Development and Verification of a Beyond-Design-Basis Accident Model for Xi’an Pulsed Reactor
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摘要: 西安脉冲堆（XAPR）采用铀氢锆燃料元件，固有安全性好。但随着XAPR长时间运行，堆芯可能发生严重事故，导致燃料元件包壳破损，因此需要开展XAPR超设计基准事故分析研究。本文基于ISAA程序，通过添加物性模型、燃料氧化模型、燃料棒力学模型和点堆动力学模型，开发了适用于XAPR的超设计基准事故一体化分析程序，并对所添加模型的准确性和适用性进行了验证。最后基于所开发的程序分别计算分析了XAPR稳态运行工况和大破口失水事故工况，并与文献结果进行比较，计算结果符合良好。本文开发的模型适用于XAPR模拟分析，为后续深入开展XAPR超设计基准事故计算分析奠定了基础。
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关键词:
- 西安脉冲堆（XAPR） /
- 严重事故 /
- 程序开发 /
- 模型验证
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.-
Key words:
- Xi’an Pulsed Reactor (XAPR) /
- Severe accident /
- Code development /
- Model validation
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表 1 中子密度计算值比较
Table 1. Comparison of Calculation Value of Neutron Density
时间/s 中子密度精确解[15]/cm−3 中子密度程序计算值/cm−3 误差/% 0.1 1.799528 1.799061 0.026 0.2 1.851268 1.853331 0.111 0.3 1.900405 1.903730 0.175 0.4 1.947593 1.952283 0.241 0.5 1.993313 1.999469 0.309 0.6 2.037894 2.045645 0.380 0.7 2.081671 2.091080 0.452 0.8 2.124816 2.135981 0.525 0.9 2.167504 2.180511 0.600 1.0 2.209841 2.224800 0.677 
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表 2 环向应变最大值计算结果对比 %
Table 2. Comparison of Maximum Circumferential Strain Results
燃料棒 实验值[6] FRAPCON计算值 本程序计算值 IRRMP-16 0.0552 0.2122 0.0683 IRRMP-18 0.0203 0.0029 0.0362
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表 3 XAPR稳态计算结果对比
Table 3. Comparison of Steady State Calculation Results of XAPR
参数 设计值 程序计算值 相对误差/% 堆芯压力 0.1627 MPa 0.16067 MPa 1.620 堆芯流量 12.1300 kg/s 12.2847 kg/s 1.278 燃料最高温度 776.9500 K 772.8324 K 0.530 包壳最高温度 405.8500 K 406.4043 K 0.137 堆芯入口温度 308.1500 K 318.1311 K 3.239 堆芯出口温度 361.380 K 356.849 K 1.254 离心泵流量 48.3300 kg/s 48.4427 kg/s 0.233 最大表面热流密度 0.7080 MW/m2 0.6778 MW/m2 4.265
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表 4 大破口失水事故工况事故序列对比
Table 4. Comparison of Large Break Loss of Coolant Acccident Sequences
事件 时间/s 功率/kW 燃料最高温度/℃ 包壳最高温度/℃ 开始失水 0.0(0.0) 2000(2000) 499.68(503.80) 133.25(132.70) 停堆信号 13.07(10.8) 2000(2000) 499.68(503.80) 133.25(132.70) 冷却剂循环中断 238.44(230) 67.102(63.100) 102.79(74.20) 90.82(69.10) 堆芯全部裸露 296.02(283) 65.327(59.7) 129.84(134.20) 128.88(132.70) 燃料温度达到最高 296.02(283) 26.71(21.1) 1188.105(1171.000) 1177.47(1169.20) 小括号内数据为文献值[2]。
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[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 ZrHx 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. -
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