Development of a Fast-Spectrum Self-Powered Neutron Detector for Molten Salt Experiments in the Versatile Test Reactor

The self-powered neutron detector (SPND) is a widely used flux monitor in thermal nuclear reactors. Although this is a mature technology, the current state of the art is tuned for a thermal neutron spectrum, so many of the devices currently in use lack sensitivity to fast neutrons. Because current in SPNDs is produced through nuclear reactions with the neutron flux inside a reactor, sensitivity in SPNDs is determined by the neutron cross section of the neutron-sensitive portion of the detector, termed the emitter. This neutron cross section drops by orders of magnitude between thermal and fast neutron energies for many emitters in currently used SPNDs, with a corresponding drop in current from the detector. This paper discusses efforts to develop a fast-spectrum self-powered neutron detector (FS-SPND) that is sensitive to neutrons with energies ranging from 0.025 eV up to 1 MeV. An in-depth analysis of Evaluated Nuclear Data File (ENDF)/B-VII.1 neutron-capture cross sections was performed, and four new materials were identified that are suitable emitter candidates for use in measuring fast neutrons. All four materials are stable mid-shell nuclei in the region between doubly magic 132Sn and 208Pb. Each candidate was simulated with the Geant4 Monte Carlo simulation toolkit to optimize overall detector efficiency.

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UKE-III: a computer program for translating neutron cross section data from the UKAEA Nuclear Data Library to the Evaluated Nuclear Data File format

10.2172/4415855 ◽

1973 ◽

Author(s):

R. Q. Wright ◽

S. N. Cramer ◽

D. C. Irving

Keyword(s):

Cross Section ◽

Neutron Cross Section ◽

Nuclear Data ◽

Data File ◽

File Format ◽

Cross Section Data ◽

Section Data ◽

Nuclear Data Library ◽

Data Library ◽

Evaluated Nuclear Data

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Cross Section and Fission Yields Induced Uncertainty in the Vver-440 Burnup Calculation

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4051613 ◽

2021 ◽

Author(s):

Branislav Vrban ◽

Stefan Cerba ◽

Jakub Luley ◽

Filip Osuský ◽

Vladimir Necas

Keyword(s):

Cross Section ◽

Neutron Cross Section ◽

Nuclear Data ◽

Reactor Operation ◽

Precise Estimation ◽

Stochastic Techniques ◽

Core Characteristics ◽

Fission Yields ◽

Operational Data ◽

Burnup Calculation

Abstract The properties of nuclear fuel depend on the actual isotopic composition which develops during a reactor operation. In practice, the prediction accuracy of burnup calculations serves as the basis for the future precise estimation of a core lifetime and other safety-based core characteristics. The present study quantifies nuclear data induced uncertainties of nuclide concentrations and multiplication factors in VVER-440 fuel depletion analysis. The well-known SCALE system and the TRITON sequence are used with the NEWT deterministic solver in the SAMPLER module that implements stochastic techniques to assess the uncertainty in computed results. The propagation of uncertainties in neutron cross section and fission yields is studied through the depletion calculation of 2D heterogeneous VVER-440 fuel assembly with an average enrichment of 4.87 wt % of 235U and six gadolinium rods with 3.35 % of Gd2O3. In the paper, fixed nominal depletion conditions are based on the real operational data of the Slovak NPP Bohunice unit 4 during cycle 30. In total 250 cases with uncertain parameters are computed and the results are evaluated by an auxiliary tool.

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New Neutron Cross-Section Measurements at ORELA for Improved Nuclear Data Calculations

10.1063/1.1945338 ◽

2005 ◽

Cited By ~ 2

Author(s):

K. H. Guber

Keyword(s):

Cross Section ◽

Neutron Cross Section ◽

Nuclear Data

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Effects of different nuclear evaluation data on the RMC keff calculation

EPJ Web of Conferences ◽

10.1051/epjconf/202023919005 ◽

2020 ◽

Vol 239 ◽

pp. 19005

Author(s):

Zhang Wenxin ◽

Qiang shenglong ◽

Yin qiang ◽

Cui Xiantao

Keyword(s):

Cross Section ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Reactor ◽

Neutron Cross Section ◽

Nuclear Data ◽

Section Data ◽

Calculation Results ◽

The Impact ◽

Physical Calculation

Neutron cross section data is the basis of nuclear reactor physical calculation and has a decisive influence on the accuracy of calculation results. AFA3Gassemble is widely used in nuclear power plants. CENACE is an ACE format multiple-temperature continuous energy cross section library that developed by China Nuclear Data Centre. In this paper, we calculated the AFA3G assemble by RMC.We respectively used ENDF6.8/, ENDF/7 and CENACE data for calculation. The impact of nuclear data on RMC calculation is studied by comparing the results of different nuclear data.

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The Development and Validation of Nuclear Cross Section Processing Code for Reactor-RXSP

Volume 5: Fuel Cycle, Radioactive Waste Management and Decommissioning; Reactor Physics and Transport Theory; Nuclear Education, Public Acceptance and Related Issues; Instrumentation and Controls; Fusion Engineering ◽

10.1115/icone21-15442 ◽

2013 ◽

Cited By ~ 1

Author(s):

Jiankai Yu ◽

Songyang Li ◽

Kan Wang ◽

Guanbo Wang ◽

Ganglin Yu

Keyword(s):

Cross Section ◽

Nuclear Data ◽

Original Data ◽

Data File ◽

Cross Section Data ◽

Reactor Physics ◽

R Matrix ◽

Section Data ◽

Nuclear Cross Section ◽

Long Time

The accuracy of the nuclear cross section data is a prerequisite for the accuracy of reactor physics calculations. The RXSP(Reactor Cross Section Processing Code) which is developed by REAL (Reactor Engineering Analysis Laboratory) of Department of Engineering Physics in Tsinghua University, has changed the situation in China that nuclear cross section processing has been dependent of NJOY for a long time. The key methods such as fast Doppler broadening, thermal libraries interpolation, and OpenMP parallel acceleration, can be achieved with RXSP. This code is able to process the original data of ENDF/B (Evaluated Nuclear Data File/B) efficiently and accurately to produce the continuous energy point cross section data which is necessary for RMC. By comparing with NJOY, The microscopic and macroscopic verification shows that RXSP has the same accuracy as NJOY while RXSP has saved greatly the processing time to meet the efficient demand in the frequent reactor physics-thermal-hydraulic coupling calculations to solve the complex questions related on a large number of materials and temperature. In addition, RXSP make it available to process the resonance parameters of the R-matrix Limited format.

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Temperature Dependence of the Westcott g-factor for Neutron Reactions in Activation Analysis

Pure and Applied Chemistry ◽

10.1351/pac199971122309 ◽

1999 ◽

Vol 71 (12) ◽

pp. 2309-2315 ◽

Cited By ~ 47

Author(s):

N. E. Holden

Keyword(s):

Cross Section ◽

Neutron Capture ◽

Nuclear Reactions ◽

Nuclear Data ◽

Reaction Rates ◽

Nuclear Data Library ◽

Neutron Reactions ◽

Using Data ◽

Data Library ◽

G Factors

The Westcott g-factors, which allow the user to determine reaction rates for nuclear reactions taking place at various temperatures, have been calculated using data from the Evaluated Neutron Nuclear Data Library, ENDF/B-VI. Nuclides chosen have g-factors which are significantly different from unity and result in different reaction rates compared to nuclides whose neutron capture cross section varies as the reciprocal of the neutron velocity. Values are presented as a function of temperature up to 673.16 K (400 °C).

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Processing and benchmarking of evaluated nuclear data file/b-viii.0β4 cross-section library by analysis of a series of critical experimental benchmark using the monte carlo code MCNP(X) and NJOY2016

Nuclear Engineering and Technology ◽

10.1016/j.net.2017.08.017 ◽

2017 ◽

Vol 49 (8) ◽

pp. 1610-1616 ◽

Cited By ~ 3

Author(s):

Kabach Ouadie ◽

Chetaine Abdelouahed ◽

Jalil Abdelhamid ◽

Darif Abdelaziz ◽

Saidi Abdelmajid

Keyword(s):

Monte Carlo ◽

Cross Section ◽

Nuclear Data ◽

Data File ◽

Monte Carlo Code ◽

Evaluated Nuclear Data

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Investigation of fast neutron shielding properties of new polyurethane-based composites loaded with B4C, BeO, WO3, ZnO, and Gd2O3 micro-and nanoparticles

Polish Journal of Medical Physics and Engineering ◽

10.2478/pjmpe-2019-0028 ◽

2019 ◽

Vol 25 (4) ◽

pp. 211-219 ◽

Cited By ~ 1

Author(s):

Asghar Mesbahi ◽

Khatibeh Verdipoor ◽

Farhad Zolfagharpour ◽

Abdolali Alemi

Keyword(s):

Cross Section ◽

Cross Sections ◽

Neutron Cross Section ◽

Radiation Shielding ◽

Fast Neutrons ◽

Nano Composites ◽

Monte Carlo Code ◽

Neutron Shielding ◽

Shielding Properties ◽

Polyurethane Matrix

Abstract The aim of the current research was to study the radiation shielding properties of polyurethane-based shielding materials filled with B4C, BeO, WO3, ZnO, and Gd2O3 particles against fast neutrons. The macroscopic cross sections of composites containing micro- and nanoparticles with a diameter of 10 µm and 50 nm were calculated using MCNPX (2.6.0) Monte Carlo code. The results showed that adding nano-scaled fillers to polyurethane matrix increases attenuation properties of neutron shields compared to micro-scaled fillers for intermediate and fast neutrons. Among the studied composites, WO3 and Gd2O3 nano-composites presented higher neutron cross section compared to others.

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Neutron Field Shaping Using Graphite for Reaction Rate Measurements

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4049725 ◽

2021 ◽

Vol 7 (2) ◽

Author(s):

Mikita Sobaleu ◽

Michal Košťál ◽

Jan Šimon ◽

Evžen Losa

Keyword(s):

Cross Section ◽

Reaction Rate ◽

Nuclear Data ◽

Data File ◽

Reaction Rates ◽

Reaction Cross ◽

Neutron Field ◽

Pure Graphite ◽

24Na Reaction ◽

Data Libraries

Abstract Neutron field shaping is the suitable method for validation of cross section in various energy regions. By increasing the share of neutrons of a certain energy interval and decreasing the share of other, a reaction becomes more sensitive to selected neutrons. As a result, reaction cross section can be validated in selected energy regions more precisely. The shaping can be carried out by both neutron filters which are materials with high absorption in some energy region, or by diffusion material changing the shape of neutron spectra by means of slowing down process. In the presented experiments, the neutron field of the light reactor 0 (LR-0) research reactor was shaped by both using graphite blocks inserted into the core and Cd cladding for increasing the epithermal reaction rate share in total reaction rates. The calculations were carried out with the Monte Carlo N-Particle Transport Code 6 (MCNP6) code and the most recent nuclear data libraries. The results in the pure graphite neutron field are in good agreement; in case of Cd cladding, significant discrepancies were reported. In case of the 23Na(n,γ)24Na reaction, overestimation by about 14% was reached in International Reactor Dosimetry and Fusion File (IRDFF-II), results in other libraries are comparable. In case of 58Fe(n,γ)59Fe, the overestimation as high as 18% is reported in IRDFF-II. For 64Zn(n,γ)65Zn reasonable agreement was reached in evaluated nuclear data file (ENDF/B-VIII), where discrepancies in pure graphite neutron field or in case of Cd cladding are about 10–15%.

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