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In order to confirm applicability and accuracy of FAC evaluation methods based on the coupled FAC model of static electrochemical analysis and dynamic oxide layer growth analysis, wall thinning rates calculated with the proposed methods were compared with those measured for the secondary piping of a PWR plant. Distributions of flow velocity and temperature along the whole system were calculated with 1D and 2D computational flow dynamics (CFD) codes and corrosive conditions were calculated with a N2H4-O2 reaction analysis code. Precise flow turbulence at major parts of the system was analyzed with 3D CFD codes to obtain mass transfer coefficients at structure surfaces. Then, wall thinning rates were calculated with the coupled FAC model by applying the mass transfer coefficients. Comparison of the calculated and measured results led to the following conclusions. 1) Structures with complicated geometry in the plant, e.g., the pair of a bend pipe and a valve, could be simplified as a combination of pipes for...

Irradiation-assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors for a long period. In-core IASCC growth tests have been carried out using the compact tension-type specimens of type 304 stainless steel that had been pre-irradiated up to a neutron fluence level around 1 × 1025 n/m2 under a pure water simulated boiling water reactor (BWR) coolant condition at the Japan Materials Testing Reactor (JMTR). In order to investigate the effect of synergy of neutron/gamma radiation and stress/water environment on SCC growth rate, we performed ex-core IASCC tests on irradiated specimens at several dissolved oxygen contents under the same electrochemical potential condition. In this paper, results of the in-core SCC growth tests are discussed and compared with the results obtained by ex-core tests from a viewpoint of the synergistic effects on IASCC. From results of in-core and ex-core tests using pre-irradiated specime...

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 go...

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 struct...

We studied the effect of the flow rate of reactor water on crack growth rates (CGRs) in a boiling water reactor (BWR) coolant system using 1/4TCT specimens made of sensitized 304 stainless steel (SUS304). Specimen electrochemical corrosion potentials (ECPs) in each oxidant concentration increased with increasing of water flow. This was caused by the increase of the mass transfer amount from bulk water to the specimen surface through a diffusion layer determined by the water flow. A relative CGR was defined as the ratio of the CGR under a given set of conditions to that under simulated normal water chemistry (NWC) at a low flow rate. The relative CGRs at high flow rates were higher than those at low flow rates under NWC. However, when the concentration of oxidants was below 10 ppb and the specimen ECP was below −0.15 V vs. Standard Hydrogen Electrode, the relative CGRs at high flow rates became as low as those at low flow rates. The mass flux of the oxidant was one dominant parameter, which controlled the CGRs at high flow rates. Although the flow rates in a BWR vessel and piping are very high, the CGRs should decrease when the oxidant concentration is adequately reduced by applying hydrogen water chemistry.

The application of chemical cleaning for dissolving and removing scale and sludge is being planned in the Japanese pressurized water reactor (PWR) plant in order to maintain high heat transfer performance and to prevent steam generator (SG) tube degradation. In this paper, the effectiveness of the Electric Power Research Institute (EPRI) and German Kraftwerk Union (KWU) processes on the integrity of structural materials other than SG tubes and the comprehensive applicability of chemical cleaning are discussed. The integrity of structural materials such as carbon steel, low-alloy steel and stainless steel was maintained after the EPRI and KWU processes. KWU chemical cleaning tailored for crevice cleaning has been studied to improve its cleaning effectiveness in crevices and to control the corrosion depth of structural materials less than the criterion for corrosion depth.

The application of the chemical cleaning process to dissolve and remove scales and sludge by chemicals is being planned at the Japanese pressurized water reactor (PWR) plant in order to maintain a designed heat transfer condition and to prevent the steam generator (SG) tube degradation. In this paper, the affects of the EPRI process and the KWU process on the long term integrity of SG tubing were investigated under the simulated SG condition using a SG model boiler test facility. No adverse effect of the both chemical cleaning processes on the long term integrity of SG tubing were observed.

Hydrogen and hydrazine co-injection into a boiling water reactor was considered as a new mitigation method of stress corrosion cracking. To confirm decrease of electrochemical corrosion potential by the reduction of oxygen and hydrogen peroxide in bulk water using reducing agent, reaction of hydrazine with oxygen or hydrogen peroxide under simulated downcomer conditions (temperature: 280°C, duration: 4.2s and gamma-irradiation) was examined. All the oxygen was consumed above the equivalent concentration for the reaction with oxygen within this short time (gamma-irradiation case) and almost all the hydrogen peroxide was consumed. Reaction rates were accelerated more than five times by gamma-irradiation in each case. Reaction rates with oxygen were compared with other reducing agents such as hydrogen, methanol and ammonia. From the viewpoint of reaction rate and formation of by-product, hydrazine was the most suitable agent. Using a simple model based on the experimental results, water chemistry for the bottom region was calculated for the case of no hydrogen injection and for the case of 0.2 mmol·kg -1 hydrogen injection into feed water. For both cases, concentrations of dissolved oxygen and hydrogen peroxide were estimated to decrease enough to mitigate SCC, with concentration of ammonia suppressed below the management criteria for reactor water chemistry.

The application of a chemical cleaning process to dissolve and remove scale and sludge by chemicals is being planned for Japanese pressurized water reactor (PWR) plants in order to maintain a high heat transfer condition and to prevent the SG tubing degradation. In this paper, the effectiveness of the EPRI process and the KWU process to maintain integrity of SG tubing were investigated under accelerated IGSCC conditions. IGSCC susceptibility and propagation ratio of degraded SG tubing with chemical cleaning, were the same as those of tubing not subjected to the chemical cleaning process. Neither the EPRI nor the KWU process adversely affected tubing integrity.

Hydrogen injection has been applied as a preventive measure against the stress corrosion cracking (SCC) phenomenon in many boiling water reactors. However, it can be applied only during normal plant operation since hydrogen is usually injected into the feedwater and this system is in standby mode during start-up operations. It is estimated that the core internals are subjected to the strain rate that may cause susceptibility to SCC initiation during start-up. Therefore, it is beneficial to perform hydrogen injection during start-up as well in order to suppress SCC initiation.For this purpose, we installed an additional hydrogen injection system to be used during plant start-up at the Tokai-2 power station. This trial Hydrogen water chemistry During Start-up (HDS) system was applied following the 19th refueling and maintenance outage in December 2002. By comparing results obtained during this start-up with HDS to previous start-up data using normal water chemistry, we made the following observations. First, as the reactor water temperature increased from initial conditions up to 180 deg. C via nuclear heating, dissolved oxygen and hydrogen peroxide concentrations decreased to levels lower than previously observed. Second, during subsequent nuclear heating, up to 250 deg. C, the dissolved oxygen concentration did not exceedmore » 1 ppb, and the electrochemical corrosion potential was maintained in a low range near -400 mV versus the standard hydrogen electrode.« less

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 bet...

Water chemistry data acquisition, processing and evaluation systems have been applied in major Japanese PWRs and BWRs to prepare for reliable and quick data accumulation with saving in manpower in plant chemical laboratories and to establish smooth and reliable information transfer from chemists to plant operators and supervisors. The following features of the systems were noted. 1. X-ray fluorescence analyses and ion chromatography were widely applied in automatic chemical analysis procedures. 2. Water chemistry data were stored in PC base-data storage systems along with other major plant data to evaluate reliabilities of the data themselves and to prepare for official documents. 3. Clear data displays were effectively applied to support quick response to instances of fuel leakage and long-term analysis for better understanding of the plant's own water chemistry responses. 4. Evaluation of SG tubing integrity is successfully carried out by estimating crevice pH based on ion chromatographic data. 5. Appli...

The application of the chemical cleaning process to dissolve and remove scales and sludge by chemicals is being planned at the Japanese pressurized water reactor (PWR) plant in order to maintain a high heat transfer condition and to prevent the steam generator tube degradation. In this paper, fundamental characteristics and cleaning effectiveness of the EPRI process and the KWU process, which are typical cleaning processes, were investigated. Both processes showed a satisfactory cleaning effectiveness for sludges and scales, and the cleaning effectiveness of the standard KWU process was improved by tailoring it to crevice cleaning conditions.

The effect of zinc addition to pressurized water reactor (PWR) primary water on primary water stress corrosion cracking (PWSCC) susceptibility of mill-annealed Alloy 600 (UNS N06600, 600MA...

Short period tests of hydrogen water chemistry (HWC) were carried out at Tsuruga-1 and Tokai-2 Nuclear Power Plants. The effects of HWC at Tsuruga-1, a non-jet pump plant, are larger than those at Tokai-2, a jet pump plant; this is caused by the much higher recombination rate of hydrogen and oxygen at the downcomer region under suitable gamma ray irradiation for the former. The effective oxygen concentrations ([O2]eff=[O2] + 1/2[H2O2]) in water measured at the reactor pressure vessel bottom are suppressed by HWC as effectively as those concentrations at the recirculating lines in both plants. The effects of HWC in both plants were analyzed and compared with literature data on HWC to evaluate the differences in the effectiveness of HWC as caused by reactor type. From the evaluation, it was concluded that the effectiveness of HWC is much more affected by the power density in the core and the average dose rate at the downcomer than by the reactor type itself. The authors propose an evaluation concept for HWC...

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors (LWRs) for a long period. In general, IASCC can be reproduced on the materials irradiated over a certain threshold fluence level of fast neutron by the post-irradiation examinations (PIEs). It is, however, considered that the reproduced IASCC by PIEs must be carefully compared with the actual IASCC in nuclear power plants, because the actual IASCC occurs in the core under simultaneous effects of radiation, stress and high temperature water environment. In the research field of IASCC, mainly PIEs for irradiated materials have been carried out, because there are many difficulties on SCC tests under neutron irradiation. Hence as a part of the key techniques for in-pile SCC tests, we have embarked on a development of the test technique to obtain information concerning effects of applied stress level, water chemistry, irradiation conditions, etc. A high temperature water loop facility was installed at the Japan Materials Testing Reactor (JMTR) to carry out the in-pile IASCC testing under a framework of cooperative research program between JAERI and the JAPC. In-pile IASCC growth tests have been successfully carried out usingmore » the compact tension (CT) type specimens of type 304 stainless steel that had been pre-irradiated up to a neutron fluence level around 1 x 10{sup 25} n/m{sup 2} before the in-pile testing since 2004. The tests were carried out in pure water simulated boiling water reactor (BWR) coolant condition. In the paper, results of the in-pile SCC growth tests will be discussed comparing with the result obtained by PIEs from a viewpoint of the synergistic effects on IASCC. (authors)« less

The relative rates of transfer of the cyclopentyl radical with molecular bromine and hydrogen bromide (k2′/k−1′) and with hydrogen tribromide and hydrogen bromide (k3′/k−1′) have been determined. The relative transfer rates are compared with the analogous values previously reported for the reactions of cyclohexyl radicals (k2/k−1 and k3/k−1). Utilizing the values of k2′/k−1 and k2/k−1 the competitive vapor phase rates of bromine atom abstraction of hydrogen from the two substrates could be obtained. An expression using a combination of the five sets of relative rate constants was used to determine the effect of competitive cage reversal which occurs in the solution phase bromination of the two substrate radicals with caged hydrogen bromide. For two structurally similar radicals, cage reversal (internal return) was found to affect the relative rates of bromination by 30%, while the relative transfer rates, although differing each by a factor of two, fortuitously nearly cancel each other's effect.The effect of both internal and external reversal reactions on the relative rates of bromination of structurally dissimilar substrates, halogenated alkanes and their parent hydrocarbons, is discussed.

A general method for the evaluation of the kinetics of the solution phase bromination of alkanes and substituted alkanes is presented. The model used considers reversible transfer reactions of the radicals formed with both hydrogen bromide and hydrogen tribromide. Cage reversal of the geminate radical – hydrogen bromide pair and reversal with hydrogen tribromide are used to rationalize the differences observed between the vapor and solution phase kinetics.The method is applied to the bromination of 1-chlorobutane.

Recoil tritium reactions in the aromatic ring were investigated with toluene and benzonitrile in the moderator-free and the C6F6-moderated systems in terms of the intramolecular tritium distribution, the relative tritiation rate, and the isotope effect in toluene. The reactions free from hexafluorobenzene are in essential agreement with non-selective nature shown in hot atom reactions. The addition of the moderator to the systems changes more markedly the uniform labeling to o,p-orientation in toluene than in benzonitrile, and the 1.17 PhCH3/PhCN tritiation rate (in PhCH3-d8/PhCN, 0.942) to the 1.89 PhCH3/PhCN one (in PhCH3-d8/PhCN, 1.51) when the mole ratio of C6F6/substrate is about 100. These findings agree with the electrophilic nature of the tritium at lower energies. The H/D isotope effect in the moderated system is given as 1.89/1.51 or 1.25 and is substantially the same as that (1.24) in the moderator-free system.

The aromatic tritiation induced by irradiation of mixtures of aromatic compounds and propanethiol[S3H] was studied with respect to the tritium distribution in the nucleus and the relative rate of tritiation. A mechanism involving a reaction of an excited aromatic molecule with the thiol is proposed.