Hydrogen and Hydrazine Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors, (VI) The Effect of Ammonia on Intergranular Stress Corrosion Cracking

Hydrogen and hydrazine co-injection into a boiling water reactor was considered as a new mitigation method of stress corrosion cracking (SCC). In this method, some amount of ammonia will be formed by the decomposition of hydrazine. The effect of ammonia on SCC susceptibility was studied over a wide range of electrochemical corrosion potentials (ECPs) in 288_C water by conducting slow strain rate technique SCC experiments (SSRTs). ECP was changed from _0:6V versus the standard hydrogen electrode (V(SHE)) to 0.1 V(SHE) by controlling dissolved oxygen concentration. Ammonia concentration was controlled to have values of 100 and 530 ppb. Similarly, sulfuric acid was injected to confirm the difference in the effect of injected chemical compounds on SCC susceptibility. The intergranular stress corrosion cracking (IGSCC) fraction, which was used as the index of SCC susceptibility, decreased with decreasing ECP for the case of no chemical injection. Sulfuric acid enhanced the IGSCC fraction. These data were in good agreement with literature data. On the other hand, ammonia at less than 530 ppb did not affect IGSCC fraction. It is expected that 51–280 ppb hydrazine and 0–53 ppb hydrogen will be injected into reactor water to mitigate SCC in BWRs. In the bottom region of the reactor pressure vessel, ECP and ammonia concentration will be _0:1 V(SHE) and 15–60 ppb, respectively. Under these conditions, ammonia did not affect SCC susceptibility. So SCC susceptibility will be mitigated by decreasing the ECP using hydrazine and hydrogen co-injection.     

Environmental controls for higher temperature direct-cycle light water reactors

Radiolysis modeling is used to estimate the minimum hydrogen concentration to activate platinum catalysts and reduce the electrochemical corrosion potential in light water reactors. Platinum catalysts are used in boiling water reactors to catalyze hydrogen and oxygen recombination, which reduces the corrosion potential and the susceptibility of austenitic structural materials to intergranular stress corrosion cracking. Two environmental challenges for material performance in higher temperature light water reactors are the increased susceptibility of austenitic materials to stress corrosion cracking and the higher production rate of oxidizing radiolytic species. For a reference supercritical water reactor, a hydrogen addition rate of 2 standard cubic feet per minute is needed to significantly reduce the susceptibility of austenitic materials to stress corrosion cracking. Also, for a reference higher temperature boiling water reactor, a hydrogen addition rate of 10 standard cubic feet per minute of hydrogen reduces the stress corrosion crack susceptibility of austenitic materials located in the lower portion of the reactor vessel.     

Predicted impact of power uprate on the water chemistry of Kuosheng boiling water reactor

A theoretical model was adapted to evaluate the impact of power uprate on the water chemistry of a commercial boiling water reactor (BWR) in this work. In principle, the power density of a nuclear reactor upon a power uprate would change immediately, followed by water chemistry variations due to enhanced radiolysis of water in the core and near-core regions. It is currently a common practice for commercial BWRs to adopt hydrogen water chemistry (HWC) for corrosion mitigation. The optimal feedwater hydrogen concentration may be different after a power uprate is implemented in a BWR. A computer code DEMACE was used in the current study to investigate the impact of various power uprate levels on major radiolytic species concentrations and electrochemical corrosion potential (ECP) behavior of components in the primary coolant circuit of a domestic BWR-6 type reactor operating under either normal water chemistry or HWC. Our analyses indicated that under a constant core flow rate the chemical species concentrations and the ECP did not vary monotonously with increases in reactor power level at a fixed feedwater hydrogen concentration. In particular, the upper plenum and the upper downcomer regions exhibited uniquely higher ECPs at 108% and 115% power levels than at the other evaluated power levels.     

Analysis of water radiolysis in relation to stress corrosion cracking of stainless steel at high temperatures - Effect of water radiolysis on limiting current densities of anodic and cathodic reactions under irradiation

Electrochemical corrosion potential (ECP) is an important measure for environmental factor in relation to stress corrosion cracking (SCC) of metal materials. In the case of SCC for in-core materials in nuclear reactors, radiolysis of coolant water decisively controls ECP of metal materials under irradiation. In the previous models for ECP evaluation of stainless steel, radiolysis of reactor water in bulk was considered to calculate the bulk concentrations of the radiolysis products. In this work, the radiolysis not only in bulk but also in the diffusion layer at the interface between stainless steel and bulk water was taken into account in the evaluation of ECP. The calculation results shows that the radiolysis in the diffusion layer give significant effects on the limiting current densities of the redox reactions of the radiolysis products, H₂O₂ and H₂, depending on dose rate, flow rate and water chemistry, and leads to the significant increase in the ECP values in some cases, especially in hydrogen water chemistry conditions.     

压水堆核电站锆水反应微观机理

压水堆核电站严重事故下的氢气行为研究需建立氢气生成的动力学模型,氢气生成反应的微观机理和反应速率常数是建立动力学模型的基础。本工作采用量子化学理论,应用量子化学软件包Gaussian03,在HF/3-21G理论模型上研究了压水堆严重事故下锆水反应的微观机理,并计算了反应速率常数。计算结果表明,锆水反应是由4个基元反应组成的总包反应。第2步基元反应的正反应速率最小,是锆水反应的速控步。在微观上研究减少或控制氢气生成的措施应从第2步基元反应入手。文中的计算结果偏于保守,以该方法建立起的动力学模型模拟压水堆核电站严重事故下的氢气行为是安全的。     

Corrosion of ZrO2 treated type 304 stainless steels in high temperature pure water with various amounts of hydrogen peroxide

As boiling water reactors (BWRs) age, intergranular stress corrosion cracking (IGSCC) of the structural materials in the reactor piping systems and vessel internals has become a major degradation problem. Several approaches to mitigating IGSCC in the structural components have been developed and investigated. Among them, the technique of inhibitive protective coatings is deemed the most promising one since it is expected to work even in the absence of the well-known hydrogen water chemistry technology.Following our earlier work on exploring the electrochemical characteristics of important oxidizing species on zirconium oxide (ZrO₂) treated Type 304 stainless steels (SSs), we targeted on the characteristics of hydrogen peroxide, which is another strongly oxidizing species in the reactor coolant other than oxygen, in this study. Tests were conducted to determine electrochemical parameters such as electrochemical corrosion potential (ECP), corrosion current density, exchange current density and Tafel constant of the reduction reaction of hydrogen peroxide on 304 SS specimens before and after the ZrO₂ treatment. The surface morphologies of the treated and untreated specimens were examined by scanning electron microscopy, energy dispersive X-ray spectroscopy, and laser Raman spectra. Furthermore, the corrosion mitigation efficiency of ZrO₂ treatment was evaluated by electrochemical polarization tests in simulated BWR environments. Test results showed that there were no significant differences in ECP between the untreated and ZrO₂ treated specimens in the test environments of various hydrogen peroxide concentrations. However, it was found via polarization analysis that the exchange current density of the reduction reaction on and the corrosion current density of the treated specimens were markedly lower than those on and of the untreated ones in the same environments. The ZrO₂ treatment was able to deter the reduction rate of hydrogen peroxide on the Type 304 SS surface.     

Effect of Hydrogen Water Chemistry on Zircaloy-2 Fuel Cladding and Structure Material Performance

The effects of hydrogen addition to the feedwater on the corrosion and hydrogen uptake performance of Zircaloy-2 fuel cladding tubes, a water rod tube and spacer materials irradiated for four cycles in a BWR were evaluated. The uniform oxide behaviors of the cladding tubes, water rod and spacer materials were not affected by hydrogen water chemistry (HWC) condition. The hydrogen uptake and pickup fractions of the water rod and spacer materials were similar to those of water rods and spacer materials under normal water chemistry (NWC) conditions. As for the fuel rods, in spite of comparably heavy crud deposition, their hydrogen uptake and pickup fractions were clearly lower than the values under NWC conditions. Overall, the results indicated that HWC had no adverse effects on fuel performance.     

BWR pipe crack control using hydrogen water chemistry: Status report on Dresden-2 program

One of the proposed remedies for intergranular stress corrosion cracking of stainless steel piping in BWRs is an alternative water chemistry called hydrogen water chemistry (H₂WC) that involves suppression of reactor water dissolved oxygen to ≤ 20 ppb via hydrogen injection to the feedwater in conjunction with control of conductivity to ≤ 0.3 μ mho/cm. A long-term verification program, over two or three 18 month fuel cycles, was started at Commonwealth Edison's Dresden-2 reactor in April 1983 (Cycle 9). This paper describes the results of the water chemistry changes, structural material and fuel evaluations, and plant radiation level changes during Cycle 9, which ended in October 1984.To date the results of the verification program are very encouraging. They indicate that the alternative water chemistry, based on hydrogen additions to the feedwater to suppress oxygen and low conductivity, can be maintained in a large operating BWR, and that it does mitigate IGSCC in stainless steel recirculation piping. Monitoring of fuel and plant materials will continue in Dresden-2 at least through Cycle 10 to confirm the absence of any unusual side effects of this remedy for IGSCC.     

Electrochemical Corrosion Potential Modeling in the Primary Heat Transport Circuit of the Chinshan Boiling Water Reactor under the Condition of Hydrogen Water Chemistry with Noble Metal Coating

The technique of noble metal treatment, such as noble metal coating (NMC) or noble metal chemical addition, accompanied by a low level hydrogen water chemistry, is being employed by a number of nuclear power plants around the world for mitigating intergranular stress corrosion cracking in the vessel internals of their boiling water reactors (BWRs). A computer model DEM ACE was expanded and employed to assess the effectiveness of NMC throughout the primary heat transport circuit (PHTC) of a BWR. The effectiveness of NMC was justified by the electrochemical corrosion potential (ECP) and crack growth rate (CGR) predictions. In calculating the ECP, enhancing factors for the exchange current densities of redox reactions available from recently published data, were employed. The Chinshan BWR was selected as a model reactor. According to the modeling results, it was found that the effectiveness of NMC in the PHTC of a BWR could vary from region to region at different feedwater hydrogen concentrations. For the selected BWR, NMC was predicted to be of little benefit when the feedwater hydrogen concentration reached 0.9 ppm or over. In particular, the NMC technique proved to be beneficial in reducing ECP and CGR along the PHTC even if the BWR was operated under normal water chemistry.     

Achievements and further plans for the OECD Halden reactor project materials programme

The materials programme at Halden, in addition to cladding corrosion studies, is aimed also at addressing the effects of operating conditions and water chemistry variables on core materials behaviour, particularly as related to reactor pressure vessel integrity and irradiation assisted stress corrosion cracking (IASCC), the materials degradation phenomenon which affects the structural integrity of stainless steel and nickel based components. The aim of the experimental work is to improve the understanding of materials ageing processes, to demonstrate the benefits of mitigation measures and to evaluate properties of materials, which have been subjected to long in-reactor service. While a number of the studies are performed in loops, which simulate light water reactor environments in terms of thermal-hydraulic, radiation and water chemistry conditions, dry irradiation facilities are also utilised, particularly in relation to studies aimed at determining the effects of fluence on material integrity.     

Water Chemistry Experience of Nuclear Power Plants in Japan

Japanese LWRs have experienced several troubles caused by corrosions of structural materials in the past ca. 20 years of their operational history, among which are increase in the occupational radiation exposures, intergranular stress corrosion cracking (IGSCC) of stainless steel piping in BWR, and steam generator corrosion problems in PWR. These problems arised partly from the improper operation of water chemistry control of reactor coolant systems. Consequently, it has been realized that water chemistry control is one of the most important factors to attain high availability and reliability of LWR, and extensive researches and developments have been conducted in Japan to achieve the optimum water chemistry control, which include the basic laboratory experiments, analyses of plant operational data, loop tests in operating plants and computer code developments. As a result of the continuing efforts, the Japanese LWR plants have currently attained a very high performance in their operation with high availability and low occupational radiation exposures. A brief review is given here on the R & D of water chemistry in Japan     

Predicted impact of core flow rate on the corrosion mitigation effectiveness of hydrogen water chemistry for Kuosheng boiling water reactor

It is currently a common practice that a boiling water reactor (BWR) adopts hydrogen water chemistry (HWC) for mitigating corrosion in structural components in its primary coolant circuit. When the core flow rate (CFR) in a BWR is changed, the coolant residence time in the primary coolant circuit would be different. The concentrations of major redox species (i.e. hydrogen, oxygen, and hydrogen peroxide) in the coolant may accordingly vary due to different durations of radiolysis in the core and other near-core regions. A theoretical model by the name of DEMACE was used in the current study to investigate the impact of various CFRs (from 100% to 80.0%) on the effectiveness of HWC in a domestic BWR. Our analyses indicated that the HWC effectiveness at some locations could be downgraded due to a decrease in CFR. However, a lower CFR was instead beneficial to the corrosion mitigation efficiency of HWC at other locations. The impact of CFR on the HWC effectiveness could vary from location to location in a BWR and eventually from plant to plant.     

In situ measurement of corrosion of type 316L stainless steel in 553 K pure water via the electrical resistance of a thin wire

A system for the in situ monitoring of corrosion depth via electrical resistance measurements was applied to study the corrosion rate of type 316L stainless steel at 553 K in pure water. Corrosion depth was measured using a 50 μm diameter wire probe mounted axially in the tube. Measurements were in good agreement with literature data for both the hydrogen water chemistry (HWC) condition and the normal water chemistry (NWC) condition. Oxide film analyses by scanning electron microscopy and laser Raman spectroscopy on the wire probe and the tube showed no effects from shape of the test specimens or the application of electric current. Corrosion kinetics was evaluated by fitting equations to the measurements. Data for the HWC condition could be fitted by a two-step logarithmic–parabolic law. A single-step logarithmic law fitted data for the NWC condition. Changes in corrosion rate by the water chemistry changes were readily detected with the technique. Corrosion depth change could be observed for the water chemistry change from the NWC condition to the HWC condition with electrochemical corrosion potential (ECP) of ?0.56 V vs. standard hydrogen electrode, which is lower than the ECP that the phase of iron oxide changes from α-Fe₂O₃ to Fe₃O₄.     

Effects of Organic Impurities on BWR Water Chemistry, (2)

In order to keep ⁶⁰Co radioactivity low without incurring serious disadvantages regarding integrity of fuel cladding and reliability of structural materials, it is desirable to relate cooling water (especially ⁶⁰Co behavior), structural materials and fuel cladding quantitatively. Organic impurities, such as fragments of cation resin, are fed into the reactor where they decompose, increasing conductivity of the reactor water and decreasing pH, which, in turn, increase ⁶⁰Co radioactivity in the reactor water and then, shutdown dose rate. The relationships between ⁶⁰Co radioactivity in the reactor water and pH have been analyzed by using water chemistry experience at operating BWR plants as well as laboratory data of resin decomposition in the water at elevated temperature, and analytical formulas to explain the shutdown dose rate increase caused by organic impurities intrusion were proposed. Effects of organic impurities on the shutdown dose rate can be moderated preventively by application of weak alkali control.     

Application of radiation chemistry research to the operation of nuclear reactors

The theme of this review is the application of radiation chemistry research to improve the operating efficiency of nuclear reactors. The intense radiation fields in reactor cores produce a hostile environment for incore materials; this report describes how recent research helped overcome the chemistry problems caused by the radiation.Examples discussed are the inhibition of graphite moderator corrosion and prevention of carbon deposition in gas-cooled reactors, suppression of radiolysis of the cooling water in concrete pressure vessels, hydrogen formation following a loss of coolant accident in a PWR and improving the stability of decontamination reagents for water reactors.     

Hydrogen water chemistry for BWRs: a status report on the EPRI development program

Many boiling water reactors (BWRs) have experienced extensive intergranular stress corrosion cracking (IGSCC) in their austenitic stainless steel reactor coolant system piping, resulting in serious adverse impacts on plant capacity factors, O&M costs, and personnel radiation exposures. A major research program to provide remedies for BWR pipe cracking was co-funded by EPRI, GE, and the BWR Owners Group for IGSCC Research between 1979 and 1988. Results from this program show that the likelihood of IGSCC depends on reactor water chemistry (particularly on the concentrations of ionic impurities and oxidizing radiolysis products) as well as on material condition and the level of tensile stress. Tests have demonstrated that the concentration of oxidizing radiolysis products in the recirculating reactor water of a BWR can be reduced substantially by injecting hydrogen into the feedwater. Recent plant data show that the use of hydrogen injection can reduce the rate of IGSCC to insignificant levels if the concentration of ionic impurities in the reactor water is kept sufficiently low. This approach to the control of BWR pipe cracking is called hydrogen water chemistry (HWC). This paper presents a review of the results of EPRI's HWC development program from 1980 to the present. In addition, plans for additional work to investigate the feasibility of adapting HWC to protect the BWR vessel and major internal components from potential stress corrosion cracking problems are summarized.     

Effects of seawater components on radiolysis of water at elevated temperature and subsequent integrity of fuel materials

Effects of seawater components on radiolysis of water at elevated temperature have been studied with a radiolysis model and a corrosion test under gamma-ray irradiation conditions to evaluate the subsequent influence on integrity of fuel materials used in an advanced boiling water reactor. In 2011, seawater flowed into the nuclear power plant system of the Hamaoka Nuclear Power Station Reactor No. 5 during the plant shutdown operation. The reactor water temperature was 250 °C and its maximum Cl^? concentration was ca. 450 ppm when seawater was mixed with reactor water. The radiolysis model predicted that the main radiolytic species were hydrogen, oxygen and hydrogen peroxide. Concentrations of radiolytic products originating from Cl^? and other seawater components were found to be rather low. The dominant product among them was ClO₃^? and its concentration was found to be below 0.01 ppm for a 10⁵ s irradiation period. No significant corrosion of zircaloy-2 and 316L stainless steel was found in the corrosion test. These results led to the conclusion that the harmful influence of radiolytic products originating from seawater components on integrity of fuel materials must be smaller than that of Cl^? which is the main ionic species in seawater.     

某三代压水堆机组停堆水化学控制工艺改进与应用

有效降低压水堆机组反应堆冷却剂系统（RCP）材料腐蚀速率的同时有效去除活化腐蚀产物,可降低堆芯外辐射场、减少工作人员受照剂量,从而确保核电机组大修工作的顺利展开。某三代PWR机组采用富集硼酸（EBA）进行反应性控制的同时,利用其在功率运行期间对RCP系统冷却剂实施水化学控制的显著优势,同时在机组首次大修期间对停堆水化学控制工艺采取的改进措施（包括碱性环境向酸性环境转换、还原环境向氧化环境转换、强制氧化期间多次向一回路添加双氧水维持氧化性、化学和容积系统混床最大流量净化等）,在机组停堆下行阶段实现了降低机组辐射剂量并减少工作人员受照剂量的目的。      

CAP1400燃料组件用新锆合金研究

在Zr-Sn-Nb系合金的基础上添加微量合金元素Ge和Si等,采用真空电弧熔炼,制备了多种新锆合金。使用透射电子显微镜(Transmission electron microscope,TEM)对合金基体进行显微组织分析,分别通过堆外高压釜腐蚀试验、定氢分析仪和万能材料试验机对合金的腐蚀、吸氢和拉伸性能进行评估。结果表明,常规工艺处理后,SZA-4和SZA-6合金均发生了完全再结晶,第二相细小、均匀弥散分布在晶粒内和晶界上;SZA-4和SZA-6合金在三种水化学条件下均具有优良的耐腐蚀性能,SZA-6合金的耐腐蚀性能优于参考合金,SZA-4合金的耐腐蚀性能略优于SZA-6合金;SZA-6合金的吸氢性能略优于SZA-4合金;两种合金的拉伸性能满足设计要求。基于SZA-4和SZA-6合金优良的耐腐蚀、吸氢和力学性能,未来将有望用于CAP1400自主化燃料组件。     

压水堆核电站铁水反应机理研究

从微观上研究压水堆核电站严重事故下减少或控制氢气生成的措施需研究氢气生成的微观机理。本工作采用量子化学理论,应用量子化学软件包Gaussian03,在B3LYP/6-311+G(d)理论模型上研究了压水堆严重事故下铁水反应的微观机理,并计算了活化能。结果表明,铁水反应是由两个基元反应组成的总包反应。第2步基元反应的正反应活化能较大,是铁水反应的速控步。在微观上研究减少或控制氢气生成的措施应从第2步基元反应入手。