Evolution of Standardized Specifications on Materials, Manufacturing and In-Service Inspection of Nuclear Reactor Vessels

The cataloguing and revision of reactor pressure vessels (RPV) manufacturing and in-service inspection codes and their standardized material specifications—as a technical heritage—are essential for understanding the historical evolution of criteria and for enabling the comparison of the various national regulations, integrating the most relevant results from the scientific research. The analysis of the development of documents including standardized requirements and the comparison of regulations is crucial to be able to implement learned lessons and comprehend the progress of increasingly stringent safety criteria, contributing to sustainable nuclear power generation in the future. A novel methodology is presented in this work where a thorough review of the regulations and technical codes for the manufacture and in-service inspection of RPVs, considering the implementation of scientific advances, is performed. In addition, an analysis focused on the differences between irradiation embrittlement prediction models and acceptance criteria for detected defects (both during manufacturing and in-service inspection) described by the different technical codes as required by different national regulations such as American, German, French or Russian is performed. The most stringent materials requirements for RPV manufacturing are provided by the American and German codes. The French code is the most stringent with respect to the reference defect size using as a criterion in the in-service inspection.

Nanoindentation Test of F321 Austenitic Stainless Steel Under Fe-ion Irradiation

Nuclear technology, as a high quality, clean and reliable energy supply, is attracting broad interest from countries across the world. F321 austenitic stainless steel (F321SS) is widely utilized in key components of nuclear power plant due to its excellent corrosion resistance and high temperature mechanical properties. Irradiation can easily lead to the degradation behaviors of materials, such as irradiation hardening, irradiation embrittlement and high-temperature He embrittlement, etc. Understanding such degradation is important for predicting the evolution of material behavior under irradiation and extending the lifespan of existing nuclear reactors. Ion irradiation is most commonly used to model neutron-induced damage since the irradiation conditions (temperature, flux, spectrum, etc.) can be regulated more accurately and flexibly. In this paper, the Fe-ion irradiation experiments of F321SS at different temperatures and doses were carried out, and the nanoindentation experiments under different conditions were further conducted. Irradiation hardening is observed in all specimens and strongly depending on irradiation temperature and damage dose. The hardness after irradiating increases with doses and saturates for at least 1dpa under low temperature regimes (< 300°C). However, at higher temperature (450°C and 560°C), nano-hardness reaches the peak at ∼0.5dpa and then declines. Moreover, the hardness of all specimens has a similar trend with temperature, that is, it first increases, reaches the peak, and then decreases.

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

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.

Development of Toughened, Fine Grained, Recrystallized W-1.1%TiC

Tungsten (W) is a principal candidate as target material because of its high density and extremely high melting point. W inherently has a critical disadvantage of its brittleness at around room temperature (low temperature brittleness), recrystallization embrittlement, and irradiation embrittlement. TFGR (Toughened, Fine Grained, Recrystallized) W-1.1%TiC has been considered as a realized solution to the embrittlement problems. We started to fabricate TFGR W-1.1%TiC in 2016 under collaboration between KEK and Metal Technology Co. LTD (MTC). The TFGR W-1.1%TiC samples were successfully fabricated in June, 2018. As a result, the specimen showed slight bend ductility and 2.6 GPa of fracture strength.

Verification of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel

Nowadays, it has been recognized that probabilistic fracture mechanics (PFM) is a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants, because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. A PFM analysis code PFM analysis of structural components in aging light water reactor (PASCAL) has been developed by the Japan Atomic Energy Agency to evaluate the through-wall cracking frequencies of domestic reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against nonductile fracture. A series of activities has been performed to verify the applicability of PASCAL. As a part of the verification activities, a working group was established with seven organizations from industry, universities, and institutes voluntarily participating as members. Through one-year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group, including the verification plan, approaches, and results.

Weld metal irradiation embrittlement analysis in the range of over-design neutron fluences

The comparison of experimental values of the critical brittle temperature ΔTF and reference temperature ΔT0 of VVER-1000 reactor vessel weld metal with an elevated content of manganese and nickel is performed. ΔTF and ΔT0 values are defined proceeding from the standard impact bend Charpy and Charpy cracked fracture toughness specimen tests, respectively. Specimens were irradiated in industrial reactors in the frame of surveillance specimen program up to the fast (E ≥ 0.5 MeV) neutron fluences corresponding to the NPP long term operation period. The research results showed the shifts ΔTF and ΔT0 to agree with each other. Besides, it was discovered that in the range of over-design fluences the design embrittlement model has a tendency to underestimate the critical brittle temperature shift.

To the issue of accuracy of determination of irradiation embrittlement coefficient

The paper presents the statistical analysis of experimental results of radiation-induced critical brittle temperature ΔTF shifts and reference temperatures ΔT0 obtained respectively from the impact bend and fracture toughness tests of the reactor vessel metal surveillance specimens to define the possibility of their mutual application for the irradiation embrittlement coefficient to be determined more accurately. The correlation between these parameters is shown to remain up to the accumulation of over-design fast neutron fluence.