CORE BURN-UP ANALYSIS OF THE RSG-GAS RESEARCH REACTOR USING DETERMINISTIC AND STOCHASTIC METHODSS

Authors

  • Tukiran Surbakti Research and Technology Center for Nuclear Reactor, Research Organization for Nuclear Energy, National Research and Innovation Agency (BRIN), Kawasan PUSPIPTEK Serpong Gd.80, Tangerang Selatan, Banten 15314, Indonesia
  • Surian Pinem Research and Technology Center for Nuclear Reactor, Research Organization for Nuclear Energy, National Research and Innovation Agency (BRIN), Kawasan PUSPIPTEK Serpong Gd.80, Tangerang Selatan, Banten 15314, Indonesia
  • Wahid Luthfi Research and Technology Center for Nuclear Reactor, Research Organization for Nuclear Energy, National Research and Innovation Agency (BRIN), Kawasan PUSPIPTEK Serpong Gd.80, Tangerang Selatan, Banten 15314, Indonesia
  • Donny Hartanto ᵇDepartment of Mechanical and Nuclear Engineering, University of Sharjah, P.O. BOX 27272, Sharjah, United Arab Emirates ᶜNuclear Energy System Simulation and Safety Research Group, Research Institute of Sciences and Engineering, University of Sharjah, P.O. BOX 27272, Sharjah, United Arab Emirates

DOI:

https://doi.org/10.11113/jurnalteknologi.v84.18425

Keywords:

RSG-GAS, WIMSD-5B, Batan-FUEL, Serpent 2, burn-up, core analysis

Abstract

Due to several characteristics, such as geometry, compact core, high coolant flow, and high neutron flux, the burn-up study of the RSG-GAS multi-purpose reactor provides challenges when employing a neutronic calculation. For the burn-up analysis, two calculating methodologies are used in the RSG-GAS: deterministic and probabilistic methods. The deterministic codes such as WIMSD-5B and Batan-FUEL are utilized, whereas the continuous-energy Monte Carlo code Serpent 2 is used for the stochastic method. WIMSD-5B is being used to produce a four-group cross-section that is needed by Batan-FUEL to do full core diffusion calculations. Burn-up calculations were performed at the whole fuel assemblies inside the core to see if the deterministic code, WIMSD-5B/Batan-FUEL, could accurately replicate the burn-up behavior of the RSG-GAS research reactor. The Serpent 2 calculation was also done with the exact models to provide a comparison. The results show that both Serpent 2 and WIMSD-5B/Batan-FUEL can perform the RSG-GAS burn-up analysis if appropriate treatments are made to the deterministic codes at both the assembly and core levels. There is a 5% difference in calculated fuel burn-up between deterministic and stochastic approaches.  

References

<http://nucleus.iaea.org/RRDB/.

Liem, P. H. 1994. Development and Verification of BATAN'S Standard, Two-Dimensional Multigroup Neutron Diffusion Codes (BATAN-2DIFF). Atom Indonesia. 20(2): 1-19.

Askew, J. R., Fayers, F. J., Kemshell, P. B. 1966. A General Description of the Lattice Code WIMS. J. Brit. l. Energy Soc. 5(4): 564.

Peng Hong Liem. 1996. Batan-FUEL: A General In-core Fuel Management Code. Atom Indonesia. 22(2): 67-80.

Liu, S. C. et al. 2015. Neutronics Comparative Analysis of Plate-type Research Reactor using Deterministic and Stochastic Methods. Ann. Nucl. Energy. 79: 133-142.

Pinem, S., et al. 2016. Fuel Element Burn-up Measurements for the Equilibrium LEU Silicide RSG GAS (MPR-30) Core Under a New Fuel Management Strategy. Annals of Nuclear Energy. 98: 211-217.

Aslina Br Ginting, Peng Hong Liem. 2015. Absolute Burnup Measurement of LEU Silicide Fuel Plate Irradiated in the RSG-GAS Multipurpose Reactor by Destructive Radiochemical Technique. Annals of Nuclear Energy. 85: 613-620.

Pinem, S., Sembiring, T. M., Liem, P. H. 2016. Neutronic and Thermal-Hydraulic Safety Analysis for the Optimization of the Uranium Foil Target in the RSG-GAS Reactor. Atom Indonesia. 42(3): 123-128.

Liem, P. H. 1997. Development of an In-Core Fuel Management Code for Searching Equilibrium Core in 2-D Reactor Geometry (BATAN-EQUIL-2D). Atom Indonesia. 23(1): 1-19.

Liem, P. H., Tagor, M. S. 2010. Design of Transition Cores of RSG GAS (MPR-30) with Higher Loading Silicide Fuel. Nuclear Engineering and Design. 240(6): 1433-1442.

Liem, P. H., et al. 2013. Nondestructive Burnup Verification by Gamma-ray Spectroscopy of LEU Silicide Fuel Plates Irradiated in the RSG GAS Multipurpose Reactor. Annals of Nuclear Energy. 56: 57-65.

Sembiring, T. M., Liem, P. H. 1999. Validation of BATAN-3DIFF Code on the 3-D Model of the IAEA 10 MWth Benchmark Core for Partially-inserted Control Rods. Atom Indonesia. 25(2): 91-100.

Liem, P. H., Surbakti, T., Hartanto, D. 2018. Kinetic Parameters Evaluation on the First Core of the RSG GAS (MPR-30) using Continuous Energy Monte Carlo Method. Progress in Nuclear Energy. 109: 196-203.

Halsall, M. J. 1997. WIMSD, A Neutronic Code for Standard Lattice Physics Analysis, Distributed by the NEA Databank, NEA-1507/04. Research Reactors, 2010. Purpose and Future. IAEA, Vienna.

Roth, M. J., Macdougall, J. D., Kemshell, P. B. 1967. The Preparation of Input Data for WIMS, AEEW-R538.

IAEA-TECDOC-233. 1980. Research Reactor Core Conversion from the Use of Highly Enriched Uranium to the Use of Low Enriched Uranium Fuel. IAEA, Vienna, Austria.

Hosoya, T., Kato, T., Murayama, Y. 2007. Investigation of JRR-3 Control Rod Worth Changed with Burn-up of Follower Fuel Elements. International Research Reactors: Safe Management and Effective Utilization, Australia, IAEA, 2010.

Askew, J. R., Fayers, F. J., Kemshell, P. B. 1966. A General Description of the Lattice Code WIMS. J. Brit. l. Energy Soc. 5(4): 564.

Iwasaki, J., Ichikawa, H., Tsuruta, H., 1984. Neutronics Calculation of Upgraded JRR-3 Research Reactor. Few-group Constants. Japan Atomic Energy Research Inst. JAERI-M–84-159.

Iwasaki, J., Tsuruta, H., Ichikawa, H. 1985. Neutronics Calculation of Upgraded JRR-3. Supplement: Fuel, Control Rod, Reflector. Japan Atomic Energy Research Inst. JAERI-M–85-062.

Liu, Y. K. 2006. Studies on the Development of Three-dimensions Neutron Transport Code. Ph.D. Thesis. Tsinghua University, China.

She, D. et al. 2013. Development of Burnup Methods and Capabilities in Monte Carlo Code RMC. Ann. Nucl. Energy. 51: 289-294.

She, D. et al. 2014. 2D full-core Monte Carlo Pin-by-pin Burnup Calculations with the RMC Code. Ann. Nucl. Energy. 64: 201-205.

Wang, K. et al. 2015. RMC – A Monte Carlo Code for Reactor Core Analysis. Ann. Nucl. Energy. 82: 121-129.

Yuan, L., Kang, Y. 1998. Problems Concerned in Fuel Design of CARR. 1998 International Meeting on Reduced Enrichment for Research and Test Reactors, Sao Paulo, Brazil.

Hellens, R. L., Price, G. A. 1964. Reactor Physics Data for Water-moderated lattices of Slightly Enriched Uranium. React. Technol. Select. Rev. 529.

Published

2022-07-29

How to Cite

Surbakti, T., Pinem, S. ., Luthfi, W., & Hartanto, D. (2022). CORE BURN-UP ANALYSIS OF THE RSG-GAS RESEARCH REACTOR USING DETERMINISTIC AND STOCHASTIC METHODSS. Jurnal Teknologi, 84(5). https://doi.org/10.11113/jurnalteknologi.v84.18425

Issue

Section

Science and Engineering