Development and Verification of a Beyond-Design-Basis Accident Model for Xi’an Pulsed Reactor

Chen Sen; Li Huaqi; Li Da; Chen Lixin; Tian Xiaoyan; Shi Leitai; Luo Xiaofei; Zhu Lei

doi:10.13832/j.jnpe.2024.060019

Volume 46 Issue 3

Jun. 2025

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Article Navigation > Nuclear Power Engineering > 2025 > 46(3): 61-67

Chen Sen, Li Huaqi, Li Da, Chen Lixin, Tian Xiaoyan, Shi Leitai, Luo Xiaofei, Zhu Lei. Development and Verification of a Beyond-Design-Basis Accident Model for Xi’an Pulsed Reactor[J]. Nuclear Power Engineering, 2025, 46(3): 61-67. doi: 10.13832/j.jnpe.2024.060019

Citation:

Chen Sen, Li Huaqi, Li Da, Chen Lixin, Tian Xiaoyan, Shi Leitai, Luo Xiaofei, Zhu Lei. Development and Verification of a Beyond-Design-Basis Accident Model for Xi’an Pulsed Reactor[J]. Nuclear Power Engineering, 2025, 46(3): 61-67. doi: 10.13832/j.jnpe.2024.060019

Citation:

Chen Sen, Li Huaqi, Li Da, Chen Lixin, Tian Xiaoyan, Shi Leitai, Luo Xiaofei, Zhu Lei. Development and Verification of a Beyond-Design-Basis Accident Model for Xi’an Pulsed Reactor[J]. Nuclear Power Engineering, 2025, 46(3): 61-67. doi: 10.13832/j.jnpe.2024.060019

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Development and Verification of a Beyond-Design-Basis Accident Model for Xi’an Pulsed Reactor

doi: 10.13832/j.jnpe.2024.060019

Northwest Institute of Nuclear Technology, Xi’an, 710024, China

Received Date: 2024-06-18
Accepted Date: 2024-08-06
Rev Recd Date: 2024-08-06

Available Online: 2025-06-09

Publish Date: 2025-06-09

Abstract

Abstract

Xi'an Pulsed Reactor (XAPR) adopts uranium-zirconium hydride fuel element, which has a good inherent safety. However, with the long-term operation of XAPR, severe accidents may occur in the core, leading to the damage of fuel element cladding, so it is necessary to carry out analysis and research on the beyond-design-basis accidents for XAPR. In this paper, based on the ISAA code, an integrated analysis code for XAPR beyond-design-basis accidents was developed by adding physical property model, fuel oxidation model, fuel rod mechanics model and point kinetics model. The accuracy and applicability of each model were validated. Besides, based on the developed code, the steady-state operation condition and large break loss of coolant accident condition were calculated and analyzed respectively, and the results were in good agreement with the literature values. The results showed that the models developed in this paper were suitable for the simulation and analysis of XAPR and could be used for the subsequent beyond-design-basis accident calculation and analysis of XAPR.
- Xi’an Pulsed Reactor (XAPR),
- Severe accident,
- Code development,
- Model validation

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References(15)

References

[1]	景春元. 脉冲反应堆动态特性与失水事故研究[D]. 西安: 西安交通大学,1999.
[2]	陈立新,赵柱民,袁建新,等. 西安脉冲堆大破口失水事故分析[J]. 原子能科学技术,2009, 43(8): 678-682. doi: 10.7538/yzk.2009.43.08.0678
[3]	朱磊,陈立新,赵柱民,等. 西安脉冲反应堆超设计基准事故动态特性分析[J]. 原子能科学技术,2012, 46(S1): 800-806.
[4]	田晓艳,陈森,李达,等. 西安脉冲堆失水事故缓解措施分析评价[J]. 现代应用物理,2023, 14(4): 040402.
[5]	田晓艳,陈森,杨宁,等. 西安脉冲堆反应性引入事故瞬态安全特性分析[J]. 原子能科学技术,2020, 54(11): 2089-2097. doi: 10.7538/yzk.2019.youxian.0854
[6]	GAO P C, ZHANG B, LI J S, et al. Development of mechanistic cladding rupture model for integrated severe accident code ISAA. Part Ⅰ: Module verification and application in CAP1400[J]. Annals of Nuclear Energy, 2021, 158: 108305. doi: 10.1016/j.anucene.2021.108305
[7]	陈伟,江新标,陈立新,等. 铀氢锆脉冲反应堆物理与安全分析[M]. 北京: 科学出版社,2018: 13-14, 81-82.
[8]	SIMNAD M T. The U ZrH_x alloy: its properties and use in TRIGA fuel[J]. Nuclear Engineering and Design, 1981, 64(3): 403-422.
[9]	SIMNAD M T, FOUSHEE F C, WEST G B. Fuel elements for pulsed TRIGA^® research reactors[J]. Nuclear Technology, 1976, 28(1): 31-56. doi: 10.13182/NT76-A31537
[10]	HILTON B A. Review of oxidation rates of DOE spent nuclear fuel: part 1: metallic fuel: ANL-00/24[R]. Argonne: Argonne National Laboratory, 2000.
[11]	陈伟东,闫淑芳,钟学奎,等. 氢化锆在450~600℃下氧化动力学的研究[J]. 稀有金属材料与工程,2011, 40(6): 1038-1040.
[12]	王建伟,黄倩,陈伟东,等. 氢化锆在高温水蒸气中的氧化行为[J]. 材料导报,2011, 25(12): 4-6.
[13]	MARCHBANKS M F, MOEN R A, IRVIN J E. Nuclear systems materials handbook: HEDL-SA-877[R]. Richland: Hanford Engineering Development Laboratory, 1976.
[14]	FANNING T H, BRUNETT A J, SUMNER T. The SAS4A/SASSYS-1 safety analysis code system, version 5: ANL/NE-16/19[R]. Argonne: Argonne National Laboratory, 2017.
[15]	袁红球,胡大璞. 高次端点浮动法-解点堆中子动力学方程[J]. 核动力工程,1995, 16(2): 124-128.