Beyond RPV Design Life

Results of impact bend tests of subsize Charpy V-notch-type (CVN) specimens are used in Russia for monitoring both re-irradiation embrittlement of the first generation of VVER-440 (including VVER-440/179 and VVER-440/230, i.e., projects B-179 and B-230) reactor pressure vessel (RPV) steels and irradiation embrittlement of VVER-440/213 RPV steels. For the first generation of VVER-440 RPVs, no surveillance program was foreseen and only subsize specimens (5 by 5 by 27.5 mm and 3 by 4 by 27 mm) can be manufactured from samples taken from the inner surface of the RPV for radiation embrittlement monitoring. For VVER-440/213 RPVs, 80 % of the surveillance specimens are irradiated with a lead factor greater than 20. Only four specimens of base metal and four specimen halves of weld seam material are irradiated with a lead factor less than 3. Taking into account that full-size (10 by 10 by 55 mm) CVN specimens were used in surveillance programs, using subsize specimens enables to solve the flux effect problem for VVER-440/213 RPV beltline materials and ensure representativity of surveillance programs. Application of subsize specimens for the estimation of radiation embrittlement requires developing a correlation between transition temperature values evaluated using test results of standard (full-size) and subsize CVN specimens. Such a correlation is discussed.

The Palisades Nuclear Generating Station (PNGS), located in Michigan, is owned and operated by Entergy. It is a Combustion Engineering 2-loop pressurized water reactor (PWR) producing 805 MWe (2,565 MWth). The PNGS was built between 1967 and 1970, with approval to operate at full power in 1973; the plant’s original license was due to expire on March 24, 2011. An application for 20-year extension was filed in 2005 with the Nuclear Regulatory Commission and was granted on January 18, 2007. Although the plant was scheduled for decommissioning by 2031, Entergy currently plans to close the PNGS in 2022. The PNGS included in its reactor pressure vessel (RPV) surveillance program a capsule, designated A-60, containing specimens of one of the vessel plates and a weld metal with nickel content of about 1.36 wt% and copper content of about 0.20 wt%. This capsule was irradiated to a fluence of 1.8 x 1020 n/cm2. The capsule was removed from its surveillance position in 1995 and has been resident in the spent fuel pool since that time. The material is also of special interest because of its very high nickel content and because of the potential for development of NiMnSi (nickel-manganese-silicon) precipitates, dubbed “late blooming phases.” The surveillance program also includes a capsule, designated T-150, dedicated specifically for thermal aging, which would provide results for at least 33 effective full power years, which is beyond the current thermal-aging database for such materials. Given that license extensions to 60 years of operation have been approved by the US Nuclear Regulatory Commission for most of the currently operating light water reactors in the United States, and that the first extension to 80 years was recently approved, there exists the probability that some RPVs will reach and possibly exceed a fast neutron fluence (> 1 MeV) of 1 x 1020 n/cm2. This is a fluence regime with no US surveillance data and very little test reactor data, except for the Light Water Reactor Sustainability Program–sponsored University of California Santa Barbara Advanced Test Reactor (ATR) ATR-2 project. Thus, the materials in the A-60 capsule represent a valuable resource for directly exploring the effects of commercial surveillance irradiation on a typical plate and a high-nickel weld with similar materials irradiated in the ATR-2 test reactor project. This report provides background information for the surveillance program, previous results of surveillance materials testing, the preliminary plan for capsule retrieval and disassembly, and the plan for testing and microstructural examination of the mechanical test specimens of these unique materials to assess the features induced by very high irradiation fluence or very long thermal aging time

In order to develop the integrity evaluation technology for aged major components of nuclear power plants, the Japan Power Engineering and Inspection Corporation (JAPEIC) has been carrying out the project named “Nuclear Power Plant Integrated Management Technology (PLIM)” entrusted by Japanese Ministry of Economy, Trade and Industry (METI) since 1996Fiscal Year (FY). One of the objectives of this project is to establish the method for integrity evaluation of aged domestic reactor pressure vessels (RPV) by developing the prediction equations of reduction of Charpy V-notch Upper Shelf Energy (USE) due to neutron irradiation embrittlement and the correlation equations between USE and fracture toughness. Because the tests are now in progress, this paper presents the following preliminary results as of the end of 2000FY, using irradiated Charpy and Compact Tension specimens of RPV materials. • Study on the effect of accelerated irradiation for test specimens. • Comparison of USE values between test results and U.S. surveillance data. • Comparison of USE values between test results and published predictions. • Correlation between USE value and fracture toughness.

Evaluation of pavement service life using AASHTO 1972 and mechanistic-empirical pavement design guides

Need for Further Development in Service Life Modelling of Concrete Structures in Chloride Environment

The service-dependent degradation (aging) of light water reactor (LWR) pressure vessels due to irradiation embrittlement is discussed in this paper. The major variables which influence the irradiation embrittlement of LWR vessels are the copper and nickel content of the vessel materials and the fluence. The vessel beltline region is subjected to the largest fluences. A surveillance program, which consists of tension and Charpy-V-notch (CVN) testing of irradiated specimens of base, heat-affected-zone, and weld materials, is required to monitor changes in their embrittlement. Three main unresolved technical issues are: (1) the limited range and accuracy of the current correlations for calculating shifts in the reference temperature for nil-ductility transition (RTNDT) and changes in the Charpy upper shelf energy; (2) the need to demonstrate the conservatism of using CVN-based RTNDTshifts for certain sensitive reactor pressure vessel materials; and (3) the type of surveillance program required for any renewed operating license period. The damage caused by irradiation embrittlement can impact plant operating procedures, including heat up/cool down and hydrostatic test procedures, as well as the acceptability of various plant transients.

Nuclear energy is highly valued in many countries for its reliability, cleanliness and cost-effectiveness. Ensuring the safe development and utilization of nuclear energy has become crucial for addressing the global energy shortage and mitigating climate change. The reactor pressure vessel (RPV) is exposed to a high-temperature, high-corrosion and high-radiation environment, the RPV materials have a significant impact on the stability and safety of a nuclear power station. In this paper, the effects of long-term high-temperature hot aging on microstructure and mechanical properties of RPV steel, as well as the research progress in embrittlement mechanism and service life prediction, are reviewed. At the same time, the existing problems and further research directions of RPV steel are pointed out.

The reactor pressure vessel (RPV) is generally considered to be the most important component in a nuclear power plant in terms of safety and overall structural integrity. An RPV is not considered to be a replaceable component, and, thus, the monitoring of the material toughness properties during long-term operation is of crucial importance. RPV surveillance programs are almost always applied for such monitoring, and the design of the capsules, operations to install and remove capsules, and testing of irradiated specimens is governed by international/national codes and standards and can differ between individual reactor designs, as well as requirements/methodologies used in different countries. Results from testing actual RPV material specimens irradiated in the surveillance capsules serve to the assessment of current and future degradation levels of RPV material toughness properties. Thus, the surveillance program results provide needed input for RPV integrity evaluations and future operating lifetime determination.

The majority of pressurized water reactors (PWR) operating in the United States have a “design basis life” of 30 to 40 years. These design basis life estimations were not based on technical studies of material degradation in general but rather on fatigue usage factors for the most part. Recognizing this fact, the subject of operating an existing nuclear steam supply system (NSSS) for longer periods than originally intended has become an important issue worldwide. Radiation embrittlement of the reactor vessel beltline region (the area surrounding the height of the reactor core) is the main concern in extending the operation of the NSSS. Radiation damage, if any, to the reactor pressure vessel material is monitored by a material radiation surveillance program. If the data from the reactor vessel materials surveillance program indicate that the reactor pressure vessel will not meet the rules of the Code of Federal Register (CFR) and various regulatory guides (RG), there are a number of options a utility may take to ensure reactor pressure vessel design life attainment or extension. This paper describes the results from 12 surveillance capsule programs, which encompass four reactor vessel materials and five reactor vessel manufacturers.

The structural integrity of a reactor pressure vessel (RPV) is important for the safety of a nuclear power plant. When the emergency core cooling system (ECCS) is operated and the coolant water is injected into the RPV due to a loss-of-coolant accident (LOCA), the pressurized thermal shock (PTS) loading takes place. With the neutron irradiation, PTS loading may lead a RPV to fracture. Therefore, it is necessary to evaluate the performance of RPV during PTS loading to keep the reactor safety. In the present study, optimization of RPV maintenance is considered, where two different attempts are made to investigate the RPV integrity during PTS loading by employing the deterministic and probabilistic methodologies. For the deterministic integrity evaluation, 3D-CFD and finite element method (FEM) simulations are performed, and stress intensity factors (SIFs) are obtained as a function of crack position inside the RPV. As to the probabilistic integrity evaluation, on the other hand, a more accurate spatial distribution of SIF on the RPV is calculated. By comparing the distribution thus obtained with the fracture toughness included as a part of the master curve, the dependence of fracture probabilities on the position inside the RPV is obtained. Using the spatial distribution of fracture probabilities in RPV, the priority of the inspection and maintenance is finally discussed.

Reactor pressure vessels (RPV) are components with the highest importance for the reactor safety and operation as they contain practically whole inventory of fission material but they are damaged/aged during their operation by an intensive reactor radiation. Surveillance specimen programs are the best method for monitoring changes in mechanical properties of reactor pressure vessel materials if they are designed and operated in such a way that they are located in conditions close to those of the vessels. Reactor Codes and standards usually included requirements and conditions for such programs to assure proper vessel monitoring. WWER (Water-Water-Energetic Reactors) reactor pressure vessels are designed according to former Russian Codes and rules with somewhat different requirements using different materials comparing e.g. with ASME Code. Two principal types of WWER reactors were designed, manufactured and are operated in several European countries (and also in China, Iran): WWER-440 and WWER-1000. Their surveillance programs were designed in quite different way, with some modifications due to the time, country of manufacturing and experience gained from their operation. The paper gives a critical comparison of these programs in both types of reactors with requirements of both Russian and ASME/ASTM Codes and Standards. Finally, information about creation of the Integral Surveillance Program for WWER-1000 type reactor pressure vessels covering vessels from several countries is described.

To ensure the structural integrity of the embrittled reactor pressure vessels (RPVs) during startup or shutdown operation, the pressure-temperature (P-T) limits are mainly determined by the fracture toughness of beltline region material with the highest level of neutron embrittlement. However, other vessel parts such as nozzles with structural discontinuities may affect the limits due to the higher stress concentration, even though the neutron embrittlement is insignificant. Therefore, not only beltline material with the highest reference temperature, but also other components with structural discontinuities have to be considered for the development of P-T limits of RPV. In the paper, the pressure-temperature operational limits of a Taiwan domestic pressurized water reactor (PWR) pressure vessel considering beltline and extended beltline regions are established per the procedure of ASME Code Section XI-Appendix G. The three-dimensional finite element models of PWR inlet and outlet nozzles above the beltline region are also built to analyze the pressure and thermal stress distributions for P-T limits calculation. The analysis results indicate that the cool-down P-T limit of the domestic PWR vessel is still dominated by the beltline region, but the heat-up limit is partially controlled by the extended beltline region. On the other hand, the relations of reference temperature between nozzles and beltline region on the P-T limits are also discussed. Present work could be a reference for the regulatory body and is also helpful for safe operation of PWRs in Taiwan.

The original Babcock & Wilcox (B&W) 177-fuel assembly (177-FA) reactor vessel surveillance programs were designed to determine the effects of neutron irradiation of B&W 177-FA plant reactor vessel materials. The B&W 177-FA plant integrated reactor vessel surveillance program was developed as the result of two events: failure of the original surveillance capsule holder tubes in B&W 177-FA-operating plants and the necessity of obtaining fracture toughness data for irradiated weld metals to ensure the continued licensability of the B&W 177-FA-operating plants. The integrated reactor vessel surveillance program assesses data from two separate but interrelated projects: the plant-specific surveillance program and the power reactor program. The plant-specific surveillance program integrates the various plant-specific surveillance programs to ensure the availability of data on a timely basis and to meet the basic requirement that each operating B&W 177-FA plant reactor has a surveillance program. The power reactor program provides the fracture toughness properties of eight weld metals that complement the data obtained from the plant-specific surveillance capsules. The Master Integrated Reactor Vessel Surveillance Program (MIRVP) builds on the earlier B&W 177-FA plant integrated reactor vessel surveillance program. MIRVP includes the integration of plant-specific reactor vessel surveillance program capsules (both original B&W 177-FA plant participants and the added Westinghouse plant participants), the existing supplemental irradiation capsules developed for the original B&W 177-FA plant integrated program, and additional supplementary irradiation surveillance capsules. The purpose of MIRVP is to provide the data required to ensure the continued licensability of pressurized water reactor vessels fabricated by B&W.

Reactor dosimetry (RD) is a tool that provides data for neutron fluence accumulated over the reactor pressure vessel (RPV) during the reactor operation. This information, however, is not sufficient for RPV lifetime assessment. The life management of RPV is a multidisciplinary task. To assess whether the RPV steel properties at the current stage (for actual accumulated neutron fluence) of reactor operation are still “safe enough,” the dependence of material properties on the fluence must be known; this is a task for material science (MS). Moreover, the mechanical loading over the RPV during normal operation and accidences have to be known, as well, for evaluation, if the RPV material integrity in this loading condition and existing cracks is provided. The crack loading path in terms of stress intensity factor is carried out by structural analyses (SA). Pressure and temperature distribution over RPV used in these analyses are obtained from a thermal hydraulic (TH) calculation. The conjunction of RD and other disciplines in RPV integrity assessment is analyzed in accordance with the FFP (fitness for purpose) approach. It could help to improve the efficiency in multi-disciplinary tasks solutions.

Chapter 4.1 - General Methodology