Nuclear Data Libraries Research Articles

GIDI+: A GNDS 2.0 suite of C++ APIs to access nuclear and atomic data for use in radiation transport codes

GIDI+ is a C++ package that reads GNDS formatted nuclear reaction and structure data, and photo-atomic reaction data as needed by radiation transport codes. As of version 3.25, GIDI+ supports reading GNDS 2.0 formatted data. GNDS is a new extensible hierarchy that has been internationally adopted as the new standard for storing nuclear and photo-atomic data, replacing ENDF-6 which has been the standard since the 1960’s. GIDI+ 3.25 supports the GNDS 2.0 map file which can contain a list of all GNDS data needed to make a complete nuclear reaction data library as needed by radiation transport codes. GIDI+ has functions that make it easy to obtain multi-group data from a GNDS file, including summed data (e.g., summing multi-group reaction cross sections to obtain the total multi-group cross section). In addition, the MCGIDI sub-package of GIDI+ can be used by Monte Carlo transport codes to look up data (e.g., cross section, deposition energy) as a function of projectile energy and material temperature. The MCGIDI sub-package can use either GNDS multi-groupor continuous energy data. MCGIDI also has functions for sampling a reaction from a GNDS reactionSuite and functions for sampling a reaction’s outgoing particle energy and angular data. Recent work on GIDI+ includes: supporting GNDS 2.0, speeding up load times and adding support for URR probability tables. GIDI+ is currently implemented in LLNL’s Mercury [4] (Monte Carlo) and Ardra [5] (deterministic) radiation transport codes. An effort to update GEANT4 to use the latest GIDI+ is underway and should be completed by March 2023. GIDI+ is released under the MIT license and can be found at github.com/LLNL/gidiplus.

NUCLEAR REACTION CROSS-SECTION EVALUATION OF NEUTRON-INDUCED REACTIONS IN SOME COMMON REACTOR FUEL ISOTOPES OF SOME ELEMENTS AT ENERGY RANGE OF 13.6-15.2 MeV

As part of the effort to enrich the nuclear data library and combat the challenge of the scarcity of data, EXIFON 2.0 code has been used to evaluate the (n, p) of 54,56Fe, 55Mn, 58,60Ni, 27Al, 50Cr, and 239Pu. Results of the evaluated excitation functions obtained with the EXIFON code were compared with the ENDF data retrieved from IAEA Nuclear Data section of the ENDF data library. Some of the evaluated excitation functions agreed well with the ENDF data, while some show some degree of variance. The good agreement was obtained for nuclides with EVEN-EVEN nuclei. The overall percentage deviation of total cross section from ENDF values obtained is less than 1% for, 50Cr, 56,54Fe, 55Mn, 60Ni, 27Al and 239Pu except for 58Ni which is 4% while this work records an overall average percentage deviation of 0.5% . However, the results show good agreement with available retrieved data and hence can be used where data is scarce or grossly inadequate.

Deuteron and alpha sub-libraries of JENDL-5

JENDL-5, the latest version of the Japanese evaluated nuclear data library, includes several sub-libraries to contribute to various applications. In this paper, we outline the evaluation and validation of the deuteron reaction sub-library developed mainly for the design of accelerator-based neutron sources and the alpha-particle reaction sub-library developed mainly for use in the back-end field. As for the deuteron sub-library, the data for 6,7Li, 9Be, and 12,13C from JENDL/DEU-2020 were partially modified and adopted. The data up to 200 MeV for 27Al, 63,65Cu, and 93Nb, which are important as accelerator structural materials, were newly evaluated based on the calculations with the DEURACS code. As for the alpha-particle sub-library, the data up to 15 MeV for 18 light nuclides from Li to Si isotopes were evaluated based on the calculations with the CCONE code, and then only the neutron production cross sections were replaced with the data of JENDL/AN-2005. Validation on neutron yield by Monte Carlo transport simulations was performed for both sub-libraries. As a result, it was confirmed that the simulations based on the sub-libraries showed good agreement with experimental data.

Nuclide production cross sections in proton-induced reactions on Bi at GeV energies

Proton-induced nuclide production cross sections for Bi at incident energies of 0.4, 1.5, and 3.0 GeV are measured with the activation method using the proton beams accelerated from the Rapid-Cycling Synchrotron of the Japan Proton Accelerator Research Complex (J-PARC). A total of 127 cross-section data are obtained. The measured data are compared with two types of Monte Carlo-based spallation models (i.e. INCL++ coupled with ABLA07 and INCL4.6 coupled with GEM), and the evaluated nuclear data library, JENDL/HE-2007. The result showed that INCL++/ABLA07 overall agrees with the experimental data, whereas INCL4.6/GEM underestimates the production of fission fragments.

Nuclear Data Uncertainty Analysis to Meet the Target Accuracy Requirements on the MYRRHAkeff

MYRRHA is a flexible experimental facility being designed at the SCK CEN, in Mol, Belgium. Cooled by lead-bismuth, it is conceived to operate both in sub-critical mode, as an accelerator driven system, and in critical mode, as a fast reactor. In order to comply with MYRRHA reactor design requirements, uncertainties due to nuclear data must be quantified. Significant gaps between the uncertainties and the target accuracies have been systematically shown in the past. In this paper, first, a Sensitivity and Uncertainty analysis with JEFF-3.3 nuclear data library of the effective neutron multiplication factorkeffof the latest MYRRHA reactor design - v1.8 - is presented. Then, since target accuracy forkeffof 300 pcm is exceeded, a Target Accuracy Requirement assessment is performed in order to find out the required accuracy on cross section data to meet the requested target accuracy. To reach the requested target accuracy, a reduction of the uncertainty in the fission and capture cross sections of240Pu JEFF-3.3 evaluation is needed.

Am-241 thermal neutron capture cross section and neutron capture resonance integral from reactor activation and oscillation measurements

In order to validate the neutron capture cross section in 241Am, comparisons of calculations with activation experiments at the TRIGA reactor at JSI and pile oscillation measurements at the MINERVE reactor at CEA Cadarache were done. The results suggest that the evaluation in the new test library JEFF-4T1 on average provide an improvement of the C/E results compared to the existing nuclear data library releases. However, due to some discrepancies, additional experiments at JSI as well as further development of the computational model of the JSI TRIGA reactor are planned to improve the accuracy and reliability of the neutron activation analysis and the nuclear data inferred from them.

New setup for measurement of prompt gammas from neutron interactions

Prompt gamma radiation from neutron interactions is an important issue as it affects the operation of nuclear facilities (radiation protection or gamma heating issues) or has various uses (non-destructive identification of elements). Despite of importance, its production is inaccurately described in present nuclear data libraries. Therefore, a set of experiments have been carried out focusing on this radiation - prompt gamma from neutron inelastic scattering on oxygen and prompt gamma from radiative capture on manganese, both present the form of an aqueous solution. The gammas were induced by the Am-Be neutron source (for oxygen) and 252Cf spontaneous fission neutron source (for manganese). In the case of oxygen, the present libraries can be used, but all the libraries tested overestimated the experiment by 20–30%, considering the experimental uncertainty of 6-9%. INDEN-4.0 library is giving the best results. In the case of manganese, none of the libraries tested give acceptable results. The INDEN updated data for gamma production on manganese give usable results, but there is still a discrepancy from -36% to 55% with uncertainties from 6-15%.

Development and Verification of Capability for Processing the Generalised Nuclear Database Structure (GNDS) in Nuclear Data Processing Code NECP-Atlas

The latest evaluated nuclear data libraries, including ENDF/B-VIII.0, TENDL2019, all have two formats, the Evaluated Nuclear Data File (ENDF-6) nuclear data format, and the Generalised Nuclear Database Structure (GNDS) format. The first official public documentation of all of the options for storing data in GNDS was released by the Working Party on International Nuclear Data Evaluation Cooperation (WPEC) of the Nuclear Energy Agency. The generalized derived class is designed to be compatible with ENDF-6 and GNDS in nuclear data processing code NECP-Atlas, and the GNDS-Toolkit is developed in NECP-Atlas to parse and convert GNDS. ACE format library and multi-group library are generated based on GNDS, and they are used in the Monte-Carlo code and deterministic based code to compare with ENDF-6. The numerical results show that GNDS libraries can be processed by NECP-Atlas competently to generate specific format library and the difference in ENDF-6 and GNDS is very small.

How can a diverse set of integral and semi-integral measurements inform identification of discrepant nuclear data?

Nuclear data are used for a variety of applications, including criticality safety, reactor performance, and material safeguards. Despite the breadth of use-cases, the effective neutron multiplication factor, keff, of ICSBEP critical assemblies are primarily used for nuclear data validation; these are sensitive to specific energy regions and nuclides and are unable to uniquely constrain nuclear data. As a consequence, general-purpose nuclear data libraries, such as ENDF/B-VIII.0 [1], may have deficiencies that, while not apparent in criticality applications, negatively impact other applications, such as non-destructive analysis of special nuclear material [2, 3] and neutron diagnosed subcritical experiments [4]. Recent work by the Experiments Underpinned by Computational Learning for Improvements in Nuclear Data (EUCLID) project developed a machine learning tool, RAFIEKI, which uses random forests and the SHAP metric to determine which nuclear data contribute most to predicted bias between measured and simulated responses (e.g. keff). This paper contrasts RAFIEKI analysis applied to keff only against RAFIEKI analysis with keff paired with either LLNL pulsed sphere measurements or subcritical benchmarks. Two examples show that a) including pulsed sphere measurements substantially increases 9Be nuclear data importance to bias between 2 and 15 MeV, and b) including subcritical benchmarks has the potential for disentangling compensating errors between 240Pu (n,el) and (n,il) cross-sections between 0.1 and 10 MeV. These results show that RAFIEKI analysis applied to response sets that include, but go beyond, keff can aid nuclear data evaluators in identifying issues in nuclear data.

Papillon Nuclear Data Library – a free and open-source C++/Python library for interacting with ACE files for continuous-energy neutron data

In this paper, a new C++ and Python library, called the Papillon Nuclear Data Library, is presented. This library gives users a simple interface to read and interact with continuous-energy neutron data in the form of ACE files. Using the library, one can easily evaluate or plot cross-sections for different nuclides and reactions, or sample scattering distributions. Currently, all continuous-energy neutron physics is implemented, and the library can read free-gas neutron and thermal scattering law ACE files. The Python interface makes using nuclear data easy and accessible for students, while the C++ interface gives experienced researchers easy access to continuous-energy physics when writing Monte Carlo codes to test new computational methods.

Modernization efforts for the R-Matrix code SAMMY

The R-Matrix code SAMMY [1] is a widely used nuclear data evaluation code focused on the resolved range, which includes corrections for experimental effects. The code is still mostly written in FORTRAN 77 and uses a memory management system suitable for the time of its initial writing in 1984. A modernization effort is underway to update the code to modern software development practices. A continuous-integration testing framework was added to automate the large existing set of test cases. Improvements in memory management were implemented to make the code easier to maintain and enable enhancements. The resonance parameters and covariance information are now stored in C++ objects shared by SAMMY and AMPX [2], which is the processing code that generates nuclear data libraries for SCALE [3]. Further plans include switching to the Evaluated Nuclear Data File (ENDF) reading and writing routines in AMPX because these routines are more robust, easier to maintain, and support more features. Support for the new Generalized Nuclear Database Structure (GNDS) format [4] is also of interest. GNDS will share not only the resonance parameters but also the parameters associated with experimental correction in GNDS. The data are currently available in a binary SAMMY format, and the ability to export them to GNDS would make them more widely available and shareable. The next step will be to use the same resonance processing code at 0K in AMPX and SAMMY as an available formalism. Then, any improvements in the formalism can immediately be tested in SCALE because the reconstruction in AMPX will use the same cross section model. The new data library can then be used for testing using the VALID Benchmark suite [5] or other suitable benchmark suites.

New experiment for measuring the delayed neutron yields of238U fission in the range 1 MeV to 19 MeV

In nuclear reactors operating in nominal conditions, delayed neutrons (DN) drive the kinetic behavior of the neutron population. Predicting this behavior requires therefore the knowledge of the DN yield per fission, average lifetime, and kinetic of their emission, which is described by precursors groups abundances. Characterizing the uncertainties of these macroscopic data is essential, especially as they are used to design the safety margins of reactors.Even for major fissile isotopes such as235U and239Pu, discrepancies still appear between the main nuclear data library. In 2018, CEA launched the ALDEN project (Average Lifetime of DElayed Neutrons), gathering several laboratories (CEA/DES and DRF, CNRS/LPSC, LP2i and LPC Caen, ENSICAEN, and University of Caen). This collaboration aims at producing new precise DN macroscopic data for reactor applications.Three experimental campaigns were conducted in 2018, 2019 and 2021 on the thermal fission of233U,235U and239Pu at Institut Laue Langevin in Grenoble (France), giving high quality results. In February 2023, a new campaign was performed at the PTB Ion Accelerator Facility in Braunschweig (Germany), focusing on the fast fission of238U. In particular, the DN yield was measured for eight energies between 1.5 and 19 MeV. This paper presents the experimental setup, that was adapted to the new facility and neutron spectrum. The methodology of analysis is also detailed, and some preliminary delayed neutron decay curves are showed.

Neutron cross section measurements for BUC approaches

Criticality safety analysis is required at various stages of the back-end of the fuel cycle, i.e. reprocessing, transport, storage and disposal of spent nuclear fuel (SNF). To account for the reduction in reactivity due to fuel burnup, the Burn-Up Credit (BUC) concept was introduced. Evidently, this concept depends on the quality of nuclear data, in particular the absorption cross sections of some key nuclides. A dedicated programme has been established at the GELINA facility of the JRC-Geel to produce accurate cross section data and validate the evaluated nuclear data libraries for neutron interactions with fission fragments that are relevant for a BUC approach. In this work, cross section data for 103Rh and 155Gd are presented and the results are compared with the main evaluation libraries, showing good agreement in the thermal energy region with ENDF/B-VIII.0 and JEFF-3.3, but not with JENDL-4.0 for 103Rh.

Control rods’ position history effect on the Indonesian multipurpose research reactor RSG-GAS operation and safety parameters

The present work extends the analysis of the first working core of RSG-GAS to the burnup calculation. During the fuel management calculations of RSG-GAS, the control rods are usually assumed to be withdrawn from the core. Due to the exclusion of the control rod, it is hypothesized that the integral control rod worth shows significant differences compared to the experimental data. Therefore, a burnup calculation was performed in this study for the first core of RSG-GAS to evaluate the impact of operating control rods. The analyses were performed using continuous-energy Monte Carlo code, Serpent 2, and ENDF-B/VIII.0 nuclear data library for both cases with and without the inclusion of control rods. As a result, it is found that the calculated control rod worth, as well as the excess reactivity at the beginning of the cycle of the RSG-GAS second core from the burnup calculation considering the control rod position, agreed with the experimental data.

Radiation and Nuclear Physics Aspects of the Use of the Thorium Fuel Cycle in a Hybrid Fusion Facility

This paper presents the results of the experimental determination and computational simulation of the ambient dose equivalent rate for a metallic thorium cylindrical miniblock and the (n,2n), (n,f), and (n,γ) reaction rates in a thin 232Th metal foil irradiated with neutrons of the NG-24M generator spectrum. The ambient dose equivalent rate was determined by dosimeters-radiometers. The reaction rates were determined by the activation method using Ge spectrometers without destroying the irradiated samples. Computational simulations of ambient dose equivalent and reaction rates were performed, respectively, using the radiation transport codes PHITS, MCNP5, and KIR2, which use various nuclear data libraries: JEFF-3.2 and -3.3; JENDL4.0; ENDF/B-VII.0, -VII.1, and -VIII.0; ROSFOND; FENDL; and TENDL. The authors give an estimate of the 232U/233U relative accumulation upon natural thorium irradiation in a fusion facility blanket with defined neutron spectrum. The nonirradiated and irradiated thorium nuclide composition change simulation and visualization were performed using analytical solutions of an ordinary system of homogeneous linear differential equations describing nuclide transmutations.

Status of JADE, an open-source software for nuclear data libraries V&V

In the last couple of years, a combined effort between NIER, Università di Bologna and Fusion For Energy led to the development of JADE, a python-based open-source software for the Verification and Validation of nuclear data libraries. Nuclear data is fundamental for particle and radiation transport simulations which, in turn, are responsible for the evaluation of key quantities for fusion-related machines design such as nuclear heating, DPA, particles production and dose rates. The aim for the project is to offer standardization and automation to the V&V process of data libraries in order to speed up their release cycles and, at the same time, improve the quality of the data. JADE takes advantage of MCNP for the particles and radiation transport simulations and, even if it is potentially applicable to the whole nuclear industry, a particular focus on fusion applications is obtained through the selections of the default benchmarks that have been implemented. The code was recently made publicly available to the community and the status of its development is summarized in this work. The more important features and benchmarks (both computational and experimental) are described, together with a brief discussion on the major case studies where JADE has been used. Lastly, the current strength and limitations of the tool are evaluated and the foreseen future developments for the project are outlined.

Farklı seviye yoğunluk modelleri kullanılarak reaksiyon eşik değerinden 20 MeV’e kadar Paladyum izotoplarının (n,2n), (n,p) ve (n,) tesir kesitlerinin hesaplanması

Bu çalışmada, Paladyum izotoplarının (102,104,106,108,110Pd) (n,2n), (n,p) ve (n,) reaksiyon tesir kesitleri Empire-3.2.3 (Malta) istatistiksel model kodu kullanılarak reaksiyon eşik değerinden 20 MeV’e kadar nötron enerjilerinde hesaplanmıştır. Hesaplamalar, dört farklı nükleer seviye yoğunluğu modeli (Genelleştirilmiş Süper Akışkan modeli, Gelişmiş Genelleştirilmiş Süper Akışkan modeli, Gilbert-Cameron modeli ve Hartree-Fock-Bogoliubov mikroskobik modeli) kullanılarak yapılmıştır. Hesaplardan elde edilen sonuçlar, EXFOR kütüphanesinde bulunan deneysel verilerle ve ENDF kütüphanesinden alınan ENDF/B-VIII.0 (USA,2018), TENDL-2019 (TALYS, 2019) ve JENDL-5 (Japan,2021) değerlendirilmiş verilerle karşılaştırılmıştır. 14.5 MeV civarında deneysel ve değerlendirilmiş verilerle karşılaştırıldığında, sonuçlar genel olarak tutarlı bir uyum göstermektedir.

Inter-Code Comparison of Computational VERA Depletion Benchmark Using OpenMC, OpenMC-ONIX and DRAGON

This research focuses on the comparative analysis of the PWR fuel assembly based on VERA depletion benchmark problems using community-developed open source Monte Carlo code OpenMC, python based burnup code system ONIX (a coupling interface for Monte Carlo code OpenMC), and deterministic DRAGON code. The depletion analysis was performed using OpenMC and ONIX with ENDF/B-VII.1 nuclear data library, and DRAGON with SHEM-361 based DRAGLIB format library (ENDF/B-VII.1). The code-to-code analysis on the evolution of k eff , atom number density, and power distribution as a function of burnup has been performed and the result shows a good agreement with the maximum difference within 200 pcm at EOC. However small discrepancy around 90 pcm has been observed in k eff calculated by DRAGON compared to OpenMC in the presence of integral fuel burnable absorbers (IFBA). The above-mentioned codes have been validated successfully for the first time against PWR fuel assembly based on VERA depletion benchmark problems. It can be concluded that initial implementation of these codes at the Department of Nuclear Science and Engineering under Military Institute of Science and Technology, Dhaka, was successful and that further research works are to be performed to utilize these codes for depletion/neutronics calculation of existing 3MW TRIGA Mark-II research reactor and VVER-type power reactor that is to be commissioned in Bangladesh.

Neutronics analysis of UN-PuN fuel for 300MW pressurized water reactor using SRAC-COREBN code

Nuclear Power Plants (NPP) is one of alternative energy that can be used to replace fossil energy. NPP is a clean energy that doesn't emit CO2 as a residue from the power plant. Pressurized Water Reactor (PWR) is the most widely used commercial reactor to date. This reactor usually uses UO2 as fuel with enrich U-235 and produce plutonium as a waste nuclear. In this study, the fuel uses natural uranium and the addition of plutonium from the spent fuel of PWR, calling it Uranium-Plutonium Nitride (UN-PuN) fuel. Plutonium from spent fuel PWR (U-fueled) consists of Pu-238, Pu-239, Pu-240, Pu-241 and Pu-242, which Pu-239 and Pu-241 are fissile material. The aim of using plutonium is to reduce the amount of nuclear waste and prevent the use of plutonium as a nuclear weapon. The addition of neptunium and protactinium is also carried out in the fuel. Neptunium is one of the minor actinide materials which have high toxicity, so add Np-237 it is very useful to reduce the amount of neptunium in the world. And also, the addition of Np-237 and Pa-231 aims to decrease the keff value both in the Beginning of Life (BOL) and End of Life (EOL) in the reactor. This research calculates the neutronic calculations for PWR use SRAC2006 with COREBN Code and Japanese Evaluated Nuclear Data Library 4.0 (JENDL-4.0) for the nuclear data library. The COREBN code is a calculation method based on interpolation of macroscopic cross-section and finite-difference diffusion theory. The optimum design reached the maximum k-eff value 1.017077 and the maximum power density was 58.47 W/cc. The results obtained are the excess reactivity value of 1.68 % smaller than previous study which produces an excess reactivity value of 11.94 %. So this design has the potential to be applied in the PWR design

Evaluation of OpenMC calculations coupling with PHITS, FLUKA, and GEANT4 for ADS study

The Accelerator-Driven Subcritical system (ADS) is a subcritical reactor driven by the spallation neutrons generated in a spallation target under the bombardment of proton beam from an accelerator. To study ADS neutronics, the external spallation neutrons should be investigated carefully for coupling with a reactor core. This work focuses on the investigation into the OpenMC calculations by using different nuclear data libraries coupling with different high-energy Monte Carlo programs such as PHITS, FLUKA, and GEANT4 for ADS neutronics simulation. MCNPX direct calculations is performed as the baseline to compare to coupling-calculations. Firstly, the information of leaking neutrons from a spallation target are calculated with different codes and physical models to evaluate difference from the direct calculations to the coupling-calculations. Secondly, the neutronics characteristics of OECD-ADS model are calculated by using OpenMC calculations coupling with a surface-source from three high-energy codes to compare to MCNPX direct calculations. The neutronics parameters of mono-energy neutron for OECD-ADS evaluated from OpenMC and MCNPX with ENDF/B-VII.0 data library or only from OpenMC with different libraries are found to be in good agreement. The effect of high-energy neutrons is evaluated to be difference of around 25% in beam requirements by MCNPX direct calculations with INCL model and PHITS-OpenMC coupling-calculation with ENDF/B-VII.0 below 20 MeV. The differences in beam requirements between different coupling-calculations with the same library are compared. Moreover, the discrepancies in beam requirements from PHITS-OpenMC with JENDL-4.0/HE from 20 MeV to 200 MeV are around 15%. Finally, both neutron profiles and neutron spectra from coupling-calculations agree well with these from MCNPX direct calculations outside the coupling region of target.