• Energy Research

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.

A methodology has been developed to quantitatively assess the suitability of use and fitness for service of candidate materials using a novel approach that includes multiple perspectives. As a case study, a carbon steel pipe has been selected for operation in the petrochemical sector. The materials studied were the following: American Petroleum Institute (API) A25, A, B, X42, X46, X52, X56, X60, X65 and X70, as well as American Society for Testing and Materials (ASTM) A-106 Gr. A, B and C. The developed model combines an analytical multiperspective approach with calculation methods based on recognized prestige standards. In the present study, the following material degradation mechanisms have been considered: generalized corrosion, fracture due to mechanical overload and high-temperature degradation. Several novel analysis elements have been incorporated into this new methodology, such as the concept of a suitability matrix and a fitness for service index. The approach allows construction of a decision diagram, and the best alternatives ordered according to the criteria and restrictions that have arisen from the analysis are obtained. Additionally, from the analysis, a series of service limitations are proposed based on the maximum hours of operation of a component. The materials ASTM A-106 Gr. A, API-A, ASTM A-106 Gr. B and API-B maintain the best balance between properties and show greater reliability versus the probability of failure due to the degradation mechanisms considered in this study. In addition, some use limitations such as critical exposure temperature have been determined for these materials (450 °C for ASTM A-106 Gr. A designation and 440 °C for API-B and ASTM A-106 Gr. B designations) to avoid the harmful effects of high-temperature operation on the material mechanical properties.

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.