Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0 and JEFF-3.3 Libraries for the Nuclear Design Calculations of the NPP Krško with the CORD-2 System

2021 ◽  
Author(s):  
Marjan Kromar ◽  
Bojan Kurincic
Keyword(s):  
Cross Sections ◽  
Fission Products ◽  
Nuclear Data ◽  
Isothermal Temperature ◽  
Nuclear Systems ◽  
Full Power ◽  
Design Calculations ◽  
Data Libraries ◽  
Control Rods ◽  
Operating Cycles

Abstract Recently, two new nuclear reaction data evaluations have been released: ENDF/B-VIII.0 and JEFF-3.3. Since the neutron nuclear data profoundly influence predictions of the nuclear systems behavior, many researchers have been investigating new data striving for more accurate predictions. The purpose of this study is to examine the effects of the neutron data libraries on the nuclear design calculations of the NPP Krško core. ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0 and JEFF-3.3 libraries are considered. In the first part of the paper the effect on the depletion of the typical NPP Krško fuel assembly in infinite geometry is investigated. In the second part, analysis of all 30 completed NPP Krško operating cycles is performed. Performed analysis has indicated differences of a few hundred pcm in multiplication factor for a fresh fuel due to differences in 235U cross sections. For a burned fuel assemblies, differences are even higher due to different rate of fission products formation, 235U burnout and Pu production. Observed differences in libraries resulted in differences of several tens of ppm in critical Boron concentration on the core level. Differences in control rods worth and Boron coefficients were inside 1 %. Some differences in isothermal temperature coefficient were observed, however they only marginally affect core power defect going from zero to full power.

scholarly journals Neutronics of Lead and Bismuth

2021 ◽  
pp. 177-213
Author(s):  
Cheol Ho Pyeon
Keyword(s):  
Numerical Simulations ◽  
Cross Section ◽  
Cross Sections ◽  
Nuclear Data ◽  
Design Parameters ◽  
Integral Parameter ◽  
Considerable Impact ◽  
Test Materials ◽  
Design Calculations ◽  
Data Libraries

AbstractCross-section uncertainties of Pb and Bi isotopes could consequently affect the precision of nuclear design calculations of preliminary analyses, before the actual operation of upcoming ADS, since Pb and Bi are composed partly of coolant material (lead-bismuth eutectic: LBE) in ADS facilities. The main characteristics of LBE in ADS are recognized as follows: chemically inactive; high boiling point mechanically; excellent neutron economy caused by large scattering cross sections. From the viewpoint of neutronics, LBE exerts considerable impact on nuclear design parameters for numerical simulations of neutron interactions of Pb and Bi isotopes. As a suitable way of investigating cross-section uncertainties, sample reactivity worth measurements in critical states are considered effective with the use of reference and test materials in a zero-power state, such as a critical assembly, because integral parameter information on cross sections of test materials can be acquired experimentally. For the required experimental study on Pb and Bi nuclear data uncertainties, the sample reactivity worth experiments are carried out at the KUCA core by the substitution of reference (aluminum) for test (Pb or Bi) materials, and numerical simulations are performed with stochastic and deterministic calculation codes together with major nuclear data libraries.

scholarly journals IMPACT OF NUCLEAR DATA LIBRARIES ON PREDICTED FAST REACTOR PERFORMANCE

2021 ◽  
Vol 247 ◽  
pp. 09026
Author(s):  
A.G. Nelson ◽  
K.M. Ramey ◽  
F. Heidet
Keyword(s):  
Cross Sections ◽  
Fast Reactor ◽  
Power Distribution ◽  
Evaluation Process ◽  
Nuclear Data ◽  
Reactor Performance ◽  
Data Set ◽  
Experimental Database ◽  
Data Libraries ◽  
Nuclear Data Libraries

The nuclear data evaluation process inherently yields a nuclear data set designed to produce accurate results for the neutron energy spectra corresponding to a specific benchmark suite of experiments. When studying reactors with spectral conditions outside of, or not well represented by, the experimental database used to evaluate the nuclear data, care should be given to the relevance of the nuclear data used. In such cases, larger biases or uncertainties may be present than in a reactor with well-represented spectra. The motivation of this work is to understand the magnitude of differences between recent nuclear data libraries to provide estimates for expected variability in criticality and power distribution results for sodiumcooled, steel-reflected, metal-fueled fast reactor designs. This work was specifically performed by creating a 3D OpenMC model of a sodium-cooled, steel-reflected, metal-fueled fast reactor similar to the FASTER design but without a thermal test region. This OpenMC model was used to compare the differences in eigenvalues, reactivity coefficients, and the spatial and energetic effects on flux and power distributions between the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.2, and JEFF-3.3 nuclear data libraries. These investigations have revealed that reactivity differences between the above libraries can vary by nearly 900 pcm and the fine-group fluxes can vary by up to 18% in individual groups. Results also show a strong variation in the flux and power distributions near the fuel/reflector interface due to the high variability in the 56Fe cross sections in the libraries examined. This indicates that core design efforts of a sodium-cooled, steel-reflected, metalfueled reactor will require the application of relatively large nuclear data uncertainties and/or the development of a representative benchmark-quality experiment.

scholarly journals Developments in New Measurements of Fission Cross-Sections, Fragment Yields, and Prompt and Quasi-Prompt Gammas for Nuclear Data Needs

2020 ◽  
Vol 242 ◽  
pp. 01002
Author(s):  
Adam Hecht ◽  
Phoenix Baldez ◽  
Baldez Baldez
Keyword(s):  
Cross Sections ◽  
Ionization Chamber ◽  
Gamma Ray ◽  
Fission Products ◽  
Product Yield ◽  
Nuclear Data ◽  
University Of New Mexico ◽  
Alpha Emission ◽  
Time Projection ◽  
The University

The University of New Mexico Fission Spectrometer was developed to measure fission product yield, as part of the LANL SPIDER collaboration. The spectrometer operates as an E-v detector to extract product mass event-by-event, with a time of flight region followed by an ionization chamber for kinetic energy measurements. By using the ionization chamber as a singlecathode/single-anode time projection chamber, stopping power and thus Z information is extracted, for coupled A and Z measurements. New work is being performed to add gamma ray detectors in the data stream, placed near the target region for prompt gammas and near the ionization chamber for quasiprompt (>50 ns) and later gammas, correlated with individual fission products. A stand-alone parallel plate ionization chamber (PPIC) is also being developed for fission tagging gamma ray data. The PPIC will also allow discrimination between charged particle out events and (n,n’γ), and discriminate between alpha emission and fission. Using layers in the PPIC, other targets can be measured simultaneously with a calibration target, giving relative fission cross sections. Past measurements with the spectrometer were performed at LANSCE and we plan to continue measurements there. The current work is supported by the NNSA Stewardship Science Academic Alliance.

scholarly journals MEASUREMENT, EVALUATION AND VALIDATION OF MOLYBDENUM CROSS SECTIONS

2021 ◽  
Vol 247 ◽  
pp. 09007
Author(s):  
Isabelle Duhamel ◽  
Nicolas Leclaire ◽  
Luiz Leal ◽  
Atsushi Kimura ◽  
Shoji Nakamura
Keyword(s):  
Cross Sections ◽  
Path Length ◽  
Measurement Instrument ◽  
Reduction Process ◽  
Nuclear Data ◽  
Flight Path ◽  
Proton Accelerator ◽  
Reactor Physics ◽  
Safety Issues ◽  
Data Libraries

Available nuclear data for molybdenum included in the nuclear data libraries are not of sufficient quality for reactor physics or criticality safety issues and indeed information about uncertainties and covariance is either missing or leaves much to be desired. Therefore, IRSN and JAEA performed experimental measurements on molybdenum at the J-PARC (Japan Proton Accelerator Research Complex) facility in Japan. The aim was to measure capture cross section and transmission of natural molybdenum at the ANNRI (Accurate Neutron-Nucleus Reaction measurement Instrument) in the MLF (Material Life and science Facility) of J-PARC. The measurements were performed on metallic natural molybdenum samples with various thicknesses. A NaI detector, placed at a flight-path length of about 28 m, was used for capture measurements and a Li-glass detector (flight-path length of about 28.7 m) for transmission measurements. Following the data reduction process, the measured data are being analyzed and evaluated to produce more accurate cross sections and associated uncertainties.

scholarly journals Evaluating nuclear data and their uncertainties

2018 ◽  
Vol 4 ◽  
pp. 29
Author(s):  
Patrick Talou
Keyword(s):  
Experimental Data ◽  
Uncertainty Quantification ◽  
Cross Sections ◽  
Coming Out ◽  
Nuclear Data ◽  
Covariance Matrices ◽  
Alternative Approaches ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Evaluated Nuclear Data

In the last decade or so, estimating uncertainties associated with nuclear data has become an almost mandatory step in any new nuclear data evaluation. The mathematics needed to infer such estimates look deceptively simple, masking the hidden complexities due to imprecise and contradictory experimental data and natural limitations of simplified physics models. Through examples of evaluated covariance matrices for the soon-to-be-released U.S. ENDF/B-VIII.0 library, e.g., cross sections, spectrum, multiplicity, this paper discusses some uncertainty quantification methodologies in use today, their strengths, their pitfalls, and alternative approaches that have proved to be highly successful in other fields. The important issue of how to interpret and use the covariance matrices coming out of the evaluated nuclear data libraries is discussed.

NEUTRONIC PERFORMANCE OF (U, Pu)C FUEL IN A LATTICE OF GFR USING SCALE 6.0

2016 ◽  
Vol 1814 ◽  
Author(s):  
A. A. P. Macedo ◽  
Carlos E. Velasquez ◽  
C. A. M. da Silva ◽  
C. Pereira
Keyword(s):  
Computer Analysis ◽  
Fast Reactor ◽  
Nuclear Data ◽  
Multiplication Factor ◽  
Assembly Model ◽  
Fast Reactors ◽  
Full Power ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Neutronic Performance

ABSTRACTThis paper studies the performance of (U, Pu)C fuel in a hexagonal assembly of a GFR (Gas Fast Reactor). The SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation version 6.0) code was used in the calculation. The goal is to evaluate the behavior of the infinite multiplication factor (kinf) for a heterogeneous assembly model using four nuclear data libraries: V6-238, V7-238, ENDF/B-VI.8 and ENDF/B-VII.0. The burnup of (U, Pu)C was performed by the TRITON-6 module, and the isotopic concentrations were evaluated during the cycle. The present work comprises calculations at Zero Power and Full Power condition. This study intends to achieve more information about different Fast Reactors.

Application of Artificial Neural Networks to Reliable Nuclear Data for Nonproliferation Modeling and Simulation

2020 ◽  
pp. 833-844
Author(s):  
Pola Lydia Lagari ◽  
Vladimir Sobes ◽  
Miltiadis Alamaniotis ◽  
Lefteri H. Tsoukalas
Keyword(s):  
Modeling And Simulation ◽  
Cross Sections ◽  
Experimental Testing ◽  
Detection Limits ◽  
Nuclear Data ◽  
Simulation Techniques ◽  
Detection And Identification ◽  
The Us ◽  
Nuclear Systems ◽  
Minimum Detection Limits

Detection and identification of special nuclear materials (SNMs) are an essential part of the US nonproliferation effort. Modern cutting-edge SNM detection methodologies rely more and more on modeling and simulation techniques. Experiments with radiological samples in realistic configurations, is the ultimate tool that establishes the minimum detection limits of SNMs in a host of different geometries. Modern modeling and simulation approaches have the potential to significantly reduce the number of experiments with radioactive sources needed to determine these detection limits and reduce the financial barrier to SNM detection. Unreliable nuclear data is one of the principal causes of uncertainty in modeling and simulating nuclear systems. In particular, nuclear cross sections introduce a significant uncertainty in the nuclear data. The goal of this research is to develop a methodology that will autonomously extract the correct nuclear resonance characteristics of experimental data in a reliable way, a task previously left to expert judgement. Accurate nuclear data will in turn allow contemporary modeling and simulation to become far more reliable, de-escalating the extent of experimental testing. Consequently, modeling and simulation techniques reduce the use and distribution of radiological sources, while at the same time increase the reliability of the currently used methods for the detection and identification of SNMs.

Application of Artificial Neural Networks to Reliable Nuclear Data for Nonproliferation Modeling and Simulation

2016 ◽  
Vol 4 (4) ◽  
pp. 54-64 ◽  
Author(s):  
Pola Lydia Lagari ◽  
Vladimir Sobes ◽  
Miltiadis Alamaniotis ◽  
Lefteri H. Tsoukalas
Keyword(s):  
Modeling And Simulation ◽  
Cross Sections ◽  
Experimental Testing ◽  
Detection Limits ◽  
Nuclear Data ◽  
Nuclear Materials ◽  
Simulation Techniques ◽  
Detection And Identification ◽  
The Us ◽  
Nuclear Systems

Detection and identification of special nuclear materials (SNMs) are an essential part of the US nonproliferation effort. Modern cutting-edge SNM detection methodologies rely more and more on modeling and simulation techniques. Experiments with radiological samples in realistic configurations, is the ultimate tool that establishes the minimum detection limits of SNMs in a host of different geometries. Modern modeling and simulation approaches have the potential to significantly reduce the number of experiments with radioactive sources needed to determine these detection limits and reduce the financial barrier to SNM detection. Unreliable nuclear data is one of the principal causes of uncertainty in modeling and simulating nuclear systems. In particular, nuclear cross sections introduce a significant uncertainty in the nuclear data. The goal of this research is to develop a methodology that will autonomously extract the correct nuclear resonance characteristics of experimental data in a reliable way, a task previously left to expert judgement. Accurate nuclear data will in turn allow contemporary modeling and simulation to become far more reliable, de-escalating the extent of experimental testing. Consequently, modeling and simulation techniques reduce the use and distribution of radiological sources, while at the same time increase the reliability of the currently used methods for the detection and identification of SNMs.

scholarly journals Infological models of the ENDF-format nuclear data

2019 ◽  
Vol 5 (1) ◽  
pp. 53-59
Author(s):  
Anatoliy G. Yuferov
Keyword(s):  
Cross Sections ◽  
Modern Science ◽  
Physical Nature ◽  
Nuclear Data ◽  
Secondary Neutron ◽  
Source Data ◽  
Presentation Formats ◽  
Hierarchical Nature ◽  
Data Libraries ◽  
Nuclear Data Libraries

Issues involved in the infologic modeling of the ENDF-format nuclear data libraries for the purpose of converting ENDF files into a relational database have been considered. The transfer to a relational format will make it possible to use standard readily available tools for nuclear data processing which simplify the conversion and operation of this data array. Infological models have been described using formulas of the “Entity (List of Attributes)” type. The proposed infological formulas are based on the physical nature of data and theoretical relations. This eliminates the need for a special notation to be introduced to describe the structure and the content of data, which, in turn, facilitates the use of relational formats in codes and solution of nuclear data evaluation problems. The concept of nuclear informatics has been formulated based on relational DBMS technologies as one of the tools for solving the “big data” problem in modern science and technology. The organizational and technological grounds for the transfer of ENDF libraries to a relational format are presented. Requirements to the nuclear data presentation formats supported by relational DBMS are listed. Peculiarities of the infological model construction, conditioned by the hierarchical nature of nuclear data, are identified. The sequence for the ENDF metadata saving is presented, which can be useful for the verification and validation (testing of the structural and syntactical validity and operability) of both source data and the procedures for the conversion to a relational format. Formulas of infological models are presented for the cross sections file, the secondary neutron energy distributions file, and the nuclear reaction product energy-angle distributions file. A complete array of infological models for ENDF libraries and the generation modules of respective relational tables are available on a public website.

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