McCARD/MIG Stochastic sampling calculations for nuclear cross section sensitivity and uncertainty analysis

박호진 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.11

In this study, a cross section stochastic sampling (S.S.) capability is implemented into both the McCARD continuous energy Monte Carlo code and MIG multiple-correlated data sampling code. The ENDF/B-VII.1 covariance data based 30 group cross section sets and the SCALE6 covariance data based 44 group cross section sets are sampled by the MIG code. Through various uncertainty quantification (UQ) benchmark calculations, the McCARD/MIG results are verified to be consistent with the McCARD stand-alone sensitivity/uncertainty (S/U) results and the XSUSA S.S. results. UQ analyses for Three Mile Island Unit 1, Peach Bottom Unit 2, and Kozloduy-6 fuel pin problems are conducted to provide the uncertainties of keff and microscopic and macroscopic cross sections by the McCARD/MIG code system. Moreover, the SNU S/U formulations for uncertainty propagation in a MC depletion analysis are validated through a comparison with the McCARD/MIG S.S. results for the UAM Exercise I-1b burnup benchmark. It is therefore concluded that the SNU formulation based on the S/U method has the capability to accurately estimate the uncertainty propagation in a MC depletion analysis.

UNCERTAINTY PROPAGATION ANALYSIS FOR YONGGWANG NUCLEAR UNIT 4 BY MCCARD/MASTER CORE ANALYSIS SYSTEM

Park, Ho Jin,Lee, Dong Hyuk,Shim, Hyung Jin,Kim, Chang Hyo Korean Nuclear Society 2014 Nuclear Engineering and Technology Vol.46 No.3

This paper concerns estimating uncertainties of the core neutronics design parameters of power reactors by direct sampling method (DSM) calculations based on the two-step McCARD/MASTER design system in which McCARD is used to generate the fuel assembly (FA) homogenized few group constants (FGCs) while MASTER is used to conduct the core neutronics design computation. It presents an extended application of the uncertainty propagation analysis method originally designed for uncertainty quantification of the FA FGCs as a way to produce the covariances between the FGCs of any pair of FAs comprising the core, or the covariance matrix of the FA FGCs required for random sampling of the FA FGCs input sets into direct sampling core calculations by MASTER. For illustrative purposes, the uncertainties of core design parameters such as the effective multiplication factor ($k_{eff}$), normalized FA power densities, power peaking factors, etc. for the beginning of life (BOL) core of Yonggwang nuclear unit 4 (YGN4) at the hot zero power and all rods out are estimated by the McCARD/MASTER-based DSM computations. The results are compared with those from the uncertainty propagation analysis method based on the McCARD-predicted sensitivity coefficients of nuclear design parameters and the cross section covariance data.

Uncertainty Propagation Analysis for Yonggwang Nuclear Unit 4 by McCARD/Master Core Analysis System

박호진,이동혁,심형진,김창효 한국원자력학회 2014 Nuclear Engineering and Technology Vol.46 No.3

This paper concerns estimating uncertainties of the core neutronics design parameters of power reactors by direct samplingmethod (DSM) calculations based on the two-step McCARD/MASTER design system in which McCARD is used to generatethe fuel assembly (FA) homogenized few group constants (FGCs) while MASTER is used to conduct the core neutronicsdesign computation. It presents an extended application of the uncertainty propagation analysis method originally designed foruncertainty quantification of the FA FGCs as a way to produce the covariances between the FGCs of any pair of FAs comprisingthe core, or the covariance matrix of the FA FGCs required for random sampling of the FA FGCs input sets into direct samplingcore calculations by MASTER. For illustrative purposes, the uncertainties of core design parameters such as the effectivemultiplication factor (keff), normalized FA power densities, power peaking factors, etc. for the beginning of life (BOL) core ofYonggwang nuclear unit 4 (YGN4) at the hot zero power and all rods out are estimated by the McCARD/MASTER-basedDSM computations. The results are compared with those from the uncertainty propagation analysis method based on theMcCARD-predicted sensitivity coefficients of nuclear design parameters and the cross section covariance data.

Advances for the time-dependent Monte Carlo neutron transport analysis in McCARD

Sang Hoon Jang,Hyung Jin Shim Korean Nuclear Society 2023 Nuclear Engineering and Technology Vol.55 No.7

For an accurate and efficient time-dependent Monte Carlo (TDMC) neutron transport analysis, several advanced methods are newly developed and implemented in the Seoul National University Monte Carlo code, McCARD. For an efficient control of the neutron population, a dynamic weight window method is devised to adjust the weight bounds of the implicit capture in the time bin-by-bin TDMC simulations. A moving geometry module is developed to model a continuous insertion or withdrawal of a control rod. Especially, the history-based batch method for the TDMC calculations is developed to predict the unbiased variance of a bin-wise mean estimate. The developed methods are verified for three-dimensional problems in the C5G7-TD benchmark, showing good agreements with results from a deterministic neutron transport analysis code, nTRACER, within the statistical uncertainty bounds. In addition, the TDMC analysis capability implemented in McCARD is demonstrated to search the optimum detector positions for the pulsed-neutron-source experiments in the Kyoto University Critical Assembly and AGN201K.

MCCARD: MONTE CARLO CODE FOR ADVANCED REACTOR DESIGN AND ANALYSIS

심형진,한범석,JONG SUNG JUNG,박호진,김창효 한국원자력학회 2012 Nuclear Engineering and Technology Vol.44 No.2

McCARD is a Monte Carlo (MC) neutron-photon transport simulation code. It has been developed exclusively for the neutronics design of nuclear reactors and fuel systems. It is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis. It has some special features such as the anterior convergence diagnostics, real variance estimation, neutronics analysis with temperature feedback, B1 theory-augmented few group constants generation, kinetics parameter generation and MC S/U analysis based on the use of adjoint flux. This paper describes the theoretical basis of these features and validation calculations for both neutronics benchmark problems and commercial PWR reactors in operation.

MCCARD: MONTE CARLO CODE FOR ADVANCED REACTOR DESIGN AND ANALYSIS

Shim, Hyung-Jin,Han, Beom-Seok,Jung, Jong-Sung,Park, Ho-Jin,Kim, Chang-Hyo Korean Nuclear Society 2012 Nuclear Engineering and Technology Vol.44 No.2

McCARD is a Monte Carlo (MC) neutron-photon transport simulation code. It has been developed exclusively for the neutronics design of nuclear reactors and fuel systems. It is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis. It has some special features such as the anterior convergence diagnostics, real variance estimation, neutronics analysis with temperature feedback, $B_1$ theory-augmented few group constants generation, kinetics parameter generation and MC S/U analysis based on the use of adjoint flux. This paper describes the theoretical basis of these features and validation calculations for both neutronics benchmark problems and commercial PWR reactors in operation.

DeCART2D/MASTER Core Follow Calculation for Hanbit Unit 3 and Comparison With McCARD Single Fuel Assembly Burnup Analyses

Jeong Woo Park,Seung-Ah Yang,Ho Jin Park 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.2

In the design of a spent-fuel (SF) storage, the consideration of burnup credit brings the benefits in safety and economic views. According to it, various SF burnup measurement systems have been developed to estimate high fidelity burnup credit, such as FORK and SMOPY. Recently, there are a few attempts to localize the SF burnup measurement system in South Korea. For the localization of SF burnup measurement systems, it is very important to build the isotope inventory data base (DB) of various kinds of SFs. In this study, we performed DeCART2D/MASTER core follow calculations and McCARD single fuel assembly (FA) burnup analyses for Hanbit unit 3 and confirmed the characteristic of the isotope inventory over burnup. Firstly, the core follow calculations for Cycles 1~7 were performed using DeCART2D/MASTER code system. The core follow calculation is very realistic and practical because it considers the design conditions from its nuclear design report (NDR). Secondly, the Monte Carlo burnup analyses for single FAs were conducted by the McCARD Monte Carlo (MC) transport code. The McCARD code can utilize continuous energy cross section library and treat complex geometric information for particle transport simulation. Accordingly, the McCARD code can provide accurate solutions for burnup analyses without approximations, but it needs huge computing resources and time burden to perform whole-core follow calculations. Therefore, we will confirm the effectiveness of the single McCARD FA burnup analyses by comparing the DeCART2D/MASTER core follow results with the McCARD solution. From the results, the use of single FA burnup analyses for the establishment of the DBs will be justified. Various FAs, that have different 235U enrichments and loading pattern of fuel rods and burnable absorbers, were considered for the burnup analyses. In addition, the results of the sensitivity analyses for power density, initial enrichment, and cooling time will be presented.

Criticality benchmark of McCARD Monte Carlo code for light-water-reactor fuel in transportation and storage packages

장준경,이호철,이현철 한국원자력학회 2018 Nuclear Engineering and Technology Vol.50 No.7

In this paper, McCARD code was verified using various models listed in the NUREG/CR-6361 benchmarkguide, which provides specifications for single pin-cells, single assemblies, and the whole core classifieddepending on the nuclear properties and structural characteristics. McCARD code was verified bycomparing its results with those of SCALE code for single pin-cell and single assembly benchmarkproblems. The difference in the multiplication factor obtained through the two codes did not exceed 90pcm. The benchmark guide treats a total of 173 whole core experiments. The experiments are categorizedas simple lattices, separator plates, reflecting walls, reflecting walls and separator plates, burnableabsorber fuel rods, water holes, poison rods, and borated moderator. As a result of numerical simulationusing McCARD, the mean value of the multiplication factors is 1.00223 and the standard deviation of themultiplication factors is 285 pcm. The difference between the multiplication factors and the experimentalvalue is in the range of -665 pcm to þ 1609 pcm. In addition, statistics of results for experimentscategorized by reactor shape, additional structure, burnable poison, etc., are detailed in the main text.

Monte Carlo burnup and its uncertainty propagation analyses for VERA depletion benchmarks by McCARD

박호진,이동혁,전병규,심형진 한국원자력학회 2018 Nuclear Engineering and Technology Vol.50 No.7

For an efficient Monte Carlo (MC) burnup analysis, an accurate high-order depletion scheme to considerthe nonlinear flux variation in a coarse burnup-step interval is crucial accompanied with an accuratedepletion equation solver. In a Seoul National University MC code, McCARD, the high-order depletionschemes of the quadratic depletion method (QDM) and the linear extrapolation/quadratic interpolation(LEQI) method and a depletion equation solver by the Chebyshev rational approximation method (CRAM)have been newly implemented in addition to the existing constant extrapolation/backward extrapolation(CEBE) method using the matrix exponential method (MEM) solver with substeps. In this paper, thequadratic extrapolation/quadratic interpolation (QEQI) method is proposed as a new high-order depletionscheme. In order to examine the effectiveness of the newly-implemented depletion modules inMcCARD, four problems in the VERA depletion benchmarks are solved by CEBE/MEM, CEBE/CRAM, LEQI/MEM, QEQI/MEM, and QDM for gadolinium isotopes. From the comparisons, it is shown that the QEQI/MEM predicts kinf's most accurately among the test cases. In addition, statistical uncertainty propagationanalyses for a VERA pin cell problem are conducted by the sensitivity and uncertainty and the stochasticsampling methods.

Dynamic Monte Carlo transient analysis for the Organization for Economic Co-operation and Development Nuclear Energy Agency (OECD/NEA) C5G7-TD benchmark

Nadeem Shaukat,류민,심형진 한국원자력학회 2017 Nuclear Engineering and Technology Vol.49 No.5

With ever-advancing computer technology, the Monte Carlo (MC) neutron transport calculation isexpanding its application area to nuclear reactor transient analysis. Dynamic MC (DMC) neutron trackingfor transient analysis requires efficient algorithms for delayed neutron generation, neutron populationcontrol, and initial condition modeling. In this paper, a new MC steady-state simulation method based ontime-dependent MC neutron tracking is proposed for steady-state initial condition modeling; during thisprocess, prompt neutron sources and delayed neutron precursors for the DMC transient simulation caneasily be sampled. The DMC method, including the proposed time-dependent DMC steady-state simulationmethod, has been implemented in McCARD and applied for two-dimensional core kineticsproblems in the time-dependent neutron transport benchmark C5G7-TD. The McCARD DMC calculationresults show good agreement with results of a deterministic transport analysis code, nTRACER.