水冷裂变和聚变反应堆中辐射溶解问题的评述:第二部分,运行反应堆腐蚀损伤的预测

D. Macdonald, G. Engelhardt
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We predict that a hydrogen concentration of 80 ppb is sufficient to reduce the ECP in the OPFA to a level (−0.324 Vshe) that is sufficient to suppress the crack growth rate (CGR) below the practical, maximum level of 10−9 cm/s (0.315 mm/a) at which SCC is considered not to be a problem in a coolant circuit but, in the PFA, the ECP is predicted to be 0.380 Vshe, which gives a calculated standard CGR of 2.7 × 10−6 cm/s. This is more than three orders in magnitude greater that the desired maximum value of 10−9 cm/s. 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引用次数: 1

摘要

水的辐射分解是水冷裂变核动力反应堆(BWRs、PWRs和candu)主热传输系统(PHTSs)腐蚀损伤的一个重要原因,并且预计将成为未来聚变反应堆(例如目前正在开发的ITER)腐蚀损伤演变的一个重要因素。在这个由两部分组成的系列的第一部分中,我们回顾了提出的水的辐射分解机制,并证明了辐射分解导致无数氧化和还原物种的形成。在第二部分中,我们回顾了放射性溶解物质在反应器phts中电化学腐蚀电位(ECP)的建立和晶间应力腐蚀开裂(IGSCC)腐蚀损伤的发展中所起的作用。我们证明,放射性氧化产物,如O2, H2O2, HO2−和OH,当摩尔过量于还原性物质(H2, H和O22−)时,其中一些(H2)在沸水堆PHTS中优先从冷却剂中剥离,例如,使许多沸水堆中的冷却剂氧化。因此,敏化奥氏体不锈钢(例如304 SS型)中IGSCC的ECP向正方向移动,其值比临界电位(Ecrit = - 0.23 Vshe,在288°C)更正。在过去的50年里,这导致了许多沸水堆运行中的IGSCC事故,给电厂运营商和电力消费者带来了巨大的成本。在压水堆的情况下,初级回路用氢气加压,使氢气浓度为10至50 cm3/kgH2O(0.89至4.46 ppm),这样就不会发生持续的沸腾,氢气抑制水的辐射分解,从而抑制水的氧化性辐射分解产物的形成。因此,ECP是由氢电极反应(HER)主导的,尽管可能会发生与HER平衡电位的重大偏差,特别是在低[H2]时。在任何情况下,ECP位移到大约- 0.85 Vshe,这低于敏化不锈钢的IGSCC临界电位,但也比铣削退火合金600的氢致开裂(HIC)的临界电位更负。这导致了pwr中的蒸汽发生器管和其他部件(例如控制棒驱动管、稳压器部件)的广泛开裂,这也给运营商和电力用户带来了高昂的成本。虽然ITER尚未运行,但拟议的冷却剂化学方案将其放置在接近正常水化学(NWC)不沸腾的沸水堆,或者如果向IBED-PHTS添加氢,接近氢水化学(HWC)沸水堆。在目前的ITER技术中,ibedphts中的H2浓度被指定为80 ppb,这是在等离子体通量区(PFA)和等离子体通量区(OPFA)中都将经历的浓度。这相当于0.90 cc(STP) H2/KgH2O,相比之下,压水堆主冷却剂回路中使用的H2/KgH2O为20-50 cc(STP) H2/KgH2O,沸水堆氢水化学(HWC)中使用的H2/KgH2O为5.5至22 cc(STP)。我们预测,80 ppb的氢浓度足以将OPFA中的ECP降低到−0.324 Vshe,足以抑制裂纹扩展速率(CGR)低于实际最大值10−9 cm/s (0.315 mm/a),此时SCC在冷却剂回路中被认为不是问题,但在PFA中,ECP预计为0.380 Vshe,计算出的标准CGR为2.7 × 10−6 cm/s。这比期望的最大值(10 - 9 cm/s)高出三个数量级。我们建议重新讨论ITER的HWC问题,根据裂变反应堆技术的经验,制定一项协议,有效地将PFA中的ECP和CGR抑制到允许IBED-PHTS运行的水平。
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A Critical Review of Radiolysis Issues in Water-Cooled Fission and Fusion Reactors: Part II, Prediction of Corrosion Damage in Operating Reactors
The radiolysis of water is a significant cause of corrosion damage in the primary heat transport systems (PHTSs) of water-cooled, fission nuclear power reactors (BWRs, PWRs, and CANDUs) and is projected to be a significant factor in the evolution of corrosion damage in future fusion reactors (e.g., the ITER that is currently under development). In Part I of this two-part series, we reviewed the proposed mechanisms for the radiolysis of water and demonstrate that radiolysis leads to the formation of a myriad of oxidizing and reducing species. In this Part II, we review the role that the radiolysis species play in establishing the electrochemical corrosion potential (ECP) and the development of corrosion damage due to intergranular stress corrosion cracking (IGSCC) in reactor PHTSs. We demonstrate, that the radiolytic oxidizing radiolysis products, such as O2, H2O2, HO2−, and OH, when in molar excess over reducing species (H2, H, and O22−), some of which (H2) are preferentially stripped from the coolant upon boiling in a BWR PHTS, for example, renders the coolant in many BWRs oxidizing, thereby shifting the ECP in the positive direction to a value that is more positive than the critical potential (Ecrit = −0.23 Vshe at 288 °C) for IGSCC in sensitized austenitic stainless steel (e.g., Type 304 SS). This has led to many IGSCC incidents in operating BWRs over the past five decades that has exacted a great cost on the plant operators and electricity consumers, alike. In the case of PWRs, the primary circuits are pressurized with hydrogen to give a hydrogen concentration of 10 to 50 cm3/kgH2O (0.89 to 4.46 ppm), such that no sustained boiling occurs, and the hydrogen suppresses the radiolysis of water, thereby inhibiting the formation of oxidizing radiolysis products from water. Thus, the ECP is dominated by the hydrogen electrode reaction (HER), although important deviations from the HER equilibrium potential may occur, particularly at low [H2]. In any event, the ECP is displaced to approximately −0.85 Vshe, which is below the critical potential for IGSCC in sensitized stainless steels but is also more negative than the critical potential for the hydrogen-induced cracking (HIC) of mill-annealed Alloy 600. This has led to extensive cracking of steam generator tubing and other components (e.g., control rod drive tubes, pressurizer components) in PWRs that has also exacted a high cost on operators and power consumers. Although the ITER has yet to operate, the proposed chemistry protocol for the coolant places it close to a BWR operating on Normal Water Chemistry (NWC) without boiling or, if hydrogen is added to the IBED-PHTS, close to a BWR on Hydrogen Water Chemistry (HWC). In the current ITER technology, the concentration of H2 in the IBED-PHTS is specified to be 80 ppb, which is the concentration that will be experienced in both the Plasma Flux Area (PFA) and in the Out of Plasma Flux Area (OPFA). That corresponds to 0.90 cc(STP) H2/KgH2O, compared with 20–50 cc(STP) H2/KgH2O employed in a PWR primary coolant circuit and 5.5 to 22 cc(STP) H2/KgH2O in a BWR on hydrogen water chemistry (HWC). We predict that a hydrogen concentration of 80 ppb is sufficient to reduce the ECP in the OPFA to a level (−0.324 Vshe) that is sufficient to suppress the crack growth rate (CGR) below the practical, maximum level of 10−9 cm/s (0.315 mm/a) at which SCC is considered not to be a problem in a coolant circuit but, in the PFA, the ECP is predicted to be 0.380 Vshe, which gives a calculated standard CGR of 2.7 × 10−6 cm/s. This is more than three orders in magnitude greater that the desired maximum value of 10−9 cm/s. We recommend that the HWC issue in ITER be revisited to develop a protocol that is effective in suppressing both the ECP and the CGR in the PFA to levels that permit the operation of the IBED-PHTS in accordance with the experience gained in fission reactor technology.
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