Preliminary neutronic analyses for annular VHTR with different coated fuel particle designs
-
Odmaa Sambuu
and
Khukhsuvd Batsaikhan
Abstract
The purpose of the present work is to determine the neutronic characteristics of annular cylindrical prismatic Very High Temperature Gas-cooled Reactor (VHTR) with different advanced fuel designs. The study includes detailed neutronic investigations of the eight types of TRISO (containing SiC layer) and TRIZO (containing ZrC layer) fuel particles for annular VHTR core. All calculations were performed using continuous energy Monte Carlo code MVP2.0 and MVPBURN with JENDL4.0 nuclear data library. The obtained results indicated that SiC + PyC and PyC layers in TRISO fuel appeared to be not excellent choice for the annular VHTR core due to its shorter operation year and, lower fuel burnup. The annular cylindrical prismatic core fueled with TRIZO fuel particle shows good neutronic characteristics compared to those designed with other TRISO fuel particles.
Funding source: National University of Mongolia
Award Identifier / Grant number: P2023- 4611
Acknowledgments
All calculations were carried out using MINATO cluster server at Nuclear Research Center, National University of Mongolia and this work was supported by the National University of Mongolia under grant number P2023- 4611.
-
Research ethics: Not applicable.
-
Informed consent: Not applicable.
-
Author contributions: All authors have accepted responsibility for the entire content of this manuscript and approved its submission.
-
Use of Large Language Models, AI and Machine Learning Tools: None declared.
-
Conflict of interests: The author states no conflict of interest.
-
Research funding: National University of Mongolia grant number P2023-4611.
-
Data availability: Not applicable.
References
Konings, R.J.M., Allen, T.R., Stoller, R.E., and Yamanaka, S. (2012). Comprehensive nuclear material, 3. Elsevier, The Netherlands, pp. 215–236.Search in Google Scholar
Miftasani, F., Su’ud, Z., and Irwanto, D. (2020). Neutronic aspect assessment on the use of ZrC triso-coated particle (TRIZO) in a high-temperature gas-cooled reactor (HTGR). J. Phys.: Conf. Ser. 1493, https://doi.org/10.1088/1742-6596/1493/1/012029.Search in Google Scholar
Nagaya, Y., Okumura, K., Mori, T., and Nakagawa, M. (2005). MVP/GMVP II: general purpose monte-carlo code for neutron and photon transport calculations based on continuous energy and multigroup methods. Jap. Atomic Energy Res. Ins. 37(3).Search in Google Scholar
Okumura, K., et al.. (2005). MVP-BURN user’s manual. Japan Atomic Energy Agency, Japan.Search in Google Scholar
Sambuu, O., Terbish, J., Obara, T., Nanzad, N., and Byambajav, M. (2018a). Design parameters in an annular, prismatic HTGR for passive decay heat removal. Ann. Nucl. Energy 111: 441–448, https://doi.org/10.1016/j.anucene.2017.09.034.Search in Google Scholar
Sambuu, O., Batsaikhan, K., Byambajav, M., Nanzad, N., and Terbish, J. (2018b). Preliminary neutronic analyses on VHTR core design. Int. Electron. J. Nucl. Saf. Simul. 9: 200–205.Search in Google Scholar
Shibata, K., Iwamoto, O., Nakagawa, T., Iwamoto, N., Ichihara, A., Kunieda, S., Chiba, S., Furutaka, K., Otuka, N., Ohsawa, T., et al.. (2011). Japanese evaluated nuclear data library-JENDL-4.0, A new library for nuclear science and engineering. J. Nucl. Sci. Tech. 48: 1–30, https://doi.org/10.1080/18811248.2011.9711675.Search in Google Scholar
Zuhair, Z., Dwijayanto, R.A.P., Adrial, H., Suwoto, S., and Setiadipura, T. (2019). Neutronic effect of utilizing TRIZO particles on core characteristics of experimental power reactor. AIP Conf. Proc. 2180: 020005, https://doi.org/10.1063/1.5135514.Search in Google Scholar
© 2024 Walter de Gruyter GmbH, Berlin/Boston
Articles in the same Issue
- Frontmatter
- Study of reliability prediction method and uncertainty analysis of the pump and valve of a floating nuclear power plant
- Thermal-hydraulic and safety analysis of alternative ceramic fuels in next generation of VVER-1000 reactor
- Transient thermal analysis for optimal armor material in upgraded mockup for PST using MOOSE Framework
- Analysis of hypothetical channel flow blockage of a tubular fuel assembly in a 10 MWth nuclear research reactor
- Preliminary neutronic analyses for annular VHTR with different coated fuel particle designs
- Improved intelligent model predictive controller for the nuclear power reactor system
- Monte Carlo simulation of photo-peak efficiency response function and calculation of total efficiency for a NaI (Tl) scintillator detector
- Calculation and application research on real-time source term diffusion in NPP auxiliary building under accident circumstances
- Calendar of events – issue 6 of KERN
Articles in the same Issue
- Frontmatter
- Study of reliability prediction method and uncertainty analysis of the pump and valve of a floating nuclear power plant
- Thermal-hydraulic and safety analysis of alternative ceramic fuels in next generation of VVER-1000 reactor
- Transient thermal analysis for optimal armor material in upgraded mockup for PST using MOOSE Framework
- Analysis of hypothetical channel flow blockage of a tubular fuel assembly in a 10 MWth nuclear research reactor
- Preliminary neutronic analyses for annular VHTR with different coated fuel particle designs
- Improved intelligent model predictive controller for the nuclear power reactor system
- Monte Carlo simulation of photo-peak efficiency response function and calculation of total efficiency for a NaI (Tl) scintillator detector
- Calculation and application research on real-time source term diffusion in NPP auxiliary building under accident circumstances
- Calendar of events – issue 6 of KERN
