Measurement of the Reactor Antineutrino Flux and Spectrum at Daya Bay

NASA Astrophysics Data System (ADS)

An, F. P.; Balantekin, A. B.; Band, H. R.; Bishai, M.; Blyth, S.; Butorov, I.; Cao, D.; Cao, G. F.; Cao, J.; Cen, W. R.; Chan, Y. L.; Chang, J. F.; Chang, L. C.; Chang, Y.; Chen, H. S.; Chen, Q. Y.; Chen, S. M.; Chen, Y. X.; Chen, Y.; Cheng, J. H.; Cheng, J.; Cheng, Y. P.; Cherwinka, J. J.; Chu, M. C.; Cummings, J. P.; de Arcos, J.; Deng, Z. Y.; Ding, X. F.; Ding, Y. Y.; Diwan, M. V.; Dove, J.; Draeger, E.; Dwyer, D. A.; Edwards, W. R.; Ely, S. R.; Gill, R.; Gonchar, M.; Gong, G. H.; Gong, H.; Grassi, M.; Gu, W. Q.; Guan, M. Y.; Guo, L.; Guo, X. H.; Hackenburg, R. W.; Han, R.; Hans, S.; He, M.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, L. M.; Hu, L. J.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. X.; Huang, X. T.; Huber, P.; Hussain, G.; Jaffe, D. E.; Jaffke, P.; Jen, K. L.; Jetter, S.; Ji, X. P.; Ji, X. L.; Jiao, J. B.; Johnson, R. A.; Kang, L.; Kettell, S. H.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lei, R. T.; Leitner, R.; Leung, K. Y.; Leung, J. K. C.; Lewis, C. A.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, S. C.; Li, W. D.; Li, X. N.; Li, X. Q.; Li, Y. F.; Li, Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, P. Y.; Lin, S. K.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, D. W.; Liu, H.; Liu, J. L.; Liu, J. C.; Liu, S. S.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, K. B.; Ma, Q. M.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Martinez Caicedo, D. A.; McDonald, K. T.; McKeown, R. D.; Meng, Y.; Mitchell, I.; Monari Kebwaro, J.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Ngai, H. Y.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevski, A.; Pan, H.-R.; Park, J.; Patton, S.; Pec, V.; Peng, J. C.; Piilonen, L. E.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, B.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Shao, B. B.; Steiner, H.; Sun, G. X.; Sun, J. L.; Tang, W.; Taychenachev, D.; Tsang, K. V.; Tull, C. E.; Tung, Y. C.; Viaux, N.; Viren, B.; Vorobel, V.; Wang, C. H.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, W. W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Wei, H. Y.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wise, T.; Wong, H. L. H.; Wong, S. C. F.; Worcester, E.; Wu, Q.; Xia, D. M.; Xia, J. K.; Xia, X.; Xing, Z. Z.; Xu, J. Y.; Xu, J. L.; Xu, J.; Xu, Y.; Xue, T.; Yan, J.; Yang, C. G.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Yeh, M.; Young, B. L.; Yu, G. Y.; Yu, Z. Y.; Zang, S. L.; Zhan, L.; Zhang, C.; Zhang, H. H.; Zhang, J. W.; Zhang, Q. M.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhao, Q. W.; Zhao, Y. F.; Zhao, Y. B.; Zheng, L.; Zhong, W. L.; Zhou, L.; Zhou, N.; Zhuang, H. L.; Zou, J. H.; Daya Bay Collaboration

2016-02-01

This Letter reports a measurement of the flux and energy spectrum of electron antineutrinos from six 2.9 GWt h nuclear reactors with six detectors deployed in two near (effective baselines 512 and 561 m) and one far (1579 m) underground experimental halls in the Daya Bay experiment. Using 217 days of data, 296 721 and 41 589 inverse β decay (IBD) candidates were detected in the near and far halls, respectively. The measured IBD yield is (1.55 ±0.04 ) ×10-18 cm2 GW-1 day-1 or (5.92 ±0.14 ) ×10-43 cm2 fission-1 . This flux measurement is consistent with previous short-baseline reactor antineutrino experiments and is 0.946 ±0.022 (0.991 ±0.023 ) relative to the flux predicted with the Huber -Mueller (ILL -Vogel ) fissile antineutrino model. The measured IBD positron energy spectrum deviates from both spectral predictions by more than 2 σ over the full energy range with a local significance of up to ˜4 σ between 4-6 MeV. A reactor antineutrino spectrum of IBD reactions is extracted from the measured positron energy spectrum for model-independent predictions.

Measurement of the reactor antineutrino flux and spectrum at Daya Bay

DOE PAGES

D. E. Jaffe; Bishai, M; Diwan, M.; ...

2016-02-12

This Letter reports a measurement of the flux and energy spectrum of electron antineutrinos from six 2.9~GW th nuclear reactors with six detectors deployed in two near (effective baselines 512~m and 561~m) and one far (1,579 m) underground experimental halls in the Daya Bay experiment. Using 217 days of data, 296,721 and 41,589 inverse beta decay (IBD) candidates were detected in the near and far halls, respectively. The measured IBD yield is (1.55 ± 0.04) × 10 –18 cm 2/GW/day or (5.92 ± 0.14) × 10 –43 cm 2/fission. This flux measurement is consistent with previous short-baseline reactor antineutrino experimentsmore » and is 0.946 ± 0.022 (0.991 ± 0.023) relative to the flux predicted with the Huber+Mueller (ILL+Vogel) fissile antineutrino model. The measured IBD positron energy spectrum deviates from both spectral predictions by more than 2σ over the full energy range with a local significance of up to ~4σ between 4-6 MeV. Furthermore, a reactor antineutrino spectrum of IBD reactions is extracted from the measured positron energy spectrum for model-independent predictions.« less

A Measurement of the Absolute Reactor Antineutrino Flux and Spectrum at Daya Bay

NASA Astrophysics Data System (ADS)

An, Fengpeng

2017-12-01

The Daya Bay Reactor Neutrino Experiment uses an array of eight underground detectors to study antineutrinos from six reactor cores with different baselines. Since the start of data-taking from late 2011, Daya Bay has collected the largest sample of reactor antineutrino events to date, and has made the most precise measurement of the neutrino oscillation parameters sin22θ13 and Δm2ee. Using the data from the four detectors in the near experimental halls, Daya Bay has made a high statistics measurement of the absolute reactor antineutrino flux and spectrum. In this paper we will present this measurement and its comparison to predictions based on different flux models.

Reactivity Initiated Accident Simulation to Inform Transient Testing of Candidate Advanced Cladding

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

Brown, Nicholas R; Wysocki, Aaron J; Terrani, Kurt A

2016-01-01

Abstract. Advanced cladding materials with potentially enhanced accident tolerance will yield different light water reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to different cladding material properties and responses to the transient, and to some extent, reactor physics, thermal, and hydraulic characteristics. Some of the differences in reactors physics characteristics will be driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and others will be driven by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermalmore » hydraulic limits after transition from the current zirconium-based cladding to the advanced materials will also affect the transient response of the integral fuel. This paper leverages three-dimensional reactor core simulation capabilities to inform on appropriate experimental test conditions for candidate advanced cladding materials in a control rod ejection event. These test conditions are using three-dimensional nodal kinetics simulations of a reactivity initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon carbide based (SiC-SiC) cladding materials. The effort yields boundary conditions for experimental mechanical tests, specifically peak cladding strain during the power pulse following the rod ejection. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus reference zirconium cladding is predominantly due to differences in: (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores due to hardened (or softened) spectrum. This study shows minimal impact of SiC-based cladding configurations on the transient response versus reference zirconium-based cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. Therefore the FeCrAl-based cases have a more rapid fuel thermal expansion rate and the resultant pellet-cladding interaction occurs more rapidly.« less

Liquid fuel molten salt reactors for thorium utilization

DOE PAGES

Gehin, Jess C.; Powers, Jeffrey J.

2016-04-08

Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with themore » online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing thorium-based MSRs.« 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).

Characterization of the Caliban and Prospero Critical Assemblies Neutron Spectra for Integral Measurements Experiments

NASA Astrophysics Data System (ADS)

Casoli, P.; Authier, N.; Jacquet, X.; Cartier, J.

2014-04-01

Caliban and Prospero are two highly enriched uranium metallic core reactors operated on the CEA Center of Valduc. These critical assemblies are suitable for integral experiments, such as fission yields measurements or perturbation measurements, which have been carried out recently on the Caliban reactor. Different unfolding methods, based on activation foils and fission chambers measurements, are used to characterize the reactor spectra and especially the Caliban spectrum, which is very close to a pure fission spectrum.

MELCOR computer code manuals

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

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

The PROSPECT physics program

NASA Astrophysics Data System (ADS)

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

2016-11-01

The precision reactor oscillation and spectrum experiment, PROSPECT, is designed to make a precise measurement of the antineutrino spectrum from a highly-enriched uranium reactor and probe eV-scale sterile neutrinos by searching for neutrino oscillations over a distance of several meters. PROSPECT is conceived as a 2-phase experiment utilizing segmented 6Li-doped liquid scintillator detectors for both efficient detection of reactor antineutrinos through the inverse beta decay reaction and excellent background discrimination. PROSPECT Phase I consists of a movable 3 ton antineutrino detector at distances of 7-12 m from the reactor core. It will probe the best-fit point of the {ν }e disappearance experiments at 4σ in 1 year and the favored region of the sterile neutrino parameter space at \\gt 3σ in 3 years. With a second antineutrino detector at 15-19 m from the reactor, Phase II of PROSPECT can probe the entire allowed parameter space below 10 eV2 at 5σ in 3 additional years. The measurement of the reactor antineutrino spectrum and the search for short-baseline oscillations with PROSPECT will test the origin of the spectral deviations observed in recent {θ }13 experiments, search for sterile neutrinos, and conclusively address the hypothesis of sterile neutrinos as an explanation of the reactor anomaly.

Spectral structure of electron antineutrinos from nuclear reactors.

PubMed

Dwyer, D A; Langford, T J

2015-01-09

Recent measurements of the positron energy spectrum obtained from inverse beta decay interactions of reactor electron antineutrinos show an excess in the 4 to 6 MeV region relative to current predictions. First-principles calculations of fission and beta decay processes within a typical pressurized water reactor core identify prominent fission daughter isotopes as a possible origin for this excess. These calculations also predict percent-level substructures in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of these substructures can elucidate the nuclear processes occurring within reactors. These substructures can be a systematic issue for measurements utilizing the detailed spectral shape.

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

Gehin, Jess C.; Powers, Jeffrey J.

Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with themore » online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing thorium-based MSRs.« less

Long lifetime fast spectrum reactor for lunar surface power system

NASA Astrophysics Data System (ADS)

Kambe, Mitsuru

1993-01-01

In the framework of innovative reactor research activities, a conceptual design study of fast spectrum reactor and primary system for 800 kWe lunar surface power system to be combined with potassium Rankine cycle power conversion has been conducted to meet the power requirements of the lunar base activities in the next century. The reactor subsystem is characterized by RAPID (Refueling by All Pins Integrated Design) concept to enhance inherent safety and to enable quick and simplifed refueling in every 10 years. RAPID concept affords power plant design lifetime of up to 30 years. Integrity of the reactor structure and replacement of failed primary circuits are also discussed. Substantial reduction in per-kWh cost on considering launch, emplacement, and final disposition can be expected by a long system lifetime.

The PROSPECT physics program

DOE PAGES

Ashenfelter, J.; Balantekin, A. B.; Band, H. R.; ...

2016-10-17

The precision reactor oscillation and spectrum experiment, PROSPECT, is designed to make a precise measurement of the antineutrino spectrum from a highly-enriched uranium reactor and probe eV-scale sterile neutrinos by searching for neutrino oscillations over a distance of several meters. The subject of this paper, PROSPECT, is conceived as a 2-phase experiment utilizing segmented 6Li-doped liquid scintillator detectors for both efficient detection of reactor antineutrinos through the inverse beta decay reaction and excellent background discrimination. PROSPECT Phase I consists of a movable 3 ton antineutrino detector at distances of 7–12 m from the reactor core. It will probe the best-fitmore » point of the ν e disappearance experiments at 4σ in 1 year and the favored region of the sterile neutrino parameter space at > 3σ in 3 years. With a second antineutrino detector at 15–19 m from the reactor, Phase II of PROSPECT can probe the entire allowed parameter space below 10 eV 2 at 5σ in 3 additional years. Finally, the measurement of the reactor antineutrino spectrum and the search for short-baseline oscillations with PROSPECT will test the origin of the spectral deviations observed in recent θ 13 experiments, search for sterile neutrinos, and conclusively address the hypothesis of sterile neutrinos as an explanation of the reactor anomaly.« less

Measurement of 89Y(n,2n) spectral averaged cross section in LR-0 special core reactor spectrum

NASA Astrophysics Data System (ADS)

Košťál, Michal; Losa, Evžen; Baroň, Petr; Šolc, Jaroslav; Švadlenková, Marie; Koleška, Michal; Mareček, Martin; Uhlíř, Jan

2017-12-01

The present paper describes reaction rate measurement of 89Y(n,2n)88Y in a well-defined reactor spectrum of a special core assembled in the LR-0 reactor and compares this value with results of simulation. The reaction rate is derived from the measurement of activity of 88Y using gamma-ray spectrometry of irradiated Y2O3 sample. The resulting cross section value averaged in spectrum is 43.9 ± 1.5 μb, averaged in the 235U spectrum is 0.172 ± 0.006 mb. This cross-section is important as it is used as high energy neutron monitor and is therefore included in the International Reactor Dosimetry and Fusion File. Calculations of reaction rates were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF nuclear data libraries. The agreement with uranium description by CIELO library is very good, while in ENDF/B-VII.0 description of uranium, underprediction about 10% in average can be observed.

Method to Reduce Long-lived Fission Products by Nuclear Transmutations with Fast Spectrum Reactors.

PubMed

Chiba, Satoshi; Wakabayashi, Toshio; Tachi, Yoshiaki; Takaki, Naoyuki; Terashima, Atsunori; Okumura, Shin; Yoshida, Tadashi

2017-10-24

Transmutation of long-lived fission products (LLFPs: 79 Se, 93 Zr, 99 Tc, 107 Pd, 129 I, and 135 Cs) into short-lived or non-radioactive nuclides by fast neutron spectrum reactors without isotope separation has been proposed as a solution to the problem of radioactive wastes disposal. Despite investigation of many methods, such transmutation remains technologically difficult. To establish an effective and efficient transmutation system, we propose a novel neutron moderator material, yttrium deuteride (YD 2 ), to soften the neutron spectrum leaking from the reactor core. Neutron energy spectra and effective half-lives of LLFPs, transmutation rates, and support ratios were evaluated with the continuous-energy Monte Carlo code MVP-II/MVP-BURN and the JENDL-4.0 cross section library. With the YD 2 moderator in the radial blanket and shield regions, effective half-lives drastically decreased from 106 to 102 years and the support ratios reached 1.0 for all six LLFPs. This successful development and implementation of a transmutation system for LLFPs without isotope separation contributes to a the ability of fast spectrum reactors to reduce radioactive waste by consuming their own LLFPs.

Regulating the Emission Spectrum of CsPbBr3 from Green to Blue via Controlling the Temperature and Velocity of Microchannel Reactor

PubMed Central

Tang, Yong; Lu, Hanguang; Rao, Longshi; Ding, Xinrui; Yan, Caiman; Yu, Binhai

2018-01-01

The ability to precisely obtain tunable spectrum of lead halide perovskite quantum dots (QDs) is very important for applications, such as in lighting and display. Herein, we report a microchannel reactor method for synthesis of CsPbBr3 QDs with tunable spectrum. By adjusting the temperature and velocity of the microchannel reactor, the emission peaks of CsPbBr3 QDs ranging from 520 nm to 430 nm were obtained, which is wider than that of QDs obtained in a traditional flask without changing halide component. The mechanism of photoluminescence (PL) spectral shift of CsPbBr3 QDs was investigated, the result shows that the supersaturation control enabled by the superior mass and heat transfer performance in the microchannel is the key to achieve the wide range of PL spectrum, with only a change in the setting of the temperature controller required. The wide spectrum of CsPbBr3 QDs can be applied to light-emitting diodes (LEDs), photoelectric sensors, lasers, etc. PMID:29498710

Measurement of the 23Na(n,2n) cross section in 235U and 252Cf fission neutron spectra

NASA Astrophysics Data System (ADS)

Košťál, Michal; Schulc, Martin; Rypar, Vojtěch; Losa, Evžen; Švadlenková, Marie; Baroň, Petr; Jánský, Bohumil; Novák, Evžen; Mareček, Martin; Uhlíř, Jan

2017-09-01

The presented paper aims to compare the calculated and experimental reaction rates of 23Na(n,2n)22Na in a well-defined reactor spectra and in the spontaneous fission spectrum of 252Cf. The experimentally determined reaction rate, derived using gamma spectroscopy of irradiated NaF sample, is used for average cross section determination.Estimation of this cross-section is important as it is included in International Reactor Dosimetry and Fusion File and is also relevant to the correct estimation of long-term activity of Na coolant in Sodium Fast Reactors. The calculations were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF nuclear data libraries. In the case of reactor spectrum, reasonable agreement was not achieved with any library. However, in the case of 252Cf spectrum agreement was achieved with IRDFF, JEFF-3.1 and JENDL libraries.

Analyzing the thermionic reactor critical experiments. [thermal spectrum of uranium 235 core

NASA Technical Reports Server (NTRS)

Niederauer, G. F.

1973-01-01

The Thermionic Reactor Critical Experiments (TRCE) consisted of fast spectrum highly enriched U-235 cores reflected by different thicknesses of beryllium or beryllium oxide with a transition zone of stainless steel between the core and reflector. The mixed fast-thermal spectrum at the core reflector interface region poses a difficult neutron transport calculation. Calculations of TRCE using ENDF/B fast spectrum data and GATHER library thermal spectrum data agreed within about 1 percent for the multiplication factor and within 6 to 8 percent for the power peaks. Use of GAM library fast spectrum data yielded larger deviations. The results were obtained from DOT R Theta calculations with leakage cross sections, by region and by group, extracted from DOT RZ calculations. Delineation of the power peaks required extraordinarily fine mesh size at the core reflector interface.

MELCOR computer code manuals: Primer and user`s guides, Version 1.8.3 September 1994. Volume 1

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

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users` Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Improved measurement of the reactor antineutrino flux and spectrum at Daya Bay

NASA Astrophysics Data System (ADS)

An, F. P.; Balantekin, A. B.; Band, H. R.; Bishai, M.; Blyth, S.; Cao, D.; Cao, G. F.; Cao, J.; Cen, W. R.; Chan, Y. L.; Chang, J. F.; Chang, L. C.; Chang, Y.; Chen, H. S.; Chen, Q. Y.; Chen, S. M.; Chen, Y. X.; Chen, Y.; Cheng, J.-H.; Cheng, J.; Cheng, Y. P.; Cheng, Z. K.; Cherwinka, J. J.; Chu, M. C.; Chukanov, A.; Cummings, J. P.; de Arcos, J.; Deng, Z. Y.; Ding, X. F.; Ding, Y. Y.; Diwan, M. V.; Dolgareva, M.; Dove, J.; Dwyer, D. A.; Edwards, W. R.; Gill, R.; Gonchar, M.; Gong, G. H.; Gong, H.; Grassi, M.; Gu, W. Q.; Guan, M. Y.; Guo, L.; Guo, R. P.; Guo, X. H.; Guo, Z.; Hackenburg, R. W.; Han, R.; Hans, S.; He, M.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. X.; Huang, X. T.; Huber, P.; Huo, W.; Hussain, G.; Jaffe, D. E.; Jaffke, P.; Jen, K. L.; Jetter, S.; Ji, X. P.; Ji, X. L.; Jiao, J. B.; Johnson, R. A.; Jones, D.; Joshi, J.; Kang, L.; Kettell, S. H.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, J. H. C.; Lei, R. T.; Leitner, R.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, S.; Li, S. C.; Li, W. D.; Li, X. N.; Li, Y. F.; Li, Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, S.; Lin, S. K.; Lin, Y.-C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, D. W.; Liu, J. L.; Liu, J. C.; Loh, C. W.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, K. B.; Lv, Z.; Ma, Q. M.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Malyshkin, Y.; Martinez Caicedo, D. A.; McDonald, K. T.; McKeown, R. D.; Mitchell, I.; Mooney, M.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Ngai, H. Y.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pan, H.-R.; Park, J.; Patton, S.; Pec, V.; Peng, J. C.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Steiner, H.; Sun, G. X.; Sun, J. L.; Tang, W.; Taychenachev, D.; Treskov, K.; Tsang, K. V.; Tull, C. E.; Viaux, N.; Viren, B.; Vorobel, V.; Wang, C. H.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Wei, H. Y.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wise, T.; Wong, H. L. H.; Wong, S. C. F.; Worcester, E.; Wu, C.-H.; Wu, Q.; Wu, W. J.; Xia, D. M.; Xia, J. K.; Xing, Z. Z.; Xu, J. Y.; Xu, J. L.; Xu, Y.; Xue, T.; Yang, C. G.; Yang, H.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Ye, Z.; Yeh, M.; Young, B. L.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, H. H.; Zhang, J. W.; Zhang, Q. M.; Zhang, X. T.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhao, Q. W.; Zhao, Y. B.; Zhong, W. L.; Zhou, L.; Zhou, N.; Zhuang, H. L.; Zou, J. H.; Daya Bay Collaboration

2017-01-01

A new measurement of the reactor antineutrino flux and energy spectrum by the Daya Bay reactor neutrino experiment is reported. The antineutrinos were generated by six 2.9 GWth nuclear reactors and detected by eight antineutrino detectors deployed in two near (560 m and 600 m flux-weighted baselines) and one far (1640 m flux-weighted baseline) underground experimental halls. With 621 days of data, more than 1.2 million inverse beta decay (IBD) candidates were detected. The IBD yield in the eight detectors was measured, and the ratio of measured to predicted flux was found to be 0.946±0.020 (0.992±0.021) for the Huber+Mueller (ILL+Vogel) model. A 2.9σ deviation was found in the measured IBD positron energy spectrum compared to the predictions. In particular, an excess of events in the region of 4-6 MeV was found in the measured spectrum, with a local significance of 4.4σ. A reactor antineutrino spectrum weighted by the IBD cross section is extracted for model-independent predictions. Supported in part by the Ministry of Science and Technology of China, the United States Department of Energy, the Chinese Academy of Sciences, the CAS Center for Excellence in Particle Physics, the National Natural Science Foundation of China, the Guangdong provincial government, the Shenzhen municipal government, the China General Nuclear Power Group, the Research Grants Council of the Hong Kong Special Administrative Region of China, the MOST and MOE in Taiwan, the U.S. National Science Foundation, the Ministry of Education, Youth and Sports of the Czech Republic, the Joint Institute of Nuclear Research in Dubna, Russia, the NSFC-RFBR joint research program, the National Commission for Scientific and Technological Research of Chile

The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

NASA Astrophysics Data System (ADS)

Wagemans, Jan; Malambu, Edouard; Borms, Luc; Fiorito, Luca

2016-02-01

The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma) irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f) prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f) prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

REACTOR PHYSICS QUARTERLY REPORT JANUARY, FEBRUARY, MARCH 1970

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

Schmid, L. C.; Clayton, E. D.; Heineman, R. E.

1970-05-01

The objective of the Reactor Physics Quarterly Report is to inform the scientific community in a timely manner of the technical progress made on the many phases of reactor physics work within the laboratory. The report contains brief technical discussions of accomplishments in all areas where significant progress has been made during the quarter.

Total Absorption Spectroscopy Study of ⁹²Rb Decay: A Major Contributor to Reactor Antineutrino Spectrum Shape

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

Sonzogni, A.; Zakari-Issoufou, A. -A.; Fallot, M.

2015-03-09

The accurate determination of the emitted reactor antineutrino flux is still a major challenge for actual and future neutrino experiments at reactors, especially after the evidence of a disagreement between the measured antineutrino energy spectrum by Double Chooz, Daya Bay, and Reno and calculated antineutrino spectra obtained from the conversion of the unique integral beta spectra measured at the ILL reactor. Using nuclear data to compute reactor antineutrino spectra may help understanding this bias, with the study of the underlying nuclear physics. Summation calculations allow identifying a list of nuclei that contribute importantly to the antineutrino energy spectra emitted aftermore » the fission of ²³⁹ ,²⁴¹Pu and ²³⁵ ,²³⁸U, and whose beta decay properties might deserve new measurements. Among these nuclei, ⁹²Rb exhausts by itself about 16% of of the antineutrino energy spectrum emitted by Pressurized Water Reactors in the 5 to 8 MeV range. In this Letter, we report new Total Absorption Spectroscopy (TAS) results for this important contributor. The obtained beta feeding from ⁹²Rb shows beta intensity unobserved before in the 4.5 to 5.5 MeV energy region and gives a ground state to ground state branch of 87.5 % ± 3%. These new data induce a dramatic change in recent summation calculations where a 51% GS to GS branch was considered for ⁹²Rb, increasing the summation antineutrino spectrum in the region nearby the observed bias.The new data still have an important impact on other summation calculations in which more recent data were considered« less

Geoantineutrino spectrum and slow nuclear burning on the boundary of the liquid and solid phases of the Earth's core

NASA Astrophysics Data System (ADS)

Rusov, V. D.; Pavlovich, V. N.; Vaschenko, V. N.; Tarasov, V. A.; Zelentsova, T. N.; Bolshakov, V. N.; Litvinov, D. A.; Kosenko, S. I.; Byegunova, O. A.

2007-09-01

We give an alternative description of the data produced in the KamLAND experiment. Assuming the existence of a natural nuclear reactor on the boundary of the liquid and solid phases of the Earth's core, a geoantineutrino spectrum is obtained. This assumption is based on the experimental results of V. Anisichkin and his collaborators on the interaction of uranium dioxide and uranium carbide with iron-nickel and silica-alumina melts at high pressure (5-10 GPa) and temperature (1600-2200°C), which led to the proposal of the existence of an actinide shell in the Earth's core. We describe the operating mechanism of this reactor as solitary waves of nuclear burning in 238U and/or 232Th medium, in particular, as neutron fission progressive waves of Feoktistov and/or Teller et al. type. Next, we propose a simplified model for the accumulation and burn-up kinetics in Feoktistov's U-Pu fuel cycle. We also apply this model for numerical simulations of neutron fission wave in a two-phase UO2/Fe medium on the surface of the Earth's solid core. The proposed georeactor model offers a mechanism for the generation of 3He. The 3He/4He distribution in the Earth's interior is calculated, which in turn can be used as a natural quantitative criterion of the georeactor thermal power. Finally, we give a tentative estimation of the geoantineutrino intensity and spectrum on the Earth's surface. For this purpose we use the O'Nions et al. geochemical model of mantle differentiation and crust growth complemented by a nuclear energy source (georeactor with power of 30 TW).

Validation of IRDFF in 252Cf standard and IRDF-2002 reference neutron fields

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

Simakov, Stanislav; Capote Noy, Roberto; Greenwood, Lawrence R.

The results of validation of the latest release of International Reactor Dosimetry and Fusion File, IRDFF-1.03, in the standard 252Cf(s.f.) and reference 235U(nth,f) neutron benchmark fields are presented. The spectrum-averaged cross sections were shown to confirm the recommended spectrum for 252Cf spontaneous fission source; that was not the case for the current recommended spectra for 235U(nth,f). IRDFF was also validated in the spectra of the research reactor facilities ISNF, Sigma-Sigma and YAYOI, which are available in the IRDF- 2002 collection. Before this analysis, the ISFN spectrum was resimulated to remove unphysical oscillations in spectrum. IRDFF-1.03 was shown to reasonably reproducemore » the spectrum-averaged data measured in these fields except for the case of YAYOI.« less

Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra.

PubMed

Ďuran, I; Bolshakova, I; Viererbl, L; Sentkerestiová, J; Holyaka, R; Lahodová, Z; Bém, P

2010-10-01

We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10(16) cm(-2) was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

Fuel assembly for the production of tritium in light water reactors

DOEpatents

Cawley, W.E.; Trapp, T.J.

1983-06-10

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

Fuel assembly for the production of tritium in light water reactors

DOEpatents

Cawley, William E.; Trapp, Turner J.

1985-01-01

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors - Volume I.

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

Gauntt, Randall O.; Mattie, Patrick D.

Sandia National Laboratories (SNL) has conducted an uncertainty analysis (UA) on the Fukushima Daiichi unit (1F1) accident progression with the MELCOR code. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). That study focused on reconstructing the accident progressions, as postulated by the limited plant data. This work was focused evaluation of uncertainty in core damage progression behavior and its effect on key figures-of-merit (e.g., hydrogen production, reactor damage state, fraction of intact fuel, vessel lower head failure). The primary intent of this studymore » was to characterize the range of predicted damage states in the 1F1 reactor considering state of knowledge uncertainties associated with MELCOR modeling of core damage progression and to generate information that may be useful in informing the decommissioning activities that will be employed to defuel the damaged reactors at the Fukushima Daiichi Nuclear Power Plant. Additionally, core damage progression variability inherent in MELCOR modeling numerics is investigated.« less

LOS ALAMOS NEUTRON SCIENCE CENTER CONTRIBUTIONS TO THE DEVELOPMENT OF FUTURE POWER REACTORS

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

GAVRON, VICTOR I.; HILL, TONY S.; PITCHER, ERIC J.

The Los Alamos Neutron Science Center (LANSCE) is a large spallation neutron complex centered around an 800 MeV high-currently proton accelerator. Existing facilities include a highly-moderated neutron facility (Lujan Center) where neutrons between thermal and keV energies are produced, and the Weapons Neutron Research Center (WNR), where a bare spallation target produces neutrons between 0.1 and several hundred MeV.The LANSCE facility offers a unique capability to provide high precision nuclear data over a large energy region, including that for fast reactor systems. In an ongoing experimental program the fission and capture cross sections are being measured for a number ofmore » minor actinides relevant for Generation-IV reactors and transmutation technology. Fission experiments makes use of both the highly moderated spallation neutron spectrum at the Lujan Center, and the unmoderated high energy spectrum at WNR. By combininb measurements at these two facilities the differential fission cross section is measured relative to the {sup 235}U(n,f) standard from subthermal energies up to about 200 MeV. An elaborate data acquisition system is designed to deal with all the different types of background present when spanning 10 energy decades. The first isotope to be measured was {sup 237}Np, and the results were used to improve the current ENDF/B-VII evaluation. Partial results have also been obtained for {sup 240}Pu and {sup 242}Pu, and the final results are expected shortly. Capture cross sections are measured at LANSCE using the Detector for Advanced Neutron Capture Experiments (DANCE). This unique instrument is highly efficient in detecting radiative capture events, and can thus handle radioactive samples of half-lives as low as 100 years. A number of capture cross sections important to fast reaction applications have been measured with DANCE. The first measurement was on {sup 237}Np(n,{gamma}), and the results have been submitted for publication. Other capture measurements in progress include {sup 240}Pu and {sup 242}Pu. The United States recently announced the Global Nuclear Energy Partnership (GNEP), with the goal of closing the commercial nuclear fuel cycle while minimizing proliferation risk. GNEP achieves these goals using fast-spectrum nuclear reactors powered by new transmutation fuels that contain significant quantities of minor actinides. The proposed Materials Test Station (MTS) will provide the GNEP with a cost-effective means of obtaining domestic fast-spectrum irradiations of advanced transmutation fuel forms and structural materials, which is an important step in the fuels qualification process. The MTS will be located at the LANSCE, and will be driven by a 1.08-MW proton beam. Th epeak neutron flux in the irradiation region is 1.67 x 10{sup 15} n/cm{sup 2}/s, and the energy spectrum is similar to that of a fast reactor, with the addition of a high-energy tail. The facility is expected to operate at least 4,400 hours per year. Fuel burnup rates will exceed 4% per year, and the radiation damage rate in iron will be 18 dpa (displacements per atom) per year. The construction cost is estimated to be $73M (including 25% contingency), with annual operating costs in the range of $6M to $10M. Appropriately funded, the MTS could begin operation in 2010.« less

Calculated power distribution of a thermionic, beryllium oxide reflected, fast-spectrum reactor

NASA Technical Reports Server (NTRS)

Mayo, W.; Lantz, E.

1973-01-01

A procedure is developed and used to calculate the detailed power distribution in the fuel elements next to a beryllium oxide reflector of a fast-spectrum, thermionic reactor. The results of the calculations show that, although the average power density in these outer fuel elements is not far from the core average, the power density at the very edge of the fuel closest to the beryllium oxide is about 1.8 times the core avearge.

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

Thorium fueled reactor

NASA Astrophysics Data System (ADS)

Sipaun, S.

2017-01-01

Current development in thorium fueled reactors shows that they can be designed to operate in the fast or thermal spectrum. The thorium/uranium fuel cycle converts fertile thorium-232 into fissile uranium-233, which fissions and releases energy. This paper analyses the characteristics of thorium fueled reactors and discusses the thermal reactor option. It is found that thorium fuel can be utilized in molten salt reactors through many configurations and designs. A balanced assessment on the feasibility of adopting one reactor technology versus another could lead to optimized benefits of having thorium resource.

New results from RENO and prospects with RENO-50

NASA Astrophysics Data System (ADS)

Kim, Soo-Bong

2015-08-01

RENO (Reactor Experiment for Neutrino Oscillation) has made a definitive measurement of the smallest mixing angle θ13 in 2012, based on the disappearance of electron antineutrinos. More precise measurements have been obtained and presented on the mixing angle and the reactor neutrino spectrum, using ˜800 days of data. We have observed an excess of inverse-beta-decay (IBD) prompt spectrum near 5 MeV with respect to the most commonly used models. The excess is found to be consistent with coming from reactors. We have also successfully measured the reactor neutrino flux with a delayed signal of neutron capture on hydrogen. A future reactor neutrino experiment of RENO-50 is proposed to determine the neutrino mass hierarchy and to make highly precise measurements of θ12, Δm212, and Δ m312. Physics goals and sensitivities of RENO-50 are presented with a strategy of obtaining an extremely good energy resolution toward the neutrino mass hierarchy.

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

DOE PAGES

An, F. P.; Balantekin, A. B.; Band, H. R.; ...

2017-06-19

Here, the Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW th reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F 239 from 0.25 to 0.35, Daya Bay measures an average IBD yield ¯σf of (5.90±0.13)×10 –43 cm 2/fission and a fuel-dependent variation in the IBDmore » yield, dσ f/dF 239, of (–1.86±0.18)×10 –43 cm 2/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10 –43 cm 2/fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.« less

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay.

PubMed

An, F P; Balantekin, A B; Band, H R; Bishai, M; Blyth, S; Cao, D; Cao, G F; Cao, J; Chan, Y L; Chang, J F; Chang, Y; Chen, H S; Chen, Q Y; Chen, S M; Chen, Y X; Chen, Y; Cheng, J; Cheng, Z K; Cherwinka, J J; Chu, M C; Chukanov, A; Cummings, J P; Ding, Y Y; Diwan, M V; Dolgareva, M; Dove, J; Dwyer, D A; Edwards, W R; Gill, R; Gonchar, M; Gong, G H; Gong, H; Grassi, M; Gu, W Q; Guo, L; Guo, X H; Guo, Y H; Guo, Z; Hackenburg, R W; Hans, S; He, M; Heeger, K M; Heng, Y K; Higuera, A; Hsiung, Y B; Hu, B Z; Hu, T; Huang, E C; Huang, H X; Huang, X T; Huang, Y B; Huber, P; Huo, W; Hussain, G; Jaffe, D E; Jen, K L; Ji, X P; Ji, X L; Jiao, J B; Johnson, R A; Jones, D; Kang, L; Kettell, S H; Khan, A; Kohn, S; Kramer, M; Kwan, K K; Kwok, M W; Langford, T J; Lau, K; Lebanowski, L; Lee, J; Lee, J H C; Lei, R T; Leitner, R; Leung, J K C; Li, C; Li, D J; Li, F; Li, G S; Li, Q J; Li, S; Li, S C; Li, W D; Li, X N; Li, X Q; Li, Y F; Li, Z B; Liang, H; Lin, C J; Lin, G L; Lin, S; Lin, S K; Lin, Y-C; Ling, J J; Link, J M; Littenberg, L; Littlejohn, B R; Liu, J L; Liu, J C; Loh, C W; Lu, C; Lu, H Q; Lu, J S; Luk, K B; Ma, X Y; Ma, X B; Ma, Y Q; Malyshkin, Y; Martinez Caicedo, D A; McDonald, K T; McKeown, R D; Mitchell, I; Nakajima, Y; Napolitano, J; Naumov, D; Naumova, E; Ngai, H Y; Ochoa-Ricoux, J P; Olshevskiy, A; Pan, H-R; Park, J; Patton, S; Pec, V; Peng, J C; Pinsky, L; Pun, C S J; Qi, F Z; Qi, M; Qian, X; Qiu, R M; Raper, N; Ren, J; Rosero, R; Roskovec, B; Ruan, X C; Steiner, H; Stoler, P; Sun, J L; Tang, W; Taychenachev, D; Treskov, K; Tsang, K V; Tull, C E; Viaux, N; Viren, B; Vorobel, V; Wang, C H; Wang, M; Wang, N Y; Wang, R G; Wang, W; Wang, X; Wang, Y F; Wang, Z; Wang, Z; Wang, Z M; Wei, H Y; Wen, L J; Whisnant, K; White, C G; Whitehead, L; Wise, T; Wong, H L H; Wong, S C F; Worcester, E; Wu, C-H; Wu, Q; Wu, W J; Xia, D M; Xia, J K; Xing, Z Z; Xu, J L; Xu, Y; Xue, T; Yang, C G; Yang, H; Yang, L; Yang, M S; Yang, M T; Yang, Y Z; Ye, M; Ye, Z; Yeh, M; Young, B L; Yu, Z Y; Zeng, S; Zhan, L; Zhang, C; Zhang, C C; Zhang, H H; Zhang, J W; Zhang, Q M; Zhang, R; Zhang, X T; Zhang, Y M; Zhang, Y X; Zhang, Y M; Zhang, Z J; Zhang, Z Y; Zhang, Z P; Zhao, J; Zhou, L; Zhuang, H L; Zou, J H

2017-06-23

The Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW_{th} reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective ^{239}Pu fission fractions F_{239} from 0.25 to 0.35, Daya Bay measures an average IBD yield σ[over ¯]_{f} of (5.90±0.13)×10^{-43}  cm^{2}/fission and a fuel-dependent variation in the IBD yield, dσ_{f}/dF_{239}, of (-1.86±0.18)×10^{-43}  cm^{2}/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the ^{239}Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes ^{235}U, ^{239}Pu, ^{238}U, and ^{241}Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10^{-43}  cm^{2}/fission have been determined for the two dominant fission parent isotopes ^{235}U and ^{239}Pu. A 7.8% discrepancy between the observed and predicted ^{235}U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

NASA Astrophysics Data System (ADS)

An, F. P.; Balantekin, A. B.; Band, H. R.; Bishai, M.; Blyth, S.; Cao, D.; Cao, G. F.; Cao, J.; Chan, Y. L.; Chang, J. F.; Chang, Y.; Chen, H. S.; Chen, Q. Y.; Chen, S. M.; Chen, Y. X.; Chen, Y.; Cheng, J.; Cheng, Z. K.; Cherwinka, J. J.; Chu, M. C.; Chukanov, A.; Cummings, J. P.; Ding, Y. Y.; Diwan, M. V.; Dolgareva, M.; Dove, J.; Dwyer, D. A.; Edwards, W. R.; Gill, R.; Gonchar, M.; Gong, G. H.; Gong, H.; Grassi, M.; Gu, W. Q.; Guo, L.; Guo, X. H.; Guo, Y. H.; Guo, Z.; Hackenburg, R. W.; Hans, S.; He, M.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hsiung, Y. B.; Hu, B. Z.; Hu, T.; Huang, E. C.; Huang, H. X.; Huang, X. T.; Huang, Y. B.; Huber, P.; Huo, W.; Hussain, G.; Jaffe, D. E.; Jen, K. L.; Ji, X. P.; Ji, X. L.; Jiao, J. B.; Johnson, R. A.; Jones, D.; Kang, L.; Kettell, S. H.; Khan, A.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, J. H. C.; Lei, R. T.; Leitner, R.; Leung, J. K. C.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, S.; Li, S. C.; Li, W. D.; Li, X. N.; Li, X. Q.; Li, Y. F.; Li, Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, S.; Lin, S. K.; Lin, Y.-C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, J. L.; Liu, J. C.; Loh, C. W.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, K. B.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Malyshkin, Y.; Martinez Caicedo, D. A.; McDonald, K. T.; McKeown, R. D.; Mitchell, I.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Ngai, H. Y.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pan, H.-R.; Park, J.; Patton, S.; Pec, V.; Peng, J. C.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Qiu, R. M.; Raper, N.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Steiner, H.; Stoler, P.; Sun, J. L.; Tang, W.; Taychenachev, D.; Treskov, K.; Tsang, K. V.; Tull, C. E.; Viaux, N.; Viren, B.; Vorobel, V.; Wang, C. H.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Wei, H. Y.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wise, T.; Wong, H. L. H.; Wong, S. C. F.; Worcester, E.; Wu, C.-H.; Wu, Q.; Wu, W. J.; Xia, D. M.; Xia, J. K.; Xing, Z. Z.; Xu, J. L.; Xu, Y.; Xue, T.; Yang, C. G.; Yang, H.; Yang, L.; Yang, M. S.; Yang, M. T.; Yang, Y. Z.; Ye, M.; Ye, Z.; Yeh, M.; Young, B. L.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, C. C.; Zhang, H. H.; Zhang, J. W.; Zhang, Q. M.; Zhang, R.; Zhang, X. T.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhou, L.; Zhuang, H. L.; Zou, J. H.; Daya Bay Collaboration

2017-06-01

The Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 G Wth reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F239 from 0.25 to 0.35, Daya Bay measures an average IBD yield σ¯f of (5.90 ±0.13 )×10-43 cm2/fission and a fuel-dependent variation in the IBD yield, d σf/d F239, of (-1.86 ±0.18 )×10-43 cm2/fission . This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1 σ . This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17 ±0.17 ) and (4.27 ±0.26 )×10-43 cm2 /fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

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

An, F. P.; Balantekin, A. B.; Band, H. R.

Here, the Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW th reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F 239 from 0.25 to 0.35, Daya Bay measures an average IBD yield ¯σf of (5.90±0.13)×10 –43 cm 2/fission and a fuel-dependent variation in the IBDmore » yield, dσ f/dF 239, of (–1.86±0.18)×10 –43 cm 2/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10 –43 cm 2/fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.« less

Fusion materials semiannual progress report for the period ending December 31, 1996

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

NONE

1997-04-01

This is the twenty-first in a series of semiannual technical progress reports on fusion materials. This report combines the full spectrum of research and development activities on both metallic and non-metallic materials with primary emphasis on the effects of the neutronic and chemical environment on the properties and performance of materials for in-vessel components. This effort forms one element of the materials program being conducted in support of the Fusion Energy Sciences Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reportedmore » separately. The report covers the following topics: vanadium alloys; silicon carbide composite materials; ferritic/martensitic steels; copper alloys and high heat flux materials; austenitic stainless steels; insulating ceramics and optical materials; solid breeding materials; radiation effects, mechanistic studies and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; and irradiation facilities, test matrices, and experimental methods.« less

Fast-spectrum space-power-reactor concepts using boron control devices

NASA Technical Reports Server (NTRS)

Mayo, W.

1973-01-01

Several fast-spectrum space power reactor concepts that use boron carbide control devices were examined to determine the neutronic feasibility of the designs. The designs considered were (1) a 199-fuel-pin, 12-poison-reflector-control-drum reactor; (2) a 232-fuel-pin reactor with 12 reflector drums and three in-core control rods; (3) a 337-fuel-pin design with 12 incore control rods; and a 181-fuel-pin design with six drums closely coupled to the core to increase reactivity per drum. Adequate reactivity control and excess reactivity could be obtained for each concept, and the goals of 50,000 hours at 2.17 thermal megawatts with a lithium-7 coolant outlet temperature of 1222 K could be met without exceeding the 1-percent-clad-creep criterion. Heating rates in the boron carbide were calculated, but a heat transfer analysis was not done.

Unfolding neutron spectrum with Markov Chain Monte Carlo at MIT research Reactor with He-3 Neutral Current Detectors [Measuring neutron spectrum at MIT research reactor utilizing He-3 Bonner Cylinder Approach with an unfolding analysis

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

Leder, A.; Anderson, A. J.; Billard, J.

Here, the Ricochet experiment seeks to measure Coherent (neutral-current) Elastic Neutrino-Nucleus Scattering (CEνNS) using dark-matter-style detectors with sub-keV thresholds placed near a neutrino source, such as the MIT (research) Reactor (MITR), which operates at 5.5 MW generating approximately 2.2 × 10 18 ν/second in its core. Currently, Ricochet is characterizing the backgrounds at MITR, the main component of which comes in the form of neutrons emitted from the core simultaneous with the neutrino signal. To characterize this background, we wrapped Bonner cylinders around a 3 2He thermal neutron detector, whose data was then unfolded via a Markov Chain Monte Carlo (MCMC) to producemore » a neutron energy spectrum across several orders of magnitude. We discuss the resulting spectrum and its implications for deploying Ricochet at the MITR site as well as the feasibility of reducing this background level via the addition of polyethylene shielding around the detector setup.« less

Unfolding neutron spectrum with Markov Chain Monte Carlo at MIT research Reactor with He-3 Neutral Current Detectors [Measuring neutron spectrum at MIT research reactor utilizing He-3 Bonner Cylinder Approach with an unfolding analysis

DOE PAGES

Leder, A.; Anderson, A. J.; Billard, J.; ...

2018-02-02

Here, the Ricochet experiment seeks to measure Coherent (neutral-current) Elastic Neutrino-Nucleus Scattering (CEνNS) using dark-matter-style detectors with sub-keV thresholds placed near a neutrino source, such as the MIT (research) Reactor (MITR), which operates at 5.5 MW generating approximately 2.2 × 10 18 ν/second in its core. Currently, Ricochet is characterizing the backgrounds at MITR, the main component of which comes in the form of neutrons emitted from the core simultaneous with the neutrino signal. To characterize this background, we wrapped Bonner cylinders around a 3 2He thermal neutron detector, whose data was then unfolded via a Markov Chain Monte Carlo (MCMC) to producemore » a neutron energy spectrum across several orders of magnitude. We discuss the resulting spectrum and its implications for deploying Ricochet at the MITR site as well as the feasibility of reducing this background level via the addition of polyethylene shielding around the detector setup.« less

Deployment history and design considerations for space reactor power systems

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.

2009-05-01

The history of the deployment of nuclear reactors in Earth orbits is reviewed with emphases on lessons learned and the operation and safety experiences. The former Soviet Union's "BUK" power systems, with SiGe thermoelectric conversion and fast neutron energy spectrum reactors, powered a total of 31 Radar Ocean Reconnaissance Satellites (RORSATs) from 1970 to 1988 in 260 km orbit. Two of the former Soviet Union's TOPAZ reactors, with in-core thermionic conversion and epithermal neutron energy spectrum, powered two Cosmos missions launched in 1987 in ˜800 km orbit. The US' SNAP-10A system, with SiGe energy conversion and a thermal neutron energy spectrum reactor, was launched in 1965 in 1300 km orbit. The three reactor systems used liquid NaK-78 coolant, stainless steel structure and highly enriched uranium fuel (90-96 wt%) and operated at a reactor exit temperature of 833-973 K. The BUK reactors used U-Mo fuel rods, TOPAZ used UO 2 fuel rods and four ZrH moderator disks, and the SNAP-10A used moderated U-ZrH fuel rods. These low power space reactor systems were designed for short missions (˜0.5 kW e and ˜1 year for SNAP-10A, <3.0 kW e and <6 months for BUK, and ˜5.5 kW e and up to 1 year for TOPAZ). The deactivated BUK reactors at the end of mission, which varied in duration from a few hours to ˜4.5 months, were boosted into ˜800 km storage orbit with a decay life of more than 600 year. The ejection of the last 16 BUK reactor fuel cores caused significant contamination of Earth orbits with NaK droplets that varied in sizes from a few microns to 5 cm. Power systems to enhance or enable future interplanetary exploration, in-situ resources utilization on Mars and the Moon, and civilian missions in 1000-3000 km orbits would generate significantly more power of 10's to 100's kW e for 5-10 years, or even longer. A number of design options to enhance the operation reliability and safety of these high power space reactor power systems are presented and discussed.

Antineutrino monitoring of thorium reactors

DOE PAGES

Akindele, Oluwatomi A.; Bernstein, Adam; Norman, Eric B.

2016-09-30

Various groups have demonstrated that antineutrino monitoring can be successful in assessing the plutonium content in water-cooled nuclear reactors for nonproliferation applications. New reactor designs and concepts incorporate nontraditional fuel types and chemistry. Understanding how these properties affect the antineutrino emission from a reactor can extend the applicability of antineutrino monitoring. Thorium molten salt reactors breed 233U, that if diverted constitute a direct use material as defined by the International Atomic Energy Agency (IAEA). The antineutrino spectrum from the fission of 233U has been estimated for the first time, and the feasibility of detecting the diversion of 8 kg ofmore » 233U, within a 30 day timeliness goal has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos at a 25 m standoff. It was found that the diversion of a significant quantity of 233U could not be detected within the current IAEA timeliness detection goal using either tests. In conclusion, a rate-time based analysis exceeded the timeliness goal by 23 days, while a spectral based analysis exceeds this goal by 31 days.« less

Antineutrino monitoring of thorium reactors

NASA Astrophysics Data System (ADS)

Akindele, Oluwatomi A.; Bernstein, Adam; Norman, Eric B.

2016-09-01

Various groups have demonstrated that antineutrino monitoring can be successful in assessing the plutonium content in water-cooled nuclear reactors for nonproliferation applications. New reactor designs and concepts incorporate nontraditional fuel types and chemistry. Understanding how these properties affect the antineutrino emission from a reactor can extend the applicability of antineutrino monitoring. Thorium molten salt reactors breed 233U, that if diverted constitute a direct use material as defined by the International Atomic Energy Agency (IAEA). The antineutrino spectrum from the fission of 233U has been estimated for the first time, and the feasibility of detecting the diversion of 8 kg of 233U, within a 30 day timeliness goal has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos at a 25 m standoff. It was found that the diversion of a significant quantity of 233U could not be detected within the current IAEA timeliness detection goal using either tests. A rate-time based analysis exceeded the timeliness goal by 23 days, while a spectral based analysis exceeds this goal by 31 days.

The U.S. RERTR program status and progress.

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

Travelli, A.

1998-01-21

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program since its inception in 1978 is described. A brief summary of the results which the RERTR Program had achieved by the end of 1996 in collaboration with its many international partners is followed by a detailed review of the major events, findings, and activities of 1997. Significant progress has been made during the past year. In the area of U.S. acceptance of spent fuel from foreign research reactors, several shipments have taken place and additional are being planned. Intense fuel development activities are in progress, including procurement ofmore » equipment, screening of candidate materials, and production of microplates. Irradiation of the first series of microplates began in August 1997 in the Advanced Test Reactor, in Idaho. Progress has been made in the Russian RERTR program, which aims to develop and demonstrate within five years the technical means needed to convert Russian-supplied research reactors to LEU fuels. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, controversial performance issues which were raised at last year's meeting. Progress was also made on several aspects of producing molybdenum-99 from fission targets utilizing LEU instead of HEU. Various types of targets and processes are being pursued, with FDA approval of an LEU process projected to occur within two years. The feasibility of LEU Fuel conversion for three important DOE research reactors (BMRR, HFBR, and HFIR) has been evaluated by the RERTR program. In spite of the many momentous events which have occurred during the intervening years, and the excellent progress achieved, the most important challenges that the RERTR program faces today are not very different in type from those that were faced during the first RERTR meeting. Now, as then, the most important task is to develop new LEU fuels satisfying requirements which cannot be satisfied by any existing fuel. These new advanced fuels will enable conversion of the reactors which cannot be converted today, ensure better efficiency and performance for all research reactors, and allow the design of more powerful new advanced LEU reactors. As in the past, the success of the RERTR program will depend on free exchange of ideas and information, and on the international friendship and cooperation that have been a trademark of the RERTR program since its inception.« less

Neutron energy spectrum influence on irradiation hardening and microstructural development of tungsten

DOE PAGES

Fukuda, Makoto; Kiran Kumar, N. A. P.; Koyanagi, Takaaki; ...

2016-07-02

We performed a neutron irradiation to single crystal pure tungsten in the mixed spectrum High Flux Isotope Reactor (HFIR). In order to investigate the influences of neutron energy spectrum, the microstructure and irradiation hardening were compared with previous data obtained from the irradiation campaigns in the mixed spectrum Japan Material Testing Reactor (JMTR) and the sodium-cooled fast reactor Joyo. The irradiation temperatures were in the range of ~90–~800 °C and fast neutron fluences were 0.02–9.00 × 10 25 n/m 2 (E > 0.1 MeV). Post irradiation evaluation included Vickers hardness measurements and transmission electron microscopy. Moreover, the hardness and microstructuremore » changes exhibited a clear dependence on the neutron energy spectrum. The hardness appeared to increase with increasing thermal neutron flux when fast fluence exceeds 1 × 10 25 n/m 2 (E > 0.1 MeV). Finally, irradiation induced precipitates considered to be χ- and σ-phases were observed in samples irradiated to >1 × 10 25 n/m 2 (E > 0.1 MeV), which were pronounced at high dose and due to the very high thermal neutron flux of HFIR. Although the irradiation hardening mainly caused by defects clusters in a low dose regime, the transmutation-induced precipitation appeared to impose additional significant hardening of the tungsten.« less

Neutron energy spectrum influence on irradiation hardening and microstructural development of tungsten

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

Fukuda, Makoto; Kiran Kumar, N. A. P.; Koyanagi, Takaaki

We performed a neutron irradiation to single crystal pure tungsten in the mixed spectrum High Flux Isotope Reactor (HFIR). In order to investigate the influences of neutron energy spectrum, the microstructure and irradiation hardening were compared with previous data obtained from the irradiation campaigns in the mixed spectrum Japan Material Testing Reactor (JMTR) and the sodium-cooled fast reactor Joyo. The irradiation temperatures were in the range of ~90–~800 °C and fast neutron fluences were 0.02–9.00 × 10 25 n/m 2 (E > 0.1 MeV). Post irradiation evaluation included Vickers hardness measurements and transmission electron microscopy. Moreover, the hardness and microstructuremore » changes exhibited a clear dependence on the neutron energy spectrum. The hardness appeared to increase with increasing thermal neutron flux when fast fluence exceeds 1 × 10 25 n/m 2 (E > 0.1 MeV). Finally, irradiation induced precipitates considered to be χ- and σ-phases were observed in samples irradiated to >1 × 10 25 n/m 2 (E > 0.1 MeV), which were pronounced at high dose and due to the very high thermal neutron flux of HFIR. Although the irradiation hardening mainly caused by defects clusters in a low dose regime, the transmutation-induced precipitation appeared to impose additional significant hardening of the tungsten.« less

Beta-spectrum shapes of forbidden β decays

NASA Astrophysics Data System (ADS)

Kostensalo, Joel; Suhonen, Jouni

2018-03-01

The neutrinoless ββ decay of atomic nuclei continues to attract fervent interest due to its potential to confirm the possible Majorana nature of the neutrino, and thus the nonconservation of the lepton number. At the same time, the direct dark matter experiments are looking for weakly interacting massive particles (WIMPs) through their scattering on nuclei. The neutrino-oscillation experiments on reactor antineutrinos base their analyses on speculations of β-spectrum shapes of nuclear decays, thus leading to the notorious “reactor antineutrino anomaly.” In all these experimental efforts, one encounters the problem of β-spectrum shapes of forbidden β decays, either as unwanted backgrounds or unknown components in the analyses of data. In this work, the problem of spectrum shapes is discussed and illustrated with a set of selected examples. The relation of the β-spectrum shapes to the problem of the effective value of the weak axial-vector coupling strength gA and the enhancement of the axial-charge matrix element is also pointed out.

Burning high-level TRU waste in fusion fission reactors

NASA Astrophysics Data System (ADS)

Shen, Yaosong

2016-09-01

Recently, the concept of actinide burning instead of a once-through fuel cycle for disposing spent nuclear fuel seems to get much more attention. A new method of burning high-level transuranic (TRU) waste combined with Thorium-Uranium (Th-U) fuel in the subcritical reactors driven by external fusion neutron sources is proposed in this paper. The thorium-based TRU fuel burns all of the long-lived actinides via a hard neutron spectrum while outputting power. A one-dimensional model of the reactor concept was built by means of the ONESN_BURN code with new data libraries. The numerical results included actinide radioactivity, biological hazard potential, and much higher burnup rate of high-level transuranic waste. The comparison of the fusion-fission reactor with the thermal reactor shows that the harder neutron spectrum is more efficient than the soft. The Th-U cycle produces less TRU, less radiotoxicity and fewer long-lived actinides. The Th-U cycle provides breeding of 233U with a long operation time (>20 years), hence significantly reducing the reactivity swing while improving safety and burnup.

Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

NASA Astrophysics Data System (ADS)

Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

2017-07-01

Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (D

Reactor antineutrino shoulder explained by energy scale nonlinearities?

NASA Astrophysics Data System (ADS)

Mention, G.; Vivier, M.; Gaffiot, J.; Lasserre, T.; Letourneau, A.; Materna, T.

2017-10-01

The Daya Bay, Double Chooz and RENO experiments recently observed a significant distortion in their detected reactor antineutrino spectra, being at odds with the current predictions. Although such a result suggests to revisit the current reactor antineutrino spectra modeling, an alternative scenario, which could potentially explain this anomaly, is explored in this letter. Using an appropriate statistical method, a study of the Daya Bay experiment energy scale is performed. While still being in agreement with the γ calibration data and 12B measured spectrum, it is shown that a O (1%) deviation of the energy scale reproduces the distortion observed in the Daya Bay spectrum, remaining within the quoted calibration uncertainties. Potential origins of such a deviation, which challenge the energy calibration of these detectors, are finally discussed.

Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

NASA Technical Reports Server (NTRS)

Roth, R. J.

1976-01-01

The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

Accelerator-driven transmutation of spent fuel elements

DOEpatents

Venneri, Francesco; Williamson, Mark A.; Li, Ning

2002-01-01

An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

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.

Submersion criticality safety of tungsten-rhenium urania cermet fuel for space propulsion and power applications

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

A.E. Craft; R. C. O'Brien; S. D. Howe

Nuclear thermal rockets are the preferred propulsion technology for a manned mission to Mars, and tungsten–uranium oxide cermet fuels could provide significant performance and cost advantages for nuclear thermal rockets. A nuclear reactor intended for use in space must remain subcritical before and during launch, and must remain subcritical in launch abort scenarios where the reactor falls back to Earth and becomes submerged in terrestrial materials (including seawater, wet sand, or dry sand). Submersion increases reflection of neutrons and also thermalizes the neutron spectrum, which typically increases the reactivity of the core. This effect is typically very significant for compact,more » fast-spectrum reactors. This paper provides a submersion criticality safety analysis for a representative tungsten/uranium oxide fueled reactor with a range of fuel compositions. Each submersion case considers both the rhenium content in the matrix alloy and the uranium oxide volume fraction in the cermet. The inclusion of rhenium significantly improves the submersion criticality safety of the reactor. While increased uranium oxide content increases the reactivity of the core, it does not significantly affect the submersion behavior of the reactor. There is no significant difference in submersion behavior between reactors with rhenium distributed within the cermet matrix and reactors with a rhenium clad in the coolant channels. The combination of the flooding of the coolant channels in submersion scenarios and the presence of a significant amount of spectral shift absorbers (i.e. high rhenium concentration) further decreases reactivity for short reactor cores compared to longer cores.« less

The RERTR Program status and progress

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

Travelli, A.

1995-12-01

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1995 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1994. The revelation that Iraq was on the verge of developing a nuclear weapon at the time of the Gulf War, and that it was planning to do so by extracting HEU from the fuel of its research reactors, has given new impetus and urgency to the RERTR commitment of eliminating HEU use in research and test reactors worldwide.more » Development of advanced LEU research reactor fuels is scheduled to begin in October 1995. The Russian RERTR program, which aims to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels, is now in operation. A Statement of Intent was signed by high US and Chinese officials, endorsing cooperative activities between the RERTR program and Chinese laboratories involved in similar activities. Joint studies of LEU technical feasibility were completed for the SAFARI-I reactor in South Africa and for the ANS reactor in the US. A new study has been initiated for the FRM-II reactor in Germany. Significant progress was made on several aspects of producing {sup 99}Mo from fission targets utilizing LEU instead of HEU. A cooperation agreements is in place with the Indonesian BATAN. The first prototypical irradiation of an LEU metal-foil target for {sup 99}Mo production was accomplished in Indonesia. The TR-2 reactor, in Turkey, began conversion. SAPHIR, in Switzerland, was shut down. LEU fuel fabrication has begun for the conversion of two more US reactors. Twelve foreign reactors and nine domestic reactors have been fully converted. Approximately 60 % of the work required to eliminate the use of HEU in US-supplied research reactors has been accomplished.« less

Low-power lead-cooled fast reactor loaded with MOX-fuel

NASA Astrophysics Data System (ADS)

Sitdikov, E. R.; Terekhova, A. M.

2017-01-01

Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

Prospective scenarios of nuclear energy evolution over the 21. century

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

Massara, S.; Tetart, P.; Garzenne, C.

2006-07-01

In this paper, different world scenarios of nuclear energy development over the 21. century are analyzed, by means of the EDF fuel cycle simulation code for nuclear scenario studies, TIRELIRE - STRATEGIE. Three nuclear demand scenarios are considered, and the performance of different nuclear strategies in satisfying these scenarios is analyzed and discussed, focusing on natural uranium consumption and industrial requirements related to the nuclear reactors and the associated fuel cycle facilities. Both thermal-spectrum systems (Pressurized Water Reactor and High Temperature Gas-cooled Reactor) and Fast Reactors are investigated. (authors)

Progressing batch hydrolysis process

DOEpatents

Wright, J.D.

1985-01-10

A progressive batch hydrolysis process is disclosed for producing sugar from a lignocellulosic feedstock. It comprises passing a stream of dilute acid serially through a plurality of percolation hydrolysis reactors charged with feed stock, at a flow rate, temperature and pressure sufficient to substantially convert all the cellulose component of the feed stock to glucose. The cooled dilute acid stream containing glucose, after exiting the last percolation hydrolysis reactor, serially fed through a plurality of pre-hydrolysis percolation reactors, charged with said feedstock, at a flow rate, temperature and pressure sufficient to substantially convert all the hemicellulose component of said feedstock to glucose. The dilute acid stream containing glucose is cooled after it exits the last prehydrolysis reactor.

Progressing batch hydrolysis process

DOEpatents

Wright, John D.

1986-01-01

A progressive batch hydrolysis process for producing sugar from a lignocellulosic feedstock, comprising passing a stream of dilute acid serially through a plurality of percolation hydrolysis reactors charged with said feedstock, at a flow rate, temperature and pressure sufficient to substantially convert all the cellulose component of the feedstock to glucose; cooling said dilute acid stream containing glucose, after exiting the last percolation hydrolysis reactor, then feeding said dilute acid stream serially through a plurality of prehydrolysis percolation reactors, charged with said feedstock, at a flow rate, temperature and pressure sufficient to substantially convert all the hemicellulose component of said feedstock to glucose; and cooling the dilute acid stream containing glucose after it exits the last prehydrolysis reactor.

Progression along the Bipolar Spectrum: A Longitudinal Study of Predictors of Conversion from Bipolar Spectrum Conditions to Bipolar I and II Disorders

PubMed Central

Alloy, Lauren B.; Urošević, Snežana; Abramson, Lyn Y.; Jager-Hyman, Shari; Nusslock, Robin; Whitehouse, Wayne G.; Hogan, Michael

2011-01-01

Little longitudinal research has examined progression to more severe bipolar disorders in individuals with “soft” bipolar spectrum conditions. We examine rates and predictors of progression to bipolar I and II diagnoses in a non-patient sample of college-age participants (n = 201) with high General Behavior Inventory scores and childhood or adolescent onset of “soft” bipolar spectrum disorders followed longitudinally for 4.5 years from the Longitudinal Investigation of Bipolar Spectrum (LIBS) project. Of 57 individuals with initial cyclothymia or bipolar disorder not otherwise specified (BiNOS) diagnoses, 42.1% progressed to a bipolar II diagnosis and 10.5% progressed to a bipolar I diagnosis. Of 144 individuals with initial bipolar II diagnoses, 17.4% progressed to a bipolar I diagnosis. Consistent with hypotheses derived from the clinical literature and the Behavioral Approach System (BAS) model of bipolar disorder, and controlling for relevant variables (length of follow-up, initial depressive and hypomanic symptoms, treatment-seeking, and family history), high BAS sensitivity (especially BAS Fun Seeking) predicted a greater likelihood of progression to bipolar II disorder, whereas early age of onset and high impulsivity predicted a greater likelihood of progression to bipolar I (high BAS sensitivity and Fun-Seeking also predicted progression to bipolar I when family history was not controlled). The interaction of high BAS and high Behavioral Inhibition System (BIS) sensitivities also predicted greater likelihood of progression to bipolar I. We discuss implications of the findings for the bipolar spectrum concept, the BAS model of bipolar disorder, and early intervention efforts. PMID:21668080

Microdosimetric investigation of the spectra from YAYOI by use of the Monte Carlo code PHITS.

PubMed

Nakao, Minoru; Baba, Hiromi; Oishi, Ayumu; Onizuka, Yoshihiko

2010-07-01

The purpose of this study was to obtain the neutron energy spectrum on the surface of the moderator of the Tokyo University reactor YAYOI and to investigate the origins of peaks observed in the neutron energy spectrum by use of the Monte Carlo Code PHITS for evaluating biological studies. The moderator system was modeled with the use of details from an article that reported a calculation result and a measurement result for a neutron spectrum on the surface of the moderator of the reactor. Our calculation results with PHITS were compared to those obtained with the discrete ordinate code ANISN described in the article. In addition, the changes in the neutron spectrum at the boundaries of materials in the moderator system were examined with PHITS. Also, microdosimetric energy distributions of secondary charged particles from neutron recoil or reaction were calculated by use of PHITS and compared with a microdosimetric experiment. Our calculations of the neutron energy spectrum with PHITS showed good agreement with the results of ANISN in terms of the energy and structure of the peaks. However, the microdosimetric dose distribution spectrum with PHITS showed a remarkable discrepancy with the experimental one. The experimental spectrum could not be explained by PHITS when we used neutron beams of two mono-energies.

Optimization of 3D Field Design

NASA Astrophysics Data System (ADS)

Logan, Nikolas; Zhu, Caoxiang

2017-10-01

Recent progress in 3D tokamak modeling is now leveraged to create a conceptual design of new external 3D field coils for the DIII-D tokamak. Using the IPEC dominant mode as a target spectrum, the Finding Optimized Coils Using Space-curves (FOCUS) code optimizes the currents and 3D geometry of multiple coils to maximize the total set's resonant coupling. The optimized coils are individually distorted in space, creating toroidal ``arrays'' containing a variety of shapes that often wrap around a significant poloidal extent of the machine. The generalized perturbed equilibrium code (GPEC) is used to determine optimally efficient spectra for driving total, core, and edge neoclassical toroidal viscosity (NTV) torque and these too provide targets for the optimization of 3D coil designs. These conceptual designs represent a fundamentally new approach to 3D coil design for tokamaks targeting desired plasma physics phenomena. Optimized coil sets based on plasma response theory will be relevant to designs for future reactors or on any active machine. External coils, in particular, must be optimized for reliable and efficient fusion reactor designs. Work supported by the US Department of Energy under DE-AC02-09CH11466.

Improved Measurement of Reactor Flux and Spectrum at Daya Bay

NASA Astrophysics Data System (ADS)

Zhan, Liang; Daya Bay Collaboration

2017-09-01

A new measurement of the reactor antineutrino flux and energy spectrum by the Daya Bay experiment is reported. With a live time of 621 days, more than 1.2 million inverse beta decay (IBD) candidates were collected by eight antineutrino detectors (ADs) deployed in two near (560 m and 600 m flux-weighted baselines) and one far (16400 m flux-weighted baseline) underground experimental halls. The IBD yield was measured and the ratio to the predicted flux using the Huber+Mueller (ILL+Vogel) model was determined to be 0.946 ± 0.020 (0.992 ± 0.021). A 2.9 σ deviation was found in the measured IBD positron energy spectrum compared to the predictions. In particular, an excess of events in the region of 4-6 MeV was found with a local significance of 4.4 σ.

Improved Delayed-Neutron Spectroscopy Using Trapped Ions

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

Norman, Eric B.

The neutrons emitted following the β decay of fission fragments (known as delayed neutrons because they are emitted after fission on a timescale of the β-decay half-lives) play a crucial role in reactor performance and control. Reviews of delayed-neutron properties highlight the need for high-quality data for a wide variety of delayed-neutron emitters to better understand the time dependence and energy spectrum of the neutrons as these properties are essential for a detailed understanding of reactor kinetics needed for reactor safety and to understand the behavior of these reactors under various accident and component-failure scenarios. For fast breeder reactors, criticalitymore » calculations require accurate delayed-neutron energy spectra and approximations that are acceptable for light-water reactors such as assuming the delayed-neutron and fission-neutron energy spectra are identical are not acceptable and improved β-delayed neutron data is needed for safety and accident analyses for these reactors. With improved nuclear data, the delayed neutrons flux and energy spectrum could be calculated from the contributions from individual isotopes and therefore could be accurately modeled for any fuel-cycle concept, actinide mix, or irradiation history. High-quality β-delayed neutron measurements are also critical to constrain modern nuclear-structure calculations and empirical models that predict the decay properties for nuclei for which no data exists and improve the accuracy and flexibility of the existing empirical descriptions of delayed neutrons from fission such as the six-group representation« less

Assessment of quasi-linear effect of RF power spectrum for enabling lower hybrid current drive in reactor plasmas

NASA Astrophysics Data System (ADS)

Cesario, Roberto; Cardinali, Alessandro; Castaldo, Carmine; Amicucci, Luca; Ceccuzzi, Silvio; Galli, Alessandro; Napoli, Francesco; Panaccione, Luigi; Santini, Franco; Schettini, Giuseppe; Tuccillo, Angelo Antonio

2017-10-01

The main research on the energy from thermonuclear fusion uses deuterium plasmas magnetically trapped in toroidal devices. To suppress the turbulent eddies that impair thermal insulation and pressure tight of the plasma, current drive (CD) is necessary, but tools envisaged so far are unable accomplishing this task while efficiently and flexibly matching the natural current profiles self-generated at large radii of the plasma column [1-5]. The lower hybrid current drive (LHCD) [6] can satisfy this important need of a reactor [1], but the LHCD system has been unexpectedly mothballed on JET. The problematic extrapolation of the LHCD tool at reactor graded high values of, respectively, density and temperatures of plasma has been now solved. The high density problem is solved by the FTU (Frascati Tokamak Upgrade) method [7], and solution of the high temperature one is presented here. Model results based on quasi-linear (QL) theory evidence the capability, w.r.t linear theory, of suitable operating parameters of reducing the wave damping in hot reactor plasmas. Namely, using higher RF power densities [8], or a narrower antenna power spectrum in refractive index [9,10], the obstacle for LHCD represented by too high temperature of reactor plasmas should be overcome. The former method cannot be used for routinely, safe antenna operations, Thus, only the latter key is really exploitable in a reactor. The proposed solutions are ultimately necessary for viability of an economic reactor.

71. ARAII. Construction progress at SL1 site near end of ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

71. ARA-II. Construction progress at SL-1 site near end of 1957. Buildings from right to left are guard house, support building, reactor building, water tank and pump house. Construction was 23 percent complete. December 20, 1957. Ineel photo no. 57-6224. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

The Fast-spectrum Transmutation Experimental Facility FASTEF: Main design achievements (part 2: Reactor building design and plant layout) within the FP7-CDT collaborative project of the European Commission

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

De Bruyn, D.; Engelen, J.; Ortega, A.

MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK-CEN in replacement of its material testing reactor BR2. SCK-CEN in association with 17 European partners from industry, research centres and academia, responded to the FP7 (Seventh Framework Programme) call from the European Commission to establish a Central Design Team (CDT) for the design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) able to demonstrate efficient transmutation and associated technology through a system working in subcritical and/or critical mode. The project has started on April 01, 2009 for a period of threemore » years. In this paper, we present the latest concept of the reactor building and the plant layout. The FASTEF facility has evolved quite a lot since the intermediate reporting done at the ICAPP'10 and ICAPP'11 conferences 1,2. Many iterations have been performed to take into account the safety requirements. The present configuration enables an easy operation and maintenance of the facility, including the possibility to change large components of the reactor. In a companion paper 3, we present the latest configuration of the reactor core and primary system. (authors)« less

A novel two-level dielectric barrier discharge reactor for methyl orange degradation.

PubMed

Tao, Xumei; Wang, Guowei; Huang, Liang; Ye, Qingguo; Xu, Dongyan

2016-12-15

A novel pilot two-level dielectric barrier discharge (DBD) reactor has been proposed and applied for degradation of continuous model wastewater. The two-level DBD reactor was skillfully realized with high space utilization efficiency and large contact area between plasma and wastewater. Various conditions such as applied voltage, initial concentration and initial pH value on methyl orange (MO) model wastewater degradation were investigated. The results showed that the appropriate applied voltage was 13.4 kV; low initial concentration and low initial pH value were conducive for MO degradation. The percentage removal of 4 L MO with concentration of 80 mg/L reached 94.1% after plasma treatment for 80min. Based on ultraviolet spectrum (UV), Infrared spectrum (IR), liquid chromatography-mass spectrometry (LC-MS) analysis of degradation intermediates and products, insights in the degradation pathway of MO were proposed. Copyright © 2016 Elsevier Ltd. All rights reserved.

Lawrence Livermore National Laboratory and Sandia National Laboratory Nuclear Accident Dosimetry Support of IER 252 and the Dose Characterization of the Flattop Reactor at the DAF

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

Hickman, D. P.; Jeffers, K. L.; Radev, R. P.

In support of IER 252 “Characterization of the Flattop Reactor at the NCERC”, LLNL performed ROSPEC measurements of the neutron spectrum and deployed 129 Personnel Nuclear Accident Dosimeters (PNAD) to establish the need for height corrections and verification of neutron spectrum evaluation of the fluences and dose. A very limited number of heights (typically only one or two heights) can be measured using neutron spectrometers, therefore it was important to determine if any height correction would be needed in future intercomparisons and studies. Specific measurement positions around the Flatttop reactor are provided in Figure 1. Table 1 provides run andmore » position information for LLNL measurements. The LLNL ROSPEC (R2) was used for run numbers 1 – 7, and vi. PNADs were positioned on trees during run numbers 9, 11, and 13.« less

An intrinsically safe facility for forefront research and training on nuclear technologies

NASA Astrophysics Data System (ADS)

Mansani, L.; Monti, S.; Ricco, G.; Ricotti, M.

2014-04-01

In this short paper the motivations for the development of fast spectrum lead-cooled reactors are briefly summarized. In particular the importance of subcritical research reactors, like the one described in this Focus Point, for the investigation of various scientifical and technological aspects and the training of students, is discussed.

Harmonic Composition of the Currents of Power Windings in 500 KV Thyristor Controlled Shunt Reactor with Split Valveside Windings

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

Matinyan, A. M., E-mail: al-drm@mail.ru; Peshkov, M. V.; Karpov, V. N.

2016-09-15

The design and current spectrum of a thyristor valve controlled shunt reactor (TCSR) with split valveside windings are described. The dependence of the amplitudes of higher-order harmonics of the power winding current on the TCSR operating regime are presented for this TCSR design.

Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Forsbacka, Matthew

2004-01-01

For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the NASA Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. "virtual" reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of van'ous core deformations. The power delivered to the SAFE-100 prototype was then dusted accordingly via kinetics calculations, The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kilowatt(sub t), to 10 kilowatts(sub t), held approximately constant at 10 kilowatts (sub t), and then allowed to decrease based on the negative thermal reactivity coefficient.

Measurement and calculation of fast neutron and gamma spectra in well defined cores in LR-0 reactor.

PubMed

Košťál, Michal; Matěj, Zdeněk; Cvachovec, František; Rypar, Vojtěch; Losa, Evžen; Rejchrt, Jiří; Mravec, Filip; Veškrna, Martin

2017-02-01

A well-defined neutron spectrum is essential for many types of experimental topics and is also important for both calibration and testing of spectrometric and dosimetric detectors. Provided it is well described, such a spectrum can also be employed as a reference neutron field that is suitable for validating selected cross sections. The present paper aims to compare calculations and measurements of such a well-defined spectra in geometrically similar cores of the LR-0 reactor with fuel containing slightly different enrichments (2%, 3.3% and 3.6%). The common feature to all cores is a centrally located dry channel which can be used for the insertion of studied materials. The calculation of neutron and gamma spectra was realized with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, ROSFOND-2010 and CENDL-3.1 nuclear data libraries. Only minor differences in neutron and gamma spectra were found in the comparison of the presented reactor cores with different fuel enrichments. One exception is the gamma spectrum in the higher energy region (above 8MeV), where more pronounced variations could be observed. Copyright © 2016 Elsevier Ltd. All rights reserved.

Shielding Analysis of a Small Compact Space Nuclear Reactor

DTIC Science & Technology

1987-08-01

RESPONSE) =4, MAXWELLIAN FISSION SPECTRUM (ILNTEGRAL RESPONSE) =5, LOS ALAMOS FISSION SPECTRUM, 1982 (INTEGRAL RESPONSE) =6, VITAMIN C NEUTRON SPECTRUM...Appendices Appendix A: Calculations of Effective Radii.. A-1 Appendix B: Atom Density Calculations for FEMPlD and FEMP2D ................ B-I Appendix C ...FEMPID and FEM22D Data........... C -i Appendix D: Energy Group Definition .......... D-I Appendix E: Transport Equation, Legendr4 Polynomial

Design Study of a Modular Gas-Cooled, Closed-Brayton Cycle Reactor for Marine Use

DTIC Science & Technology

1989-06-01

materials in the core and surroundings. To investigate this design point in the marine variant I developed the program HEAT.BAS to perform a one-dimensional...helium as the working fluid. The core is a graphite moderated, epithermal spectrum reactor, using TRISO fuel particles in extruded graphite fuel elements...The fuel is highly enriched U2315 . The containment is shaped in an inverted ’T’ with two sections. The upper section contains the reactor core

Spectrum and density of neutron flux in the irradiation beam line no. 3 of the IBR-2 reactor

NASA Astrophysics Data System (ADS)

Shabalin, E. P.; Verkhoglyadov, A. E.; Bulavin, M. V.; Rogov, A. D.; Kulagin, E. N.; Kulikov, S. A.

2015-03-01

Methodology and results of measuring the differential density of the neutron flux in irradiation beam line no. 3 of the IBR-2 reactor using neutron activation analysis (NAA) are presented in the paper. The results are compared to the calculation performed on the basis of the 3D MCNP model. The data that are obtained are required to determine the integrated radiation dose of the studied samples at various distances from the reactor.

Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

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

Worrall, Andrew; Todosow, Michael

2016-01-01

Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include:more » increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle performance metrics for a small modular reactor are compared to a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. Metrics performance for a small modular reactor are degraded for mass of spent nuclear fuel and high level waste disposed, mass of depleted uranium disposed, land use per energy generated, and carbon emission per energy generated« less

A ``NEW'' Solid-Core Reactor Fuel Form that Maximizes the Performance of Nuclear Thermal and Electric Rockets

NASA Astrophysics Data System (ADS)

Rom, Frank E.; Finnegan, Patrick M.

1994-07-01

The ``NEW'' solid-core fuel form is the old Vapor Transport (VT) fuel pin investigated at NASA about 30 years ago. It is simply a tube sealed at both ends partially filled with UO2. During operation the UO2 forms an annular layer on the inside of the tube by vaporization and condensation. This form is an ideal structure for overall strength and retention of fission products. All of the structural material lies between the fuel (including fission products) and the reactor coolant. The isothermal inside fuel surface temperature that results from the vaporization and condensation of fuel during operation eliminates hotspots, significantly increasing the design fuel pin surface temperature. For NTP, W-UO2 fuel pins yield higher operating temperatures than for other fuel forms, because W has about a ten-fold lower vaporization rate compared to any other known material. The use of perigee propulsion using W-UO2 fuel pins can result in a more than ten-fold reduction in reactor power. Lower reactor power, together with zero fission product release potential, and the simplicity of fabrication of VT fuel pins should greatly simplify and reduce the cost of development of NTP. For NEP, VT fuel pins can increase fast neutron spectrum reactor life with no fission product release. Thermal spectrum NEP reactors using W184 or Mo VT fuel pins, with only small amounts of high neutron absorbing additives, offer benefits because of much lower fissionable fuel requirements. The VT fuel pin has application to commercial power reactors with similar benefits.

Total absorption spectroscopy of fission fragments relevant for reactor antineutrino spectra

DOE PAGES

Fallot, M.; Porta, A.; Meur, L. Le; ...

2017-09-13

Here, the accurate determination of reactor antineutrino spectra remains a very active research topic for which new methods of study have emerged in recent years. Indeed, following the long-recognized reactor anomaly (measured antineutrino deficit in short baseline reactor experiments when compared with spectral predictions), the three international reactor neutrino experiments Double Chooz, Daya Bay and Reno have recently demonstrated the existence of spectral distortions in their measurements with respect to the same predictions. These spectral predictions were obtained through the conversion of integral beta-energy spectra obtained at the ILL research reactor. Several studies have shown that the underlying nuclear physicsmore » required for the conversion of these spectra into antineutrino spectra is not totally understood. An alternative to such converted spectra is a complementary approach that consists of determining the antineutrino spectrum by means of the measurement and processing of nuclear data. The beta properties of some key fission products suffer from the pandemonium effect which can be circumvented by the use of the Total Absorption Gamma-ray Spectroscopy technique (TAGS). The two main contributors to the Pressurized Water Reactor antineutrino spectrum in the region where the spectral distortion has been observed are 92Rb and 142Cs, which have been measured at the radioactive beam facility of the University of Jyvaskyla in two TAGS experiments. We present the results of the analysis of the TAGS measurements of the β-decay properties of 92Rb along with preliminary results on 142Cs and report on the measurements already performed.« less

Total absorption spectroscopy of fission fragments relevant for reactor antineutrino spectra

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

Fallot, M.; Porta, A.; Meur, L. Le

Here, the accurate determination of reactor antineutrino spectra remains a very active research topic for which new methods of study have emerged in recent years. Indeed, following the long-recognized reactor anomaly (measured antineutrino deficit in short baseline reactor experiments when compared with spectral predictions), the three international reactor neutrino experiments Double Chooz, Daya Bay and Reno have recently demonstrated the existence of spectral distortions in their measurements with respect to the same predictions. These spectral predictions were obtained through the conversion of integral beta-energy spectra obtained at the ILL research reactor. Several studies have shown that the underlying nuclear physicsmore » required for the conversion of these spectra into antineutrino spectra is not totally understood. An alternative to such converted spectra is a complementary approach that consists of determining the antineutrino spectrum by means of the measurement and processing of nuclear data. The beta properties of some key fission products suffer from the pandemonium effect which can be circumvented by the use of the Total Absorption Gamma-ray Spectroscopy technique (TAGS). The two main contributors to the Pressurized Water Reactor antineutrino spectrum in the region where the spectral distortion has been observed are 92Rb and 142Cs, which have been measured at the radioactive beam facility of the University of Jyvaskyla in two TAGS experiments. We present the results of the analysis of the TAGS measurements of the β-decay properties of 92Rb along with preliminary results on 142Cs and report on the measurements already performed.« less

ANALYTICAL CHEMISTRY DIVISION ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31, 1961

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

None

1962-02-01

Research and development progress is reported on analytlcal instrumentation, dlssolver-solution analyses, special research problems, reactor projects analyses, x-ray and spectrochemical analyses, mass spectrometry, optical and electron microscopy, radiochemical analyses, nuclear analyses, inorganic preparations, organic preparations, ionic analyses, infrared spectral studies, anodization of sector coils for the Analog II Cyclotron, quality control, process analyses, and the Thermal Breeder Reactor Projects Analytical Chemistry Laboratory. (M.C.G.)

Measurements of SNAC2 area dosimeters placed in different configurations around the PROSPERO reactor and comparison with TRIPOLI-4 calculations

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

Rousseau, G.; Chambru, L.; Authier, N.

2015-07-01

In the context of criticality accident alarm system tests, several experiments were carried out in 2013 on the PROSPERO reactor to study the response to neutron and gamma of different devices and dosimeters, particularly on the SNAC2 dosimeter. This article presents the results of this criticality dosimeter in different configurations, and compares the experimental measurements with the results of calculation performed with the TRIPOLI-4 Monte-Carlo Neutral Particles transport code. PROSPERO is a metallic critical assembly managed by the Criticality, Neutron Science and Measurement Department located at the French CEA Research Center of Valduc. The core, surrounded by a reflector ofmore » depleted uranium, is composed of 2 horizontal cylindrical blocks made of a highly enriched uranium alloy which can be placed in contact, and of 4 depleted uranium control rods which allow the reactor to be driven. This reactor, placed in a cell 10 m x 8 m x 6 m high, with 1.4-meter-thick concrete walls, is used as a fast neutron spectrum source and is operated at stable power level in delayed critical state, which can vary from 3 mW to 3 kW. PROSPERO is extensively used for electronic hardening or to study the effect of the neutrons on various materials. The SNAC2 criticality dosimeter is a zone dosimeter allowing the off line measurement of criticality accident neutron doses. This dosimeter consists of the pile up of seven activation foils embedded into a 23 mm diameter x 21 mm height cadmium container. The activation measurement of each foil, using a gamma spectroscopy technique, gives information about the neutron reaction rates. The SNAC2 software allows the spectrum unfolding from these values, taking into account the hypothesis of a particular spectrum shape, in three components: a Maxwell spectrum component for the thermal range, a 1/E component for the epithermal range, and a Watt spectrum component for the high energy range. Moreover, from the neutron spectrum, the SNAC software can calculate the neutron fluence integrated by the dosimeter and the neutron dose. During the 3 weeks measurement campaign many radioprotection devices were used. To modify the spectrum seen by these devices, several shields of various thicknesses made of concrete or polyethylene, with or without cadmium covers, were placed in the PROSPERO cell. These devices allow the study of criticality accident spectra in several environments: from metal to pseudo liquid. The fluxes measured by the SNAC2 devices were compared with TRIPOLI-4 calculations. (authors)« less

Analysis of a boron-carbide-drum-controlled critical reactor experiment

NASA Technical Reports Server (NTRS)

Mayo, W. T.

1972-01-01

In order to validate methods and cross sections used in the neutronic design of compact fast-spectrum reactors for generating electric power in space, an analysis of a boron-carbide-drum-controlled critical reactor was made. For this reactor the transport analysis gave generally satisfactory results. The calculated multiplication factor for the most detailed calculation was only 0.7-percent Delta k too high. Calculated reactivity worth of the control drums was $11.61 compared to measurements of $11.58 by the inverse kinetics methods and $11.98 by the inverse counting method. Calculated radial and axial power distributions were in good agreement with experiment.

Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

NASA Astrophysics Data System (ADS)

Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

2006-01-01

A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

The reactor antineutrino anomaly and low energy threshold neutrino experiments

NASA Astrophysics Data System (ADS)

Cañas, B. C.; Garcés, E. A.; Miranda, O. G.; Parada, A.

2018-01-01

Short distance reactor antineutrino experiments measure an antineutrino spectrum a few percent lower than expected from theoretical predictions. In this work we study the potential of low energy threshold reactor experiments in the context of a light sterile neutrino signal. We discuss the perspectives of the recently detected coherent elastic neutrino-nucleus scattering in future reactor antineutrino experiments. We find that the expectations to improve the current constraints on the mixing with sterile neutrinos are promising. We also analyze the measurements of antineutrino scattering off electrons from short distance reactor experiments. In this case, the statistics is not competitive with inverse beta decay experiments, although future experiments might play a role when compare it with the Gallium anomaly.

Characterization of gamma field in the JSI TRIGA reactor

NASA Astrophysics Data System (ADS)

Ambrožič, Klemen; Radulović, Vladimir; Snoj, Luka; Gruel, Adrien; Guillou, Mael Le; Blaise, Patrick; Destouches, Christophe; Barbot, Loïc

2018-01-01

Research reactors such as the "Jožzef Stefan" Institute TRIGA reactor have primarily been designed for experimentation and sample irradiation with neutrons. However recent developments in incorporating additional instrumentation for nuclear power plant support and with novel high flux material testing reactor designs, γ field characterization has become of great interest for the characterization of the changes in operational parameters of electronic devices and for the evaluation of γ heating of MTR's structural materials in a representative reactor Γ spectrum. In this paper, we present ongoing work on γ field characterization both experimentally, by performing γ field measurements, and by simulations, using Monte Carlo particle transport codes in conjunction with R2S methodology for delayed γ field characterization.

[Progressive damage monitoring of corrugated composite skins by the FBG spectral characteristics].

PubMed

Zhang, Yong; Wang, Bang-Feng; Lu, Ji-Yun; Gu, Li-Li; Su, Yong-Gang

2014-03-01

In the present paper, a method of monitoring progressive damage of composite structures by non-uniform fiber Bragg grating (FBG) reflection spectrum is proposed. Due to the finite element analysis of corrugated composite skins specimens, the failure process under tensile load and corresponding critical failure loads of corrugated composite skin was predicated. Then, the non-uniform reflection spectrum of FBG sensor could be reconstructed and the corresponding relationship between layer failure order sequence of corrugated composite skin and FBG sensor reflection spectrums was acquired. A monitoring system based on FBG non-uniform reflection spectrum, which can be used to monitor progressive damage of corrugated composite skins, was built. The corrugated composite skins were stretched under this FBG non-uniform reflection spectrum monitoring system. The results indicate that real-time spectrums acquired by FBG non-uniform reflection spectrum monitoring system show the same trend with the reconstruction reflection spectrums. The maximum error between the corresponding failure and the predictive value is 8.6%, which proves the feasibility of using FBG sensor to monitor progressive damage of corrugated composite skin. In this method, the real-time changes in the FBG non-uniform reflection spectrum within the scope of failure were acquired through the way of monitoring and predicating, and at the same time, the progressive damage extent and layer failure sequence of corru- gated composite skin was estimated, and without destroying the structure of the specimen, the method is easy and simple to operate. The measurement and transmission section of the system are completely composed of optical fiber, which provides new ideas and experimental reference for the field of dynamic monitoring of smart skin.

Precision measurement of neutrino oscillation parameters with KamLAND.

PubMed

Abe, S; Ebihara, T; Enomoto, S; Furuno, K; Gando, Y; Ichimura, K; Ikeda, H; Inoue, K; Kibe, Y; Kishimoto, Y; Koga, M; Kozlov, A; Minekawa, Y; Mitsui, T; Nakajima, K; Nakajima, K; Nakamura, K; Nakamura, M; Owada, K; Shimizu, I; Shimizu, Y; Shirai, J; Suekane, F; Suzuki, A; Takemoto, Y; Tamae, K; Terashima, A; Watanabe, H; Yonezawa, E; Yoshida, S; Busenitz, J; Classen, T; Grant, C; Keefer, G; Leonard, D S; McKee, D; Piepke, A; Decowski, M P; Detwiler, J A; Freedman, S J; Fujikawa, B K; Gray, F; Guardincerri, E; Hsu, L; Kadel, R; Lendvai, C; Luk, K-B; Murayama, H; O'Donnell, T; Steiner, H M; Winslow, L A; Dwyer, D A; Jillings, C; Mauger, C; McKeown, R D; Vogel, P; Zhang, C; Berger, B E; Lane, C E; Maricic, J; Miletic, T; Batygov, M; Learned, J G; Matsuno, S; Pakvasa, S; Foster, J; Horton-Smith, G A; Tang, A; Dazeley, S; Downum, K E; Gratta, G; Tolich, K; Bugg, W; Efremenko, Y; Kamyshkov, Y; Perevozchikov, O; Karwowski, H J; Markoff, D M; Tornow, W; Heeger, K M; Piquemal, F; Ricol, J-S

2008-06-06

The KamLAND experiment has determined a precise value for the neutrino oscillation parameter Deltam21(2) and stringent constraints on theta12. The exposure to nuclear reactor antineutrinos is increased almost fourfold over previous results to 2.44 x 10(32) proton yr due to longer livetime and an enlarged fiducial volume. An undistorted reactor nu[over]e energy spectrum is now rejected at >5sigma. Analysis of the reactor spectrum above the inverse beta decay energy threshold, and including geoneutrinos, gives a best fit at Deltam21(2)=7.58(-0.13)(+0.14)(stat) -0.15+0.15(syst) x 10(-5) eV2 and tan2theta12=0.56(-0.07)+0.10(stat) -0.06+0.10(syst). Local Deltachi2 minima at higher and lower Deltam21(2) are disfavored at >4sigma. Combining with solar neutrino data, we obtain Deltam21(2)=7.59(-0.21)+0.21 x 10(-5) eV2 and tan2theta12=0.47(-0.05)+0.06.

Energy spectrum of 208Pb(n,x) reactions

NASA Astrophysics Data System (ADS)

Tel, E.; Kavun, Y.; Özdoǧan, H.; Kaplan, A.

2018-02-01

Fission and fusion reactor technologies have been investigated since 1950's on the world. For reactor technology, fission and fusion reaction investigations are play important role for improve new generation technologies. Especially, neutron reaction studies have an important place in the development of nuclear materials. So neutron effects on materials should study as theoretically and experimentally for improve reactor design. For this reason, Nuclear reaction codes are very useful tools when experimental data are unavailable. For such circumstances scientists created many nuclear reaction codes such as ALICE/ASH, CEM95, PCROSS, TALYS, GEANT, FLUKA. In this study we used ALICE/ASH, PCROSS and CEM95 codes for energy spectrum calculation of outgoing particles from Pb bombardment by neutron. While Weisskopf-Ewing model has been used for the equilibrium process in the calculations, full exciton, hybrid and geometry dependent hybrid nuclear reaction models have been used for the pre-equilibrium process. The calculated results have been discussed and compared with the experimental data taken from EXFOR.

An assessment of the attractiveness of material associated with thorium/uranium and uranium closed fuel cycles from a safeguards perspective

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

Bathke, Charles Gary; Wallace, Richard K; Hase, Kevin R

2010-01-01

This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with various proposed nuclear fuel cycles. Specifically, this paper examines two closed fuel cycles. The first fuel cycle examined is a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of plutonium/thorium and {sup 233}U/thorium. The used fuel is then reprocessed using the THOREX process and the actinides are recycled. The second fuel cycle examined consists of conventional light water reactors (LWR) whose fuel is reprocessed for actinides that are then fed to and recycled untilmore » consumed in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). As reprocessing of LWR fuel has already been examined, this paper will focus on the reprocessing of the scheme's fast-spectrum reactors' fuel. This study will indicate what is required to render these materials as having low utility for use in nuclear weapons. Nevertheless, the results of this paper suggest that all reprocessing products evaluated so far need to be rigorously safeguarded and provided high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE). The methodology and key findings will be presented.« less

OBJECT KINETIC MONTE CARLO SIMULATIONS OF RADIATION DAMAGE IN TUNGSTEN SUBJECTED TO NEUTRON FLUX WITH PKA SPECTRUM CORRESPONDING TO THE HFIR

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

Nandipati, Giridhar; Setyawan, Wahyu; Heinisch, Howard L.

2015-12-31

The objective of this work is to study the damage accumulation in pure tungsten (W) subjected to neutron bombardment with a primary knock-on atom (PKA) spectrum corresponding to the High Flux Isotope Reactor (HFIR), using the object kinetic Monte Carlo (OKMC) method.

Data acquisition system for segmented reactor antineutrino detector

NASA Astrophysics Data System (ADS)

Hons, Z.; Vlášek, J.

2017-01-01

This paper describes the data acquisition system used for data readout from the PMT channels of a segmented detector of reactor antineutrinos with active shielding. Theoretical approach to the data acquisition is described and two possible solutions using QDCs and digitizers are discussed. Also described are the results of the DAQ performance during routine data taking operation of DANSS. DANSS (Detector of the reactor AntiNeutrino based on Solid Scintillator) is a project aiming to measure a spectrum of reactor antineutrinos using inverse beta decay (IBD) in a plastic scintillator. The detector is located close to an industrial nuclear reactor core and is covered by passive and active shielding. It is expected to have about 15000 IBD interactions per day. Light from the detector is sensed by PMT and SiPM.

Dual annular rotating "windowed" nuclear reflector reactor control system

DOEpatents

Jacox, Michael G.; Drexler, Robert L.; Hunt, Robert N. M.; Lake, James A.

1994-01-01

A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core.

Neutronic experiments with fluorine rich compounds at LR-0 reactor

DOE PAGES

Losa, Evzen; Kostal, Michal; Czakoj, T.; ...

2018-06-06

Here, research on molten salt reactor (MSR) neutronics continues in Research Centre Rez (Czech Republic) with experimental work being conducted using fluoride salt that was originally used in the Molten Salt Reactor Experiment (MSRE). Previous results identified significant variations in the neutron spectrum measured in LiF-NaF salt. These variations could originate from the fluorine description in current nuclear data sets. Subsequent experiments were performed to try to confirm this phenomenon. Therefore, another fluorine-rich compound, Teflon, was used for testing. Critical experiments showed slight discrepancies in C/E-1 for both compounds, Teflon and FLIBE, and systematic overestimation of criticality was observed inmore » calculations. Different nuclear data libraries were used for data set testing. For Teflon, the overestimation is higher when using JENDL-3.3, JENDL-4, and RUSFOND-2010 libraries, all three of which share the same inelastic-to-elastic scattering cross section ratio. Calculations using other libraries (ENDF/B-VII.1, ENDF/B-VII.0, JEFF-3.2, JEFF-3.1, and CENDL-3.1) tend to be closer to the experimental value. Neutron spectrum measurement in both substances revealed structure similar to that seen in previous measurements using LiF-NaF salt, which indicates that the neutron spectrum seems to be strongly shaped by fluorine. Discrepancies between experimental and calculational results seem to be larger in the neutron energy range of 100–1300 keV than in higher energies. In the case of neutron spectrum calculation, none of the tested libraries gives overall better results than the others.« less

Neutronic experiments with fluorine rich compounds at LR-0 reactor

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

Losa, Evzen; Kostal, Michal; Czakoj, T.

Here, research on molten salt reactor (MSR) neutronics continues in Research Centre Rez (Czech Republic) with experimental work being conducted using fluoride salt that was originally used in the Molten Salt Reactor Experiment (MSRE). Previous results identified significant variations in the neutron spectrum measured in LiF-NaF salt. These variations could originate from the fluorine description in current nuclear data sets. Subsequent experiments were performed to try to confirm this phenomenon. Therefore, another fluorine-rich compound, Teflon, was used for testing. Critical experiments showed slight discrepancies in C/E-1 for both compounds, Teflon and FLIBE, and systematic overestimation of criticality was observed inmore » calculations. Different nuclear data libraries were used for data set testing. For Teflon, the overestimation is higher when using JENDL-3.3, JENDL-4, and RUSFOND-2010 libraries, all three of which share the same inelastic-to-elastic scattering cross section ratio. Calculations using other libraries (ENDF/B-VII.1, ENDF/B-VII.0, JEFF-3.2, JEFF-3.1, and CENDL-3.1) tend to be closer to the experimental value. Neutron spectrum measurement in both substances revealed structure similar to that seen in previous measurements using LiF-NaF salt, which indicates that the neutron spectrum seems to be strongly shaped by fluorine. Discrepancies between experimental and calculational results seem to be larger in the neutron energy range of 100–1300 keV than in higher energies. In the case of neutron spectrum calculation, none of the tested libraries gives overall better results than the others.« less

Determination of neutron multiplication coefficients for fuel elements irradiated by spallation neutrons

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

Bhatia, Chitra; Kumar, V.

2010-02-15

A neutron multiplication coefficient, k{sub eff}, has been estimated for spallation neutron flux using the data of spectrum average cross sections of all absorption, fission, and nonelastic reaction channels of {sup 232}Th, {sup 238}U, {sup 235}U, and {sup 233}U fuel elements. It has been revealed that in spallation neutron flux (i) nonfission, nonabsorption reactions play an important role in the calculation of k{sub eff}, (ii) one can obtain a high value of k{sub eff} even for fertile {sup 232}Th fuel, which is hardly possible in a conventional fast reactor, and (iii) spectrum average absorption cross sections of neutron poisons ofmore » a conventional reactor are relatively very small.« less

RER SPECTRA OBTAINED WITH A MULTICRYSTAL SPECTROMETER

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

Austin, W.E.; Champion, W.R.

1959-11-01

Relative gamma spectra were obtained twenty feet from the Hadiation Effects Reactor. The measurements were made using a multicry-stal spectrometer. This design incorporates pair and anticompton spectrometers in combination. Two reactor configurations were used; with shield tanks empty- and water filled. The spectra were obtained before the fuel elements were run at high power. Consequently very little of the fission product spectrum is tntermined. (J.R.D.)

Unfolding Neutron Spectrum with Markov Chain Monte Carlo at MIT Research Reactor with He-3 Neutral Current Detectors

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

Leder, A.; Anderson, A. J.; Billard, J.

2018-02-02

The Ricochet experiment seeks to measure Coherent (neutral-current) Elastic Neutrino-Nucleus Scattering (CEνNS) using dark-matter-style detectors with sub-keV thresholds placed near a neutrino source, such as the MIT (research) Reactor (MITR), which operates at 5.5 MW generating approximately 2.2 × 1018 ν/second in its core. Currently, Ricochet is characterizing the backgrounds at MITR, the main component of which comes in the form of neutrons emitted from the core simultaneous with the neutrino signal. To characterize this background, we wrapped Bonner cylinders around a 32He thermal neutron detector, whose data was then unfolded via a Markov Chain Monte Carlo (MCMC) to produce a neutron energymore » spectrum across several orders of magnitude. We discuss the resulting spectrum and its implications for deploying Ricochet at the MITR site as well as the feasibility of reducing this background level via the addition of polyethylene shielding around the detector setup.« less

Unfolding neutron spectrum with Markov Chain Monte Carlo at MIT research Reactor with He-3 Neutral Current Detectors

NASA Astrophysics Data System (ADS)

Leder, A.; Anderson, A. J.; Billard, J.; Figueroa-Feliciano, E.; Formaggio, J. A.; Hasselkus, C.; Newman, E.; Palladino, K.; Phuthi, M.; Winslow, L.; Zhang, L.

2018-02-01

The Ricochet experiment seeks to measure Coherent (neutral-current) Elastic Neutrino-Nucleus Scattering (CEνNS) using dark-matter-style detectors with sub-keV thresholds placed near a neutrino source, such as the MIT (research) Reactor (MITR), which operates at 5.5 MW generating approximately 2.2 × 1018 ν/second in its core. Currently, Ricochet is characterizing the backgrounds at MITR, the main component of which comes in the form of neutrons emitted from the core simultaneous with the neutrino signal. To characterize this background, we wrapped Bonner cylinders around a 32He thermal neutron detector, whose data was then unfolded via a Markov Chain Monte Carlo (MCMC) to produce a neutron energy spectrum across several orders of magnitude. We discuss the resulting spectrum and its implications for deploying Ricochet at the MITR site as well as the feasibility of reducing this background level via the addition of polyethylene shielding around the detector setup.

Measuring Neutron Spectrum at MIT Research Reactor Utilizing He-3 Bonner Cylinder Approach with an Unfolding Analysis

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

Leder, A.; Anderson, A. J.; Billard, J.

2017-10-02

The Ricochet experiment seeks to measure Coherent (neutral-current) Elastic Neutrino-Nucleus Scattering using dark-matter-style detectors with sub-keV thresholds placed near a neutrino source, such as the MIT (research) Reactor (MITR), which operates at 5.5 MW generating approximately 2.2e18 neutrinos/second at the core. Currently, Ricochet is characterizing the backgrounds at MITR, the main component of which comes in the form of neutrons emitted from the core simultaneous with the neutrino signal. To characterize this background, we wrapped a Bonner cylinder around a He-3 thermal neutron detector, whose data was then unfolded to produce a neutron energy spectrum across several orders of magnitude.more » We discuss the resulting spectrum and its implications for deploying Ricochet in the future at the MITR site as well as the feasibility of reducing this background level via the addition of polyethylene shielding around the detector setup.« less

Reactor operations informal monthly report, May 1, 1995--May 31, 1995

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

Hauptman, H.M.; Petro, J.N.; Jacobi, O.

1995-05-01

This document is an informal progress report for the operational performance of the Brookhaven Medical Research Reactor, and the Brookhaven High Flux Beam Reactor, for the month of May, 1995. Both machines ran well during this period, with no reportable instrumentation problems, all scheduled maintenance performed, and only one reportable occurance, involving a particle on Vest Button, Personnel Radioactive Contamination.

A roadmap for nuclear energy technology

NASA Astrophysics Data System (ADS)

Sofu, Tanju

2018-01-01

The prospects for the future use of nuclear energy worldwide can best be understood within the context of global population growth, urbanization, rising energy need and associated pollution concerns. As the world continues to urbanize, sustainable development challenges are expected to be concentrated in cities of the lower-middle-income countries where the pace of urbanization is fastest. As these countries continue their trajectory of economic development, their energy need will also outpace their population growth adding to the increased demand for electricity. OECD IEA's energy system deployment pathway foresees doubling of the current global nuclear capacity by 2050 to reduce the impact of rapid urbanization. The pending "retirement cliff" of the existing U.S. nuclear fleet, representing over 60 percent of the nation's emission-free electricity, also poses a large economic and environmental challenge. To meet the challenge, the U.S. DOE has developed the vision and strategy for development and deployment of advanced reactors. As part of that vision, the U.S. government pursues programs that aim to expand the use of nuclear power by supporting sustainability of the existing nuclear fleet, deploying new water-cooled large and small modular reactors to enable nuclear energy to help meet the energy security and climate change goals, conducting R&D for advanced reactor technologies with alternative coolants, and developing sustainable nuclear fuel cycle strategies. Since the current path relying heavily on water-cooled reactors and "once-through" fuel cycle is not sustainable, next generation nuclear energy systems under consideration aim for significant advances over existing and evolutionary water-cooled reactors. Among the spectrum of advanced reactor options, closed-fuel-cycle systems using reactors with fast-neutron spectrum to meet the sustainability goals offer the most attractive alternatives. However, unless the new public-private partnership models emerge to tackle the licensing and demonstration challenges for these advanced reactor concepts, realization of their enormous potential is not likely, at least in the U.S.

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

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

Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.

1983-06-01

During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

Sodium leak detection system for liquid metal cooled nuclear reactors

DOEpatents

Modarres, Dariush

1991-01-01

A light source is projected across the gap between the containment vessel and the reactor vessel. The reflected light is then analyzed with an absorption spectrometer. The presence of any sodium vapor along the optical path results in a change of the optical transmissivity of the media. Since the absorption spectrum of sodium is well known, the light source is chosen such that the sensor is responsive only to the presence of sodium molecules. The optical sensor is designed to be small and require a minimum of amount of change to the reactor containment vessel.

Staged depressurization system

DOEpatents

Schulz, T.L.

1993-11-02

A nuclear reactor having a reactor vessel disposed in a containment shell is depressurized in stages using depressurizer valves coupled in fluid communication with the coolant circuit. At least one sparger submerged in the in-containment refueling water storage tank which can be drained into the containment sump communicates between one or more of the valves and an inside of the containment shell. The depressurizer valves are opened in stages, preferably at progressively lower coolant levels and for opening progressively larger flowpaths to effect depressurization through a number of the valves in parallel. The valves can be associated with a pressurizer tank in the containment shell, coupled to a coolant outlet of the reactor. At least one depressurization valve stage openable at a lowest pressure is coupled directly between the coolant circuit and the containment shell. The reactor is disposed in the open sump in the containment shell, and a further valve couples the open sump to a conduit coupling the refueling water storage tank to the coolant circuit for adding water to the coolant circuit, whereby water in the containment shell can be added to the reactor from the open sump. 4 figures.

Staged depressurization system

DOEpatents

Schulz, Terry L.

1993-01-01

A nuclear reactor having a reactor vessel disposed in a containment shell is depressurized in stages using depressurizer valves coupled in fluid communication with the coolant circuit. At least one sparger submerged in the in-containment refueling water storage tank which can be drained into the containment sump communicates between one or more of the valves and an inside of the containment shell. The depressurizer valves are opened in stages, preferably at progressively lower coolant levels and for opening progressively larger flowpaths to effect depressurization through a number of the valves in parallel. The valves can be associated with a pressurizer tank in the containment shell, coupled to a coolant outlet of the reactor. At least one depressurization valve stage openable at a lowest pressure is coupled directly between the coolant circuit and the containment shell. The reactor is disposed in the open sump in the containment shell, and a further valve couples the open sump to a conduit coupling the refueling water storage tank to the coolant circuit for adding water to the coolant circuit, whereby water in the containment shell can be added to the reactor from the open sump.

Correlating Fast Fluence to dpa in Atypical Locations

NASA Astrophysics Data System (ADS)

Drury, Thomas H.

2016-02-01

Damage to a nuclear reactor's materials by high-energy neutrons causes changes in the ductility and fracture toughness of the materials. The reactor vessel and its associated piping's ability to withstand stress without brittle fracture are paramount to safety. Theoretically, the material damage is directly related to the displacements per atom (dpa) via the residual defects from induced displacements. However in practice, the material damage is based on a correlation to the high-energy (E > 1.0 MeV) neutron fluence. While the correlated approach is applicable when the material in question has experienced the same neutron spectrum as test specimens which were the basis of the correlation, this approach is not generically acceptable. Using Monte Carlo and discrete ordinates transport codes, the energy dependent neutron flux is determined throughout the reactor structures and the reactor vessel. Results from the models provide the dpa response in addition to the high-energy neutron flux. Ratios of dpa to fast fluence are calculated throughout the models. The comparisons show a constant ratio in the areas of historical concern and thus the validity of the correlated approach to these areas. In regions above and below the fuel however, the flux spectrum has changed significantly. The correlated relationship of material damage to fluence is not valid in these regions without adjustment. An adjustment mechanism is proposed.

Dual annular rotating [open quotes]windowed[close quotes] nuclear reflector reactor control system

DOEpatents

Jacox, M.G.; Drexler, R.L.; Hunt, R.N.M.; Lake, J.A.

1994-03-29

A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core. 4 figures.

SoLid Detector Technology

NASA Astrophysics Data System (ADS)

Labare, Mathieu

2017-09-01

SoLid is a reactor anti-neutrino experiment where a novel detector is deployed at a minimum distance of 5.5 m from a nuclear reactor core. The purpose of the experiment is three-fold: to search for neutrino oscillations at a very short baseline; to measure the pure 235U neutrino energy spectrum; and to demonstrate the feasibility of neutrino detectors for reactor monitoring. This report presents the unique features of the SoLid detector technology. The technology has been optimised for a high background environment resulting from low overburden and the vicinity of a nuclear reactor. The versatility of the detector technology is demonstrated with a 288 kg detector prototype which was deployed at the BR2 nuclear reactor in 2015. The data presented includes both reactor on, reactor off and calibration measurements. The measurement results are compared with Monte Carlo simulations. The 1.6t SoLid detector is currently under construction, with an optimised design and upgraded material technology to enhance the detector capabilities. Its deployement on site is planned for the begin of 2017 and offers the prospect to resolve the reactor anomaly within about two years.

Is the Autism Treatment Evaluation Checklist a Useful Tool for Monitoring Progress in Children with Autism Spectrum Disorders?

ERIC Educational Resources Information Center

Magiati, I.; Moss, J.; Yates, R.; Charman, T.; Howlin, P.

2011-01-01

Background: There are few well validated brief measures that can be used to assess the general progress of young children with autism spectrum disorders (ASD) over time. In the present study, the Autism Treatment Evaluation Checklist (ATEC) was used as part of a comprehensive assessment battery to monitor the progress of 22 school-aged children…

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

Li, C.; Yu, G.; Wang, K.

The physical designs of the new concept reactors which have complex structure, various materials and neutronic energy spectrum, have greatly improved the requirements to the calculation methods and the corresponding computing hardware. Along with the widely used parallel algorithm, heterogeneous platforms architecture has been introduced into numerical computations in reactor physics. Because of the natural parallel characteristics, the CPU-FPGA architecture is often used to accelerate numerical computation. This paper studies the application and features of this kind of heterogeneous platforms used in numerical calculation of reactor physics through practical examples. After the designed neutron diffusion module based on CPU-FPGA architecturemore » achieves a 11.2 speed up factor, it is proved to be feasible to apply this kind of heterogeneous platform into reactor physics. (authors)« less

Current Results of NEUTRINO-4 Experiment

NASA Astrophysics Data System (ADS)

Serebrov, A.; Ivochkin, V.; Samoilov, R.; Fomin, A.; Polyushkin, A.; Zinoviev, V.; Neustroev, P.; Golovtsov, V.; Chernyj, A.; Zherebtsov, O.; Martemyanov, V.; Tarasenkov, V.; Aleshin, V.; Petelin, A.; Izhutov, A.; Tuzov, A.; Sazontov, S.; Ryazanov, D.; Gromov, M.; Afanasiev, V.; Zaytsev, M.; Chaikovskii, M.

2017-12-01

The main goal of experiment “Neutrino-4” is to search for the oscillation of reactor antineutrino to a sterile state. Experiment is conducted on SM-3 research reactor (Dimitrovgrad, Russia). Data collection with full-scale detector with liquid scintillator volume of 3m3 was started in June 2016. We present the results of measurements of reactor antineutrino flux dependence on the distance in range 6- 12 meters from the center of the reactor. At that distance range, the fit of experimental dependence has good agreement with the law 1/L2. Which means, at achieved during the data collecting accuracy level oscillations to sterile state are not observed. In addition, the spectrum of prompt signals of neutrino-like events at different distances have been presented.

Summary of space nuclear reactor power systems, 1983 - 1992

NASA Astrophysics Data System (ADS)

Buden, D.

1993-08-01

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987-88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

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

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. Themore » objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.« less

PATHFINDER ATOMIC POWER PLANT TECHNICAL PROGRESS REPORT FOR JULY 1, 1959- SEPTEMBER 30, 1959

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

None

1960-10-31

ABS>Fuel Element Research and Development. Dynamic and static corrosion tests on 8001 Al were completed. Annealmmmg of 1100 cladding on 5083 and M400 cladding on X2219 were tested at 500 deg C, and investigation continued on producing X8101 Al alloy cladding in tube plates by extrusion. Boiler fuel element capsule irradiation tests and subassembly tests are described Heat transfer loop studies and fuel fabrication for the critical facility are reported. Boiler fuel element mechanical design and testing progress is desc ribed. and the superheater fuel element temperature evaluating routine is discussed. Low- enrichment superheater fuel element development included design studiesmore » and stainless steel powder and UO/sub 2/ powder fabrication studies Reactor Mechanical Studies. Research is reported on vessel and structure design, fabrication, and testing, recirculation system design, steam separator tests, and control rod studies. Nuclear Analysis. Reactor physics studies are reported on nuclear constants, baffle plate analysis, comparison of core representations, delayed neutron fraction. and shielding analysis of the reactor building. Reactor and system dynamics and critical experiments were also studied. Chemistry. Progress is reported on recombiner. radioactive gas removal and storage, ion exchanger and radiochemical processing. (For preceding period see ACNP-5915.) (T.R.H.)« less

Exploratory study of several advanced nuclear-MHD power plant systems.

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.; Rosa, R. J.; Yang, Y. Y.

1973-01-01

In order for efficient multimegawatt closed cycle nuclear-MHD systems to become practical, long-life gas cooled reactors with exit temperatures of about 2500 K or higher must be developed. Four types of nuclear reactors which have the potential of achieving this goal are the NERVA-type solid core reactor, the colloid core (rotating fluidized bed) reactor, the 'light bulb' gas core reactor, and the 'coaxial flow' gas core reactor. Research programs aimed at developing these reactors have progressed rapidly in recent years so that prototype power reactors could be operating by 1980. Three types of power plant systems which use these reactors have been analyzed to determine the operating characteristics, critical parameters and performance of these power plants. Overall thermal efficiencies as high as 80% are projected, using an MHD turbine-compressor cycle with steam bottoming, and slightly lower efficiencies are projected for an MHD motor-compressor cycle.

Schedule and status of irradiation experiments

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

Rowcliffe, A.F.; Grossbeck, M.L.; Robertson, J.P.

1998-09-01

The current status of reactor irradiation experiments is presented in tables summarizing the experimental objectives, conditions, and schedule. Currently, the program has one irradiation experiment in reactor and five experiments in the design or construction stages. Postirradiation examination and testing is in progress on ten experiments.

A prototype experiment for cooperative monitoring of nuclear reactors with cubic meter scale antineutrino detectors

NASA Astrophysics Data System (ADS)

Bernstein, A.; Allen, M.; Bowden, N.; Brennan, J.; Carr, D. J.; Estrada, J.; Hagmann, C.; Lund, J. C.; Madden, N. W.; Winant, C. D.

2005-09-01

Our Lawrence Livermore National Laboratory/Sandia National Laboratories collaboration has deployed a cubic-meter-scale antineutrino detector to demonstrate non-intrusive and automatic monitoring of the power levels and plutonium content of a nuclear reactor. Reactor monitoring of this kind is required for all non-nuclear weapons states under the Nuclear Nonproliferation Treaty (NPT), and is implemented by the International Atomic Energy Agency (IAEA). Since the antineutrino count rate and energy spectrum depend on the relative yields of fissioning isotopes in the reactor core, changes in isotopic composition can be observed without ever directly accessing the core. Data from a cubic meter scale antineutrino detector, coupled with the well-understood principles that govern the core's evolution in time, can be used to determine whether the reactor is being operated in an illegitimate way. Our group has deployed a detector at the San Onofre reactor site in California to demonstrate this concept. This paper describes the concept and shows preliminary results from 8 months of operation.

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

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

Characterization of fast neutron spectrum in the TRIGA for hardness testing of electronic components

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

Nelson, George W.

1986-07-01

Argonne National Laboratory-West, operated by the University of Chicago, is located near Idaho Falls, ID, on the Idaho National Engineering Laboratory Site. ANL-West performs work in support of the Liquid Metal Fast Breeder Reactor Program (LMFBR) sponsored by the United States Department of Energy. The NRAD reactor is located at the Argonne Site within the Hot Fuel Examination Facility/North, a large hot cell facility where both non-destructive and destructive examinations are performed on highly irradiated reactor fuels and materials in support of the LMFBR program. The NRAD facility utilizes a 250-kW TRIGA reactor and is completely dedicated to neutron radiographymore » and the development of radiography techniques. Criticality was first achieved at the NRAD reactor in October of 1977. Since that time, a number of modifications have been implemented to improve operational efficiency and radiography production. This paper describes the modifications and changes that significantly improved operational efficiency and reliability of the reactor and the essential auxiliary reactor systems. (author)« less

Neutron Capture and the Antineutrino Yield from Nuclear Reactors.

PubMed

Huber, Patrick; Jaffke, Patrick

2016-03-25

We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ∼0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

Hanford Works monthly report, December 1950

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

Prout, G.R.

1951-01-22

This is a progress report of the production reactors on the Hanford Reservation for the month of December 1950. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, April 1952

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

Prout, G.R.

1952-05-20

This is a progress report of the production reactors on the Hanford Reservation for the month of April 1952. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, August 1950

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

Prout, G.R.

1950-09-18

This is a progress report of the production reactors on the Hanford Reservation for the month of August 1950. This report takes each division (e.g. manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, May 1951

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

Prout, G.R.

1951-06-21

This is a progress report of the production reactors on the Hanford Reservation for the month of May 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, December 1951

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

Prout, G.R.

1952-01-22

This is a progress report of the production reactors on the Hanford Reservation for the month of December 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, March 1951

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

Prout, G.R.

1951-04-20

This is a progress report of the production reactors on the Hanford Reservation for the month of March 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, July 1951

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

Prout, G.R.

This is a progress report of the production reactors on the Hanford Reservation for the month of July 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, March 1952

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

Prout, G.R.

1952-04-18

This is a progress report of the production reactors on the Hanford Reservation for the month of April 1952. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, July 1952

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

Prout, G.R.

1952-08-15

This is a progress report of the production reactors on the Hanford Reservation for the month of July 1952. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford works monthly report, September 1951

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

Prout, G.R.

This is a progress report of the production reactors on the Hanford Reservation for the month of September 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, January 1952

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

Prout, G.R.

This is a progress report of the production reactors on the Hanford Reservation for the month of January 1952. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, August 1951

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

Not Available

1951-09-24

This is a progress report of the production reactors on the Hanford Reservation for the month of August 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, July 1950

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

Prout, G.R.

1950-08-18

This is a progress report of the production reactors on the Hanford Reservation for the month of July 1950. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, November 1951

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

Prout, G.R.

1951-12-21

This is a progress report of the production reactors on the Hanford Reservation for the month of November 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, October 1950

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

Prout, G.R.

1950-11-20

This is a progress report of the production reactors on the Hanford Reservation for the month of October 1950. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, September 1950

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

Prout, G.R.

1950-10-20

This is a progress report of the production reactors on the Hanford Reservation for the month of September 1950. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, November 1950

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

Prout, G.R.

1950-12-20

This is a progress report of the production reactors on the Hanford Reservation for the month of November 1950. This report takes each division (e.g. manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

FERRET-SAND II physics-dosimetry analysis for N Reactor Pressure Tubes 2954, 3053 and 1165 using a WIMS calculated input spectrum

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

McElroy, W.N.; Kellogg, L.S.; Matsumoto, W.Y.

1988-05-01

This report is in response to a request from Westinghouse Hanford Company (WHC) that the PNL National Dosimetry Center (NDC) perform physics-dosimetry analyses (E > MeV) for N Reactor Pressure Tubes 2954 and 3053. As a result of these analyses, and recommendations for additional studies, two physics-dosimetry re-evaluations for Pressure Tube 1165 were also accomplished. The primary objective of Pacific Northwest Laboratories' (PNL) National Dosimetry Center (NDC) physics-dosimetry work for N Reactor was to provide FERRET-SAND II physics-dosimetry results to assist in the assessment of neutron radiation-induced changes in the physical and mechanical properties of N Reactor pressure tubes. 15more » refs., 6 figs., 5 tabs.« less

NEOS Data and the Origin of the 5 MeV Bump in the Reactor Antineutrino Spectrum

NASA Astrophysics Data System (ADS)

Huber, Patrick

2017-01-01

We perform a combined analysis of recent NEOS and Daya Bay data on the reactor antineutrino spectrum. This analysis includes approximately 1.5 million antineutrino events, which is the largest neutrino event sample analyzed to date. We use a double ratio which cancels flux model dependence and related uncertainties as well as the effects of the detector response model. We find at 3-4 standard deviation significance level, that plutonium-239 and plutonium-241 are disfavored as the single source for the so-called 5 MeV bump. This analysis method has general applicability and, in particular, with higher statistics data sets, will be able to shed significant light on the issue of the bump. With some caveats, this should also allow us to improve the sensitivity for sterile neutrino searches in NEOS.

Search for eV Sterile Neutrinos - The Stereo Experiment

NASA Astrophysics Data System (ADS)

Haser, J.; Stereo Collaboration

2017-07-01

In the recent years, major milestones in neutrino physics were accomplished at nuclear reactors: the smallest neutrino mixing angle $\\theta_{13}$ was determined with high precision and the emitted antineutrino spectrum was measured at unprecedented resolution. However, two anomalies, the first one related to the absolute flux and the second one to the spectral shape, have yet to be solved. The flux anomaly is known as the Reactor Antineutrino Anomaly and could be caused by the existence of a light sterile neutrino participating in the neutrino oscillation phenomenon. Introducing a sterile state implies the presence of a fourth mass eigenstate, global fits favour oscillation parameters around $\\sin^2({2\\theta}) \\approx 0.09$ and $\\Delta m^2 \\approx 1\\,\\mathrm{eV}^2$. The Stereo experiment was built to finally solve this puzzle. It is one of the first running experiments built to search for eV sterile neutrinos and takes data since end of 2016 at ILL Grenoble (France). At a short baseline of 10 metres, it measures the antineutrino flux and spectrum emitted by a compact research reactor. The segmentation of the detector in six target cells allows for measurements of the neutrino spectrum at multiple baselines. An active-sterile flavour oscillation could be unambiguously detected, as it distorts the spectral shape of each cell's measurement differently. This contribution gives an overview on the Stereo experiment, along with details on the detector design, detection principle and the current status of data analysis.

Experimental detailed power distribution in a fast spectrum thermionic reactor fuel element at the core/BeO reflector interface region

NASA Technical Reports Server (NTRS)

Klann, P. G.; Lantz, E.

1973-01-01

A zero-power critical assembly was designed, constructed, and operated for the prupose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power conversion system. The critical assembly was modified to simulate a fast spectrum advanced thermionics reactor by: (1) using BeO as a reflector in place of some of the existing molybdenum, (2) substituting Nb-1Zr tubing for some of the existing Ta tubing, and (3) inserting four full-scale mockups of thermionic type fuel elements near the core and BeO reflector boundary. These mockups were surrounded with a buffer zone having the equivalent thermionic core composition. In addition to measuring the critical mass of this thermionic configuration, a detailed power distribution in one of the thermionic element stages in the mixed spectrum region was measured. A power peak to average ratio of two was observed for this fuel stage at the midplane of the core and adjacent to the reflector. Also, the power on the outer surface adjacent to the BeO was slightly more than a factor of two larger than the power on the inside surface of a 5.08 cm (2.0 in.) high annular fuel segment with a 2.52 cm (0.993 in. ) o.d. and a 1.86 cm (0.731 in.) i.d.

Neutron flux and gamma dose measurement in the BNCT irradiation facility at the TRIGA reactor of the University of Pavia

NASA Astrophysics Data System (ADS)

Bortolussi, S.; Protti, N.; Ferrari, M.; Postuma, I.; Fatemi, S.; Prata, M.; Ballarini, F.; Carante, M. P.; Farias, R.; González, S. J.; Marrale, M.; Gallo, S.; Bartolotta, A.; Iacoviello, G.; Nigg, D.; Altieri, S.

2018-01-01

University of Pavia is equipped with a TRIGA Mark II research nuclear reactor, operating at a maximum steady state power of 250 kW. It has been used for many years to support Boron Neutron Capture Therapy (BNCT) research. An irradiation facility was constructed inside the thermal column of the reactor to produce a sufficient thermal neutron flux with low epithermal and fast neutron components, and low gamma dose. In this irradiation position, the liver of two patients affected by hepatic metastases from colon carcinoma were irradiated after borated drug administration. The facility is currently used for cell cultures and small animal irradiation. Measurements campaigns have been carried out, aimed at characterizing the neutron spectrum and the gamma dose component. The neutron spectrum has been measured by means of multifoil neutron activation spectrometry and a least squares unfolding algorithm; gamma dose was measured using alanine dosimeters. Results show that in a reference position the thermal neutron flux is (1.20 ± 0.03) ×1010 cm-2 s-1 when the reactor is working at the maximum power of 250 kW, with the epithermal and fast components, respectively, 2 and 3 orders of magnitude lower than the thermal component. The ratio of the gamma dose with respect to the thermal neutron fluence is 1.2 ×10-13 Gy/(n/cm2).

INCAS: an analytical model to describe displacement cascades

NASA Astrophysics Data System (ADS)

Jumel, Stéphanie; Claude Van-Duysen, Jean

2004-07-01

REVE (REactor for Virtual Experiments) is an international project aimed at developing tools to simulate neutron irradiation effects in Light Water Reactor materials (Fe, Ni or Zr-based alloys). One of the important steps of the project is to characterise the displacement cascades induced by neutrons. Accordingly, the Department of Material Studies of Electricité de France developed an analytical model based on the binary collision approximation. This model, called INCAS (INtegration of CAScades), was devised to be applied on pure elements; however, it can also be used on diluted alloys (reactor pressure vessel steels, etc.) or alloys composed of atoms with close atomic numbers (stainless steels, etc.). INCAS describes displacement cascades by taking into account the nuclear collisions and electronic interactions undergone by the moving atoms. In particular, it enables to determine the mean number of sub-cascades induced by a PKA (depending on its energy) as well as the mean energy dissipated in each of them. The experimental validation of INCAS requires a large effort and could not be carried out in the framework of the study. However, it was verified that INCAS results are in conformity with those obtained from other approaches. As a first application, INCAS was applied to determine the sub-cascade spectrum induced in iron by the neutron spectrum corresponding to the central channel of the High Flux Irradiation Reactor of Oak Ridge National Laboratory.

Hanford Atomic Products Operation monthly report, April 1954

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

McCune, F.K.

1954-05-21

This is a progress report of the production reactors on the Hanford Reservation for the month of April 1954. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Atomic Products Operation monthly report, March 1953

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

McCune, F.K.

1953-04-22

This is a progress report of the production reactors on the Hanford Reservation for the month of March 1953. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

A PROGRAM OF RESEARCH ON MECHANICAL METALLURGY AS RELATED TO FUEL-ELEMENT FABRICATION. Summary Report, January 1-September 30, 1961

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

Trozera, T.A.; White, J.L.; Chambers, R.H.

Research progress on mechanical metallurgy of reactor materials is reported in three sections: deformation characteristics of reactor materials, stored energy of cold work, and microplastic propenties and mechanical relaxation spectra of very pure refractory bcc metals. (M.C.G.)

Hanford Atomic Products Operation monthly report, April 1953

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

McCune, F.K.

1953-05-20

This is a progress report of the production reactors on the Hanford Reservation for the month of April 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Atomic Products Operation monthly report, January 1954

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

McCune, F.K.

1954-02-25

This is a progress report of the production reactors on the Hanford Reservation for the month of January 1954. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes the accomplishments and employee relations for that month.

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

None

Progress is reported on fundamental research in: crystal physics, reactions at metal surfaces, spectroscopy of ionic media, structure of metals, theory of alloying, physical properties, sintering, deformation of crystalline solids, x ray diffraction, metallurgy of superconducting materials, and electron microscope studies. Long-randge applied research studies were conducted for: zirconium metallurgy, materials compatibility, solid reactions, fuel element development, mechanical properties, non-destructive testing, and high-temperature materials. Reactor development support work was carried out for: gas-cooled reactor program, molten-salt reactor, high-flux isotope reactor, space-power program, thorium-utilization program, advanced-test reactor, Army Package Power Reactor, Enrico Fermi fast-breeder reactor, and water desalination program. Other programmore » activities, for which research was conducted, included: thermonuclear project, transuraniunn program, and post-irradiation examination laboratory. Separate abstracts were prepared for 30 sections of the report. (B.O.G.)« less

AGM2015: Antineutrino Global Map 2015

PubMed Central

Usman, S.M.; Jocher, G.R.; Dye, S.T.; McDonough, W.F.; Learned, J.G.

2015-01-01

Every second greater than 1025 antineutrinos radiate to space from Earth, shining like a faint antineutrino star. Underground antineutrino detectors have revealed the rapidly decaying fission products inside nuclear reactors, verified the long-lived radioactivity inside our planet, and informed sensitive experiments for probing fundamental physics. Mapping the anisotropic antineutrino flux and energy spectrum advance geoscience by defining the amount and distribution of radioactive power within Earth while critically evaluating competing compositional models of the planet. We present the Antineutrino Global Map 2015 (AGM2015), an experimentally informed model of Earth’s surface antineutrino flux over the 0 to 11 MeV energy spectrum, along with an assessment of systematic errors. The open source AGM2015 provides fundamental predictions for experiments, assists in strategic detector placement to determine neutrino mass hierarchy, and aids in identifying undeclared nuclear reactors. We use cosmochemically and seismologically informed models of the radiogenic lithosphere/mantle combined with the estimated antineutrino flux, as measured by KamLAND and Borexino, to determine the Earth’s total antineutrino luminosity at . We find a dominant flux of geo-neutrinos, predict sub-equal crust and mantle contributions, with ~1% of the total flux from man-made nuclear reactors. PMID:26323507

AGM2015: Antineutrino Global Map 2015.

PubMed

Usman, S M; Jocher, G R; Dye, S T; McDonough, W F; Learned, J G

2015-09-01

Every second greater than 10(25) antineutrinos radiate to space from Earth, shining like a faint antineutrino star. Underground antineutrino detectors have revealed the rapidly decaying fission products inside nuclear reactors, verified the long-lived radioactivity inside our planet, and informed sensitive experiments for probing fundamental physics. Mapping the anisotropic antineutrino flux and energy spectrum advance geoscience by defining the amount and distribution of radioactive power within Earth while critically evaluating competing compositional models of the planet. We present the Antineutrino Global Map 2015 (AGM2015), an experimentally informed model of Earth's surface antineutrino flux over the 0 to 11 MeV energy spectrum, along with an assessment of systematic errors. The open source AGM2015 provides fundamental predictions for experiments, assists in strategic detector placement to determine neutrino mass hierarchy, and aids in identifying undeclared nuclear reactors. We use cosmochemically and seismologically informed models of the radiogenic lithosphere/mantle combined with the estimated antineutrino flux, as measured by KamLAND and Borexino, to determine the Earth's total antineutrino luminosity at . We find a dominant flux of geo-neutrinos, predict sub-equal crust and mantle contributions, with ~1% of the total flux from man-made nuclear reactors.

On the possible use of the MASURCA reactor as a flexible, high-intensity, fast neutron beam facility

NASA Astrophysics Data System (ADS)

Dioni, Luca; Jacqmin, Robert; Sumini, Marco; Stout, Brian

2017-09-01

In recent work [1, 2], we have shown that the MASURCA research reactor could be used to deliver a fairly-intense continuous fast neutron beam to an experimental room located next to the reactor core. As a consequence of the MASURCA favorable characteristics and diverse material inventories, the neutron beam intensity and spectrum can be further tailored to meet the users' needs, which could be of interest for several applications. Monte Carlo simulations have been performed to characterize in detail the extracted neutron (and photon) beam entering the experimental room. These numerical simulations were done for two different bare cores: A uranium metallic core (˜30% 235U enriched) and a plutonium oxide core (˜25% Pu fraction, ˜78% 239Pu). The results show that the distinctive resonance energy structures of the two core leakage spectra are preserved at the channel exit. As the experimental room is large enough to house a dedicated set of neutron spectrometry instruments, we have investigated several candidate neutron spectrum measurement techniques, which could be implemented to guarantee well-defined, repeatable beam conditions to users. Our investigation also includes considerations regarding the gamma rays in the beams.

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

Meadows, J.; Smith, D.; Greenwood, L.

Four sample packets containing elemental Ti, Fe, Ni, Cu, Nb, Ag, Eu, Tb and Hf have been irradiated in three distinct accelerator neutron fields, at Argonne National Laboratory and Los Alamos National Laboratory, USA, and Japan Atomic Energy Research Institute, Tokai, Japan. The acquired experimental data include differential cross sections and integral cross sections for the continuum neutron spectrum produced by 7-MeV deuterons incident on thick Be-metal target. The U-238(n,f) cross section was also measured at 10.3 MeV as a consistency check on the experimental technique. This the third progress report on a project which has been carried out undermore » the auspices of an IAEA Coordinated Research Program entitled ``Activation Cross Sections for the Generation Of Long-lived Radionuclides of Importance in Fusion Reactor Technology``. The present report provides the latest results from this work. Comparison is made between the 14.7-MeV cross-section values obtained from the separate investigations at Argonne and JAERI. Generally, good agreement observed within the experimental errors when consistent sample parameters, radioactivity decay data and reference cross values are employed. A comparison is also made between the experimental results and those derived from calculations using a nuclear model. Experimental neutron information on the Be(d,n) neutron spectrum was incorporated in the comparisons for the integral results. The agreement is satisfactory considering the various uncertainties that are involved.« less

Measuring Developmental Progress of Children with Autism Spectrum Disorder on School Entry Using Parent Report

ERIC Educational Resources Information Center

Charman, Tony; Howlin, Patricia; Berry, Bryony; Prince, Emily

2004-01-01

Increasing numbers of children with autism spectrum disorder (ASD) are diagnosed in the preschool years, and their educational progress must be monitored. Parent questionnaire data can augment psychometric assessments and individual planning at low cost. One hundred and twenty-five parents of UK children who entered dedicated autism primary…

Reactivity effects of moderator voids

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

Ahlfeld, C.E.; Pryor, R.J.

1975-01-01

Reactivity worths for large moderator voids similar to those produced by steaming in postulated reactor transients were measured in the Process Development Pile (PDP) reactor. The experimental results were compared to the computed void worths obtained from techniques currently used in routine safety analyses. Neutron energy spectrum measurements were used to verify a modified lattice pattern that correctly computed the measured spectrum, and consequently, improved macroscopic cross sections. In addition, a special two-dimensional transport calculation was performed to obtain an axially defined diffusion coefficient for the void region. The combination of the modified lattice calculations and the axial diffusion coefficientmore » yielded void reactivity worths which agreed very well with experiment. It was concluded that the computational modules available in the JOSHUA system (GLASS, GRIMHX) would yield accurate void reactivity worths in SLR--SRP safety analysis studies, provided the above mentioned modifications were made.« less

A flavor symmetry model for bilarge leptonic mixing and the lepton masses

NASA Astrophysics Data System (ADS)

Ohlsson, Tommy; Seidl, Gerhart

2002-11-01

We present a model for leptonic mixing and the lepton masses based on flavor symmetries and higher-dimensional mass operators. The model predicts bilarge leptonic mixing (i.e., the mixing angles θ12 and θ23 are large and the mixing angle θ13 is small) and an inverted hierarchical neutrino mass spectrum. Furthermore, it approximately yields the experimental hierarchical mass spectrum of the charged leptons. The obtained values for the leptonic mixing parameters and the neutrino mass squared differences are all in agreement with atmospheric neutrino data, the Mikheyev-Smirnov-Wolfenstein large mixing angle solution of the solar neutrino problem, and consistent with the upper bound on the reactor mixing angle. Thus, we have a large, but not close to maximal, solar mixing angle θ12, a nearly maximal atmospheric mixing angle θ23, and a small reactor mixing angle θ13. In addition, the model predicts θ 12≃ {π}/{4}-θ 13.

Search for eV sterile neutrinos at a nuclear reactor — the Stereo project

NASA Astrophysics Data System (ADS)

Haser, J.; Stereo Collaboration

2016-05-01

The re-analyses of the reference spectra of reactor antineutrinos together with a revised neutrino interaction cross section enlarged the absolute normalization of the predicted neutrino flux. The tension between previous reactor measurements and the new prediction is significant at 2.7 σ and is known as “reactor antineutrino anomaly”. In combination with other anomalies encountered in neutrino oscillation measurements, this observation revived speculations about the existence of a sterile neutrino in the eV mass range. Mixing of this light sterile neutrino with the active flavours would lead to a modification of the detected antineutrino flux. An oscillation pattern in energy and space could be resolved by a detector at a distance of few meters from a reactor core: the neutrino detector of the Stereo project will be located at about 10 m distance from the ILL research reactor in Grenoble, France. Lengthwise separated in six target cells filled with 2 m3 Gd-loaded liquid scintillator in total, the experiment will search for a position-dependent distortion in the energy spectrum.

Current Development in Treatment and Hydrogen Energy Conversion of Organic Solid Waste

NASA Astrophysics Data System (ADS)

Shin, Hang-Sik

2008-02-01

This manuscript summarized current developments on continuous hydrogen production technologies researched in Korea advanced institute of science and technology (KAIST). Long-term continuous pilot-scale operation of hydrogen producing processes fed with non-sterile food waste exhibited successful results. Experimental findings obtained by the optimization processes of growth environments for hydrogen producing bacteria, the development of high-rate hydrogen producing strategies, and the feasibility tests for real field application could contribute to the progress of fermentative hydrogen production technologies. Three major technologies such as controlling dilution rate depending on the progress of acidogenesis, maintaining solid retention time independently from hydraulic retention time, and decreasing hydrogen partial pressure by carbon dioxide sparging could enhance hydrogen production using anaerobic leaching beds reactors and anaerobic sequencing batch reactors. These findings could contribute to stable, reliable and effective performances of pilot-scale reactors treating organic wastes.

Status of the US RERTR Program

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

Travelli, A.

1995-02-01

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1994 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1993 in collaboration with its many international partners. The RERTR Program has moved aggressively to support President Clinton`s nonproliferation policy and his goal {open_quotes}to minimize the use of highly-enriched uranium in civil nuclear programs{close_quotes}. An Environmental Assessment which addresses the urgent-relief acceptance of 409 spent fuel elements was completed, and the first shipment of spent fuel elements is scheduledmore » for this month. An Environmental Impact Statement addressing the acceptance of spent research reactor fuel containing enriched uranium of U.S. origin is scheduled for completion by the end of June 1995. The U.S. administration has decided to resume development of high-density LEU research reactor fuels. DOE funding and guidance are expected to begin soon. A preliminary plan for the resumption of fuel development has been prepared and is ready for implementation. The scope and main technical activities of a plan to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels was agreed upon by the RERTR Program and four Russian institutes lead by RDIPE. Both Secretary O`Leary and Minister Michailov have expressed strong support for this initiative. Joint studies have made significant progress, especially in assessing the technical and economic feasibility of using reduced enrichment fuels in the SAFARI-I reactor in South Africa and in the Advanced Neutron Source reactor under design at ORNL. Significant progress was achieved on several aspects of producing {sup 99}Mo from fission targets utilizing LEU instead of HEU to the achievement of the common goal.« less

Corrections on energy spectrum and scatterings for fast neutron radiography at NECTAR facility

NASA Astrophysics Data System (ADS)

Liu, Shu-Quan; Bücherl, Thomas; Li, Hang; Zou, Yu-Bin; Lu, Yuan-Rong; Guo, Zhi-Yu

2013-11-01

Distortions caused by the neutron spectrum and scattered neutrons are major problems in fast neutron radiography and should be considered for improving the image quality. This paper puts emphasis on the removal of these image distortions and deviations for fast neutron radiography performed at the NECTAR facility of the research reactor FRM- II in Technische Universität München (TUM), Germany. The NECTAR energy spectrum is analyzed and established to modify the influence caused by the neutron spectrum, and the Point Scattered Function (PScF) simulated by the Monte-Carlo program MCNPX is used to evaluate scattering effects from the object and improve image quality. Good analysis results prove the sound effects of the above two corrections.

Design of megawatt power level heat pipe reactors

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

Mcclure, Patrick Ray; Poston, David Irvin; Dasari, Venkateswara Rao

An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors.more » The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.« less

Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

NASA Astrophysics Data System (ADS)

Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

The slightly-enriched spectral shift control reactor. Final report, September 30, 1988--September 30, 1991

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

Martin, W.R.; Lee, J.C.; Larsen, E.W.

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technologymore » retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.« less

The slightly-enriched spectral shift control reactor

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

Martin, W.R.; Lee, J.C.; Larsen, E.W.

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technologymore » retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.« less

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Fielding, R.; Gan, J.

2007-09-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

Vibro-acoustic Imaging at the Breazeale Reactor

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

Smith, James Arthur; Jewell, James Keith; Lee, James Edwin

2016-09-01

The INL is developing Vibro-acoustic imaging technology to characterize microstructure in fuels and materials in spent fuel pools and within reactor vessels. A vibro-acoustic development laboratory has been established at the INL. The progress in developing the vibro-acoustic technology at the INL is the focus of this report. A successful technology demonstration was performed in a working TRIGA research reactor. Vibro-acoustic imaging was performed in the reactor pool of the Breazeale reactor in late September of 2015. A confocal transducer driven at a nominal 3 MHz was used to collect the 60 kHz differential beat frequency induced in a spentmore » TRIGA fuel rod and empty gamma tube located in the main reactor water pool. Data was collected and analyzed with the INLDAS data acquisition software using a short time Fourier transform.« less

Renewing Liquid Fueled Molten Salt Reactor Research and Development

NASA Astrophysics Data System (ADS)

Towell, Rusty; NEXT Lab Team

2016-09-01

Globally there is a desperate need for affordable, safe, and clean energy on demand. More than anything else, this would raise the living conditions of those in poverty around the world. An advanced reactor that utilizes liquid fuel and molten salts is capable of meeting these needs. Although, this technology was demonstrated in the Molten Salt Reactor Experiment (MSRE) at ORNL in the 60's, little progress has been made since the program was cancelled over 40 years ago. A new research effort has been initiated to advance the technical readiness level of key reactor components. This presentation will explain the motivation and initial steps for this new research initiative.

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

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

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

Development and application of the dynamic system doctor to nuclear reactor probabilistic risk assessments.

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

Kunsman, David Marvin; Aldemir, Tunc; Rutt, Benjamin

2008-05-01

This LDRD project has produced a tool that makes probabilistic risk assessments (PRAs) of nuclear reactors - analyses which are very resource intensive - more efficient. PRAs of nuclear reactors are being increasingly relied on by the United States Nuclear Regulatory Commission (U.S.N.R.C.) for licensing decisions for current and advanced reactors. Yet, PRAs are produced much as they were 20 years ago. The work here applied a modern systems analysis technique to the accident progression analysis portion of the PRA; the technique was a system-independent multi-task computer driver routine. Initially, the objective of the work was to fuse the accidentmore » progression event tree (APET) portion of a PRA to the dynamic system doctor (DSD) created by Ohio State University. Instead, during the initial efforts, it was found that the DSD could be linked directly to a detailed accident progression phenomenological simulation code - the type on which APET construction and analysis relies, albeit indirectly - and thereby directly create and analyze the APET. The expanded DSD computational architecture and infrastructure that was created during this effort is called ADAPT (Analysis of Dynamic Accident Progression Trees). ADAPT is a system software infrastructure that supports execution and analysis of multiple dynamic event-tree simulations on distributed environments. A simulator abstraction layer was developed, and a generic driver was implemented for executing simulators on a distributed environment. As a demonstration of the use of the methodological tool, ADAPT was applied to quantify the likelihood of competing accident progression pathways occurring for a particular accident scenario in a particular reactor type using MELCOR, an integrated severe accident analysis code developed at Sandia. (ADAPT was intentionally created with flexibility, however, and is not limited to interacting with only one code. With minor coding changes to input files, ADAPT can be linked to other such codes.) The results of this demonstration indicate that the approach can significantly reduce the resources required for Level 2 PRAs. From the phenomenological viewpoint, ADAPT can also treat the associated epistemic and aleatory uncertainties. This methodology can also be used for analyses of other complex systems. Any complex system can be analyzed using ADAPT if the workings of that system can be displayed as an event tree, there is a computer code that simulates how those events could progress, and that simulator code has switches to turn on and off system events, phenomena, etc. Using and applying ADAPT to particular problems is not human independent. While the human resources for the creation and analysis of the accident progression are significantly decreased, knowledgeable analysts are still necessary for a given project to apply ADAPT successfully. This research and development effort has met its original goals and then exceeded them.« less

ITER activities and fusion technology

NASA Astrophysics Data System (ADS)

Seki, M.

2007-10-01

At the 21st IAEA Fusion Energy Conference, 68 and 67 papers were presented in the categories of ITER activities and fusion technology, respectively. ITER performance prediction, results of technology R&D and the construction preparation provide good confidence in ITER realization. The superconducting tokamak EAST achieved the first plasma just before the conference. The construction of other new experimental machines has also shown steady progress. Future reactor studies stress the importance of down sizing and a steady-state approach. Reactor technology in the field of blanket including the ITER TBM programme and materials for the demonstration power plant showed sound progress in both R&D and design activities.

Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

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

J. Rempe; D. Knudson; J. Daw

2014-01-01

The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation.more » To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.« less

THE ARMOUR DUST FUELED REACTOR (ADFR). Quarterly Progress Report No. 2 For the Period May 21, 1958 to August 21, 1958

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

Loewe, W.E.; Krucoff, D.

1958-10-31

Study of the ADFR concept included experimental work on fuel dust suspension stability and redispersibility, erosion, and dust deposition using the fuel dust circulation loop. Some theoretical work was done in the areas of reactor safety and breeding. (For preceding period -see AECU-3827.) (auth)

Current status of SPINNORs designs

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

Su'ud, Zaki

2010-06-22

This study discuss about the SPINNOR (Small Power Reactor, Indonesia, No On-site Refuelling) and the VSPINNOR (Very Small Power Reactor, Indonesia, No On-site Refuelling) which are small lead-bismuth cooled nuclear power reactors with fast neutron spectrum that could be operated for more than 10 or 15 years without on-site refuelling. They are based on the concept of a long-life core reactor developed in Indonesia since early 1990 in collaboration with the Research Laboratory for Nuclear Reactors of the Tokyo Institute of Technology (RLNR TITech). The reactor cores are designed to have near zero (less then one effective delayed neutron fraction)more » burn-up reactivity swing during the whole course of their operation to avoid a possibility of prompt criticality accident. The basic concept is that central region of the reactor core is filled with fertile (blanket) material. During the reactor operation fissile material accumulates in this central region, which helps to compensate fissile material loss in the peripheral core region and also contributes to negative coolant loss reactivity effect. A concept of high fuel volume fraction in the core is applied to achieve smaller size of a critical reactor. In this paper we consider to add Np-237 to the fuel to enhance non proliferation characteristics of the systems. The effect of Np-237 amount variation is discussed.« less

Latest results from Daya Bay

NASA Astrophysics Data System (ADS)

Vorobel, Vit; Daya Bay Collaboration

2017-07-01

The Daya Bay Reactor Neutrino Experiment was designed to measure θ 13, the smallest mixing angle in the three-neutrino mixing framework, with unprecedented precision. The experiment consists of eight functionally identical detectors placed underground at different baselines from three pairs of nuclear reactors in South China. Since Dec. 2011, the experiment has been running stably for more than 4 years, and has collected the largest reactor anti-neutrino sample to date. Daya Bay is able to greatly improve the precision on θ 13 and to make an independent measurement of the effective mass splitting in the electron antineutrino disappearance channel. Daya Bay can also perform a number of other precise measurements, such as a high-statistics determination of the absolute reactor antineutrino flux and spectrum, as well as a search for sterile neutrino mixing, among others. The most recent results from Daya Bay are discussed in this paper, as well as the current status and future prospects of the experiment.

Electron emission produced by photointeractions in a slab target

NASA Technical Reports Server (NTRS)

Thinger, B. E.; Dayton, J. A., Jr.

1973-01-01

The current density and energy spectrum of escaping electrons generated in a uniform plane slab target which is being irradiated by the gamma flux field of a nuclear reactor are calculated by using experimental gamma energy transfer coefficients, electron range and energy relations, and escape probability computations. The probability of escape and the average path length of escaping electrons are derived for an isotropic distribution of monoenergetic photons. The method of estimating the flux and energy distribution of electrons emerging from the surface is outlined, and a sample calculation is made for a 0.33-cm-thick tungsten target located next to the core of a nuclear reactor. The results are to be used as a guide in electron beam synthesis of reactor experiments.

Reactivity change in a fast-spectrum space power reactor due to a 328-meter-per-second (1075-ft/sec) impact

NASA Technical Reports Server (NTRS)

Peoples, J. A., Jr.; Puthoff, R. L.

1973-01-01

Application of nuclear reactors in space will present operational problems. One such problem is the possibility of an earth impact at velocities in excess of 305 m/sec (1000 ft/sec). This report shows the results of an impact against concrete at 328 m/sec (1075 ft/sec) and examines the deformed core to estimate the range of activity inserted as a result of the impact. The results of this examination are that the deformation of the reactor core within the containment vessel left only an estimated 2.7 percent void in the core and that the reactivity inserted due to this impact deformation could be from 4.0 to 10.25 dollars.

Aerosol reactor production of uniform submicron powders

NASA Technical Reports Server (NTRS)

Flagan, Richard C. (Inventor); Wu, Jin J. (Inventor)

1991-01-01

A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

Aerosol reactor production of uniform submicron powders

DOEpatents

Flagan, Richard C.; Wu, Jin J.

1991-02-19

A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

Phenomena Important in Molten Salt Reactor Simulations

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

Diamond, David J.; Brown, Nicholas R.; Denning, Richard

The U.S. Nuclear Regulatory Commission (NRC) is preparing for the future licensing of advanced reactors that will be very different from current light water reactors. Part of the NRC preparation strategy is to identify the simulation tools that will be used for confirmatory safety analysis of normal operation and abnormal situations in those reactors. This report advances that strategy for reactors that will use molten salts (MSRs). This includes reactors with the fuel within the salt as well as reactors using solid fuel. Although both types are discussed in this report, the emphasis is on those reactors with liquid fuelmore » because of the perception that solid-fuel MSRs will be significantly easier to simulate. These liquid-fuel reactors include thermal and fast neutron spectrum alternatives. The specific designs discussed in the report are a subset of many designs being considered in the U.S. and elsewhere but they are considered the most likely to submit information to the NRC in the near future. The objective herein, is to understand the design of proposed molten salt reactors, how they will operate under normal or transient/accident conditions, and what will be the corresponding modeling needs of simulation tools that consider neutronics, heat transfer, fluid dynamics, and material composition changes in the molten salt. These tools will enable the NRC to eventually carry out confirmatory analyses that examine the validity and accuracy of applicant’s calculations and help determine the margin of safety in plant design.« less

FUEL ASSAY REACTOR

DOEpatents

Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

1962-12-25

A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

Electronics for the STEREO experiment

NASA Astrophysics Data System (ADS)

HÉLAINE, Victor; STEREO Collaboration

2017-09-01

The STEREO experiment, aiming to probe short baseline neutrino oscillations by precisely measuring reactor anti-neutrino spectrum, is currently under installation. It is located at short distance from the compact research reactor core of the Institut Laue-Langevin, Grenoble, France. Dedicated electronics, hosted in a single µTCA crate, were designed for this experiment. In this article, the electronics requirements, architecture and the performances achieved are described. It is shown how intrinsic Pulse Shape Discrimination properties of the liquid scintillator are preserved and how custom adaptable logic is used to improve the muon veto efficiency.

R and D program for French sodium fast reactor: On the description and detection of sodium boiling phenomena during sub-assembly blockages

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

Vanderhaegen, M.; Laboratory of Waves and Acoustic, Institut Langevin, ESPCI ParisTech, 10 rue Vauquelin, 75005 Paris; Paumel, K.

2011-07-01

In support of the French ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) reactor program, which aims to demonstrate the industrial applicability of sodium fast reactors with an increased level of safety demonstration and availability compared to the past French sodium fast reactors, emphasis is placed on reactor instrumentation. It is in this framework that CEA studies continuous core monitoring to detect as early as possible the onset of sodium boiling. Such a detection system is of particular interest due to the rapid progress and the consequences of a Total Instantaneous Blockage (TIB) at a subassembly inlet, where sodium boilingmore » intervenes in an early phase. In this paper, the authors describe all the particularities which intervene during the different boiling stages and explore possibilities for their detection. (authors)« less

THE ARMOUR DUST FUELED REACTOR (ADFR). Quarterly Progress Report No. 1 for the Period February 21, 1958 to May 21, 1958

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

Loewe, W.E.; Krucoff, D.

1958-10-31

Work has begun on the ADFR, a reactor using a new fuel form -- fissionable dust carried in an inent gas. Temperatures in the range 2,000 to 3,000 deg F appear feasible in an all-ceramic system. Experimental study of the fuel form was initiated, and a loop to circulate the fuel dust was constructed. Initial operation is encouraging. Theoretical studies were carried on in the areas of reactor physics, heat transfer, and safety. (auth)

System and method for generating steady state confining current for a toroidal plasma fusion reactor

DOEpatents

Fisch, Nathaniel J.

1981-01-01

A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to establish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated in the plasma.

System and method for generating steady state confining current for a toroidal plasma fusion reactor

DOEpatents

Bers, Abraham

1981-01-01

A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to estalish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated inthe plasma.

Multi-reactor power system configurations for multimegawatt nuclear electric propulsion

NASA Technical Reports Server (NTRS)

George, Jeffrey A.

1991-01-01

A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

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

Block, R.C.; Feiner, F.

This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Stacey, W. M.

2009-09-01

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

Progress in the RAMI analysis of a conceptual LHCD system for DEMO

NASA Astrophysics Data System (ADS)

Mirizzi, F.

2014-02-01

Reliability, Availability, Maintainability and Inspectability (RAMI) concepts and techniques, that acquired great importance during the first manned space missions, have been progressively extended to industrial, scientific and consumer equipments to assure them satisfactory performances and lifetimes. In the design of experimental facilities, like tokamaks, mainly aimed at demonstrating validity and feasibility of scientific theories, RAMI analysis has been often left aside. DEMO, the future prototype fusion reactors, will be instead designed for steadily delivering electrical energy to commercial grids, so that the RAMI aspects will assume an absolute relevance since their initial design phases. A preliminary RAMI analysis of the LHCD system for the conceptual EU DEMO reactor is given in the paper.

Physical particularities of nuclear reactors using heavy moderators of neutrons

NASA Astrophysics Data System (ADS)

Kulikov, G. G.; Shmelev, A. N.

2016-12-01

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using 233U as a fissile nuclide and 232Th and 231Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

Renewable Wood Pulp Paper Reactor with Hierarchical Micro/Nanopores for Continuous-Flow Nanocatalysis.

PubMed

Koga, Hirotaka; Namba, Naoko; Takahashi, Tsukasa; Nogi, Masaya; Nishina, Yuta

2017-06-22

Continuous-flow nanocatalysis based on metal nanoparticle catalyst-anchored flow reactors has recently provided an excellent platform for effective chemical manufacturing. However, there has been limited progress in porous structure design and recycling systems for metal nanoparticle-anchored flow reactors to create more efficient and sustainable catalytic processes. In this study, traditional paper is used for a highly efficient, recyclable, and even renewable flow reactor by tailoring the ultrastructures of wood pulp. The "paper reactor" offers hierarchically interconnected micro- and nanoscale pores, which can act as convective-flow and rapid-diffusion channels, respectively, for efficient access of reactants to metal nanoparticle catalysts. In continuous-flow, aqueous, room-temperature catalytic reduction of 4-nitrophenol to 4-aminophenol, a gold nanoparticle (AuNP)-anchored paper reactor with hierarchical micro/nanopores provided higher reaction efficiency than state-of-the-art AuNP-anchored flow reactors. Inspired by traditional paper materials, successful recycling and renewal of AuNP-anchored paper reactors were also demonstrated while high reaction efficiency was maintained. © 2017 The Authors. Published by Wiley-VCH Verlag GmbH & Co. KGaA.

Preliminary Study on LiF4-ThF4-PuF4 Utilization as Fuel Salt of miniFUJI Molten Salt Reactor

NASA Astrophysics Data System (ADS)

Waris, Abdul; Aji, Indarta K.; Pramuditya, Syeilendra; Widayani; Irwanto, Dwi

2016-08-01

miniFUJI reactor is molten salt reactor (MSR) which is one type of the Generation IV nuclear energy systems. The original miniFUJI reactor design uses LiF-BeF2-ThF4-233UF4 as a fuel salt. In the present study, the use of LiF4-ThF4-PuF4 as fuel salt instead of LiF-BeF2-ThF4-UF4 will be discussed. The neutronics cell calculation has been performed by using PIJ (collision probability method code) routine of SRAC 2006 code, with the nuclear data library is JENDL-4.0. The results reveal that the reactor can attain the criticality condition with the plutonium concentration in the fuel salt is equal to 9.16% or more. The conversion ratio diminishes with the enlarging of plutonium concentration in the fuel. The neutron spectrum of miniFUJI MSR with plutonium fuel becomes harder compared to that of the 233U fuel.

New results from RENO and future RENO-50 project

NASA Astrophysics Data System (ADS)

Kim, S. B.

2017-07-01

RENO (Reactor Experiment for Neutrino Oscillation) has obtained a more precise value of the smallest mixing angle θ_{13} and the first result on neutrino squared-mass difference \\vertΔ m_{ee}^2\\vert from an energy- and baseline-dependent disappearance of reactor electron antineutrinos (overline{ν}_e) using 500 days of data. Based on the ratio of inverse-beta-decay (IBD) prompt spectra measured between two identical far and near detectors, we obtain sin^2 2 θ_{13} = 0.082 ± 0.009({stat.}) ± 0.006({syst.}) and \\vertΔ m_{ee}^2\\vert =[2.62_{-0.23}^{+0.21}({stat.}) _{-0.13} ^{+0.12}({syst.})]× 10^{-3} eV2. An excess of reactor antineutrinos near 5MeV is observed in the measured prompt spectrum with respect to the most commonly used models. The excess is found to be consistent with coming from reactors. A future reactor experiment of RENO-50 is proposed to determine the neutrino mass hierarchy and to make highly precise measurements of θ_{12} , Δ m_{21}^2 , and \\vertΔ m_{ee}^2\\vert.

Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

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

Heeger, Karsten M.

2014-09-13

This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zeromore » $$\\theta_{13}$$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.« less

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2009-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2010-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

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

Brown, Nicholas R.; Powers, Jeffrey J.; Mueller, Don

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy researchmore » and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.« less

(Boiling water reactor (BWR) CORA experiments)

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

Ott, L.J.

To participate in the 1990 CORA Workshop at Kernforschungszentrum Karlsruhe (KfK) GmbH, Karlsruhe, FRG, on October 1--4, and to participate in detailed discussions on October 5 with the KfK CORA Boiling Water Reactor (BWR) experiments. The traveler attended the 1990 CORA Workshop at KfK, FRG. Participation included the presentation of a paper on work performed by the Boiling Water Reactor Core Melt Progression Phenomena Program at Oak Ridge National Laboratory (ORNL) on posttest analyses of CORA BWR experiments. The Statement of Work (November 1989) for the BWR Core Melt Progression Phenomena Program provides for pretest and posttest analyses of themore » BWR CORA experiments performed at KfK. Additionally, it is intended that ORNL personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to discuss such topics as (1) experimental test schedule, (2) BWR test conduct, (3) perceived BWR experimental needs, and (4) KfK operational staff needs with respect to ORNL support. 19 refs.« less

Status and progress of the RERTR program in the year 2002.

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

Travelli, A.; Technology Development

2003-01-01

Following the cancellation of the 2001 International RERTR Meeting, which had been planned to occur in Bali, Indonesia, this paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the years 2001 and 2002, and discusses the main activities planned for the year 2003. The past two years have been characterized by very important achievements of the RERTR program, but these technical achievements have been overshadowed by the terrible events of September 11, 2001. Those events have caused the U.S. Government to reevaluate the importance andmore » urgency of the RERTR program goals. A recommendation made at the highest levels of the government calls for an immediate acceleration of the program activities, with the goal of converting all the world's research reactors to low-enriched fuel at the earliest possible time, and including both Soviet-designed and United States-designed research reactors.« less

Transport of Neutrons and Photons Through Iron and Water Layers

NASA Astrophysics Data System (ADS)

Košťál, Michal; Cvachovec, František; Ošmera, Bohumil; Hansen, Wolfgang; Noack, Klaus

2009-08-01

The neutron and photon spectra were measured after iron and water plates placed at the horizontal channel of the Dresden University reactor AK-2. The measurements have been performed with the multiparameter spectrometer [1] with a stilbene cylindrical crystal, 10 × 10 mm or 45 × 45 mm; the neutron and photon spectra have been measured simultaneously. The calculations were performed with the MCNP code and nuclear data libraries ENDF/B VI.2, ENDF/BVII.0, JENDL 3.3 and JEFF 3.1. The measured channel leakage spectrum was used as the input spectrum for the transport calculation. Photons, the primary photons from the reactor - as well as the ones induced by neutron interaction - were calculated. The comparison of the measurements and calculations through 10 cm of iron and 20 cm thickness of water are presented. Besides that, the attenuation of the radiation mixed field by iron layers from 5 to 30 cm is presented; the measured and calculated data are compared.

Development of Ultra-Fine Multigroup Cross Section Library of the AMPX/SCALE Code Packages

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

Jeon, Byoung Kyu; Sik Yang, Won; Kim, Kang Seog

The Consortium for Advanced Simulation of Light Water Reactors Virtual Environment for Reactor Applications (VERA) neutronic simulator MPACT is being developed by Oak Ridge National Laboratory and the University of Michigan for various reactor applications. The MPACT and simplified MPACT 51- and 252-group cross section libraries have been developed for the MPACT neutron transport calculations by using the AMPX and Standardized Computer Analyses for Licensing Evaluations (SCALE) code packages developed at Oak Ridge National Laboratory. It has been noted that the conventional AMPX/SCALE procedure has limited applications for fast-spectrum systems such as boiling water reactor (BWR) fuels with very highmore » void fractions and fast reactor fuels because of its poor accuracy in unresolved and fast energy regions. This lack of accuracy can introduce additional error sources to MPACT calculations, which is already limited by the Bondarenko approach for resolved resonance self-shielding calculation. To enhance the prediction accuracy of MPACT for fast-spectrum reactor analyses, the accuracy of the AMPX/SCALE code packages should be improved first. The purpose of this study is to identify the major problems of the AMPX/SCALE procedure in generating fast-spectrum cross sections and to devise ways to improve the accuracy. For this, various benchmark problems including a typical pressurized water reactor fuel, BWR fuels with various void fractions, and several fast reactor fuels were analyzed using the AMPX 252-group libraries. Isotopic reaction rates were determined by SCALE multigroup (MG) calculations and compared with continuous energy (CE) Monte Carlo calculation results. This reaction rate analysis revealed three main contributors to the observed differences in reactivity and reaction rates: (1) the limitation of the Bondarenko approach in coarse energy group structure, (2) the normalization issue of probability tables, and (3) neglect of the self-shielding effect of resonance-like cross sections at high energy range such as (n,p) cross section of Cl35. The first error source can be eliminated by an ultra-fine group (UFG) structure in which the broad scattering resonances of intermediate-weight nuclides can be represented accurately by a piecewise constant function. A UFG AMPX library was generated with modified probability tables and tested against various benchmark problems. The reactivity and reaction rates determined with the new UFG AMPX library agreed very well with respect to Monte Carlo Neutral Particle (MCNP) results. To enhance the lattice calculation accuracy without significantly increasing the computational time, performing the UFG lattice calculation in two steps was proposed. In the first step, a UFG slowing-down calculation is performed for the corresponding homogenized composition, and UFG cross sections are collapsed into an intermediate group structure. In the second step, the lattice calculation is performed for the intermediate group level using the condensed group cross sections. A preliminary test showed that the condensed library reproduces the results obtained with the UFG cross section library. This result suggests that the proposed two-step lattice calculation approach is a promising option to enhance the applicability of the AMPX/SCALE system to fast system analysis.« less

On the electronic properties of GaSb irradiated with reactor neutrons and its charge neutrality level

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

Boiko, V. M.; Brudnii, V. N., E-mail: brudnyi@mail.tsu.ru; Ermakov, V. S.

2015-06-15

The electronic properties and the limiting position of the Fermi level in p-GaSb crystals irradiated with full-spectrum reactor neutrons at up to a fluence of 8.6 × 10{sup 18} cm{sup −2} are studied. It is shown that the irradiation of GaSb with reactor neutrons results in an increase in the concentration of free holes to p{sub lim} = (5−6) × 10{sup 18} cm{sup −3} and in pinning of the Fermi level at the limiting position F{sub lim} close to E{sub V} + 0.02 eV at 300 K. The effect of the annealing of radiation defects in the temperature range 100–550°Cmore » is explored.« less

Careful measurement of first hyperpolarizability spectrum by hyper-Rayleigh scattering (HRS)

NASA Astrophysics Data System (ADS)

Zhu, Jing; Lu, Changgui; Cui, Yiping

2008-01-01

The first hyperpolarizability (β) spectrum of an azobenzene derivative around its two-photon resonance region is detected carefully by hyper-Rayleigh scattering. The present work uses a fluorescence spectrometer (Edinburgh instruments, F900) as the detector instead of interference filter and photoelectric multiplier tube (PMT). For each wavelength, HRS emission spectrum accompanied with two-photon fluorescence (TPF) is carefully detected by changing the detection wavelength around half of the incident wavelength. Full width to half maximum (FWHM) of the spectrum is about 0.4nm, which is similar to that of the laser. When the incident wavelength moves into the two-photon resonance region, TPF signal increases quickly and should be eliminated. In order to receive accurate β spectrum, the data detected by the oscillograph should be made some emendations, such as TPF, incident energy, absorption and pulse width. Compared with the β spectrum detected in previous works, the spectrum received in this work presents a clearer profile. The β spectrum exhibits a similar profile as its UV-visible spectrum just with blue-shift of wavelength. It could be explained that the electronic vibration structure in two-photon progress is different from that in one-photon progress, while the broadening mechanism may be similar, considering the resonant two-state model (RTSM).

Summary of NR Program Prometheus Efforts

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

J Ashcroft; C Eshelman

2006-02-08

The Naval Reactors Program led work on the development of a reactor plant system for the Prometheus space reactor program. The work centered on a 200 kWe electric reactor plant with a 15-20 year mission applicable to nuclear electric propulsion (NEP). After a review of all reactor and energy conversion alternatives, a direct gas Brayton reactor plant was selected for further development. The work performed subsequent to this selection included preliminary nuclear reactor and reactor plant design, development of instrumentation and control techniques, modeling reactor plant operational features, development and testing of core and plant material options, and development ofmore » an overall project plan. Prior to restructuring of the program, substantial progress had been made on defining reference plant operating conditions, defining reactor mechanical, thermal and nuclear performance, understanding the capabilities and uncertainties provided by material alternatives, and planning non-nuclear and nuclear system testing. The mission requirements for the envisioned NEP missions cannot be accommodated with existing reactor technologies. Therefore concurrent design, development and testing would be needed to deliver a functional reactor system. Fuel and material performance beyond the current state of the art is needed. There is very little national infrastructure available for fast reactor nuclear testing and associated materials development and testing. Surface mission requirements may be different enough to warrant different reactor design approaches and development of a generic multi-purpose reactor requires substantial sacrifice in performance capability for each mission.« less

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

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

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

2015-07-01

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

Study of the Photon Strength Functions for Gadolinium Isotopes with the DANCE Array

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

Dashdorj, D.; Mitchell, G. E.; Baramsai, B.

2009-03-10

The gadolinium isotopes are interesting for reactor applications as well as for medicine and astrophysics. The gadolinium isotopes have some of the largest neutron capture cross sections. As a consequence they are used in the control rod in reactor fuel assembly. From the basic science point of view, there are seven stable isotopes of gadolinium with varying degrees of deformation. Therefore they provide a good testing ground for the study of deformation dependent structure such as the scissors mode. Decay gamma rays following neutron capture on Gd isotopes are detected by the DANCE array, which is located at flight pathmore » 14 at the Lujan Neutron Scattering Center at Los Alamos National Laboratory. The high segmentation and close packing of the detector array enable gamma-ray multiplicity measurements. The calorimetric properties of the DANCE array coupled with the neutron time-of-flight technique enables one to gate on a specific resonance of a specific isotope in the time-of-flight spectrum and obtain the summed energy spectrum for that isotope. The singles gamma-ray spectrum for each multiplicity can be separated by their DANCE cluster multiplicity. Various photon strength function models are used for comparison with experimentally measured DANCE data and provide insight for understanding the statistical decay properties of deformed nuclei.« less

Light-Water-Reactor safety research program. Quarterly progress report, January--March 1977

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

None

The report summarizes the Argonne National Laboratory work performed during January, February, and March 1977 on water-reactor-safety problems. The following research and development areas are covered: (1) loss-of-coolant accident research: heat transfer and fluid dynamics; (2) transient fuel response and fission-product release program; (3) mechanical properties of zircaloy containing oxygen; and (4) steam-explosion studies.

16TH Annual Review of Progress in Applied Computational Electromagnetics of the Naval Postgraduate School, Monterey, CA. Volume I

DTIC Science & Technology

2000-03-24

Element Method for Designing Plasma Reactors" Leo Kempel, Paul Rummel, Tim Grotjohn and John Amrhein ............................................ 28...34Finite Element Method for Designing Plasma Reactors" Leo Kempel, Paul Rummel, Tim Grotjohn and John Amrhein...548 "lime-Domain Simulation of Electromagnetic Wave Propagation in a Magnetized Plasma" J. Paul , C. Christopoulos, and

Status of liquid metal fast breeder reactor fuel development in Japan

NASA Astrophysics Data System (ADS)

Katsuragawa, M.; Kashihara, H.; Akebi, M.

1993-09-01

The mixed-oxide fuel technology for a liquid metal fast breeder reactor (LMFBR) in Japan is progressing toward commercial deployment of LMFBR. Based on accumulated experience in Joyo and Monju fuel development, efforts for large scale LMFBR fuel development are devoted to improved irradiation performance, reliability and economy. This paper summarizes accomplishments, current activities and future plans for LMFBR fuel development in Japan.

Year One Summary of X-energy Pebble Fuel Development at ORNL

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

Helmreich, Grant W.; Hunn, John D.; McMurray, Jake W.

2017-06-01

The Advanced Reactor Concepts X-energy (ARC-Xe) Pebble Fuel Development project at Oak Ridge National Laboratory (ORNL) has successfully completed its first year, having made excellent progress in accomplishing programmatic objectives. The primary focus of research at ORNL in support of X-energy has been the training of X-energy fuel fabrication engineers and the establishment of US pebble fuel production capabilities able to supply the Xe-100 pebble-bed reactor. These efforts have been strongly supported by particle fuel fabrication and characterization expertise present at ORNL from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program.

Metal Hall sensors for the new generation fusion reactors of DEMO scale

NASA Astrophysics Data System (ADS)

Bolshakova, I.; Bulavin, M.; Kargin, N.; Kost, Ya.; Kuech, T.; Kulikov, S.; Radishevskiy, M.; Shurygin, F.; Strikhanov, M.; Vasil'evskii, I.; Vasyliev, A.

2017-11-01

For the first time, the results of on-line testing of metal Hall sensors based on nano-thickness (50-70) nm gold films, which was conducted under irradiation by high-energy neutrons up to the high fluences of 1 · 1024 n · m-2, are presented. The testing has been carried out in the IBR-2 fast pulsed reactor in the neutron flux with the intensity of 1.5 · 1017 n · m-2 · s-1 at the Joint Institute for Nuclear Research. The energy spectrum of neutron flux was very close to that expected for the ex-vessel sensors locations in the ITER experimental reactor. The magnetic field sensitivity of the gold sensors was stable within the whole fluence range under research. Also, sensitivity values at the start and at the end of irradiation session were equal within the measurement error (<1%). The results obtained make it possible to recommend gold sensors for magnetic diagnostics in the new generation fusion reactors of DEMO scale.

Determining Reactor Fuel Type from Continuous Antineutrino Monitoring

NASA Astrophysics Data System (ADS)

Jaffke, Patrick; Huber, Patrick

2017-09-01

We investigate the ability of an antineutrino detector to determine the fuel type of a reactor. A hypothetical 5-ton antineutrino detector is placed 25 m from the core and measures the spectral shape and rate of antineutrinos emitted by fission fragments in the core for a number of 90-d periods. Our results indicate that four major fuel types can be differentiated from the variation of fission fractions over the irradiation time with a true positive probability of detection at approximately 95%. In addition, we demonstrate that antineutrinos can identify the burnup at which weapons-grade mixed-oxide (MOX) fuel would be reduced to reactor-grade MOX, on average, providing assurance that plutonium-disposition goals are met. We also investigate removal scenarios where plutonium is purposefully diverted from a mixture of MOX and low-enriched uranium fuel. Finally, we discuss how our analysis is impacted by a spectral distortion around 6 MeV observed in the antineutrino spectrum measured from commercial power reactors.

DOPPLER CALCULATIONS FOR LARGE FAST CERAMIC REACTORS--EFFECTS OF IMPROVED METHODS AND RECENT CROSS SECTION INFORMATION

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

Greebler, P.; Goldman, E.

1962-12-19

Doppler calculations for large fast ceramic reactors (FCR), using recent cross section information and improved methods, are described. Cross sections of U/sup 238/, Pu/sup 239/, and Pu/sup 210/ with fuel temperature variations needed for perturbation calculations of Doppler reactivity changes are tabulated as a function of potential scattering cross section per absorber isotope at energies below 400 kev. These may be used in Doppler calculations for anv fast reactor. Results of Doppler calculations on a large fast ceramic reactor are given to show the effects of the improved calculation methods and of recent cross secrion data on the calculated Dopplermore » coefficient. The updated methods and cross sections used yield a somewhat harder spectrum and accordingly a somewhat smaller Doppler coefficient for a given FCR core size and composition than calculated in earlier work, but they support the essential conclusion derived earlier that the Doppler effect provides an important safety advantage in a large FCR. 28 references. (auth)« less

Summary of space nuclear reactor power systems, 1983--1992

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

Buden, D.

1993-08-11

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressedmore » from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.« less

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

Zabelin, A.I.; Shmelev, V.E.

Radiolysis of the coolant proceeds at a higher rate in a boiling water reactor as compared to a water-moderated, water-cooled reactor. The radiolytic gases (hydrogen and oxygen) exiting the reactor together with steam can form a potentially explosive mixture. Special interest attaches to the results obtained under the codnitions of prolonged operation of the VK-50 reactor. Tests of various water-chemistry conditions which were performed in the experimental reactor showed their critical influence on the rate of progress of radiolytic processes. The entire period of operation of the reactor may be arbitrarily divided into three stages, each of which is characterizedmore » by its own peculiar conditions of water chemistry and range of thermal power. From stage to stage, there is a noticeable improvement in the coolant quality which to a limited extent is reflected in the exit of radiolytic gases with the steam. The concentration of radiolytic gases increases with decreased power and with an increased content of corrosion products and other contaminants in the coolant.« less

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

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

Block, R.C.; Feiner, F.

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers dealmore » with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately.« less

Nuclear reactor safety research since three mile island.

PubMed

Mynatt, F R

1982-04-09

The Three Mile Island nuclear power plant accident has resulted in redirection of reactor safety research priorities. The small release to the environment of radioactive iodine-13 to 17 curies in a total radioactivity release of 2.4 million to 13 million curies-has led to a new emphasis on the physical chemistry of fission product behavior in accidents; the fact that the nuclear core was severely damaged but did not melt down has opened a new accident regime-that of the degraded core; the role of the operators in the progression and severity of the accident has shifted emphasis from equipment reliability to human reliability. As research progresses in these areas, the technical base for regulation and risk analysis will change substantially.

Planning and supervision of reactor defueling using discrete event techniques

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

Garcia, H.E.; Imel, G.R.; Houshyar, A.

1995-12-31

New fuel handling and conditioning activities for the defueling of the Experimental Breeder Reactor II are being performed at Argonne National Laboratory. Research is being conducted to investigate the use of discrete event simulation, analysis, and optimization techniques to plan, supervise, and perform these activities in such a way that productivity can be improved. The central idea is to characterize this defueling operation as a collection of interconnected serving cells, and then apply operational research techniques to identify appropriate planning schedules for given scenarios. In addition, a supervisory system is being developed to provide personnel with on-line information on themore » progress of fueling tasks and to suggest courses of action to accommodate changing operational conditions. This paper provides an introduction to the research in progress at ANL. In particular, it briefly describes the fuel handling configuration for reactor defueling at ANL, presenting the flow of material from the reactor grid to the interim storage location, and the expected contributions of this work. As an example of the studies being conducted for planning and supervision of fuel handling activities at ANL, an application of discrete event simulation techniques to evaluate different fuel cask transfer strategies is given at the end of the paper.« less

Top shield temperatures, C and K Reactors

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

Agar, J.D.

1964-12-28

A modification program is now in progress at the C and K Reactors consisting of an extensive renovation of the graphite channels in the vertical safety rod ststems. The present VSR channels are being enlarged by a graphite coring operation and channel sleeves will be installed in the larger channels. One problem associated with the coring operation is the danger of damaging top thermal shield cooling tubes located close to the VSR channels to such an extent that these tubes will have to be removed from service. If such a condition should exist at one or a number of locationsmore » in the top shield of the reactors after reactor startup, the question remains -- what would the resulting temperatures be of the various components of the top shields? This study was initiated to determine temperature distributions in the top shield complex at the C and K Reactors for various top thermal shield coolant system conditions. Since the top thermal shield cooling system at C Reactor is different than those at the K Reactors, the study was conducted separately for the two different systems.« less

Renewable Wood Pulp Paper Reactor with Hierarchical Micro/Nanopores for Continuous‐Flow Nanocatalysis

PubMed Central

Namba, Naoko; Takahashi, Tsukasa; Nogi, Masaya; Nishina, Yuta

2017-01-01

Abstract Continuous‐flow nanocatalysis based on metal nanoparticle catalyst‐anchored flow reactors has recently provided an excellent platform for effective chemical manufacturing. However, there has been limited progress in porous structure design and recycling systems for metal nanoparticle‐anchored flow reactors to create more efficient and sustainable catalytic processes. In this study, traditional paper is used for a highly efficient, recyclable, and even renewable flow reactor by tailoring the ultrastructures of wood pulp. The “paper reactor” offers hierarchically interconnected micro‐ and nanoscale pores, which can act as convective‐flow and rapid‐diffusion channels, respectively, for efficient access of reactants to metal nanoparticle catalysts. In continuous‐flow, aqueous, room‐temperature catalytic reduction of 4‐nitrophenol to 4‐aminophenol, a gold nanoparticle (AuNP)‐anchored paper reactor with hierarchical micro/nanopores provided higher reaction efficiency than state‐of‐the‐art AuNP‐anchored flow reactors. Inspired by traditional paper materials, successful recycling and renewal of AuNP‐anchored paper reactors were also demonstrated while high reaction efficiency was maintained. PMID:28394501

Progress in space nuclear reactor power systems technology development - The SP-100 program

NASA Technical Reports Server (NTRS)

Davis, H. S.

1984-01-01

Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

Uranium plasma radiates in the UV spectrum

NASA Technical Reports Server (NTRS)

Williams, M. D.

1973-01-01

Description of an experiment designed to produce and spectroscopically analyze a simulated gas core reactor plasma in the spectral range from 300 to 1300 A. The plasma was produced by focusing the radiation of a Q-spoiled ruby laser onto the flat surface of a pure U-238 specimen.

Real-time pH monitoring of industrially relevant enzymatic reactions in a microfluidic side-entry reactor (μSER) shows potential for pH control.

PubMed

Gruber, Pia; Marques, Marco P C; Sulzer, Philipp; Wohlgemuth, Roland; Mayr, Torsten; Baganz, Frank; Szita, Nicolas

2017-06-01

Monitoring and control of pH is essential for the control of reaction conditions and reaction progress for any biocatalytic or biotechnological process. Microfluidic enzymatic reactors are increasingly proposed for process development, however typically lack instrumentation, such as pH monitoring. We present a microfluidic side-entry reactor (μSER) and demonstrate for the first time real-time pH monitoring of the progression of an enzymatic reaction in a microfluidic reactor as a first step towards achieving pH control. Two different types of optical pH sensors were integrated at several positions in the reactor channel which enabled pH monitoring between pH 3.5 and pH 8.5, thus a broader range than typically reported. The sensors withstood the thermal bonding temperatures typical of microfluidic device fabrication. Additionally, fluidic inputs along the reaction channel were implemented to adjust the pH of the reaction. Time-course profiles of pH were recorded for a transketolase and a penicillin G acylase catalyzed reaction. Without pH adjustment, the former showed a pH increase of 1 pH unit and the latter a pH decrease of about 2.5 pH units. With pH adjustment, the pH drop of the penicillin G acylase catalyzed reaction was significantly attenuated, the reaction condition kept at a pH suitable for the operation of the enzyme, and the product yield increased. This contribution represents a further step towards fully instrumented and controlled microfluidic reactors for biocatalytic process development. © 2017 The Authors. Biotechnology Journal published by WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Design improvement and performance evaluation of solar photocatalytic reactor for industrial effluent treatment.

PubMed

Nair, Ranjith G; Bharadwaj, P J; Samdarshi, S K

2016-12-01

This work reports the details of the design components and materials used in a linear compound parabolic trough reactor constructed with an aim to use the photocatalyst for solar photocatalytic applications. A compound parabolic trough reactor has been designed and engineered to exploit both UV and visible part of the solar irradiation. The developed compound parabolic trough reactor could receive almost 88% of UV radiation along with a major part of visible radiation. The performance of the reactor has been evaluated in terms of degradation of a probe pollutant using the parameters such as rate constant, residence time and photonic efficiency. An attempt has been made to assess the performance in different ranges of solar spectrum. Finally the developed reactor has been employed for the photocatalytic treatment of a paper mill effluent using Degussa P25 as the photocatalyst. The paper mill effluent collected from Nagaon paper mill, Assam, India has been treated under both batch mode and continuous mode using Degussa P25 photocatalyst under artificial and natural solar radiation, respectively. The photocatalytic degradation kinetics of the paper mill effluent has been determined using the reduction in total organic carbon (TOC) values of the effluent. Copyright © 2015 Elsevier Inc. All rights reserved.

Sterile neutrinos or flux uncertainties? — Status of the reactor anti-neutrino anomaly

NASA Astrophysics Data System (ADS)

Dentler, Mona; Hernández-Cabezudo, Álvaro; Kopp, Joachim; Maltoni, Michele; Schwetz, Thomas

2017-11-01

The ˜ 3 σ discrepancy between the predicted and observed reactor anti-neutrino flux, known as the reactor anti-neutrino anomaly, continues to intrigue. The recent discovery of an unexpected bump in the reactor anti-neutrino spectrum, as well as indications that the flux deficit is different for different fission isotopes seems to disfavour the explanation of the anomaly in terms of sterile neutrino oscillations. We critically review this conclusion in view of all available data on electron (anti)neutrino disappearance. We find that the sterile neutrino hypothesis cannot be rejected based on global data and is only mildly disfavored compared to an individual rescaling of neutrino fluxes from different fission isotopes. The main reason for this is the presence of spectral features in recent data from the NEOS and DANSS experiments. If state-of-the-art predictions for reactor fluxes are taken at face value, sterile neutrino oscillations allow a consistent description of global data with a significance close to 3 σ relative to the no-oscillation case. Even if reactor fluxes and spectra are left free in the fit, a 2 σ hint in favour of sterile neutrinos remains, with allowed parameter regions consistent with an explanation of the anomaly in terms of oscillations.

Reactor on-off antineutrino measurement with KamLAND

NASA Astrophysics Data System (ADS)

Gando, A.; Gando, Y.; Hanakago, H.; Ikeda, H.; Inoue, K.; Ishidoshiro, K.; Ishikawa, H.; Koga, M.; Matsuda, R.; Matsuda, S.; Mitsui, T.; Motoki, D.; Nakamura, K.; Obata, A.; Oki, A.; Oki, Y.; Otani, M.; Shimizu, I.; Shirai, J.; Suzuki, A.; Takemoto, Y.; Tamae, K.; Ueshima, K.; Watanabe, H.; Xu, B. D.; Yamada, S.; Yamauchi, Y.; Yoshida, H.; Kozlov, A.; Yoshida, S.; Piepke, A.; Banks, T. I.; Fujikawa, B. K.; Han, K.; O'Donnell, T.; Berger, B. E.; Learned, J. G.; Matsuno, S.; Sakai, M.; Efremenko, Y.; Karwowski, H. J.; Markoff, D. M.; Tornow, W.; Detwiler, J. A.; Enomoto, S.; Decowski, M. P.

2013-08-01

The recent long-term shutdown of Japanese nuclear reactors has resulted in a significantly reduced reactor ν¯e flux at KamLAND. This running condition provides a unique opportunity to confirm and constrain backgrounds for the reactor ν¯e oscillation analysis. The data set also has improved sensitivity for other ν¯e signals, in particular ν¯e’s produced in β-decays from U238 and Th232 within the Earth’s interior, whose energy spectrum overlaps with that of reactor ν¯e’s. Including constraints on θ13 from accelerator and short-baseline reactor neutrino experiments, a combined three-flavor analysis of solar and KamLAND data gives fit values for the oscillation parameters of tan⁡2θ12=0.436-0.025+0.029, Δm212=7.53-0.18+0.18×10-5eV2, and sin⁡2θ13=0.023-0.002+0.002. Assuming a chondritic Th/U mass ratio, we obtain 116-27+28 ν¯e events from U238 and Th232, corresponding to a geo ν¯e flux of 3.4-0.8+0.8×106cm-2s-1 at the KamLAND location. We evaluate various bulk silicate Earth composition models using the observed geo ν¯e rate.

US Efforts in Support of Examinations at Fukushima Daiichi – 2016 Evaluations

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

Amway, P.; Andrews, N.; Bixby, Willis

Although it is clear that the accident signatures from each unit at the Fukushima Daiichi Nuclear Power Station (NPS) [Daiichi] differ, much is not known about the end-state of core materials within these units. Some of this uncertainty can be attributed to a lack of information related to cooling system operation and cooling water injection. There is also uncertainty in our understanding of phenomena affecting: a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs), and b) accident progression after vessel failure (ex-vessel progression) for BWRs and Pressurized Water Reactors (PWRs). These uncertainties arise due to limitedmore » full scale prototypic data. Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents. Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company Holdings (TEPCO) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document reports recent results from the US Forensics Effort to use information obtained by TEPCO to enhance the safety of existing and future nuclear power plant designs. This Forensics Effort, which is sponsored by the Reactor Safety Technologies Pathway of the Department of Energy Office of Nuclear Energy Light Water Reactor (LWR) Sustainability Program, consists of a group of US experts in LWR safety and plant operations that have identified examination needs and are evaluating TEPCO information from Daiichi that address these needs. Examples presented in this report demonstrate that significant safety insights are being obtained in the areas of component performance, fission product release and transport, debris end-state location, and combustible gas generation and transport. In addition to reducing uncertainties related to severe accident modeling progression, these insights are being used to update guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, reduced uncertainties in modeling the events at Daiichi will improve the realism of reactor safety evaluations and inform future D&D activities by improving the capability for characterizing potential hazards to workers involved with cleanup activities.« less

Analysis of C/E results of fission rate ratio measurements in several fast lead VENUS-F cores

NASA Astrophysics Data System (ADS)

Kochetkov, Anatoly; Krása, Antonín; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente; Bianchini, Giancarlo; Fabrizio, Valentina; Carta, Mario; Firpo, Gabriele; Fridman, Emil; Sarotto, Massimo

2017-09-01

During the GUINEVERE FP6 European project (2006-2011), the zero-power VENUS water-moderated reactor was modified into VENUS-F, a mock-up of a lead cooled fast spectrum system with solid components that can be operated in both critical and subcritical mode. The Fast Reactor Experiments for hybrid Applications (FREYA) FP7 project was launched in 2011 to support the designs of the MYRRHA Accelerator Driven System (ADS) and the ALFRED Lead Fast Reactor (LFR). Three VENUS-F critical core configurations, simulating the complex MYRRHA core design and one configuration devoted to the LFR ALFRED core conditions were investigated in 2015. The MYRRHA related cores simulated step by step design peculiarities like the BeO reflector and in pile sections. For all of these cores the fuel assemblies were of a simple design consisting of 30% enriched metallic uranium, lead rodlets to simulate the coolant and Al2O3 rodlets to simulate the oxide fuel. Fission rate ratios of minor actinides such as Np-237, Am-241 as well as Pu-239, Pu-240, Pu-242 and U-238 to U-235 were measured in these VENUS-F critical assemblies with small fission chambers in specially designed locations, to determine the spectral indices in the different neutron spectrum conditions. The measurements have been analyzed using advanced computational tools including deterministic and stochastic codes and different nuclear data sets like JEFF-3.1, JEFF-3.2, ENDF/B7.1 and JENDL-4.0. The analysis of the C/E discrepancies will help to improve the nuclear data in the specific energy region of fast neutron reactor spectra.

Physical particularities of nuclear reactors using heavy moderators of neutrons

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

Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Shmelev, A. N.

2016-12-15

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using {sup 233}U as a fissile nuclide and {sup 232}Th and {sup 231}Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program packagemore » for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.« less

Use of boron nitride for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements

NASA Astrophysics Data System (ADS)

Radulović, Vladimir; Trkov, Andrej; Jaćimović, Radojko; Gregoire, Gilles; Destouches, Christophe

2016-12-01

A recent experimental irradiation and measurement campaign using containers made from boron nitride (BN) at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor in Ljubljana, Slovenia, has shown the applicability of BN for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements. The first part of the paper focuses on the determination of the transmission function of a BN container through Monte Carlo calculations and experimental measurements. The second part presents the process of tayloring the sensitivity of integral activation measurements to specific needs and a selection of suitable radiative capture reactions for neutron spectrum characterization in the epithermal range. A BN container used in our experiments and its qualitative effect on the neutron spectrum in the irradiation position employed is displayed in the Graphical abstract.

Joule-Heated Molten Regolith Electrolysis Reactor Concepts for Oxygen and Metals Production on the Moon and Mars

NASA Technical Reports Server (NTRS)

Sibille, Laurent; Dominguez, Jesus A.

2012-01-01

The technology of direct electrolysis of molten lunar regolith to produce oxygen and molten metal alloys has progressed greatly in the last few years. The development of long-lasting inert anodes and cathode designs as well as techniques for the removal of molten products from the reactor has been demonstrated. The containment of chemically aggressive oxide and metal melts is very difficult at the operating temperatures ca. 1600 C. Containing the molten oxides in a regolith shell can solve this technical issue and can be achieved by designing a Joule-heated (sometimes called 'self-heating') reactor in which the electrolytic currents generate enough Joule heat to create a molten bath. Solutions obtained by multiphysics modeling allow the identification of the critical dimensions of concept reactors.

Engineering calculations for communications satellite systems planning

NASA Technical Reports Server (NTRS)

Levis, C. A.; Martin, C. H.; Reilly, C. H.; Gonsalvez, D. J.; Yamaura, Y.

1985-01-01

An extended gradient search code for broadcasting satellite service (BSS) spectrum/orbit assignment synthesis is discussed. Progress is also reported on both single-entry and full synthesis computational aids for fixed satellite service (FSS) spectrum/orbit assignment purposes.

Treatment of Copper Contaminated Municipal Wastewater by Using UASB Reactor and Sand-Chemically Carbonized Rubber Wood Sawdust Column

PubMed Central

Biswas, Swarup; Mishra, Umesh

2016-01-01

The performance of a laboratory scale upflow anaerobic sludge blanket (UASB) reactor and its posttreatment unit of sand-chemically carbonized rubber wood sawdust (CCRWSD) column system for the treatment of a metal contaminated municipal wastewater was investigated. Copper ion contaminated municipal wastewater was introduced to a laboratory scale UASB reactor and the effluent from UASB reactor was then followed by treatment with sand-CCRWSD column system. The laboratory scale UASB reactor and column system were observed for a period of 121 days. After the posttreatment column the average removal of monitoring parameters such as copper ion concentration (91.37%), biochemical oxygen demand (BODT) (93.98%), chemical oxygen demand (COD) (95.59%), total suspended solid (TSS) (95.98%), ammonia (80.68%), nitrite (79.71%), nitrate (71.16%), phosphorous (44.77%), total coliform (TC) (99.9%), and fecal coliform (FC) (99.9%) was measured. The characterization of the chemically carbonized rubber wood sawdust was done by scanning electron microscope (SEM), X-ray fluorescence spectrum (XRF), and Fourier transforms infrared spectroscopy (FTIR). Overall the system was found to be an efficient and economical process for the treatment of copper contaminated municipal wastewater. PMID:26904681

Measurements Methods for the analysis of Nuclear Reactors Thermal Hydraulic in Water Scaled Facilities

NASA Astrophysics Data System (ADS)

Spaccapaniccia, C.; Planquart, P.; Buchlin, J. M. AB(; ), AC(; )

2018-01-01

The Belgian nuclear research institute (SCK•CEN) is developing MYRRHA. MYRRHA is a flexible fast spectrum research reactor, conceived as an accelerator driven system (ADS). The configuration of the primary loop is pool-type: the primary coolant and all the primary system components (core and heat exchangers) are contained within the reactor vessel, while the secondary fluid is circulating in the heat exchangers. The primary coolant is Lead Bismuth Eutectic (LBE). The recent nuclear accident of Fukushima in 2011 changed the requirements for the design of new reactors, which should include the possibility to remove the residual decay heat through passive primary and secondary systems, i.e. natural convection (NC). After the reactor shut down, in the unlucky event of propeller failures, the primary and secondary loops should be able to remove the decay heat in passive way (Natural Convection). The present study analyses the flow and the temperature distribution in the upper plenum by applying laser imaging techniques in a laboratory scaled water model. A parametric study is proposed to study stratification mitigation strategies by varying the geometry of the buffer tank simulating the upper plenum.

Treatment of Copper Contaminated Municipal Wastewater by Using UASB Reactor and Sand-Chemically Carbonized Rubber Wood Sawdust Column.

PubMed

Biswas, Swarup; Mishra, Umesh

2016-01-01

The performance of a laboratory scale upflow anaerobic sludge blanket (UASB) reactor and its posttreatment unit of sand-chemically carbonized rubber wood sawdust (CCRWSD) column system for the treatment of a metal contaminated municipal wastewater was investigated. Copper ion contaminated municipal wastewater was introduced to a laboratory scale UASB reactor and the effluent from UASB reactor was then followed by treatment with sand-CCRWSD column system. The laboratory scale UASB reactor and column system were observed for a period of 121 days. After the posttreatment column the average removal of monitoring parameters such as copper ion concentration (91.37%), biochemical oxygen demand (BODT) (93.98%), chemical oxygen demand (COD) (95.59%), total suspended solid (TSS) (95.98%), ammonia (80.68%), nitrite (79.71%), nitrate (71.16%), phosphorous (44.77%), total coliform (TC) (99.9%), and fecal coliform (FC) (99.9%) was measured. The characterization of the chemically carbonized rubber wood sawdust was done by scanning electron microscope (SEM), X-ray fluorescence spectrum (XRF), and Fourier transforms infrared spectroscopy (FTIR). Overall the system was found to be an efficient and economical process for the treatment of copper contaminated municipal wastewater.

Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

NASA Astrophysics Data System (ADS)

Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

2010-06-01

In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

Progress in the development of the neutron flux monitoring system of the French GEN-IV SFR: simulations and experimental validations [ANIMMA--2015-IO-98

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

Jammes, C.; Filliatre, P.; De Izarra, G.

The neutron flux monitoring system of the French GEN-IV sodium-cooled fast reactor will rely on high temperature fission chambers installed in the reactor vessel and capable of operating over a wide-range neutron flux. The definition of such a system is presented and the technological solutions are justified with the use of simulation and experimental results. (authors)

Progress in the Production of JP-8 Based Hydrogen and Advanced Tactical Fuels for Military Applications

DTIC Science & Technology

2011-02-01

of a multi- year program to develop, optimize, and demonstrate the military viability of a technology for on-demand production of high...continuous reactor system used for kinetic rate data experiment 86 52 Schematic of a differential reactor. The catalyst bed is kept small , and...program to develop, optimize, and demonstrate the military viability of a technology for on-demand production of high-pressure hydrogen for fuel

Li/sub 2/O microsphere fabrication. Monthly letter progress report No. 1 for the period ending May 31, 1977

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

Schluderberg, D C

1977-06-01

The purpose of this task is to establish process parameters for the fabrication of lithium oxide (Li/sub 2/O) microspheres having properties which as closely as possible approximate those required for the design characteristics of the University of Wisconsin design for a TOKAMAK-type fusion reactor utilizing the moving bed of Li/sub 2/O microspheres as both reactor coolant and tritium breeder.

Two detector arrays for fast neutrons at LANSCE

NASA Astrophysics Data System (ADS)

Haight, R. C.; Lee, H. Y.; Taddeucci, T. N.; O'Donnell, J. M.; Perdue, B. A.; Fotiades, N.; Devlin, M.; Ullmann, J. L.; Laptev, A.; Bredeweg, T.; Jandel, M.; Nelson, R. O.; Wender, S. A.; White, M. C.; Wu, C. Y.; Kwan, E.; Chyzh, A.; Henderson, R.; Gostic, J.

2012-03-01

The neutron spectrum from neutron-induced fission needs to be known in designing new fast reactors, predicting criticality for safety analyses, and developing techniques for global security application. The experimental data base of fission neutron spectra is very incomplete and most present evaluated libraries are based on the approach of the Los Alamos Model. To validate these models and to provide improved data for applications, a program is underway to measure the fission neutron spectrum for a wide range of incident neutron energies using the spallation source of fast neutrons at the Weapons Neutron Research (WNR) facility at the Los Alamos Neutron Science Center (LANSCE). In a double time-of-flight experiment, fission neutrons are detected by arrays of neutron detectors to increase the solid angle and also to investigate possible angular dependence of the fission neutrons. The challenge is to measure the spectrum from low energies, down to 100 keV or so, to energies over 10 MeV, where the evaporation-like spectrum decreases by 3 orders of magnitude from its peak around 1 MeV. For these measurements, we are developing two arrays of neutron detectors, one based on liquid organic scintillators and the other on 6Li-glass detectors. The range of fission neutrons detected by organic liquid scintillators extends from about 600 keV to well over 10 MeV, with the lower limit being defined by the limit of pulse-shape discrimination. The 6Li-glass detectors have a range from very low energies to about 1 MeV, where their efficiency then becomes small. Various considerations and tests are in progress to understand important contributing factors in designing these two arrays and they include selection and characterization of photomultiplier tubes (PM), the performance of relatively thin (1.8 cm) 6Li-glass scintillators on 12.5 cm diameter PM tubes, use of 17.5 cm diameter liquid scintillators with 12.5 cm PM tubes, measurements of detector efficiencies with tagged neutrons from the WNR/LANSCE neutron beam, and efficiency calibration with 252Cf spontaneous fission neutrons. Design considerations and test results are presented.

NE-213-scintillator-based neutron detection system for diagnostic measurements of energy spectra for neutrons having energies greater than or equal to 0.8 MeV created during plasma operations at the Princeton Tokamak Fusion Test Reactor

NASA Astrophysics Data System (ADS)

Dickens, J. K.; Hill, N. W.; Hou, F. S.; McConnell, J. W.; Spencer, R. R.; Tsang, F. Y.

1985-08-01

A system for making diagnostic measurements of the energy spectra of greater than or equal to 0.8-MeV neutrons produced during plasma operations of the Princeton Tokamak Fusion Test Reactor (TFTR) has been fabricated and tested and is presently in operation in the TFTR Test Cell Basement. The system consists of two separate detectors, each made up of cells containing liquid NE-213 scintillator attached permanently to RCA-8850 photomultiplier tubes. Pulses obtained from each photomultiplier system are amplified and electronically analyzed to identify and separate those pulses due to neutron-induced events in the detector from those due to photon-induced events in the detector. Signals from each detector are routed to two separate Analog-to-Digital Converters, and the resulting digitized information, representing: (1) the raw neutron-spectrum data; and (2) the raw photon-spectrum data, are transmited to the CICADA data-acquisition computer system of the TFTR. Software programs have been installed on the CICADA system to analyze the raw data to provide moderate-resolution recreations of the energy spectrum of the neutron and photon fluences incident on the detector during the operation of the TFTR. A complete description of, as well as the operation of, the hardware and software is given in this report.

Statistical Model Analysis of (n,p) Cross Sections and Average Energy For Fission Neutron Spectrum

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

Odsuren, M.; Khuukhenkhuu, G.

2011-06-28

Investigation of charged particle emission reaction cross sections for fast neutrons is important to both nuclear reactor technology and the understanding of nuclear reaction mechanisms. In particular, the study of (n,p) cross sections is necessary to estimate radiation damage due to hydrogen production, nuclear heating and transmutations in the structural materials of fission and fusion reactors. On the other hand, it is often necessary in practice to evaluate the neutron cross sections of the nuclides for which no experimental data are available.Because of this, we carried out the systematical analysis of known experimental (n,p) and (n,a) cross sections for fastmore » neutrons and observed a systematical regularity in the wide energy interval of 6-20 MeV and for broad mass range of target nuclei. To explain this effect using the compound, pre-equilibrium and direct reaction mechanisms some formulae were deduced. In this paper, in the framework of the statistical model known experimental (n,p) cross sections averaged over the thermal fission neutron spectrum of U-235 are analyzed. It was shown that the experimental data are satisfactorily described by the statistical model. Also, in the case of (n,p) cross sections the effective average neutron energy for fission spectrum of U-235 was found to be around 3 MeV.« less

Cultivation of E. coli in single- and ten-stage tower-loop reactors.

PubMed

Adler, I; Schügerl, K

1983-02-01

E. Coli was cultivated in batch and continuous operations in the presence of an antifoam agent in stirred-tank and in single- and ten-stage airlift tower reactors with an outer loop. The maximum specific growth rate, mu(m), the substrate yield coefficient, Y(x/s), the respiratory quotient, RQ, substrate conversion, U(s), the volumetric mass transfer coefficient, K(L)a, the specific interfacial area, a, and the specific power input, P/V(L), were measured and compared. If a medium is used with a concentration of complex substrates (extracts) 2.5 times higher than that of glucose, a spectrum of C sources is available and cell regulation influences reactor performance. Both mu(m) and Y(X/S), which were evaluated in batch reactors, cannot be used for continuous reactors or, when measured in stirred-tank reactors, cannot be employed for tower-loop reactors: mu(m) is higher in the stirred-tank batch than in the tower-loop batch reactor, mu(m) and Y(x/s) are higher in the continuous reactor than in the batch single-stage tower-loop reactor. The performance of the single-stage is better than that of the ten-stage reactor due to the inefficient trays employed. A reduction of the medium recirculation rate reduces OTR, U(s), Pr, and Y(X/S) and causes cell sedimentation and flocculation. The volumetric mass transfer coefficient is reduced with increasing cultivation time; the Sauter bubble diameter, d(s), remains constant and does not depend on operational conditions. An increase in the medium recirculation rate reduces k(L)a. The specific power input, P/V(L), for the single-stage tower loop is much lower with the same k(L)a value than for a stirred tank. The relationship k(L)a vs. P/V(L) evaluated for model media in stirred tanks, can also be used for cultivations in these reactors.

Does electromagnetic radiation accelerate galactic cosmic rays

NASA Technical Reports Server (NTRS)

Eichler, D.

1977-01-01

The 'reactor' theories of Tsytovich and collaborators (1973) of cosmic-ray acceleration by electromagnetic radiation are examined in the context of galactic cosmic rays. It is shown that any isotropic synchrotron or Compton reactors with reasonable astrophysical parameters can yield particles with a maximum relativistic factor of only about 10,000. If they are to produce particles with higher relativistic factors, the losses due to inverse Compton scattering of the electromagnetic radiation in them outweigh the acceleration, and this violates the assumptions of the theory. This is a critical restriction in the context of galactic cosmic rays, which have a power-law spectrum extending up to a relativistic factor of 1 million.

PRELIMINARY HAZARDS SUMMARY REPORT FOR THE VALLECITOS SUPERHEAT REACTOR

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

Murray, J.L.

1961-02-01

BS>The Vallecitos Superheat Reactor (VSR) is a light-watermoderated, thermal-spectrum reactor, cooled by a combination of moderator boiling and forced convection cooling with saturated steam. The reactor core consists of 32 fuel hurdles containing 5300 lb of UO/sub 2/ enriched in U/sub 235/ to 3.6%. The fuel elements are arranged in individual process tubes that direct the cooling steam flow and separate the steam from the water moderator. The reactor vessel is designed for 1250 psig and operates at 960 to 1000 psig. With the reactor operating at 12.5 Mw(t), the maximum fuel cladding temperature is 1250 deg F and themore » cooling steam is superheated to an average temperature of about 810 deg F at 905 psig. Nu clear operation of the reactor is controlled by 12 control rods, actuated by drives mounted on the bottom of the reactor vessel. The water moderator recirculates inside the reactor vessel and through the core region by natural convection. Inherent safety features of the reactor include the negative core reactivity effects upon heating the UO/sub 2/ fuel (Doppler effect), upon increasing the temperature or void content of the moderator in the operating condition, and upon unflooding the fuel process tubes in the hot condition. Snfety features designed into the reactor and plant systems include a system of sensors and devices to detect petentially unsafe operating conditions and to initiate automatically the appropriate countermeasures, a set of fast and reliable control rods for scramming the reactor if a potentially unsafe condition occurs, a manually-actuated liquid neutron poison system, and an emergency cooling system to provide continued steam flow through the reactor core in the event the reactor becomes isolated from either its normal source of steam supply or discharge. The release of radioactivity to unrestricted areas is maintained within permissible limits by monitoring the radioactivity of wastes and controlling their release. The reactor and many of its auxiliaries are housed within a high-integrity essentially leak-tight containment vessel. (auth)« less

Experimental investigation of dynamo effect in the secondary pumps of the fast breeder reactor Superphenix

NASA Astrophysics Data System (ADS)

Alemany, A.; Marty, Ph.; Plunian, F.; Soto, J.

2000-01-01

The fast breeder reactors (FBR) BN600 (Russia) and Phenix (France) have been the subject of several experimental studies aimed at the observation of dynamo action. Though no dynamo effect has been identified, the possibility was raised for the FBR Superphenix (France) which has an electric power twice that of BN600 and five times larger than Phenix. We present the results of a series of experimental investigations on the secondary pumps of Superphenix. The helical sodium flow inside one pump corresponds to a maximum magnetic Reynolds number (Rm) of 25 in the experimental conditions (low temperature). The magnetic field was recorded in the vicinity of the pumps and no dynamo action has been identified. An estimate of the critical flow rate necessary to reach dynamo action has been found, showing that the pumps are far from producing dynamo action. The magnetic energy spectrum was also recorded and analysed. It is of the form k[minus sign]11/3, suggesting the existence of a large-scale magnetic field. Following Moffatt (1978), this spectrum slope is also justified by a phenomenological approach.

Light-water-reactor safety research program. Quarterly progress report, July--September 1975

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

Not Available

1975-01-01

Progress is summarized in the following research and development areas: (1) loss-of-coolant accident research; heat transfer and fluid dynamics; (2) transient fuel response and fission-product release; and (3) mechanical properties of Zircaloy containing oxygen. Also included is an appendix on Kinetics of Fission Gas and Volatile Fission-product Behavior under Transient Conditions in LWR Fuel.

The Behavior of Matter under Nonequilibrium Conditions: Fundamental Aspects and Applications in Energy-oriented Problems: Progress Report for Period September 1984--November 1987

DOE R&D Accomplishments Database

Prigogine, I.

1987-10-07

This report briefly discusses progress on the following topics: state selection dynamics; polymerization under nonequilibrium conditions; inhomogeneous fluctuations in hydrodynamics and in completely mixed reactors; homoclinic bifurcations and mixed-mode oscillations; intrinsic randomness and spontaneous symmetry breaking in explosive systems; and microscopic means of irreversibility.

Lattice Studies of Hyperon Spectroscopy

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

Richards, David G.

2016-04-01

I describe recent progress at studying the spectrum of hadrons containing the strange quark through lattice QCD calculations. I emphasise in particular the richness of the spectrum revealed by lattice studies, with a spectrum of states at least as rich as that of the quark model. I conclude by prospects for future calculations, including in particular the determination of the decay amplitudes for the excited states.

Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments

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

Miller, Marcelo E.; Sztejnberg, Manuel L.; Gonzalez, Sara J.

2011-12-15

Purpose: A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comision Nacional de Energia Atomica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. Methods: The first calibration of the detector was done with themore » well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Results: Local mixed-field thermal neutron sensitivities and global thermal and mixed-field thermal neutron sensitivities derived from measurements performed at the RA-6 were compared and no significant differences were found. Global RA-6-based thermal neutron sensitivity showed agreement with pure thermal neutron sensitivity measurements performed in the RA-3 spectrum. Additionally, the detector response proved nearly unchanged by differences in neutron spectra from real (RA-6 BNCT beam) and ideal (considered for calibration calculations at RA-3) neutron source descriptions. The results confirm that the special design of the Rh SPND can be considered as having a pure thermal response for neutron spectra with epithermal-to-thermal flux ratios up to 12%. In addition, the linear response of the detector to thermal flux allows the use of a mixed-field thermal neutron sensitivity of 1.95 {+-} 0.05 x 10{sup -21} A n{sup -1}{center_dot}cm{sup 2}{center_dot}s. This sensitivity can be used in spectra with up to 21% epithermal-to-thermal flux ratio without significant error due to epithermal neutron and gamma induced effects. The values of the measured fluxes in clinical applications had discrepancies with calculated results that were in the range of -25% to +30%, which shows the importance of a local on-line independent measurement as part of a treatment planning quality control system. Conclusions: The usefulness of the CNEA Rh SPND for the on-line local measurement of thermal neutron flux on BNCT patients has been demonstrated based on an appropriate neutron spectra calibration and clinical applications.« less

Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments.

PubMed

Miller, Marcelo E; Sztejnberg, Manuel L; González, Sara J; Thorp, Silvia I; Longhino, Juan M; Estryk, Guillermo

2011-12-01

A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comisión Nacional de Energía Atómica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Local mixed-field thermal neutron sensitivities and global thermal and mixed-field thermal neutron sensitivities derived from measurements performed at the RA-6 were compared and no significant differences were found. Global RA-6-based thermal neutron sensitivity showed agreement with pure thermal neutron sensitivity measurements performed in the RA-3 spectrum. Additionally, the detector response proved nearly unchanged by differences in neutron spectra from real (RA-6 BNCT beam) and ideal (considered for calibration calculations at RA-3) neutron source descriptions. The results confirm that the special design of the Rh SPND can be considered as having a pure thermal response for neutron spectra with epithermal-to-thermal flux ratios up to 12%. In addition, the linear response of the detector to thermal flux allows the use of a mixed-field thermal neutron sensitivity of 1.95 ± 0.05 × 10(-21) A n(-1)[middle dot]cm² [middle dot]s. This sensitivity can be used in spectra with up to 21% epithermal-to-thermal flux ratio without significant error due to epithermal neutron and gamma induced effects. The values of the measured fluxes in clinical applications had discrepancies with calculated results that were in the range of -25% to +30%, which shows the importance of a local on-line independent measurement as part of a treatment planning quality control system. The usefulness of the CNEA Rh SPND for the on-line local measurement of thermal neutron flux on BNCT patients has been demonstrated based on an appropriate neutron spectra calibration and clinical applications.

Analysis of the chemiluminescence from electronically excited lead oxide generated in a flow tube reactor

NASA Astrophysics Data System (ADS)

Dorko, E. A.; Glessner, J. W.; Ritchey, C. M.; Rutger, L. L.; Pow, J. J.; Brasure, L. D.; Duray, J. P.; Snyder, S. R.

1986-03-01

The chemiluminescence from electronically excited lead oxide formed during the reaction between lead vapor and either 3Σ O 2 or 1Δ O 2 has been studied. The reactions were accomplished in a flow tube reactor. A microwave discharge was used to generate 1Δ O 2. The vibronic spectrum was analyzed and the band head assignments were used in a linear least-squares calculation to obtain the vibronic molecular constants for the X, a, b, A, B, C, C', D, and E electronic states of lead oxide. Based on these and other molecular constants, Franck-Condon factors were calculated for the transitions to the ground state and also for the A-a and D-a transitions. Evidence was presented to support a kinetic analysis of the mechanism leading to chemiluminescence under the experimental conditions encountered in the flow tube reactor. Mechanisms presented earlier were verified by the present data.

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

Bustraan, M.; Coehoorn, J.; Veenema, J.J.

This report is a collection of separate contributions on some aspects of the work done on STEK up to February 1970. A description is given of STEK together with the philosophy of its design, i.e. integral measurements of fission product cross sections by a sample oscillator technique in fast reactor spectra. The influences fission products may have on fast breeder reactors are briefly demonstrated by an example. A description of the facility and of the sample oscillator and sample exchange mechanism is given. Some preliminary results of measurements of reactor parameters and the neutron spectrum in the first fast zonemore » in STEK are given. For the use of lead as material for the buffer an argumentation is given. The proposed program for the measurements of the integral fission product cross sections is outlined. The procurement of some, highly active, samples of actual fission products is briefly sketched. (auth)« less

Deflection Measurements of a Thermally Simulated Nuclear Core Using a High-Resolution CCD-Camera

NASA Technical Reports Server (NTRS)

Stanojev, B. J.; Houts, M.

2004-01-01

Space fission systems under consideration for near-term missions all use compact. fast-spectrum reactor cores. Reactor dimensional change with increasing temperature, which affects neutron leakage. is the dominant source of reactivity feedback in these systems. Accurately measuring core dimensional changes during realistic non-nuclear testing is therefore necessary in predicting the system nuclear equivalent behavior. This paper discusses one key technique being evaluated for measuring such changes. The proposed technique is to use a Charged Couple Device (CCD) sensor to obtain deformation readings of electrically heated prototypic reactor core geometry. This paper introduces a technique by which a single high spatial resolution CCD camera is used to measure core deformation in Real-Time (RT). Initial system checkout results are presented along with a discussion on how additional cameras could be used to achieve a three- dimensional deformation profile of the core during test.

Design of a Low Power, Fast-Spectrum, Liquid-Metal Cooled Surface Reactor System

NASA Astrophysics Data System (ADS)

Marcille, T. F.; Dixon, D. D.; Fischer, G. A.; Doherty, S. P.; Poston, D. I.; Kapernick, R. J.

2006-01-01

In the current 2005 US budget environment, competition for fiscal resources make funding for comprehensive space reactor development programs difficult to justify and accommodate. Simultaneously, the need to develop these systems to provide planetary and deep space-enabling power systems is increasing. Given that environment, designs intended to satisfy reasonable near-term surface missions, using affordable technology-ready materials and processes warrant serious consideration. An initial lunar application design incorporating a stainless structure, 880 K pumped NaK coolant system and a stainless/UO2 fuel system can be designed, fabricated and tested for a fraction of the cost of recent high-profile reactor programs (JIMO, SP-100). Along with the cost reductions associated with the use of qualified materials and processes, this design offers a low-risk, high-reliability implementation associated with mission specific low temperature, low burnup, five year operating lifetime requirements.

Measurement of the^ 235U(n,n')^235mU Integral Cross Section in a Pulsed Reactor

NASA Astrophysics Data System (ADS)

Vieira, D. J.; Bond, E. M.; Belier, G.; Meot, V.; Becker, J. A.; Macri, R. A.; Authier, N.; Hyneck, D.; Jacquet, X.; Jansen, Y.; Legrendre, J.

2009-10-01

We will present the integral measurement of the neutron inelastic cross section of ^235U leading to the 26-minute, E*=76.5 eV isomer state. Small samples (5-20 microgm) of isotope-enriched ^235U were activated in the central cavity of the CALIBAN pulsed reactor at Valduc where a nearly pure fission neutron spectrum is produced with a typical fluence of 3x10^14 n/cm^2. After 30 minutes the samples were removed from the reactor and counted in an electrostatic-deflecting electron spectrometer that was optimized for the detection of ^235mU conversion electrons. From the decay curve analysis of the data, the 26-minute ^235mU component was extracted. Preliminary results will be given and compared to gamma-cascade calculations assuming complete K-mixing or with no K-mixing.

Vibration and acoustic noise emitted by dry-type air-core reactors under PWM voltage excitation

NASA Astrophysics Data System (ADS)

Li, Jingsong; Wang, Shanming; Hong, Jianfeng; Yang, Zhanlu; Jiang, Shengqian; Xia, Shichong

2018-05-01

According to coupling way between the magnetic field and the structural order, structure mode is discussed by engaging finite element (FE) method and both natural frequency and modal shape for a dry-type air-core reactor (DAR) are obtained in this paper. On the basis of harmonic response analysis, electromagnetic force under PWM (Pulse Width Modulation) voltage excitation is mapped with the structure mesh, the vibration spectrum is gained and the consequences represents that the whole structure vibration predominates in the radial direction, with less axial vibration. Referring to the test standard of reactor noise, the rules of emitted noise of the DAR are measured and analyzed at chosen switching frequency matches the sample resonant frequency and the methods of active vibration and noise reduction are put forward. Finally, the low acoustic noise emission of a prototype DAR is verified by measurement.

Calculation of energetic characteristics of C-14 emitted from Beloyarsk nuclear power plant plume with fast neutron reactor

NASA Astrophysics Data System (ADS)

Kolotkov, Gennady A.; Penin, Sergei

2017-11-01

The paper examines an update of comparative analysis of radionuclides released into the atmosphere from Beloyarsk nuclear power plant with fast-neutron reactor for nine years in a row, from 2008 to 2016. It has been shown that the main radionuclides throw out into the atmosphere from Beloyarsk nuclear power plant are beta-active radionuclides. Based on data releases of the RPA "Typhoon", it has been conclude that radiation situation become worse insignificantly; beside on the new reactor BN-800 was put in operation in 2016. Using Spencer-Fano's equation, it was carried out the summary spectrum of emitted radionuclides. On example of Beloyarsk nuclear power plant, it was considered a question about ability of remote detection of raised radioactivity in the atmospheric radioactive plume. It has been shown that it possible to detect raised radioactivity in the emission plume from Beloyarsk nuclear power plant.

MYRRHA: A multipurpose nuclear research facility

NASA Astrophysics Data System (ADS)

Baeten, P.; Schyns, M.; Fernandez, Rafaël; De Bruyn, Didier; Van den Eynde, Gert

2014-12-01

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a multipurpose research facility currently being developed at SCK•CEN. MYRRHA is based on the ADS (Accelerator Driven System) concept where a proton accelerator, a spallation target and a subcritical reactor are coupled. MYRRHA will demonstrate the ADS full concept by coupling these three components at a reasonable power level to allow operation feedback. As a flexible irradiation facility, the MYRRHA research facility will be able to work in both critical as subcritical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV and fusion reactors, and radioisotope production for medical and industrial applications. MYRRHA will be cooled by lead-bismuth eutectic and will play an important role in the development of the Pb-alloys technology needed for the LFR (Lead Fast Reactor) GEN IV concept. MYRRHA will also contribute to the study of partitioning and transmutation of high-level waste. Transmutation of minor actinides (MA) can be completed in an efficient way in fast neutron spectrum facilities, so both critical reactors and subcritical ADS are potential candidates as dedicated transmutation systems. However critical reactors heavily loaded with fuel containing large amounts of MA pose reactivity control problems, and thus safety problems. A subcritical ADS operates in a flexible and safe manner, even with a core loading containing a high amount of MA leading to a high transmutation rate. In this paper, the most recent developments in the design of the MYRRHA facility are presented.

Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

NASA Astrophysics Data System (ADS)

Nur Krisna, Dwita; Su'ud, Zaki

2017-01-01

Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

Reactor Monitoring with Antineutrinos - A Progress Report

NASA Astrophysics Data System (ADS)

Bernstein, Adam

2012-08-01

The Reactor Safeguards regime is the name given to a set of protocols and technologies used to monitor the consumption and production of fissile materials in nuclear reactors. The Safeguards regime is administered by the International Atomic Energy Agency (IAEA), and is an essential component of the global Treaty on Nuclear Nonproliferation, recently renewed by its 189 remaining signators. (The 190th, North Korea, withdrew from the Treaty in 2003). Beginning in Russia in the 1980s, a number of researchers worldwide have experimentally demonstrated the potential of cubic meter scale antineutrino detectors for non-intrusive real-time monitoring of fissile inventories and power output of reactors. The detectors built so far have operated tens of meters from a reactor core, outside of the containment dome, largely unattended and with remote data acquisition for an entire 1.5 year reactor cycle, and have achieved levels of sensitivity to fissile content of potential interest for the IAEA safeguards regime. In this article, I will describe the unique advantages of antineutrino detectors for cooperative monitoring, consider the prospects and benefits of increasing the range of detectability for small reactors, and provide a partial survey of ongoing global research aimed at improving near-field and far field monitoring and discovery of nuclear reactors.

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

Venkataraman, M.; Natarajan, R.; Raj, Baldev

The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR)more » spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)« less

Improvement of yields and rates during enzymatic hydrolysis of cellulose to glucose. Progress report, March 1, 1979-May 31, 1979

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

Sundstrom, D W; Klei, H E; Coughlin, R W

1979-05-01

The objective of this program is to show that the conversion of cellulose to glucose can be significantly increased by enzymatically removing the inhibitory cellobiose from the reaction system using immobilized ..beta..-glucosidase (..beta..-G). An enzymatic catalyst was prepared and used in a fluidized bed with cellobiose as the substrate, only a 10% loss of activity was observed during a 500 hour period. Cellulose was hydrolyzed in two batch reactors operated side-by-side, with one reactor containing immobilized ..beta..-G and cellulose and the other reactor containing an equal amount of cellulose only. After 30 hours the reactor containing the immobilized ..beta..-G hadmore » 100% more glucose, indicating that the catalytic removal of the cellobiose had a significant effect upon the production of glucose.« less

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

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

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

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

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

NASA Technical Reports Server (NTRS)

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

1976-01-01

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

Key Assets for a Sustainable Low Carbon Energy Future

NASA Astrophysics Data System (ADS)

Carre, Frank

2011-10-01

Since the beginning of the 21st century, concerns of energy security and climate change gave rise to energy policies focused on energy conservation and diversified low-carbon energy sources. Provided lessons of Fukushima accident are evidently accounted for, nuclear energy will probably be confirmed in most of today's nuclear countries as a low carbon energy source needed to limit imports of oil and gas and to meet fast growing energy needs. Future challenges of nuclear energy are then in three directions: i) enhancing safety performance so as to preclude any long term impact of severe accident outside the site of the plant, even in case of hypothetical external events, ii) full use of Uranium and minimization long lived radioactive waste burden for sustainability, and iii) extension to non-electricity energy products for maximizing the share of low carbon energy source in transportation fuels, industrial process heat and district heating. Advanced LWRs (Gen-III) are today's best available technologies and can somewhat advance nuclear energy in these three directions. However, breakthroughs in sustainability call for fast neutron reactors and closed fuel cycles, and non-electric applications prompt a revival of interest in high temperature reactors for exceeding cogeneration performances achievable with LWRs. Both types of Gen-IV nuclear systems by nature call for technology breakthroughs to surpass LWRs capabilities. Current resumption in France of research on sodium cooled fast neutron reactors (SFRs) definitely aims at significant progress in safety and economic competitiveness compared to earlier reactors of this type in order to progress towards a new generation of commercially viable sodium cooled fast reactor. Along with advancing a new generation of sodium cooled fast reactor, research and development on alternative fast reactor types such as gas or lead-alloy cooled systems (GFR & LFR) is strategic to overcome technical difficulties and/or political opposition specific to sodium. In conclusion, research and technology breakthroughs in nuclear power are needed for shaping a sustainable low carbon future. International cooperation is key for sharing costs of research and development of the required novel technologies and cost of first experimental reactors needed to demonstrate enabling technologies. At the same time technology breakthroughs are developed, pre-normative research is required to support codification work and harmonized regulations that will ultimately apply to safety and security features of resulting innovative reactor types and fuel cycles.

Measurement of neutrino mixing angle θ13 and mass difference Δ mee2 from reactor antineutrino disappearance in the RENO experiment

NASA Astrophysics Data System (ADS)

Kim, Soo-Bong

2016-07-01

RENO (Reactor Experiment for Neutrino Oscillation) made a definitive measurement of the smallest neutrino mixing angle θ13 in 2012, based on the disappearance of reactor electron antineutrinos. The experiment has obtained a more precise value of the mixing angle and the first result on neutrino mass difference Δ mee2 from an energy and baseline dependent reactor neutrino disappearance using ∼500 days of data. Based on the ratio of inverse-beta-decay (IBD) prompt spectra measured in two identical far and near detectors, we obtain sin2 ⁡ (2θ13) = 0.082 ± 0.009 (stat .) ± 0.006 (syst .) and | Δ mee2 | = [2.62-0.23+0.21 (stat.)-0.13+0.12 (syst .) ] ×10-3 eV2. An excess of reactor antineutrinos near 5 MeV is observed in the measured prompt spectrum with respect to the most commonly used models. The excess is found to be consistent with coming from reactors. A successful measurement of θ13 is also made in an IBD event sample with a delayed signal of neutron capture on hydrogen. A precise value of θ13 would provide important information on determination of the leptonic CP phase if combined with a result of an accelerator neutrino beam experiment.

Current status of the Double Chooz experiment

NASA Astrophysics Data System (ADS)

Haser, J.; Double Chooz Collaboration

2016-04-01

The Double Chooz reactor antineutrino experiment aims for a precision measurement of the neutrino mixing angle θ13. Located at the Chooz nuclear power plant in France, it observes an energy dependent deficit in the electron antineutrino spectrum, currently with one detector filled with gadolinium-loaded liquid scintillator at a baseline of 1.05 km. The Double Chooz analysis utilizes different approaches to extract θ13: A combined rate and spectral shape fit as well as a background-model-independent analysis based on reactor power variations are performed, giving consistent results. Among the recent reactor-based oscillation experiments with comparable baseline it was the only one to observe reactor shutdown phases, during which all reactors are turned off. These enabled to measure the backgrounds solely, allowing to crosscheck the background models used in the oscillation analysis. At present an improved analysis was put forward with twice as much data statistics collected compared to the last publication. Revised selection criteria and background studies enhance the signal to background ratio while a decrease in the corresponding uncertainties is achieved. Along with an improved energy calibration the overall systematic uncertainty on θ13 is reduced, preparing for a two detector analysis. The new analysis obtains from 467.90 live days with 66.5 GW-ton-years of exposure (reactor power × detector mass × live time) a value of sin2 ⁡ 2θ13 =0.090-0.029+0.032(stat + syst).

Progression In The Concepts Of Cognitive Sense Wireless Networks - An Analysis Report

NASA Astrophysics Data System (ADS)

Ajay, V. P.; Nesasudha, M.

2017-10-01

This paper illustrates the conception of networks, their primary goals (from day one to the present), the changes it had to endure to get to its present form and the developments which are in progress and in store for further standardization. The analysis gives more importance to the specifics of the Cognitive Radio Networks, which makes use of the dynamic spectrum access procedures, framed for better utilization of our available spectrum resources. The main conceptual difficulties and current research trends are also discussed in terms of real time implementation.

An accurate redetermination of the 118Sn binding energy

NASA Astrophysics Data System (ADS)

Borzakov, S. B.; Chrien, R. E.; Faikow-Stanczyk, H.; Grigoriev, Yu. V.; Panteleev, Ts. Ts.; Pospisil, S.; Smotritsky, L. M.; Telezhnikov, S. A.

2002-03-01

The energy of well-known strong γ line from 198Au, the "gold standard", has been modified in the light of new adjustments in the fundamental constants and the value of 411.80176(12) keV was determined, which is 0.29 eV lower than the latest 1999 value. An energy calibration procedure for determining the neutron binding energy, Bn, from complicated (n, γ) spectra has been developed. A mathematically simple minimization function consisting only of terms having as parameters the coefficients of the energy calibration curve (polynomial) is used. A priori information about the relationships among the energies of different peaks on the spectrum is taken into account by a Monte-Carlo simulation. The procedure was used in obtaining Bn for 118Sn. The γ-ray spectrum from thermal neutron radiative capture by 117Sn has been measured on the IBR-2 pulsed reactor. γ-rays were detected by a 72 cm 3 HPGe detector. For a better determination of Bn it was important to determine Bn for 64Cu. This value was obtained from two γ-spectra. One spectrum was measured on the IBR-2 by the same detector. The other spectrum was measured with a pair spectrometer at the Brookhaven High Flux Beam Reactor. From these two spectra, Bn for 64Cu was determined to be equal to 7915.52(8) keV. This result essentially differs from the previous value of 7915.96(11) keV. The mean value of the two most precise results of the Bn for 118Sn, was determined to be 9326.35(9) keV. The Bn for 57Fe was determined to be 7646.08(9) keV.

Characterization of a novel micro-pressure swirl reactor for removal of chemical oxygen demand and total nitrogen from domestic wastewater at low temperature.

PubMed

Ren, Qingkai; Yu, Yang; Zhu, Suiyi; Bian, Dejun; Huo, Mingxin; Zhou, Dandan; Huo, Hongliang

2017-06-01

A novel micro-pressure swirl reactor (MPSR) was designed and applied to treat domestic wastewater at low temperature by acclimating microbial biomass with steadily decreasing temperature from 15 to 3 °C. Chemical oxygen demand (COD) was constantly removed by 85% and maintained below 50 mg L -1 in the effluent during the process. When the air flow was controlled at 0.2 m 3  h -1 , a swirl circulation was formed in the reactor, which created a dissolved oxygen (DO) gradient with a low DO zone in the center and a high DO zone in the periphery for denitrification and nitrification. 81% of total nitrogen was removed by this reactor, in which ammonium was reduced by over 90%. However, denitrification was less effective because of the presence of low levels of oxygen. The progressively decreasing temperature favored acclimation of psychrophilic bacteria in the reactor, which replaced mesophilic bacteria in the process of treatment.

Subtyping of Toddlers with ASD Based on Patterns of Social Attention Deficits

DTIC Science & Technology

2015-10-30

autism spectrum . Initial major tasks of the research project included regulatory review and approval and preparation of the eye tracking experiment...situation, we need to discover how to better classify and categorize patterns of behavior and development within the autism spectrum . Our subgrouping...and clinically relevant subgroups within the autism spectrum . Summary of Results, Progress, and Accomplishments with Discussion: Initial major

Summary of the Advanced Reactor Design Criteria (ARDC) Phase 2 Activities

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

Holbrook, Mark Raymond

This report provides an end-of-year summary reflecting the progress and status of proposed regulatory design criteria for advanced non-LWR designs in accordance with the Level 3 milestone in M3AT-15IN2001017 in work package AT-15IN200101. These criteria have been designated as ARDC, and they provide guidance to future applicants for addressing the GDC that are currently applied specifically to LWR designs. The report provides a summary of Phase 2 activities related to the various tasks associated with ARDC development and the subsequent development of example adaptations of ARDC for Sodium Fast Reactor (SFR) and modular High Temperature Gas-cooled Reactor (HTGR) designs.

Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

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

Not Available

1993-07-01

The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meetingmore » have been cataloged separately.« less

Neutron Physics Division progress report for period ending February 28, 1977

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

Maienschein, F.C.

1977-05-01

Summaries are given of research progress in the following areas: (1) measurements of cross sections and related quantities, (2) cross section evaluations and theory, (3) cross section processing, testing, and sensitivity analysis, (4) integral experiments and their analyses, (5) development of methods for shield and reactor analyses, (6) analyses for specific systems or applications, and (7) information analysis and distribution. (SDF)

Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

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

Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

1976-04-01

Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

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

Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.

2008-06-23

This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been mademore » at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics computer codes and methodologies applied to SSTAR. Another section of the report details the SSTAR safety design approach which is based upon defense-in-depth providing multiple levels of protection against the release of radioactive materials and how the inherent safety features of the lead coolant, nitride fuel, fast neutron spectrum core, pool vessel configuration, natural circulation, and containment meet or exceed the requirements for each level of protection. The report also includes recent results of a systematic analysis by LANL of data on corrosion of candidate cladding and structural material alloys of interest to SSTAR by LBE and Pb coolants; the data were taken from a new database on corrosion by liquid metal coolants created at LANL. The analysis methodology that considers penetration of an oxidation front into the alloy and dissolution of the trailing edge of the oxide into the coolant enables the long-term corrosion rate to be extracted from shorter-term corrosion data thereby enabling an evaluation of alloy performance over long core lifetimes (e.g., 30 years) that has heretofore not been possible. A number of candidate alloy specimens with special treatments or coatings which might enhance corrosion resistance at the temperatures at which SSTAR would operate were analyzed following testing in the DELTA loop at LANL including steels that were treated by laser peening at LLNL; laser peening is an approach that alters the oxide-metal bonds which could potentially improve corrosion resistance. LLNL is also carrying out Multi-Scale Modeling of the Fe-Cr system with the goal of assisting in the development of cladding and structural materials having greater resistance to irradiation.« less

Heavy-section steel irradiation program. Progress report, April 1996--September 1996

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

Corwin, W.R.

1997-09-01

The Heavy-Section Steel Irradiation Program was established to quantitatively assess the effects of neutron irradiation on the material behavior of typical reactor pressure vessel (RPV) steels. During this period, fracture mechanics testing of specimens of the irradiated low upper shelf (LUS) weld were completed and analyses performed. Heat treatment of five RPV plate materials was initiated to examine phosphorus segregation effects on the fracture toughness of the heat affected zone of welds. Initial results show that all five materials exhibited very large prior austenite grain sizes as a consequence of the initial heat treatment. Irradiated and annealed specimens of LUSmore » weld material were tested and analyzed. Four sets of Charpy V-notch (CVN) specimens were aged at various temperatures and tested to examine the reason for overrecovery of upper shelf energy that has been observed. Molecular dynamics cascade simulations were extended to 40 keV and have provided information representative of most of the fast neutron spectrum. Investigations of the correlation between microstructural changes and hardness changes in irradiated model alloys was also completed. Preliminary planning for test specimen machining for the Japan Power Development Reactor was completed. A database of Charpy impact and fracture toughness data for RPV materials that have been tested in the unirradiated and irradiated conditions is being assembled and analyzed. Weld metal appears to have similar CVN and fracture toughness transition temperature shifts, whereas the fracture toughness shifts are greater than CVN shifts for base metals. Draft subcontractor reports on precracked cylindrical tensile specimens were completed, reviewed, and are being revised. Testing on precracked CVN specimens, both quasi-static and dynamic, was evaluated. Additionally, testing of compact specimens was initiated as an experimental comparison of constraint limitations. 16 figs., 2 tabs.« less

Isotopic composition and neutronics of the Okelobondo natural reactor

NASA Astrophysics Data System (ADS)

Palenik, Christopher Samuel

The Oklo-Okelobondo and Bangombe uranium deposits, in Gabon, Africa host Earth's only known natural nuclear fission reactors. These 2 billion year old reactors represent a unique opportunity to study used nuclear fuel over geologic periods of time. The reactors in these deposits have been studied as a means by which to constrain the source term of fission product concentrations produced during reactor operation. The source term depends on the neutronic parameters, which include reactor operation duration, neutron flux and the neutron energy spectrum. Reactor operation has been modeled using a point-source computer simulation (Oak Ridge Isotope Generation and Depletion, ORIGEN, code) for a light water reactor. Model results have been constrained using secondary ionization mass spectroscopy (SIMS) isotopic measurements of the fission products Nd and Te, as well as U in uraninite from samples collected in the Okelobondo reactor zone. Based upon the constraints on the operating conditions, the pre-reactor concentrations of Nd (150 ppm +/- 75 ppm) and Te (<1 ppm) in uraninite were estimated. Related to the burnup measured in Okelobondo samples (0.7 to 13.8 GWd/MTU), the final fission product inventories of Nd (90 to 1200 ppm) and Te (10 to 110 ppm) were calculated. By the same means, the ranges of all other fission products and actinides produced during reactor operation were calculated as a function of burnup. These results provide a source term against which the present elemental and decay abundances at the fission reactor can be compared. Furthermore, they provide new insights into the extent to which a "fossil" nuclear reactor can be characterized on the basis of its isotopic signatures. In addition, results from the study of two other natural systems related to the radionuclide and fission product transport are included. A detailed mineralogical characterization of the uranyl mineralogy at the Bangombe uranium deposit in Gabon, Africa was completed to improve geochemical models of the solubility-limiting phase. A study of the competing effects of radiation damage and annealing in a U-bearing crystal of zircon shows that low temperature annealing in actinide-bearing phases is significant in the annealing of radiation damage.

The Effects of Repeated Low-Dose Sarin Exposure

DTIC Science & Technology

2005-08-01

support the idea that there is a triphasic NT model for onset and progression of seizures and subsequent brain damage upon acute exposure to OP ChE...An ACh microbore column (1x530 mm ID, 10 µm UniJet, BAS #MF-8904) coupled with AChE/choline oxidase immobilized enzyme reactor (BAS # MF-8903) was...peroxidase to the electrode for the reduction of hydrogen peroxide that was generated from the immobilized enzyme reactor . The detector was set at (-)1.0

Status and progress of the RERTR program in the year 2000.

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

Travelli, A.

2000-09-28

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the year 2000 and discusses the main activities planned for the year 2001. The past year was characterized by important accomplishments and events for the RERTR program. Four additional shipments containing 503 spent fuel assemblies from foreign research reactors were accepted by the U.S. Altogether, 3,740 spent fuel assemblies from foreign research reactors have been received by the U.S. under the acceptance policy. Postirradiation examinations of three batches of microplates have continued to reveal excellentmore » irradiation behavior of U-MO dispersion fuels in a variety of compositions and irradiating conditions. h-radiation of two new batches of miniplates of greater sizes is in progress in the ATR to investigate me swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g /cm{sup 3} range. Qualification of the U-MO dispersion fuels is proceeding on schedule. Test fuel elements with 6 gU/cm{sup 3} are being fabricated by BWXT and are scheduled to begin undergoing irradiation in the HFR-Petten in the spring of 2001, with a goal of qualifying this fuel by the end of 2003. U-Mo with 8-9 gU/cm{sup 3} is planned to be qualified by the end of 2005. Joint LEU conversion feasibility studies were completed for HFR-Petten and for SAFARI-1. Significant improvements were made in the design of LEU metal-foil annular targets that would allow efficient production of fission {sup 99}Mo. Irradiations in the RAS-GAS reactor showed that these targets can formed from aluminum tubes, and that the yield and purity of their product from the acidic process were at least as good as those from the HEU Cintichem targets. Progress was made on irradiation testing of LEU UO{sub 2} dispersion fuel and on LEU conversion feasibility studies in the Russian RERTR program. Conversion of the BER-11reactor in Berlin, Germany, was completed and conversion of the La Reins reactor in Santiago, Chile, began. These are exciting times for the program. In the fuel development area, the RERTR program is aggressively pursuing qualification of high-density LEU U-Mo dispersion fuels, with the dual goal of enabling fi.uther conversions and of developing a substitute for LEU silicide fuels that can be more easily disposed of after expiration of the FRR SNF Acceptance Program. The {sup 99}Mo effort has reached the point where it appears feasible for all the {sup 99}Mo producers of the world to agree jointly to a common course of action leading to the elimination of HEU use in their processes. As in the past, the success of the RERTR program will depend on the international friendship and cooperation that has always been its trademark.« less

Simulation numerique de l'effet du reflecteur radial sur les cellules rep en utilisant les codes DRAGON et DONJON

NASA Astrophysics Data System (ADS)

Bejaoui, Najoua

The pressurized water nuclear reactors (PWRs) is the largest fleet of nuclear reactors in operation around the world. Although these reactors have been studied extensively by designers and operators using efficient numerical methods, there are still some calculation weaknesses, given the geometric complexity of the core, still unresolved such as the analysis of the neutron flux's behavior at the core-reflector interface. The standard calculation scheme is a two steps process. In the first step, a detailed calculation at the assembly level with reflective boundary conditions, provides homogenized cross-sections for the assemblies, condensed to a reduced number of groups; this step is called the lattice calculation. The second step uses homogenized properties in each assemblies to calculate reactor properties at the core level. This step is called the full-core calculation or whole-core calculation. This decoupling of the two calculation steps is the origin of methodological bias particularly at the interface core reflector: the periodicity hypothesis used to calculate cross section librairies becomes less pertinent for assemblies that are adjacent to the reflector generally represented by these two models: thus the introduction of equivalent reflector or albedo matrices. The reflector helps to slowdown neutrons leaving the reactor and returning them to the core. This effect leads to two fission peaks in fuel assemblies localised at the core/reflector interface, the fission rate increasing due to the greater proportion of reentrant neutrons. This change in the neutron spectrum arises deep inside the fuel located on the outskirts of the core. To remedy this we simulated a peripheral assembly reflected with TMI-PWR reflector and developed an advanced calculation scheme that takes into account the environment of the peripheral assemblies and generate equivalent neutronic properties for the reflector. This scheme is tested on a core without control mechanisms and charged with fresh fuel. The results of this study showed that explicit representation of reflector and calculation of peripheral assembly with our advanced scheme allow corrections to the energy spectrum at the core interface and increase the peripheral power by up to 12% compared with that of the reference scheme.

Non-Destructive Analysis of Natural Uranium Pellet

NASA Astrophysics Data System (ADS)

Wigley, Samantha; Glennon, Kevin; Kitcher, Evans; Folden, Cody

2017-09-01

As part of ongoing nuclear forensics research, samples of natUO2 have been irradiated in a thermal neutron spectrum at the University of Missouri Research Reactor (MURR) with the goal of simulating a pressurized heavy water reactor. Non-destructive gamma ray analysis has been performed on the samples to assay various nuclides in order to determine the burnup and time since irradiation. The quantity of 137Cs was used to determine the burnup directly, and a maximum likelihood method has been used to estimate both the burnup and the time since irradiation. This poster will discuss the most recent results of these analyses. National Science Foundation (PHY-1659847), Department of Energy (DE-FG02-93ER40773).

Search for neutrino oscillations at the palo verde nuclear reactors

PubMed

Boehm; Busenitz; Cook; Gratta; Henrikson; Kornis; Lawrence; Lee; McKinny; Miller; Novikov; Piepke; Ritchie; Tracy; Vogel; Wang; Wolf

2000-04-24

We report on the initial results from a measurement of the antineutrino flux and spectrum at a distance of about 800 m from the three reactors of the Palo Verde Nuclear Generating Station using a segmented gadolinium-loaded scintillation detector. We find that the antineutrino flux agrees with that predicted in the absence of oscillations excluding at 90% C.L. nu;(e)-nu;(x) oscillations with Deltam(2)>1.12x10(-3) eV(2) for maximal mixing and sin (2)2straight theta>0.21 for large Deltam(2). Our results support the conclusion that the atmospheric neutrino oscillations observed by Super-Kamiokande do not involve nu(e).

Evaluation of B&W UO2/ThO2 VIII experimental core: criticality and thermal disadvantage factor analysis

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

Carlo Parisi; Emanuele Negrenti

2017-02-01

In the framework of the OECD/NEA International Reactor Physics Experiment (IRPHE) Project, an evaluation of core VIII of the Babcock & Wilcox (B&W) Spectral Shift Control Reactor (SSCR) critical experiment program was performed. The SSCR concept, moderated and cooled by a variable mixture of heavy and light water, envisaged changing of the thermal neutron spectrum during the operation to encourage breeding and to sustain the core criticality. Core VIII contained 2188 fuel rods with 93% enriched UO2-ThO2 fuel in a moderator mixture of heavy and light water. The criticality experiment and measurements of the thermal disadvantage factor were evaluated.

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

Performance analysis of fusion nuclear-data benchmark experiments for light to heavy materials in MeV energy region with a neutron spectrum shifter

NASA Astrophysics Data System (ADS)

Murata, Isao; Ohta, Masayuki; Miyamaru, Hiroyuki; Kondo, Keitaro; Yoshida, Shigeo; Iida, Toshiyuki; Ochiai, Kentaro; Konno, Chikara

2011-10-01

Nuclear data are indispensable for development of fusion reactor candidate materials. However, benchmarking of the nuclear data in MeV energy region is not yet adequate. In the present study, benchmark performance in the MeV energy region was investigated theoretically for experiments by using a 14 MeV neutron source. We carried out a systematical analysis for light to heavy materials. As a result, the benchmark performance for the neutron spectrum was confirmed to be acceptable, while for gamma-rays it was not sufficiently accurate. Consequently, a spectrum shifter has to be applied. Beryllium had the best performance as a shifter. Moreover, a preliminary examination of whether it is really acceptable that only the spectrum before the last collision is considered in the benchmark performance analysis. It was pointed out that not only the last collision but also earlier collisions should be considered equally in the benchmark performance analysis.

Progress and challenges of nuclear science development in Vietnam - an outlook on the occassion of the 10-th anniversary of the Dalat Nuclear Research Reactor

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

Hien, P.D.

1994-12-31

Over ten years since the commissioning of the Dalat nuclear research reactor a number of nuclear techniques have been developed and applied in Vietnam Manufacturing of radioisotopes and nuclear instruments, development of isotope tracer and nuclear analytical techniques for environmental studies, exploitation of filtered neutron beams, ... have been major activities of reactor utilizations. Efforts made during ten years of reactor operation have resulted also in establishing and sustaining the applications of nuclear techniques in medicine, industry, agriculture, etc. The successes achieved and lessons teamed over the past ten years are discussed illustrating the approaches taken for developing the nuclearmore » science in the conditions of a country having a very low national income and experiencing a transition from a centrally planned to a market-oriented economic system.« less

SP-100 ground engineering system test site description and progress update

NASA Astrophysics Data System (ADS)

Baxter, William F.; Burchell, Gail P.; Fitzgibbon, Davis G.; Swita, Walter R.

1991-01-01

The SP-100 Ground Engineering System Test Site will provide the facilities for the testing of an SP-100 reactor, which is technically prototypic of the generic design for producing 100 kilowatts of electricity. This effort is part of the program to develop a compact, space-based power system capable of producing several hundred kilowatts of electrical power. The test site is located on the U.S. Department of Energy's Hanford Site near Richland, Washington. The site is minimizing capital equipment costs by utilizing existing facilities and equipment to the maximum extent possible. The test cell is located in a decommissioned reactor containment building, and the secondary sodium cooling loop will use equipment from the Fast Flux Test Facility plant which has never been put into service. Modifications to the facility and special equipment are needed to accommodate the testing of the SP-100 reactor. Definitive design of the Ground Engineering System Test Site facility modifications and systems is in progress. The design of the test facility and the testing equipment will comply with the regulations and specifications of the U.S. Department of Energy and the State of Washington.

Technology development for iron Fischer-Tropsch catalysts. Technical progress report No. 8, July 1, 1992--September 30, 1992

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

Frame, R.R.; Gala, H.B.

1992-12-31

The objectives of this contract are to develop a technology for the production of active and stable iron Fischer-Tropsch catalysts for use in slurry-phase synthesis reactors and to develop a scaleup procedure for large-scale synthesis of such catalysts for process development and long-term testing in slurry bubble-column reactors. With a feed containing hydrogen and carbon monoxide in the molar ratio of 0.5 to 1.0 to the slurry bubble-column reactor, the catalyst performance target is 88% CO + H{sub 2} conversion at a minimum space velocity of 2.4 NL/hr/gFe. The desired sum of methane and ethane selectivities is no more thanmore » 4%, and the conversion loss per week is not to exceed 1%. Contract Tasks are as follows: 1.0--Catalyst development, 1.1--Technology assessment, 1.2--Precipitated catalyst preparation method development, 1.3--Novel catalyst preparation methods investigation, 1.4--Catalyst pretreatment, 1.5--Catalyst characterization, 2.0--Catalyst testing, 3.0--Catalyst aging studies, and 4.0--Preliminary design and cost estimate of a catalyst synthesis facility. This paper reports progress made on Task 1.« less

Absence of a dose-fractionation effect on neoplastic transformation induced by fission-spectrum neutrons in C3H 10T1/2 cells

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

Saran, A.; Pazzaglia, S.; Coppola, M.

1991-06-01

We have investigated the effect of fission-spectrum neutron dose fractionation on neoplastic transformation of exponentially growing C3H 10T1/2 cells. Total doses of 10.8, 27, 54, and 108 cGy were given in single doses or in five equal fractions delivered at 24-h intervals in the biological channel of the RSV-TAPIRO reactor at CRE-Casaccia. Both cell inactivation and neoplastic transformation were more effectively induced by fission neutrons than by 250-kVp X rays. No significant effect on cell survival or neoplastic transformation was observed with split doses compared to single doses of fission-spectrum neutrons. Neutron RBE values relative to X rays determined frommore » data for survival and neoplastic transformation were comparable.« less

CY2013 Annual Report for DOE-ITU INERI 2010-006-E

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

Kennedy, J. Rory; Rondinella, Vincenzo V.

2014-12-01

New concepts for nuclear energy development are considered in both the USA and Europe within the framework of the Generation-IV International Forum (GIF) as well as in various US-DOE programs (e.g. the Fuel Cycle Research and Development - FCRD) and as part of the European Sustainable Nuclear Energy Technology Platform (SNE-TP). Since most new fuel cycle concepts envisage the adoption of a closed nuclear fuel cycle employing fast reactors, the fuel behavior characteristics of the various proposed advanced fuel forms must be effectively investigated using state of the art experimental techniques before implementation. More rapid progress can be achieved ifmore » effective synergy with advanced (multi-scale) modeling efforts can be achieved. The fuel systems to be considered include minor actinide (MA) transmutation fuel types such as advanced MOX, advanced metal alloy, inert matrix fuel (IMF), and other ceramic fuels like nitrides, carbides, etc., for fast neutronic spectrum conditions. Most of the advanced fuel compounds have already been the object of past examination programs, which included irradiations in research reactors. The knowledge derived from previous experience constitutes a significant, albeit incomplete body of data. New or upgraded experimental tools are available today that can extend the scientific and technological knowledge towards achieving the objectives associated with the new generation of nuclear reactors and fuels. The objectives of this project will be three-fold: (1) to extend the available knowledge on properties and irradiation behavior of high burnup and minor actinide bearing advanced fuel systems; (2) to establish a synergy with multi-scale and code development efforts in which experimental data and expertise on the irradiation behavior of nuclear fuels is properly conveyed for the upgrade/development of advanced modeling tools; (3) to promote the effective use of international resources to the characterization of irradiated fuel through exchange of expertise and information among leading experimental facilities. The priorities in this project will be set according to the down selection procedure of U.S. and European development programs.« less

NE-213-scintillator-based neutron detection system for diagnostic measurements of energy spectra for neutrons having energies greater than or equal to 0. 8 MeV created during plasma operations at the Princeton Tokamak Fusion Test Reactor

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

Dickens, J.K.; Hill, N.W.; Hou, F.S.

1985-08-01

A system for making diagnostic measurements of the energy spectra of greater than or equal to 0.8-MeV neutrons produced during plasma operations of the Princeton Tokamak Fusion Test Reactor (TFTR) has been fabricated and tested and is presently in operation in the TFTR Test Cell Basement. The system consists of two separate detectors, each made up of cells containing liquid NE-213 scintillator attached permanently to RCA-8850 photomultiplier tubes. Pulses obtained from each photomultiplier system are amplified and electronically analyzed to identify and separate those pulses due to neutron-induced events in the detector from those due to photon-induced events in themore » detector. Signals from each detector are routed to two separate Analog-to-Digital Converters, and the resulting digitized information, representing: (1) the raw neutron-spectrum data; and (2) the raw photon-spectrum data, are transmited to the CICADA data-acquisition computer system of the TFTR. Software programs have been installed on the CICADA system to analyze the raw data to provide moderate-resolution recreations of the energy spectrum of the neutron and photon fluences incident on the detector during the operation of the TFTR. A complete description of, as well as the operation of, the hardware and software is given in this report.« less

Progress of the RERTR program in 2001.

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

Travelli, A.

2002-03-07

This paper describes the 2001 progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners. Postirradiation examinations of microplates have continued to reveal excellent irradiation behavior of U-Mo dispersion fuels in a variety of compositions and irradiating conditions. Irradiation of two new batches of miniplates of greater sizes was completed in the ATR to investigate the swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g/cm{sup 3} range. Qualificationmore » of the U-Mo dispersion fuels has been delayed by a patent issue involving KAERI. Test fuel elements with uranium density of 6 g/cm{sup 3} are being fabricated by BWXT and are expected to begin undergoing irradiation in the HFR-Petten reactor around March 2003, with a goal of qualifying this fuel by mid-2005. U-Mo fuel with uranium density of 8-9 g/cm{sup 3} is expected to be qualified by mid-2007. Final irradiation tests of LEU {sup 99}Mo targets in the RAS-GAS reactor at BATAN, in Indonesia, had to be postponed because of the 9/11 attacks, but the results collected to date indicate that these targets will soon be ready for commercial production. Excellent cooperation is also in progress with the CNEA in Argentina, MDSN/AECL in Canada, and ANSTO in Australia. Irradiation testing of five WWR-M2 tube-type fuel assemblies fabricated by the NZChK and containing LEU UO{sub 2} dispersion fuel was successfully completed within the Russian RERTR program. A new LEU U-Mo pin-type fuel that could be used to convert most Russian-designed research reactors has been developed by VNIINM and is ready for testing. Four additional shipments containing 822 spent fuel assemblies from foreign research reactors were accepted by the U.S. by September 30, 2001. Altogether, 4,562 spent fuel assemblies from foreign research reactors had been received by that date by the U.S. under the FRR SNF acceptance policy. The RERTR program is aggressively pursuing qualification of high-density LEU U-Mo dispersion fuels, with the dual goal of enabling further conversions and of developing a substitute for LEU silicide fuels that can be more easily disposed of after expiration of the U.S. FRR SNF Acceptance Program. As in the past, the success of the RERTR program will depend on the international friendship and cooperation that has always been its trademark.« less

VVER Reactor Safety in Eastern Europe and Former Soviet Union

NASA Astrophysics Data System (ADS)

Papadopoulou, Demetra

2012-02-01

VVER Soviet-designed reactors that operate in Eastern Europe and former Soviet republics have heightened international concern for years due to major safety deficiencies. The governments of countries with VVER reactors have invested millions of dollars toward improving the safety of their nuclear power plants. Most of these reactors will continue to operate for the foreseeable future since they provide urgently-needed electrical power. Given this situation, this paper assesses the radiological consequences of a major nuclear accident in Eastern Europe. The paper also chronicles the efforts launched by the international nuclear community to improve the safety of the reactors and notes the progress made so far through extensive collaborative efforts in Armenia, Bulgaria, the Czech Republic, Hungary, Kazakhstan, Lithuania, Russia, Slovakia, and Ukraine to reduce the risks of nuclear accidents. Western scientific and technical staff collaborated with these countries to improve the safety of their reactor operations by strengthening the ability of the regulator to perform its oversight function, installing safety equipment and technologies, investing time in safety training, and working diligently to establish an enduring safety culture. Still, continued safety improvement efforts are necessary to ensure safe operating practices and achieve timely phase-out of older plants.

Development of advanced strain diagnostic techniques for reactor environments.

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

Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.

2013-02-01

The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding.more » During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.« less

Multiphysics Modeling for Dimensional Analysis of a Self-Heated Molten Regolith Electrolysis Reactor for Oxygen and Metals Production on the Moon and Mars

NASA Technical Reports Server (NTRS)

Dominguez, Jesus; Sibille, Laurent

2010-01-01

The technology of direct electrolysis of molten lunar regolith to produce oxygen and molten metal alloys has progressed greatly in the last few years. The development of long-lasting inert anodes and cathode designs as well as techniques for the removal of molten products from the reactor has been demonstrated. The containment of chemically aggressive oxide and metal melts is very difficult at the operating temperatures ca. 1600 C. Containing the molten oxides in a regolith shell can solve this technical issue and can be achieved by designing a self-heating reactor in which the electrolytic currents generate enough Joule heat to create a molten bath.

Parents' Educational Expectations for Young Children with Autism Spectrum Disorder

ERIC Educational Resources Information Center

Bush, Hillary H.; Cohen, Shana R.; Eisenhower, Abbey S.; Blacher, Jan

2017-01-01

Among typically developing children, many characteristics have been associated with parents' expectations for their children's adjustment to school and academic progress. Despite the history of increased parental involvement in the education of children with autism spectrum disorder (ASD) relative to parents of children without ASD, there is…

Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

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

S. Bragg-Sitton; J. Bess; J. Werner

2011-09-01

Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al.,more » 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).« less

Browns Ferry Unit-3 cavity neutron spectral analysis. Final report

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

Martin, G.C.

1981-08-01

This report describes neutron dosimetry measurements performed in the Browns Ferry Unit-3 reactor cavity using multiple dosimeter and spectrum unfolding techniques to assess radiation-induced degradation of nuclear plant pressure vessels. Test results and conclusions indicating the feasibility of determining neutron flux spectra and the densities in the pressure vessel cavity region via dosimetric measurements are presented.

Fabrication, Quality Assurance, and Quality Control for PROSPECT Detector Component Production

NASA Astrophysics Data System (ADS)

Gustafson, Ian; Prospect (The Precision Reactor Oscillation; Spectrum Experiment) Collaboration

2017-09-01

The Precision Reactor Oscillation and Spectrum Experiment (PROSPECT) is an electron antineutrino (νe) detector intended to make a precision measurement of the 235U neutrino spectrum and to search for the possible existence of sterile neutrinos with a mass splitting of Δm2 on the order of 1 eV2 . As a short baseline detector, PROSPECT will be located less than 10 meters from the High Flux Isotope Reactor at Oak Ridge National Laboratory. As PROSPECT intends to search for baseline-dependent oscillations, physical segmentation is needed to better measure the interaction position. PROSPECT will therefore be a segmented detector in two dimensions, thereby improving position measurements. PROSPECT will be segmented into 154 (11×14) 1.2-meter long rectangular tubes, using optical separators. Each separator will consist of a carbon fiber core, laminated with optical reflector (to increase light collection) and Teflon (to ensure compatibility with the scintillator). These optical separators will be held in place via strings of 3D printed PLA rods called `pinwheels.' This poster discusses the fabrication and quality assurance (QA) procedures used in the production of both the PROSPECT optical separators and pinwheels. For the PROSPECT collaboration.

Characterization of an LED based photoreactor to degrade 4-chlorophenol in an aqueous medium using coumarin (C-343) sensitized TiO2.

PubMed

Ghosh, Jyoti P; Langford, Cooper H; Achari, Gopal

2008-10-16

A detailed performance evaluation of a simple high intensity LED based photoreactor exploiting a narrow wavelength range of the LED to match the spectrum of a dye in a photocatalysis system is reported. A dye sensitized (coumarin-343, lambda max = 446 nm) TiO 2 photocatalyst was used for the degradation of 4-chlorophenol (4-CP) in an aqueous medium using the 436 nm LED based photoreactor. The LED reactor performed competitively with a conventional multilamp reactor and sunlight in the degradation of 4-CP. Light intensities entering the reaction vessel were measured by conventional ferrioxalate actinometry. The results can be fitted by approximate first order kinetic behavior in this system. Hydroxyl radicals were detected by spin trapping EPR, and effects of OH radical quenchers on kinetics suggest that the reaction is initiated by these radicals or their equivalents. LEDs operating at competitive intensities offer a number of advantages to the photochemist or the environmental engineer via long life, efficient current to light conversion, narrow bandwidth, forward directed output, and direct current power for remote operation. Matching light source spectrum to chromophore is a key.

The RERTR Program : a status report.

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

Travelli, A.

1998-10-19

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners since its inception in 1978. A brief summary of the results that the program had attained by the end of 1997 is followed by a detailed review of the major events, findings, and activities that took place in 1998. The past year was characterized by exceptionally important accomplishments and events for the RERTR program. Four additional shipments of spent fuel from foreign research reactors were accepted by the U.S. Altogether, 2,231 spent fuel assemblies from foreignmore » research reactors have been received by the U.S. under the acceptance policy. Fuel development activities began to yield solid results. Irradiations of the first two batches of microplates were completed. Preliminary postirradiation examinations of these microplates indicate excellent irradiation behavior of some of the fuel materials that were tested. These materials hold the promise of achieving the pro am goal of developing LEU research reactor fuels with uranium density in the 8-9 g /cm{sup 3} range. Progress was made in the Russian RERTR program, which aims to develop and demonstrate the technical means needed to convert Russian-supplied research reactors to LEU fuels. Feasibility studies for converting to LEU fuel four Russian-designed research reactors (IR-8 in Russia, Budapest research reactor in Hungary, MARIA in Poland, and WWR-SM in Uzbekistan) were completed. A new program activity began to study the feasibility of converting three Russian plutonium production reactors to the use of low-enriched U0{sub 2}-Al dispersion fuel, so that they can continue to produce heat and electricity without producing significant amounts of plutonium. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, the transient performance of the core under hypothetical accident conditions. A major milestone was accomplished in the development of a process to produce molybdenum-99 from fission targets utilizing LEU instead of HEU. Targets containing LEU metal foils were irradiated in the RAS-GAS reactor at BATAN, Indonesia, and molybdenum-99 was successfully extracted through the ensuing process. These are exciting times for the program and for all those involved in it, and last year's successes augur well for the future. However, as in the past, the success of the RERTR program will depend on the international friendship and cooperation that have always been its trademark.« less

Anaerobic treatment of sludge from a nitrification-denitrification landfill leachate plant

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

Maranon, E.; Castrillon, L.; Fernandez, Y.

2006-07-01

The viability of anaerobic digestion of sludge from a MSW landfill leachate treatment plant, with COD values ranging between 15,000 and 19,400 mg O{sub 2} dm{sup -3}, in an upflow anaerobic sludge blanket reactor was studied. The reactor employed had a useful capacity of 9 l, operating at mesophilic temperature. Start-up of the reactor was carried out in different steps, beginning with diluted sludge and progressively increasing the amount of sludge fed into the reactor. The study was carried out over a period of 7 months. Different amounts of methanol were added to the feed, ranging between 6.75 and 1more » cm{sup 3} dm{sup -3} of feed in order to favour the growth of methanogenic flora. The achieved biodegradation of the sludge using an upflow anaerobic sludge blanket Reactor was very high for an HRT of 9 days, obtaining decreases in COD of 84-87% by the end of the process. Purging of the digested sludge represented {approx}16% of the volume of the treated sludge.« less

NASA's Kilopower Reactor Development and the Path to Higher Power Missions

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Oleson, Steven R.; Poston, David I.; McClure, Patrick

2017-01-01

The development of NASAs Kilopower fission reactor is taking large strides toward flight development with several successful tests completed during its technology demonstration trials. The Kilopower reactors are designed to provide 1-10 kW of electrical power to a spacecraft which could be used for additional science instruments as well as the ability to power electric propulsion systems. Power rich nuclear missions have been excluded from NASA proposals because of the lack of radioisotope fuel and the absence of a flight qualified fission system. NASA has partnered with the Department of Energy's National Nuclear Security Administration to develop the Kilopower reactor using existing facilities and infrastructure to determine if the design is ready for flight development. The 3-year Kilopower project started in 2015 with a challenging goal of building and testing a full-scale flight prototypic nuclear reactor by the end of 2017. As the date approaches, the engineering team shares information on the progress of the technology as well as the enabling capabilities it provides for science and human exploration.

NASA's Kilopower Reactor Development and the Path to Higher Power Missions

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Oleson, Steven R.; Poston, Dave I.; McClure, Patrick

2017-01-01

The development of NASA's Kilopower fission reactor is taking large strides toward flight development with several successful tests completed during its technology demonstration trials. The Kilopower reactors are designed to provide 1-10 kW of electrical power to a spacecraft which could be used for additional science instruments as well as the ability to power electric propulsion systems. Power rich nuclear missions have been excluded from NASA proposals because of the lack of radioisotope fuel and the absence of a flight qualified fission system. NASA has partnered with the Department of Energy's National Nuclear Security Administration to develop the Kilopower reactor using existing facilities and infrastructure to determine if the design is ready for flight development. The 3-year Kilopower project started in 2015 with a challenging goal of building and testing a full-scale flight prototypic nuclear reactor by the end of 2017. As the date approaches, the engineering team shares information on the progress of the technology as well as the enabling capabilities it provides for science and human exploration.

Status and problems of fusion reactor development.

PubMed

Schumacher, U

2001-03-01

Thermonuclear fusion of deuterium and tritium constitutes an enormous potential for a safe, environmentally compatible and sustainable energy supply. The fuel source is practically inexhaustible. Further, the safety prospects of a fusion reactor are quite favourable due to the inherently self-limiting fusion process, the limited radiologic toxicity and the passive cooling property. Among a small number of approaches, the concept of toroidal magnetic confinement of fusion plasmas has achieved most impressive scientific and technical progress towards energy release by thermonuclear burn of deuterium-tritium fuels. The status of thermonuclear fusion research activity world-wide is reviewed and present solutions to the complicated physical and technological problems are presented. These problems comprise plasma heating, confinement and exhaust of energy and particles, plasma stability, alpha particle heating, fusion reactor materials, reactor safety and environmental compatibility. The results and the high scientific level of this international research activity provide a sound basis for the realisation of the International Thermonuclear Experimental Reactor (ITER), whose goal is to demonstrate the scientific and technological feasibility of a fusion energy source for peaceful purposes.

Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

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

Ragusa, Jean; Vierow, Karen

2011-09-01

The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzedmore » advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.« less

Mechanical property changes induced in structural alloys by neutron irradiations with different helium to displacement ratios*1

NASA Astrophysics Data System (ADS)

Mansur, L. K.; Grossbeck, M. L.

1988-07-01

Effects of helium on mechanical properties of irradiated structural materials are reviewed. In particular, variations in response to the ratio of helium to displacement damage serve as the focus. Ductility in creep and tensile tests is emphasized. A variety of early work has led to the current concentration on helium effects for fusion reactor materials applications. A battery of techniques has been developed by which the helium to displacement ratio can be varied. Our main discussion is devoted to the techniques of spectral tailoring and isotopic alloying currently of interest for mixed-spectrum reactors. Theoretical models of physical mechanisms by which helium interacts with displacement damage have been developed in terms of hardening to dislocation motion and grain boundary cavitation. Austenitic stainless steels, ferritic/martensitic steels and vanadium alloys are considered. In each case, work at low strain rates, where the main problems may lie, at the helium to displacement ratios appropriate to fusion reactor materials is lacking. Recent experimental evidence suggests that both in-reactor and high helium results may differ substantially from post-irradiation or low helium results. It is suggested that work in these areas is especially needed.

Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

NASA Technical Reports Server (NTRS)

Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

1972-01-01

A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (2) Critical experiment of lithium-6 used in LEM and LIM

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

Tsunoda, Hirokazu; Sato, Osamu; Okajima, Shigeaki

2002-07-01

In order to achieve fully automated reactor operation of RAPID-L reactor, innovative reactivity control systems LEM, LIM, and LRM are equipped with lithium-6 as a liquid poison. Because lithium-6 has not been used as a neutron absorbing material of conventional fast reactors, measurements of the reactivity worth of Lithium-6 were performed at the Fast Critical Assembly (FCA) of Japan Atomic Energy Research Institute (JAERI). The FCA core was composed of highly enriched uranium and stainless steel samples so as to simulate the core spectrum of RAPID-L. The samples of 95% enriched lithium-6 were inserted into the core parallel to themore » core axis for the measurement of the reactivity worth at each position. It was found that the measured reactivity worth in the core region well agreed with calculated value by the method for the core designs of RAPID-L. Bias factors for the core design method were obtained by comparing between experimental and calculated results. The factors were used to determine the number of LEM and LIM equipped in the core to achieve fully automated operation of RAPID-L. (authors)« less

Novel spectrophotometric methods for simultaneous determination of Amlodipine, Valsartan and Hydrochlorothiazide in their ternary mixture

NASA Astrophysics Data System (ADS)

Lotfy, Hayam M.; Hegazy, Maha A.; Mowaka, Shereen; Mohamed, Ekram Hany

2015-04-01

This work represents a comparative study of two smart spectrophotometric techniques namely; successive resolution and progressive resolution for the simultaneous determination of ternary mixtures of Amlodipine (AML), Hydrochlorothiazide (HCT) and Valsartan (VAL) without prior separation steps. These techniques consist of several consecutive steps utilizing zero and/or ratio and/or derivative spectra. By applying successive spectrum subtraction coupled with constant multiplication method, the proposed drugs were obtained in their zero order absorption spectra and determined at their maxima 237.6 nm, 270.5 nm and 250 nm for AML, HCT and VAL, respectively; while by applying successive derivative subtraction they were obtained in their first derivative spectra and determined at P230.8-246, P261.4-278.2, P233.7-246.8 for AML, HCT and VAL respectively. While in the progressive resolution, the concentrations of the components were determined progressively from the same zero order absorption spectrum using absorbance subtraction coupled with absorptivity factor methods or from the same ratio spectrum using only one divisor via amplitude modulation method can be used for the determination of ternary mixtures using only one divisor where the concentrations of the components are determined progressively. The proposed methods were checked using laboratory-prepared mixtures and were successfully applied for the analysis of pharmaceutical formulation containing the cited drugs. Moreover comparative study between spectrum addition technique as a novel enrichment technique and a well established one namely spiking technique was adopted for the analysis of pharmaceutical formulations containing low concentration of AML. The methods were validated as per ICH guidelines where accuracy, precision and specificity were found to be within their acceptable limits. The results obtained from the proposed methods were statistically compared with the reported one where no significant difference was observed.

Design of the DEMO Fusion Reactor Following ITER.

PubMed

Garabedian, Paul R; McFadden, Geoffrey B

2009-01-01

Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task.

Design of the DEMO Fusion Reactor Following ITER

PubMed Central

Garabedian, Paul R.; McFadden, Geoffrey B.

2009-01-01

Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task. PMID:27504224

Implementation of k0-INAA standardisation at ITU TRIGA Mark II research reactor, Turkey based on k0-IAEA software

NASA Astrophysics Data System (ADS)

Esen, Ayse Nur; Haciyakupoglu, Sevilay

2016-02-01

The purpose of this study is to test the applicability of k0-INAA method at the Istanbul Technical University TRIGA Mark II research reactor. The neutron spectrum parameters such as epithermal neutron flux distribution parameter (α), thermal to epithermal neutron flux ratio (f) and thermal neutron flux (φth) were determined at the central irradiation channel of the ITU TRIGA Mark II research reactor using bare triple-monitor method. HPGe detector calibrations and calculations were carried out by k0-IAEA software. The α, f and φth values were calculated to be -0.009, 15.4 and 7.92·1012 cm-2 s-1, respectively. NIST SRM 1633b coal fly ash and intercomparison samples consisting of clay and sandy soil samples were used to evaluate the validity of the method. For selected elements, the statistical evaluation of the analysis results was carried out by z-score test. A good agreement between certified/reported and experimental values was obtained.

Recent Results from the Daya Bay Experiment

NASA Astrophysics Data System (ADS)

Yu, Z. Y.; Daya Bay Collaboration

2017-09-01

The Daya Bay reactor neutrino experiment was designed to precisely measure the neutrino oscillation parameter θ 13 via the relative comparison of neutrino rates and spectra at different baselines. Eight identically designed detectors were deployed in two near experimental halls and a far hall. Six 2.9 GWth nuclear power reactors served as intense {\\bar ν _e} sources. Since Dec. 2011, the experiment has been running stably. The latest neutrino oscillation results were based on 1230 days of data. Analysis using a three-flavor oscillation model yielded sin22θ 13 = 0.0841 ± 0.0027(stat.) ± 0.0019(syst.), and effective neutrino mass-squared difference ≤ft| {Δ mee^2} \\right| = ≤ft( {2.50 +/- 0.06≤ft( {stat.} \\right) +/- 0.06≤ft( {syst.} \\right)} \\right) × {10 - 3}e{V^2}. Besides, results from the absolute measurement of reactor {\\bar ν _e} flux and energy spectrum, and a search for a light sterile neutrino are also presented.

Solar spectral conversion for improving the photosynthetic activity in algae reactors.

PubMed

Wondraczek, Lothar; Batentschuk, Miroslaw; Schmidt, Markus A; Borchardt, Rudolf; Scheiner, Simon; Seemann, Benjamin; Schweizer, Peter; Brabec, Christoph J

2013-01-01

Sustainable biomass production is expected to be one of the major supporting pillars for future energy supply, as well as for renewable material provision. Algal beds represent an exciting resource for biomass/biofuel, fine chemicals and CO2 storage. Similar to other solar energy harvesting techniques, the efficiency of algal photosynthesis depends on the spectral overlap between solar irradiation and chloroplast absorption. Here we demonstrate that spectral conversion can be employed to significantly improve biomass growth and oxygen production rate in closed-cycle algae reactors. For this purpose, we adapt a photoluminescent phosphor of the type Ca0.59Sr0.40Eu0.01S, which enables efficient conversion of the green part of the incoming spectrum into red light to better match the Qy peak of chlorophyll b. Integration of a Ca0.59Sr0.40Eu0.01S backlight converter into a flat panel algae reactor filled with Haematococcus pluvialis as a model species results in significantly increased photosynthetic activity and algae reproduction rate.

Americium-241 integral radiative capture cross section in over-moderated neutron spectrum from pile oscillator measurements in the Minerve reactor

NASA Astrophysics Data System (ADS)

Geslot, Benoit; Gruel, Adrien; Ros, Paul; Blaise, Patrick; Leconte, Pierre; Noguere, Gilles; Mathieu, Ludovic; Villamarin, David; Becares, Vicente; Plompen, Arjan; Kopecky, Stefan; Schillebeeckx, Peter

2017-09-01

An experimental program, called AMSTRAMGRAM, was recently conducted in the Minerve low power reactor operated by CEA Cadarache within the frame of the CHANDA initiative (Solving CHAllenges in Nuclear Data). Its aim was to measure the integral capture cross section of 241Am in the thermal domain. Motivation of this work is driven by large differences in this actinide thermal point reported by major nuclear data libraries. The AMSTRAMGRAM experiment, that made use of well characterized EC-JRC americium samples, was based on the oscillation technique commonly implemented in the Minerve reactor. First results are presented and discussed in this article. A preliminary calculation scheme was used to compare measured and calculated results. It is shown that this work confirms a bias previously observed with JEFF-3.1.1 (C/E-1 = -10.5 ± 2%). On the opposite, the experiment is in close agreement with 241Am thermal point reported in JEFF-3.2 (C/E-1 = 0.5 ± 2%).

Needs of Accurate Prompt and Delayed γ-spectrum and Multiplicity for Nuclear Reactor Designs

NASA Astrophysics Data System (ADS)

Rimpault, G.; Bernard, D.; Blanchet, D.; Vaglio-Gaudard, C.; Ravaux, S.; Santamarina, A.

The local energy photon deposit must be accounted accurately for Gen-IV fast reactors, advanced light-water nuclear reactors (Gen-III+) and the new experimental Jules Horowitz Reactor (JHR). The γ energy accounts for about 10% of the total energy released in the core of a thermal or fast reactor. The γ-energy release is much greater in the core of the reactor than in its structural sub-assemblies (such as reflector, control rod followers, dummy sub-assemblies). However, because of the propagation of γ from the core regions to the neighboring fuel-free assemblies, the contribution of γ energy to the total heating can be dominant. For reasons related to their performance, power reactors require a 7.5% (1σ) uncertainty for the energy deposition in non-fuelled zones. For the JHR material-testing reactor, a 5% (1 s) uncertainty is required in experimental positions. In order to verify the adequacy of the calculation of γ-heating, TLD and γ-fission chambers were used to derive the experimental heating values. Experimental programs were and are still conducted in different Cadarache facilities such as MASURCA (for SFR), MINERVE and EOLE (for JHR and Gen-III+ reactors). The comparison of calculated and measured γ-heating values shows an underestimation in all experimental programs indicating that for the most γ-production data from 239Pu in current nuclear-data libraries is highly suspicious.The first evaluation priority is for prompt γ-multiplicity for U and Pu fission but similar values for otheractinides such as Pu and U are also required. The nuclear data library JEFF3.1.1 contains most of the photon production data. However, there are some nuclei for which there are missing or erroneous data which need to be completed or modified. A review of the data available shows a lack of measurements for conducting serious evaluation efforts. New measurements are needed to guide new evaluation efforts which benefit from consolidated modeling techniques.

Microstructural evolution of pure tungsten neutron irradiated with a mixed energy spectrum

NASA Astrophysics Data System (ADS)

Koyanagi, Takaaki; Kumar, N. A. P. Kiran; Hwang, Taehyun; Garrison, Lauren M.; Hu, Xunxiang; Snead, Lance L.; Katoh, Yutai

2017-07-01

Microstructures of single-crystal bulk tungsten (W) and polycrystalline W foil with a strong grain texture were investigated using transmission electron microscopy following neutron irradiation at ∼90-800 °C to 0.03-4.6 displacements per atom (dpa) in the High Flux Isotope Reactor with a mixed energy spectrum. The dominant irradiation defects were dislocation loops and small clusters at ∼90 °C. Additional voids were formed in W irradiated at above 460 °C. Voids and precipitates involving transmutation rhenium and osmium were the dominant defects at more than ∼1 dpa. We found a new phenomenon of microstructural evolution in irradiated polycrystalline W: Re- and Os-rich precipitation along grain boundaries. Comparison of results between this study and previous studies using different irradiation facilities revealed that the microstructural evolution of pure W is highly dependent on the neutron energy spectrum in addition to the irradiation temperature and dose.

Frequency spectrum of tantalum at temperatures of 293-2300 K

NASA Astrophysics Data System (ADS)

Semenov, V. A.; Kozlov, Zh. A.; Krachun, L.; Mateescu, G.; Morozov, V. M.; Oprea, A. I.; Oprea, K.; Puchkov, A. V.

2010-05-01

The temperature dependence of the frequency spectrum of tantalum in the temperature range from room temperature to 2300 K has been studied for the first time using inelastic slow-neutron scattering. The inelastic slow-neutron scattering spectra have been measured at different temperatures on a DIN-2PI time-of-flight spectrometer installed at the IBR-2 nuclear reactor (Joint Institute for Nuclear Research, Dubna, Russia) with the use of a TS3000K high-temperature thermostat. From the measured spectra, the frequency spectra of the tantalum crystal lattice have been determined at temperatures of 293, 1584, and 2300 K by the iteration method. As the temperature increases, the frequency spectrum, on the whole, is softened and the specific features manifested themselves at room temperature are smoothed. The variations observed have been explained by the increase in the role of the effects of vibration anharmonism at high temperatures.

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

Rodriguez, Salvador B.

SNL has a combination of experimental facilities, nuclear engineering, nuclear security, severe nuclear accidents, and nuclear safeguards expertise that can enable significant progress towards molten salts and fuels for Molten Salt Reactors (MSRs). The following areas and opportunities are discussed in more detail in this white paper.

Impact of Nuclear Data Uncertainties on Advanced Fuel Cycles and their Irradiated Fuel - a Comparison between Libraries

NASA Astrophysics Data System (ADS)

Díez, C. J.; Cabellos, O.; Martínez, J. S.

2014-04-01

The uncertainties on the isotopic composition throughout the burnup due to the nuclear data uncertainties are analysed. The different sources of uncertainties: decay data, fission yield and cross sections; are propagated individually, and their effect assessed. Two applications are studied: EFIT (an ADS-like reactor) and ESFR (Sodium Fast Reactor). The impact of the uncertainties on cross sections provided by the EAF-2010, SCALE6.1 and COMMARA-2.0 libraries are compared. These Uncertainty Quantification (UQ) studies have been carried out with a Monte Carlo sampling approach implemented in the depletion/activation code ACAB. Such implementation has been improved to overcome depletion/activation problems with variations of the neutron spectrum.

Advanced propulsion engine assessment based on a cermet reactor

NASA Technical Reports Server (NTRS)

Parsley, Randy C.

1993-01-01

A preferred Pratt & Whitney conceptual Nuclear Thermal Rocket Engine (NTRE) has been designed based on the fundamental NASA priorities of safety, reliability, cost, and performance. The basic philosophy underlying the design of the XNR2000 is the utilization of the most reliable form of ultrahigh temperature nuclear fuel and development of a core configuration which is optimized for uniform power distribution, operational flexibility, power maneuverability, weight, and robustness. The P&W NTRE system employs a fast spectrum, cermet fueled reactor configured in an expander cycle to ensure maximum operational safety. The cermet fuel form provides retention of fuel and fission products as well as high strength. A high level of confidence is provided by benchmark analysis and independent evaluations.

Current drive for stability of thermonuclear plasma reactor

NASA Astrophysics Data System (ADS)

Amicucci, L.; Cardinali, A.; Castaldo, C.; Cesario, R.; Galli, A.; Panaccione, L.; Paoletti, F.; Schettini, G.; Spigler, R.; Tuccillo, A.

2016-01-01

To produce in a thermonuclear fusion reactor based on the tokamak concept a sufficiently high fusion gain together stability necessary for operations represent a major challenge, which depends on the capability of driving non-inductive current in the hydrogen plasma. This request should be satisfied by radio-frequency (RF) power suitable for producing the lower hybrid current drive (LHCD) effect, recently demonstrated successfully occurring also at reactor-graded high plasma densities. An LHCD-based tool should be in principle capable of tailoring the plasma current density in the outer radial half of plasma column, where other methods are much less effective, in order to ensure operations in the presence of unpredictably changes of the plasma pressure profiles. In the presence of too high electron temperatures even at the periphery of the plasma column, as envisaged in DEMO reactor, the penetration of the coupled RF power into the plasma core was believed for long time problematic and, only recently, numerical modelling results based on standard plasma wave theory, have shown that this problem should be solved by using suitable parameter of the antenna power spectrum. We show here further information on the new understanding of the RF power deposition profile dependence on antenna parameters, which supports the conclusion that current can be actively driven over a broad layer of the outer radial half of plasma column, thus enabling current profile control necessary for the stability of a reactor.

A microprocessor tester for the treat upgrade reactor trip system

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

Lenkszus, F.R.; Bucher, R.G.

1985-02-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. To improve the analytical extrapolation of test results to full-size assembly bundles, the facility upgrade will increase the maximum size of the test bundle from 7 to 37 fuel pins. By creating a core convertor zone around the test location, the neutron spectrum incident on the test assembly will be hardened and the maximum energy deposited in the sample will be increased. In addition, a programmable Automated Reactor Control System (ARCS) willmore » permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations. A quantitative reliability analysis of the RTS shows that the unreliability, that is, the probability of failure, is acceptable for a 10 hour mission time or risk interval.« less

2010 Summary of Advances in Autism Spectrum Disorder Research

ERIC Educational Resources Information Center

Interagency Autism Coordinating Committee, 2010

2010-01-01

As part of the Combating Autism Act of 2006, the members of the Interagency Autism Coordinating Committee (IACC) are required to develop an annual "Summary of Advances" to describe each year's top advances in autism spectrum disorder (ASD) research. These advances represent significant progress in the early diagnosis of ASD, understanding the…

Language Acquisition in Autism Spectrum Disorders: A Developmental Review

ERIC Educational Resources Information Center

Eigsti, Inge-Marie; de Marchena, Ashley B.; Schuh, Jillian M.; Kelley, Elizabeth

2011-01-01

This paper reviews the complex literature on language acquisition in the autism spectrum disorders (ASD). Because of the high degree of interest in ASD in the past decade, the field has been changing rapidly, with progress in both basic science and applied clinical areas. In addition, psycholinguistically-trained researchers have increasingly…

Measurement of the prompt fissionγ-ray spectrum of 242Pu

NASA Astrophysics Data System (ADS)

Urlass, Sebastian; Beyer, Roland; Junghans, Arnd Rudolf; Kögler, Toni; Schwengner, Ronald; Wagner, Andreas

2018-03-01

The prompt γ-ray spectrum of fission fragments is important in understanding the dynamics of the fission process, as well as for nuclear engineering in terms of predicting the γ-ray heating in nuclear reactors. The γ-ray spectrum measured from the fission fragments of the spontaneous fission of 242Pu will be presented here. A fission chamber containing in total 37mg of 242Pu was used as active sample. The γ-quanta were detected with high time- and energy-resolution using LaBr3 and HPGe detectors, respectively, in coincidence with spontaneous fission events detected by the fission chamber. The acquired γ-ray spectra were corrected for the detector response using the spectrum stripping method. About 70 million fission events were detected which results in a very low statistical uncertainty and a wider energy range covered compared to previous measurements. The prompt fission γ-ray spectrum measured with the HPGe detectors shows structures that allow conclusions about the nature of γ-ray transitions in the fission fragments. The average photon multiplicity of 8.2 and the average total energy release by prompt photons per fission event of about 6.8 MeV were determined for both detector types.

Isotopic Transmutations in Irradiated Beryllium and Their Implications on MARIA Reactor Operation

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

Andrzejewski, Krzysztof J.; Kulikowska, Teresa A

2004-04-15

Beryllium irradiated by neutrons with energies above 0.7 MeV undergoes (n,{alpha}) and (n,2n) reactions. The Be(n,{alpha}) reaction results in subsequent buildup of {sup 6}Li and {sup 3}He isotopes with large thermal neutron absorption cross sections causing poisoning of irradiated beryllium. The amount of the poison isotopes depends on the neutron flux level and spectrum. The high-flux MARIA reactor operated in Poland since 1975 consists of a beryllium matrix with fuel channels in cutouts of beryllium blocks. As the experimental determination of {sup 6}Li, {sup 3}H, and {sup 3}He content in the operational reactor is impossible, a systematic computational study ofmore » the effect of {sup 3}He and {sup 6}Li presence in beryllium blocks on MARIA reactor reactivity and power density distribution has been undertaken. The analysis of equations governing the transmutation has been done for neutron flux parameters typical for MARIA beryllium blocks. Study of the mutual influence of reactor operational parameters and the buildup of {sup 6}Li, {sup 3}H, and {sup 3}He in beryllium blocks has shown the necessity of a detailed spatial solution of transmutation equations in the reactor, taking into account the whole history of its operation. Therefore, fuel management calculations using the REBUS code with included chains for Be(n,{alpha})-initiated reactions have been done for the whole reactor lifetime. The calculated poisoning of beryllium blocks has been verified against the critical experiment of 1993. Finally, the current {sup 6}Li, {sup 3}H, and {sup 3}He contents, averaged for each beryllium block, have been calculated. The reactivity drop caused by this poisoning is {approx}7%.« less

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

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

Sun, Kaichao; Hu, Lin-wen; Newton, Thomas

2017-05-01

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of nonproliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel basedmore » on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem ¼ 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines.« less

Safety and Regulatory Issues of the Thorium Fuel Cycle

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

Ade, Brian; Worrall, Andrew; Powers, Jeffrey

2014-02-01

Thorium has been widely considered an alternative to uranium fuel because of its relatively large natural abundance and its ability to breed fissile fuel (233U) from natural thorium (232Th). Possible scenarios for using thorium in the nuclear fuel cycle include use in different nuclear reactor types (light water, high temperature gas cooled, fast spectrum sodium, molten salt, etc.), advanced accelerator-driven systems, or even fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on concepts that mix thorium with uranium (UO2 + ThO2),more » add fertile thorium (ThO2) fuel pins to LWR fuel assemblies, or use mixed plutonium and thorium (PuO2 + ThO2) fuel assemblies. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts on the nuclear fuel. Thorium and its irradiation products have nuclear characteristics that are different from those of uranium. In addition, ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These aspects are key to reactor safety-related issues. The primary objectives of this report are to summarize historical, current, and proposed uses of thorium in nuclear reactors; provide some important properties of thorium fuel; perform qualitative and quantitative evaluations of both in-reactor and out-of-reactor safety issues and requirements specific to a thorium-based fuel cycle for current LWR reactor designs; and identify key knowledge gaps and technical issues that need to be addressed for the licensing of thorium LWR fuel in the United States.« less

Impact of thermal spectrum small modular reactors on performance of once-through nuclear fuel cycles with low-enriched uranium

DOE PAGES

Brown, Nicholas R.; Worrall, Andrew; Todosow, Michael

2016-11-18

Small modular reactors (SMRs) offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary example reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies ismore » a cartridge type one-batch core with slightly less than 5.0% enrichment. Challenges associated with SMRs include increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burnup in the reactor and the fuel cycle performance. This paper summarizes a list of the factors relevant to SMR fuel, core, and operation that will impact fuel cycle performance. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for a hypothetical example SMR are compared with those for a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of, mass of depleted uranium disposed of, land use per energy generated, and carbon emissions per energy generated. Finally, it is noted that the features of some SMR designs impact three main aspects of fuel cycle performance: (1) small cores which means high leakage (there is a radial and axial component), (2) no boron which means heterogeneous core and extensive use of control rods and BPs, and (3) single batch cores. But not all of the SMR designs have all of these traits. As a result, the approach used in this study is therefore a bounding case and not all SMRs may be affected to the same extent.« less

Study of the Open Loop and Closed Loop Oscillator Techniques

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

Imel, George R.; Baker, Benjamin; Riley, Tony

This report presents the progress and completion of a five-year study undertaken at Idaho State University of the measurement of very small worth reactivity samples comparing open and closed loop oscillator techniques.The study conclusively demonstrated the equivalency of the two techniques with regard to uncertainties in reactivity values, i.e., limited by reactor noise. As those results are thoroughly documented in recent publications, in this report we will concentrate on the support work that was necessary. For example, we describe in some detail the construction and calibration of a pilot rod for the closed loop system. We discuss the campaign tomore » measure the required reactor parameters necessary for inverse-kinetics. Finally, we briefly discuss the transfer of the open loop technique to other reactor systems.« less

Beryllium for fusion application - recent results

NASA Astrophysics Data System (ADS)

Khomutov, A.; Barabash, V.; Chakin, V.; Chernov, V.; Davydov, D.; Gorokhov, V.; Kawamura, H.; Kolbasov, B.; Kupriyanov, I.; Longhurst, G.; Scaffidi-Argentina, F.; Shestakov, V.

2002-12-01

The main issues for the application of beryllium in fusion reactors are analyzed taking into account the latest results since the ICFRM-9 (Colorado, USA, October 1999) and presented at 5th IEA Be Workshop (10-12 October 2001, Moscow Russia). Considerable progress has been made recently in understanding the problems connected with the selection of the beryllium grades for different applications, characterization of the beryllium at relevant operational conditions (irradiation effects, thermal fatigue, etc.), and development of required manufacturing technologies. The key remaining problems related to the application of beryllium as an armour in near-term fusion reactors (e.g. ITER) are discussed. The features of the application of beryllium and beryllides as a neutron multiplier in the breeder blanket for power reactors (e.g. DEMO) in pebble-bed form are described.

Research Investigation Directed Toward Extending the Useful Range of the Electromagnetic Spectrum

DTIC Science & Technology

1988-12-31

W. Holber, D. Gaines, C. F. Yu, R. M. Osgood, "Laser Desorption of Polymer in a Plasma Reactor," Appl. Phys. Lett. 52, 11 (1988). vii G. V. Treyz, R...and C. Wittig, Chem. Phys. Lett. 67, 48 (1979). 5 P.B. Beeken , E.A. Hanson, and G.W. Flynn, J. Chem. Phys. 78, 5892 (1983). 6 M.C. Heaven, AFOSR Report

High-Energy Electron Confinement in a Magnetic Cusp Configuration

NASA Astrophysics Data System (ADS)

Park, Jaeyoung; Krall, Nicholas A.; Sieck, Paul E.; Offermann, Dustin T.; Skillicorn, Michael; Sanchez, Andrew; Davis, Kevin; Alderson, Eric; Lapenta, Giovanni

2015-04-01

We report experimental results validating the concept that plasma confinement is enhanced in a magnetic cusp configuration when β (plasma pressure/magnetic field pressure) is of order unity. This enhancement is required for a fusion power reactor based on cusp confinement to be feasible. The magnetic cusp configuration possesses a critical advantage: the plasma is stable to large scale perturbations. However, early work indicated that plasma loss rates in a reactor based on a cusp configuration were too large for net power production. Grad and others theorized that at high β a sharp boundary would form between the plasma and the magnetic field, leading to substantially smaller loss rates. While not able to confirm the details of Grad's work, the current experiment does validate, for the first time, the conjecture that confinement is substantially improved at high β . This represents critical progress toward an understanding of the plasma dynamics in a high-β cusp system. We hope that these results will stimulate a renewed interest in the cusp configuration as a fusion confinement candidate. In addition, the enhanced high-energy electron confinement resolves a key impediment to progress of the Polywell fusion concept, which combines a high-β cusp configuration with electrostatic fusion for a compact, power-producing nuclear fusion reactor.

Progress and prospect of true steady state operation with RF

NASA Astrophysics Data System (ADS)

Jacquinot, Jean

2017-10-01

Operation of fusion confinement experiments in full steady state is a major challenge for the development towards fusion energy. Critical to achieving this goal is the availability of actively cooled plasma facing components and auxiliary systems withstanding the very harsh plasma environment. Equally challenging are physics issues related to achieving plasma conditions and current drive efficiency required by reactor plasmas. RF heating and current drive systems have been key instruments for obtaining the progress made until today towards steady state. They hold all the records of long pulse plasma operation both in tokamaks and in stellarators. Nevertheless much progress remains to be made in particular for integrating all the requirements necessary for maintaining in steady state the density and plasma pressure conditions of a reactor. This is an important stated aim of ITER and of devices equipped with superconducting magnets. After considering the present state of the art, this review will address the key issues which remain to be solved both in physics and technology for reaching this goal. They constitute very active subjects of research which will require much dedicated experimentation in the new generation of superconducting devices which are now in operation or becoming close to it.

Progress in catalytic ignition fabrication and modeling : fabrication part 2.

DOT National Transportation Integrated Search

2012-06-01

The ignition temperature and heat generation from oxidation of methane on a platinum catalyst were : determined experimentally. A 127 micron diameter platinum coiled wire was placed crosswise in a : quartz tube of a plug flow reactor. A source meter ...

Development status of rotary engine at Toyo Kogyo. [for general aviation aircraft

NASA Technical Reports Server (NTRS)

Yamamoto, K.

1978-01-01

Progress in the development of rotary engines which use a thermal reactor as the primary part of the exhaust emission control system is reviewed. Possibilities of further improvements in fuel economy of future rotary engines are indicated.

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

Rempe, Joy; Farmer, Mitchell; Corradini, Michael

The Three Mile Island Unit 2 (TMI-2) accident, which occurred on March 28, 1979, led industry and regulators to enhance strategies to protect against severe accidents in commercial nuclear power plants. Investigations in the years after the accident concluded that at least 45% of the core had melted and that nearly 19 tonnes of the core material had relocated to the lower head. Postaccident examinations indicate that about half of that material formed a solid layer near the lower head and above it was a layer of fragmented rubble. As discussed in this paper, numerous insights related to pressurized watermore » reactor accident progression were gained from postaccident evaluations of debris, reactor pressure vessel (RPV) specimens, and nozzles taken from the RPV. In addition, information gleaned from TMI-2 specimen evaluations and available data from plant instrumentation were used to improve severe accident simulation models that form the technical basis for reactor safety evaluations. Finally, the TMI-2 accident led the nuclear community to dedicate considerable effort toward understanding severe accident phenomenology as well as the potential for containment failure. Because available data suggest that significant amounts of fuel heated to temperatures near melting, the events at Fukushima Daiichi Units 1, 2, and 3 offer an unexpected opportunity to gain similar understanding about boiling water reactor accident progression. To increase the international benefit from such an endeavor, we recommend that an international effort be initiated to (a) prioritize data needs; (b) identify techniques, samples, and sample evaluations needed to address each information need; and (c) help finance acquisition of the required data and conduct of the analyses.« less

From Confrontation to Cooperation: 8th International Seminar on Nuclear War

NASA Astrophysics Data System (ADS)

Zichichi, A.; Dardo, M.

1992-09-01

The Table of Contents for the full book PDF is as follows: * OPENING SESSION * A. Zichichi: Opening Statements * R. Nicolosi: Opening Statements * MESSAGES * CONTRIBUTIONS * "The Contribution of the Erice Seminars in East-West-North-South Scientific Relations" * 1. LASER TECHNOLOGY * "Progress in laser technology" * "Progress in laboratory high gain ICF: prospects for the future" * "Applications of laser in metallurgy" * "Laser tissue interactions in medicine and surgery" * "Laser fusion" * "Compact X-ray lasers in the laboratory" * "Alternative method for inertial confinement" * "Laser technology in China" * 2. NUCLEAR AND CHEMICAL SAFETY * "Reactor safety and reactor design" * "Thereotical analysis and numerical modelling of heat transfer and fuel migration in underlying soils and constructive elements of nuclear plants during an accident release from the core" * "How really to attain reactor safely" * "The problem of chemical weapons" * "Long terms genetic effects of nuclear and chemical accidents" * "Features of the brain which are of importance in understanding the mode of operation of toxic substances and of radiation" * "CO2 and ultra safe reactors" * 3. USE OF MISSILES * "How to convert INF technology for peaceful scientific purposes" * "Beating words into plowshares: a proposal for the peaceful uses of retired nuclear warheads" * "Some thoughts on the peaceful use of retired nuclear warheads" * "Status of the HEFEST project" * 4. OZONE * "Status of the ozone layer problem" * 5. CONVENTIONAL AND NUCLEAR FORCE RESTRUCTURING IN EUROPE * 6. CONFLICT AVOIDANCE MODEL * 7. GENERAL DISCUSSION OF THE WORLD LAB PROJECTS * "East-West-North-South Collaboration in Subnuclear Physics" * "Status of the World Lab in the USSR" * CLOSING SESSION

BNL program in support of LWR degraded-core accident analysis

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

Ginsberg, T.; Greene, G.A.

1982-01-01

Two major sources of loading on dry watr reactor containments are steam generatin from core debris water thermal interactions and molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON (Muir, 1981) and MARCH (Wooton, 1980) computer codes. Progress in the two programs is described in this paper. 8 figures.

Advances in boron neutron capture therapy (BNCT) at kyoto university - From reactor-based BNCT to accelerator-based BNCT

NASA Astrophysics Data System (ADS)

Sakurai, Yoshinori; Tanaka, Hiroki; Takata, Takushi; Fujimoto, Nozomi; Suzuki, Minoru; Masunaga, Shinichiro; Kinashi, Yuko; Kondo, Natsuko; Narabayashi, Masaru; Nakagawa, Yosuke; Watanabe, Tsubasa; Ono, Koji; Maruhashi, Akira

2015-07-01

At the Kyoto University Research Reactor Institute (KURRI), a clinical study of boron neutron capture therapy (BNCT) using a neutron irradiation facility installed at the research nuclear reactor has been regularly performed since February 1990. As of November 2014, 510 clinical irradiations were carried out using the reactor-based system. The world's first accelerator-based neutron irradiation system for BNCT clinical irradiation was completed at this institute in early 2009, and the clinical trial using this system was started in 2012. A shift of BCNT from special particle therapy to a general one is now in progress. To promote and support this shift, improvements to the irradiation system, as well as its preparation, and improvements in the physical engineering and the medical physics processes, such as dosimetry systems and quality assurance programs, must be considered. The recent advances in BNCT at KURRI are reported here with a focus on physical engineering and medical physics topics.

Phenol removal from hypersaline wastewaters in a Membrane Biological Reactor (MBR): operation and microbiological characterisation.

PubMed

Dosta, J; Nieto, J M; Vila, J; Grifoll, M; Mata-Álvarez, J

2011-03-01

In this study, two Membrane Biological Reactors (MBR) with submerged flat membranes, one at lab-scale conditions and the other at pilot-plant conditions, were operated at environmental temperature to treat an industrial wastewater characterised by low phenol concentrations (8-16 mg L(-1)) and high salinity (∼ 150-160 mS cm(-1)). During the operation of both reactors, the phenol loading rate was progressively increased and less than 1mg phenol L(-1) was detected even at very low HRTs (0.5-0.7 days). Membrane fouling was minimized by the cross flow aeration rate inside the MBRs and by intermittent permeation. Microbial community analysis of both reactors revealed that members of the genera Halomonas and Marinobacter (gammaproteobacteria) were major components. Growth-linked phenol degradation by pure cultures of Marinobacter isolates demonstrated that this bacterium played a major role in the removal of phenol from the bioreactors. Copyright © 2010 Elsevier Ltd. All rights reserved.

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.

IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

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

Yamada, K.; Aksan, S. N.

The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present,more » 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)« less

NUCLEAR NEW BUILD-INTEGRATING CULTURAL DIFFERENCES IN RADIATION PROTECTION.

PubMed

Haemmerli, Valentin; Bryant, Peter A; Cole, Peter

2017-04-01

Across the world, we are seeing a resurgence in Nuclear New Build. In the UK alone, plans are under way for the construction of 10 new reactors, using 4 different reactor designs all of which are to be provided by foreign vendors, and operated by 3 newly formed licensees within the UK. As these new licensees embark on the task of establishing themselves and progressing the design and build of these reactors, there are challenges faced in integrating the Radiation Protection Requirements and Culture from the various Foreign Investors and Vendors into the UK 'Context'. The following paper identifies the origin of the Radiation Protection Requirements within the UK and foreign investor/vendor countries, in an attempt to integrate them into the UK licensing and approval process. Thus, allowing due credit to be taken for the regulatory regime of the foreign countries where these reactors originate. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

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.

An Evaluation of a Behaviorally Based Social Skills Group for Individuals Diagnosed with Autism Spectrum Disorder

ERIC Educational Resources Information Center

Leaf, Justin B.; Leaf, Jeremy A.; Milne, Christine; Taubman, Mitchell; Oppenheim-Leaf, Misty; Torres, Norma; Townley-Cochran, Donna; Leaf, Ronald; McEachin, John; Yoder, Paul

2017-01-01

In this study we evaluated a social skills group which employed a progressive applied behavior analysis model for individuals diagnosed with autism spectrum disorder. A randomized control trial was utilized; eight participants were randomly assigned to a treatment group and seven participants were randomly assigned to a waitlist control group. The…

A Comparison of Methods for Teaching Receptive Labeling to Children with Autism Spectrum Disorders: A Systematic Replication

ERIC Educational Resources Information Center

Grow, Laura L.; Kodak, Tiffany; Carr, James E.

2014-01-01

Previous research has demonstrated that the conditional-only method (starting with a multiple-stimulus array) is more efficient than the simple-conditional method (progressive incorporation of more stimuli into the array) for teaching receptive labeling to children with autism spectrum disorders (Grow, Carr, Kodak, Jostad, & Kisamore, 2011).…

Increasing Instructional Efficiency by Presenting Additional Stimuli in Learning Trials for Children with Autism Spectrum Disorders

ERIC Educational Resources Information Center

Vladescu, Jason C.; Kodak, Tiffany M.

2013-01-01

The current study examined the effectiveness and efficiency of presenting secondary targets within learning trials for 4 children with an autism spectrum disorder. Specifically, we compared 4 instructional conditions using a progressive prompt delay. In 3 conditions, we presented secondary targets in the antecedent or consequence portion of…

Postsecondary Pathways and Persistence for STEM versus Non-STEM Majors: Among College Students with an Autism Spectrum Disorder

ERIC Educational Resources Information Center

Wei, Xin; Christiano, Elizabeth R. A.; Yu, Jennifer W.; Blackorby, Jose; Shattuck, Paul; Newman, Lynn A.

2014-01-01

Little is known about postsecondary pathways and persistence among college students with an autism spectrum disorder (ASD). This study analyzed data from the National Longitudinal Transition Study-2, 2001-2009, a nationally representative sample of students in special education with an ASD who progressed from high school to postsecondary…

Postsecondary Pathways and Persistence for STEM versus Non-STEM Majors; Among College Students with an Autism Spectrum Disorder

ERIC Educational Resources Information Center

Wei, Xin; Christiano, Elizabeth R.; Yu, Jennifer W.; Blackorby, Jose; Shattuck, Paul; Newman, Lynn

2014-01-01

Little is known about postsecondary pathways and persistence among college students with an Autism Spectrum Disorder (ASD). This study analyzed data from the National Longitudinal Transition Study-2, 2001-2009, a nationally representative sample of students in special education with an ASD who progressed from high school to postsecondary…

A RADIAL REACTOR FOR TRICHLOROETHYLENE STEAM REFORMING. (R822721C633)

EPA Science Inventory

The perspectives, information and conclusions conveyed in research project abstracts, progress reports, final reports, journal abstracts and journal publications convey the viewpoints of the principal investigator and may not represent the views and policies of ORD and EPA. Concl...

Rebuilding the Brookhaven high flux beam reactor: A feasibility study

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

Brynda, W.J.; Passell, L.; Rorer, D.C.

1995-01-01

After nearly thirty years of operation, Brookhaven`s High Flux Beam Reactor (HFBR) is still one of the world`s premier steady-state neutron sources. A major center for condensed matter studies, it currently supports fifteen separate beamlines conducting research in fields as diverse as crystallography, solid-state, nuclear and surface physics, polymer physics and structural biology and will very likely be able to do so for perhaps another decade. But beyond that point the HFBR will be running on borrowed time. Unless appropriate remedial action is taken, progressive radiation-induced embrittlement problems will eventually shut it down. Recognizing the HFBR`s value as a nationalmore » scientific resource, members of the Laboratory`s scientific and reactor operations staffs began earlier this year to consider what could be done both to extend its useful life and to assure that it continues to provide state-of-the-art research facilities for the scientific community. This report summarizes the findings of that study. It addresses two basic issues: (i) identification and replacement of lifetime-limiting components and (ii) modifications and additions that could expand and enhance the reactor`s research capabilities.« less

New results from RENO & prospects with RENO-50

NASA Astrophysics Data System (ADS)

Joo, K. K.

2017-09-01

This paper briefly describes recent progress of RENO and next generation future prospect of the reactor neutrino oscillation experiment, RENO-50. Recently the RENO experiment has updated its latest value on sin22θ 13 and provided new results on 5 MeV excess, Δm2 ee, θ 13 with n-H analysis, absolute antineutrino flux measurement, and sterile neutrino search. It gives rich programs of neutrino properties, detector development, nuclear monitoring and application. Using reactor neutrinos, the future RENO-50 experiment will search for more precise measurement of θ 12, Δm 2 12 and mass hierarchy.

Comparison of Fission Product Yields and Their Impact

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

S. Harrison

2006-02-01

This memorandum describes the Naval Reactors Prime Contractor Team (NRPCT) Space Nuclear Power Program (SNPP) interest in determining the expected fission product yields from a Prometheus-type reactor and assessing the impact of these species on materials found in the fuel element and balance of plant. Theoretical yield calculations using ORIGEN-S and RACER computer models are included in graphical and tabular form in Attachment, with focus on the desired fast neutron spectrum data. The known fission product interaction concerns are the corrosive attack of iron- and nickel-based alloys by volatile fission products, such as cesium, tellurium, and iodine, and the radiologicalmore » transmutation of krypton-85 in the coolant to rubidium-85, a potentially corrosive agent to the coolant system metal piping.« less

Versatile in situ gas analysis apparatus for nanomaterials reactors.

PubMed

Meysami, Seyyed Shayan; Snoek, Lavina C; Grobert, Nicole

2014-09-02

We report a newly developed technique for the in situ real-time gas analysis of reactors commonly used for the production of nanomaterials, by showing case-study results obtained using a dedicated apparatus for measuring the gas composition in reactors operating at high temperature (<1000 °C). The in situ gas-cooled sampling probe mapped the chemistry inside the high-temperature reactor, while suppressing the thermal decomposition of the analytes. It thus allows a more accurate study of the mechanism of progressive thermocatalytic cracking of precursors compared to previously reported conventional residual gas analyses of the reactor exhaust gas and hence paves the way for the controlled production of novel nanomaterials with tailored properties. Our studies demonstrate that the composition of the precursors dynamically changes as they travel inside of the reactor, causing a nonuniform growth of nanomaterials. Moreover, mapping of the nanomaterials reactor using quantitative gas analysis revealed the actual contribution of thermocatalytic cracking and a quantification of individual precursor fragments. This information is particularly important for quality control of the produced nanomaterials and for the recycling of exhaust residues, ultimately leading toward a more cost-effective continuous production of nanomaterials in large quantities. Our case study of multiwall carbon nanotube synthesis was conducted using the probe in conjunction with chemical vapor deposition (CVD) techniques. Given the similarities of this particular CVD setup to other CVD reactors and high-temperature setups generally used for nanomaterials synthesis, the concept and methodology of in situ gas analysis presented here does also apply to other systems, making it a versatile and widely applicable method across a wide range of materials/manufacturing methods, catalysis, as well as reactor design and engineering.

Engineering and Physics Optimization of Breed and Burn Fast Reactor Systems; NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) QUARTERLY PROGRESS REPORT

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

Pavel Hejzlar, Peter Yarsky, Mike Driscoll, Dan Wachs, Kevan Weaver, Ken Czerwinski, Mike Pope, James Parry, Theron D. Marshall, Cliff B. Davis, Dustin Crawford, Thomas Hartmann, Pradip Saha; Hejzlar, Pavel; Yarsky, Peter

2005-01-31

This project is organized under four major tasks (each of which has two or more subtasks) with contributions among the three collaborating organizations (MIT, INEEL and ANL-West): Task A: Core Physics and Fuel Cycle; Task B: Core Thermal Hydraulics; Task C: Plant Design; Task D: Fuel Design The lead PI, Michael J. Driscoll, has consolidated and summarized the technical progress submissions provided by the contributing investigators from all sites, under the above principal task headings.

BNL severe-accident sequence experiments and analysis program. [PWR; BWR

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

Greene, G.A.; Ginsberg, T.; Tutu, N.K.

1983-01-01

In the analysis of degraded core accidents, the two major sources of pressure loading on light water reactor containments are: steam generation from core debris-water thermal interactions; and molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON (Muir, 1981) and MARCH (Wooton, 1980) computer codes. Progress in the two programs is described.

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

Andrews, Nathan; Faucett, Christopher; Haskin, Troy Christopher

Following the conclusion of the first phase of the crosswalk analysis, one of the key unanswered questions was whether or not the deviations found would persist during a partially recovered accident scenario, similar to the one that occurred in TMI - 2. In particular this analysis aims to compare the impact of core degradation morphology on quenching models inherent within the two codes and the coolability of debris during partially recovered accidents. A primary motivation for this study is the development of insights into how uncertainties in core damage progression models impact the ability to assess the potential for recoverymore » of a degraded core. These quench and core recovery models are of the most interest when there is a significant amount of core damage, but intact and degraded fuel still remain in the cor e region or the lower plenum. Accordingly this analysis presents a spectrum of partially recovered accident scenarios by varying both water injection timing and rate to highlight the impact of core degradation phenomena on recovered accident scenarios. This analysis uses the newly released MELCOR 2.2 rev. 966 5 and MAAP5, Version 5.04. These code versions, which incorporate a significant number of modifications that have been driven by analyses and forensic evidence obtained from the Fukushima - Daiichi reactor site.« less

Heavy-section steel irradiation program. Semiannual progress report, October 1995--March 1996

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

Corwin, W.R.

1997-04-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents which have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the capabilities and limitations of the integrity inherent in the RPV. In particular, it is vital to fully understand the degree of irradiation-induced degradation of the RPVs fracture resistance which occurs during service, since without thatmore » radiation damage, it is virtually impossible to postulate a realistic scenario that would result in RPV failure. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established with its primary goal to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior and, in particular, the fracture toughness properties of typical pressure-vessel steels as they relate to light-water RPV integrity. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are being examined on a wide range of fracture properties.« less

Shapes matter: examining the optical response evolution in stretched aluminium nanoparticles via time-dependent density functional theory.

PubMed

Mokkath, Junais Habeeb

2017-12-20

Using first-principles time-dependent density functional theory calculations, we investigate the shape-anisotropy effects on the optical response of a spherical aluminium nanoparticle subjected to a stretching process in different directions. Progressively increased stretching in one direction resulted in prolate spheroid (nanorice) geometries and produced a couple of well-distinguishable dominant peaks together with some satellite peaks in the UV-visible region of the electromagnetic spectrum. On the other hand, progressively increased stretching in two directions caused multiple peaks to appear in the UV-visible region of the electromagnetic spectrum. We believe that our findings can be beneficial for the emerging and potentially far-reaching field of aluminum plasmonics.

Validation of DRAGON4/DONJON4 simulation methodology for a typical MNSR by calculating reactivity feedback coefficient and neutron flux

NASA Astrophysics Data System (ADS)

Al Zain, Jamal; El Hajjaji, O.; El Bardouni, T.; Boukhal, H.; Jaï, Otman

2018-06-01

The MNSR is a pool type research reactor, which is difficult to model because of the importance of neutron leakage. The aim of this study is to evaluate a 2-D transport model for the reactor compatible with the latest release of the DRAGON code and 3-D diffusion of the DONJON code. DRAGON code is then used to generate the group macroscopic cross sections needed for full core diffusion calculations. The diffusion DONJON code, is then used to compute the effective multiplication factor (keff), the feedback reactivity coefficients and neutron flux which account for variation in fuel and moderator temperatures as well as the void coefficient have been calculated using the DRAGON and DONJON codes for the MNSR research reactor. The cross sections of all the reactor components at different temperatures were generated using the DRAGON code. These group constants were used then in the DONJON code to calculate the multiplication factor and the neutron spectrum at different water and fuel temperatures using 69 energy groups. Only one parameter was changed where all other parameters were kept constant. Finally, Good agreements between the calculated and measured have been obtained for every of the feedback reactivity coefficients and neutron flux.

Extension of the Bgl Broad Group Cross Section Library

NASA Astrophysics Data System (ADS)

Kirilova, Desislava; Belousov, Sergey; Ilieva, Krassimira

2009-08-01

The broad group cross-section libraries BUGLE and BGL are applied for reactor shielding calculation using the DOORS package based on discrete ordinates method and multigroup approximation of the neutron cross-sections. BUGLE and BGL libraries are problem oriented for PWR or VVER type of reactors respectively. They had been generated by collapsing the problem independent fine group library VITAMIN-B6 applying PWR and VVER one-dimensional radial model of the reactor middle plane using the SCALE software package. The surveillance assemblies (SA) of VVER-1000/320 are located on the baffle above the reactor core upper edge in a region where geometry and materials differ from those of the middle plane and the neutron field gradient is very high which would result in a different neutron spectrum. That is why the application of the fore-mentioned libraries for the neutron fluence calculation in the region of SA could lead to an additional inaccuracy. This was the main reason to study the necessity for an extension of the BGL library with cross-sections appropriate for the SA region. Comparative analysis of the neutron spectra of the SA region calculated by the VITAMIN-B6 and BGL libraries using the two-dimensional code DORT have been done with purpose to evaluate the BGL applicability for SA calculation.

Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

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

Clarno, Kevin T; Palmtag, Scott; Davidson, Gregory G

2014-01-01

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly casesmore » are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.« less

Neutron Reference Benchmark Field Specification: ACRR Free-Field Environment (ACRR-FF-CC-32-CL).

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

Vega, Richard Manuel; Parma, Edward J.; Griffin, Patrick J.

2015-07-01

This report was put together to support the International Atomic Energy Agency (IAEA) REAL- 2016 activity to validate the dosimetry community’s ability to use a consistent set of activation data and to derive consistent spectral characterizations. The report captures details of integral measurements taken in the Annular Core Research Reactor (ACRR) central cavity free-field reference neutron benchmark field. The field is described and an “a priori” calculated neutron spectrum is reported, based on MCNP6 calculations, and a subject matter expert (SME) based covariance matrix is given for this “a priori” spectrum. The results of 31 integral dosimetry measurements in themore » neutron field are reported.« less

LCRE and SNAP 50-DR-1 programs. Engineering progress report, October 1, 1962--December 31, 1962

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

None

Declassified 5 Sep 1973. Information is presented concerning LCRE specifications, reactor kinetics, fuel elements, primary coolant circuit, secondary coolant circuit, materials development, and fabrication; and SNAP50-DR- 1 specifications, primary pump, and materials development. (DCC)

Building on knowledge base of sodium cooled fast spectrum reactors to develop materials technology for fusion reactors

NASA Astrophysics Data System (ADS)

Raj, Baldev; Rao, K. Bhanu Sankara

2009-04-01

The alloys 316L(N) and Mod. 9Cr-1Mo steel are the major structural materials for fabrication of structural components in sodium cooled fast reactors (SFRs). Various factors influencing the mechanical behaviour of these alloys and different modes of deformation and failure in SFR systems, their analysis and the simulated tests performed on components for assessment of structural integrity and the applicability of RCC-MR code for the design and validation of components are highlighted. The procedures followed for optimal design of die and punch for the near net shape forming of petals of main vessel of 500 MWe prototype fast breeder reactor (PFBR); the safe temperature and strain rate domains established using dynamic materials model for forming of 316L(N) and 9Cr-1Mo steels components by various industrial processes are illustrated. Weldability problems associated with 316L(N) and Mo. 9Cr-1Mo are briefly discussed. The utilization of artificial neural network models for prediction of creep rupture life and delta-ferrite in austenitic stainless steel welds is described. The usage of non-destructive examination techniques in characterization of deformation, fracture and various microstructural features in SFR materials is briefly discussed. Most of the experience gained on SFR systems could be utilized in developing science and technology for fusion reactors. Summary of the current status of knowledge on various aspects of fission and fusion systems with emphasis on cross fertilization of research is presented.

Impact of fission neutron energies on reactor antineutrino spectra

NASA Astrophysics Data System (ADS)

Littlejohn, B. R.; Conant, A.; Dwyer, D. A.; Erickson, A.; Gustafson, I.; Hermanek, K.

2018-04-01

Recent measurements of reactor-produced antineutrino fluxes and energy spectra are inconsistent with models based on measured thermal fission beta spectra. In this paper, we examine the dependence of antineutrino production on fission neutron energy. In particular, the variation of fission product yields with neutron energy has been considered as a possible source of the discrepancies between antineutrino observations and models. In simulations of low-enriched and highly-enriched reactor core designs, we find a substantial fraction of fissions (from 5% to more than 40%) are caused by nonthermal neutrons. Using tabulated evaluations of nuclear fission and decay, we estimate the variation in antineutrino emission by the prominent fission parents U 235 , Pu 239 , and Pu 241 versus neutron energy. The differences in fission neutron energy are found to produce less than 1% variation in detected antineutrino rate per fission of U 235 , Pu 239 , and Pu 241 . Corresponding variations in the antineutrino spectrum are found to be less than 10% below 7 MeV antineutrino energy, smaller than current model uncertainties. We conclude that insufficient modeling of fission neutron energy is unlikely to be the cause of the various reactor anomalies. Our results also suggest that comparisons of antineutrino measurements at low-enriched and highly-enriched reactors can safely neglect the differences in the distributions of their fission neutron energies.

TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

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

Ryan, B.C.

1997-05-01

This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely themore » total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.« less

Core Dynamics Analysis for Reactivity Insertion and Loss of Coolant Flow Tests Using the High Temperature Engineering Test Reactor

NASA Astrophysics Data System (ADS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-temperature Gas-cooled Reactors (HTGRs). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named-ACCORD-, was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel channels and twenty temperature coefficients. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results. Moreover, the effect of the model is formulated quantitatively with our proposed equation. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed.

Parents Suggest Which Indicators of Progress and Outcomes Should Be Measured in Young Children with Autism Spectrum Disorder

ERIC Educational Resources Information Center

McConachie, Helen; Livingstone, Nuala; Morris, Christopher; Beresford, Bryony; Le Couteur, Ann; Gringras, Paul; Garland, Deborah; Jones, Glenys; Macdonald, Geraldine; Williams, Katrina; Parr, Jeremy R.

2018-01-01

Evaluation of interventions for children with autism spectrum disorder (ASD) is hampered by the multitude of outcomes measured and tools used. Measurement in research with young children tends to focus on core impairments in ASD. We conducted a systematic review of qualitative studies of what matters to parents. Parent advisory groups completed…

Using Systematic Instruction and Graphic Organizers to Teach Science Concepts to Students with Autism Spectrum Disorders and Intellectual Disability

ERIC Educational Resources Information Center

Knight, Victoria F.; Spooner, Fred; Browder, Diane M.; Smith, Bethany R.; Wood, Charles L.

2013-01-01

Literacy in science is important for all students and is one component of access and progress in the general education curriculum. One barrier to science literacy for students with autism spectrum disorders (ASD) is the extensive amount of vocabulary involved in comprehending science content. Based on the inherent link between vocabulary knowledge…

Microglia as Biosensors and Effectors of Neurodysfunction

DTIC Science & Technology

2010-04-01

induce or exacerbate the onset and progression of autism spectrum disorders. Dendritic spines receive the majority of excitatory synapses in the brain... autism spectrum disorders, synaptogenesis 15. NUMBER OF PAGES 24 synaptogenesis 16. PRICE CODE 17. SECURITY CLASSIFICATION OF REPORT...4/1/09-4/30/10 INTRODUCTION: Autism occurs during the post-natal period that neurons form new experience-dependent synaptic connections

Cognitive Ability Is Associated with Different Outcome Trajectories in Autism Spectrum Disorders

ERIC Educational Resources Information Center

Ben-Itzchak, Esther; Watson, Linda R.; Zachor, Ditza A.

2014-01-01

Variability in clinical expression and in intervention outcome has been described in autism spectrum disorder (ASD). The study examined progress after 1 and 2 years of intervention and compared the impact of baseline cognitive ability on outcome trajectories in 46 children (m = 25.5 months) with ASD. The entire group showed a gradual decrease in…

Factors Impacting on the Outcomes of Greek Intervention Programmes for Children with Autistic Spectrum Disorders

ERIC Educational Resources Information Center

Makrygianni, Maria K.; Reed, Phil

2010-01-01

This study explored the best predictors of the progress of children with autistic spectrum disorders (ASD), on some developmental domains (autistic severity, language, communication and socialisation), which are related to the core features of ASD. Eighty-six children (2.5-14 years old) with ASD, from 10 schools in Greece, were included in the…

Effects of a Prototypical Training Program on the Implementation of Systematic Observational Data Collection on IEP Objectives for the Core Deficits of Autism Spectrum Disorders

ERIC Educational Resources Information Center

Harkins, Jessica L.

2013-01-01

Legal mandates and best practice recommendations for the education of students with autism spectrum disorders (ASD) emphasize the importance of systematic, ongoing observational data collection in order to monitor progress and demonstrate accountability. The absence of such documentation in decision-making on instructional objectives indicates a…

Impairments in Monkey and Human Face Recognition in 2-Year-Old Toddlers with Autism Spectrum Disorder and Developmental Delay

ERIC Educational Resources Information Center

Chawarska, Katarzyna; Volkmar, Fred

2007-01-01

Face recognition impairments are well documented in older children with Autism Spectrum Disorders (ASD); however, the developmental course of the deficit is not clear. This study investigates the progressive specialization of face recognition skills in children with and without ASD. Experiment 1 examines human and monkey face recognition in…

Transmutation of Isotopes --- Ecological and Energy Production Aspects

NASA Astrophysics Data System (ADS)

Gudowski, Waclaw

2000-01-01

This paper describes principles of Accelerator-Driven Transmutation of Nuclear Wastes (ATW) and gives some flavour of the most important topics which are today under investigations in many countries. An assessment of the potential impact of ATW on a future of nuclear energy is also given. Nuclear reactors based on self-sustained fission reactions --- after spectacular development in fifties and sixties, that resulted in deployment of over 400 power reactors --- are wrestling today more with public acceptance than with irresolvable technological problems. In a whole spectrum of reasons which resulted in today's opposition against nuclear power few of them are very relevant for the nuclear physics community and they arose from the fact that development of nuclear power had been handed over to the nuclear engineers and technicians with some generically unresolved problems, which should have been solved properly by nuclear scientists. In a certain degree of simplification one can say, that most of the problems originate from very specific features of a fission phenomenon: self-sustained chain reaction in fissile materials and very strong radioactivity of fission products and very long half-life of some of the fission and activation products. And just this enormous concentration of radioactive fission products in the reactor core is the main problem of managing nuclear reactors: it requires unconditional guarantee for the reactor core integrity in order to avoid radioactive contamination of the environment; it creates problems to handle decay heat in the reactor core and finally it makes handling and/or disposal of spent fuel almost a philosophical issue, due to unimaginable long time scales of radioactive decay of some isotopes. A lot can be done to improve the design of conventional nuclear reactors (like Light Water Reactors); new, better reactors can be designed but it seems today very improbable to expect any radical change in the public perception of conventional nuclear power. In this context a lot of hopes and expectations have been expressed for novel systems called Accelerator-Driven Systems, Accelerator-Driven Transmutation of Waste or just Hybrid Reactors. All these names are used for description of the same nuclear system combining a powerful particle accelerator with a subcritical reactor. A careful analysis of possible environmental impact of ATW together with limitation of this technology is presented also in this paper.

Synapses as therapeutic targets for autism spectrum disorders: an international symposium held in pavia on july 4th, 2014.

PubMed

Curatolo, Paolo; Ben-Ari, Yehezkel; Bozzi, Yuri; Catania, Maria Vincenza; D'Angelo, Egidio; Mapelli, Lisa; Oberman, Lindsay M; Rosenmund, Christian; Cherubini, Enrico

2014-01-01

New progresses into the molecular and cellular mechanisms of autism spectrum disorders (ASDs) have been discussed in 1 day international symposium held in Pavia (Italy) on July 4th, 2014 entitled "synapses as therapeutic targets for autism spectrum disorders" (satellite of the FENS Forum for Neuroscience, Milan, 2014). In particular, world experts in the field have highlighted how animal models of ASDs have greatly advanced our understanding of the molecular pathways involved in synaptic dysfunction leading sometimes to "synaptic clinical trials" in children.

Recent Results from the Daya Bay Reactor Neutrino Experiment

NASA Astrophysics Data System (ADS)

Huang, En-Chuan

2016-11-01

The Daya Bay Reactor Neutrino Experiment is designed to precisely measure the mixing parameter sin2 2θ13 via relative measurements with eight functionally identical antineutrino detectors (ADs). In 2012, Daya Bay has first measured a non-zero sin2 2θ13 value with a significance larger than 5σ with the first six ADs. With the installation of two new ADs to complete the full configuration, Daya Bay has continued to increase statistics and lower systematic uncertainties for better precision of sin2 2θ13 and for the exploration of other physics topics. In this proceeding, the latest analysis results of sin2 2θ13 and |Δm 2 ee|, including a measurement made with neutron capture on Gadolinium and an independent measurement made with neutron capture on hydrogen are presented. The latest results of the search for sterile neutrino in the mass splitting range of 10-3 eV2 < |Δm 2 41| < 0.3 eV2 and the absolute measurement of the rate and energy spectrum of reactor antineutrinos will also be presented.

Sterile Neutrino Search with the PROSPECT Experiment

NASA Astrophysics Data System (ADS)

Surukuchi Venkata, Pranava Teja

2017-01-01

PROSPECT is a multi-phased short-baseline reactor antineutrino experiment with primary goals of performing a search for sterile neutrinos and making a precise measurement of 235U reactor antineutrino spectrum from the High Flux Isotope Reactor at Oak Ridge National Laboratory. PROSPECT will provide a model independent oscillation measurement of electron antineutrinos by performing relative spectral comparison between a wide range of baselines. By covering the baselines of 7-12 m with Phase-I and extending the coverage to 19m with Phase-II, the PROSPECT experiment will be able to address the current eV-scale sterile neutrino oscillation best-fit region within a single year of data-taking and covers a major portion of suggested parameter space within 3 years of Phase-II data-taking. Additionally, with a Phase-II detector PROSPECT will be able to distinguish between 3+1 mixing, 3+N mixing and other non-standard oscillations. In this talk, we describe the PROSPECT oscillation fitting framework and expected detector sensitivity to the oscillations arising from eV-scale sterile neutrinos. DOE

Application of gaseous core reactors for transmutation of nuclear waste

NASA Technical Reports Server (NTRS)

Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

1976-01-01

An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

Study for requirement of advanced long life small modular fast reactor

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

Tak, Taewoo, E-mail: ttwispy@unist.ac.kr; Choe, Jiwon, E-mail: chi91023@unist.ac.kr; Jeong, Yongjin, E-mail: yjjeong09@unist.ac.kr

2016-01-22

To develop an advanced long-life SMR core concept, the feasibility of the long-life breed-and-burn core concept has been assessed and the preliminary selection on the reactor design requirement such as fuel form, coolant material has been performed. With the simplified cigar-type geometry of 8m-tall CANDLE reactor concept, it has demonstrated the strengths of breed-and-burn strategy. There is a saturation region in the graph for the multiplication factors, which means that a steady breeding is being proceeded along the axial direction. The propagation behavior of the CANDLE core can be also confirmed through the evolution of the axial power profile. Coolantmore » material is expected to have low melting point, density, viscosity and absorption cross section and a high boiling point, specific heat, and thermal conductivity. In this respect, sodium is preferable material for a coolant of this nuclear power plant system. The metallic fuel has harder spectrum compared to the oxide and carbide fuel, which is favorable to increase the breeding and extend the cycle length.« less

Study for requirement of advanced long life small modular fast reactor

NASA Astrophysics Data System (ADS)

Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung; Kim, T. K.

2016-01-01

To develop an advanced long-life SMR core concept, the feasibility of the long-life breed-and-burn core concept has been assessed and the preliminary selection on the reactor design requirement such as fuel form, coolant material has been performed. With the simplified cigar-type geometry of 8m-tall CANDLE reactor concept, it has demonstrated the strengths of breed-and-burn strategy. There is a saturation region in the graph for the multiplication factors, which means that a steady breeding is being proceeded along the axial direction. The propagation behavior of the CANDLE core can be also confirmed through the evolution of the axial power profile. Coolant material is expected to have low melting point, density, viscosity and absorption cross section and a high boiling point, specific heat, and thermal conductivity. In this respect, sodium is preferable material for a coolant of this nuclear power plant system. The metallic fuel has harder spectrum compared to the oxide and carbide fuel, which is favorable to increase the breeding and extend the cycle length.

Cross-Section Measurements in the Fast Neutron Energy Range

NASA Astrophysics Data System (ADS)

Plompen, Arjan

2006-04-01

Generation IV focuses research for advanced nuclear reactors on six concepts. Three of these concepts, the lead, gas and sodium fast reactors (LFR, GFR and SFR) have fast neutron spectra, whereas a fourth, the super-critical water reactor (SCWR), can be configured to have a fast spectrum. Such fast neutron spectra are essential to meet the sustainability objective of GenIV. Nuclear data requirements for GenIV concepts will therefore emphasize the energy region from about 1 keV to 10 MeV. Here, the potential is illustrated of the GELINA neutron time-of-flight facility and the Van de Graaff laboratory at IRMM to measure the relevant nuclear data in this energy range: the total, capture, fission and inelastic-scattering cross sections. In particular, measurement results will be shown for lead and bismuth inelastic scattering for which the need was recently expressed in a quantitative way by Aliberti et al. for Accelerator Driven Systems. Even without completion of the quantitative assessment of the data needs for GenIV concepts at ANL it is clear that this particular effort is of relevance to LFR system studies.

Qualification of heavy water based irradiation device in the JSI TRIGA reactor for irradiations of FT-TIMS samples for nuclear safeguards

NASA Astrophysics Data System (ADS)

Radulović, Vladimir; Kolšek, Aljaž; Fauré, Anne-Laure; Pottin, Anne-Claire; Pointurier, Fabien; Snoj, Luka

2018-03-01

The Fission Track Thermal Ionization Mass Spectrometry (FT-TIMS) method is considered as the reference method for particle analysis in the field of nuclear Safeguards for measurements of isotopic compositions (fissile material enrichment levels) in micrometer-sized uranium particles collected in nuclear facilities. An integral phase in the method is the irradiation of samples in a very well thermalized neutron spectrum. A bilateral collaboration project was carried out between the Jožef Stefan Institute (JSI, Slovenia) and the Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA, France) to determine whether the JSI TRIGA reactor could be used for irradiations of samples for the FT-TIMS method. This paper describes Monte Carlo simulations, experimental activation measurements and test irradiations performed in the JSI TRIGA reactor, firstly to determine the feasibility, and secondly to design and qualify a purpose-built heavy water based irradiation device for FT-TIMS samples. The final device design has been shown experimentally to meet all the required performance specifications.

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.; Morton, T. J.

2006-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG testing will utilize a higher fidelity point kinetics model to control core power transients, and reactivity feedback will be based on localized feedback coefficients and several independent temperature measurements taken within the core block. This paper presents preliminary test results and discusses the methodology that will be implemented in follow-on DDG testing and the additional instrumentation required to implement high fidelity dynamic testing.

Continuous optical measurement system of hemolysis during a photosensitization reaction using absorption spectrum

NASA Astrophysics Data System (ADS)

Hamada, R.; Ogawa, E.; Arai, T.

2018-02-01

To investigate hemolysis phenomena during a photosensitization reaction with the reaction condition continuously and simultaneously for a safety assessment of hemolysis side effect, we constructed an optical system to measure blood sample absorption spectrum during the reaction. Hemolysis degree might be under estimated in general evaluation methods because there is a constant oxygen pressure assumption in spite of oxygen depression take place. By investigating hemoglobin oxidation and oxygen desorption dynamics obtained from the contribution of the visible absorption spectrum and multiple regression analysis, both the hemolysis phenomena and its oxygen environment might be obtained with time. A 664 nm wavelength laser beam for the reaction excitation and 475-650 nm light beam for measuring the absorbance spectrum were arranged perpendicularly crossing. A quartz glass cuvette with 1×10 mm in dimensions for the spectrum measurement was located at this crossing point. A red blood cells suspension medium was arranged with low hematocrit containing 30 μg/ml talaporfin sodium. This medium was irradiated up to 40 J/cm2 . The met-hemoglobin, oxygenatedhemoglobin, and deoxygenated-hemoglobin concentrations were calculated by a multiple regression analysis from the measured spectra. We confirmed the met-hemoglobin concentration increased and oxygen saturation decreased with the irradiation time, which seems to indicate the hemolysis progression and oxygen consumption, respectively. By using our measuring system, the hemolysis progression seems to be obtained with oxygen environment information.

Thermal neutron filter design for the neutron radiography facility at the LVR-15 reactor

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

Soltes, Jaroslav; Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague,; Viererbl, Ladislav

2015-07-01

In 2011 a decision was made to build a neutron radiography facility at one of the unused horizontal channels of the LVR-15 research reactor in Rez, Czech Republic. One of the key conditions for operating an effective radiography facility is the delivery of a high intensity, homogeneous and collimated thermal neutron beam at the sample location. Additionally the intensity of fast neutrons has to be kept as low as possible as the fast neutrons may damage the detectors used for neutron imaging. As the spectrum in the empty horizontal channel roughly copies the spectrum in the reactor core, which hasmore » a high ratio of fast neutrons, neutron filter components have to be installed inside the channel in order to achieve desired beam parameters. As the channel design does not allow the instalment of complex filters and collimators, an optimal solution represent neutron filters made of large single-crystal ingots of proper material composition. Single-crystal silicon was chosen as a favorable filter material for its wide availability in sufficient dimensions. Besides its ability to reasonably lower the ratio of fast neutrons while still keeping high intensities of thermal neutrons, due to its large dimensions, it suits as a shielding against gamma radiation from the reactor core. For designing the necessary filter dimensions the Monte-Carlo MCNP transport code was used. As the code does not provide neutron cross-section libraries for thermal neutron transport through single-crystalline silicon, these had to be created by approximating the theory of thermal neutron scattering and modifying the original cross-section data which are provided with the code. Carrying out a series of calculations the filter thickness of 1 m proved good for gaining a beam with desired parameters and a low gamma background. After mounting the filter inside the channel several measurements of the neutron field were realized at the beam exit. The results have justified the expected calculated values. After the successful filter installing and a series of measurements, first test neutron radiography attempts with test samples could been carried out. (authors)« less

Reliability and safety of the electrical power supply complex of the Hanford production reactors

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

Robbins, F.D.

Safety has been and must continue to be the inviolable modulus by which the operation of a nuclear reactor must be judged. A malfunction in any reactor may well result in a release of fission products which may dissipate over a wide geographical area. Such dissipation may place the health, happiness and even the lives of the people in the region in serious jeopardy. As a result, the property damage and liability cost may reach astronomical values in the order of magnitude of billions of dollars. Reliability of the electrical network is an indispensable factor in attaining a high ordermore » of safety assurance. Progress in the peaceful use of atomic energy may take the form of electrical power generation using the nuclear reactor as a source of thermal energy. In view of these factors it seems appropriate and profitable that a critical engineering study be made of the safety and reliability of the Hanford reactors without regard to cost economics. This individual and independent technical engineering analysis was made without regard to Hanford traditional engineering and administration assignments. The main objective has been to focus attention on areas which seem to merit further detailed study on conditions which seem to need adjustment but most of all on those changes which will improve reactor safety. This report is the result of such a study.« less

Using the sound of nuclear energy

DOE PAGES

Garrett, Steven; Smith, James; Smith, Robert; ...

2016-08-01

The generation of sound by heat has been documented as an “acoustical curiosity” since a Buddhist monk reported the loud tone generated by a ceremonial rice-cooker in his diary, in 1568. Over the last four decades, significant progress has been made in understanding “thermoacoustic processes,” enabling the design of thermoacoustic engines and refrigerators. Motivated by the Fukushima nuclear reactor disaster, we have developed and tested a thermoacoustic engine that exploits the energy-rich conditions in the core of a nuclear reactor to provide core condition information to the operators without a need for external electrical power. The heat engine is self-poweredmore » and can wirelessly transmit the temperature and reactor power level by generation of a pure tone which can be detected outside the reactor. We report here the first use of a fission-powered thermoacoustic engine capable of serving as a performance and safety sensor in the core of a research reactor and present data from the hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These measurements confirmed that the frequency of the sound produced indicates the reactor’s coolant temperature and that the amplitude (above an onset threshold) is related to the reactor’s operating power level. Furthermore, these signals can be detected even in the presence of substantial background noise generated by the reactor’s fluid pumps.« less

Using the sound of nuclear energy

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

Garrett, Steven; Smith, James; Smith, Robert

The generation of sound by heat has been documented as an “acoustical curiosity” since a Buddhist monk reported the loud tone generated by a ceremonial rice-cooker in his diary, in 1568. Over the last four decades, significant progress has been made in understanding “thermoacoustic processes,” enabling the design of thermoacoustic engines and refrigerators. Motivated by the Fukushima nuclear reactor disaster, we have developed and tested a thermoacoustic engine that exploits the energy-rich conditions in the core of a nuclear reactor to provide core condition information to the operators without a need for external electrical power. The heat engine is self-poweredmore » and can wirelessly transmit the temperature and reactor power level by generation of a pure tone which can be detected outside the reactor. We report here the first use of a fission-powered thermoacoustic engine capable of serving as a performance and safety sensor in the core of a research reactor and present data from the hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These measurements confirmed that the frequency of the sound produced indicates the reactor’s coolant temperature and that the amplitude (above an onset threshold) is related to the reactor’s operating power level. Furthermore, these signals can be detected even in the presence of substantial background noise generated by the reactor’s fluid pumps.« less

Supercell Depletion Studies for Prismatic High Temperature Reactors

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

J. Ortensi

2012-10-01

The traditional two-step method of analysis is not accurate enough to represent the neutronic effects present in the prismatic high temperature reactor concept. The long range coupling of the various regions in high temperature reactors poses a set of challenges that are not seen in either LWRs or fast reactors. Unlike LWRs, which exhibit large, localized effects, the dominant effects in PMRs are, for the most part, distributed over larger regions, but with lower magnitude. The 1-D in-line treatment currently used in pebble bed reactor analysis is not sufficient because of the 2-D nature of the prismatic blocks. Considerable challengesmore » exist in the modeling of blocks in the vicinity of reflectors, which, for current small modular reactor designs with thin annular cores, include the majority of the blocks. Additional challenges involve the treatment of burnable poisons, operational and shutdown control rods. The use of a large domain for cross section preparation provides a better representation of the neutron spectrum, enables the proper modeling of BPs and CRs, allows the calculation of generalized equivalence theory parameters, and generates a relative power distribution that can be used in compact power reconstruction. The purpose of this paper is to quantify the effects of the reflector, burnable poison, and operational control rods on an LEU design and to delineate an analysis approach for the Idaho National Laboratory. This work concludes that the use of supercells should capture these long-range effects in the preparation of cross sections and along with a set of triangular meshes to treat BPs, and CRs a high fidelity neutronics computation is attainable.« less

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.

Microstructural evolution of pure tungsten neutron irradiated with a mixed energy spectrum

DOE PAGES

Koyanagi, Takaaki; Kumar, N. A. P. Kiran; Hwang, Taehyun; ...

2017-04-13

Here, microstructures of single-crystal bulk tungsten (W) and polycrystalline W foil with a strong grain texture were investigated using transmission electron microscopy following neutron irradiation at ~90–800 °C to 0.03–4.6 displacements per atom (dpa) in the High Flux Isotope Reactor with a mixed energy spectrum. The dominant irradiation defects were dislocation loops and small clusters at ~90 °C. Additional voids were formed in W irradiated at above 460 °C. Voids and precipitates involving transmutation rhenium and osmium were the dominant defects at more than ~1 dpa. We found a new phenomenon of microstructural evolution in irradiated polycrystalline W: Re- andmore » Os-rich precipitation along grain boundaries. Comparison of results between this study and previous studies using different irradiation facilities revealed that the microstructural evolution of pure W is highly dependent on the neutron energy spectrum in addition to the irradiation temperature and dose.« less

Microstructural evolution of pure tungsten neutron irradiated with a mixed energy spectrum

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

Koyanagi, Takaaki; Kumar, N. A. P. Kiran; Hwang, Taehyun

Here, microstructures of single-crystal bulk tungsten (W) and polycrystalline W foil with a strong grain texture were investigated using transmission electron microscopy following neutron irradiation at ~90–800 °C to 0.03–4.6 displacements per atom (dpa) in the High Flux Isotope Reactor with a mixed energy spectrum. The dominant irradiation defects were dislocation loops and small clusters at ~90 °C. Additional voids were formed in W irradiated at above 460 °C. Voids and precipitates involving transmutation rhenium and osmium were the dominant defects at more than ~1 dpa. We found a new phenomenon of microstructural evolution in irradiated polycrystalline W: Re- andmore » Os-rich precipitation along grain boundaries. Comparison of results between this study and previous studies using different irradiation facilities revealed that the microstructural evolution of pure W is highly dependent on the neutron energy spectrum in addition to the irradiation temperature and dose.« less

Determining the wavelength spectrum of neutrons on the NG6 beam line at NCNR

NASA Astrophysics Data System (ADS)

Ivanov, Juliet

2016-09-01

Historically, in-beam experiments and bottle experiments have been performed to determine the lifetime of a free neutron. However, these two different experimental techniques have provided conflicting results. It is crucial to precisely and accurately elucidate the neutron lifetime for Big Bang Nucleosynthesis calculations and to investigate physics beyond the Standard Model. Therefore, we aimed to understand and minimize systematic errors present in the neutron beam experiment at the NIST Center for Neutron Research (NCNR). In order to reduce the uncertainty related to wavelength dependent corrections present in previous beam experiments, the wavelength spectrum of the NCNR reactor cold neutron beam must be known. We utilized a beam chopper and lithium detector to characterize the wavelength spectrum on the NG6 beam line at the NCNR. The experimental design and techniques employed will be discussed, and our results will be presented. Future plans to utilize our findings to improve the neutron lifetime measurement at NCNR will also be described.

The Reliability and Validity of a Progress-Monitoring Tool: A Psychometric Examination of the Phonological Awareness Skills of Preschoolers with ASD

ERIC Educational Resources Information Center

Martini, Jay R.

2017-01-01

The purpose of this study was to conduct a psychometric evaluation the "Sound Beginning" phonological awareness progress monitoring tool. This assessment was used to track emergent literacy skills of preschoolers with autism spectrum disorder who were participating in a randomized trial studying early literacy interventions. Research…

Using Parents and Teachers to Monitor Progress among Children with ASD: A Review of Intervention Research

ERIC Educational Resources Information Center

Witmer, Sara E.; Nasamran, Amy; Parikh, Purvi J.; Schmitt, Heather A.; Clinton, Marianne C.

2015-01-01

Despite growing knowledge of the effectiveness of various interventions for children with autism spectrum disorders (ASD), it is never clear whether a particular intervention will be effective for a specific child with ASD. Careful monitoring of an individual child's progress is necessary to know whether an intervention is effective. In this…

Studying the Earth from space

USGS Publications Warehouse

,

1981-01-01

scene, as man sees it, on film sensitive to that part of the electromagnetic spectrum called visible energy. Electromagnetic energy travels in waves of various lengths; most are invisible to the human eye. Wavelengths progressively longer than those that the eye can see are infrared and microwave. Wavelengths progressively shorter than the eye can see are ultraviolet, X-rays, and gamma rays.

Comparison of SAND-II and FERRET

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

Wootan, D.W.; Schmittroth, F.

1981-01-01

A comparison was made of the advantages and disadvantages of two codes, SAND-II and FERRET, for determining the neutron flux spectrum and uncertainty from experimental dosimeter measurements as anticipated in the FFTF Reactor Characterization Program. This comparison involved an examination of the methodology and the operational performance of each code. The merits of each code were identified with respect to theoretical basis, directness of method, solution uniqueness, subjective influences, and sensitivity to various input parameters.

High Power Mid Wave Infrared Semiconductor Lasers

DTIC Science & Technology

2006-06-15

resonance and the gain spectrum. The devices were grown using solid source molecular beam epitaxy (MBE) in a V80 reactor. Two side polished, undoped...verify the inherent low activation energy. N-type and P-type AISb, and various compositions of InxAl 1xSb, were grown by solid-source molecular beam ...level monitoring. Advances in epitaxial growth of semiconductor materials have allowed the development of Arsenic- free optically-pumped MWIR lasers on

Multi-phase model development to assess RCIC system capabilities under severe accident conditions

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

Kirkland, Karen Vierow; Ross, Kyle; Beeny, Bradley

The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that themore » system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.« less

Numerical simulations of epitaxial growth process in MOVPE reactor as a tool for design of modern semiconductors for high power electronics

NASA Astrophysics Data System (ADS)

Skibinski, Jakub; Caban, Piotr; Wejrzanowski, Tomasz; Kurzydlowski, Krzysztof J.

2014-10-01

In the present study numerical simulations of epitaxial growth of gallium nitride in Metal Organic Vapor Phase Epitaxy reactor AIX-200/4RF-S is addressed. Epitaxial growth means crystal growth that progresses while inheriting the laminar structure and the orientation of substrate crystals. One of the technological problems is to obtain homogeneous growth rate over the main deposit area. Since there are many agents influencing reaction on crystal area such as temperature, pressure, gas flow or reactor geometry, it is difficult to design optimal process. According to the fact that it's impossible to determine experimentally the exact distribution of heat and mass transfer inside the reactor during crystal growth, modeling is the only solution to understand the process precisely. Numerical simulations allow to understand the epitaxial process by calculation of heat and mass transfer distribution during growth of gallium nitride. Including chemical reactions in numerical model allows to calculate the growth rate of the substrate and estimate the optimal process conditions for obtaining the most homogeneous product.

How Parents Home Educate Their Children with an Autism Spectrum Disorder with the Support of the Schools of Isolated and Distance Education

ERIC Educational Resources Information Center

McDonald, Jasmine; Lopes, Elaine

2014-01-01

Students with an autism spectrum disorder (ASD) often cannot access reliable mainstream inclusive practice that maximises their progress over time. In response to this, some parents have chosen to home educate their children. Limited research indicates that while parents find the experience beneficial for their child, there is a need for…

Different Responses to Trauma in Two Children with Autistic Spectrum Disorder: The Mouth as Crossroads for the Sense of Self

ERIC Educational Resources Information Center

Rhode, Maria

2004-01-01

Two contrasting cases are discussed of boys with autistic spectrum disorder who had suffered cumulative trauma. Although their material was similar in many respects, the 9-year-old made excellent progress during therapy, while the 4-year-old developed much less in spite of being in intensive treatment. This contrast is discussed with regard to…

To the Maximum Extent Appropriate: Determining Success and the Least Restrictive Environment for a Student with Autism Spectrum Disorder

ERIC Educational Resources Information Center

McCloskey, Erin

2016-01-01

This case study uses an institutional ethnography approach to investigate the experiences of the parents of a son who has autism spectrum disorder (ASD) as they engaged in meetings related to his progress and placement with school personnel over the course of one school year. Through an analysis of the discourse used in these meetings and…

Light Water Reactor Sustainability Program, U.S. Efforts in Support of Examinations at Fukushima Daiichi-2017 Evaluations

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

Farmer, Mitchell T.

Although the accident signatures from each unit at the Fukushima Daiichi Nuclear Power Station (NPS) [Daiichi] differ, much is not known about the end-state of core materials within these units. Some of this uncertainty can be attributed to a lack of information related to cooling system operation and cooling water injection. There is also uncertainty in our understanding of phenomena affecting: a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs), and b) accident progression after vessel failure (ex-vessel progression) for BWRs and Pressurized Water Reactors (PWRs). These uncertainties arise due to limited full scale prototypic data.more » Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents. Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document, which has been updated to include FY2017 information, summarizes results from U.S. efforts to use information obtained by TEPCO Holdings to enhance the safety of existing and future nuclear power plant designs. This effort, which was initiated in 2014 by the Reactor Safety Technologies Pathway of the Department of Energy Office of Nuclear Energy Light Water Reactor (LWR) Sustainability Program, consists of a group of U.S. experts in LWR safety and plant operations that have identified examination needs and are evaluating TEPCO Holdings information from Daiichi that address these needs. Each year, annual reports include examples demonstrating that significant safety insights are being obtained in the areas of component performance, fission product release and transport, debris end-state location, and combustible gas generation and transport. In addition to reducing uncertainties related to severe accident modeling progression, these insights are being used to update guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, reduced uncertainties in modeling the events at Daiichi will improve the realism of reactor safety evaluations and inform future D&D activities by improving the capability for characterizing potential hazards to workers involved with cleanup activities. Highlights in this FY2017 report include new insights with respect to the forces required to produce the observed Daiichi Unit 1 (1F1) shield plug endstate, the observed leakage from 1F1 components, and the amount of combustible gas generation required to produce the observed explosions in Daiichi Units 3 and 4 (1F3 and 1F4). This report contains an appendix with a list of examination needs that was updated after U.S. experts reviewed recently obtained information from examinations at Daiichi. Additional details for higher priority, near-term, examination activities are also provided. This report also includes an appendix with a description of an updated website that has been reformatted to better assist U.S. experts by providing information in an archived retrievable location, as well as an appendix summarizing U.S. Forensics activities to host the TMI-2 Knowledge Transfer and Relevance to Fukushima Meeting that was held in Idaho Falls, ID, on October 10-14, 2016.« less

State of the art of aerobic granulation in continuous flow bioreactors.

PubMed

Kent, Timothy R; Bott, Charles B; Wang, Zhi-Wu

In the wake of the success of aerobic granulation in sequential batch reactors (SBRs) for treating wastewater, attention is beginning to turn to continuous flow applications. This is a necessary step given the advantages of continuous flow treatment processes and the fact that the majority of full-scale wastewater treatment plants across the world are operated with aeration tanks and clarifiers in a continuous flow mode. As in SBRs, applying a selection pressure, based on differences in either settling velocity or the size of the biomass, is essential for successful granulation in continuous flow reactors (CFRs). CFRs employed for aerobic granulation come in multiple configurations, each with their own means of achieving such a selection pressure. Other factors, such as bioaugmentation and hydraulic shear force, also contribute to aerobic granulation to some extent. Besides the formation of aerobic granules, long-term stability of aerobic granules is also a critical issue to be addressed. Inorganic precipitation, special inocula, and various operational optimization strategies have been used to improve granule long-term structural integrity. Accumulated studies reviewed in this work demonstrate that aerobic granulation in CFRs is capable of removing a wide spectrum of contaminants and achieving properties generally comparable to those in SBRs. Despite the notable research progress made toward successful aerobic granulation in lab-scale CFRs, to the best of our knowledge, there are only three full-scale tests of the technique, two being seeded with anammox-supported aerobic granules and the other with conventional aerobic granules; two other process alternatives are currently in development. Application of settling- or size-based selection pressures and feast/famine conditions are especially difficult to implement to these and similar mainstream systems. Future research efforts needs to be focused on the optimization of the granule-to-floc ratio, enhancement of granule activity, improvement of long-term granule stability, and a better understanding of aerobic granulation mechanisms in CFRs, especially in full-scale applications. Copyright © 2018 Elsevier Inc. All rights reserved.

Severe early-onset epileptic encephalopathy due to mutations in the KCNA2 gene: Expansion of the genotypic and phenotypic spectrum.

PubMed

Hundallah, Khaled; Alenizi, Asma'a; AlHashem, Amal; Tabarki, Brahim

2016-07-01

Recently, de novo loss- or gain-of-function mutations in the KCNA2 gene; have been described in individuals with epileptic encephalopathy, ataxia or intellectual disability. In this report, we describe a further case of KCNA2-early-onset epileptic encephalopathy. The patient presented since birth with intractable seizures, progressive microcephaly, developmental delay, and progressive brain atrophy. Whole-exome sequencing showed a novel de novo mutation in the KCNA2 gene: c.1120A > G (p.Thr374Ala). This case expands the genotypic and phenotypic disease spectrum of this genetic form of KCNA2-early onset epileptic encephalopathy. Copyright © 2016 European Paediatric Neurology Society. Published by Elsevier Ltd. All rights reserved.

A progressive model for teaching children with autism to follow gaze shift.

PubMed

Gunby, Kristin V; Rapp, John T; Bottoni, Melissa M

2018-06-06

Gunby, Rapp, Bottoni, Marchese and Wu () taught three children with autism spectrum disorder to follow an instructor's gaze shift to select a specific item; however, Gunby et al. used different types of prompts with each participant. To address this limitation, we used a progressive training model for increasing gaze shift for three children with autism spectrum disorder. Results show that each participant learned to follow an adult's shift in gaze to make a correct selection. In addition, two participants displayed the skill in response to a parent's gaze shift and with only social consequences; however, the third participant required verbal instruction and tangible reinforcement to demonstrate the skill outside of training sessions. © 2018 Society for the Experimental Analysis of Behavior.

A comparison of methods for teaching receptive labeling to children with autism spectrum disorders: a systematic replication.

PubMed

Grow, Laura L; Kodak, Tiffany; Carr, James E

2014-01-01

Previous research has demonstrated that the conditional-only method (starting with a multiple-stimulus array) is more efficient than the simple-conditional method (progressive incorporation of more stimuli into the array) for teaching receptive labeling to children with autism spectrum disorders (Grow, Carr, Kodak, Jostad, & Kisamore,). The current study systematically replicated the earlier study by comparing the 2 approaches using progressive prompting with 2 boys with autism. The results showed that the conditional-only method was a more efficient and reliable teaching procedure than the simple-conditional method. The results further call into question the practice of teaching simple discriminations to facilitate acquisition of conditional discriminations. © Society for the Experimental Analysis of Behavior.

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

Monteleone, S.

This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updatedmore » and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.« less

Overview of International Thermonuclear Experimental Reactor (ITER) engineering design activities*

NASA Astrophysics Data System (ADS)

Shimomura, Y.

1994-05-01

The International Thermonuclear Experimental Reactor (ITER) [International Thermonuclear Experimental Reactor (ITER) (International Atomic Energy Agency, Vienna, 1988), ITER Documentation Series, No. 1] project is a multiphased project, presently proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement among the European Atomic Energy Community (EC), the Government of Japan (JA), the Government of the Russian Federation (RF), and the Government of the United States (US), ``the Parties.'' The ITER project is based on the tokamak, a Russian invention, and has since been brought to a high level of development in all major fusion programs in the world. The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER design is being developed, with support from the Parties' four Home Teams and is in progress by the Joint Central Team. An overview of ITER Design activities is presented.

SP-100 system thaw and start-up strategies

NASA Astrophysics Data System (ADS)

Kirpich, A.; Choe, H.

The authors review several strategies that have been considered during calendar year 1990 for SP-100 system start-up and thaw, and provide results of screening analyses. Screening studies have identified three concepts which are capable of thawing the SP-100 system: (1) a heat pipe approach in which reactor-heated heat pipes are routed to equipment enclosed within a thaw cavity; (2) a bleed-tube approach in which perforated tubes routed through ducting and components achieve thaw progressively as the result of successively thawing bleed-holes; and (3) a NaK traceline approach in which reactor-heated NaK is routed along ducting and components and achieves lithium thaw by both radiative and conductive coupling methods. Key issues have been identified in connection with pump start-up, reactor and power converter transient behavior, and hydraulic circuit stability. Pump start-up stability is accomplished by a linked-pump hydraulic arrangement.

Microscale immobilized enzyme reactors in proteomics: latest developments.

PubMed

Safdar, Muhammad; Spross, Jens; Jänis, Janne

2014-01-10

Enzymatic digestion of proteins is one of the key steps in proteomic analyses. There has been a steady progress in the applied digestion protocols in the past, starting from conventional time-consuming in-solution or in-gel digestion protocols to rapid and efficient methods utilizing different types of microscale enzyme reactors. Application of such microreactors has been proven beneficial due to lower sample consumption, higher sensitivity and straightforward coupling with LC-MS set-ups. Novel stationary phases, immobilization techniques and device formats are being constantly developed and tested to optimize digestion efficiency of proteolytic enzymes. This review focuses on the latest developments associated with the preparation and application of microscale enzyme reactors for proteomics applications since 2008 onwards. A special attention has been paid to the discussion of different stationary phases applied for immobilization purposes. Copyright © 2013 Elsevier B.V. All rights reserved.

Overview of the present progress and activities on the CFETR

NASA Astrophysics Data System (ADS)

Wan, Yuanxi; Li, Jiangang; Liu, Yong; Wang, Xiaolin; Chan, Vincent; Chen, Changan; Duan, Xuru; Fu, Peng; Gao, Xiang; Feng, Kaiming; Liu, Songlin; Song, Yuntao; Weng, Peide; Wan, Baonian; Wan, Farong; Wang, Heyi; Wu, Songtao; Ye, Minyou; Yang, Qingwei; Zheng, Guoyao; Zhuang, Ge; Li, Qiang; CFETR Team

2017-10-01

The China Fusion Engineering Test Reactor (CFETR) is the next device in the roadmap for the realization of fusion energy in China, which aims to bridge the gaps between the fusion experimental reactor ITER and the demonstration reactor (DEMO). CFETR will be operated in two phases. Steady-state operation and self-sufficiency will be the two key issues for Phase I with a modest fusion power of up to 200 MW. Phase II aims for DEMO validation with a fusion power over 1 GW. Advanced H-mode physics, high magnetic fields up to 7 T, high frequency electron cyclotron resonance heating and lower hybrid current drive together with off-axis negative-ion neutral beam injection will be developed for achieving steady-state advanced operation. The recent detailed design, research and development (R&D) activities including integrated modeling of operation scenarios, high field magnet, material, tritium plant, remote handling and future plans are introduced in this paper.

A COST BENEFIT APPROACH TO REACTOR SIZING AND NUTRIENT SUPPLY FOR BIOTRICKLING FILTERS FOR AIR POLLUTION CONTROL. (R825392)

EPA Science Inventory

The perspectives, information and conclusions conveyed in research project abstracts, progress reports, final reports, journal abstracts and journal publications convey the viewpoints of the principal investigator and may not represent the views and policies of ORD and EPA. Concl...

EXPERIMENTAL INVESTIGATION OF MERCURY TRANSFORMATIONS IN PILOT-SCALE COMBUSTION SYSTEMS AND A BENCH-SCALE ENTRAINED FLOW REACTOR. (R827649)

EPA Science Inventory

The perspectives, information and conclusions conveyed in research project abstracts, progress reports, final reports, journal abstracts and journal publications convey the viewpoints of the principal investigator and may not represent the views and policies of ORD and EPA. Concl...

POLYMER PYROLYSIS AND OXIDATION STUDIES IN A CONTINUOUS FEED AND FLOW REACTOR: CELLULOSE AND POLYSTYRENE. (R824970)

EPA Science Inventory

The perspectives, information and conclusions conveyed in research project abstracts, progress reports, final reports, journal abstracts and journal publications convey the viewpoints of the principal investigator and may not represent the views and policies of ORD and EPA. Concl...

Infrared identification of the Criegee intermediates syn- and anti-CH3CHOO, and their distinct conformation-dependent reactivity

PubMed Central

Lin, Hui-Yu; Huang, Yu-Hsuan; Wang, Xiaohong; Bowman, Joel M.; Nishimura, Yoshifumi; Witek, Henryk A.; Lee, Yuan-Pern

2015-01-01

The Criegee intermediates are carbonyl oxides that play critical roles in ozonolysis of alkenes in the atmosphere. So far, the mid-infrared spectrum of only the simplest Criegee intermediate CH2OO has been reported. Methyl substitution of CH2OO produces two conformers of CH3CHOO and consequently complicates the infrared spectrum. Here we report the transient infrared spectrum of syn- and anti-CH3CHOO, produced from CH3CHI + O2 in a flow reactor, using a step-scan Fourier-transform spectrometer. Guided and supported by high-level full-dimensional quantum calculations, rotational contours of the four observed bands are simulated successfully and provide definitive identification of both conformers. Furthermore, anti-CH3CHOO shows a reactivity greater than syn-CH3CHOO towards NO/NO2; at the later period of reaction, the spectrum can be simulated with only syn-CH3CHOO. Without NO/NO2, anti-CH3CHOO also decays much faster than syn-CH3CHOO. The direct infrared detection of syn- and anti-CH3CHOO should prove useful for field measurements and laboratory investigations of the Criegee mechanism. PMID:25959902

INTEGRAL REACTION RATES AND NEUTRON ENERGY SPECTRA IN A WELL MODERATED REACTOR

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

Connolly, J.W.; Rose, A.; Wall, T.

1963-04-01

Cadmium ratio measurements in the internal reflector of MOATA were made with gold, indium, tungsten, manganese, molybdenum, and copper detectors. These measurements were analyzed on the assumption that the neutron spectrum consists of a Maxwellian distribution to which is smoothly joined a 1/E slowing down spectrum, the cross sections being averaged according to the methods of Westcott. A search through recent literature suggests that the s factors for gold and indium listed by Westcott are in error. If this is accepted, then it appears that the measured epithermal spectrum is closely 1/E in form for neutron energies between 1 andmore » 600 ev. The corrections to be applied when foils of finite thickness are used in cadmium ratio measuremerts are discussed, and the spectrum derived from these measurements was used to calculate reaction rate ratios of copper: indium and copper: gold alloy foils. These ratios were compared with measured values. Values of the effective resonance integral of Pt/sup 198/ wire detectors were measured, and from these values an estimate was made of the infinitely dilute resonance integral of this isotope. (auth)« less

NPR design basis

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

Locke, G.L.

1958-09-08

The design basis is composed of requirements and conditions for the design of the reactor plant (composed of the reactor and heat dissipation system). Its intent is to insure that the final product meets the economic, safety, and technical objectives of the project. The design basis is dependent on the ground rules, objectives, technical criteria, and practical design considerations. This document is being issued with the understanding that these items are not yet firmly established in all respects, and therefore, the numbers put down here are subject to change. Consideration of the spectrum of probable changes that might be mademore » leads to the conclusion that the numbers here are close to the final ones and are satisfactory as a basis for the initial stages of design. Some numbers are omitted because of insufficient data at this time.« less

Integral cross section measurement of the U 235 ( n , n ' ) U 235 m reaction in a pulsed reactor

DOE PAGES

Bélier, G.; Bond, E. M.; Vieira, D. J.; ...

2015-04-08

The integral measurement of the neutron inelastic cross section leading to the 26-minute half-life 235mU isomer in a fission-like neutron spectrum is presented. The experiment has been performed at a pulsed reactor, where the internal conversion decay of the isomer was measured using a dedicated electron detector after activation. The sample preparation, efficiency measurement, irradiation, radiochemistry purification, and isomer decay measurement will be presented. We determined the integral cross section for the ²³⁵U(n,n') 235mU reaction to be 1.00±0.13b. This result supports an evaluation performed with TALYS-1.4 code with respect to the isomer excitation as well as the total neutron inelasticmore » scattering cross section.« less

Analysis of the Daya Bay Reactor Antineutrino Flux Changes with Fuel Burnup

DOE PAGES

Hayes, A. C.; Ricard-McCutchan, E. A.; Jungman, Gerard; ...

2018-01-12

We investigate the recent Daya Bay results on the changes in the antineutrino flux and spectrum with the burnup of the reactor fuel. We find that the discrepancy between current model predictions and the Daya Bay results can be traced to the original measured 235U/ 239Pu ratio of the fission beta spectra that were used as a base for the expected antineutrino fluxes. An analysis of the antineutrino spectra that is based on a summation over all fission fragment beta-decays, using nuclear database input, explains all of the features seen in the Daya Bay evolution data. However, this summation methodmore » still predicts an anomaly. Thus, we conclude that there is currently not enough information to use the antineutrino flux changes to rule out the possible existence of sterile neutrinos.« less

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

Rice, S.; Algora, A.; Tain, J. L.

The beta decays of Br-86 and Rb-91 have been studied using the total absorption spectroscopy technique. The radioactive nuclei were produced at the Ion Guide Isotope Separator On-Line facility in Jyvaskyla and further purified using the JYFLTRAP. Br-86 and Rb-91 are considered to be major contributors to the decay heat in reactors. In addition, Rb-91 was used as a normalization point in direct measurements of mean gamma energies released in the beta decay of fission products by Rudstam et al. assuming that this decaywas well known from high-resolution measurements. Our results show that both decays were suffering from the Pandemoniummore » effect and that the results of Rudstam et al. should be renormalized. The relative impact of the studied decays in the prediction of the decay heat and antineutrino spectrum from reactors has been evaluated.« less

Neutrino masses and mixing from S4 flavor twisting

NASA Astrophysics Data System (ADS)

Ishimori, Hajime; Shimizu, Yusuke; Tanimoto, Morimitsu; Watanabe, Atsushi

2011-02-01

We discuss a neutrino mass model based on the S4 discrete symmetry where the symmetry breaking is triggered by the boundary conditions of the bulk right-handed neutrino in the fifth spacial dimension. The three generations of the left-handed lepton doublets and the right-handed neutrinos are assigned to be the triplets of S4. The magnitudes of the lepton mixing angles, especially the reactor angle, are related to the neutrino mass patterns, and the model will be tested in future neutrino experiments, e.g., an early discovery of the reactor angle favors the normal hierarchy. For the inverted hierarchy, the lepton mixing is predicted to be almost the tribimaximal mixing. The size of the extra dimension has a connection to the possible mass spectrum; a small (large) volume corresponds to the normal (inverted) mass hierarchy.

Coupling Schemes for Multiphysics Reactor Simulation

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

Vijay Mahadeven; Jean Ragusa

2007-11-01

This report documents the progress of the student Vijay S. Mahadevan from the Nuclear Engineering Department of Texas A&M University over the summer of 2007 during his visit to the INL. The purpose of his visit was to investigate the physics-based preconditioned Jacobian-free Newton-Krylov method applied to physics relevant to nuclear reactor simulation. To this end he studied two test problems that represented reaction-diffusion and advection-reaction. These two test problems will provide the basis for future work in which neutron diffusion, nonlinear heat conduction, and a twophase flow model will be tightly coupled to provide an accurate model of amore » BWR core.« less

Energy efficient continuous flow ash lockhopper

NASA Technical Reports Server (NTRS)

Collins, Earl R., Jr. (Inventor); Suitor, Jerry W. (Inventor); Dubis, David (Inventor)

1989-01-01

The invention relates to an energy efficient continuous flow ash lockhopper, or other lockhopper for reactor product or byproduct. The invention includes an ash hopper at the outlet of a high temperature, high pressure reactor vessel containing heated high pressure gas, a fluidics control chamber having an input port connected to the ash hopper's output port and an output port connected to the input port of a pressure letdown means, and a control fluid supply for regulating the pressure in the control chamber to be equal to or greater than the internal gas pressure of the reactor vessel, whereby the reactor gas is contained while ash is permitted to continuously flow from the ash hopper's output port, impelled by gravity. The main novelty resides in the use of a control chamber to so control pressure under the lockhopper that gases will not exit from the reactor vessel, and to also regulate the ash flow rate. There is also novelty in the design of the ash lockhopper shown in two figures. The novelty there is the use of annular passages of progressively greater diameter, and rotating the center parts on a shaft, with the center part of each slightly offset from adjacent ones to better assure ash flow through the opening.

An RF-Powered Micro-Reactor for Efficient Extraction and Hydrolysis

NASA Astrophysics Data System (ADS)

Scott, V.

2014-12-01

An RF sample-processing micro-reactor that was developed as part of potential in situ Exploration Missions to inner- and outer-planetary bodies was designed to utilize aqueous solutions subjected to 60 GHz radiation at 730 mW of input power to extract target organic compounds and molecular and inorganic ions as well as to hydrolyze complex polymeric materials. Successful identification and characterization of these molecules relies on the sample-processing techniques utilized alongside state-of-the-art detection and analysis. For mass and power restrictions put on space exploration missions, smaller and more efficient instruments are highly desirable. The RF micro-reactor potentially offers a simplified alternative to the typical gold-standard extractions that often use solvents, chemicals, and conditions that can vary wildly and depend on the targeted molecules. Instead, this instrument uses a single solvent ­— water — that can be "tuned" under the different experimental conditions, leveraging the operating principles of the Sub-Critical Water Extractor. Proof-of-concept experiments examining the hydrolysis of glycosidic and peptide bonds were successful in demonstrating the RF micro-reactor's capabilities. Progress toward coupling the reactor with a micro-scale sample-handling system enabling slurry delivery has been made and preliminary results on heterogeneous reactions and extractions will be presented.

The analysis and forecasting of male cycling time trial records established within England and Wales.

PubMed

Dyer, Bryce; Hassani, Hossein; Shadi, Mehran

2016-01-01

The format of cycling time trials in England, Wales and Northern Ireland, involves riders competing individually over several fixed race distances of 10-100 miles in length and using time constrained formats of 12 and 24 h in duration. Drawing on data provided by the national governing body that covers the regions of England and Wales, an analysis of six male competition record progressions was undertaken to illustrate its progression. Future forecasts are then projected through use of the Singular Spectrum Analysis technique. This method has not been applied to sport-based time series data before. All six records have seen a progressive improvement and are non-linear in nature. Five records saw their highest level of record change during the 1950-1969 period. Whilst new record frequency generally has reduced since this period, the magnitude of performance improvement has generally increased. The Singular Spectrum Analysis technique successfully provided forecasted projections in the short to medium term with a high level of fit to the time series data.

The Role of ERK1/2 in the Progression of Anti-Androgen Resistance of MtDNA Deficient Prostate Cancer

DTIC Science & Technology

2012-03-01

proto-oncogenic pathway, it is plausible that the mitoGPS is a ubiquitous (patho) physiological response to the etiology and/or progression of a broad...the mitochondrion as a direct physiological source of hypoxia in an in vitro system. Our results demonstrate that the reduction of the mitochondrial...ubiquitous (patho) physiological response to the etiology and/or progression of a broad spectrum of human diseases that are attributed to respiratory

Progress report No. 2, design of new pile area {open_quotes}G,{close_quotes} Project C-300, Atomic Energy Commission Directive HW-104

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

Lowe, P.E.

1950-01-16

This document is the progress report for the design and development of Pile Area {open_quotes}G{close_quotes}, to cover the period June 1, 1949 to December 31, 1949. This project represents the major effort of the Reactor Division of the Design and Construction Divisions. It is being pursued with the aim of being able to incorporate the major project objectives in a finished pile design if that design is begun about January, 1951.

Index to Nuclear Safety. A technical progress review by chronology, permuted title, and author. Vol. 11, No. 1--Vol. 17, No. 6

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

Cottrell, W.B.; Klein, A.

1977-02-23

This index to Nuclear Safety covers articles in Nuclear Safety Vol. 11, No. 1 (Jan.-Feb. 1970), through Vol. 17, No. 6 (Nov.-Dec. 1976). The index includes a chronological list of articles (including abstract) followed by KWIC and Author Indexes. Nuclear Safety, a bimonthly technical progress review prepared by the Nuclear Safety Information Center, covers all safety aspects of nuclear power reactors and associated facilities. The index lists over 350 technical articles in the last six years of publication.

Progress Report on SAM Reduced-Order Model Development for Thermal Stratification and Mixing during Reactor Transients

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

Hu, R.

This report documents the initial progress on the reduced-order flow model developments in SAM for thermal stratification and mixing modeling. Two different modeling approaches are pursued. The first one is based on one-dimensional fluid equations with additional terms accounting for the thermal mixing from both flow circulations and turbulent mixing. The second approach is based on three-dimensional coarse-grid CFD approach, in which the full three-dimensional fluid conservation equations are modeled with closure models to account for the effects of turbulence.

Neutron Reference Benchmark Field Specification: ACRR 44 Inch Lead-Boron (LB44) Bucket Environment (ACRR-LB44-CC-32-CL).

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

Vega, Richard Manuel; Parma, Edward J.; Griffin, Patrick J.

2015-07-01

This report was put together to support the International Atomic Energy Agency (IAEA) REAL- 2016 activity to validate the dosimetry community’s ability to use a consistent set of activation data and to derive consistent spectral characterizations. The report captures details of integral measurements taken in the Annular Core Research Reactor (ACRR) central cavity with the 44 inch Lead-Boron (LB44) bucket, reference neutron benchmark field. The field is described and an “a priori” calculated neutron spectrum is reported, based on MCNP6 calculations, and a subject matter expert (SME) based covariance matrix is given for this “a priori” spectrum. The results ofmore » 31 integral dosimetry measurements in the neutron field are reported.« less

Neutron Reference Benchmark Field Specifications: ACRR Polyethylene-Lead-Graphite (PLG) Bucket Environment (ACRR-PLG-CC-32-CL).

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

Vega, Richard Manuel; Parm, Edward J.; Griffin, Patrick J.

2015-07-01

This report was put together to support the International Atomic Energy Agency (IAEA) REAL- 2016 activity to validate the dosimetry community’s ability to use a consistent set of activation data and to derive consistent spectral characterizations. The report captures details of integral measurements taken in the Annular Core Research Reactor (ACRR) central cavity with the Polyethylene-Lead-Graphite (PLG) bucket, reference neutron benchmark field. The field is described and an “a priori” calculated neutron spectrum is reported, based on MCNP6 calculations, and a subject matter expert (SME) based covariance matrix is given for this “a priori” spectrum. The results of 37 integralmore » dosimetry measurements in the neutron field are reported.« less

The MSFR as a flexible CR reactor: the viewpoint of safety

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

Fiorina, C.; Cammi, A.; Franceschini, F.

2013-07-01

In this paper, the possibility has first been discussed of using the liquid-fuelled Molten Salt Fast Reactor (MSFR) as a flexible conversion ratio (CR) reactor without design modification. By tuning the reprocessing rate it is possible to determine the content of fission products in the core, which in turn can significantly affect the neutron economy without incurring in solubility problems. The MSFR can thus be operated as U-233 breeder (CR>1), iso-breeder (CR=1) and burner reactor (CR<1). In particular a 40 year doubling time can be achieved, as well as a considerable Transuranics and MA (minor actinide) burning rate equal tomore » about 150 kg{sub HN}/GWE-yr. The safety parameters of the MSFR have then been evaluated for different fuel cycle strategies. Th use and a softer spectrum combine to give a strong Doppler coefficient, one order of magnitude higher compared to traditional fast reactors (FRs). The fuel expansion coefficient is comparable to the Doppler coefficient and is only mildly affected by core compositions, thus assisting the fuel cycle flexibility of the MSFR. βeff and generation time are comparable to the case of traditional FRs, if a static fuel is assumed. A notable reduction of βeff is caused by salt circulation, but a low value of this parameter is a limited concern in the MSFR thanks to the lack of a burnup reactivity swing and of positive feedbacks. A simple approach has also been developed to evaluate the MSFR capabilities to withstand all typical double-fault accidents, for different fuel cycle options.« less

Activation product transport in fusion reactors. [RAPTOR

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

Klein, A.C.

1983-01-01

Activated corrosion and neutron sputtering products will enter the coolant and/or tritium breeding material of fusion reactor power plants and experiments and cause personnel access problems. Radiation levels around plant components due to these products will cause difficulties with maintenance and repair operations throughout the plant. Similar problems are experienced around fission reactor systems. The determination of the transport of radioactive corrosion and neutron sputtering products through the system is achieved using the computer code RAPTOR. This code calculates the mass transfer of a number of activation products based on the corrosion and sputtering rates through the system, the depositionmore » and release characteristics of various plant components, the neturon flux spectrum, as well as other plant parameters. RAPTOR assembles a system of first order linear differential equations into a matrix equation based upon the reactor system parameters. Included in the transfer matrix are the deposition and erosion coefficients, and the decay and activation data for the various plant nodes and radioactive isotopes. A source vector supplies the corrosion and neutron sputtering source rates. This matrix equation is then solved using a matrix operator technique to give the specific activity distribution of each radioactive species throughout the plant. Once the amount of mass transfer is determined, the photon transport due to the radioactive corrosion and sputtering product sources can be evaluated, and dose rates around the plant components of interest as a function of time can be determined. This method has been used to estimate the radiation hazards around a number of fusion reactor system designs.« less

Test of a prototype neutron spectrometer based on diamond detectors in a fast reactor

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

Osipenko, M.; Ripani, M.; Ricco, G.

2015-07-01

A prototype of neutron spectrometer based on diamond detectors has been developed. This prototype consists of a {sup 6}Li neutron converter sandwiched between two CVD diamond crystals. The radiation hardness of the diamond crystals makes it suitable for applications in low power research reactors, while a low sensitivity to gamma rays and low leakage current of the detector permit to reach good energy resolution. A fast coincidence between two crystals is used to reject background. The detector was read out using two different electronic chains connected to it by a few meters of cable. The first chain was based onmore » conventional charge-sensitive amplifiers, the other used a custom fast charge amplifier developed for this purpose. The prototype has been tested at various neutron sources and showed its practicability. In particular, the detector was calibrated in a TRIGA thermal reactor (LENA laboratory, University of Pavia) with neutron fluxes of 10{sup 8} n/cm{sup 2}s and at the 3 MeV D-D monochromatic neutron source named FNG (ENEA, Rome) with neutron fluxes of 10{sup 6} n/cm{sup 2}s. The neutron spectrum measurement was performed at the TAPIRO fast research reactor (ENEA, Casaccia) with fluxes of 10{sup 9} n/cm{sup 2}s. The obtained spectra were compared to Monte Carlo simulations, modeling detector response with MCNP and Geant4. (authors)« less

Progression of Challenging Behaviors in Children and Adolescents with Autism Spectrum Disorders as Measured by the Autism Spectrum Disorders-Problem Behaviors for Children (ASD-PBC)

ERIC Educational Resources Information Center

Matson, Johnny L.; Mahan, Sara; Hess, Julie A.; Fodstad, Jill C.; Neal, Daniene

2010-01-01

This study examined the effect of age on challenging behaviors among 167 children, ages 3-14 years, with Autistic Disorder, Pervasive Developmental Disorder Not Otherwise Specified, or Asperger's syndrome. Results of a MANOVA indicated that there were no significant differences between young children, children, and young adolescents on any of the…

Update on autism spectrum disorder: vaccines, genomes, and social skills training.

PubMed

McGuinness, Teena M

2015-04-01

Despite making significant progress in understanding autism spectrum disorder (ASD) and its genetic underpinnings, controversy remains regarding ASD and its historical, erroneous association with vaccines. This controversy includes the latest anti-vaccine movement that caused a recurrence of the almost vanquished measles and mumps diseases. The history of ASD, complexities of research involving ASD genetics, and benefits of social skills training are explored. Copyright 2015, SLACK Incorporated.

The Effects of Intertrial Interval and Instructional Format on Skill Acquisition and Maintenance for Children with Autism Spectrum Disorders

ERIC Educational Resources Information Center

Cariveau, Tom; Kodak, Tiffany; Campbell, Vincent

2016-01-01

We replicated and extended the study by Koegel, Dunlap, and Dyer (1980) by examining the effects of 3 intertrial-interval (ITI) durations on skill acquisition in 2 children with autism spectrum disorders. Specifically, we compared the effect of short (2 s), progressive (2 s to 20 s), and long (20 s) ITIs on participants' mastery of tacts or…

Quarterly Progress Report (January 1 to March 31, 1950)

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

Brookhaven National Laboratory

This is the first of a series of Quarterly Reports. These reports will deal primarily with the progress made in our scientific program during a three months period. Those interested in matters pertaining to organization, administration, complete scientific program, personnel and other matters not directly involved in current scientific progress are referred to our Annual Progress Report which is issued in January. We have attempted to describe new information that appears significant, or of interest, to other scientists within the Atomic Energy Commission Laboratories. No effort has been made, however, to detail progress in each and every research project. Littlemore » or no reference will therefore be found to the projects in which progress during the current period is considered too inconclusive. Since our organizational structure is departmental, the work described herein is arranged in the following sequence: (1) Accelerator Project; (2) Biology Department; (3) Chemistry Department; (4) Instrumentation and Health Physic8 Department; (5) Medical Department; (6) Physics Department; and (7) Reactor Science and Engineering Department.« less

Status and progress of the RERTR program in the year 2003.

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

Travelli, A.; Nuclear Engineering Division

2003-01-01

One of the most important events affecting the RERTR program during the past year was the decision by the U.S. Department of Energy to request the U.S. Congress to significantly increase RERTR program funding. This decision was prompted, at least in part, by the terrible events of September 11, 2001, and by a high-level U.S./Russian Joint Expert Group recommendation to immediately accelerate RERTR program activities in both countries, with the goal of converting all the world's research reactors to low-enriched fuel at the earliest possible time, and including both Soviet-designed and United States-designed research reactors. The U.S. Congress is expectedmore » to approve this request very soon, and the RERTR program has prepared itself well for the intense activities that the 'Accelerated RERTR Program' will require. Promising results have been obtained in the development of a fabrication process for monolithic LEU U-Mo fuel. Most existing and future research reactors could be converted to LEU with this fuel, which has a uranium density between 15.4 and 16.4 g/cm{sup 3} and yielded promising irradiation results in 2002. The most promising method hinges on producing the monolithic meat by cold-rolling a thin ingot produced by casting. The aluminum clad and the meat are bonded by friction stir welding and the cladding surface is finished by a light cold roll. This method can be applied to the production of miniplates and appears to be extendable to the production of full-size plates, possibly with intermediate anneals. Other methods planned for investigation include high temperature bonding and hot isostatic pressing. The progress achieved within the Russian RERTR program, both for the traditional tube-type elements and for the new 'universal' LEU U-Mo pin-type elements, promises to enable soon the conversion of many Russian-designed research and test reactors. Irradiation testing of both fuel types with LEU U-Mo dispersion fuels has begun. Detailed studies are in progress to define the feasibility of converting each Russian-designed research and test reactor to either fuel type. The plan for the Accelerated RERTR Program is structured to achieve LEU conversion of all HEU research reactors supplied by the United States and Russia during the next nine years. This effort will address, in addition to the fuel development and qualification, the analyses and performance/economic/safety evaluations needed to implement the conversions. In combination with this over-arching goal, the RERTR program plans to achieve at the earliest possible date qualification of LEU U-Mo dispersion fuels with uranium densities of 6 g/cm{sup 3} and 7 g/cm{sup 3}. Reactors currently using or planning to use LEU silicide fuel will rely on this fuel after termination of the FRRSNFA program, because it is acceptable to COGEMA for reprocessing. Qualification of LEU U-Mo dispersion fuels has suffered some unavoidable delays but, to accelerate it as much as possible, the RERTR program, the French CEA, and the Australian ANSTO have agreed to jointly pursue a two-element qualification test of LEU U-Mo dispersion fuel with uranium density of 7.0 g/cm{sup 3} to be performed in the Osiris reactor during 2004. The RERTR program also intends to eliminate all obstacles to the utilization of LEU in targets for isotope production, so that this important function can be performed without the need for weapons-grade materials. All of us, working together as we have for many years, can ensure that all these goals will be achieved. By promoting the efficiency and safety of research reactors while eliminating the traffic in weapons-grade uranium, we can prevent the possibility that some of this material might fall in the wrong hands. Few causes can be more deserving of our joint efforts.« less

Rett syndrome and MeCP2.

PubMed

Liyanage, Vichithra R B; Rastegar, Mojgan

2014-06-01

Rett syndrome (RTT) is a severe and progressive neurological disorder, which mainly affects young females. Mutations of the methyl-CpG binding protein 2 (MECP2) gene are the most prevalent cause of classical RTT cases. MECP2 mutations or altered expression are also associated with a spectrum of neurodevelopmental disorders such as autism spectrum disorders with recent links to fetal alcohol spectrum disorders. Collectively, MeCP2 relation to these neurodevelopmental disorders highlights the importance of understanding the molecular mechanisms by which MeCP2 impacts brain development, mental conditions, and compromised brain function. Since MECP2 mutations were discovered to be the primary cause of RTT, a significant progress has been made in the MeCP2 research, with respect to the expression, function and regulation of MeCP2 in the brain and its contribution in RTT pathogenesis. To date, there have been intensive efforts in designing effective therapeutic strategies for RTT benefiting from mouse models and cells collected from RTT patients. Despite significant progress in MeCP2 research over the last few decades, there is still a knowledge gap between the in vitro and in vivo research findings and translating these findings into effective therapeutic interventions in human RTT patients. In this review, we will provide a synopsis of Rett syndrome as a severe neurological disorder and will discuss the role of MeCP2 in RTT pathophysiology.

Neutron spectrum measurements using proton recoil proportional counters: results of measurements of leakage spectra for the Little Boy assembly

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

Bennett, E.F.; Yule, T.J.

1984-01-01

Measurements of degraded fission-neutron spectra using recoil proportional counters are done routinely for studies involving fast reactor mockups. The same techniques are applicable to measurements of neutron spectra required for personnel dosimetry in fast neutron environments. A brief discussion of current applications of these methods together with the results of a measurement made on the LITTLE BOY assembly at Los Alamos are here described.

Background evaluation for the neutron sources in the Daya Bay experiment

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

Gu, W. Q.; Cao, G. F.; Chen, X. H.

2016-07-06

Here, we present an evaluation of the background induced by 241Am–13C neutron calibration sources in the Daya Bay reactor neutrino experiment. Furthermore, as a significant background for electron-antineutrino detection at 0.26 ± 0.12 detector per day on average, it has been estimated by a Monte Carlo simulation that was benchmarked by a special calibration data set. This dedicated data set also provides the energy spectrum of the background.

Broad Energy Range Neutron Spectroscopy using a Liquid Scintillator and a Proportional Counter: Application to a Neutron Spectrum Similar to that from an Improvised Nuclear Device

PubMed Central

Randers-Pehrson, Gerhard; Marino, Stephen A.; Garty, Guy; Harken, Andrew; Brenner, David J.

2015-01-01

A novel neutron irradiation facility at the Radiological Research Accelerator Facility (RARAF) has been developed to mimic the neutron radiation from an Improvised Nuclear Device (IND) at relevant distances (e.g. 1.5 km) from the epicenter. The neutron spectrum of this IND-like neutron irradiator was designed according to estimations of the Hiroshima neutron spectrum at 1.5 km. It is significantly different from a standard reactor fission spectrum, because the spectrum changes as the neutrons are transported through air, and it is dominated by neutron energies from 100 keV up to 9 MeV. To verify such wide energy range neutron spectrum, detailed here is the development of a combined spectroscopy system. Both a liquid scintillator detector and a gas proportional counter were used for the recoil spectra measurements, with the individual response functions estimated from a series of Monte Carlo simulations. These normalized individual response functions were formed into a single response matrix for the unfolding process. Several accelerator-based quasi-monoenergetic neutron source spectra were measured and unfolded to test this spectroscopy system. These reference neutrons were produced from two reactions: T(p,n)3He and D(d,n)3He, generating neutron energies in the range between 0.2 and 8 MeV. The unfolded quasi-monoenergetic neutron spectra indicated that the detection system can provide good neutron spectroscopy results in this energy range. A broad-energy neutron spectrum from the 9Be(d,n) reaction using a 5 MeV deuteron beam, measured at 60 degrees to the incident beam was measured and unfolded with the evaluated response matrix. The unfolded broad neutron spectrum is comparable with published time-of-flight results. Finally, the pair of detectors were used to measure the neutron spectrum generated at the RARAF IND-like neutron facility and a comparison is made to the neutron spectrum of Hiroshima. PMID:26273118

Broad Energy Range Neutron Spectroscopy using a Liquid Scintillator and a Proportional Counter: Application to a Neutron Spectrum Similar to that from an Improvised Nuclear Device.

PubMed

Xu, Yanping; Randers-Pehrson, Gerhard; Marino, Stephen A; Garty, Guy; Harken, Andrew; Brenner, David J

2015-09-11

A novel neutron irradiation facility at the Radiological Research Accelerator Facility (RARAF) has been developed to mimic the neutron radiation from an Improvised Nuclear Device (IND) at relevant distances (e.g. 1.5 km) from the epicenter. The neutron spectrum of this IND-like neutron irradiator was designed according to estimations of the Hiroshima neutron spectrum at 1.5 km. It is significantly different from a standard reactor fission spectrum, because the spectrum changes as the neutrons are transported through air, and it is dominated by neutron energies from 100 keV up to 9 MeV. To verify such wide energy range neutron spectrum, detailed here is the development of a combined spectroscopy system. Both a liquid scintillator detector and a gas proportional counter were used for the recoil spectra measurements, with the individual response functions estimated from a series of Monte Carlo simulations. These normalized individual response functions were formed into a single response matrix for the unfolding process. Several accelerator-based quasi-monoenergetic neutron source spectra were measured and unfolded to test this spectroscopy system. These reference neutrons were produced from two reactions: T(p,n) 3 He and D(d,n) 3 He, generating neutron energies in the range between 0.2 and 8 MeV. The unfolded quasi-monoenergetic neutron spectra indicated that the detection system can provide good neutron spectroscopy results in this energy range. A broad-energy neutron spectrum from the 9 Be(d,n) reaction using a 5 MeV deuteron beam, measured at 60 degrees to the incident beam was measured and unfolded with the evaluated response matrix. The unfolded broad neutron spectrum is comparable with published time-of-flight results. Finally, the pair of detectors were used to measure the neutron spectrum generated at the RARAF IND-like neutron facility and a comparison is made to the neutron spectrum of Hiroshima.

Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

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

Beyer, Brian David; Beddingfield, David H; Durst, Philip

2010-01-01

The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguardsmore » criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.« less

NRC ARDC Guidance Support Status Report

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

Holbrook, Mark R.

This report provides a summary that reflects the progress and status of proposed regulatory design criteria for advanced non-light water reactor (LWR) designs in accordance with the Level 3 milestone M3AT-17IN2001013 in work package AT-17IN200101. These criteria have been designated as advanced reactor design criteria (ARDC) and they provide guidance to future applicants for addressing the general design criteria (GDC) that are currently applied specifically to LWR designs. This report provides a summary of Phase 2 activities related to the various tasks associated with ARDC development and the subsequent development of ARDC regulatory guidance for sodium fast reactor (SFR) andmore » modular high-temperature gas-cooled reactor (HTGR) designs. Status Report Organization: Section 2 discusses the origin of the GDC and their application to LWRs. Section 3 addresses the objective of this initiative and how it benefits the advanced non-LWR reactor vendors. Section 4 discusses the scope and structure of the initiative. Section 5 provides background on the U.S. Department of Energy (DOE) ARDC team’s original development of the proposed ARDC that were submitted to the NRC for consideration. Section 6 provides a summary of recent ARDC Phase 2 activities. Appendices A through E document the DOE ARDC team’s public comments on various sections of the NRC’s draft regulatory guide DG–1330, “Guidance for Developing Principal Design Criteria for Non-Light Water Reactors.”« less

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

Batchelor, D.B.; Carreras, B.A.; Hirshman, S.P.

Significant progress has been made in the development of new modest-size compact stellarator devices that could test optimization principles for the design of a more attractive reactor. These are 3 and 4 field period low-aspect-ratio quasi-omnigenous (QO) stellarators based on an optimization method that targets improved confinement, stability, ease of coil design, low-aspect-ratio, and low bootstrap current.

Solid State Division progress report for period ending March 31, 1992

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

Green, P.H.; Hinton, L.W.

1992-09-01

During this period, the division conducted a broad, interdisciplinary materials research program with emphasis on theoretical solid state physics, superconductivity, neutron scattering, synthesis and characterization of materials, ion beam and laser processing, and the structure of solids and surfaces. The High Flux Isotope Reactor was returned to full operation.

JPL in-house fluidized bed reactor research

NASA Technical Reports Server (NTRS)

Rohatgi, N. K.

1985-01-01

The progress in the in-house program on the silane fluidized-bed system is reported. A seed-particle cleaning procedure was developed to obtain material purity near the level required to produce a semiconductor-grade product. The liner-seal design was consistently proven to withstand heating/cooling cycles in all of the experimental runs.

Neuroleptic-induced catatonia: clinical presentation, response to benzodiazepines, and relationship to neuroleptic malignant syndrome.

PubMed

Lee, Joseph W Y

2010-02-01

Neuroleptic-induced catatonia (NIC), manifested in an extrapyramidal-catatonic syndrome, has been sporadically reported in the literature. Confusion surrounds its relationship to neuroleptic malignant syndrome (NMS) and extrapyramidal reactions to neuroleptics. This study examined (a) its clinical presentation and response to benzodiazepines, (b) the hypothesis that NIC and NMS are on the same spectrum with a continuum of symptom progression, and (c) its possible relationship to extrapyramidal reactions. Of 127 episodes of acute catatonia prospectively identified, 18 were diagnosed with NIC. All catatonia episodes received benzodiazepines. The NIC episodes were analyzed noting their clinical presentations, laboratory findings, and responses to treatments. Their responses to benzodiazepines were compared, with retrospective rating on a 7-point scale, to that for catatonia episodes associated with mania and schizophrenia. The progression of symptoms in each NIC episode was reviewed. The NIC episodes presented predominantly in the stuporous form associated with parkinsonism. Delirium, autonomic abnormality, and elevated serum creatine phosphokinase were all common. Neuroleptic malignant syndrome was diagnosed in 3 episodes (17%). The 3 catatonia groups did not differ significantly in their benzodiazepines responses: 78% (14/18) of NIC, 75% (12/16) of manic catatonia, and 67% (34/51) of schizophrenic catatonia episodes showed full responses. A spectrum of presentation across episodes was noted with simple NIC without delirium, autonomic disturbances, or fever at one end and NMS or malignant NIC at the other end. Symptoms in individual episodes showed a similar continuum progression. No extrapyramidal reactions immediately preceded the NIC episodes. Findings of this study support the hypothesis that NIC and NMS are disorders on the same spectrum and reveal no indication that extrapyramidal reactions progress to NIC.

Jordanian Parents' Beliefs about the Causes of Disability and the Progress of Their Children with Disabilities: Insights on Mainstream Schools and Segregated Centres

ERIC Educational Resources Information Center

Al-Dababneh, Kholoud Adeeb; Al-Zboon, Eman K.; Baibers, Haitham

2017-01-01

This study aims to identify the beliefs of Jordanian parents of children with disabilities (CWD), including intellectual disabilities, specific learning disorders and Autism Spectrum Disorder: both in terms of the causes of these disabilities, and the ability of their children to make progress. A qualitative interpretive methodology was employed.…

Front-end Design and Characterization for the ν-Angra Nuclear Reactor Monitoring Detector

NASA Astrophysics Data System (ADS)

Dornelas, T. I.; Araújo, F. T. H.; Cerqueira, A. S.; Costa, J. A.; Nóbrega, R. A.

2016-07-01

The Neutrinos Angra (ν-Angra) Experiment aims to construct an antineutrinos detection device capable of monitoring the Angra dos Reis nuclear reactor activity. Nuclear reactors are intense sources of antineutrinos, and the thermal power released in the fission process is directly related to the flow rate of these particles. The antineutrinos energy spectrum also provides valuable information on the nuclear source isotopic composition. The proposed detector will be equipped with photomultipliers tubes (PMT) which will be readout by a custom Amplifier-Shaper-Discriminator circuit designed to condition its output signals to the acquisition modules to be digitized and processed by an FPGA. The readout circuit should be sensitive to single photoelectron signals, process fast signals, with a full-width-half-amplitude of about 5 ns, have a narrow enough output pulse width to detect both particles coming out from the inverse beta decay (bar nue+p → n + e+), and its output amplitude should be linear to the number of photoelectrons generated inside the PMT, used for energy estimation. In this work, some of the main PMT characteristics are measured and a new readout circuit is proposed, described and characterized.

Multifaceted effects of HZSM-5 (Proton-exchanged Zeolite Socony Mobil-5) on catalytic cracking of pinewood pyrolysis vapor in a two-stage fixed bed reactor.

PubMed

Wang, Yimeng; Wang, Jie

2016-08-01

The pinewood was pyrolyzed in the first reactor at a heating rate of 10°Cmin(-1) from room temperature to 700°C, and the vapor was allowed to be cracked through the second reactor in a temperature range of 450-750°C without and with HZSM-5. Attempts were made to determine a wide spectrum of gaseous and liquid products, as well as the mass and element partitions to gas, water, bio-oil, coke and char. HZSM-5 showed a preferential deoxygenation effect via the facilitated decarbonylation and decarboxylation with the inhibited dehydration at 550-600°C. This catalyst also displayed a high selectivity for the formations of aromatic hydrocarbons and olefins by the promoted hydrogen transfer to these products at 550-600°C. The bio-oil produced with HZSM-5 at 500-600°C had the yields of 14.5-16.8%, the high heat values of 39.1-42.4MJkg(-1), and the energy recoveries of 33-35% (all dry biomass basis). Copyright © 2016 Elsevier Ltd. All rights reserved.

[Studies on photo-electron-chemical catalytic degradation of the malachite green].

PubMed

Li, Ming-yu; Diao, Zeng-hui; Song, Lin; Wang, Xin-le; Zhang, Yuan-ming

2010-07-01

A novel two-compartment photo-electro-chemical catalytic reactor was designed. The TiO2/Ti thin film electrode thermally formed was used as photo-anode, and graphite as cathode and a saturated calomel electrode (SCE) as the reference electrode in the reactor. The anode compartment and cathode compartment were connected with the ionic exchange membrane in this reactor. Effects of initial pH, initial concentration of malachite green and connective modes between the anode compartment and cathode compartment on the decolorization efficiency of malachite green were investigated. The degradation dynamics of malachite green was studied. Based on the change of UV-visible light spectrum, the degradation process of malachite green was discussed. The experimental results showed that, during the time of 120 min, the decolouring ratio of the malachite green was 97.7% when initial concentration of malachite green is 30 mg x L(-1) and initial pH is 3.0. The catalytic degradation of malachite green was a pseudo-first order reaction. In the degradation process of malachite green the azo bond cleavage and the conjugated system of malachite green were attacked by hydroxyl radical. Simultaneity, the aromatic ring was oxidized. Finally, malachite green was degraded into other small molecular compounds.

Determination of uncertainties of PWR spent fuel radionuclide inventory based on real operational history data

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

Fast, Ivan; Bosbach, Dirk; Aksyutina, Yuliya

A requisite for the official approval of the safe final disposal of SNF is a comprehensive specification and declaration of the nuclear inventory in SNF by the waste supplier. In the verification process both the values of the radionuclide (RN) activities and their uncertainties are required. Burn-up (BU) calculations based on typical and generic reactor operational parameters do not encompass any possible uncertainties observed in real reactor operations. At the same time, the details of the irradiation history are often not well known, which complicates the assessment of declared RN inventories. Here, we have compiled a set of burnup calculationsmore » accounting for the operational history of 339 published or anonymized real PWR fuel assemblies (FA). These histories were used as a basis for a 'SRP analysis', to provide information about the range of the values of the associated secondary reactor parameters (SRP's). Hence, we can calculate the realistic variation or spectrum of RN inventories. SCALE 6.1 has been employed for the burn-up calculations. The results have been validated using experimental data from the online database - SFCOMPO-1 and -2. (authors)« less

Validation of the U.S. NRC NGNP evaluation model with the HTTR

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

Saller, T.; Seker, V.; Downar, T.

2012-07-01

The High Temperature Test Reactor (HTTR) was modeled with TRITON/PARCS. Traditional light water reactor (LWR) homogenization methods rely on the short mean free paths of neutrons in LWR. In gas-cooled, graphite-moderated reactors like the HTTR neutrons have much longer mean free paths and penetrate further into neighboring assemblies than in LWRs. Because of this, conventional lattice calculations with a single assembly may not be valid. In addition to difficulties caused by the longer mean free paths, the HTTR presents unique axial and radial heterogeneities that require additional modifications to the single assembly homogenization method. To handle these challenges, the homogenizationmore » domain is decreased while the computational domain is increased. Instead of homogenizing a single hexagonal fuel assembly, the assembly is split into six triangles on the radial plane and five blocks axially in order to account for the placement of burnable poisons. Furthermore, the radial domain is increased beyond a single fuel assembly to account for spectrum effects from neighboring fuel, reflector, and control rod assemblies. A series of five two-dimensional cases, each closer to the full core, were calculated to evaluate the effectiveness of the homogenization method and cross-sections. (authors)« less

AGC-2 Graphite Pre-irradiation Data Package

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

David Swank; Joseph Lord; David Rohrbaugh

2010-08-01

The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterizedmore » prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.« less

Improved Search for a Light Sterile Neutrino with the Full Configuration of the Daya Bay Experiment

NASA Astrophysics Data System (ADS)

An, F. P.; Balantekin, A. B.; Band, H. R.; Bishai, M.; Blyth, S.; Cao, D.; Cao, G. F.; Cao, J.; Cen, W. R.; Chan, Y. L.; Chang, J. F.; Chang, L. C.; Chang, Y.; Chen, H. S.; Chen, Q. Y.; Chen, S. M.; Chen, Y. X.; Chen, Y.; Cheng, J.-H.; Cheng, J.; Cheng, Y. P.; Cheng, Z. K.; Cherwinka, J. J.; Chu, M. C.; Chukanov, A.; Cummings, J. P.; de Arcos, J.; Deng, Z. Y.; Ding, X. F.; Ding, Y. Y.; Diwan, M. V.; Dolgareva, M.; Dove, J.; Dwyer, D. A.; Edwards, W. R.; Gill, R.; Gonchar, M.; Gong, G. H.; Gong, H.; Grassi, M.; Gu, W. Q.; Guan, M. Y.; Guo, L.; Guo, R. P.; Guo, X. H.; Guo, Z.; Hackenburg, R. W.; Han, R.; Hans, S.; He, M.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. X.; Huang, X. T.; Huber, P.; Huo, W.; Hussain, G.; Jaffe, D. E.; Jaffke, P.; Jen, K. L.; Jetter, S.; Ji, X. P.; Ji, X. L.; Jiao, J. B.; Johnson, R. A.; Joshi, J.; Kang, L.; Kettell, S. H.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, J. H. C.; Lei, R. T.; Leitner, R.; Leung, J. K. C.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, S.; Li, S. C.; Li, W. D.; Li, X. N.; Li, Y. F.; Li, Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, S.; Lin, S. K.; Lin, Y.-C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, D. W.; Liu, J. L.; Liu, J. C.; Loh, C. W.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, K. B.; Lv, Z.; Ma, Q. M.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Malyshkin, Y.; Martinez Caicedo, D. A.; McDonald, K. T.; McKeown, R. D.; Mitchell, I.; Mooney, M.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Ngai, H. Y.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pan, H.-R.; Park, J.; Patton, S.; Pec, V.; Peng, J. C.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Steiner, H.; Sun, G. X.; Sun, J. L.; Tang, W.; Taychenachev, D.; Treskov, K.; Tsang, K. V.; Tull, C. E.; Viaux, N.; Viren, B.; Vorobel, V.; Wang, C. H.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Wei, H. Y.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wise, T.; Wong, H. L. H.; Wong, S. C. F.; Worcester, E.; Wu, C.-H.; Wu, Q.; Wu, W. J.; Xia, D. M.; Xia, J. K.; Xing, Z. Z.; Xu, J. Y.; Xu, J. L.; Xu, Y.; Xue, T.; Yang, C. G.; Yang, H.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Ye, Z.; Yeh, M.; Young, B. L.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, H. H.; Zhang, J. W.; Zhang, Q. M.; Zhang, X. T.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhao, Q. W.; Zhao, Y. B.; Zhong, W. L.; Zhou, L.; Zhou, N.; Zhuang, H. L.; Zou, J. H.; Daya Bay Collaboration

2016-10-01

This Letter reports an improved search for light sterile neutrino mixing in the electron antineutrino disappearance channel with the full configuration of the Daya Bay Reactor Neutrino Experiment. With an additional 404 days of data collected in eight antineutrino detectors, this search benefits from 3.6 times the statistics available to the previous publication, as well as from improvements in energy calibration and background reduction. A relative comparison of the rate and energy spectrum of reactor antineutrinos in the three experimental halls yields no evidence of sterile neutrino mixing in the 2 ×10-4≲|Δ m412|≲0.3 eV2 mass range. The resulting limits on sin22 θ14 are improved by approx imately a factor of 2 over previous results and constitute the most stringent constraints to date in the |Δ m412|≲0.2 eV2 region.

Pulsed plasma chemical synthesis of SixCyOz composite nanopowder

NASA Astrophysics Data System (ADS)

Kholodnaya, G.; Sazonov, R.; Ponomarev, D.; Remnev, G.

2017-05-01

SixCyOz composite nanopowder with an average size of particles about 10-50 nm was produced using the pulsed plasma chemical method. The experiments on the synthesis of nanosized composite were carried out using a TEA-500 pulsed electron accelerator. To produce a composite, SiCl4, O2, and CH4 were used. The major part of experiments was conducted using a plasma chemical reactor (quartz, 140 mm diameter, 6 l volume). The initial reagents were injected into the reactor, then a pulsed electron beam was injected which initiated the chemical reactions whose products were the SixCyOz composite nanopowder. To define the morphology of the particles, the JEOL-II-100 transmission electron microscope (TEM) with an accelerating voltage of 100 kV was used. The substances in the composition of the composite nanopowder were identified using the infrared absorption optical spectrum. To conduct this analysis, the Nicolet 5700 FT-IR spectrometer was used.

Analysis of the Daya Bay Reactor Antineutrino Flux Changes with Fuel Burnup

NASA Astrophysics Data System (ADS)

Hayes, A. C.; Jungman, Gerard; McCutchan, E. A.; Sonzogni, A. A.; Garvey, G. T.; Wang, X. B.

2018-01-01

We investigate the recent Daya Bay results on the changes in the antineutrino flux and spectrum with the burnup of the reactor fuel. We find that the discrepancy between current model predictions and the Daya Bay results can be traced to the original measured U 235 /Pu 239 ratio of the fission β spectra that were used as a base for the expected antineutrino fluxes. An analysis of the antineutrino spectra that is based on a summation over all fission fragment β decays, using nuclear database input, explains all of the features seen in the Daya Bay evolution data. However, this summation method still allows for an anomaly. We conclude that there is currently not enough information to use the antineutrino flux changes to rule out the possible existence of sterile neutrinos.

Performance of the electronic personal dosemeter for neutron 'Saphydose-N' at different workplaces of nuclear facilities.

PubMed

Lahaye, T; Chau, Q; Ménard, S; Lacoste, V; Muller, H; Luszik-Bhadra, M; Reginatto, M; Bruguier, P

2006-01-01

This paper mainly aims at presenting the measurements and the results obtained with the electronic personal neutron dosemeter Saphydose-N at different facilities. Three campaigns were led in the frame of the European contract EVIDOS ('Evaluation of Individual Dosimetry in Mixed Neutron and Photon Radiation Fields'). The first one consisted in the measurements at the IRSN French research laboratory in reference neutron fields generated by a thermal facility (SIGMA), radionuclide ISO sources ((241)AmBe; (252)Cf; (252)Cf(D(2)O)\\Cd) and a realistic spectrum (CANEL/T400). The second one was performed at the Krümmel Nuclear Power Plant (Germany) close to the boiling water reactor and to a spent fuel transport cask. The third one was realised at Mol (Belgium), at the VENUS Research Reactor and at Belgonucléaire, a fuel processing factory.

Testing FLUKA on neutron activation of Si and Ge at nuclear research reactor using gamma spectroscopy

NASA Astrophysics Data System (ADS)

Bazo, J.; Rojas, J. M.; Best, S.; Bruna, R.; Endress, E.; Mendoza, P.; Poma, V.; Gago, A. M.

2018-03-01

Samples of two characteristic semiconductor sensor materials, silicon and germanium, have been irradiated with neutrons produced at the RP-10 Nuclear Research Reactor at 4.5 MW. Their radionuclides photon spectra have been measured with high resolution gamma spectroscopy, quantifying four radioisotopes (28Al, 29Al for Si and 75Ge and 77Ge for Ge). We have compared the radionuclides production and their emission spectrum data with Monte Carlo simulation results from FLUKA. Thus we have tested FLUKA's low energy neutron library (ENDF/B-VIIR) and decay photon scoring with respect to the activation of these semiconductors. We conclude that FLUKA is capable of predicting relative photon peak amplitudes, with gamma intensities greater than 1%, of produced radionuclides with an average uncertainty of 13%. This work allows us to estimate the corresponding systematic error on neutron activation simulation studies of these sensor materials.

Experimental physics characteristics of a heavy-metal-reflected fast-spectrum critical assembly

NASA Technical Reports Server (NTRS)

Heneveld, W. H.; Paschall, R. K.; Springer, T. H.; Swanson, V. A.; Thiele, A. W.; Tuttle, R. J.

1972-01-01

A zero-power critical assembly was designed, constructed, and operated for the purpose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power electric conversion system. The range of the previous experimental investigations has been expanded to include the reactivity effects of:(1) surrounding the reactor with 15.24 cm (6 in.) of polyethylene, (2) reducing the heights of a portion of the upper and lower axial reflectors by factors of 2 and 4, (3) adding 45 kg of W to the core uniformly in two steps, (4) adding 9.54 kg of Ta to the core uniformly, and (5) inserting 2.3 kg of polyethylene into the core proper and determining the effect of a Ta addition on the polyethylene worth.

Diamond tool wear detection method using cutting force and its power spectrum analysis in ultra-precision fly cutting

NASA Astrophysics Data System (ADS)

Zhang, G. Q.; To, S.

2014-08-01

Cutting force and its power spectrum analysis was thought to be an effective method monitoring tool wear in many cutting processes and a significant body of research has been conducted on this research area. However, relative little similar research was found in ultra-precision fly cutting. In this paper, a group of experiments were carried out to investigate the cutting forces and its power spectrum characteristics under different tool wear stages. Result reveals that the cutting force increases with the progress of tool wear. The cutting force signals under different tool wear stages were analyzed using power spectrum analysis. The analysis indicates that a characteristic frequency does exist in the power spectrum of the cutting force, whose power spectral density increases with the increasing of tool wear level, this characteristic frequency could be adopted to monitor diamond tool wear in ultra-precision fly cutting.

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

Bakosi, Jozsef; Christon, Mark A.; Francois, Marianne M.

Progress is reported on computational capabilities for the grid-to-rod-fretting (GTRF) problem of pressurized water reactors. Numeca's Hexpress/Hybrid mesh generator is demonstrated as an excellent alternative to generating computational meshes for complex flow geometries, such as in GTRF. Mesh assessment is carried out using standard industrial computational fluid dynamics practices. Hydra-TH, a simulation code developed at LANL for reactor thermal-hydraulics, is demonstrated on hybrid meshes, containing different element types. A series of new Hydra-TH calculations has been carried out collecting turbulence statistics. Preliminary results on the newly generated meshes are discussed; full analysis will be documented in the L3 milestone, THM.CFD.P5.05,more » Sept. 2012.« less

Recent Advances in Understanding and Managing Autism Spectrum Disorders

PubMed Central

Germain, Blair; Eppinger, Melissa A.; Mostofsky, Stewart H.; DiCicco-Bloom, Emanuel; Maria, Bernard L.

2017-01-01

Autism spectrum disorder in children is a group of neurodevelopmental disorders characterized by difficulties with social communication and behavior. Growing scientific evidence in addition to clinical practice has led the Diagnostic and Statistical Manual of Mental Disorders (DSM-5) to categorize several disorders into the broader category of autism spectrum disorder. As more is learned about how autism spectrum disorder manifests, progress has been made toward better clinical management including earlier diagnosis, care, and when specific interventions are required. The 2014 Neurobiology of Disease in Children symposium, held in conjunction with the 43rd annual meeting of the Child Neurology Society, aimed to (1) describe the clinical concerns involving diagnosis and treatment, (2) review the current status of understanding in the pathogenesis of autism spectrum disorder, (3) discuss clinical management and therapies for autism spectrum disorder, and (4) define future directions of research. The article summarizes the presentations and includes an edited transcript of question-and-answer sessions. PMID:26336201

Heavy-section steel irradiation program. Semiannual progress report, October 1996--March 1997

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

Rosseel, T.M.

1998-02-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. Because the RPV is the only key safety-related component of the plant for which a redundant backup system does not exist, it is imperative to fully understand the degree of irradiation-induced degradation of the RPV`s fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established. Its primary goal is to provide a thorough, quantitative assessment of the effects of neutron irradiationmore » on the material behavior and, in particular, the fracture toughness properties of typical pressure-vessel steels as they relate to light-water RPV integrity. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are being examined on a wide range of fracture properties. The HSSI Program is arranged into eight tasks: (1) program management, (2) irradiation effects in engineering materials, (3) annealing, (4) microstructural analysis of radiation effects, (5) in-service irradiated and aged material evaluations, (6) fracture toughness curve shift method, (7) special technical assistance, and (8) foreign research interactions. The work is performed by the Oak Ridge National Laboratory.« less

A summary of sodium-cooled fast reactor development

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

Aoto, Kazumi; Dufour, Philippe; Hongyi, Yang

Much of the basic technology for the Sodium-cooled fast Reactor (SFR) has been established through long term development experience with former fast reactor programs, and is being confirmed by the Phénix end-of-life tests in France, the restart of Monju in Japan, the lifetime extension of BN-600 in Russia, and the startup of the China Experimental Fast Reactor in China. Planned startup in 2014 for new SFRs: BN-800 in Russia and PFBR in India, will further enhance the confirmation of the SFR basic technology. Nowadays, the SFR development has advanced to aiming at establishment of the Generation-IV system which is dedicatedmore » to sustainable energy generation and actinide management, and several advanced SFR concepts are under development such as PRISM, JSFR, ASTRID, PGSFR, BN-1200, and CFR-600. Generation-IV International Forum is an international collaboration framework where various R&D activities are progressing on design of system and component, safety and operation, advanced fuel, and actinide cycle for the Generation-IV SFR development, and will play a beneficial role of promoting them thorough providing an opportunity to share the past experience and the latest data of design and R&D among countries developing SFR.« less