Evaluation of the CNESTEN's TRIGA Mark II research reactor physical parameters with TRIPOLI-4® and MCNP
2
조회
0
다운로드
0
인용
0
참고문헌
키워드
초록
This paper focuses on the development of a new computational model of the CNESTEN’s TRIGA Mark II research reactor using the 3D continuous energy Monte-Carlo code TRIPOLI-4 (T4). This new model was developed to assess neutronic simulations and determine quantities of interest such as kinetic parameters of the reactor, control rods worth, power peaking factors and neutron flux distributions. This model is also a key tool used to accurately design new experiments in the TRIGA reactor, to analyze these experiments and to carry out sensitivity and uncertainty studies. The geometry and materials data, as part of the MCNP reference model, were used to build the T4 model. In this regard, the differences between the two models are mainly due to mathematical approaches of both codes. Indeed, the study presented in this article is divided into two parts: the first part deals with the development and the validation of the T4 model. The results obtained with the T4 model were compared to the existing MCNP reference model and to the experimental results from the Final Safety Analysis Report (FSAR). Different core configurations were investigated via simulations to test the computational model reliability in predicting the physical parameters of the reactor. As a fairly good agreement among the results was deduced, it seems reasonable to assume that the T4 model can accurately reproduce the MCNP calculated values. The second part of this study is devoted to the sensitivity and uncertainty (S/U) studies that were carried out to quantify the nuclear data uncertainty in the multiplication factor keff . For that purpose, the T4 model was used to calculate the sensitivity profiles of the keff to the nuclear data. The integrated-sensitivities were compared to the results obtained from the previous works that were carried out with MCNP and SCALE-6.2 simulation tools and differences of less than 5% were obtained for most of these quantities except for the C-graphite sensitivities. Moreover, the nuclear data uncertainties in the keff were derived using the COMAC-V2.1 covariance matrices library and the calculated sensitivities. The results have shown that the total nuclear data uncertainty in the keff is around 585 pcm using the COMAC-V2.1. This study also demonstrates that the contribution of zirconium isotopes to the nuclear data uncertainty in the keff is not negligible and should be taken into account when performing S/U analysis
Abstract
This paper focuses on the development of a new computational model of the CNESTEN’s TRIGA Mark II research reactor using the 3D continuous energy Monte-Carlo code TRIPOLI-4 (T4). This new model was developed to assess neutronic simulations and determine quantities of interest such as kinetic parameters of the reactor, control rods worth, power peaking factors and neutron flux distributions. This model is also a key tool used to accurately design new experiments in the TRIGA reactor, to analyze these experiments and to carry out sensitivity and uncertainty studies. The geometry and materials data, as part of the MCNP reference model, were used to build the T4 model. In this regard, the differences between the two models are mainly due to mathematical approaches of both codes. Indeed, the study presented in this article is divided into two parts: the first part deals with the development and the validation of the T4 model. The results obtained with the T4 model were compared to the existing MCNP reference model and to the experimental results from the Final Safety Analysis Report (FSAR). Different core configurations were investigated via simulations to test the computational model reliability in predicting the physical parameters of the reactor. As a fairly good agreement among the results was deduced, it seems reasonable to assume that the T4 model can accurately reproduce the MCNP calculated values. The second part of this study is devoted to the sensitivity and uncertainty (S/U) studies that were carried out to quantify the nuclear data uncertainty in the multiplication factor keff . For that purpose, the T4 model was used to calculate the sensitivity profiles of the keff to the nuclear data. The integrated-sensitivities were compared to the results obtained from the previous works that were carried out with MCNP and SCALE-6.2 simulation tools and differences of less than 5% were obtained for most of these quantities except for the C-graphite sensitivities. Moreover, the nuclear data uncertainties in the keff were derived using the COMAC-V2.1 covariance matrices library and the calculated sensitivities. The results have shown that the total nuclear data uncertainty in the keff is around 585 pcm using the COMAC-V2.1. This study also demonstrates that the contribution of zirconium isotopes to the nuclear data uncertainty in the keff is not negligible and should be taken into account when performing S/U analysis
참고문헌 (36)
- Aaron E. Craft, "Characterization of a Neutron Beam Following Reconfi guration of the Neutron Radiography Reactor (NRAD) Core and Addition of New Fuel Elements", Nuclear Engineering and Technology, 48(1), pp.200~210 (2016), 한국원자력학회
- Sarah W. Morgan, "Beam characterization at the Neutron Radiography Reactor", Nuclear Engineering and Design, 265, pp.639~653 (2013), Elsevier BV[DOI]
- Vladimir Radulović, "Silicon carbide neutron detector testing at the JSI TRIGA reactor for enhanced border and port security", Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 972, pp.164122~ (2020), Elsevier BV[DOI]
- A. Gruel, "Gamma-Heating and Gamma Flux Measurements in the JSI TRIGA Reactor: Results and Prospects", IEEE Transactions on Nuclear Science, 67(4), pp.559~567 (2020), Institute of Electrical and Electronics Engineers (IEEE)[DOI]
- B. El Bakkari, "Feasibility analysis of I-131 production in the Moroccan TRIGA research reactor", Annals of Nuclear Energy, 78, pp.140~145 (2015), Elsevier BV[DOI]
- M.Q. Huda, "Studies on the overall safety aspects during irradiation of TeO2 in the central thimble of the TRIGA research reactor", Annals of Nuclear Energy, 36(2), pp.199~212 (2009), Elsevier BV[DOI]
- B. El Bakkari, "Monte Carlo modelling of TRIGA research reactor", Radiation Physics and Chemistry, 79(10), pp.1022~1030 (2010), Elsevier BV[DOI]
- B. El Bakkari, "Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises", Annals of Nuclear Energy, 36(5), pp.626~633 (2009), Elsevier BV[DOI]
- B. El Bakkari, "Fuel burnup analysis for the Moroccan TRIGA research reactor", Annals of Nuclear Energy, 51, pp.112~119 (2013), Elsevier BV[DOI]
- H. Ghninou, "Characterization of Neutron Reaction Rates in Different Irradiation Channels of the CNESTEN’s TRIGA Mark II Research Reactor", IEEE Transactions on Nuclear Science, 69(7), pp.1806~1814 (2022), Institute of Electrical and Electronics Engineers (IEEE)[DOI]
- E. Brun, "TRIPOLI-4®, CEA, EDF and AREVA reference Monte Carlo code", Annals of Nuclear Energy, 82, pp.151~160 (2015), Elsevier BV[DOI]
- P. Archier, "COMAC:nuclear data covariance matrices library for reactor applications", PHYSOR - role React, Phys. Sustain. Future (2014)
- 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(12), 2887~2996[DOI]
- The JEFF-3.1.1, "A. Santamarina", Nuclear Data Library, ~, pp.JEFF Report 22 (2009)
- CALENDF-2010: User Manual, "J.C. Sublet", ~, pp.Rapport CEA-R-6277 (2011)
- Luka Snoj, "Calculation of kinetic parameters for mixed TRIGA cores with Monte Carlo", Annals of Nuclear Energy, 37(2), pp.223~229 (2010), Elsevier BV[DOI]
- Yasushi NAUCHI, "Development of Calculation Technique for Iterated Fission Probability and Reactor Kinetic Parameters Using Continuous-Energy Monte Carlo Method", Journal of Nuclear Science and Technology, 47(11), pp.977~990 (2010), Informa UK Limited[DOI]
- Guillaume Truchet, "Computing adjoint-weighted kinetics parameters in Tripoli-4® by the Iterated Fission Probability method", Annals of Nuclear Energy, 85, pp.17~26 (2015), Elsevier BV[DOI]
- Theory, Interface, Verification, Validation, and Performance of the Adjoint-Weighted Point Reactor Kinetics Parameter Calculations in MCNP, "C. Brian Kiedrowski", Los Alamos Natl. Lab, ~, pp.XCP-3 Monte Carlo Codes -UR-10-01700 (2010)
- Yasushi NAUCHI, "Proposal of Direct Calculation of Kinetic Parameters β<sub>eff<", sub>and Based on Continuous Energy Monte Carlo Method, 2005, pp.42(6), 503~514[DOI]
- Evaluation of Reactor Kinetic Parameters without the Need for Perturbation Codes, "M.M. Bretscher", ~ (1997)
- M.Q. Huda, "Benchmark analysis of the TRIGA MARK II research reactor using Monte Carlo techniques", Annals of Nuclear Energy, 31(11), pp.1299~1313 (2004), Elsevier BV[DOI]
- Tetsuo MATSUMOTO, "Benchmark Analysis of TRIGA Mark II Reactivity Experiment Using a Continuous Energy Monte Carlo Code MCNP", Journal of Nuclear Science and Technology, 37(12), pp.1082~1087 (2000), Informa UK Limited[DOI]
- Haseebur Rehman, "Neutronics analysis of TRIGA Mark II research reactor", Nuclear Engineering and Technology, 50(1), pp.35~42 (2018), 한국원자력학회
- Luka Snoj, "Power peakings in mixed TRIGA cores", Nuclear Engineering and Design, 238(9), pp.2473~2479 (2008), Elsevier BV[DOI]
- H. Ziani, "Eigenvalue sensitivity and nuclear data uncertainty analysis for the Moroccan TRIGA Mark II research reactor using SCALE6.2 and MCNP6.2", Nuclear Engineering and Design, 378, pp.111160~ (2021), Elsevier BV[DOI]
- Christopher M. Perfetti, "SCALE Continuous-Energy Eigenvalue Sensitivity Coefficient Calculations", Nuclear Science and Engineering, 182(3), pp.332~353 (2016), Informa UK Limited[DOI]
- Yishu Qiu, "Computing eigenvalue sensitivity coefficients to nuclear data based on the CLUTCH method with RMC code", Annals of Nuclear Energy, 88, pp.237~251 (2016), Elsevier BV[DOI]
- N. Hfaiedh, "Determination of the optimized SHEM mesh for neutron transport calculations", M C 2005 Int. Top. Meet. Math. Comput. Supercomput. React. Phys. Nucl. Biol. Appl (2005)
- G. Rimpault, "The ERANOS Code and Data System for Fast Reactor Neutronic Analyses" (2002)
- G. Rimplaut, "Algorithmic features of the ECCO Cell Code for treating heterogeneous fast reactor subassemblies", Proc. Int. Conf. Math. Comput. React. Phys. Environ. Anal. (1995)
- C. De Saint Jean, "Evaluation of Cross Section Uncertainties Using Physical Constraints: Focus on Integral Experiments", Nucl. Data Sheets, Special Issue on International Workshop on Nuclear Data Covariances, pp.178~184 (2015)
- A. Santamarina, "Re-estimation of nuclear data and reliable covariances using integral experiments. application to jeff3 library", MandC2017 -International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering. MandC2017 - International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (2017)
- M. Makhloul, "Sensitivity and Uncertainty quantification of neutronic integral data in the TRIGA Mark II Research Reactor", Nuclear Engineering and Technology, 54(2), pp.523~531 (2022), 한국원자력학회[DOI]
- B.T. Rearden, "SCALE Code System", ORNLTM-200539 Version 62, ~ (2016)
- X-5 Monte Carlo Team, "MCNP – A General Monte Carlo N-Particle Transport Code", Version 5, LA-UR-03-1987, ~ (2004)