Japanese Pressurized Water Reactor Research Articles

Determination and Sensitivity Analysis of Target Flaw Sizes for Performance Demonstration of Cast Austenitic Stainless Steel Pipe by Probabilistic Fracture Mechanics

Abstract The material of primary coolant pipe of nuclear power plants is cast austenitic stainless steel (CASS), which has a coarse crystal structure and its anisotropy causes low permeability of ultrasonic waves and low detectability. Also, the material strength and fracture toughness of CASS are affected by thermal aging depending on ferrite content, chemical compositions and casting process. The ASME Section XI Code Case N-838 provides the target flaw sizes for performance demonstration (PD) examination system of ultrasonic testing (UT) based on probabilistic fracture mechanics (PFM) analysis incorporated a thermal aging model. In this study, the target flaw sizes for Japanese pressurized water reactor (PWR) plants were investigated by using the PFM analysis code “PREFACE.” Thermal aging prediction models for tensile strength and fracture toughness of CASS were incorporated into the PREFACE Code depending on not only ferrite content and chemical compositions but also the casting process, centrifugal casting, or static casting. Those parameters were input as probabilistic variables, which were based on the material database from the Japanese PWR plants. The PFM analysis revealed that static CASS pipes have smaller target flaw sizes than centrifugal CASS pipes and the obtained target flaw sizes were very close to those in ASME Code Case N-838.

The role of silicon on solute clustering and embrittlement in highly neutron-irradiated pressurized water reactor surveillance test specimens

• The role of Si content on irradiation embrittlement was clarified. • Solute atom clusters of highly neutron-irradiated Japanese pressurized water reactor surveillance test specimens were analyzed by atom probe tomography. • The cluster radius and number density decreased and increased, respectively, with increasing Si content, resulting in a constant volume fraction of SCs. • Increase in Si reduces the degree of irradiation embrittlement. This is consistent with the results of our previous study based on Bayesian statistical analysis. Solute clusters (SCs) formed in pressurized water reactor surveillance test specimens neutron-irradiated to a fluence of 1 × 10 20 n/cm 2 were analyzed via atom probe tomography to understand the effect of silicon on solute clustering and irradiation embrittlement of reactor pressure vessel steels. In high-Cu bearing materials, Cu atoms were aggregated at the center of cluster surrounded by the Mn, Ni, and Si atoms like a core-shell structure. In low-Cu bearing materials, Mn, Ni, and Si atoms formed cluster and these solutes were not comprised core-shell structure in SCs. While the number of Cu atoms in clusters was decreased with decreasing nominal Cu content, the number of Si atoms had clearly increased. The cluster radius ( r ) and number density ( N d ) decreased and increased with increasing nominal Si content, respectively. The shift in the reference temperature for nil-ductility transition ( ΔRT NDT ) showed a good correlation with the square root of volume fraction ( V f ) multiplied by ( V f × r ) . The negative relation between the nominal Si content and ΔRT NDT indicated that increasing of nominal Si content reduces the degree of embrittlement.

Bayesian Analysis of Japanese Pressurized Water Reactor Surveillance Data for Irradiation Embrittlement Prediction

Abstract The goal of this study was to identify the chemical component variables that should be used in irradiation embrittlement prediction and to determine the uncertainty of prediction of irradiation embrittlement of reactor pressure vessel (RPV) steels. To this end, statistical analysis using a Bayesian nonparametric (BNP) method was performed for Japanese pressurized water reactor (PWR) surveillance test data whose neutron fluence ranged from 3 × 1018 to 1.2 × 1020 n/cm2 (E > 1 MeV). The BNP method is a machine learning statistical method that takes the complexity and uncertainty of input variables into account. Statistical analysis using an index to select the most suitable combination of input variables revealed that four variables, namely, neutron fluence and Cu, Ni, and Si contents, were the most effective combination for embrittlement prediction. Cu content had the largest effect on the degree of embrittlement, followed by Ni and Si, in that order. The shift in the reference nil-ductility temperature (ΔRTNDT) was also calculated using the probability distribution obtained by the BNP method. The overall standard deviation of the residuals between the calculated and measured values of ΔRTNDT was 8.4 °C, which was comparable to that of the current Japanese embrittlement correlation method (JEAC4201-2013). The 95% credible interval (CI) of the posterior distribution of ΔRTNDT (i.e., the range in which data can exist when the uncertainty of input data is taken into consideration) calculated by the BNP method was identical to or smaller than the margin in the current Japanese embrittlement correlation method described in JEAC4201-2013. This result indicates that an adequate margin is provided in JEAC4201-2013.

Effect of Coolant Water Temperature of Emergency Core Cooling System on Failure Probability of Reactor Pressure Vessel

Abstract Structural integrity assessment of reactor pressure vessel (RPV) is important for the safe operation of nuclear power plant. For an RPV in a pressurized water reactor (PWR), pressurized thermal shock (PTS) resulted from rapid coolant water injection due to a loss-of-coolant accident is an issue of particular concern. The coolant water temperature in the emergency core cooling system (ECCS) can influence the integrity of RPV subjected to PTS events; thus, this paper is focused on investigating the effect of coolant water temperature of ECCS on failure probability of an RPV. First, thermal-hydraulic (TH) analyses were conducted for a Japanese PWR model plant by using RELAP5, and different coolant water temperatures in ECCS were considered to investigate the effect of coolant water temperature on TH behaviors during a PTS event. Using the TH analysis results, probabilistic fracture mechanics (PFM) analyses were performed for the RPV of the Japanese model plant. Based on the PFM analysis results, the effect of coolant water temperature on failure probability of the RPV was quantified.

PWR secondary water chemistry guidelines in Japan - Purpose and technical background

In the more than 40 years of operational history of pressurized water reactors (PWRs) in Japan, sustainable development of water chemistry technologies has resulted in the world's highest secondary system component integrity; additionally, secondary system components, especially steam generator (SG) tubing, with excellent material integrity have been developed to prevent leakage of radioactive contamination from the primary to the secondary system and to maintain the heat removal function of the secondary system. Although reasonable control and diagnostic parameters for water chemistry are utilized by each PWR owner, the specific values are not shared.To ensure reliable PWR operation and to achieve the highest safety level, relevant members of the Standards Committee and the related committee organized by the Atomic Energy Society of Japan (AESJ) decided to establish water chemistry guidelines for PWRs. The Japanese PWR secondary water chemistry guidelines provide strategies for improving material integrity and the heat removal function. The guidelines also provide reasonable “action levels” for control parameters and “control values” and “diagnostic values” for multiple parameters, and they stipulate the responses when these levels are exceeded. Specifically, “conditioning parameters” are adopted in the guidelines. Good operational practice conditions are also discussed with reference to long-term experience.This paper introduces the purpose, technical background and framework of preliminary secondary water chemistry guidelines for Japanese PWRs. Addition, the differences in bases of parameter settings between the Japanese and overseas guidelines are discussed.

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

The thermal hydraulic (TH) behavior of coolant water is a key factor in the structural integrity assessments on reactor pressure vessels (RPVs) of pressurized water reactors (PWRs) under pressurized thermal shock (PTS) events, because the TH behavior may affect the loading conditions in the assessment. From the viewpoint of TH behavior, configuration of plant equipment and their dimensions, and operator action time considerably influence various parameters, such as the temperature and flow rate of coolant water and inner pressure. In this study, to investigate the influence of the operator action time on TH behavior during a PTS event, we developed an analysis model for a typical Japanese PWR plant, including the RPV and the main components of both primary and secondary systems, and performed TH analyses by using a system analysis code called RELAP5. We applied two different operator action times based on the Japanese and the United States (US) rules: Operators may act after 10 min (Japanese rules) and 30 min (the US rules) after the occurrence of PTS events. Based on the results of TH analysis with different operator action times, we also performed structural analyses for evaluating thermal-stress distributions in the RPV during PTS events as loading conditions in the structural integrity assessment. From the analysis results, it was clarified that differences in operator action times significantly affect TH behavior and loading conditions, as the Japanese rule may lead to lower stresses than that under the US rule because an earlier operator action caused lower pressure in the RPV.

BWR water chemistry guidelines and PWR primary water chemistry guidelines in Japan – Purpose and technical background

After 40years of light water reactor (LWR) operations in Japan, the sustainable development of water chemistry technologies has aimed to ensure the highest coolant system component integrity and fuel reliability performance for maintaining LWRs in the world; additionally, it aimed to achieve an excellent dose rate reduction. Although reasonable control and diagnostic parameters are utilized by each boiling water reactor (BWR) and pressurized water reactor (PWR) owner, it is recognized that specific values are not shared among everyone involved.To ensure the reliability of BWR and PWR operation and maintenance, relevant members of the Atomic Energy Society of Japan (AESJ) decided to establish guidelines for water chemistry. The Japanese BWR and PWR water chemistry guidelines provide strategies to improve material and fuel reliability performance as well as to reduce dosing rates. The guidelines also provide reasonable “control values”, “diagnostic values” and “action levels” for multiple parameters, and they stipulate responses when these levels are exceeded. Specifically, “conditioning parameters” are adopted in the Japanese PWR primary water chemistry guidelines. Good practices for operational conditions are also discussed with reference to long-term experience.This paper presents the purpose, technical background and framework of the preliminary water chemistry guidelines for Japanese BWRs and PWRs. It is expected that the guidelines will be helpful as an introduction to achieve safety and reliability during operations.

Determination of 79Se and 135Cs in Spent Nuclear Fuel for Inventory Estimation of High-Level Radioactive Wastes

79Se and 135Cs are long-lived fission products and are found in high-level radioactive waste (HLW). The estimation of their inventories in HLW is essential for the safety assessment of geological disposal, owing to their mobility in the strata. In this study, the amounts of 79Se and 135Cs in a spent nuclear fuel solution were measured. About 5 g of irradiated UO2 fuel discharged from a commercial Japanese pressurized water reactor (PWR) with a burn-up of 44.9 GWd/t was sampled and dissolved with 50mL of 4M nitric acid in a hot cell for 2 h. After Se and Cs were chemically separated, the amounts of 79Se and 135Cs in the spent nuclear fuel solution were measured by inductively coupled plasma quadrupole mass spectrometry (ICP-QMS). The amounts of 79Se and 135Cs were 5:2 ± 1:5 and 447 ± 40 g/MTU, respectively. The results presented in this study, which are the first postirradiation experimental data in Japan, showed good agreement with those obtained by the ORIGEN2 code using the data library of JENDL-3.3.

Numerical Simulation on Effect of Methanol Addition on Coolant Radiolysis in Pressurized Water Reactor

The addition of alcohol into the primary coolant of pressurized water reactors (PWRs) was proposed as an alternative of hydrogen addition. The radiolysis of the coolant in the presence of methanol was simulated, assuming a batch system and changing dose rate. At 300°C, the addition of 4.0 ppm methanol shows an equivalent scavenging capacity toward the OH radical to a typical dissolved hydrogen concentration in Japanese PWR plants. At low dose rate, the addition of 1.0 ppm methanol suppresses considerably the radiolysis of the coolant. At high dose rate, the addition of at least 10 ppm methanol is necessary to lower concentrations of the radiolytic oxidizing products. Concentrations of the radiolytic oxidizing products are determined not only by the scavenging of the OH radical but also by secondary reactions. In the presence of fast neutron irradiation, methanol acts to decompose the radiolytic oxidizing products. The fact that steady-state concentrations of the final organic products depend on dose rate means that complicated secondary reactions contribute to the decomposition of methanol. The major final organic product is formed through a reaction of ethylene glycol with OH. Methanol addition may have the same effect as ethylene glycol addition, and radiation chemical reactions of ethylene glycol systems should be elucidated in future.

Numerical Simulation on Effect of Methanol Addition on Coolant Radiolysis in Pressurized Water Reactor

The addition of alcohol into the primary coolant of pressurized water reactors (PWRs) was proposed as an alternative of hydrogen addition. The radiolysis of the coolant in the presence of methanol was simulated, assuming a batch system and changing dose rate. At 300°C, the addition of 4.0 ppm methanol shows an equivalent scavenging capacity toward the OH radical to a typical dissolved hydrogen concentration in Japanese PWR plants. At low dose rate, the addition of 1.0 ppm methanol suppresses considerably the radiolysis of the coolant. At high dose rate, the addition of at least 10 ppm methanol is necessary to lower concentrations of the radiolytic oxidizing products. Concentrations of the radiolytic oxidizing products are determined not only by the scavenging of the OH radical but also by secondary reactions. In the presence of fast neutron irradiation, methanol acts to decompose the radiolytic oxidizing products. The fact that steady-state concentrations of the final organic products depend on dose rate means that complicated secondary reactions contribute to the decomposition of methanol. The major final organic product is formed through a reaction of ethylene glycol with OH. Methanol addition may have the same effect as ethylene glycol addition, and radiation chemical reactions of ethylene glycol systems should be elucidated in future.

Applicability of Chemical Cleaning Process to Steam Generator Secondary Side, (IV)

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.

Applicability of Chemical Cleaning Process to Steam Generator Secondary Side, (III)

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.

Applicability of Chemical Cleaning Process to Steam Generator Secondary Side, (II)

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.

Applicability of Chemical Cleaning Process to Steam Generator Secondary Side, (I)

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.

On-line fatigue-monitoring system for nuclear power plant

Fatigue is one of the most severe aging factors for nuclear components and it is very important to monitor the degradation due to thermal fatigue in order to keep the integrity of components when the plant life extension is performed. From this viewpoint, the Japanese pressurized-water reactor (PWR) group has developed an advanced transient and fatigue usage monitoring system which can cope with thermal stratification pheomenon and inner surface heat transfer coefficient change on the critical locations during plant operations. The system estimates the thermal transient changes making use of plant process data and calculates the stress vs. time history and fatigue usage factor on the critical locations. In this paper, the outline of the fatigue-monitoring system of the Japanese PWR group is described and the system software logic is verified using measured data obtained from a PWR plant and detailed computer simulations such as three-dimensional finite element analyses.

Designing a Knowledge-Based Operator Support System for Practical Applications

A practical knowledge-based operator support system is being developed for Japanese pressurized water reactors. This system will be implemented at real power plants in the near future. The difficulty of realizing a practically usable system using normative models based on deep knowledge is discussed. Instead of adopting the normative approach, the system introduces a hierarchically organized model, called a “plant abnormality model,” to its diagnosis part. With its ability to envelope unforeseen events, it avoids the use of imperfect deep knowledge and the scenario dependability that is considered an inherent problem in abnormality models. Existing operational procedures are broken down into functionally independent task units and specified as knowledge sources for operational guidance. Depending on the plant status, relevant task units are dynamically integrated to synthesize operational procedures that are provided as operational guidance. Estimated information on unobservable or predictive plant status is used to enable flexible and timely synthesis of the procedures. An attempt is made to organize the information so that it is better understood by the operators by adopting the hypothesis-and-test scheme as a framework for the inference control mechanisms of the diagnosis system.