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Examples of Use of SINBAD Database for Nuclear Data and Code Validation
Benchmark experiments can serve several purposes and the availability of suitable experimental data is fundamental to ensure that modelling tools meet the requirements of nuclear industry and licensing needs. Essentially, they are used for: - validation of nuclear cross-section data, - validation of nuclear codes and models used, - qualification and training of users through benchmarking and code comparison exercises. The Shielding Integral Benchmark Archive and Database (SINBAD) project was started in the early 1990's at the OECD Nuclear Energy Agency Data Bank (NEA DB) and the Radiation Safety Information Computational Center (RSICC) as a collection of quality experiments for validating and benchmarking computer codes and data for radiation transport and shielding applications. Currently, the SINBAD database comprises over 100 benchmark compilations and evaluations of relevance to: - reactor shielding, pressure vessel dosimetry (48); - fusion blanket neutronics (31); - accelerator shielding (23). In order to facilitate their use, the data are organized in the standardized format structure. The index and the master document (called abstract) use the Hyper Text Markup Language (HTML) format with links to text files, figures, Portable Document Format (pdf) documents and computer code inputs. The format allows easy access and forward/backward navigation among the data and has proved to be suitable, general enough and above all easy (and low-cost) to maintain. In addition to the characterization of the radiation source, description of the shielding set-up, the instrumentation and the relevant detector measurements, most sets in SINBAD contain also the deterministic or Monte Carlo radiation transport computer model used for the interpretation of the experiment and, where available, results from uncertainty analysis. The set of primary documents used for the benchmark compilation and evaluation are provided in computer readable form. The database is intended for different types of users, including nuclear data evaluators, computer code developers, experiment designers and university students. SINBAD is available from the NEA Data Bank and RSICC (see https://www.oecdnea.org/science/wprs/shielding/sinbad/). Up to now several hundreds of SINBAD packages were distributed from the NEA DB and RSICC. Since 2011 the maintenance and further development of the database is coordinated in the scope of the OECD NEA Nuclear Science Committee (NSC) Working Party on Scientific Issues of Reactor Systems (WPRS) Expert Group on Radiation Transport and Shielding (EGRTS). A key objective of the group is to identify, evaluate and preserve experimental data on shielding benchmarks. The Expert Group maintains close links with the International Reactor Physics Experiment Evaluation (IRPhE) and International Criticality Safety Benchmark Evaluation (ICSBEP) projects, as well as in co-ordination with the JEFF and EFF NEA Data Bank projects and with the SATIF community of accelerator shielding specialists. In the last years, much effort was devoted to the quality review of the available benchmark information.
NCSP Analytical Methods Subtask 3, AMPX Development and Maintenance, and NCSP Nuclear Data Subtask 6, SAMMY Modernization
The modernization of SAMMY continued with the removal of all container array access for the main program. The container array was used due to the computer limitations at the time of the initial writing of SAMMY, but it prevented the use of some modern code testing options. Additionally, the authors continued to make the code more modular by shifting the storage of the energy grid and the covariance information to C++ storage, thus allowing for better unit testing and the elimination of scratch files. For AMPX, the authors continued to implement the new Evaluated Nuclear Data File format to allow AMPX to read new evaluations as they become available.
Re-analysis of HCPB/HCLL Blanket Mock-up Experiments Using Recent Nuclear Data Libraries
We have re-analysed the two breeding blankets experiments performed previously in the frame of the European fusion program on two mock-ups of the European Helium-Cooled-Lithiium Lead (HCLL) and Helium-Cooled-Pebble-Bed (HCPB) test blanket modules for ITER. The tritium production rate and the neutron and photon spectra measured in these mock-ups were compared with calculations using FENDL-3 Starter Library, release 4 and state-of-the-art nuclear data evaluations, JEFF-3.1.2, JENDL-4.0 and ENDF/B-VII.0. The tritium production calculated for the HCPB mock-up underestimates the experimental result by about 10%. The result calculated with FENDL-3/SLIB4 gives slightly smaller tritium production by 2% than the one with FENDL-2.1. The difference attributes to the slight modification of the total and elastic scattering cross section of Be. For the HCLL experiment, all libraries reproduce the experimental results well. FENDL-3/SLIB4 gives better result both for the measured spectra and the tritium production compared to FENDL-2.1.
The Half Monte Carlo Method: Combining Total Monte Carlo with Nuclear Data Sensitivity Profiles
Both Total Monte Carlo (TMC) and Generalized Linear Least Squares (GLLS) algorithms have been used extensively to perform nuclear uncertainty propagation. Both methods have strengths and weaknesses, for example, the time required to perform TMC limits its attractiveness, while GLLS cannot generate higher order moments of output distributions and also has trouble dealing with nonlinear distributions. This work uses a new method that overcomes many of the aforementioned limitations of TMC and GLLS, while at the same time reasonably reproduces the TMC result. The concept is a simple combination of the two methods i.e., generate the random nuclear data files using the TMC approach, and then propagate the effect on the integral parameters like k{sub eff} using sensitivity coefficients, as opposed to cumbersome neutron transport calculations. Agreement is demonstrated for the Jezebel sphere ICSBEP benchmark. Since the method is part TMC part GLLS, the term Half Monte Method (HMM) will be used.
EXCALIBUR-at-CALIBAN: a neutron transmission experiment for {sup 238}U(n,n'{sub continuum}γ) nuclear data validation
Two recent papers justified a new experimental program to give a new basis for the validation of {sup 238}U nuclear data, namely neutron induced inelastic scattering and transport codes at neutron fission energies. The general idea is to perform a neutron transmission experiment through natural uranium material. As shown by Hans Bethe, neutron transmissions measured by dosimetric responses are linked to inelastic cross sections. This paper describes the principle and the results of such an experience called EXCALIBUR performed recently (January and October 2014) at the CALIBAN reactor facility. (authors)
Our Future Nuclear Data Needs
A well-established knowledge of nuclear phenomena including fission, reaction cross sections, and structure/decay properties is critical for applications ranging from the design of new reactors to nonproliferation to the production of radioisotopes for the diagnosis and treatment of illness. However, the lack of a well-quantified, predictive theoretical capability means that most nuclear observables must be measured directly and used to calibrate empirical models, which in turn provide the data needed for these applications. In many cases, either there is a lack of data needed to guide the models or the results of the different measurements are discrepant, leading to the development of evaluation methodologies to provide recommended values and uncertainties. In this review, we describe the nuclear data evaluation process and the international community that carries it out. We then discuss new measurements and improved theory and/or modeling needed to address future challenges in applied nuclear science.
Improving the Nuclear Data on Fission Product Decays at CARIBU: LL18-FissProdCARIBU-PD3Jb TRA 2018
The scope of the project is to improve the fundamental nuclear data of fission products relevant to applications. Per the TRA guidance document “We will improve the precision of the absolute g-ray branching ratios for longlived fission products by nearly an order of magnitude to 1-2% precision” describes the goal of the project. This goal is unlikely to change during the course of the project.
Public Release of the MENDF80 and MT80 Nuclear Data Libraries for NDI
This document describes the MENDF80 and MT80 data libraries, which are multi-group neutron cross section libraries based on ENDF/B-VIII.0 for LANL’s Nuclear Data Interface (NDI). MENDF80 is a downscatter-only library, while MT80 is multi-temperature. Both libraries also have 30-group pre-collapsed versions, MENDF80 30 and MT80 30.
Chlorine Nuclear Data Needs for Advanced Reactors [Slides]
The key physics parameters of the SSR-W have been shown to exhibit high sensitivity to chlorine nuclear data, particularly the Cl-35 (n,p) cross section. This sensitivity will be a feature for any fast chloride reactor. The sensitivity, together with large uncertainty in the fast region, results in significant uncertainty in important output reactor physics parameters, such as reactivity, and hence the required fuel loading for criticality.
Generalized Nuclear Data - WPEC Subgroup 43
This repository describes the goals and governance of WPEC Subgroup 43 (SG43). This subgroup is tasked with creating infrastructure for the Generalized Nuclear Data (GND) format.
Sample Generation for Nuclear Data
This report summarizes a NEAMS (Nuclear Energy Advanced Modeling and Simulation) project focused on developing a sampling capability that can handle the challenges of generating samples from nuclear cross-section data. The covariance information between energy groups tends to be very ill-conditioned and thus poses a problem using traditional methods for generated correlated samples. This report outlines a method that addresses the sample generation from cross-section matrices. The treatment allows one to assume the cross sections are distributed with a multivariate normal distribution, lognormal distribution, or truncated normal distribution.
United States Nuclear Data Program Work Plan for Fiscal Year 2025
The work plan described in this document has been developed to cover work to be performed by the U. S. Nuclear Data Program (USNDP) during Fiscal Year (FY) 2025 that begins on October 1, 2024. This plan has been prepared in consultation with the members of the Coordinating Committee who represent the organizations participating in the program. Each Coordinating Committee member prepared a draft plan for his/her organization. Each contribution was integrated into a unified work plan. The draft plan was then circulated to the Coordinating Committee for comments and corrections before the final document was submitted to the U.S. Department of Energy (DOE).
Nuclear Data Sheets for A=194
Experimental nuclear structure and decay data are evaluated for all of 14 known nuclides of mass 194 (Ta, W, Re, Os, Ir, Pt, Au, Hg, Tl, Pb, Bi, Po, At, Rn). Detailed evaluated spectroscopic information is presented for each reaction and decay experiment, and by combining all the available data, recommended values are provided for energies, spins and parities and half-lives of levels, together with energies, branching ratios and multipolarities of gamma radiations from the levels, and characteristics of beta and alpha radiations in radioactive decays. No excited states have yet been identified in 194 Ta, 194 W, and 194 Rn. Only limited spectroscopic data are available for 194 Re, with a 45–μs isomer identified, but its absolute energy yet to be determined. Only two levels at unknown absolute energies are reported for 194 At, both of which are isomeric but neither can be unambiguously identified as the ground state. Data are scarce for 194 Os, 194 Bi and 194 Po, while 194 Pt, 194 Au, 194 Hg, 194 Tl and 194 Pb are well studied through decays and reactions. For 194 Ir, although a major (n, γ), E=thermal study and two particle-transfer studies provide large amounts of spectral data, still several problematic issues are found in the assignment of multipolarities for the secondary gamma transitions in the (n, γ) study by 2008Ba25, 1998Ba85 and 1998Ba42, all from the same group. Some issues have been resolved by the evaluators, but we suggest that further experimental work is needed on 194 Ir structure to resolve standing issues, and verify the conclusions made by 2008Ba25. Furthermore, our attempt to contact the prime authors of the 2008Ba25 paper was not successful. The present work supersedes all the earlier evaluations of A=194 nuclides by 2006Si17, 1996Br26, 1989Si01 and 1972Au11.
Nuclear Data Sheets for A=123
Experimental nuclear structure and decay data are evaluated for all of 15 known nuclides of mass 123 (Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce). For each nuclide, detailed evaluated spectroscopic information is presented in each reaction and decay, and the best values combining all available data are recommended for level properties, γ and β radiations, and other spectroscopic properties. No excited states have been identified in 123 Ru, 123 Rh and 123 Pd. For 123 Ag, the long-predicted 1/2 – β-emitting isomer has been identified at 60-keV by 2019Ch24 recently, resolving unknown excitation energies in the level scheme that was previously available only from isomeric decays of two isomers (202 ns and 393 ns) with the position and spin-parity of the former remaining unknown. Significant discrepancies exist between data on high-spin sequences based on 11/2 (–) isomer in 123 Cd (2002Hw01 and 2016Re05), which needs to be resolved with further experimental investigation. In 123 Cs, the 114-ns isomer as the πg 9/2 bandhead proposed at 231.6+x by 2000Gi12 has been resolved by 2004Si26 and 2004Si27 to be the 328-keV level that is proposed by 2000Gi12 as a separate level. Excited states in 123 La and 123 Ce have only been studied via (HI, xnγ) reactions, with their base levels and thus excitation energies remaining unknown. The β– decay schemes for daughter nuclide 123 Cd, 123 In and 123 Sn and the ε decay schemes for 123 Xe, 123 Cs and 123 Ba are considered incomplete due to large gaps between the highest observed excited levels and the Q-values. 123 Sn, 123 Sb, 123 Te and 123 I are the most extensively studied nuclides via various reactions and decays. Furthermore, this work supersedes earlier full evaluations of A=123 by 2004Oh11, 1993Oh12, 1980Ta02 and 1972Au10.
Nuclear Data Sheets for A=153
The experimental results published before August 2020 from the various reaction and decay studies leading to nuclides of Z=55 to Z=72, 153 Ba, 153 La, 153 Ce, 153 Pr, 153 Nd, 153 Pm, 153 Sm, 153 Eu, 153 Gd, 153 Tb, 153 Dy, 153 Ho, 153 Er, 153 Tm, 153 Yb, 153 Lu, 153 Hf in the A=153 mass chain have been reviewed. These data are summarized and presented, together with adopted level schemes and properties. This work is intended to supersede the previous evaluation of the A=153 nuclides by R.G. Helmer (2006He06), which was published in Nuclear Data Sheets 107, 507 (2006).
Dimensionality Reduction Algorithms for Big Nuclear Data Streams with Application to SCALE Cross Sections
A central problem in performing accurate nuclear reactor simulations is the ability to capture the changes in the few group microscopic cross sections as functions of system state, e.g. moderator density, fuel temperature, or neutron spectrum, and time and/or burnup of the fuel. For most reactor simulation scenarios, this results in huge data stream, which must be interpolated as needed by downstream calculations (e.g., core-wide simulation). This is typically accomplished by using sophisticated interpolating functions to capture the few-group variations over the wide range of conditions expected during operation. To address handling this big data, this paper explores the use of dimensionality reduction (DR) techniques to identify the true dimensionality of the data prior to its use in the calculations. This could significantly reduce the computational cost required for data processing/manipulation, thus reducing a significant portion of the total calculations time. By way of an example, in the current version of the ORIGEN-ARP module, the assembly-averaged microscopic cross sections at different conditions, e.g., fuel enrichments, irradiation history and water density, which are calculated using the TRITON sequence and interpolated on-demand by the ARP module to model assembly depletion for a wide range of assembly types (e.g. Westinghouse 17x17 and GE 9x9) over a wide range of conditions. This approach typically results in ∼1 GB per assembly type, which although not insurmountable, can be impractical for distribution. and it is the premise of this paper that most of this data is heavily correlated and hence can take great benefit of condensation via DR. To demonstrate DR, the proposed algorithm is applied to the nuclear data streams generated by SCALE 6.1. Using the depletion sequence (TRITON), microscopic cross sections are calculated for a mixed oxide (MOX) fuel assembly design at a representative range of fuel enrichments and fuel burnup. In conventional core simulators, multi dimensional interpolation of these cross sections is used to provides neutronics feedback due to local conditions without returning to expensive TRITON calculations for every change. This paper shows that dimensionality reduction can help compress the size of the cross section data without sacrificing the accuracy of their reconstruction at all conditions of interest. This work explores the application of dimensionality reduction algorithms to nuclear data streams typically employed in support of conventional full-core reactor calculations which rely on interpolation of pre-calculated microscopic cross sections. In particular, analysis of the few group MOX cross sections revealed that a large degree of dependencies exist between the few-group cross sections, which can be revealed via range-finding algorithms. Specifically, correlations between different reaction types for the same isotope have been identified, and correlations between different enrichments and correlations across different isotopes have been identified. This paper demonstrates that the reduction errors can be quantified, and they are uncorrelated with the range of variations for the various cross sections. In future work, we will focus on identifying the smallest rank that takes into account all types of correlations across the range of conditions used to model cross section variations. (authors)
Oak Ridge National Laboratory NCSP Analytical Methods Subtask 3, AMPX Development and Maintenance, and NCSP Nuclear Data Subtask 6, SAMMY Modernization
The modernization of SAMMY continued with consolidation of access to covariance information for adjusted parameters and data in SAMMY. This consolidation allowed for removal of many scratch files. In addition, work was initiated to make the 0K cross section calculation more modular and less dependent on SAMMY global parameters. An initial application programming interface (API) was added to expose cross sections (including resolution broadening) generated by SAMMY to external fitting routines. The processing for thermal moderators in AMPX was updated for selected moderators for which the generated grid was not fine enough. Updated libraries were generated for SCALE. In addition, work continued to fully support new Evaluated Nuclear Data File (ENDF) formats, including the Generalized Nuclear Database Structure (GNDS) in AMPX.