Development of uncertainty quantification module for VVER analysis in STREAM/RAST-V two-step method
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This paper introduces the creation of a module for Uncertainty Quantification (UQ) specifically designed for VVER analysis through the implementation of the STREAM/RAST-V two-step approach. The aim was to expand the range of use by developing a UQ module tailored for analyzing VVER. This research presents two innovative computational functionalities: (1) development of a library for the pin-based pointwise energy slowing down method (PSM), and (2) extension of the analysis area to study hexagonal-geometry fuel assemblies. The proposed UQ scheme was evaluated through verification using UAM benchmark, and comparative analysis between codes using SCALE 6.2.2 for. STREAM provides an accuracy comparable to that of SCALE 6.2.2. Additionally, a PSM covariance library was utilized in the calculations, achieving 0.7941% and 0.7907% accuracies in the hot full power and hot zero power calculations, respectively. To assess the UQ sequences in the two-step method, the STREAM/RAST-V calculation scheme was verified using the STREAM lattice code. To conclude, this study furnishes comprehensive insights into the development of the UQ module within the two-step method for VVER analysis, and it validates its performance through utilization of the UAM benchmark
Abstract
This paper introduces the creation of a module for Uncertainty Quantification (UQ) specifically designed for VVER analysis through the implementation of the STREAM/RAST-V two-step approach. The aim was to expand the range of use by developing a UQ module tailored for analyzing VVER. This research presents two innovative computational functionalities: (1) development of a library for the pin-based pointwise energy slowing down method (PSM), and (2) extension of the analysis area to study hexagonal-geometry fuel assemblies. The proposed UQ scheme was evaluated through verification using UAM benchmark, and comparative analysis between codes using SCALE 6.2.2 for. STREAM provides an accuracy comparable to that of SCALE 6.2.2. Additionally, a PSM covariance library was utilized in the calculations, achieving 0.7941% and 0.7907% accuracies in the hot full power and hot zero power calculations, respectively. To assess the UQ sequences in the two-step method, the STREAM/RAST-V calculation scheme was verified using the STREAM lattice code. To conclude, this study furnishes comprehensive insights into the development of the UQ module within the two-step method for VVER analysis, and it validates its performance through utilization of the UAM benchmark
참고문헌 (42)
- J. Jang, "Uncertainties of PWR spent nuclear fuel isotope inventory for back-end cycle analysis with Stream", RAST-K, 2021, pp.158, ~
- J. Jang, "Uncertainty quantification in decay heat calculation of spent nuclear fuel by Stream", RAST-K, 2021, pp.53, 2803~2815
- S. Choi, "Resonance treatment using pin-based pointwise energy slowing-down method", J. Comp. Physiol, 330, pp.134~155 (2017)[DOI]
- B. T. Rearden, "SCALE Code System", Oak Ridge National Laboratory, ~ (2017)
- M. B. Chadwick, "ENDF", B-VII.1 nuclear data for science and technology: cross sections, covariances, fission product yields and decay data, 2011, pp.112, 2887~2996
- "CASL: The consortium for advanced simulation of light water reactors A DOE energy innovation hub"[URL]
- R. N. Bratton, "OECD", NEA benchmark for uncertainty analysis in modeling(UAM)for lwrs – summary and discussion of neutronics cases(phase I), 2014, pp.46, 313~342
- J. Jang, "Analysis of Rostov-II benchmark using conventional two-step code systems", Energies, 15, pp.3318~ (2022)[DOI]
- J. Jang, "Development of nodal diffusion code RAST-V for Vodo–Vodyanoi Energetichesky reactor analysis", Nucl. Eng. Technol, 54, pp.3494~3515 (2022)[DOI]
- S. Choi, "Development of high-fidelity neutron transport code STREAM", Comput. Phys. Commun, 264 (2021)[DOI]
- J. Park, "RAST-K v2—three-dimensional nodal diffusion code for pressurized water reactor core analysis", Energies, 13, pp.6324~ (2020)[DOI]
- R. J. J. Stamm’ler, "Methods of Steady-State Reactor Physics in Nuclear Design" (1983), Academic Press Inc
- D. Knott, "Handbook of Nuclear Engineering", pp.913~1239 (2010), Springer
- J. Jang, "Development of Nodal Diffusion Code for VVER and HTGR Analysis with Advanced Semi Analytic Nodal Method", doctor, pp.2023
- The NJOY Nuclear Data Processing System, Version 2016, "D. W. Muir", Los Alamos National Laboratory, ~, pp.LA-UR-17-20093 (2016)
- J. Jang, "Verification and validation of isotope inventory prediction for back-end cycle management using two-step method", Nucl. Eng. Technol, 53, pp.2104~2125 (2021)[DOI]
- J. Jang, "Validation of spent nuclear fuel decay heat calculation by a two-step method", Nucl. Eng. Technol, 53, pp.44~60 (2021)[DOI]
- J. Choe, "Verification and validation of Stream", RAST-K for PWR analysis, 2019, pp.51, 356~368
- R. J. J. Stamm’ler, "Methods of Steady-State Reactor Physics in Nuclear Design" (1983), Academic Press Inc
- D. Knott, "Handbook of Nuclear Engineering", pp.913~1239 (2010), Springer
- A. Yamamoto, "Uncertainty quantification of LWR core characteristics using random sampling method", Nucl. Sci. Eng, 181, pp.160~174 (2015)[DOI]
- Chiba: A Code to Process Neutron-Nuclide Reaction Cross Section Covariance version 2.3., "G. Errorj", Japan Atomic Energy Agency, ~ (2007)
- "Preparation of a covariance processing system for the evaluated nuclear data file, JENDL, (III),, JNC TJ 9440 99–003, vol.3" (1999)[URL]
- AMPX-6:a modular code system for processing ENDF, "B", D. Wiarda, 2016, ORNL[DOI]
- G. Ilas, "Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties", Nucl. Eng. Des, 319, pp.176~184 (2017)[DOI]
- V. C. Klema, "The singular value decomposition: its computation and some applications", IEEE Trans. Automat. Control, 25 (1980)
- T. Zhu, "Sampling-Based Nuclear Data Uncertainty Quantification for Continuous Energy Monte Carlo Codes", pp.2015
- N. J. Quartemont, "Nuclear data covariance analysis in radiation-trransport simulations utilizing SCALE sampler and the IRDFF nuclear data library", IEE, transactions on nuclear, Science, 67, pp.482~491 (2020)
- K. A. Kinoshita, "Uncertainty Quantification of Nutronics Characteristics Using Latin HyperCube Sampling Method", PHYSOR (2014)
- "Covariance plots for ENDF/B-VII.1"[URL]
- J. Jang, "Development of Nodal Diffusion Code for VVER and HTGR Analysis with Advanced Semi Analytic Nodal Method", doctor, pp.2023
- Benchmark for Uncertainty Analysis in Modelling (Uam) for Design, Operation and Safety Analysis of LWRs, I: Specification and Support Data for the Neutronics Cases (Phase I), "K. M. Ivanov", OECD Nuclear Energy Agency, ~ (2016)
- Y. Bilodid, "X2 VVER-1000 benchmark revision: fresh HZP core state and the reference Monte Carlo solution", Ann. Nucl. Energy, 144 (2020)[DOI]
- J. Lepp¨anen, "The Serpent Monte Carlo code: status, development and applications in 2013", Ann. Nucl. Energy, 82, pp.142~150 (2015)[DOI]
- M. B. Chadwick, "ENDF", B-VII.0: next generation evaluated nuclear data library for nuclear science and technology, nuclear data sheets, next generation evaluated nuclear data library for nuclear science and technology, 2006, pp.107, 2931~3060
- E. Canuti, "Sensitivity studies for the exercise I-1 of the OECD", UAM benchmark, 2012, pp.2012, 10~
- H. J. Park, "McCARD", MIG stochastic sampling calculations for nuclear cross section sensitivity and uncertainty analysis, 2022, pp.54, 4272~4279
- A. Labarile, "Comparison of SERPENT and SCALE methodology for LWRs transport calculations and additionally uncertainty analysis for cross section perturbation with SAMPLER module", EPJ Nuclear, Sci. Technol, 2 (2016)[DOI]
- L. Mercatali, "SCALE modeling of selected neutronics test problems within the OECD UAM LWR’s benchmark", Sci. Technol. Nucl. Installations, 2013, pp.1~11 (2013)[DOI]
- C. Arenas, "Uncertainty analysis of light water reactor fuel lattices", Sci. Technol. Nucl. Installations, 2013, pp.1~10 (2013)[DOI]
- B. T. Rearden, "Sensitivity and Uncertainty Analysis Capabilities and Data in SCALE", 174 (2011)[DOI]
- SCALE, A modular code system for performing standardized computer analyses licensing evaluation, Rev.4 (ORNL, "NUREG", CSD-2R4), Vols. and 111 (April 1995). RSIC code package CCC-545, version 4.2, pp.1995, ~[DOI]
