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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

A New Method for Generating Probability Tables in the Unresolved Resonance Region

One new method for constructing probability tables in the unresolved resonance region (URR) has been developed. This new methodology is an extensive modification of the single-level Breit-Wigner (SLBW) pseudo-resonance pair sequence method commonly used to generate probability tables in the URR. The new method uses a Monte Carlo process to generate many pseudo-resonance sequences by first sampling the average resonance parameter data in the URR and then converting the sampled resonance parameters to the more robust R-matrix limited (RML) format. Furthermore, for each sampled set of pseudo-resonance sequences, the temperature-dependent cross sections are reconstructed on a small grid around the energy of reference using the Reich-Moore formalism and the Leal-Hwang Doppler broadening methodology. We then use the effective cross sections calculated at the energies of reference to construct probability tables in the URR. The RML cross-section reconstruction algorithm has been rigorously tested for a variety of isotopes, including 16 O, 19 F, 35 Cl, 56 Fe, 63 Cu, and 65 Cu. The new URR method also produced normalized cross-section factor probability tables for 238 U that were found to be in agreement with current standards. The modified 238 U probability tables were shown to produce results in excellent agreement with several standard benchmarks, including the IEU-MET-FAST-007 (BIG TEN), IEU-MET-FAST-003, and IEU-COMP-FAST-004 benchmarks.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

On-the-fly Doppler Broadening with Probability Table Interpolation for Unresolved Resonance Region in RMC

With the rapid development of computational power and parallel algorithms, Monte Carlo method has been widely investigated and used in neutron transport simulations due to its advantages in geometry modelling and usage of continuous energy point-wise cross sections. Nowadays the road has been paved to couple Monte Carlo codes with thermal-hydraulics codes to make multi-physics calculations. One of the most important aspects of thermal-hydraulics feedback is the detailed temperature distribution, which demands temperature dependent cross sections in neutron transport simulations. The conventional pre-generated cross section libraries at each temperature point are no longer feasible for detailed temperature distributions because of tremendous memory footprint. Hence, different on-the-fly Doppler broadening methods have been proposed for resolved resonance energy region and thermal energy region, such as the interpolation, regress model fitting, multipole representation (MPR), target motion sampling (TMS) and different integration algorithms based on Doppler broadening equation. Compared to resolved resonance energy region and thermal energy region, the representation of cross sections in unresolved resonance region is slightly different. As a result of the limitation of experiment measurements, the overlapped resonances cannot be distinguished from each other and the statistic average resonance parameters are given in the evaluated nuclear data file. In early applications, infinite-dilute pointwise cross sections were obtained from average parameters and were directly used in neutron simulations. However, it was proven that these infinite-dilute pointwise cross sections cannot capture the energy self-shielding effect, thus leading to an overestimate of reaction rates and usually an underestimate of k{sub eff} owing to the increased absorption rates and increased scattering probability to energy below unresolved resonance region where the capture-to-fission ratio is typically higher. This phenomenon is of great importance especially for problems where there is a considerable flux in unresolved resonance region. Then the probability table method was put forward by Levitt to take energy self-shielding effect into account, which provides probability distribution function of cross sections in unresolved resonance region instead of specific cross section values. The probability distribution function is presented in the form of tables and the mean values are equal to the infinite dilution cross sections. To obtain the temperature dependent cross sections in unresolved energy region, there are two different methods. One is to calculate cross sections directly from average resonance parameters and their statistical distributions , which is effective and precise but time-consuming. The other is to interpolate between pre-generated probability table files, which is much more efficient. In this paper, on-the-fly Doppler broadening of cross sections in unresolved resonance regions based on probability tables interpolation was implemented in RMC codes as a supplement to on-the-fly Doppler broadening for resolved resonance energy region and thermal region, which is based on the integration of Doppler broadening equation. The results of criticality calculation and depletion calculation show that on-the-fly Doppler broadening for the whole energy region is accurate and efficient. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Development of nuclear data processing code FRENDY version 2

Nuclear data processing is an important interface between an evaluated nuclear data library and neutronics calculation codes. JAEA has been developed the new nuclear data processing code FRENDY from 2013. FRENDY version 1 generates the ACE files used for the continuous-energy Monte Carlo codes including PHITS, Solomon, Serpent,and MCNP and it was released as the open-source software under the 2-clause BSD license in 2019.After we released FRENDY version 1, many functions, e.g., the multi-group neutron cross-section library generation, the statistical uncertainty quantification of the probability tables for unresolved resonance cross-section, the 'perturbation' of the ACE file for uncertainty quantification, and the modification of the ENDF-6 formatted nuclear data file, have been developed. We released FRENDY version 2 including these functions. The present paper gives an overview of FRENDY and presents features of new functions implemented in FRENDY version 2. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

FISPACT-II: An Advanced Simulation System for Activation, Transmutation and Material Modelling

Fispact-II is a code system and library database for modelling activation-transmutation processes, depletion-burn-up, time dependent inventory and radiation damage source terms caused by nuclear reactions and decays. The Fispact-II code, written in object-style Fortran, follows the evolution of material irradiated by neutrons, alphas, gammas, protons, or deuterons, and provides a wide range of derived radiological output quantities to satisfy most needs for nuclear applications. It can be used with any ENDF-compliant group library data for nuclear reactions, particle-induced and spontaneous fission yields, and radioactive decay (including but not limited to TENDL-2015, ENDF/B-VII.1, JEFF-3.2, JENDL-4.0u, CENDL-3.1 processed into fine-group-structure files, GEFY-5.2 and UKDD-16), as well as resolved and unresolved resonance range probability tables for self-shielding corrections and updated radiological hazard indices. The code has many novel features including: extension of the energy range up to 1 GeV; additional neutron physics including self-shielding effects, temperature dependence, thin and thick target yields; pathway analysis; and sensitivity and uncertainty quantification and propagation using full covariance data. The latest ENDF libraries such as TENDL encompass thousands of target isotopes. Nuclear data libraries for Fispact-II are prepared from these using processing codes PREPRO, NJOY and CALENDF. These data include resonance parameters, cross sections with covariances, probability tables in the resonance ranges, PKA spectra, kerma, dpa, gas and radionuclide production and energy-dependent fission yields, supplemented with all 27 decay types. All such data for the five most important incident particles are provided in evaluated data tables. The Fispact-II simulation software is described in detail in this paper, together with the nuclear data libraries. The Fispact-II system also includes several utility programs for code-use optimisation, visualisation and production of secondary radiological quantities. Included in the paper are summaries of results from the suite of verification and validation reports available with the code.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

The SCALE/AMPX multigroup cross section processing for fast reactor analysis

The SCALE/AMPX multigroup (MG) cross section processing procedure has been updated to minimize reactivity differences for fast reactor designs and boiling water reactors (BWRs) with very high void fractions to provide excellent agreement with continuous-energy reference calculations. SCALE MG calculations are widely applied to thermal spectrum light-water reactor (LWR) systems as well as fast spectrum metallic systems of interest to National Nuclear Security Administration. With growing interest from industry and regulators in applying SCALE for the design of fast spectrum reactors, both sodium and molten salt, and in the licensing of power up rates for BWRs, it is desirable to review the SCALE/AMPX procedure for unresolved resonance self-shielded data and high energy neutron spectra. The data were improved by generating MG unresolved resonance data based on the analytic probability table method with the narrow resonance approximation as well as the use of a very fine group structure typically used in fast system analysis. This study focused on verifying the probability table and the SCALE/AMPX MG cross section processing procedure by performing reaction rate analysis and benchmark calculations for various fast reactor and high void BWR systems. Here, results indicate that the improved SCALE/AMPX MG cross section processing provides excellent results for the fast reactor analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Summary: Verification of the ENDF/B-VII.1 and VIII.0 AMPX 1597-group Libraries for Advanced Reactor Analysis

The SCALE/AMPX multigroup (MG) cross section processing procedure has been updated to minimize reactivity differences for various advanced thermal and fast reactor designs, as observed for the current MG libraries, resulting in excellent agreement between the calculations with the new MG libraries and the continuous-energy reference calculations. The SCALE MG calculations are widely applied to thermal spectrum light-water reactor systems, as well as fast spectrum metallic systems. Due to growing interest from industry and regulators in applying SCALE for the design of fast spectrum reactors—both sodium and molten salt—it was desirable to review and improve the SCALE/AMPX procedure for unresolved resonance self-shielded data and high-energy neutron spectra. The data were improved by generating MG unresolved resonance data based on the analytic probability table method with the narrow resonance approximation and by using very fine and intermediate group structures that are typical for fast system analysis. This study focused on verifying the improved SCALE/AMPX MG cross section processing procedure and the new AMPX 1597-group library with the ENDF/B-VII.1 and VIII.0 evaluated nuclear data. The verification was made by performing reaction rate analysis and benchmark calculations for various thermal and fast reactor systems. Results indicate that the improved SCALE/AMPX MG cross section processing and libraries provide excellent results for advanced reactor analysis.

Kim, Kang Seog↗
DOE OSTI2019

Verification of the ENDF/B-VII.1 and VIII.0 AMPX 1597-Group Libraries for Advanced Reactor Analysis

The SCALE/AMPX multigroup (MG) cross section processing procedure has been updated to minimize reactivity differences for various advanced thermal and fast reactor designs, as observed for the current MG libraries, resulting in excellent agreement between the calculations with the new MG libraries and the continuous-energy reference calculations. The SCALE MG calculations are widely applied to thermal spectrum light-water reactor systems, as well as fast spectrum metallic systems. Due to growing interest from industry and regulators in applying SCALE for the design of fast spectrum reactors—both sodium and molten salt—it was desirable to review and improve the SCALE/AMPX procedure for unresolved resonance self-shielded data and high-energy neutron spectra. The data were improved by generating MG unresolved resonance data based on the analytic probability table method with the narrow resonance approximation and by using very fine and intermediate group structures that are typical for fast system analysis. This study focused on verifying the improved SCALE/AMPX MG cross section processing procedure and the new AMPX 1597-group library with the ENDF/B-VII.1 and VIII.0 evaluated nuclear data. The verification was made by performing reaction rate analysis and benchmark calculations for various thermal and fast reactor systems. Results indicate that the improved SCALE/AMPX MG cross section processing and libraries provide excellent results for advanced reactor analysis.

Kim, Kang Seog↗
DOE OSTI2025

Investigation of the approach used in the unresolved resonance region

The typical practice used in Monte Carlo neutron transport codes in the unresolved resonance region (URR) is to take advantage of the probability table (PT) approach. Cross sections are sampled from PTs and used as needed, along with generated average cross sections. The PTs are generated based on cross-section calculations performed using the single-level Breit–Wigner approximation. Although the approach used in the URR seems plausible, a detailed examination was needed to understand the benchmark results obtained using specific tests.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

On-the-Fly Treatment of Temperature Dependent Cross Sections in the Unresolved Resonance Regions in RMC Code

Monte Carlo method can provide high fidelity neutronics analysis of different nuclear reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. It can also be coupled with thermal-hydraulics codes to consider the feedback. For thermal reactors such as PWR and HTGR, two main temperature effects should be considered. The first is the effect of thermal motion of target nuclei in resolved resonance energy regions, which is known as Doppler effect. The second is the thermal scattering and bound effect in thermal energy regions. The traditional approach of pre-generated cross sections has difficulty in memory footprint for detailed temperature modelling in multi-physics calculations. Recently, on-the-fly (OTF) technique has been proposed in order to reduce the memory consumption for both resolved resonance energy and thermal energy. For fast reactors calculations, the temperature dependence of cross sections in the unresolved resonance energy regions is also important. However, less attention was paid to the unresolved resonance regions compared with resolved resonance energy and thermal energy. In this paper, the on-the-fly temperature treatment of cross sections in unresolved resonance regions was proposed and developed in RMC codes. The results show that the on-the-fly treatment has high efficiency and satisfactory fidelity. The on-the-fly treatment of unresolved resonance region was also combined with the target motion sampling (TMS) method for resolved resonance energy. The formats of libraries were changed, in which the probability tables were separated from the cross sections. Then the interpolation of probability table was performed on-the-fly during the transport routine. The results show that the on-the-fly treatment has high efficiency and satisfactory fidelity. The on-the-fly treatment of unresolved resonance region was also combined with the target motion sampling (TMS) method for resolved resonance energy, showing that TMS method is effective and accurate for cross sections treatment in resolved resonance energy regions. The memory saving of the on-the-fly treatment is also considerable. For the high-fidelity 'transport-burnup-thermal-hydraulics' coupling calculation, the memory requirement is 60 Mb. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Advancements in nuclear data evaluations in the unresolved resonance region [Slides]

Minimal updates to current capabilities (SAMMY/AMPX) and to the ENDF format are proposed to include threshold reaction for outgoing charged-particle channels in the URR formalism. If average resonance parameters are reported, probability tables for newly defined reactions channels may be needed, implemented, and tested. Nuclear data evaluation of n+ 35 Cl reactions can be considered the best case to apply the proposed updates.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

On-the-fly Doppler broadening of unresolved resonance region cross sections

In this paper, two methods for computing temperature-dependent unresolved resonance region cross sections on-the-fly within continuous-energy Monte Carlo neutron transport simulations are presented. The first method calculates Doppler broadened cross sections directly from zero-temperature average resonance parameters. In a simulation, at each event that requires cross section values, a realization of unresolved resonance parameters is generated about the desired energy and temperature-dependent single-level Breit-Wigner resonance cross sections are computed directly via the analytical Ψ-x Doppler integrals. The second method relies on the generation of equiprobable cross section magnitude bands on an energy-temperature mesh. Within a simulation, the bands are sampled and interpolated in energy and temperature to obtain cross section values on-the-fly. Both of the methods, as well as their underlying calculation procedures, are verified numerically in extensive code-to-code comparisons. Energy-dependent pointwise cross sections calculated with the newly-implemented procedures are shown to be in excellent agreement with those calculated by a widely-used nuclear data processing code. Relative differences at or below 0.1% are observed. Integral criticality benchmark results computed with the proposed methods are shown to reproduce those computed with a state-of-the-art processed nuclear data library very well. In simulations of fast spectrum systems which are highly-sensitive to the representation of cross section data in the unresolved region, k-eigenvalue and neutron flux spectra differences of <10 pcm and <1.0% are observed, respectively. The direct method is demonstrated to be well-suited to the calculation of reference solutions — against which results obtained with a discretized representation may be assessed — as a result of its treatment of the energy, temperature, and cross section magnitude variables as continuous. Also, because there is no pre-processed data to store (only temperature-independent average resonance parameters) the direct method is very memory-efficient. Typically, only a few kB of memory are needed to store all required unresolved region data for a single nuclide. However, depending on the details of a particular simulation, performing URR cross section calculations on-the-fly can significantly increase simulation times. Alternatively, the method of interpolating equiprobable probability bands is demonstrated to produce results that are as accurate as the direct reference solutions, to within arbitrary precision, with high computational efficiency in terms of memory requirements and simulation time. Analyses of a fast spectrum system show that interpolation on a coarse energy-temperature mesh can be used to reproduce reference k-eigenvalue results obtained with cross sections calculated continuously in energy and directly at an exact temperature to within <10 pcm. Probability band data on a mesh encompassing the range of temperatures relevant to reactor analysis usually require around 100 kB of memory per nuclide. Finally, relative to the case in which probability table data generated at a single, desired temperature are used, minor increases in simulation times are observed when probability band interpolation is employed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Calculation and use of pointwise effective cross-sections in the unresolved resonance region - 14605

The pointwise neutron elastic scattering, absorption, fission and total cross-sections in the Unresolved Resonance Region (URR) can be calculated with a code C-007 [1]. The volume of these data is large and estimated to make about 106 numbers for each of the reactions. These data volumes may cause time and memory problems in solving radiation transport by the Monte Carlo method. There are methods to thin the data. These are, for example, the widely used probability tables [2] and subgroups [3]. The paper offers another data thinning approach which was successfully used for the Resolved Resonance Region (RRR) [4,5]. It transforms the pointwise resonance cross-sections into the pointwise monotone ones within some energy intervals with the use of cross-section probability distribution functions P(σ). A similar approach was used to calculate pointwise cross-sections in the URR [6]. (authors)

42 ENGINEERING↗
DOE OSTI2018

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

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 high 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.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

About the Impact of the Unresolved Resonance Region in Monte Carlo Simulations of Sodium Fast Reactors

In the last few years, and within the framework of different European projects, KENO-VI code from SCALE system has been employed to perform detailed continuous-energy Mon-te Carlo transport calculations for advanced fast reactors. The core characterization of both the sodium-cooled ASTRID and the lead-cooled ALFRED reactors was performed during the FP7 cross-cutting ESNII+ project; more recently, core calculations for the sodium-cooled Superphénix reactor and the improved European Sodium Fast Reactor design were performed within the HORIZON2020 ESFR-SMART project. In all cases, the effective multiplication factor predicted by KENO-VI was systematically higher (around 400-500 pcm) than the values computed by MCNP and Serpent Monte Carlo codes, using the same nu-clear data library.In order to provide insight into the origin of the observed discrepancies, a simplified 2D MOX-fueled SFR pin-cell benchmark has been launched. The multiplication factor, as well as 1-group and VITAMINJ 175-group cross-sections computed by KENO-VI, Serpent and MCNP codes employing ENDF/B-VII.1 data library, have been compared.Significant differences between KENO-VI and the other codes have been found in the un-resolved resonance regions of 239Pu and 241Pu capture and production cross sections, while negligible differences appeared outside those energy ranges. On the other hand, calculations without using probability tables have shown very good agreement. Quantita-tive comparison is presented and analyzed, along with a discussion of the impact of the probability-table treatment in the three codes for MOX-fueled systems with typical SFR spectrum.

Jimenez carrascosa, A↗