On Use of Multi-Chambered Fission Detectors for In-Core, Neutron Spectroscopy

Presented is a short, computational study on the potential use of multichambered fission detectors for in-core, neutron spectroscopy. Motivated by the development of very small fission chambers at CEA in France and at Kansas State University in the U.S., it was assumed in this preliminary analysis that devices can be made small enough to avoid flux perturbations and that uncertainties related to measurements can be ignored. It was hypothesized that a sufficient number of chambers with unique reactants can act as a real-time, foilactivation experiment. An unfolding scheme based on maximizing (Shannon) entropy was used to produce a flux spectrum from detector signals that requires no prior information. To test the method, integral, detector responses were generated for singleisotope detectors of various Th, U, Np, Pu, Am, and Cs isotopes using a simplified, pressurized-water reactor spectrum and fluxweighted, microscopic, fission cross sections, in the WIMS-69 multigroup format. An unfolded spectrum was found from subsets of these responses that had a maximum entropy while reproducing the responses considered and summing to one (that is, they were normalized). Several nuclide subsets were studied, and, as expected, the results indicate inclusion of more nuclides leads to better spectra but with diminishing improvements, with the best-case spectrum having an average, relative, group-wise error of approximately 51%. Furthermore, spectra found from minimum-norm and Tihkonov-regularization inversion were of lower quality than the maximum entropy solutions. Finally, the addition of thermal-neutron filters (here, Cd and Gd) provided substantial improvement over unshielded responses alone. The results, as a whole, suggest that in-core, neutron spectroscopy is at least marginally feasible.

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Analyses of the OSU-MASLWR Experimental Test Facility

Science and Technology of Nuclear Installations ◽

10.1155/2012/528241 ◽

2012 ◽

Vol 2012 ◽

pp. 1-19 ◽

Cited By ~ 6

Author(s):

F. Mascari ◽

G. Vella ◽

B. G. Woods ◽

F. D'Auria

Keyword(s):

Natural Circulation ◽

Developed Countries ◽

System Level ◽

Test Facility ◽

State University ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Standard Problem ◽

Oregon State University

Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU) has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR) design, a small modular pressurized water reactor (PWR), relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and to describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility. A summary of the best estimate thermal hydraulic system code analyses, already performed, to analyze the codes capability in predicting the phenomena typical of the MASLWR prototype, thermal hydraulically characterized in the OSU-MASLWR facility, is presented as well.

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Comparison of Countercurrent Flow Limitation Experiments Performed in Two Different Models of the Hot Leg of a Pressurized Water Reactor With Rectangular Cross Section

Journal of Engineering for Gas Turbines and Power ◽

10.1115/1.4002884 ◽

2010 ◽

Vol 133 (5) ◽

Cited By ~ 6

Author(s):

Christophe Vallée ◽

Tobias Seidel ◽

Dirk Lucas ◽

Akio Tomiyama ◽

Michio Murase

Keyword(s):

Cross Section ◽

Flow Rate ◽

Cross Sections ◽

Characteristic Length ◽

Rectangular Cross Section ◽

Channel Height ◽

Countercurrent Flow ◽

Flow Limitation ◽

Pressurized Water Reactor ◽

Pressurized Water

In order to investigate the two-phase flow behavior during countercurrent flow limitation in the hot leg of a pressurized water reactor, two test models were built: one at the Kobe University and the other at the TOPFLOW test facility of Forschungszentrum Dresden-Rossendorf (FZD). Both test facilities are devoted to optical measurement techniques; therefore, a flat hot leg test section design was chosen. Countercurrent flow limitation (CCFL) experiments were performed, simulating the reflux condenser cooling mode appearing in some accident scenarios. The fluids used were air and water, both at room temperature. The pressure conditions were varied from atmospheric at Kobe to 3.0 bars absolute at TOPFLOW. According to the presented review of literature, very few data are available on flooding in channels with a rectangular cross section, and no experiments were performed in the past in such flat models of a hot leg. Commonly, the macroscopic effects of CCFL are represented in a flooding diagram, where the gas flow rate is plotted versus the discharge water flow rate, using the nondimensional superficial velocity (also known as Wallis parameter) as coordinates. However, the classical definition of the Wallis parameter contains the pipe diameter as characteristic length. In order to be able to perform comparisons with pipe experiments and to extrapolate to the power plant scale, the appropriate characteristic length should be determined. A detailed comparison of the test facilities operated at the Kobe University and at FZD is presented. With respect to the CCFL behavior, it is shown that the essential parts of the two hot leg test sections are very similar. This geometrical analogy allows us to perform meaningful comparisons. However, clear differences in the dimensions of the cross section (H×W=150×10 mm2 in Kobe, 250×50 mm2 at FZD) make it possible to point out the right characteristic length for hot leg models with rectangular cross sections. The hydraulic diameter, the channel height, and the Laplace critical wavelength (leading to the Kutateladze number) were tested. A comparison of our own results with similar experimental data and empirical correlations for pipes available in literature shows that the channel height is the characteristic length to be used in the Wallis parameter for channels with rectangular cross sections. However, some limitations were noticed for narrow channels, where CCFL is reached at lower gas fluxes, as already observed in small scale hot legs with pipe cross sections.

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The RELAP5 Simulation of 2″ LOCA of APEX Facility

18th International Conference on Nuclear Engineering: Volume 3 ◽

10.1115/icone18-29900 ◽

2010 ◽

Cited By ~ 1

Author(s):

Heng Xie

Keyword(s):

Test Facility ◽

State University ◽

Pressurized Water Reactor ◽

Reduced Pressure ◽

Pressurized Water ◽

Sequence Of Events ◽

Comparison Results ◽

Integral Systems ◽

Oregon State University ◽

Good Agreement

The RELAP5/SCDAP Mod3.2(am5) code is employed to simulate the OSU-AP1000-05 test conducted in the Advanced Plant Experimental (APEX) test facility at Oregon State University (OSU). The APEX-1000 test facility is an one-fourth height, one-half time scale, and reduced pressure integral systems facility to simulate the Westinghouse Advanced Passive 1000 MW (AP1000) pressurized water reactor. OSU-AP1000-05 is a two-inch break at the bottom of cold leg #4 with 3 out of 4 ADS-4 valves of OSU-APEX-1000 facility. RELAP5 predictions are compared to the experimental data generated by the test. The comparison shows good agreement between the predicted and measured sequence of events of some key parameters during the transient. From the comparison results, it could be preliminary concluded that the RELAP5/SCDAP Mod3.2(am5) code are suitable to simulate the small LOCA of APEX.

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Effective Diffusion Homogenization of Cross Sections for Pressurized Water Reactor Core Calculations

Nuclear Science and Engineering ◽

10.13182/nse94-a21485 ◽

1994 ◽

Vol 116 (2) ◽

pp. 86-95 ◽

Cited By ~ 6

Author(s):

A. Trkov ◽

M. Ravnik

Keyword(s):

Cross Sections ◽

Effective Diffusion ◽

Reactor Core ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor

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DEVELOPMENT OF THE MPACT 69-GROUP LIBRARY FOR MAGNOX REACTOR ANALYSIS USING CASL VERA

EPJ Web of Conferences ◽

10.1051/epjconf/202124710016 ◽

2021 ◽

Vol 247 ◽

pp. 10016

Author(s):

Kang Seog Kim ◽

Brian J. Ade ◽

Nicholas P. Luciano

Keyword(s):

Cross Section ◽

Cross Sections ◽

Group Structure ◽

Homogenization Method ◽

Light Water Reactors ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Neutron Leakage ◽

Water Reactors ◽

Neutronic Characteristics

The Consortium for Advanced Simulation of Light Water Reactors (CASL) has developed the CASL toolset, Virtual Environment for Reactor Analysis (VERA), for pressurized water reactor (PWR) analysis. Recently the CASL VERA was improved for Magnox reactor analysis, which required the development of a new cross section library and new geometrical and thermal feedback capabilities for graphite-moderated Magnox reactors. The MPACT neutronics module of the CASL core simulator is a 3D whole core transport code, which requires a new cross section library with a different energy group structure due to the different neutronic characteristics of Magnox compared with PWR. A new 69-group structure was developed based on the MPACT 51-group structure to have more thermal energy groups and to be a subset of the SCALE 252-group structure. The ENDF/B-VII.1 MPACT 69-group library was developed for Magnox reactor analysis using the SCALE/AMPX and VERA-XSTools for which a super-homogenization method was applied, and transport cross sections were generated for graphite using a neutron leakage conservation method. Benchmark results show that new MPACT 69-group library works reasonably well for Magnox reactor analysis.

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Uncertainty Analyses Applied to the UAM/TMI-1 Lattice Calculations Using the DRAGON (Version 4.05) Code and Based on JENDL-4 and ENDF/B-VII.1 Covariance Data

Science and Technology of Nuclear Installations ◽

10.1155/2013/437854 ◽

2013 ◽

Vol 2013 ◽

pp. 1-21 ◽

Cited By ~ 2

Author(s):

Augusto Hernández-Solís ◽

Christophe Demazière ◽

Christian Ekberg

Keyword(s):

Uncertainty Analysis ◽

Covariance Matrix ◽

Cross Sections ◽

Nuclear Data ◽

Uncertainty Assessment ◽

Test Case ◽

Pressurized Water Reactor ◽

Cross Sectional ◽

Pressurized Water ◽

Water Reactor

The OECD/NEA Uncertainty Analysis in Modeling (UAM) expert group organized and launched the UAM benchmark. Its main objective is to perform uncertainty analysis in light water reactor (LWR) predictions at all modeling stages. In this paper, multigroup microscopic cross-sectional uncertainties are propagated through the DRAGON (version 4.05) lattice code in order to perform uncertainty analysis on and 2-group homogenized macroscopic cross-sections. The chosen test case corresponds to the Three Mile Island-1 (TMI-1) lattice, which is a 15 15 pressurized water reactor (PWR) fuel assembly segment with poison and at full power conditions. A statistical methodology is employed for the uncertainty assessment, where cross-sections of certain isotopes of various elements belonging to the 172-group DRAGLIB library format are considered as normal random variables. Two libraries were created for such purposes, one based on JENDL-4 data and the other one based on the recently released ENDF/B-VII.1 data. Therefore, multigroup uncertainties based on both nuclear data libraries needed to be computed for the different isotopic reactions by means of ERRORJ. The uncertainty assessment performed on and macroscopic cross-sections, that is based on JENDL-4 data, was much higher than the assessment based on ENDF/B-VII.1 data. It was found that the computed Uranium 235 fission covariance matrix based on JENDL-4 is much larger at the thermal and resonant regions than, for instance, the covariance matrix based on ENDF/B-VII.1 data. This can be the main cause of significant discrepancies between different uncertainty assessments.

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A New Method to Evaluate the Correlation Coefficients Between Nuclides in Uncertainty and Sensitivity Analysis

Volume 3: Nuclear Fuel and Material, Reactor Physics and Transport Theory; Innovative Nuclear Power Plant Design and New Technology Application ◽

10.1115/icone25-66777 ◽

2017 ◽

Author(s):

Xu Jia Yi ◽

Ma Xu Bo ◽

Shen Jing Wen ◽

Liu Jia Yi ◽

Chen Yi Xue

Keyword(s):

Sensitivity Analysis ◽

Correlation Coefficient ◽

Cross Section ◽

Cross Sections ◽

Correlation Coefficients ◽

Atomic Density ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Uncertainty And Sensitivity Analysis ◽

Cross Section Uncertainty

Uncertainty and sensitivity analysis is an essential component of nuclear engineering calculations. Uncertainties in the cross-section input data directly affect uncertainties in the results. The covariance values between different types of cross-sections are considered in the NJOY covariance library. However, the correlation coefficient between isotopes can depend on the specific problem. The correlation coefficient between 235U and 238U in a pressurized water reactor (PWR) might be different from that in a fast reactor. In this study, a new Monte Carlo-based method is proposed for calculating this effect. The correlation coefficients between different isotopes are calculated using a problem-dependent fraction parameter. The correlation coefficients between the capture cross-sections of 235U, 238U, 239Pu, and 241Pu are calculated. The same method can be extended to other reaction types. The correlation coefficients as a function of the isotopic atomic density uncertainty and the average one-group microscopic cross-section uncertainty are also studied. It is shown that the correlation coefficients vary very little with the uncertainty in the average one-group microscopic cross-section. The correlation coefficient of an isotope pair changes slightly over the course of a cycle because of atomic density and microscopic cross-section changes.

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Leaching Studies of Cement-Glass Package Containing Sodium Borate

MRS Proceedings ◽

10.1557/proc-84-393 ◽

1986 ◽

Vol 84 ◽

Cited By ~ 3

Author(s):

Masahiro Okamoto ◽

Koichi Chino ◽

Tsutomu Baba ◽

Tatsuo Izumida ◽

Fumio Kawamura ◽

...

Keyword(s):

Sodium Silicate ◽

Liquid Waste ◽

Sodium Borate ◽

Pressurized Water Reactor ◽

X Ray Diffraction ◽

Adsorption Ability ◽

Pressurized Water ◽

X Ray ◽

Solidification Reaction ◽

Main Components

AbstractA new solidification technique using cement-glass, which is a mixture of sodium silicate, cement, additives, and initiator of the solidification reaction, was developed for sodium borate liquid waste generated from pressurized water reactor (PWR) plants. The cement-glass could solidify eight times as much sodium borate as cement could, because the solidifying reaction of the cement-glass is not hindered by borate ions.The reaction mechanism of sodium silicate and phosphoric silicate (initiator), the main components of cement-glass, was studied through X-ray diffraction and compressive strength measurements. It was found that three- dimensionally bonded silicon dioxide was produced by polymerization of the two silicates. The leaching ratio of cesium from the cement-glass package was one-tenth that of the cement one. This low value was attributed to a high cesium adsorption ability of the cement-glass and it could be theoretically predicted accordingly.

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