SDDR neutron MC calculation by 69%. The MS-CADIS method also increased the fraction of nonzero scoring mesh tally elements in the space-energy regions of high importance to the final SDDR.« less

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.

Measurement of the neutron-capture cross section on 63,65Cu between 0.4 and 7.5 MeV

NASA Astrophysics Data System (ADS)

Bray, Isabel; Bhike, Megha; Krishichayan, (None); Tornow, W.

2015-10-01

Copper is currently being used as a cooling and shielding material in most experimental searches for 0 ν β β decay. In order to accurately interpret background events in these experiments, the cross section of neutron-induced reactions on copper must be known. The purpose of this work was to measure the cross section of the 63,65Cu(n, γ)64,66Cu reactions. Data were collected through the activation method at a range of energies from approximately 0.4 MeV to 7.5 MeV, employing the neutron production reactions 3H(p,n)3Heand2H(d,n)3He. Previous data were limited to energies below approximately 3 MeV. The results are compared to predictions from the nuclear data libraries ENDF/B-VII.1 and TENDL-2014.

Advanced Non-Destructive Assessment Technology to Determine the Aging of Silicon Containing Materials for Generation IV Nuclear Reactors

NASA Astrophysics Data System (ADS)

Koenig, T. W.; Olson, D. L.; Mishra, B.; King, J. C.; Fletcher, J.; Gerstenberger, L.; Lawrence, S.; Martin, A.; Mejia, C.; Meyer, M. K.; Kennedy, R.; Hu, L.; Kohse, G.; Terry, J.

2011-06-01

To create an in-situ, real-time method of monitoring neutron damage within a nuclear reactor core, irradiated silicon carbide samples are examined to correlate measurable variations in the material properties with neutron fluence levels experienced by the silicon carbide (SiC) during the irradiation process. The reaction by which phosphorus doping via thermal neutrons occurs in the silicon carbide samples is known to increase electron carrier density. A number of techniques are used to probe the properties of the SiC, including ultrasonic and Hall coefficient measurements, as well as high frequency impedance analysis. Gamma spectroscopy is also used to examine residual radioactivity resulting from irradiation activation of elements in the samples. Hall coefficient measurements produce the expected trend of increasing carrier concentration with higher fluence levels, while high frequency impedance analysis shows an increase in sample impedance with increasing fluence.

Studies of neutron and proton nuclear activation in low-Earth orbit

NASA Technical Reports Server (NTRS)

Laird, C. E.

1982-01-01

The expected induced radioactivity of experimental material in low Earth orbit was studied for characteristics of activating particles such as cosmic rays, high energy Earth albedo neutrons, trapped protons, and secondary protons and neutrons. The activation cross sections for the production of long lived radioisotopes and other existing nuclear data appropriate to the study of these reactions were compiled. Computer codes which are required to calculate the expected activation of orbited materials were developed. The decreased computer code used to predict the activation of trapped protons of materials placed in the expected orbits of LDEF and Spacelab II. Techniques for unfolding the fluxes of activating particles from the measured activation of orbited materials are examined.

Research on water discharge characteristics of PEM fuel cells by using neutron imaging technology at the NRF, HANARO.

PubMed

Kim, TaeJoo; Sim, CheulMuu; Kim, MooHwan

2008-05-01

An investigation into the water discharge characteristics of proton exchange membrane (PEM) fuel cells is carried out by using a feasibility test apparatus and the Neutron Radiography Facility (NRF) at HANARO. The feasibility test apparatus was composed of a distilled water supply line, a compressed air supply line, heating systems, and single PEM fuel cells, which were a 1-parallel serpentine type with a 100 cm(2) active area. Three kinds of methods were used: compressed air supply-only; heating-only; and a combination of the methods of a compressed air supply and heating, respectively. The resultant water discharge characteristics are different according to the applied methods. The compressed air supply only is suitable for removing the water at a flow field and a heating only is suitable for water at the MEA. Therefore, in order to remove all the water at PEM fuel cells, the combination method is needed at the moment.

Preface: Proceedings of the 4th European Conference on Neutron Scattering (Lund, Sweden, 25 29 June 2007)

NASA Astrophysics Data System (ADS)

Rennie, Adrian R.

2008-03-01

Approximately 700 delegates came to the small university city of Lund in southern Sweden at the end of June 2007 to attend the 4th European Conference on Neutron Scattering. The majority of these participants are primarily interested in specific areas of condensed matter science and use neutron techniques as a powerful tool to study the structure and dynamic behaviour of materials. These range from liquids, superconductors, magnetic materials and archaeological artefacts. The diversity of scientific problems is reflected by the attendance of many laboratories with specializations in numerous different disciplines. The maturity of the technique is shown by the fact that neutron scattering is now applied widely in so many areas. Most results from neutron scattering experiements are published as articles that primarily relate to a specific scientific discipline in the context of problem oriented research. The neutron scattering conference provided an opportunity to exchange ideas between different fields. It is hoped that this collection of papers, from participants that submitted articles on applications of neutron scattering, will continue to promote the exchange of ideas for new studies, as was seen at the conference. The papers that describe instrumentation and advances in methods of neutron scattering will appear separately in Measurement Science and Technology Worldwide activity in developing new facilities for neutron scattering and the motivation for substantial projects, such as the new target station at the ISIS facility in the UK or the proposed European Spallation Source, comes from unique information obtained from working with neutrons. The results reported in the following papers show that there is substantial exciting work still to be performed as the community of users expands into new fields. The participants, as well as the organizers, are extremely grateful to the numerous sponsors that helped to make the conference a resounding success. We are grateful to IOP Publishing for agreeing to publish these papers, and for their friendly service, and prompt and efficient organization of the refereeing and editorial process.

Au Foil Activation Measurement and Simulation of the Concrete Neutron Shielding Ability for the Proposed New SANRAD Facility

NASA Astrophysics Data System (ADS)

Radebe, M. J.; Korochinsky, S.; Strydom, W. J.; De Beer, F. C.

The purpose of this study was to measure the effective neutron shielding characteristics of the new shielding material designed and manufactured to be used for the construction of the new SANRAD facility at Necsa, South Africa, through Au foil activation as well as MCNP simulations. The shielding capability of the high density shielding material was investigated in the worst case region (the neutron beam axis) of the experimental chamber for two operational modes. The everyday operational mode includes the 15 cm thick poly crystalline Bismuth filter at room temperature (assumed) to filter gamma-rays and some neutron spectrum energies. The second mode, dynamic imaging, will be conducted without the Bi-filter. The objective was achieved through a foil activation measurement at the current SANRAD facility and MCNP calculations. Several Au foilswere imbedded at different thicknesses(two at each position) of shielding material up to 80 cm thick to track the attenuation of the neutron beam over distance within the shielding material. The neutron flux and subsequently the associated dose rates were calculated from the activation levels of the Au foils. The concrete shielding material was found to provide adequate shielding for all energies of neutrons emerging from beam port no-2 of the SAFARI-1 research reactorwithin a thickness of 40 cm of concrete.

Neutronic investigation and activation calculation for CFETR HCCB blankets

NASA Astrophysics Data System (ADS)

Shuling, XU; Mingzhun, LEI; Sumei, LIU; Kun, LU; Kun, XU; Kun, PEI

2017-12-01

The neutronic calculations and activation behavior of the proposed helium cooled ceramic breeder (HCCB) blanket were predicted for the Chinese Fusion Engineering Testing Reactor (CFETR) design model using the MCNP multi-particle transport code and its associated data library. The tritium self-sufficiency behavior of the HCCB blanket was assessed, addressing several important breeding-related arrangements inside the blankets. Two candidate first wall armor materials were considered to obtain a proper tritium breeding ratio (TBR). Presentations of other neutronic characteristics, including neutron flux, neutron-induced damages in terms of the accumulated dpa and helium production were also conducted. Activation, decay heat levels and contact dose rates of the components were calculated to estimate the neutron-induced radioactivity and personnel safety. The results indicate that neutron radiation is efficiently attenuated and slowed down by components placed between the plasma and toroidal field coil. The dominant nuclides and corresponding isotopes in the structural steel were discussed. A radioactivity comparison between pure beryllium and beryllium with specific impurities was also performed. After a millennium cooling time, the decay heat of all the concerned components and materials is less than 1 × 10-4 kW, and most associated in-vessel components qualify for recycling by remote handling. The results demonstrate that acceptable hands-on recycling and operation still require a further long waiting period to allow the activated products to decay.

NOTE: Total body-calcium measurements: comparison of two delayed-gamma neutron activation facilities

NASA Astrophysics Data System (ADS)

Ma, R.; Ellis, K. J.; Yasumura, S.; Shypailo, R. J.; Pierson, R. N., Jr.

1999-06-01

This study compares two independently calibrated delayed-gamma neutron activation (DGNA) facilities, one at the Brookhaven National Laboratory (BNL), Upton, New York, and the other at the Children's Nutrition Research Center (CNRC), Houston, Texas that measure total body calcium (TBCa). A set of BNL phantoms was sent to CNRC for neutron activation analysis, and a set of CNRC phantoms was measured at BNL. Both facilities showed high precision (<2%), and the results were in good agreement, within 5%.

Measurement techniques for trace metals in coal-plant effluents: A brief review

NASA Technical Reports Server (NTRS)

Singh, J. J.

1979-01-01

The strong features and limitations of techniques for determining trace elements in aerosols emitted from coal plants are discussed. Techniques reviewed include atomic absorption spectroscopy, charged particle scattering and activation, instrumental neutron activation analysis, gas/liquid chromatography, gas chromatographic/mass spectrometric methods, X-ray fluorescence, and charged-particle-induced X-ray emission. The latter two methods are emphasized. They provide simultaneous, sensitive multielement analyses and lend themselves readily to depth profiling. It is recommended that whenever feasible, two or more complementary techniques should be used for analyzing environmental samples.

Radiocarbon Production by Thunderstorms

NASA Astrophysics Data System (ADS)

Babich, L. P.

2017-11-01

In view of the neutron flux enhancements observed in thunderstorms, a contribution of thunderstorm neutrons to atmospheric radiocarbon (isotope 614C) production is analyzed in connection with the archaeometry. Herein, estimates of neutron fluence per lightning electromagnetic pulse in regions with severe thunderstorm activity, at which a local rate of the 614C production is comparable to the observed rates, are shown to be consistent with the measured magnitudes of thunderstorm neutron fluence. At present, available observations of atmospheric neutron and parent gamma ray flashes correlated with thunderstorms do not allow making final conclusions about thunderstorm contributions to 614C production. For this, numerous studies of high-energy phenomena in thunderstorms are required, especially in the tropical belt where the thunderstorm activity is especially severe and where the 614C production by galactic cosmic rays is almost independent of the solar activity disturbing the Earth's magnetic field shielding the Earth from cosmic rays.

In-vivo assessment of total body protein in rats by prompt-γ neutron activation analysis

NASA Astrophysics Data System (ADS)

Stamatelatos, Ion E.; Boozer, Carol N.; Ma, Ruimei; Yasumura, Seiichi

1997-02-01

A prompt-(gamma) neutron activation analysis facility for in vivo determination of total body protein (TBP) in rats has been designed. TBP is determined in vivo by assessment of total body nitrogen. The facility is based on a 252Cf radionuclide neutron source within a heavy water moderator assembly and two NaI(Tl) scintillation detectors. The in vivo precision of the technique, as estimated by three repeated measurements of 15 rats is 6 percent, for a radiation dose equivalent of 60 mSv. The radiation dose per measurement is sufficiently low to enable serial measurements on the same animal. MCNP-4A Monte Carlo transport code was utilized to calculate thermal neutron flux correction factors to account for differences in size and shape of the rats and calibration phantoms. Good agrement was observed in comparing body nitrogen assessment by prompt-(gamma) neutron activation and chemical carcass analysis.

Experimental Measurements at the MASURCA Facility

NASA Astrophysics Data System (ADS)

Assal, W.; Bosq, J. C.; Mellier, F.

2012-12-01

Dedicated to the neutronics studies of fast and semi-fast reactor lattices, MASURCA (meaning “mock-up facility for fast breeder reactor studies at CADARACHE”) is an airflow cooled fast reactor operating at a maximum power of 5 kW playing an important role in the CEA research activities. At this facility, a lot of neutron integral experimental programs were undertaken. The purpose of this poster is to show a panorama of the facility from this experimental measurement point of view. A hint at the forthcoming refurbishment will be included. These programs include various experimental measurements (reactivity, distributions of fluxes, reaction rates), performed essentially with fission chambers, in accordance with different methods (noise methods, radial or axial traverses, rod drops) and involving several devices systems (monitors, fission chambers, amplifiers, power supplies, data acquisition systems ...). For this purpose are implemented electronics modules to shape the signals sent from the detectors in various mode (fluctuation, pulse, current). All the electric and electronic devices needed for these measurements and the relating wiring will be fully explained through comprehensive layouts. Data acquired during counting performed at the time of startup phase or rod drops are analyzed by the mean of a Neutronic Measurement Treatment (TMN in French) programmed on the basis of the MATLAB software. This toolbox gives the opportunity of data files management, reactivity valuation from neutronics measurements and transient or divergence simulation at zero power. Particular TMN using at MASURCA will be presented.

FAST NEUTRON DOSIMETER FOR HIGH TEMPERATURE OPERATION BY MEASUREMENT OF THE AMOUNT OF CESIUM 137 FORMED FROM A THORIUM WIRE

DOEpatents

McCune, D.A.

1964-03-17

A method and device for measurement of integrated fast neutron flux in the presence of a large thermal neutron field are described. The device comprises a thorium wire surrounded by a thermal neutron attenuator that is, in turn, enclosed by heat-resistant material. The method consists of irradiating the device in a neutron field whereby neutrons with energies in excess of 1.1 Mev cause fast fissions in the thorium, then removing the thorium wire, separating the cesium-137 fission product by chemical means from the thorium, and finally counting the radioactivity of the cesium to determine the number of fissions which have occurred so that the integrated fast flux may be obtained. (AEC)

Particle discrimination of NaI(Tl) scintillator under high-energy neutron field to measure the photon energy spectrum

NASA Astrophysics Data System (ADS)

Kamada, So; Takada, Masashi; Suzuki, Toshikazu

2014-09-01

Photons are measured separately from neutrons in high-energy neutron fields using a NaI(Tl) scintillator, 7.62 cm in diameter and 7.62 cm in length, combined with a pulse-shape discrimination method. The particle discrimination capability for this scintillator is confirmed using a time-of-flight method. Neutron fields were produced by irradiating Li targets with 40 and 80 MeV proton beams at the cyclotron facility in the National Institute of Radiological Sciences. Figures of merit corresponding to particle discrimination for the scintillator at the two neutron fields are improved with higher neutron energies. Photon energy spectra for energies over 6.5 MeV can be measured using the NaI(Tl) scintillator.

Thermal neutron shield and method of manufacture

DOEpatents

Brindza, Paul Daniel; Metzger, Bert Clayton

2013-05-28

A thermal neutron shield comprising concrete with a high percentage of the element Boron. The concrete is least 54% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of Boron loaded concrete which includes enriching the concrete mixture with varying grit sizes of Boron Carbide.

Optimization of the beam shaping assembly in the D-D neutron generators-based BNCT using the response matrix method.

PubMed

Kasesaz, Y; Khalafi, H; Rahmani, F

2013-12-01

Optimization of the Beam Shaping Assembly (BSA) has been performed using the MCNP4C Monte Carlo code to shape the 2.45 MeV neutrons that are produced in the D-D neutron generator. Optimal design of the BSA has been chosen by considering in-air figures of merit (FOM) which consists of 70 cm Fluental as a moderator, 30 cm Pb as a reflector, 2mm (6)Li as a thermal neutron filter and 2mm Pb as a gamma filter. The neutron beam can be evaluated by in-phantom parameters, from which therapeutic gain can be derived. Direct evaluation of both set of FOMs (in-air and in-phantom) is very time consuming. In this paper a Response Matrix (RM) method has been suggested to reduce the computing time. This method is based on considering the neutron spectrum at the beam exit and calculating contribution of various dose components in phantom to calculate the Response Matrix. Results show good agreement between direct calculation and the RM method. Copyright © 2013 Elsevier Ltd. All rights reserved.

New applications and developments in the neutron shielding

NASA Astrophysics Data System (ADS)

Uğur, Fatma Aysun

2017-09-01

Shielding neutrons involve three steps that are slowing neutrons, absorption of neutrons, and impregnation of gamma rays. Neutrons slow down with thermal energy by hydrogen, water, paraffin, plastic. Hydrogenated materials are also very effective for the absorption of neutrons. Gamma rays are produced by neutron (radiation) retention on the neutron shield, inelastic scattering, and degradation of activation products. If a source emits gamma rays at various energies, high-energy gamma rays sometimes specify shielding requirements. Multipurpose Materials for Neutron Shields; Concrete, especially with barium mixed in, can slow and absorb the neutrons, and shield the gamma rays. Plastic with boron is also a good multipurpose shielding material. In this study; new applications and developments in the area of neutron shielding will be discussed in terms of different materials.

A method for moisture measurement in porous media based on epithermal neutron scattering.

PubMed

El Abd, A

2015-11-01

A method for moisture measurement in porous media was proposed. A wide beam of epithermal neutrons was obtained from a Pu-Be neutron source immersed in a cylinder made of paraffin wax. (3)He detectors (four or six) arranged in the backward direction of the incident beam were used to record scattered neutrons from investigated samples. Experiments of water absorption into clay and silicate bricks, and a sand column were investigated by neutron scattering. While the samples were absorbing water, scattered neutrons were recorded from fixed positions along the water flow direction. It was observed that, at these positions scattered neutrons increase as the water uptake increases. Obtained results are discussed in terms of the theory of macroscopic flow in porous media. It was shown that, the water absorption processes were Fickian and non Fickian in the sand column and brick samples, respectively. The advantages of applying the proposed method to study fast as well as slow flow processes in porous media are discussed. Copyright © 2015 Elsevier Ltd. All rights reserved.

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