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• China Nuclear Data Center developed China Evaluated Photonuclear Data Library CENDL/PD-b for nuclear engineering needs. • CENDL/PD-b contains 266 materials with photon incident energy range from photonuclear threshold energies to 200 MeV. • The nuclei in CENDL/PD-b are evaluated by the Chinese theoretical model codes to describe the experimental data available. • The global model parametersare applied to generate the photonuclear data, which exhibit good data systematic in physics. • For the first time this work systematically tests the CENDL/PD-b. This study generates the ACE-formatted library for the test edition of China Evaluated Photonuclear Data Library CENDL/PD-b by using NJOY2016, and verifies the applicability of the ACE-formatted library against the simple transport problem. This study also validates the performance of CENDL/PD-b on Barber and George benchmark, the photo-neutron yields of C, Al, Cu, Ta, and Pb targets were calculated by using the Monte Carlo code. The verification and benchmarking results show that CENDL/PD-b ACE-formatted library can be applied for Monte Carlo transport simulation, and the photo-neutron yields of C, Cu, Ta are in good agreement with the experimental values, all within the uncertainty of the experimental values. However, the photo-neutron yields of Al and Pb are lower than the experimental values, and more experimental validation for those photonuclear data are still needed in future. [ABSTRACT FROM AUTHOR]

This work covers the operational use of the Aerojet General Nucleonics Model 201 Reactor (AGN-201M) at the University of New Mexico (UNM) in Albuquerque, New Mexico. The AGN-201M is a research and test reactor that operates at 5-watts (thermal) and utilizes HALEU fuel encased in polyethylene fuel disks. Operational usage in hours and watt-hours for the 2021 and 2022 fiscal year is provided, as well as licensing data (including training usage), notable research projects, outreach, and educational use in the Nuclear Engineering program at UNM. All research projects described herein have published conference papers, journal articles, or in-progress conference papers, journal articles and benchmarks. [ABSTRACT FROM AUTHOR]

This work provides fully reproducible benchmark models of the Systems for Nuclear Auxiliary Power (SNAP) 8 Experimental Reactor. Validation benchmarks of the criticality configuration experiments under dry conditions with no coolant are presented. In addition, relevant reactivity worth experiments, and measurement of power distributions are validated and presented here. Discrepancies between modeled and experimental results are at most 110 pcm for fuel and poison worths, and critical configurations that do not manipulate control elements are within 50 pcm. Larger discrepancies found in reflector element worth experiments are due to experimental methodology which was noted as being simplified due to limited calculational capabilities at the time, while discrepancy in the control element manipulated critical configuration is due the absence of structural components for modeling simplification. Power distributions closely follow what was observed in experiment with expected peaking in reflecting elements. All models are well documented with cited references describing and justifying assumptions, simplifications, and adjustments to reproduce the models with any nuclear safety codes; model inputs and outputs are stored in the snapReactors GitHub repository. [ABSTRACT FROM AUTHOR]

• Using adding and doubling. • High precision with simple methods. • Convergence acceleration. • Pade approximants. In the past, the method of adding and doubling has received considerable attention as an effective approach to solving the linear transport equation. Here, we consider doubling for the problem of thermal creep in Rarefied Gas Dynamics. The method, called the Discrete/Discrete Ordinates (D/DO) method is in contrast to the well-known Analytical/Discrete Ordinates (A/DO) method. The D/DO method solves the fully discretized transport equation to demonstrate that straightforward numerical approximation also yields high quality numerical solutions. The method combines the fundamental forward Euler evaluation of the matrix exponential with Richardson extrapolation to generate a high quality benchmark. We then generalize the method using Padé approximants for the matrix exponential and present a 7-place benchmark. [ABSTRACT FROM AUTHOR]

• A transient analysis program HC-PK-CFD is developed. • Verification of HC-PK-CFD is performed based on a benchmark on the beam trip transients for XADS reactor. • It is demonstrated that the HC-PK-CFD program is capable and of enough accuracy to perform thermal-hydraulics/safety analysis problems. It is important to employ multi-physics coupling strategy in thermal-hydraulics/safety analysis of nuclear reactors. In general, it is based on couplings of system codes with neutron dynamics codes, which is incompetent to cope with multi-dimensional thermal-hydraulics/safety problems involved in liquid metal cooled pool-type reactors. To cope with it, Computational Fluid Dynamics tools are good choices which have attracted considerable attentions in reactor simulation these years. In order to apply a commercial CFD program in thermal-hydraulics/safety analyses, the research on couplings of Point Kinetics (PK) model and a fuel pin Heat Conduction (HC) model to the commercial CFD platform FLUENT abbreviated as HC-PK-CFD is performed in this study. Verifications are performed by the benchmark simulation on beam interruption transients for an accelerator-driven sub-critical reactor system. It is demonstrated that the results simulated by HC-PK-CFD agrees well with the ones by other ten multi-physics coupling codes participating in this benchmark. [ABSTRACT FROM AUTHOR]

• The MPACT library was developed using the SCALE and VERA XSTools codes. • Eight kinds of the benchmark problems were developed for verification. • The ENDF/B-VII.1 MPACT library was verified through a code-to-code comparison. The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3D whole core transport code being developed for the CASL toolset, known as the Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include a subgroup method for resonance self-shielding and a whole-core transport solver with a 2D/1D synthesis method. Depletion capability is supported by coupling SCALE-ORIGEN with MPACT. A new ENDF/B-VII.1 MPACT 51-group library has been developed by using the AMPX/SCALE and CASL-XSTools code packages based on the procedure introduced in the companion paper, "Development of the Multigroup Cross Section Library for the CASL Neutronics Simulator MPACT: Method and Procedure." The cross section library generation procedure, cross section library, and MPACT transport capability have been verified by performing various benchmark calculations and comparing the benchmark results to the continuous-energy (CE) Monte Carlo results. The code-to-code comparison for the benchmarks shows that the CASL neutronics simulator MPACT and its 51-group library are working reasonably. [ABSTRACT FROM AUTHOR]

Highlights • Benchmark for space-time coupled transport/depletion. • Analytical solution for fixed-source and eigenvalue problems. • Simple to implement and evaluate the analytical solution. • Benchmark has been implemented and demonstrated with MC21 code. • Convergence of coupled MC21 depletion calculations is assessed. Abstract Formal analysis of the convergence and stability of coupled neutron transport/depletion calculations is difficult due to the complicated nature of the nonlinear system of equations that must be solved. In this paper, we present a new benchmark problem for coupled neutron transport/depletion calculations along with its fully analytic solution. The benchmark problem – which we call the "Candlestick benchmark" because it models a progressive burnup from one end of a slab to another – is straightforward to implement and can be used to quantify the accuracy and stability of coupled transport/depletion calculations. Descriptions and solutions are provided for both a fixed-source and eigenvalue version of the benchmark. The Candlestick benchmark is a first-of-its-kind analytic solution to a nonlinear spatial depletion problem, and it is useful for the purpose of validating the solution methods used by reactor depletion solvers. As a demonstration of the usefulness of the Candlestick benchmark, we have used it to assess some of the convergence behaviors of Monte Carlo depletion calculations, including the effects of spatial discretization, temporal discretization, and stochastic error on the results of the calculation. [ABSTRACT FROM AUTHOR]

Highlights • A new benchmark for neutronic code validation for severe accidents in fast reactors. • The benchmark is based on analysis of the plutonium-fuel based core SNEAK-12B. • The analyzed experiments include core degradation and axial/radial material relocation. • The analysis is based on MCNPX2.7, Serpent-2.1.29, and ERANOS 2.4 codes. • The results include criticality values and axial distribution of reaction rates and flux. Abstract The present work details information (core geometry, material balance, and criticality measurements and calculations) regarding a new benchmark to be introduced to the international community, for dealing with neutronic code validation in the frame of the analysis of severe accidents in fast reactors leading to core degradation and material relocation. This specific benchmark is based on analysis of selected experiments performed at the Schnelle Null-Energie-Anordnung Karlsruhe (SNEAK). Unlike the previously analyzed SNEAK-12A core, which was loaded with enriched metallic uranium fuel, the core analyzed in this study, the SNEAK-12B core, was loaded with plutonium fuel to better represent future fast systems, and the experiments that were considered include fuel relocation and redistribution of structural material. In this paper, the experimental results are analyzed by computational tools such as MCNPX2.7 and Serpent-2.1.29 Monte Carlo codes, and the ERANOS 2.4 system code for deterministic calculations, all based on JEFF-3.1.1 nuclear data libraries. The paper provides a complete and detailed specification for the benchmark problem. Preliminary results of available experimental results ( k eff and axial distribution of reaction rates) are given and additional quantities are presented (such as axial flux distribution). The benchmark offers an excellent opportunity to validate calculation schemes for strongly heterogeneous configurations, in particular the preparation of homogenized self-shielded neutron cross-sections for deterministic core calculations, as well as leakage treatment in locally voided zones. The analyses of SNEAK experiments, presented in this paper, provide grounds for the design of innovative experimental capabilities in severe accident modeling in Zero Power Reactors (ZPR), such as the Zero-power Experimental PHYsics Reactor (ZEPHYR) project led independently by the Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA). This paper is completed by a comprehensive nuclear data sensitivity and uncertainty analysis of the reactivity coefficients and k eff in a companion paper. [ABSTRACT FROM AUTHOR]

The design of a molten salt reactor is largely based on CFD simulations. Phase change plays an important role in the safety of the reactor, but numerical modelling of phase change is particularly challenging. Therefore, the knowledge of the margin of error of CFD simulations involving phase change is very important. Relevant experimental validation data is lacking. For this reason, a numerical benchmark designed after the freeze valve is proposed. The benchmark consists of five stages, where with each step more complexity is added. The step-wise addition of complexity allows for pinpointing potential sources of discrepancy. Results were obtained with three different codes: STAR-CCM+, OpenFOAM, and DGFlows. The results were found to be largely consistent between the codes, however the addition of conjugate heat transfer introduced some discrepancies. These results indicate that careful consideration is needed when coupling conjugate heat transfer solvers with solid–liquid phase change models. • Numerical benchmark for modelling phase change in molten salt reactors was created. • Step-wise addition of complexity allows for pinpointing where models differ. • Three CFD codes were compared using phase change and conjugate heat transfer. • Conjugate heat transfer modelling approach introduced discrepancies. [ABSTRACT FROM AUTHOR]

This paper presents nuclear reactor transient measurements with thermal feedback and corresponding coupled neutronic–thermohydraulic simulations performed for the natural circulation cooled Training Reactor of Budapest University of Technology and Economics. During the experiments, the outlet temperatures of several fuel assemblies were measured with thermocouples, while the reactor power was simultaneously logged by an ex-core neutron detector. The numerical simulations were carried out with the internally coupled TRACE/Point Kinetics and TRACE/PARCS code systems. The parameterised group constant library of the PARCS neutronic model, as well as the reactivity coefficients of the Point Kinetic model, were generated by the Serpent 2 Monte Carlo code. In the former case, special attention was given to the applied diffusion coefficients. The 3D/1D TRACE calculation model was prepared with the SNAP graphical interface. It was shown that after applying the necessary corrections, related to the peculiar natural circulation flow and reactor parameters, the measurement and both simulation results showed good agreement in terms of peak nominal power and average assembly outlet temperature. Although several anomalies are presented and modelling details that need improvement are highlighted, this study suggests that the used code systems could be appropriate tools to reproduce and analyse transients performed in small-core, high-leakage, natural circulation cooled pool-type reactors. [ABSTRACT FROM AUTHOR]

Spent fuel pools (SFPs) are large structures equipped with storage racks designed to temporarily store irradiated nuclear fuel removed from the reactor. SFP severe accidents have long been considered as highly improbable since the accident progression is slow (in comparison with reactor core accidents) and let time to corrective operator actions. However, the accident at the Fukushima Dai-ichi Nuclear Power Plants has highlighted the vulnerability of nuclear fuels that are stored in SFPs in case of prolonged loss-of-cooling accidents and consequently renewed international interest in the safety of SFPs. In this context, the AIR-SFP project, funded by the Euratom 7th FP in the frame of the NUGENIA+ project, was launched in May 2015 with 15 participants. One of the objectives was to assess the applicability of Severe Accident (SA) codes, which were initially developed for reactor applications, to the calculation of transients in SFPs. To reach this objective, a benchmark, including a criticality risk assessment, was carried out. The degradation progression was computed by 14 participants with 6 different SA codes and 5 have participated to the criticality risk assessment. Main results are presented as well as conclusions that have been drawn concerning SA codes readiness to address these “beyond-scope” scenarios. [ABSTRACT FROM AUTHOR]

In the paper the development and validation of a 3-D computational model of VENUS-2 benchmark using Serpent 2 code is presented. This paper provides the comparative analyses of calculated results obtained with Serpent and MCNP against experimental results (measured data). Overall the Serpent results are in almost perfect agreement with MCNP results and in good agreement with experimental results. An average deviation of pin power distribution from the experiment in the case of 3.3 wt% UO 2 fuel is 1.5%, 2.1% for 4.0 wt% UO 2 fuel and 2.8% in the MOX fuel. [ABSTRACT FROM AUTHOR]

Nuclear data, such as neutron cross sections, fission yield data and decay data, are essential to perform neutron transport calculations. This data is constantly being refined as new measurements are performed and computational power increases allowing for the calculation of higher fidelity nuclear models. In this work the nuclear data libraries ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0 are benchmarked against experimental data obtained from the Research Reactor OPAL in Sydney, Australia. ENDF/B-VIII.0 is currently in the beta stage of testing, though all the criticality-sensitive data has been frozen. Calculations of critical configurations, control rod worths and burnup of 6 operating cycles of OPAL were performed. This new library yields results that are consistent with both the experimental data available and calculated data from previous work, with the results produced using the new library providing similar or improved agreement with experimental data when compared to other code benchmarks of the OPAL Reactor. Isotopic evolution of key species within the core over these 6 cycles is also presented, including important fission product poisons, fissile isotopes and fertile isotopes. While no experimental data exists on the specific composition of any spent fuel from these cycles, differences in the material inventory are noted between calculations using the new library and older ENDF/B evaluations. [ABSTRACT FROM AUTHOR]

The assessment of point kinetics models validity is particularly important for highly heterogeneous cores, such as the novel low void fast reactor concept ASTRID. This article presents a benchmark study between CEA and IAEC on the importance of spatial effects during unprotected transients in the ASTRID core. The analysis is performed using two different neutronics spatial models: one based on a deterministic, discrete ordinate neutronics code, while the other employs the Transient Fission Matrix approach, which allows for fast estimation of both reactivity and flux redistribution during the transient with a precision similar to that of the Monte Carlo code. Both models have been coupled to the thermal-hydraulics and applied to an ASTRID representative geometry. This application case is supposed to be sensitive to power redistributions. Even so, the analysis shows good agreement between spatial and point kinetics for unprotected loss of flow and unprotected top of power transients. Some small discrepancies between the results of both codes are found to originate from differences in the calculated spatial distribution of the neutronic reactivity feedback coefficients during stabilization, highlighting the importance of exact spatial description of neutronic feedback coefficients. [ABSTRACT FROM AUTHOR]

A new utility has been developed with extensive capabilities in identifying nuclide decay and transmutation characteristics, allowing for accurate and efficient tracking of the change in isotopic concentrations in nuclear reactor fuel over time when coupled with a transport solution method. This tool, named the Application Programming Interface for Depletion Analysis (APIDA), employs both a matrix exponential method and a linear chain method to solve for the end-of-time-step nuclide concentrations for all isotopes relevant to nuclear reactors. The Chebyshev Rational Approximation Method (CRAM) was utilized to deal with the ill-conditioned matrices generated during lattice depletion calculations, and a complex linear chain solver was developed to handle isotopes reduced from the burnup matrix due to either radioactive stability or a sufficiently low neutron-induced reaction cross section. The entire tool is housed in a robust but simple application programming interface (API). The development of this API allows other codes, particularly numerical neutron transport solvers, to incorporate APIDA as the burnup solver in a lattice depletion code or reactor core analysis code in memory, without the need to write or read from the hard disk. The APIDA code was benchmarked using several decay and transmutation chains. Burnup solutions produced by APIDA were shown to provide material concentrations comparable to the analytically solved Bateman equations – well below 0.01% relative error for even the most extreme cases using isotopes with vastly different decay constants. As a first order demonstration of the API, APIDA was coupled with the transport solver in the SERPENT code for a fuel pin cell depletion problem. Results show APIDA to be effective and efficient in solving lattice depletion problems, in addition to being successful in terms of portability for users to implement via the API. [ABSTRACT FROM AUTHOR]

A comprehensive set of transmutation chain benchmark problems for numerically validating methods for solving burnup equations was created. These benchmark problems were designed to challenge both traditional and modern numerical methods used to solve the complex set of ordinary differential equations used for tracking the change in nuclide concentrations over time due to nuclear phenomena. Given the development of most burnup solvers is done for the purpose of coupling with an established transport solution method, these problems provide a useful resource in testing and validating the burnup equation solver before coupling for use in a lattice or core depletion code. All the relevant parameters for each benchmark problem are described. Results are also provided in the form of reference solutions generated by the Mathematica tool, as well as additional numerical results from MATLAB. [ABSTRACT FROM AUTHOR]

• Impact of heterogeneous modeling on burnt fuel vector is evaluated. • The effects of fully heterogeneous model based burnt fuel composition on the core integral parameters is analyzed. • Using the respective assembly type burnt vector, instead of the generic burnt vector, shows a significant difference in assembly power fractions. Computational benchmarks are an important source for verifying the codes and modeling methodology in nuclear reactor analysis. The current study aims at utilizing the newly developed burnup capability of open source code OpenMC to perform analyses of the IAEA 10-MW MTR benchmark reactor. The whole core model developed for the benchmark is verified against the Serpent results from an earlier study. The effect of the volume-weighted homogenized (VWH) unit-cell model against a fully heterogeneous fuel assembly model to obtain burnup dependent number densities is determined. It was observed that the comparison of four factors (resonance escape, fast fission, thermal utilization, and reproduction factor) for two modeling options shows relative differences which increase with burnup. These differences translate to about a 350 pcm difference in the HEU case while the LEU core shows a relatively larger difference of about 500 pcm in integral core parameters. Root Mean Square (RMS) and maximum relative error in assembly power as large as 0.83 and 2 % respectively are seen for HEU and LEU cases. Although standard fuel element (SFE) and control fuel element (CFE) are geometrically significantly different, some studies have used VWH unit-cell based on SFE to generate and use number densities for CFE as well. The effect of this modeling choice is also analyzed. This led to a relative difference of more than 20 % in the concentration of Pu-239 for both HEU and LEU studies, with SFE based number density being overpredicted. Xe-135 concentration also shows about a 20 % difference for the HEU case. The difference in the whole core neutron multiplication factor, SFE vs SFE + CFE individual burnt fuel vector, is negligible. On the other hand, power RMS error and maximum relative difference for power fraction exhibit a maximum value of 0.84 and 2.24 respectively. This study can serve as a reference for future verification efforts using the IAEA 10-MW MTR benchmark reactor. [ABSTRACT FROM AUTHOR]

• The Combined Matrix Iterative Algorithm (MMPA-CMIA) based on the MMPA method is proposed for solving burnup equations. • The numerical precision of three algorithms based on the MMPA method is compared with that of 16th-order CRAM in detail. • The accuracy and efficiency of MMPA-CMIA are verified by the benchmarks. In this work, a new efficient solution algorithm named Combined Matrix Iterative Algorithm based on the Mini-Max Polynomial Approximation (MMPA) method is proposed for burnup calculations. As a new matrix exponential method in recent years, MMPA method has good numerical stability and computational accuracy, and all calculation are real number operations. As the same with Chebyshev Rational Approximation Method (CRAM) which is widely and successfully used in solving burnup equations, the method can also directly deal with rigid and sparse burnup matrices. Compared with the traditional solution algorithms based on MMPA method, the Combined Matrix Iterative Algorithm (MMPA-CMIA) proposed in this paper can effectively reduce the computational complexity and improve the computational efficiency after equivalently transforming the expression of the MMPA method. Then the MMPA-CMIA is used to theoretically demonstrate and analyze the solution efficiency of random matrices with different orders which has the similar distribution characteristics to burnup matrices. Compared with the 16-order CRAM method and the other two traditional solution algorithms based on MMPA method, the algorithm shows good acceleration effect and computational precision. Finally, the algorithm is used in the actual burnup benchmarks to verify and validate the accuracy and efficiency. The burnup matrices of different orders are constructed through detailed burnup chain and simplified burnup chain. The results show that the algorithm has good accuracy and numerical stability. And comparing with 16-order CRAM, the algorithm can save about 30% of computing time for solving burnup equations. It shows that the algorithm is a potentially powerful method for accelerating the solution of large-scale burnup calculations. [ABSTRACT FROM AUTHOR]

This work uses the 2-D C5G7 benchmark to verify the accuracy and efficiency of the MOCUM code, a parallel neutronics program based on the method of characteristics (MOC) for arbitrary core geometries. Compared to references, MOCUM k eff , assembly and pin power maximum percentage errors are 0.02%, −0.06%, and 0.64%, respectively. The parallel algorithm achieves speed-ups of more than 17 times, compared to the calculation speed using one thread. The sensitivity study of various MOC parameters reveals that the C5G7 benchmark problem requires 48 or more azimuthal angles. The strong flux gradient and the heterogeneous effects demand fine unstructured meshes to yield accurate flux solution. The simulation uses 258 million zones with an average mesh size of 0.016 cm 2 . The investigation of the polar angle quadrature indicates that Leonard polar angle performs slightly better and more than three polar angles are not necessary. In addition, parameter sensitivity study shows that coarse parameters are prone to introduce error to the neutron flux but not k eff . [ABSTRACT FROM AUTHOR]

Advances in reactor physics have led to the development of new computational technologies and upgraded cross-section libraries so as to produce an accurate approximation to the true solution for the problem. Thus it is necessary to revisit the benchmark problems with the advanced computational code system and upgraded cross-section libraries to see how far they are in agreement with the earlier reported values. Present study is one such analysis with the DRAGON code employing advanced self shielding models like USS and 172 energy group ‘JEFF3.1’ cross-section library in DRAGLIB format. Although DRAGON code has already demonstrated its capability for heavy water moderator systems, it is now tested for light water reactor (LWR) and fast reactor systems. As a part of validation of DRAGON for LWR, a VVER computational benchmark titled “Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel-Volume 3” submitted by the Russian Federation has been taken up. Presently, pincell and assembly calculations are carried out considering variation in fuel temperature (both fresh and spent), moderator temperatures and boron content in the moderator. Various parameters such as infinite neutron multiplication ( k ∞ ) factor, one group integrated flux, few group homogenized cross-sections (absorption, nu-fission) and reaction rates (absorption, nu-fission) of individual isotopic nuclides are calculated for different reactor states. Comparisons of results are made with the reported Monte Carlo (MCU) values of the benchmark. Maximum deviation of 1.8% in k ∞ is observed for variants with spent fuel and for the states with control rod whereas all the other results are in par with the results reported in the benchmark document. The few and multi-group macroscopic cross-sections and flux of all the nuclides also compare well with the benchmark results except for the 11 B macroscopic absorption cross section, which is further compared with the XNWLUP software. Inter-comparison of results with the generalized self-shielding model SHI of DRAGON code employing the traditional WIMSD formatted 172 group cross-section library has also been made to highlight the improvements made in computational schemes and cross-section library format. [ABSTRACT FROM AUTHOR]

The importance of computer simulations in the assessment of nuclear plant safety systems has increased dramatically during the last three decades. The systems of interest include existing or proposed systems that operate, for example, normal operation, in design basis accident conditions, and in severe accident scenario beyond the design basis. The role of computer simulations is especially critical if one is interested in the reliability, robustness, or safety of high consequence systems that cannot be physically tested in a fully representative environment. In the European 7th Framework SARNET project, European Commission (EC) co-funded from 2008 to 2013, the Phébus FPT3 experiment was chosen as a code benchmark exercise to assess the status of the various codes used for severe accident analyses in light water reactors. The aim of the benchmark was to assess the capability of computer codes to model in an integral way the physical processes taking place during a severe accident in a pressurised water reactor (PWR), starting from the initial stages of core degradation, fission product, actinide and structural material release, their transport through the primary circuit up to the behaviour of the released fission products in the containment. The FPT3 benchmark was well supported, with participation from 16 organisations in 11 countries, using 8 different codes. The temperature history of the fuel bundle and the total hydrogen production were well captured. No code was able to reproduce accurately the final bundle state, using as bulk fuel relocation temperature, the temperature of the first significant material relocation observed during the experiment. The total volatile fission product release was well simulated, but the kinetics were generally overestimated. Concerning the modelling of semi-volatile, low-volatile and structural material release, the models need improvement, notably for Mo and Ru for which a substantial difference between bundle and fuel release was experimentally observed, due to retention in the cooler upper part of the bundle. The retention in the primary circuit was not well predicted, this was due mainly the non-prototypic formation of a boron-rich blockage in the rising line of the FPT3 steam generator, simulated in the circuit as a single external cooled U tube. The deposition mechanism and the volatility of some elements (Te, Cs, I) could be better predicted. Containment vessel thermal hydraulics, designed in the experiment to be well-mixed, were well calculated. Concerning the containment aerosol depletion rate, only stand-alone cases (in which the input data were derived from experimental data) provided acceptable results, whilst the integral cases (in which the input data came from circuit calculations) tended to largely overestimate the total aerosol airborne mass entering the containment. The disagreement of the calculated total aerosol airborne mass in the containment vessel with the measured one is due to the combination of a general underestimation of the overall circuit retention and overestimation of fission product and structural material release. Calculation of iodine chemistry in the containment turned out to be a major challenge. Its quality strongly depends on the correct prediction of chemistry speciation in the integral codes. The major difficulties are related to the presence of high fraction of iodine in gaseous form in the primary circuit during the test, which is not correctly reproduced by the codes. This inability of the codes compromised simulation of the observed iodine behaviour in the containment. In the benchmark a significant user effect was detected (different results being obtained by different users of the same code) which had to be taken into account in analysing the results. This article reports the benchmark results comparing the main parameters calculated and observed, summarising the results achieved, and identifying the areas in which understanding needs to be improved. Relevant experimental and theoretical work is under way to resolve the issues raised. [ABSTRACT FROM AUTHOR]

• The deterministic, response-matrix-based solver REMS for reactor dosimetry applications and its associated methodology are presented. • The ongoing validation of the Serpent 2 Monte Carlo code for reactor dosimetry applications is further extended by simulation of the VENUS-3 benchmark experiment. • The validation of REMS is started by simulation of the VENUS-3 benchmark. VTT is currently working on the development of the deterministic computer code REMS for reactor dosimetry calculations. The transport solution in REMS is based on a response-matrix formalism, with multi-group responses pre-computed by Serpent 2. In this sense, the Serpent-REMS sequence embodies a two-step approach. Reaction rates for three threshold nuclear reactions were calculated by REMS and compared against measured values in the context of the VENUS-3 benchmark experiment. For the 383 experimental values available, the average and standard deviation of calculated-to-experimental reaction rate ratios were (1.03 ± 0.11). This work also reports on the ongoing validation of the Monte Carlo code Serpent 2 for reactor dosimetry applications. The VENUS-3 benchmark problem was also calculated with Serpent 2. The average reaction rate ratios and standard deviations were (1.02 ± 0.06). Execution times for REMS are considerably lower than those of Serpent 2. Furthermore, parallel scalablitity tests suggest promising prospects for the solution of multi-step irradiation scenarios with REMS. Future work will focus on means of improving the accuracy of the REMS solution by refining the Serpent 2 models used in response-matrix generation, as well as on uncertainty qualification. [ABSTRACT FROM AUTHOR]

• The direct transport code SHARK was verified with the plate-type JRR-3M macro benchmark with a series of cases from 2D lattice cases to 3D whole-core cases. • The detailed geometry modeling was conducted to show the high-fidelity calculation ability in SHARK. • Sensitivity analyses were applied in the plate-type calculation with the 2D/1D transport method. The amount azimuthal angles needs to be increased compared to the traditional pin-cell cases. The JRR-3M reactor is a water-cooled plate-type reactor for research. A macro benchmark based on the JRR-3M reactor is established with the Monte-Carlo code RMC and deterministic direct transport code SHARK with accurate and explicit geometry modeling. A series of cases from 2D lattice to 3D whole-core are adopted for the verification of SHARK. In the verification, sensitivity analyses are conducted with the 2D lattice case and applied in the 3D whole-core calculation. Numerical results indicate the good accuracy of SHARK for the plate-type JRR-3M macro benchmark. [ABSTRACT FROM AUTHOR]

The development of multi-physics multi-scale coupled methodologies for Light Water Reactor (LWR) analysis requires comprehensive validation and verification procedures, which include well-established benchmarks developed in international cooperation. The Nuclear Energy Agency (NEA) of the Organization for Economic Co-operation and Development (OECD) has provided such framework, and over the years a number of LWR benchmarks have been developed and successfully conducted. The first set of NEA/OECD benchmarks that permits testing of the neutronics/thermal–hydraulics coupling, and verifying the capability of the coupled codes to analyze complex transients with coupled core/plant interactions have been completed and documented. These benchmarks provided a validation basis for the new generation of coupled “best-estimate” codes. The above mentioned OECD/NEA LWR benchmark activities have also stimulated follow up developments and benchmarks to test these developments. The models utilized have been improved when moving from one benchmark to the next and this created a need to validate them using high-quality experimental data. Second set of the NEA/OECD benchmarks have been initiated by the Expert Group on Uncertainty Analysis in Modelling (EGUAM) at the Nuclear Science Committee (NSC), NEA/OECD to address the current trends in the development of LWR multi-physics and multi-scale modeling and simulation. These benchmarks include the following common features, which address some of the issues identified in the first set of OECD/NEA benchmarks: (a) utilization of high-quality experimental data; (b) refined local scale modeling in addition to global predictions; (c) more detailed comparisons and analysis; (d) including uncertainty and sensitivity analysis of modeling predictions. The paper presents each of these new benchmarks by providing description and discussion of comparative analysis of obtained results. Special attention is devoted to uncertainty propagation in LWR multi-physics and multi-scale simulations for design and safety evaluations. [ABSTRACT FROM AUTHOR]

The Neutronics and Thermal-hydraulics Coupled Simulation program (NTC) is developed by FDS Team, which is a code used for transient analysis of advanced reactors. To investigate the capacity and calculation correctness of NTC for transient simulation, a benchmark on beam interruptions in an 80 MWth LBE-cooled and MOX-fuelled experimental accelerator-driven sub-critical system XADS was carried out by NTC. The benchmark on beam interruptions used in this paper was developed by the OECD/NEA Working Party on Scientific Issues in Partitioning and Transmutation (WPPT). The calculation model had the minimum phenomenological and computational complexity which concerned a simple model (single fuel channel thermal-hydraulics) of the average fuel pin corresponding to the BOL fuel condition. This benchmark was designed to investigate the temperature and power responses caused by beam interruption of different durations, which aimed at comparative assessment of NTC and other computation methods. A comparison of NTC and other ten sets of temperature and power was provided, which showed that the results had good agreement. [ABSTRACT FROM AUTHOR]

• Hybrid finite element SP 3 solutions of three-dimensional VVER benchmarks are presented. • High-fidelity continuous Galerkin finite element SP 3 results applied as references. • The effect of polynomial order applied for axial source fitting is analyzed. • Radially applying higher-order triangular elements is advised. • Discontinuity factors are calculated based on Serpent output parameters. • New SP 3 results are presented for a two-dimensional VVER-1000 benchmark. As a follow-up to the paper previously published by the authors detailing the hybrid finite element (HYBFEM) solution of the SP 3 equations, in this paper, besides a semi-analytical solver verification with an analysis on the applied source fitting polynomial order, three-dimensional VVER-440, and VVER-1000 benchmarks are reproduced with the continuous Galerkin (CGFEM) and HYBFEM SP 3 modules of the SPNDYN code. The effect of the applied axial nodalisation and radial element order on the calculation results is discussed. The HYBFEM solver extended with a discontinuity factor (DCF) calculation module based on Serpent 2 output parameters is verified on a two-dimensional VVER-1000 benchmark. The SPNDYN CGFEM and HYBFEM results show very good agreement. We also verify that by introducing DCFs in SP 3 calculations, the reference transport solution can be reproduced with high accuracy. We highlight the effect of using scalar flux weighted linearly anisotropic scattering matrices in higher-order transport calculations. [ABSTRACT FROM AUTHOR]

The QUENCH-10 (Q-10) and QUENCH-16 (Q-16) experiments were chosen as a SARNET2 code benchmark (SARNET2-COOL-D5.4) exercise to assess the status of modelling air ingress sequences and to compare the capabilities of the various codes used for accident analyses, specifically ATHLET-CD (GRS and RUB), ICARE-CATHARE (IRSN), MAAP (EDF), MELCOR (INRNE and PSI), SOCRAT (IBRAE), and RELAP/SCDAPSim (PSI). Both experiments addressed air ingress into an overheated core following earlier partial oxidation in steam. Q-10 was performed with extensive preoxidation, moderate/high air flow rate and high temperatures at onset of reflood (max T pct = 2200 K), while Q-16 was performed with limited preoxidation, low air flow rate and relative low temperatures at reflood initiation (max T pct = 1870 K). Variables relating to the major signatures (thermal response, hydrogen generation, oxide layer development, oxygen and nitrogen consumption and reflood behaviour) were compared globally and/or at selected locations. In each simulation, the same input models and assumptions are used for both experiments, differing only in respect of the boundary conditions. However, some slight idealisations were made to the assumed boundary conditions in order to avoid ambiguities in the code-to-code comparisons; in this way, it was possible to focus more easily on the key phenomena and hence make the results of the exercise more transparent. Remarks are made concerning the capability of physical modelling within the codes, description of the experiment facility and test conduct as specified in the code input, and code limitations that might warrant additional research to support model improvements, especially the modelling of nitride formation and melt oxidation. [ABSTRACT FROM AUTHOR]

Highlights: [•] MVP was used for SPERT-IV core modeling. [•] Neutronics analysis of SPERT-IV reactor was performed. [•] Calculation performed to estimate critical rod height, excess reactivity. [•] Neutron flux, time integrated neutron flux and Cd-ratio also calculated. [•] Calculated values agree with experimental data. [ABSTRACT FROM AUTHOR]

Abstract: The MCNP distributions include sets of pre-calculated thermal scattering libraries but these libraries are available for several temperature steps only. In order to achieve reliable results it is suitable to process the cross section libraries for the desired temperature. In general, there are three methods to process these thermal scattering libraries for the desired temperatures. This paper deals with the comparison of these three methods on the basis of several benchmarks and on the basis of a thermal transient experiment of a WWER-440 reactor. The choice is up to the MCNP user but unfortunately very few studies concerning the comparison have been published so far. Therefore conclusions and results presented in this paper may help the user to choose the most appropriate method for his calculation. [Copyright &y& Elsevier]

Abstract: Experiments for the determination of the three-dimensional fission density distribution were successfully carried out at the IPEN/MB-01 research reactor facility. The experiment is of very good quality for utilization in the benchmark of computer codes and related nuclear data library commonly used for the calculation of fission density distribution in reactor cores. The complete experimental data set comprises a very massive set of measured fission density distribution. The experiment was evaluated and included in the IRPhE handbook. Two calculation methodologies were employed for the theoretical analysis of the proposed benchmark; one stochastic (MCNP-5) and the other one deterministic (NJOY/AMPX-II/TORT); both utilizing ENDF/B-VII.0 as the source of nuclear data. The theory/experiment comparison reveals in general a good agreement for both methodologies. However, MCNP-5 results are in a better shape than those of TORT. TORT underestimates the relative power distribution mainly in the axial upper part of the fuel rod. This effect is mostly credited to the over-prediction of the control bank worth. [ABSTRACT FROM AUTHOR]

Abstract: Monte Carlo calculation methods to estimate the effective delayed neutron fraction β eff are investigated: one is proposed by Meulekamp et al. and the other is by Nauchi et al. It is revealed that both the methods calculate the delayed neutron fraction weighted with the importance functions defined by Kobayashi. The accuracy of the methods are also examined for several simple benchmark systems. Consequently, it is found that Meulekamp’s method causes ∼5% discrepancies in the β eff values for fast systems; Nauchi’s method gives good results for fast bare systems but ∼10% discrepancies for fast reflected systems. Both the methods calculate the β eff values approximately within the accuracy of ∼2% for thermal systems. [Copyright &y& Elsevier]

Uncertainty in nuclear reaction cross sections, angular distributions, and other nuclear data directly impact how well simulations of nuclear systems represent physical observations. To determine how well the nuclear data in ENDF/B-VIII.0, JEFF-3.3, and JENDL-4.0 evaluations describe the physical behaviour of 235U, and 239Pu when subjected to a neutron flux, the neutron emission spectrum was measured for carbon, 93.0% 235U, and 93.9% 239Pu samples, and compared against detailed MCNP6 simulations. The measurements were performed at the Los Alamos Neutron Science Center using a quasi-differential method previously developed at Rensselaer Polytechnic Institute. The measurement spanned 0.82–20 MeV and 30–150 degrees. The measurements show there are a significant number of discrepancies between library predictions of the neutron yield and physical observation. A few of the main discrepancies found are described in this paper. Based on these results a new evaluation utilizing these results for carbon, 235U, and 239Pu is recommended. [ABSTRACT FROM AUTHOR]

• The Transient Fission Matrix model was used to analyze a fuel assembly benchmark. • The matrices can be used to access fission source modes. • The fission source modes allow the investigation of the system's coupling behavior. • The approach can be connected to the dominance ratio or the eigenvalue separation. • The coupling behavior in the case of reactivity/fuel changes was also studied. Traditionally, stochastic analysis of reactor kinetics behavior had always been a prohibitively expensive problem, in terms of the necessary calculation time. Recent advances in computational methods, however, have made it possible for several widely used Monte Carlo codes to incorporate methods that enable the study and simulation of such topics, with one of them being reactor space–time kinetics or STK, which until recently was limited to deterministic codes only. The present paper uses the Transient Fission Matrix (TFM) model, a hybrid method that uses a system response obtained through Monte Carlo and stored in fission and time matrices as input for deterministic calculations. The result enables a view of the system's behavior in terms of its neutronic coupling and its kinetic parameters, namely the neutron propagation probability, generation time, lifetime and delayed neutron fraction. The TFM method is applied to a coupled core configuration with the purpose of generating a numerical benchmark. The Serpent 2 Monte Carlo code is used for the stochastic part of the calculation. The geometry consists of two fuel assemblies placed in a light water tank, with a water blade of varying width between them. The process consists of adjusting the width (thus the separation) of the two assemblies and then extracting TFM, flux and fission rate results. The blade width ranges between 0 cm and 20 cm. The system behavior, as well as the coupling effects between the two assemblies, are then analyzed and discussed. By obtaining the system's eigenvalues from the transient fission matrices, the degree of coupling can be assessed. As the assemblies move further apart, the system slowly transitions from a single core, to two only loosely coupled ones. This study enables to produce a benchmark to compare future calculations schemes against and predefine an innovative way of designing high dominant ratio configurations. An actual experimental program could be led in ad hoc zero power reactor able to perform them, such as the KUCA reactor of Kyoto University, where similar (in their philosophy) experiments were done in the 90's. [ABSTRACT FROM AUTHOR]

• Massimo Salvatores was the key persons in the promotion of integral experiments. • He initiated many international programs of integral experiments in the world. • He expressed a great interest to the integral experiments in the VENUS-F facility. • VENUS-F experiments to support ADS MYRRHA and LFR core designs are presented. In spite of the remarkable progress of computational tools in the last decades, integral experiments in zero-power facilities are still of a great importance in the world for nuclear data validation, licensing of new core designs and education and training. Massimo Salvatores was one of the key persons in the promotion of integral experiments and their analysis since the early 1980s. He initiated many international programs of integral experiments in different zero-power facilities in the world. He expressed a great interest to the integral experiments in the VENUS-F facility that have been performed since 2011 to support ADS MYRRHA and LFR core designs. Some of these experiments and associated calculations are presented in this paper. [ABSTRACT FROM AUTHOR]

This paper describes steady-state reactor physics measurements and calculations that were performed for the Training Reactor of Budapest University of Technology and Economics (BME TR) with the purpose of benchmarking. Based on the available geometry specifications and material compositions a model of BME TR was created with the well-validated, general-purpose Serpent 2 Monte Carlo code. Uncertain parameters (such as fuel density and control rod positions) were adjusted to related measurements. The Serpent 2 model was used for the generation of group constants, examining several homogenization schemes. Models were created in the PARCS diffusion code, the SPNDYN diffusion and SP3 code and the PARTISN discrete ordinates code. Various Monte Carlo and deterministic calculations were performed with the adjusted models and the results were then compared with actual measured data. The calculations and measurements show good agreement, this way the Serpent model was successfully validated, while the deterministic models make a good basis for more complex benchmarks in the future, such as transients with thermal–hydraulic feedback. [ABSTRACT FROM AUTHOR]

• A neutronic benchmark of the megawatt heat pipe reactor is detailed. • The main neutronic characteristics are investigated using OpenMC. • Cross-sections, feedback effects, power distributions, and burn-up are studied. • All modeling required and all comparison results are accessible. Megawatt-class heat pipe reactors (HPRs) have been identified as a candidate for the decentralized electricity markets, for instance, remote regions, island communities, and military bases. HPRs exhibit some unique designs, such as the encapsulated solid core with fuel pellets and heat pipes, control drums, and irregular reflectors. The neutronic tools are under development to investigate core characteristics and serve multiphysics coupling. This paper proposes a public neutronic benchmark of megawatt HPR of which all modeling parameters are available. This paper uses open-source Monte-Carlo code OpenMC to carry out the preliminary neutronic analysis of this benchmark. The influence of the cross-section library can be >750 pcm. The addition of axial structures zone can lead to >300 pcm reactivity increase. The Doppler effect of fuel exactly matches the oxide-fueled fast reactor Doppler constant formula. The reactivity worth of the control devices, flux distribution, power distribution, and depletion results are presented and accessible for future benchmarks. [ABSTRACT FROM AUTHOR]

Polaris is a 2D multigroup lattice physics capability in the SCALE code system used for the analysis of light water reactor fuel designs. Verification and validation results of Polaris in SCALE 6.2 are presented. The SCALE/KENO continuous-energy Monte Carlo code is used as a reference to compare the results of more than 2,000 combinations of pressurized water reactor and boiling water reactor lattice designs and state points. Radiochemical assay measurements were also used to assess the accuracy of depletion calculations. Results show that Polaris, with ENDF/B-VII.1 cross section library, produces accurate lattice physics data for core simulators. [ABSTRACT FROM AUTHOR]

• We present a multi-physics benchmark for codes targeting molten salt fast reactors. • The benchmark is general and includes steady-state and transient calculations. • We present the results of the benchmarking campaign of four multi-physics tools. • Results show excellent agreement among all partners. • The provided result set is useful to test and develop similar multi-physics codes. Verification and validation of multi-physics codes dedicated to fast-spectrum molten salt reactors (MSR) is a very challenging task. Existing benchmarks are meant for single-physics codes, while experimental data for validation are absent. This is concerning, given the importance numerical simulations have in the development of fast MSR designs. Here, we propose the use of a coupled numerical benchmark specifically designed to assess the physics-coupling capabilities of the aforementioned codes. The benchmark focuses on the specific characteristics of fast MSRs and features a step-by-step approach, where physical phenomena are gradually coupled to easily identify sources of error. We collect and compare the results obtained during the benchmarking campaign of four multi-physics tools developed within the SAMOFAR project. Results show excellent agreement for all the steps of the benchmark. The benchmark generality and the broad spectrum of results provided constitute a useful tool for the testing and development of similar multi-physics codes. [ABSTRACT FROM AUTHOR]