Nuclear Data Libraries Research Articles

The description of giant dipole resonance key parameters with multitask neural networks

Giant dipole resonance (GDR) is one of the fundamental collective excitation modes in nucleus. Continuous efforts have been made to the evaluation of GDR key parameters in different nuclear data libraries. We introduced multitask learning (MTL) approach to learn and reproduce the evaluated experimental data of GDR key parameters, including both GDR energies and widths. Compared to the theoretical GDR parameters in RIPL-3 library, the accuracies of MTL approach are almost doubled for 129 nuclei with experimental data. The significant improvement is largely due to the right classification of unimodal nuclei and bimodal nuclei by the classification neural network. Based on the good performance of the neural network approach, an extrapolation to 79 nuclei around the β-stability line without experimental data is made, which provides an important reference to future experiments and data evaluations. The successful application of MTL approach in this work further proofs the feasibility of studying multi-output physical problems with multitask neural network in nuclear physics domain.

Neutron Field Shaping Using Graphite for Reaction Rate Measurements

Abstract Neutron field shaping is the suitable method for validation of cross section in various energy regions. By increasing the share of neutrons of a certain energy interval and decreasing the share of other, a reaction becomes more sensitive to selected neutrons. As a result, reaction cross section can be validated in selected energy regions more precisely. The shaping can be carried out by both neutron filters which are materials with high absorption in some energy region, or by diffusion material changing the shape of neutron spectra by means of slowing down process. In the presented experiments, the neutron field of the light reactor 0 (LR-0) research reactor was shaped by both using graphite blocks inserted into the core and Cd cladding for increasing the epithermal reaction rate share in total reaction rates. The calculations were carried out with the Monte Carlo N-Particle Transport Code 6 (MCNP6) code and the most recent nuclear data libraries. The results in the pure graphite neutron field are in good agreement; in case of Cd cladding, significant discrepancies were reported. In case of the 23Na(n,γ)24Na reaction, overestimation by about 14% was reached in International Reactor Dosimetry and Fusion File (IRDFF-II), results in other libraries are comparable. In case of 58Fe(n,γ)59Fe, the overestimation as high as 18% is reported in IRDFF-II. For 64Zn(n,γ)65Zn reasonable agreement was reached in evaluated nuclear data file (ENDF/B-VIII), where discrepancies in pure graphite neutron field or in case of Cd cladding are about 10–15%.

JENDL/DEU-2020: deuteron nuclear data library for design studies of accelerator-based neutron sources

ABSTRACT Intensive fast neutron sources using deuteron accelerators have been proposed for various applications. To contribute to the design study of such neutron sources, a deuteron nuclear data library for 6,7Li, 9Be, and 12,13C up to 200 MeV, JENDL/DEU-2020 is developed. Evaluation of JENDL/DEU-2020 is performed by employing the code system DEURACS with particular attention to neutron production data. Toward the evaluation of JENDL/DEU-2020, some modifications are made to DEURACS. Validation of the library is performed through simulation with the Monte Carlo transport calculation codes. From the simulation, it is shown that calculation results based on JENDL/DEU-2020 reproduce measured neutron production data well in incident energies up to 200 MeV. The new library is expected to make a large contribution to diverse design studies of deuteron accelerator neutron sources.

Verification of CENDL-3.2 Nuclear Data on VENUS-3 Shielding Benchmark by ARES Transport Code

The recently released CENDL-3.2 nuclear data library is deemed as an important achievement in the field of nuclear data research in China. To verify the applicability of the library to the shielding calculation of PWR and analyze the influence of multigroup cross-section parameters on the shielding calculation, ARES-MACXS module is used to process the MATXS format multigroup library based on CENDL-3.2 to generate multigroup working cross sections for PWR shielding calculation. VENUS-3 experimental facility has a clear and complete geometry. It is often used to test the ability of the advanced transport calculation method of calculating RPV fast neutron flux and to evaluate the accuracy of cross-section library. Different cross-section parameters are chosen for ARES to calculate VENUS-3 benchmark, and equivalent neutron flux of 58Ni(n,p)58Co, 115In(n,n′)115mIn and 27Al(n,α)24Na detectors is calculated according to the data provided by the benchmark report. The numerical results demonstrate that almost all the relative deviations between the calculated results and the experimental results are within 20%, which satisfies the requirement of shielding calculation. CENDL-3.2 is suitable for PWR shielding calculation. The comparison of various cross-section parameters results indicates that multigroup cross-section parameters have large effects on the transport calculation results.

Natural Uranium Utilization in FUJI-U3 Molten Salt Reactor

Molten Salt Reactor (MSR) is one of the next-generation nuclear power systems known as Generation IV. In this paper, the FUJI-U3 MSR type is designed for neutronic analysis that natural uranium is used as a fuel. The standard reactor core model is adopted due to the flux flattening effect in the active core, as reported from previous research. FLiBe eutectic is used both as the coolant salt and moderator in the reactor core. The neutronic calculation is performed using the SRAC 2006 program: PIJ for pin cell calculation and Citation for core calculation. JENDL 4.0 is used as a nuclear data library that provides the cross-section data of the nuclides. The neutronic parameter results, such as the effective multiplication factor, conversion ratio, neutron spectrum, and power density, are obtained from the calculation. The reactor can achieve the criticality condition with a minimum loading Uranium Tetrafluoride (UF4) of 3.6%mol for 2000 effective full power days (EFPD). The results calculation provide some parameter survey for the research and the future development of MSR.

Excess reactivity optimization on High Temperature Gas Reactor (HTGR) design for remote area

East Nusa Tenggara (NTT) is an island that has an electrification ratio of 85.84%. Small power nuclear power plant could be applied in such a remote area, and High-Temperature Gas Reactor (HTGR) is one of the potential options. Based on data obtained from the Central Statistics Agency (BPS), an HTGR was designed. The purpose of the present study was to find the optimal excess reactivity for HTGR using the basic model of Japan’s HTTR (High-Temperature Engineering Test Reactor). The calculation was conducted using the Standard Reactor Analysis Code System (SRAC) with the Japanese Evaluated Nuclear Data Library (JENDL) 4.0 as the nuclear data library. Calculations for modified HTTR geometry and fuel configuration was performed and analyzed. As a result, for 1.6 times the HTTR geometry model, (7.00 - 8.75)% fissile enrichment with the addition of 0.065% B4C material is the best configuration to obtain the optimal excess reactivity. Meanwhile, for 1.5, 1.4 and 1.3 times HTTR geometry the best configuration values were (5.00 - 6.75)% fissile enrichment with 0.035% B4C, (6.00 - 7.75)% fissile enrichment with 0.045% B4C and (7.00 - 8.75)% fissile enrichment with 0.050% B4C, respectively.

Comparatives Studies in Molten Salt Reactor FUJI-U3 with Various Power

Indonesia has a resource of nuclear fuel which spread around the country, especially in Bangka Belitung, West Kalimantan, West Sulawesi, and Papua. Therefore, Indonesia has a potential in the nuclear energy development. The Generation IV nuclear power system has been developed by the Generation IV Forum (GIF) with its benefits such as highly economical, enhanced safety, low waste production, and proliferation resistant. MSR is one of Generation IV nuclear power system which uses molten salt as a fuel and a coolant. Based on RUPTL (Electricity Supply Business Plan) PLN 2018-2027, Indonesia needs 100 MWe power harness from nuclear energy. This study focuses on a FUJI-U3 reactor, a type of MSR which developed in Japan. Power operation varied from 120-150 MWth 60-100 MWe and using FLiBe as molten salt. The calculation was performed with PIJ and CITATION modules in SRAC 2006 and JENDL 4.0 as nuclear data library.

Conservative covariance for general-purpose nuclear data evaluation

General-purpose evaluated nuclear data libraries requires the production of covariance data that can be safely used for any applications. Propagated uncertainty must be conservative regardless on which specific nuclear data an application is sensitive to. In this paper, the usual analytical and Bayesian marginalization techniques used for the production of covariance data are reviewed, their mathematical justification and practical implementation are discussed. An alternative technique is also presented, it is based on a spectral decomposition of the posterior Bayesian correlation matrix. This new method is shown to be more valid than the usual analytical and Bayesian marginalization techniques and exhibits the required properties for general-purpose applications. Several options in the spectral decomposition method are detailed, allowing a “strict conservatism” to a “best-estimate” justified when experimental data are well reproduced. These options are illustrated on a test-case and compared with the usual methods. The spectral decomposition method is also shown to be able to identify outliers in experimental data.

Study on Control Rod Reactivity of Small Pebble Bed Reactor with Wallpaper Fuel Design

The control rod reactivity is a crucial reactor parameter designed to shut down the reactor and to control the safety of reactor core under any operation conditions. This paper discusses the control rod reactivity of small pebble bed reactor with wallpaper fuel design. The geometric configuration of RDE was selected for the reactor core model. A number of simulations with Monte Carlo n-particle transport code MCNP6 based on evaluated nuclear data library ENDF/B-VII were employed to investigate the total and individual control rod worth and reactivity in stuck rod condition. In the simulation model, the total amount of uranium per pebble was kept constant at 5 g by varying the hexagonal lattice pitch calculated for different radius of the central graphite zone. The simulation results show that the total control rod achieves the best effectiveness on the central graphite radius of 1.0 cm. The highest worth of individual rods in the wallpaper fueled core are almost greater than that of standard fueled core. The stuck rod criterion in all simulation results confirm that, the failure of one rod with highest worth will not prevent the control system from shutting down the reactor. These results conclude that selecting a central graphite radius of wallpaper fuel is one of the strategies to be carefully considered and a radius of 1.0 cm is the best one in controlling reactor safety and ensuring the control system could shut down the reactor safely.

Analysis of water ingress accident in the RDE core using MCNPX

The 10 MW Indonesia’s Reaktor Daya Eksperimental (RDE) is a Pebble Bed High Temperature Reactor-type and to be constructed in Serpong Nuclear Zone, Puspiptek, Indonesia. The reactor applies helium gas coolant, graphite moderator, and low uranium enriched fuel (17%). A water-ingress accident may take place in the RDE core and such the accident is caused by a pipe break in the steam generator system of the reactor and hence the accident influences the reactivity of the core. To begin with, an initial core of the RDE should be modelled and the model contains the identification of material zones in the reactor core. To estimate the reactivity change due to the water-ingress accident in the initial RDE core, a void core condition from any coolant should be firstly deliberated. The core reactivity change is then estimated based on effective multiplication factor (k eff) as a function of water density which comes into void cells among pebble containing helium coolant in the RDE core and that of the variation of 235U enrichment. The data library of thermal scattering S(α,β) as a function of temperature has been applied. The data base of thermal scattering for graphite was also utilized to comprehend the validity calculation using MCNPX. The ENDF/B-VII continuous energy nuclear data library was also smeared for all calculations at temperature of 900 0K. For the calculation using MCNPX, two options of KCODE and KSRC were taken into account to estimate k eff of the RDE core. To achieve the accuracy of keff in the RDE reactor core, the amount of 20,000 neutrons per cycle for 150 non-active cycles and 650 active cycles was taken into account. The results showed that water-ingress accident in the RDE core is a dynamical behaviour which may result in reactivity core change of the RDE core. Finally, from all caclulated results taking into account the varietes of water density and 235U fuel enrichment, the RDE reactor core is totally safe in the event of water-ingress accident ensuing during the RDE reactor operation.

Critical analysis of light-water moderation UO2 and uranium-plutonium mixed oxide lattices by continuous-energy Monte Carlo calculation

ABSTRACT Critical analysis of 120 light-water moderation lattices composed of 2.6 wt% UO2 fuel rods or 3.0 wt%-PuO2 uranium-plutonium mixed oxide (MOX) fuel rods was performed using a continuous-energy Monte Carlo calculation code to validate the JENDL-4.0 nuclear data library. As a result, the average neutron multiplication factor (keff ) of the 35 UO2 lattices (hydrogen-to-heavy-metal atomic ratio (H/HM): 4.3–8.7) was 0.99930 with the uncertainty of 0.00005 (1 σ), and that of the 85 MOX lattices (H/HM: 12.0–27.6) was 0.99856 with the uncertainty of 0.00006 (1 σ). The keff values of the UO2 lattices partly showed a decreasing trend with H/HM, while those of the MOX lattices an increasing trend. The keff values of the MOX lattices presented an increasing trend with the measurement dates. Assuming that the trend was relating to the uncertainties in the neutron capture cross-section of 241Am, a simulation to modify the cross-section was performed by increasing the atomic number densities of 241Am in the MOX fuel. It indicated that the neutron-flux-average capture cross-sections of 241Am for the MOX lattices should be increased by 10% to 20% from those calculated by the cross-section in JENDL-4.0.

Waste expectations of fusion steels under current waste repository criteria

During operation fusion reactor components will be exposed to long periods of neutron irradiation. As such, a reactor’s structural steels will become activated and need to be disposed of as radioactive waste. Previous studies have shown that such wastes can struggle to meet low level waste (LLW) requirements meaning that costly geological disposal may be required. In order to explore the waste expectations of steels from European DEMO-like fusion reactors, several radioactive waste management systems have been investigated. This includes their LLW criteria, currently available disposal sites and planned future developments. This information was used to analyse the results of DEMO-like inventory simulations of potential reactor steels. The simulations were performed with the inventory code FISPACT-II and the TENDL2017 nuclear data library. The results suggest that when steels are exposed to near plasma neutron fluxes they will struggle to meet the majority of LLW requirements. For lower neutron fluxes, typical of reactor containment vessels, the waste expectations can be more positive, with several steels able to meet some low level criteria. It can be concluded that steels should not be expected to be consistently internationally classified as LLW 100 years after reactor shut down. As all activated fusion waste cannot be disposed of in a single location, it is recommended that waste disposal strategies are included in any fusion reactor proposal before construction begins. These strategies need to align with the radioactive waste regulations the proposed reactor will be subject to.

Neutronic and burn‐up calculations of the ( ThO2‐UO2 ) pin cell benchmark using DRAGON5 and MCNP6 .2 codes with ENDF / B‐VIII .0 nuclear data library

A (ThO2-UO2) fuel pin benchmark, consisting of 75 w/o Th and 25 w/o U, was utilized for analysis of depletion and inventory of the fuel for the pressurized water reactor (PWR) pin-cell model. To perform this study and evaluate cross-section data, the new version VIII.0 of the ENDF/B nuclear data library was utilized to generate a data set in ACE (A Compact ENDF) format at different temperatures (583.1, 600, 621.1, and 900 K), utilizing NJOY2016 and MAKXSF6.2 codes to process the data at the specific temperatures. In this paper, the infinite multiplication factor (kinf), neutron energy flux spectrum (utilizing 172 energy groups), the concentrations and activities of the most important products of fission, the actinide radionuclides accumulated in the pin-cell, and the total radioactivity were calculated and compared, utilizing the MCNP6.2 and DRAGON5 codes. Furthermore, the burn-up (BU) dependent behavior of the PWR thorium pin-cell was validated and compared with the reference results obtained utilizing the CASMO-4, MIT-MOCUP, and INEEL-MOCUP codes. The results of this study show good agreements between all codes analysed.

MRCsC: A user-friendly software for predicting shielding effectiveness against fast neutrons

A new user-friendly software called MRCsC was developed to accurately and precisely predict macroscopic effective removal cross-section, ΣR, (in cm−1) of fast neutrons for different shielding materials. The program includes a revised calculation model and the latest data published by Evaluated Nuclear Data Library “ENDF/B-VIII”. Experimental data and NXCom results were used to comparatively and statistically assess MRCsC predictions. The results reveal that both MRCsC and NXCom programs provide accurate and precise results when compared with the experimental data and that only MRCsC results are statistically equal to the experimental ones to the 95% confidence.

Analysis of the excess reactivity and control rod worth of RSG-GAS equilibrium silicide core using Continuous-Energy Monte Carlo Serpent2 code

This present work is a continuation of our previous work to verify the RSG-GAS equilibrium silicide core parameters. Earlier, the burnup of 22 fuel elements in the 88th equilibrium core has been measured, and it showed a good agreement between the experimental and calculation results. In this paper, several important neutronics parameters documented from the experiment at the same core, such as the excess reactivity, integral control rod worth, and total control rod worth, are confirmed using the full core calculation results of Monte Carlo Serpent2 code in conjunction with ENDF/B-VII.1 and the most recent ENDF/B-VIII.0 nuclear data libraries. The excess reactivity and integral control rod worth were measured by the positive and negative reactivity compensation method. It is worth mentioning that the measured integral control rod worth hasn’t been verified because the in-core fuel management code of RSG-GAS (BATAN-FUEL) solves only for a 2-dimensional configuration. The fuel element burnup at the end of cycle calculated by BATAN-FUEL and Serpent2 is also compared and discussed. The 3-D Serpent2 model of RSG-GAS first core is modified, and several improvements in the models are proposed, including the explicit modeling of 6 neutron beam tubes in the beryllium block reflector. It is found that the calculated integral control rod worth has a good agreement with the experiment, while Serpent2 overestimates the excess reactivity by a maximum of 661 pcm. The impact of nuclear data is insignificant for the selected core parameters.

Shielding design and neutronics calculation of the GDT based fusion neutron source ALIANCE

This paper presents the high flux neutron shielding design and extensive neutronics calculations of GDT based fusion neutron source ALIANCE. Neutron distribution of ALIANCE is strongly inhomogeneous along the axis: significant portion of the neutron flux is generated near the two mirrors, while the rest of it is spread over the remaining central volume of plasma. The shielding design includes 40 cm stainless steel as the main shielding layer and an additional 5 cm tungsten carbide shielding layer at mirror plugs to protect superconducting coils from neutron damage and reduce nuclear heating. The simulations have been carried out by using Monte Carlo transport code SuperMC with nuclear data library FENDL 3.1. Results show that the nuclear heating on the mirror coils can be reduced by more than two thirds with additional tungsten carbide shield, and fast neutron fluence by 30 %. The highest nuclear heating and the highest fast neutron fluence zones are located at the mirror coils, and the values are about 300 W/m3 and 9 × 1018 n/cm2 respectively, which meets the threshold of ITER superconducting coils. The specific activities of shielding layers are of order of 1012 Bq/kg. The structural materials’ specific activities will decrease to 4 × 1011 Bq/kg in one year after shutdown, and their decay heat will quickly drop below 2 kW/m3 after one day. Besides, all the structural materials of ALIANCE can be recycled by different recycling technologies. The modeling and calculations reported in this paper will be beneficial for the pre-conceptual engineering design of ALIANCE.

IDENTIFICATION OF REACTOR PHYSICS BENCHMARKS FOR NUCLEAR DATA TESTING: TOOLS AND EXAMPLES

Measurements of reactor physics quantities aimed at identifying the reactivity worth of materials, spectral ratios of cross-sections, and reactivity coefficients have ensured reactor physics codes can accurately predict nuclear reactor systems. These measurements were critical in the absence of sufficiently accurate differential data, and underpinned the need for experiments through the 50s, 60s, 70s and 80s. Data from experimental campaigns were routinely incorporated into nuclear data libraries either through changes to general nuclear data libraries, or more commonly in the local libraries generated by a particular institution or consortium interested in accurately predicting a specific nuclear system (e.g. fast reactors) or parameters (e.g. fission gas release, yields). Over the last three decades, the model has changed. In tandem access to computing power and monte carlo codes rose dramatically. The monte carlo codes were well suited to computing k-eff, and owing to the availability of high quality criticality benchmarks and these benchmarks were increasing used to test the nuclear data. Meanwhile, there was a decline in the production of local libraries as new nuclear systems were not being built, and the existing systems were considered adequately predicted. The cost-to-benefit ratio of validating new libraries relative to their improved prediction capability was less attractive. These trends have continued. It is widely acknowledged that the checking of new nuclear data libraries is highly skewed towards testing against criticality benchmarks, ignoring many of the high quality reactor physics benchmarks during the testing and production of general-purpose nuclear data libraries. However, continued increases in computing power, methodology (GPT), and additional availability reactor physics experiments from sources such as the International Handbook of Evaluated Reactor Physics Experiments should result in better testing of new libraries and ensured applicability to a wide variety of nuclear systems. It often has not. Leveraging the wealth of historical reactor physics measurements represents perhaps the simplest way to improve the quality of nuclear data libraries in the coming decade. Resources at the Nuclear Energy Agency can be utilized to assist in interrogating available identify benchmarks in the reactor physics experiments handbook, and expediting their use in verification and validation. Additionally, high quality experimental campaigns that should be examined in validation will be highlighted to illustrate potential improvements in the verification and validation process.

BEAVRS BENCHMARK EVALUATION WITH CASMO5 AND SIMULATE5

The MIT BEAVRS benchmark problem, which was primarily setup for the verification and validation of high-fidelity tools that have coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion models, has also found extensive usage in the reactor physics community for validating core analysis tools. The primary purpose of this paper is to provide an accurate, comprehensive evaluation of the BEAVRS benchmark with CASMO5 and SIMULATE5 codes. The CMS5 calculated results for low-power physics tests (hot zero power critical boron, control rod worth and isothermal temperature coefficients) and full power operation (boron let-down and flux map reaction rate distributions) are compared to plant measured data provided in the benchmark specification. The CMS5 model, using ENDF/BVII.1 nuclear data library, predicts HZP critical boron concentration for all-rods-out conditions within 10 ppm for Cycle-1, and 25 ppm in Cycle-2; the control rod worth is predicted with a difference of 0.7% ± 3.8%, where the maximum difference is less than 10%. For the core follow calculations at the hot full power condition, the average difference in predicting the critical boron concentration is less than 20 ppm. In addition, the radial and nodal reaction rate distributions are predicted with a mean difference of about 1.6% and 3.8%, respectively. The CMS5 calculations are repeated using the most recent ENDF/B-VIII.0 library. No significant difference is observed in predicting measured plant parameters with different nuclear data libraries. Additionally, the impact of various modeling options, which are typically employed with nodal diffusion codes, on the predictions of important core parameters are presented as part of the benchmark evaluation.

VALIDATION OF THE WIMS/PANTHER EMBEDDED SUPERCELL METHOD

The WIMS/PANTHER Embedded Supercell Method (ESM) provides a significant improvement in prediction accuracy for radial power distributions for PWR reactors compared to the standard “two-step” approach, without the need for a significant increase in computational resource. Recent papers at PHYSOR conferences have outlined the details of the method and demonstrated its operation, and the accuracy improvements possible, by means of benchmarking calculations.This paper applies the method to a 4-loop PWR in the U.K, and three PWRs (3-loop and 2-loop) in Belgium. Comparisons are made against measured data from the start-of-cycle physics testing performed for each cycle, and power-shape measurements collected during the cycle using a conventional “two-step” nodal reactor solution, and with the ESM. All results will be presented with the JEF2.2 nuclear data library, for ease of comparison between the methods and previously reported results, although the effects of more modern evaluations will be commented upon.The benchmark calculations referred to above studied a challenging MOX/UO2 benchmark core akin to an SMR. The four reactors studied here include conventional UO2 only core designs and cycles with UO2/MOX mixed cores. A variety of boron-and gadolinium-based burnable absorbers are also present. The data is used to show that the method both operates successfully for real reactor problems, and delivers improvements in the prediction accuracy of measured parameters.

NUCLEAR DATA V&V ANALYSIS FOR FUSION APPLICATIONS: INTEGRAL BENCHMARKS AND DECAY DATA

A reliable estimation of the operational parameters is one of the primary concerns in the design of magnetic fusion devices such as ITER and DEMO. Methods of diagnostics and control over the critical plasma parameters determining its stability and efficiency rely on the high-energy neutron field monitoring. Extreme operational environment, such as high-energy neutron flux, electromagnetic radiation, and high temperatures might reduce the performance of the detector systems. Therefore, research and development activities in detector prototyping are carried out to address this problem. To predict the performance of the detector materials, simulations using the latest releases of the nuclear data libraries as input for the inventory codes are carried out. This paper describes the latest validation and verification (V&V) benchmark exercise for FISPACT-II & TENDL-2017 based on the fusion decay heat measurements performed at the Japanese FNS facility for the materials in the diagnostic components for the radiation measurements. The breakdown of decay-heat contributions from individual radionuclides have been employed to interpret the simulated results, benchmark the data against the experimental measurements, and revise the neutron-induced reactions cross-section and decay data for the associated radionuclides for the upcoming release of the TENDL-2019 nuclear data library.