Development of a PARCS/Serpent model for neutronics analysis of the dalat nuclear research reactor

Viet-Phu Tran; Kien-Cuong Nguyen; Donny Hartanto; Hoai-Nam Tran; Vinh Thanh Tran; Van-Khanh Hoang; Pham Nhu Viet Ha

doi:10.1007/s41365-021-00855-5

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Development of a PARCS/Serpent model for neutronics analysis of the dalat nuclear research reactor

NUCLEAR ENERGY SCIENCE AND ENGINEERING | Updated：2021-03-12

- Development of a PARCS/Serpent model for neutronics analysis of the dalat nuclear research reactor
- Nuclear Science and Techniques Vol. 32, Issue 2, Article number: 15(2021)
- Affiliations：
  
  1.Institute for Nuclear Science and Technology (INST), Vietnam Atomic Energy Institute (VINATOM), 179 Hoang Quoc Viet str., Cau Giay dist., Hanoi 100000, Vietnam
  2.Nuclear Research Institute (NRI), Vietnam Atomic Energy Institute (VINATOM), 01 Nguyen Tu Luc, Dalat, Lamdong 670000, Vietnam
  3.Department of Mechanical and Nuclear Engineering, University of Sharjah, PO. BOX 27272, Sharjah, United Arab Emirates
  4.Institute of Fundamental and Applied Sciences, Duy Tan University, Ho Chi Minh city 700000, Vietnam
  5.Faculty of Natural Science, Duy Tan University, Danang 550000, Vietnam
  6.Nuclear Safety Division, Technical Support Center for Radiation and Nuclear Safety and Emergency Response, Vietnam Agency for Radiation and Nuclear Safety (VARANS), 113 Tran Duy Hung str., Cau Giay dist., Hanoi 100000, Vietnam
- Author bio：
  
  cuongnk.re@dnri.vn
  cuongnk.re@dnri.vn
- Funds：
- DOI：10.1007/s41365-021-00855-5
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Viet-Phu Tran, Kien-Cuong Nguyen, Donny Hartanto, et al. Development of a PARCS/Serpent model for neutronics analysis of the dalat nuclear research reactor. [J]. Nuclear Science and Techniques 32(2):15(2021)
DOI：

Viet-Phu Tran, Kien-Cuong Nguyen, Donny Hartanto, et al. Development of a PARCS/Serpent model for neutronics analysis of the dalat nuclear research reactor. [J]. Nuclear Science and Techniques 32(2):15(2021) DOI： 10.1007/s41365-021-00855-5.

摘要

Abstract

Cross-section homogenization for full-core calculations of small and complex reactor configurations, such as research reactors, has been recently recognized as an interesting and challenging topic. This paper presents the development of a PARCS/Serpent model for the neutronics analysis of a research reactor type TRIGA Mark-II loaded with Russian VVR-M2 fuel (known as the Dalat Nuclear Research Reactor or DNRR). The full-scale DNRR model and a supercell model for a shim/safety rod and its surrounding fuel bundles with the Monte Carlo code Serpent 2 were proposed to generate homogenized few-group cross-sections for full-core diffusion calculations with PARCS. The full-scale DNRR model with Serpent 2 was also utilized as a reference to verify the PARCS/Serpent calculations. Comparison of the effective neutron multiplication factors, radial and axial core power distributions, and control rod worths showed a generally good agreement between PARCS and Serpent 2. In addition, the discrepancies between the PARCS and Serpent 2 results are also discussed. Consequently, the results indicate the applicability of the PARCS/Serpent model for further steady state and transient analyses of the DNRR.

关键词

Keywords

PARCSSerpent 2Group constantDNRR

references

J. Leppänen, M. Pusa, T. Viitanen et al., The Serpent Monte Carlo code: Status, development and applications in 2013. Ann. Nucl. Energy 82, 142-150 (2015). doi: 10.1016/j.anucene.2014.08.024http://doi.org/10.1016/j.anucene.2014.08.024

R.S.I.C. Center, A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design. Oak Ridge, Tennessee, June (2011).

R.J. Stammler, Helios Methods. Studsvik Scandpower (2009).

X. Qin, M. Li, H. Liao et al., Neutronics analysis of commercial pressurized water reactor loaded with FCM fuel. Nuclear Techniques 43 (8): 080007 (2020). doi: 10.11889/j.0253-3219.2020.hjs.43.080007http://doi.org/10.11889/j.0253-3219.2020.hjs.43.080007 (in Chinese)

T. Downar, Y. Xu, V. Seker, PARCS v3.0 US NRC Core Neutronics Simulator - Theory Manual. RES/U.S. NRC, Rockville, Md (2012).

A. Rais, D. Siefman, M. Hursin et al., Neutronics modeling of the CROCUS reactor with SERPENT and PARCS codes, in Proc. M&C 2017, Jeju, Korea, April 16-20 (2017).

F. Fejt, J. Frybort, Analysis of a small-scale reactor core with PARCS/Serpent. Ann. Nucl. Energy 117, 25-31 (2018). doi: 10.1016/j.anucene.2018.03.002http://doi.org/10.1016/j.anucene.2018.03.002

A. Dambrosio, M. Rušcˇák, G. Mazzini et al., Neutronic analysis of the LVR-15 research reactor using the PARCS code. Ann. Nucl. Energy 117, 145-154 (2018). doi: 10.1016/j.anucene.2018.03.009http://doi.org/10.1016/j.anucene.2018.03.009

G. Baiocco, A. Petruzzi, S. Bznuni et al., Analysis of a small PWR core with the PARCS/Helios and PARCS/Serpent code systems. Ann. Nucl. Energy 107, 42-48 (2018). doi: 10.1016/j.anucene.2017.04.008http://doi.org/10.1016/j.anucene.2017.04.008

A. Rais, D. Siefman, G. Girardin et al., PARCS few-group homogenized parameters generation using Serpent Monte Carlo code at the CROCUS reactor, in Proc. IGORR 2014, 1-11 (2014).

A. Rais, D. Siefman, G. Girardin et al., Methods and models for the coupled neutronics and thermal-hydraulics analysis of the CROCUS reactor at EFPL. Sci. Technol. Nucl. Ins., 2015, 237646 (2015). doi: 10.1155/2015/237646http://doi.org/10.1155/2015/237646

D. Siefman, G. Girardin, A. Rais et al., Full core modeling techniques for research reactors with irregular geometries using Serpent and PARCS applied to the CROCUS reactor. Ann. Nucl. Energy 85, 434-443 (2015). doi: 10.1016/j.anucene.2015.05.004http://doi.org/10.1016/j.anucene.2015.05.004

J. Leppänen, M. Pusa, E. Fridman, Overview of methodology for spatial homogenization in the Serpent 2 Monte Carlo code. Ann. Nucl. Energy 96, 126-136 (2016). doi: 10.1016/j.anucene.2016.06.007http://doi.org/10.1016/j.anucene.2016.06.007

G. Zhang, J. Liu, W.S. Yang, Application of PROTEUS-MOC for the steady state simulations of TREAT. Nuclear Techniques 43 (4): 040008 (2020). doi: 10.11889/j.0253-3219.2020.hjs.43.040008http://doi.org/10.11889/j.0253-3219.2020.hjs.43.040008 (in Chinese)

S.B. Alam, D. Kumar, B. Almutairi et al., Lattice benchmarking of deterministic, Monte Carlo and hybrid Monte Carlo reactor physics codes for the soluble-boron-free SMR cores. Nucl. Eng. Des. 356 110350 (2020). doi: 10.1016/j.nucengdes.2019.110350http://doi.org/10.1016/j.nucengdes.2019.110350

T. Li, K. Zhuang, W. Shang et al., Preliminary neutronics design of space nuclear reactor based on molten salt cooling. Nuclear Techniques 43 (8): 080006 (2020). doi: 10.11889/j.0253-3219.2020.hjs.43.080006http://doi.org/10.11889/j.0253-3219.2020.hjs.43.080006 (in Chinese)

NRI, Safety Analysis Report for the Dalat Nuclear Research Reactor. Nuclear Research Institute, Dalat, Viet Nam (2012).

IAEA, History, Development, and Future of TRIGA Research Reactors. Technical Reports Series No. 482 (2016).

D.B . (editor) Pelowitz, MCNP6TM User’s Manual, Version 1.0. LA-CP-13-00634, Rev. 0, Los Alamos National Laboratory, May (2013).

Q.B. Do, HN. Tran, Q.H. Ngo et al., Determination of fuel burnup distribution of a research reactor based on measurements at subcritical conditions. Nucl. Sci. Tech. 29, 174 (2018). doi: 10.1007/s41365-018-0511-0http://doi.org/10.1007/s41365-018-0511-0

G.T.T. Phan, Q.B. Do, Q.H. Ngo et al., Application of differential evolution algorithm for fuel loading optimization of the DNRR research reactor. Nucl. Eng. Des. 362, 110582 (2020). doi: 10.1016/j.nucengdes.2020.110582http://doi.org/10.1016/j.nucengdes.2020.110582

J. Rhodes, K. Smith, D. Lee, Casmo-5 development and applications, in Proc. PHYSOR-2006, ANS Topical Meeting on Reactor Physics (2006).

D. Ćalić, G. Žerovnik, A. Trkov et al., Validation of the Serpent 2 code on TRIGA Mark II benchmark experiments. Appl. Radiat. Isot. 107, 165-170 (2016). doi: 10.1016/j.apradiso.2015.10.022http://doi.org/10.1016/j.apradiso.2015.10.022

C. Castagna, D. Chiesa, A. Cammi et al., A new model with Serpent for the first criticality benchmarks of the TRIGA Mark II reactor. Ann. Nucl. Energy 113, 171-176 (2018). doi: 10.1016/j.anucene.2017.11.011http://doi.org/10.1016/j.anucene.2017.11.011

M.B. Chadwick, M. Herman, P. Obložinský, et al., ENDF/B-VII.1 nuclear data for science and technology: Cross sections, covariances, fission product yields and decay data. Nucl. Data Sheets 112 (12), 2887-2996 (2011). doi: 10.1016/j.nds.2011.11.002http://doi.org/10.1016/j.nds.2011.11.002

J. Leppänen, WIMS 172-group structure. Serpent Wiki (2020). http://serpent.vtt.fi/mediawiki/index.php/WIMS_172-group_structure. Accessed: 2020-06-20

N.K. Cuong, N.T. Dung, T.V. Phu et al., Modeling of the Dalat nuclear research reactor (DNRR) with the Serpent 2 Monte Carlo code. Nucl. Sci. Technol. 9 (3), 21-29 (2019).

L.B. Vien, L.V. Vinh, H.T. Nghiem et al., Design analyses for full core conversion of the Dalat nuclear research reactor. Nucl. Sci. Technol. 4 (1), 10-25 (2014).

G. Phan, H.N. Tran, K.C. Nguyen et al., Comparative analysis of the Dalat nuclear research reactor with HEU fuel using SRAC and MCNP5. Sci. Technol. Nucl. Ins., 2017, 2615409 (2017). doi: 10.1155/2017/2615409http://doi.org/10.1155/2017/2615409

Q. B. Do, G.T.T. Phan, K.C. Nguyen et al., Criticality and rod worth analysis of the DNRR research reactor using the SRAC and MCNP5 codes. Nucl. Eng. Des. 343, 197-209 (2019). doi: 10.1016/j.nucengdes.2019.01.011http://doi.org/10.1016/j.nucengdes.2019.01.011

J. Leppänen, Boundary conditions. Serpent Wiki (2020). http://serpent.vtt.fi/mediawiki/index.php/Boundary_conditions. Accessed: 2020-10-20

T. Clerc, A. Hbert, H. Leroyer et al., An advanced computational scheme for the optimization of 2D radial reflector calculations in pressurized water reactors. Nucl. Eng. Des. 273, 560-575 (2014). doi: 10.1016/j.nucengdes.2014.03.044http://doi.org/10.1016/j.nucengdes.2014.03.044

P. Suk, Radial reflector discontinuity factors iteration scheme at VR-1. Nucl. Eng. Des. 366 110737 (2020). doi: 10.1016/j.nucengdes.2020.110737http://doi.org/10.1016/j.nucengdes.2020.110737

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