Water Chemistry in a Crack Tip under Irradiation, (II)

Under neutron and gamma-ray irradiations, radiolytic species are generated directly in the crack tip, which causes higher oxidant concentrations and subsequently influences crack propagation rate.

A crevice radiolysis model was proposed to estimate the oxidant concentrations in the crack tip water under gamma-ray irradiation. Direct generation of radiolytic species in the crevice water, and their secondary generation and disappearance caused by their interaction with the crevice surface as well as species in the crevice water were included in the model. The diffusion of the radiolytic species through the narrow gap from the bulk water to the crack tip and vice versa were also considered.

Calculation results confirmed that the concentrations of H₂O₂, one of the most important oxidants in BWR environments, in both bulk water and crack tip water under irradiation (energy deposition rate: 0.1 W/cm) were high enough to show high local ECP in both regions under NWC, but were high in the bulk water and low in the crack tip water under HWC. A high H₂ diffusion rate from the bulk to the crack tip enhanced the recombination reaction of H₂O₂ and H₂.     

International Conference on Water Chemistry in Nuclear Power Plants

Minor actinide (MA) transmutation performances were evaluated for BWR cores with mixed oxide (MOX) fuels of various hydrogen-to-heavy metal atom number ratios (H/HM) to decrease the work for high-level radioactive wastes (HLW) management. One effective approach to increase the MA transmutation ratio is a multi-recycle core of h4OX fuel with MA. The decrease of the MA-to-fissile Plutonium (Puf) amount ratio makes it possible to recycle MA without additional plutonium (Pu) obtained from reprocessing of LWR fuel. For this purpose, a T-ratio, which is the MA-to-Puf amount ratio change during burnup, of less than unity is required, and H/HM must be less than 3 when MA enrichment is more than 2 wt%. For multi-recycle cores with the H/HM of 3 and the initial MA enrichment of 5 wt%, more than 50% of MA load initially can be transmuted by the third recycle.     

Optimization of Dissolved Hydrogen Concentration for Control of Primary Coolant Radiolysis in Pressurized Water Reactors

Optimization of the dissolved hydrogen concentration in primary coolant of pressurized water reactors has several potential advantages in material integrity and dose reduction. To assess the threshold value of dissolved hydrogen for reducing condition, in-pile loop experiments and radiolysis model calculations were performed. Both experiments and model calculations indicate that the threshold value of dissolved hydrogen for radiolysis is much less than the present control level in pressurized water reactors, and the in-core region is more easily affected by reduced dissolved hydrogen level than the out-of-core region. The measurement of electrochemical corrosion potential of structural material, while varying dissolved hydrogen levels, is desirable in the future to compensate for the uncertainty of the model calculation. For this purpose, electrochemical corrosion potential of stainless steel at the top-of-core, in which the reference electrode could be inserted, was estimated based on the correlation between the experiments and the modeling for the loop.     

Some Aspects of LWR's Water Chemistry

The hypothesis laid down in this paper offers an alternative to the current interpretation of the processes: hideout (hideout-return), crud deposition and change of the coolants activity level in the nuclear power reactors under different operating conditions.

This alternative is based on the supposition that the heat flux has not a direct effect on the processes mentioned above, but acts through the heat transfer mechanism in the boundary, caused by itself.

The boundary influenced by heat flow is in non-equilibrium state and in such system states (at adequate heat flux) non-equilibrium structures called dissipative structures arise which ale closely connected with heat transfer mechanism. The transport and the location of the colloidal corrosion products dispersed in steam generators-or reactor water during the units operation are strongly influenced by the existence of dissipative structures. The transport and location of the main part of ion species depend also on the existence of these structures because the colloidal particles act like collectors of the ions dissolved in the water (The ions are inserted in the colloidal particles double layer).

The hideout and hideout-return phenomena are interpreted as closely connected with the existence of the above mentioned dissipative structures. It was attempted to consider the changes in nuclide concentrations in the LWR coolant upon start-up and shutdown as hideout respectively hideout-return processes. The recent shutdown chemistry aspects are discussed also.     

Hydrazine and Hydrogen Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors (IV)

A calculation model has been developed in order to evaluate effectiveness of hydrazine and hydrogen co-injection (HHC) into reactor water for mitigation of intergranular stress corrosion cracking of structural materials used in boiling water reactors (BWRs). The HHC uses the strong reducing power of hydrazine radical, which is produced in the downcomer region under irradiation by γ-rays and neutrons. Some reactions and their reaction rate constants were determined based on experiments which were carried out in aerated water, hydrogenated water, and deaerated water. The calculated results were in good agreement with experimental data by a factor of two. The model was applied to a BWR and it was found that the HHC cut oxygen and hydrogen peroxide amounts dissolved in reactor water more effectively than hydrogen water chemistry alone. Thus, the required amount of hydrogen for hydrazine injection was much lower than that for hydrogen water chemistry. Consequently, electrochemical corrosion potential of structural materials could be lowered below–0:1V vs. SHE without any increase of MS line dose rate, which has been a limitation of the conventional hydrogen water chemistry. The HHC was predicted to decrease crack growth rate of structural materials by a factor of 10.     

Structural Material Anomaly Detection System Using Water Chemistry Data

The concept of an advanced water chemistry diagnosis system for detection of anomalies and preventive maintenance of system components is proposed and put into a concrete form. Using the analogy to a medical inspection system, analyses of water chemistry change will make it possible to detect symptoms of anomalies in system components.

Then, correlations between water chemistry change and anomaly occurrence in the components of the BWR primary cooling system are analyzed theoretically. These fragmentary correlations are organized and reduced to an algorithm for the on-line diagnosis system using on-line monitoring data, pH and conductivity. By using actual plant data, the on-line diagnosis model system is verified to be applicable for early and automatic finding of the anomaly cause and for timely supply of much diagnostic information to plant operators.     

The Efficiency of Noble Metals in Reducing the Corrosion Potential in the Primary Coolant Circuits of Boiling Water Reactors Operating under Hydrogen Water Chemistry Operation

In order to promote the effectiveness of hydrogen water chemistry (HWC) and to achieve a more effective reduction in electrochemical corrosion potential (ECP) in the primary coolant circuits of boiling water reactors (BWRs), the technology of noble metal chemical addition (NMCA) was brought into practice about 10 years ago. NMCA aims at enhancing the oxidation of hydrogen on metal surfaces and lowering the concentrations of the oxidants (oxygen and hydrogen peroxide) via recombination with hydrogen on the catalyzed surfaces, and therefore reducing the corrosion potentials of the structural alloys in a BWR primary heat transport circuit. Previous research indicates that the effectiveness of NMCA in combination with a low HWC might be evaluated via model predictions of the hydrogen-to-oxidant molar ratio (M_H/O) in the primary coolant circuit. If the M_H/O at a certain location is calculated to be greater than 2, it is justified that the NMCA would be effective in reducing the ECP to much below the critical potential for Intergranular Stress Corrosion Cracking (IGSCC), E_IGSCC, of --0.23 V_SHE. However, this statement is true only when the recombination efficiency of hydrogen with oxygen and/or hydrogen peroxide at the location of interest is 100%. Otherwise, significant amounts of oxidants may still be present, even with a stoichiometric M_H/O of greater than 2. With the aid of a computer model DEMACE, we explored the impact of incomplete recombination and found that the ECP might be reduced under given circumstances, but not to a great extent, and might remain well above E_IGSCC. Accordingly, considerable caution should be exercised upon using the M_H/O as a sole indicator for evaluating the effectiveness of NMCA with low HWC as a means of mitigating IGSCC in a BWR. An important finding of this study is that it is necessary to quantify the recombination efficiencies of hydrogen with oxygen and/or hydrogen peroxide on the noble metal treated stainless steel surfaces in order to qualify the use of M_H/O as an indicator for NMCA effectiveness in the primary coolant circuit of a BWR.     

Effects of Hydrogen Peroxide on Intergranular Stress Corrosion Cracking of Stainless Steel in High Temperature Water, (III)

The stress corrosion cracking (SCC) of structural materials used in boiling water reactors has been studied at relatively low hydrogen peroxide (H₂O₂) concentrations, around lOppb, which was assumed to be representative of the corrosion environment formed in hydrogen water chemistry (HWC). The 1/4T compact tension specimen was used for measurement of crack growth rates (CGRs) of sensitized type 304 stainless steel in high temperature and high purity water. Crack length was monitored by a reversing direct current potential drop method. Since H2O2 is easily decomposed thermally, a polytetrafluoroethylene-lined autoclave was used to minimize its decomposition on the autoclave surface. The CGR in the H₂O₂ environment differed from that in the O₂ environment even though the electrochemical corrosion potential (ECP) for both conditions was the same. The data implied that the ECP could not be used as a common environmental deterministic parameter for SCC behavior at higher potentials for different oxidant conditions. The corrosion current density was found to play an important role as an environmental index for SCC, which was given as just the current density at the ECP at a specific oxidant concentration. The CGRs were found to be written as CGR = (3.8±0.6)xl0^-3 _icor +(l-5±1.6) x 10^-8mm/s using the calculated corrosion current density i_corbelow 10^-4 A-cm^-2.     

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.     

压水堆核电厂失水事故后安全壳内产氢量计算研究

采用ORIGEN2程序对压水堆核电厂失水事故工况下堆芯区和地坑区氢气的产生量进行计算,以合理减少安全壳内可燃气体的控制设计评价的保守性.通过冷却剂的辐照分解产氢以及其他相关计算模型,对600MW(电功率)级压水堆核电厂失水事故工况下的氢气产生量进行计算.计算结果表明原评价结果过于保守,在核电厂失水事故后仍有充分的时间准备投入安全壳内氢气复合器.     

Structural Material Anomaly Detection System Using Water Chemistry Data, (II)

A calculation model was developed to predict the dose rate, caused by ⁵⁸CO which is formed by the activation of ⁵⁸Ni, around the recirculation pipes m boiling water reactors (BWRs). The model is characterized by considering direct deposition of Ni ion on the nuclear fuel cladding surface and the geometrical contact probability for the ferrite formation reaction between deposited Ni(Co) and Fe₂O₃ on fuel cladding surface.

This model showed the important role of the amount of Fe crud on the surface to reduce ⁵⁸CO ion concentration in the reactor water. And the necessary Fe concentration in the feedwater for reducing the dose rate in the primary system was estimated as a function of the operating time. This model also enables the quantitative predictions of the effect of prefilming treatment of the feedwater heater tubes or another methods to reduce dose rate in an Fe crud suppressed BWR.     

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.     

Influence of Dissolved Hydrogen on Structure of Oxide Film on Alloy 600 Formed in Primary Water of Pressurized Water Reactors

In order to investigate the relationship between the susceptibility of primary water stress corrosion cracking (PWSCC) in Alloy 600 and the content of dissolved hydrogen (DH) in the primary water of pressurized water reactors (PWR), structural analysis of oxide films formed under four different DH conditions in simulated primary water of PWR was carried out using a grazing incidence X-ray diffractometer (GIXRD), a scanning electron microscope (SEM) and a transmission electron microscope (TEM). In particular, to perform accurate analysis of the thin oxide films, the synchrotron radiation of SPring-8 was used for GIXRD.

It has been observed that the oxide film is mainly composed of nickel oxide, under the condition without hydrogen. On the other hand, needle-like oxides are formed at 1.0 ppm of DH. In the environment of 2.75 ppm of DH, the oxide film has thin spinel structures. From these results and phase diagram consideration, the condition around 1.0 ppm of DH corresponds to the boundary between stable NiO and spinel oxides, and also to the peak range of PWSCC susceptibility. This suggests that the boundary between NiO and spinel oxides may affect the SCC susceptibility.     

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.     

Formation of platinum nanoparticle colloidal solution by gamma-ray irradiation

Platinum nanoparticle colloidal solution with a small amount of impurities was formed from a suspension of hexahydroxy platinic acid (SHHPA) by gamma-ray irradiation to suppress changes in water chemistry such as electrical conductivity and concentration of impurities in the reactor water during noble metal chemical addition in plant operation. The SHHPA was prepared from sodium hexahydroxyplatinate solution by using an H-type cation exchange resin. Optimum conditions for formation of the platinum nanoparticle colloidal solution were the following: absorbed dose of gamma-ray irradiation, >6 kGy; pH of solution, >8.2; air saturation; no methanol addition. Characteristics of the formed platinum nanoparticles were as follows: mean particle size, 2.3 ± 0.5 nm; particle charge, negative; isoelectric point at a pH of 3.5 ± 0.1; the chemical compound consisted mainly of platinum dioxide without platinum metal. No precipitation of platinum nanoparticles was observed after storage time of 1 year without any stirring in a room where the temperature varied from about 10 to about 35 °C.     

Suboptimal Control of Pressurized Water Reactor Power Plant Using Approximate Model-Following Method

We attempted to develop an effective control system that can successfully manage the nuclear steam supply (NSS) system of a PWR power plant in an operational mode requiring relatively small variations of power. A procedure is proposed for synthesizing control system that is a simple, yet practiced, suboptimal control system. The suboptimal control system is designed in two steps; application of the optimal control theory, based on the linear state-feedback control and the use of an approximate model-following method. This procedure can appreciably reduce the complexity of the structure of the controller by accepting a slight deviation from the optimality and by the use of the output-feedback control. This eliminates the engineering difficulty caused by an incompletely state-feedback that is sometimes encountered in practical applications of the optimal state-feedback control theory to complex large-scale dynamical systems. Digital simulations and graphical studies based on the Bode-diagram demonstrate the effectiveness of the suboptimal control, and the applicability of the proposed design method as well.     

Effects of Hydrogen Peroxide on Corrosion of Stainless Steel, (I)

In order to evaluate the effects of hydrogen peroxide (H₂O₂) on intergranular stress corrosion cracking, a high temperature high pressure water loop, which can control H₂O₂ concentration with minimal oxygen (O₂) co-existence, is required. This loop is characterized by

1. A once-through type loop to prevent accumulation of decomposed O₂ in the loop

2. Minimized autoclave volume to prevent bulk thermal decomposition of H₂O₂

3. A polytetrafluoroethylene (PTFE) lining to prevent surface decomposition of H₂O₂, and

4. A H₂O₂ monitoring system with an off-line H₂O₂ detector to determine concentration in the sampled water which is combined with an in-line dissolved O₂ detector to determine the decomposed O₂ concentration.

The authors developed such a loop previously. In the present work, performance tests were carried out and measured data were evaluated by comparing with predicted values to verify whether the target characteristics were met. The measured H₂O₂ remaining in the sampled water agreed with the predicted amount within 5%. It was confirmed that the ratio of H₂O₂ remaining in the loop autoclave was more than 90% and the concentration could be monitored continuously with the in-line dissolved O₂ detector installed after the cooler in the loop. Electrochemical corrosion potential (ECP) and frequency dependent complex impedance were measured successfully by changing H₂O₂ concentration.     

Japanese Contributions to Fields of Predicting and Combating Environmentally Assisted Cracking (EAC) of Steels in LWR Water

Two major currents of the EAC relevant activities in Japan are reviewed in some chronological manner. One is the studies on the acceleration of fatigue crack growth in LWR. The work was first pioneered early in Japan, and extensive collaborative works have followed interacting with the international activities, which have yielded a clear views on the potential issues of the RPV structural steels used in the domestic NPPs. Another is the development of materials and techniques to combat the stress corrosion cracking of the structural materials in LWR water environment. Extensive collaboration among the government agencies, utilities, industries and academic societies lead the issues to a dramatic solution in rather short period. Both of those two aspects are stressed to have generic significance in the present and future nuclear technology.     

Water Chemistry in a Crack Tip under Irradiation, (I)

In order to evaluate crack propagation rate for irradiation assisted stress corrosion cracking (IASCC) of stainless steel, accurate estimation of the concentrations of radiolytic species in the crack tip water is important. These concentrations are affected by energy deposition in the crack tip water. Concentrations were determined by measurements with thin thermo luminescence dosimeters (TLDs) and calculations based on a Monte Carlo method. The measured and the calculated results confirmed that the energy deposition rate in the crack tip water was larger than that in the bulk water. This was caused by higher back-scattered radiation from the surrounding steel. The energy absorption of the water in the crevice with a 1 μm width was 1.3 times as high as that in the bulk water.     

Chemistry of Radioactive Nitrogen in BWR Primary System

Numerical model calculations and evaluation of plant experience with chemical behaviors of radioactive nitrogen (13N and 16N) in BWR primary systems have suggested the followings :

(1) Calculated results showed that the major species of radioactive nitrogen released to the turbine system was nitrogen mono-oxide.

(2) Based on ¹³N measurement in the main steam of a BWR, 85% of ¹³N was found to be anion. This result supported the prediction above.

(3) Calculated results showed that chemical behaviors of ¹³N and ¹⁶N were not the same because of the difference in half-lives of ¹³N and ¹⁶N; 13N was easily accumulated in the core by-pass channels where the residence time of water was several times longer than in fuel channels, but the main source of ¹⁶N was the boiling channels in the reactor core.

(4) Nitrogen-16 released to the main steam line could be decreased by lowering the hydrogen atom concentration in the water; typically to around 1/10 of the normal level by the addition of nitrous acid of ~1×10~₆M at the reactor core inlet.