Analysis of several VERA benchmark problems with the photon transport capability of STREAM

STREAM - a lattice transport calculation code with method of characteristics for the purpose of light water reactor analysis - has been developed by the Computational Reactor Physics and Experiment laboratory (CORE) of the Ulsan National Institute of Science and Technology (UNIST). Recently, efforts have been taken to develop a photon module in STREAM to assess photon heating and the influence of gamma photon transport on power distributions, as only neutron transport was considered in previous STREAM versions. A multi-group photon library is produced for STREAM based on the ENDF/B-VII.1 library with the use of the library-processing code NJOY. The developed photon solver for the computation of 2D and 3D distributions of photon flux and energy deposition is based on the method of characteristics like the neutron solver. The photon library and photon module produced and implemented for STREAM are verified on VERA pin and assembly problems by comparison with the Monte Carlo code MCS - also developed at UNIST. A short analysis of the impact of photon transport during depletion and thermal hydraulics feedback is presented for a 2D core also from the VERA benchmark. (C) 2022 Korean Nuclear Society, Published by Elsevier Korea LLC

Establishment and Application of Diamond Detector Analysis System

Department of Nuclear EngineeringIn this study, a diamond detector in a mixed neutron-photon field of the CROCUS research reactor at ??cole Polytechnique F??d??rale de Lausanne (EPFL) is modeled. Simulations are carried out to analyze pulses from the diamond detector in more detail, which induce a novel discovery. Through a code-to-code comparison, the Monte Carlo codes SERPENT v2.1.29 and GEANT4 v10.04.p02 are selected for the CROCUS whole core calculation and the detailed physics modeling in the diamond crystal, respectively. The neutron and prompt gamma ray contributions to the detector are modeled by a two-step procedure (SERPENT2/GEANT4), and the simulation of the delayed gamma ray contribution is carried out by a three-step procedure (SERPENT2/STREAM-SNF/GEANT4). The simulations show that the fraction of the gamma-to-neutron fluxes in the diamond detector is approximately 91.4%, and that of the delayed-to-prompt gamma fluxes is approximately 47.2%. By using the flux spectra calculated at the location of the detector, the physics of particle interactions with the diamond crystal is investigated. The contributions of the neutrons and gamma rays to the diamond detector signal amount to approximately 27% and 73%, respectively. The energies and positions of the particles contributing to the detector signal as tallied in GEANT4 are employed to reconstruct numerical pulses and create a scatter plot. In the scatter plot, pulses are arranged according to the energy for each calculation width, which is defined as the width at 0% of the maximum amplitude. The proton recoil plot shows two bands, one due to protons impacting the anode and the other by protons impacting the cathode, thus showing that protons do not have sufficient energy to penetrate the diamond crystal and have the same probability of interacting with the anode and cathode. This tendency also appears as a high-energy tail in a pulse energy spectrum consisting of the number of pulses according to the energy distribution. Meanwhile, neutron scattering collisions have a homogeneous distribution in the crystal. Hence, a structure with a higher count at the ballistic center region (BCR) is observed and is probably related to the amplitude of the BCR pulses being higher. Thus, it is possible to observe better pulses resulting from the energy depositions at the BCR. Finally, the modeling performance is assessed by comparing the calculated results with the experimental data. In the pulse energy spectrum, a curve produced by the simulations matches with that produced by the measurements. The slope of the curves between 1 MeV and 2 MeV is mainly produced by gamma interactions. The high-energy tail is produced by neutron interactions, especially, the proton recoil. The lithium converter reactions in the diamond detector account for 14.31% and 15.13% beyond 1.34 MeV for the measurement and simulation, respectively, showing consistency.clos

Verification and validation of monte carlo code MCS for the multi-physics high-fidelity analysis of OPR-1000 multi-cycle operation

Department of Nuclear EngineeringWorldwide demand for high-fidelity simulation tools for large scale power reactor analysis resulted in the coupling of neutronics, thermal hydraulics, and fuel mechanics in nuclear reactor cores. The Computational Reactor Physics and Experiment (CORE) laboratory at Ulsan National Institute of Science and Technology (UNIST) has developed an inhouse Monte Carlo code MCS coupled with multi-physics (MP) tools such as a one-dimensional (1D) single-phase closed-channel thermal-hydraulic code (TH1D), a sub-channel two-phase thermal-hydraulic code (CTF) and a fuel performance code (FRAPCON) to provide thermal-hydraulic and fuel behavior feedbacks for realistic applications. This thesis describes the 3D whole-core analysis with pin-wise resolution in which neutron transport, depletion, thermal-hydraulic, and fuel behavior calculations are performed using the coupled MCS/TH1D, MCS/CTF, and MCS/FRAPCON tools. The OPR-1000 PWR core operated for 2 consecutive cycles is selected as a target for MP coupling analysis, including verification and validation. The OPR-1000 PWR is a Generation II nuclear reactor in South Korea with 2815 MW thermal power modeled explicitly in MCS. The MCS simulation of OPR core during the zero-power physics testing was evaluated for critical boron concentration (CBC) and control rod worth, and the results show good agreement with the references. The verification and validation of MCS MP coupling were conducted at hot full power conditions along with various feedback required for reactor power simulation such as depletion, equilibrium xenon update, CBC search, and on-the-fly cross-section reconstruction. The influence on other parameters, including CBC, axial shape index, pin- and assembly-wise radial/axial power profiles, fuel temperatures, and moderator temperatures/densities, were investigated. The MCS MP coupled results were also compared against the experimental data for validation. Additional comparisons were made with the data from the plant???s nuclear design report and result from the deterministic two-step code STREAM/RASTK 2.0 (ST/R2) and 3D Method of Characteristic (MOC) direct neutron transport code STREAM. For the MCS MP coupling tool with thermal-hydraulic and fuel performance feedback, excellent agreement is observed with the measured values with a root mean square (RMS) error of 26 ppm for CBC and 1.8% for assembly power. Compared to other codes, MCS MP coupling results have an RMS error of 16 ppm for CBC and 1.8% for assembly power. Larger discrepancies in relative assembly power between MCS based MP tool and measured data occur in the core-periphery where the power is relatively low, while at the end of cycles, the discrepancies are still within 1 standard deviation of about 2.4%. This study demonstrates MCS???s capability to perform high-fidelity simulation of a practical light water reactor core.clos

Development of a New Monte Carlo Code for High-Fidelity Power Reactor Analysis

Department of Nuclear EngineeringThe high-fidelity multiphysics simulation using transport codes are being mainstream in the reactor physics society. Monte Carlo neutron transport code is one of the most promising candidates of neutronics code of multiphysics simulation since it has advantages of using continuous energy cross section and explicit geometry modeling. Most of the methods required for the Monte Carlo multiphysics simulation has been developed and studied well individually. However, Monte Carlo method have not been able to be applied for large-scale multiphysics simulation such as Pressurized Water Reactor analysis because of the limited computing power, memory storage and especially lack of Monte Carlo codes adapted for large-scale power reactor simulation. Development of Monte Carlo multiphysics code is a challenging due to two aspect: implementing various state of the art techniques into one single code system and making it feasible running simulations on practical computing machines. A new Monte Carlo multiphysics code named MCS was developed for large-scale power reactor analysis. Various state-of-the art techniques were implemented to make it practical tool for multiphysics simulation including thermal hydraulics, depletion, equilibrium xenon, eigenvalue search, on-the-fly cross section generation, hash-indexing, parallel fission bank. The high performance of MCS was achieved and demonstrated. The test result confirmed that the overhead of massive number of tallies is only a one percent up to 13M tally bins, and the parallel efficiency was maintained above 90% up to 1,120 processors when solving power reactor simulation. The fundamental study of power reactor analysis was performed to decide calculation condition including burnup sensitivity, mesh sensitivity, history sensitivity against pressurized water reactor benchmark problem BEAVRS Cycle 1. Finally, capability of power reactor analysis was demonstrated against BEAVRS Cycle 1 and 2.ope

Overcoming the challenges of Monte Carlo depletion: Application to a material-testing reactor with the MCS code

The theoretical aspects behind the reactor depletion capability of the Monte Carlo code MCS developed at the Ulsan National Institute of Science and Technology (UNIST) and practical results of this depletion feature for a Material-Testing Reactor (MTR) with plate-type fuel are described in this paper. A verification of MCS results is first performed against MCNP6 to confirm the suitability of MCS for the criticality and depletion analysis of the MTR. Then, the dependence of the effective neutron multiplication factor to the number of axial and radial depletion cells adopted in the fuel plates is performed with MCS in order to determine the minimum spatial segmentation of the fuel plates. Monte Carlo depletion results with 37,800 depletion cells are provided by MCS within acceptable calculation time and memory usage. The results show that at least 7 axial meshes per fuel plate are required to reach the same precision as the reference calculation whereas no significant differences are observed when modeling 1 or 10 radial meshes per fuel plate. This study demonstrates that MCS can address the need for Monte Carlo codes capable of providing reference solutions to complex reactor depletion problems with refined meshes for fuel management and research reactor applications. (c) 2020 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/)

Interpretation of two SINBAD photon-leakage benchmarks with nuclear library ENDF/B-VIII.0 and Monte Carlo code MCS

A review of the documentation and an interpretation of the NEA-1517/74 and NEA-1517/80 shielding benchmarks (measurements of photon leakage flux from a hollow sphere with a central 14 MeV neutron source) from the SINBAD database with the Monte Carlo code MCS and the most up-to-date ENDF/B-VIII.0 neutron data library are conducted. The two analyzed benchmarks describe satisfactorily the energy resolution of the photon detector and the geometry of the spherical samples with inner beam tube, tritium target and cooling water circuit, but lack information regarding the detector geometry and the distances of shields and collimators relatively to the neutron source and the detector. Calculations are therefore conducted for a sphere model only. A preliminary verification of MCS neutron-photon calculations against MCNP6.2 is first conducted, then the impact of modelling the inner beam tube, tritium target and cooling water circuit is assessed. Finally, a comparison of calculated results with the libraries ENDF/B-VII.1 and ENDF/B-VIII.0 against the measurements is conducted and shows reasonable agreement. The MCS and MCNP inputs used for the interpretation are available as supplementary material of this article. (C) 2019 Korean Nuclear Society, Published by Elsevier Korea LLC

Validation of nuclide depletion capabilities in Monte Carlo code MCS

In this work, the depletion capability implemented in Monte Carlo code MCS is investigated to predict the isotopic compositions of spent nuclear fuel (SNF). By comparison of MCS calculation results to post irradiation examination (PIE) data obtained from one pressurized water reactor (PWR), the validation of this capability is conducted. The depletion analysis is performed with the ENDF/B-VII.1 library and a fuel assembly model. The transmutation equation is solved by the Chebyshev Rational Approximation Method (CRAM) with a depletion chain of 3820 isotopes. 18 actinides and 19 fission products are analyzed in 14 SNF samples. The effect of statistical uncertainties on the calculated number densities is discussed. On average, most of the actinides and fission products analyzed are predicted within +/- 6% of the experiment. MCS depletion results are also compared to other depletion codes based on publicly reported information in literature. The code-to-code analysis shows comparable accuracy. Overall, it is demonstrated that the depletion capability in MCS can be reliably applied in the prediction of SNF isotopic inventory. (c) 2020 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/)

Propagation of radiation source uncertainties in spent fuel cask shielding calculations

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose rate coming from two sources: nuclear data and modeling parameters, are quantified. The nuclear data uncertainties are obtained from the stochastic sampling of the cross-section covariance and perturbed fission product yields. Uncertainties induced by perturbed modeling parameters consider the design parameters and operating conditions. Uncertainties coming from the two sources result in perturbed depleted nuclide inventories and radiation source terms which are then propagated to the dose rate on the cask surface. The uncertainty analysis results show that the neutron and secondary photon dose have uncertainties which are dominated by the cross section and modeling parameters, while the fission yields have relatively insignificant effect. Besides, the primary photon dose is mostly influenced by the fission yield and modeling parameters, while the cross-section data have a relatively negligible effect. Moreover, the neutron, secondary photon, and primary photon dose can have uncertainties up to about 13%, 14%, and 6%, respectively

Analysis of VVER-1000 mock-up criticality experiments with nuclear data library ENDF/B-VIII.0 and Monte Carlo code MCS

The criticality analysis of VVER-1000 mock-up benchmark experiments from the LR-0 research reactor operated by the Research Center Rez in the Czech Republic has been conducted with the MCS Monte Carlo code developed at the Computational Reactor Physics and Experiment laboratory of the Ulsan National Institute of Science and Technology. The main purpose of this work is to evaluate the newest ENDF/B-VIII.0 nuclear data library against the VVER-1000 mock-up integral experiments and to validate the criticality analysis capability of MCS for light water reactors with hexagonal fuel lattices. A preliminary code/code comparison between MCS and MCNP6 is first conducted to verify the suitability of MCS for the benchmark interpretation, then the validation against experimental data is performed with both ENDF/B-VII.1 and ENDF/B-VIII.0 libraries. The investigated experimental data comprises six experimental critical configurations and four experimental pin-by-pin power maps. The MCS and MCNP6 inputs used for the criticality analysis of the VVER-1000 mock-up are available as supplementary material of this article

Verification of photon transport capability of UNIST Monte Carlo code MCS

A photon transport capability has been implemented and verified in the Monte Carlo code MCS of Ulsan National Institute of Science and Technology for the purpose of radiation shielding studies. The MCS photon fixed-source mode simulates the transport of photons between 1 keV and 20 MeV for all elements from hydrogen to fermium. The specific physics for the main four photo-atomic reactions (Rayleigh scattering, Compton scattering, photoelectric effect and pair production) and three secondary processes of photon production (positron-electron annihilation, atomic relaxation and electron/positron bremsstrahlung) are reviewed. Verification results against Monte Carlo codes MCNP6.1 and SERPENT2.1.29 are presented. The verification cases include the comparison of energy distributions of photon flux in an infinite medium, of spatial distributions of photon flux in a cylinder, of the spatial distribution of photon body -equivalent dose in a spent nuclear fuel transport cask, and of photon KERMA (Kinetic Energy Released per MAss) in photon detector calibration geometries. Good calculation/calculation agreement is observed overall, with some marked differences in the detailed photon flux comparison at given energy ranges traced back to differences in photon physics implementation. MCS can from now on be applied for the purpose of advanced photon studies and corresponding validation against experimental shielding benchmarks will follow in the future