重水堆压力管延迟氢化物开裂应力强度因子门槛值测试方法研究

doi:10.11902/1005.4537.2024.221

中国腐蚀与防护学报

2025, Vol. 45

Issue (2): 381-387 CSTR: 32134.14.1005.4537.2024.221 DOI: 10.11902/1005.4537.2024.221

临氢关键材料服役行为研究专刊

本期目录 | 过刊浏览

重水堆压力管延迟氢化物开裂应力强度因子门槛值测试方法研究

鲍一晨¹, 石秀强¹, 孟凡江¹, 潘春婷², 明洪亮²(

)

^1.上海核工程研究设计院股份有限公司上海 200233
^2.中国科学院金属研究所沈阳 110016

Assessment Method of Threshold Stress Intensity Factor of Delayed Hydride Cracking for Pressure Tube in Heavy Water Reactor

BAO Yichen¹, SHI Xiuqiang¹, MENG Fanjiang¹, PAN Chunting², MING Hongliang²(

)

^1.Shanghai Nuclear Engineering Research and Design Institute Co., Ltd., Shanghai 200233, China
^2.Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016, China

引用本文:

鲍一晨, 石秀强, 孟凡江, 潘春婷, 明洪亮. 重水堆压力管延迟氢化物开裂应力强度因子门槛值测试方法研究[J]. 中国腐蚀与防护学报, 2025, 45(2): 381-387.
Yichen BAO, Xiuqiang SHI, Fanjiang MENG, Chunting PAN, Hongliang MING. Assessment Method of Threshold Stress Intensity Factor of Delayed Hydride Cracking for Pressure Tube in Heavy Water Reactor[J]. Journal of Chinese Society for Corrosion and protection, 2025, 45(2): 381-387.

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摘要:

在重水堆高温、高压、高辐照运行工况下，压力管材料性能会逐渐发生老化劣化，尤其当锆合金吸收冷却剂中的氘/氢后，其易发生延迟氢化物开裂(DHC)从而威胁压力管的边界完整性。根据加拿大标准CSA N285.8的要求，需要对DHC的应力强度因子门槛值K_IH进行评估。针对这一评估需求，对压力管K_IH的测试方法进行了研究。使用紧凑拉伸试样进行K_IH的测定，测定前使用电化学方法对试样预充氢约180 mg/kg，并分别在250、180、150和120 ℃下测定了其K_IH值。测试结果表明，在150~250 ℃之间使用降K法能够较准确地测定压力管锆合金材料的K_IH值，且其对测试温度无明显的依赖性。

关键词 ：压力管, 延迟氢化物开裂, 应力强度因子门槛值, 重水堆, 测试方法

Abstract：

The material properties of pressure tube will gradually deteriorate under high temperature, high pressure and high irradiation operating conditions of heavy water reactor (HWR), especially when the Zr-alloy absorbs deuterium/hydrogen from the coolant, it will become susceptive to the delayed hydride cracking (DHC), thus threatening the boundary integrity of the pressure tube. According to the Canadian Standard CSA N285.8, the threshold stress intensity factor (K_IH) for DHC needs to be evaluated. In response to this demand, the K_IH measurement method of pressure tube materials was studied. The K_IH was determined using compact tensile specimens, which were pre-charged with hydrogen by electrochemical method for about 180 mg/kg before tensile tests, and their K_IH values were measured at 250, 180, 150 and 120 ^oC respectively. The test results showed that the K_IH value of pressure tube Zr-2.5NbZr-alloy can be determined more accurately by using the K-reduction method at temperatures between 150 and 250 ^oC, and the measured values have no obvious dependence on the test temperatures.

Key words： pressure tube delayed hydride cracking threshold stress intensity factor heavy water reactor testing method

收稿日期: 2024-07-24 32134.14.1005.4537.2024.221

ZTFLH:

TG172.1

基金资助:国家重点研发计划(2019YFB1900902)

通讯作者: 明洪亮，E-mail：hlming12s@imr.ac.cn，研究方向为核电厂材料腐蚀与水化学

Corresponding author: MING Hongliang, E-mail: hlming12s@imr.ac.cn

作者简介: 鲍一晨，男，1986年生，硕士，高级工程师

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https://www.jcscp.org/CN/10.11902/1005.4537.2024.221 或 https://www.jcscp.org/CN/Y2025/V45/I2/381

图1 Zr-2.5Nb合金CCT试样取样位置示意图和试样尺寸图[13]

图2 轴向-径向和切向-径向的截面微观组织形貌

图3 充氢时间与试样中氢浓度关系

图4 充氢试样中的氢化物分布

表1 不同温度下有效电位增量与实际裂纹长度统计

图5 KIH测试温度曲线

图6 断裂区域面积计算示意图

图7 不同温度下的加载曲线

图8 不同温度下所测KIH值

图9 不同温度下的试样断口形貌

图10 250 ℃下高倍断口形貌

图11 升K和降K模式下所得KIH值对比[8]

图12 不同温度下断口条纹间距变化[13]

图13 KIH测值与超出TSSD氢浓度的关系[8]

1	Sagat S, Coleman C E, Griffiths M, et al. The effect of fluence and irradiation temperature on delayed hydride cracking in Zr-2.5Nb [A]. GardeA M, BradleyE R. Zirconium in the Nuclear Industry: Tenth International Symposium [M]. West Conshocken: ASTM, 1994: 35
2	Bao Y C, Shi X Q, Zhao C L. Hydrogen corrosion-uptake analysis and modeling for heavy water reactor Zr-2.5Nb pressure tube [J]. Corros. Prot., 2020, 41(11): 22
2	鲍一晨, 石秀强, 赵传礼. 重水堆Zr-2.5Nb压力管腐蚀吸氢分析与建模 [J]. 腐蚀与防护, 2020, 41(11): 22
3	Shi S Q, Puls M P. Criteria for fracture initiation at hydrides in zirconium alloys I. Sharp crack tip [J]. J. Nucl. Mater., 1994, 208: 232
4	Simpson L A, Puls M P. The effects of stress, temperature and hydrogen content on hydride-induced crack growth in Zr-2.5 Pct Nb [J]. Metall. Trans., 1979, 10A: 1093
5	IAEA. Delayed hydride cracking in zirconium alloys in pressure tube nuclear reactors [R]. Vienna: IAEA, 2004
6	Shek G K, Metzger D R. Effect of hydrogen concentration on the threshold stress intensity factor for delayed hydride cracking in Zr-2.5Nb pressure tubes [A]. Proceedings of ASME 2011 Pressure Vessels and Piping Conference [C]. Baltimore, 2011: 1287
7	Kim Y S, Ahn S B, Kim K S, et al. Temperature dependence of threshold stress intensity factor, K_IH in Zr-2.5Nb alloy and its effect on temperature limit for delayed hydride cracking [J]. Key Eng. Mater., 2006, 326-328: 919
8	Kim Y S, Park S S, Kwun S I. Threshold stress intensity factor, K_IH for delayed hydride cracking of a Zr-2.5Nb tube with loading mode [J]. J. Alloy. Compd., 2008, 462: 367
9	Sun C, Tan J, Ying S H, et al. Threshold stress intensity factor for delayed hydride cracking of a recrystallized N18 alloy plate along the rolling direction [J]. J. Nucl. Mater., 2010, 406: 212
10	Shek G K, Jovanoviċ M T, Seahra H, et al. Hydride morphology and striation formation during delayed hydride cracking in Zr-2.5%Nb [J]. J. Nucl. Mater., 1996, 231: 221
11	Yan D, Eadie R. The critical length of the hydride cluster in delayed hydride cracking of Zr-2.5wt%Nb [J]. J. Mater. Sci., 2000, 35: 5667
12	Kim Y S, Cheong Y M. Anisotropic delayed hydride cracking velocity of CANDU Zr-2.5Nb pressure tubes [J]. J. Nucl. Mater., 2008, 373: 179
13	Pan C T, Zhao G N, Bao Y C, et al. Effect of temperature on the delayed hydride cracking rate of Zr-2.5Nb alloy pressure tubes [J]. J. Nucl. Mater., 2023, 588: 154778
14	Puls M P. Effects of crack tip stress states and hydride-matrix interaction stresses on delayed hydride cracking [J]. Metall. Trans., 1990, 21A: 2905
15	Shmakov A A, Singh R N, Yan D, et al. A combined SIF and temperature model of delayed hydride cracking in zirconium materials [J]. Comput. Mater. Sci., 2007, 39: 237
16	Kim Y S, Matvienko Y G, Cheong Y M, et al. A model of the threshold stress intensity factor, K_IH, for delayed hydride cracking of Zr-2.5Nb alloy [J]. J. Nucl. Mater., 2000, 278: 251
17	Kim Y S, Kwon S C, Kim S S. Crack growth pattern and threshold stress intensity factor, K_IH, of Zr-2.5Nb alloy with the notch direction [J]. J. Nucl. Mater., 2000, 280: 304
18	Kim S S. The texture dependence of K_IH in Zr-2.5%Nb pressure tube materials [J]. J. Nucl. Mater., 2006, 349: 83
19	CSA. CSA N285.8:21 Technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactor [S]. Canadian Standards Association, 2021
20	Shi S Q, Puls M P. Dependence of the threshold stress intensity factor on hydrogen concentration during delayed hydride cracking in zirconium alloys [J]. J. Nucl. Mater., 1995, 218: 30

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