不同类型核燃料对热管冷却反应堆燃耗性能的影响

秦凯文; 杨波; 王子鸣; 钱云琛; 刘豪杰; 刘义保

doi:10.11884/HPLPB202234.220156

不同类型核燃料对热管冷却反应堆燃耗性能的影响

doi: 10.11884/HPLPB202234.220156

秦凯文^1,,
杨波^{1, 2, ,},
王子鸣¹,
钱云琛¹,
刘豪杰¹,
刘义保^{1, 2, ,}

1.
东华理工大学核科学与工程学院，南昌 330013
2.
东华理工大学核资源与环境国家重点实验室，南昌 330013

基金项目: 国家自然科学基金项目(11965001); 江西省教育厅科技计划重点项目(GJJ170428); 江西省自然科学基金项目(20212BAB201004)

详细信息

作者简介:
秦凯文，2549820280@qq.com

通讯作者:
杨　波，Boyang@ecut.edu.cn

刘义保，ybliu@ecut.edu.cn

中图分类号: TL32
计量
- 文章访问数: 623
- HTML全文浏览量: 250
- PDF下载量: 101
- 被引次数: 0
出版历程
- 收稿日期: 2022-05-16
- 修回日期: 2022-09-26
- 网络出版日期: 2022-11-02
- 刊出日期: 2022-11-02

Influence of different types of nuclear fuel on burnup performance of heat pipe cooled reactor

Qin Kaiwen^1
,,
Yang Bo^{1, 2
, ,},
Wang Ziming¹,
Qian Yunchen¹,
Liu Haojie¹,
Liu Yibao^{1, 2
, ,}

1.
Nuclear Science and Technology, East China University of Technology, Nanchang 330013, China
2.
National Key Laboratory of Nuclear Resources and Environment, East China University of Technology, Nanchang 330013, China

摘要

摘要: 热管冷却反应堆采用固态反应堆设计理念，具有功率密度高、结构紧凑、固有安全性高等特点，在深空探索、深海勘探、偏远地区等场景中具有广阔的应用前景。核燃料作为热管冷却反应堆的重要组成部分，不同类型核燃料在堆芯燃耗分析时会呈现不同的中子学性能。基于美国爱达荷国家实验室(INL)提出的热管冷却反应堆INL Design A，利用清华大学蒙特卡罗中子输运程序RMC (Reactor Monte Carlo code)建立堆芯物理模型，选取UO₂，(U_0.9Pu_0.1)O₂，U-10Zr，U-8Pu-10Zr，UN，UC这6种核燃料开展燃耗计算，分析了不同核燃料、不同功率水平对热管冷却反应堆堆芯燃耗性能的影响。计算结果表明：在堆芯燃耗深度相同情况下(20.8 GW·d·t⁻¹)，装载U-8Pu-10Zr燃料的堆芯所需²³⁵U富集度最低(9.8%)，具有较好的U-Pu增殖性能。堆芯功率处于5 MW的热管冷却反应堆，燃料中²⁴¹Pu的存在不仅没起到增大堆芯燃耗深度的作用，反而导致堆芯剩余反应性和堆芯寿期末次锕系核素(MAs)的产量增大，影响反应堆的安全性与经济性。因此，对于装载含有Pu燃料的小功率长寿期热管冷却反应堆，需重点关注²⁴¹Pu对堆芯燃耗性能的影响。
- 热管冷却反应堆 /
- 燃耗计算 /
- RMC程序 /
- ²⁴¹Pu核素
Abstract: The heat pipe cooled reactor adopts the solid-state reactor design concept, and it has the characteristics of high power density, compact structure and high inherent safety. It has been extensively used for deep space exploration, deep sea exploration, remote areas electricity markets and other scenarios. Nuclear fuel is an important part of the heat pipe cooling reactor, different types of nuclear fuel will reflect different neutronics performance on the reactor burnup analysis. In this paper, based on the heat pipe cooled reactor INL Design A proposed by the Idaho National Laboratory (INL), the burnup calculation is done by selecting six nuclear fuels : UO₂, (U_0.9Pu_0.1)O₂, U-10Zr, U-8Pu-10Zr, UN and UC. The effects of different nuclear fuel and power levels on the burnup performance of heat pipe cooled reactor core were analyzed. The calculation results show that under the same core burnup depth (20.8 GW·d·t⁻¹), the core loaded with U-8Pu-10Zr fuel requires the lowest ²³⁵U enrichment (9.8%), and has better U-Pu breeding. For the heat pipe cooling reactor with the core power of 5 MW, the presence of ²⁴¹Pu in the fuel does not increase the core burnup depth, but leads to the increase of residual reactivity of the core and the yield of the secondary actinides nuclides (MAs) in the core end of life, which affects the safety and economy of the reactor. Therefore, for the low-power and long-life heat pipe cooled reactor loaded with Pu fuel, it is necessary to focus on the influence of ²⁴¹Pu on the core burnup performance.
- heat pipe cooled reactor /
- burnup calculation /
- RMC code /
- ²⁴¹Pu nuclide

HTML全文

图 1 RMC输出INL Design A几何结构图

Figure 1. Geometries of INL Design A by RMC

下载: 全尺寸图片幻灯片

图 2 不同核燃料堆芯k_eff随燃耗深度的变化

Figure 2. Core k_eff of different nuclear fuel varies with burnup depth

下载: 全尺寸图片幻灯片

图 3 堆芯k_eff随控制鼓旋转的变化

Figure 3. Core k_eff change with rotation of control drum

下载: 全尺寸图片幻灯片

图 4 不同功率反应堆堆芯燃耗

Figure 4. Reactor core burnup of different power

下载: 全尺寸图片幻灯片

图 5 不同功率下不含²⁴¹Pu反应堆堆芯燃耗

Figure 5. Reactor core burnup without ²⁴¹Pu at different powers

下载: 全尺寸图片幻灯片

图 6 U-Pu燃料循环燃耗链

Figure 6. Burnup chain of U-Pu cycle

下载: 全尺寸图片幻灯片

表 1 有效增值系数k_eff计算结果

Table 1. Calculation results of effective increment coefficient k_eff

control condition	calculated value of k_eff in this paper	calculated value of k_eff of INL	difference of k_eff /10⁻⁵
all poisons out	1.028 82±0.000 33	1.028 25	57
control drums rotation 180°	0.950 98±0.000 33	0.950 42	56
annular shutdown rod in	0.945 89±0.000 33	0.945 55	34
solid shutdown in	0.959 30±0.000 34	0.959 33	−3
all poisons in	0.845 04±0.000 33	0.845 94	−90

下载: 导出CSV

表 2 典型压水堆乏燃料中钚的含量

Table 2. Plutonium composition in a typical PWR spent fuel

plutonium isotope	mass fraction/%
²³⁸Pu	2.332
²³⁹Pu	56.873
²⁴⁰Pu	26.997
²⁴¹Pu	6.105
²⁴²Pu	7.693

下载: 导出CSV

表 3 不同功率水平反应堆燃料Pu同位素分析

Table 3. Pu isotope analysis of reactor fuel at different power levels

nuclide	quantity of fissile nuclides and fissionable nuclides/kg		difference/kg
nuclide	at 5 MW	at 200 MW	difference/kg
²³⁸Pu	8.319	10.422	2.103
²³⁹Pu	311.710	311.134	−0.576
²⁴⁰Pu	137.792	138.676	0.884
²⁴¹Pu	3.533	29.835	26.302
²⁴²Pu	37.129	37.447	0.318

下载: 导出CSV

表 4 不同功率水平反应堆乏燃料分析

Table 4. Spent fuel analysis of reactor at different power levels

spent nuclear fuel	nuclide	half-life/a	quantity of fissile nuclides and fissionable nuclides/kg		difference/kg
spent nuclear fuel	nuclide	half-life/a	at 5 MW	at 200 MW	difference/kg
MAs	²³⁷Np	2.14×10⁶	2.71	0.71	2
	²⁴¹Am	432.2	27.46	2.15	25.31
	²⁴³Am	7 380	1.80	1.72	0.08
	²⁴³Cm	8 500	3.72×10⁻⁴	0	3.72×10⁻⁴
	²⁴⁴Cm	18.1	0.08	0.14	−0.06
	²⁴⁵Cm	28.5	5.43×10⁻⁴	4.25×10⁻⁴	1.18×10⁻⁴
LLFPs	⁹⁹Tc	2.11×10⁵	2.70	2.70	0
	¹²⁹I	1.27×10⁷	0.61	0.62	−0.01
	¹³⁵Cs	2.30×10⁶	3.91	4.34	−0.43

下载: 导出CSV

参考文献(17)

[1]	余红星, 马誉高, 张卓华, 等. 热管冷却反应堆的兴起和发展[J]. 核动力工程, 2019, 40(4):1-8 doi: 10.13832/j.jnpe.2019.04.0001 Yu Hongxing, Ma Yugao, Zhang Zhuohua, et al. Initiation and development of heat pipe cooled reactor[J]. Nuclear Power Engineering, 2019, 40(4): 1-8 doi: 10.13832/j.jnpe.2019.04.0001
[2]	王傲, 申凤阳, 胡古, 等. 热管空间核反应堆电源的研究进展[J]. 核技术, 2020, 43:060002 doi: 10.11889/j.0253-3219.2020.hjs.43.060002 Wang Ao, Shen Fengyang, Hu Gu, et al. A survey of heatpipe space nuclear reactor power supply[J]. Nuclear Techniques, 2020, 43: 060002 doi: 10.11889/j.0253-3219.2020.hjs.43.060002
[3]	McClure P R, Poston D I, Dasari V R, et al. Design of megawatt power level heat pipe reactors[R]. Los Alamos: Los Alamos National Laboratory, 2015.
[4]	Sterbentz J W, Werner J E, McKellar M G, et al. Special purpose nuclear reactor (5 MW) for reliable power at remote sites assessment report[R]. Idaho Falls: Idaho National Laboratory, 2017.
[5]	Sterbentz J W, Werner J E, Hummel A J, et al. Preliminary assessment of two alternative core design concepts for the special purpose reactor[R]. Idaho Falls: Idaho National Laboratory, 2018.
[6]	屈伸, 曹良志, 郑琪, 等. 热管堆高温数据库的制作及堆芯初步物理计算[J]. 现代应用物理, 2017, 8:041202 doi: 10.12061/j.issn.2095-6223.2017.041202 Qu Shen, Cao Liangzhi, Zheng Qi, et al. Development of high-temperature nuclear database and preliminary physical computation of a heat pipe reactor[J]. Modern Applied Physics, 2017, 8: 041202 doi: 10.12061/j.issn.2095-6223.2017.041202
[7]	李冠兴, 周邦新, 肖岷, 等. 中国新一代核能核燃料总体发展战略研究[J]. 中国工程科学, 2019, 21(1):6-11 Li Guanxing, Zhou Bangxin, Xiao Min, et al. Overall development strategy of China’s new-generation nuclear fuel[J]. Strategic Study of CAE, 2019, 21(1): 6-11
[8]	Fütterer M A, D’Agata E, Laurie M, et al. Next generation fuel irradiation capability in the High Flux Reactor Petten[J]. Journal of Nuclear Materials, 2009, 392(2): 184-191. doi: 10.1016/j.jnucmat.2009.03.030
[9]	Greenquist I, Powers J J. Sensitivity and uncertainty of the IFR-1 BISON benchmark[R]. Oak Ridge: Oak Ridge National Laboratory, 2022.
[10]	IAEA. Thermophysical properties of materials for nuclear engineering: a tutorial and collection of data[M]. Vienna: IAEA, 2008: 92-110.
[11]	Wang Kan, Li Zeguang, She Ding, et al. RMC—A Monte Carlo code for reactor core analysis[J]. Annals of Nuclear Energy, 2015, 82: 121-129. doi: 10.1016/j.anucene.2014.08.048
[12]	刘晓波, 胡泽华. 蒙卡程序计算临界基准题测试检验ENDF/B-VIII. 0核数据库[J]. 强激光与粒子束, 2022, 34:026003 doi: 10.11884/HPLPB202234.210366 Liu Xiaobo, Hu Zehua. Monte Carlo calculation of critical benchmarking models for testing ENDF/B-VIII. 0 nuclear data[J]. High Power Laser and Particle Beams, 2022, 34: 026003 doi: 10.11884/HPLPB202234.210366
[13]	胡赟, 徐銤. 快堆金属燃料的发展[J]. 原子能科学技术, 2008, 42(9):810-815 Hu Yun, Xu Mi. Development of metallic fuel for fast reactor[J]. Atomic Energy Science and Technology, 2008, 42(9): 810-815
[14]	Gao Yucui, Cao Liangzhi, Yang Yongwei, et al. Physical study of an ultra-long-life small modular fast reactor loaded with U-Pu-Zr fuel[J]. Annals of Nuclear Energy, 2020, 142: 107390. doi: 10.1016/j.anucene.2020.107390
[15]	Luzzi L, Cammi A, Di Marcello V, et al. Application of the TRANSURANUS code for the fuel pin design process of the ALFRED reactor[J]. Nuclear Engineering and Design, 2014, 277: 173-187. doi: 10.1016/j.nucengdes.2014.06.032
[16]	Liu Bin, Wang Kai, Tu Jing, et al. Transmutation of minor actinides in the pressurized water reactors[J]. Annals of Nuclear Energy, 2014, 64: 86-92. doi: 10.1016/j.anucene.2013.09.042
[17]	Yang W S, Kim Y, Hill R N, et al. Long-lived fission product transmutation studies[J]. Nuclear Science and Engineering, 2004, 146(3): 291-318. doi: 10.13182/NSE04-A2411