Surveillance Specimen Programs for WWER Type Reactors

Four types of surveillance programs were (are) realized in Slovak NPP’s: • “Standard Surveillance Specimen Program” (SSSP) was finished in Jaslovske´ Bohunice V-2 Nuclear Power Plant (NPP) Units 3 and 4; • “Extended Surveillance Specimen Program” (ESSP), was prepared for Jaslovske´ Bohunice NPP V-2 with aim to validate the SSSP results; • For the Mochovce NPP Unit 1 and 2 was prepared completely new surveillance program “Modern Surveillance Specimen Program” (MSSP), based on the philosophy that the results of MSSP must be available during all NPP service life; • For the Bohunice V-1 NPP was finished “New Surveillance Specimen Program” (NSSP) coordinated by IAEA, which gave arguments for prolongation of service life these units for minimum 20 years; • New Advanced Surveillance Specimen Program (ASSP) for Bohunice V-2 NPP (units 3 and 4) and Mochovce NPP (units 1, 2) is approved now. ASSP is dealing with the irradiation embrittlement of heat affected zone (HAZ) and RPV’s austenitic cladding, which were not evaluated till this time in surveillance programs. SSSP started in 1979 and was finished in 1990. ESSP program started in 1995 and will be finished in 2007, was prepared with aim of: • increasing of neutron fluence measurement accuracy; • substantial improvement the irradiation temperature measurement; • fixed orientation of samples to the centre of the reactor core; • minimum differences of neutron dose for all the Charpy-V notch and COD specimens; • the dose rate effect evaluation. In the year 1996 was started the new surveillance specimen program for the Mochovce RPV’s unit-1 and 2, based on the fundamental postulate — to provide the irradiation embrittlement monitoring till the end of units operation. The “New Surveillance Specimen Program” (NSSP) prepared in the year 1999 for the Bohunice V-1 NPP was finished in the year 2004. Main goal of this program was to evaluate the weld material properties degradation due to the irradiation and recovery efficiency by annealing too. The results showed us that the NPP V-1 RPV’s can be operated without any technical limitation for minimum next 20 years. Advanced Surveillance Specimen Program (ASSP) for Bohunice V-2 NPP (units 3 and 4) and Mochovce NPP (units 1 and 2) was prepared as the part of research project dealing with the WWER-440 units ageing management.

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.

Water-water energetic reactors (WWERs) are, in principle, pressurized water reactors but designed and manufactured from Russian-type materials and operated according to former Soviet/Russian rules, standards, and requirements. Their surveillance specimen programs are designed, manufactured, operated, tested, and evaluated with the same purpose but with some differences given by different codes and approaches used in individual countries. This paper describes standard as well as all supplementary and extended surveillance programs in WWER-440/V-213- and WWER-1000/V-320C-type reactors.

High-energy gamma rays emitted from the core of a nuclear reactor produce displacement damage in the reactor pressure vessel (RPV). The contribution of gamma damage to RPV embrittlement has in the past been discounted. However, in certain reactor designs the gamma flux at the RPV is sufficiently large that its contribution to displacement damage can be substantial. For example, gamma rays have been implicated in the RPV embrittlement observed in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory.In the present study, mechanical property changes induced by 10-MeV electron irradiation of a model Fe alloy and an RPV alloy of interest to the HFIR were examined. Tensile specimens were irradiated with high-energy electrons to reproduce damage characteristic of the Compton recoil-electrons induced by gamma bombardment. Substantial increases in yield and ultimate stress were observed in the alloys after irradiation to doses up to 5.3x10−3 dpa at temperatures (~50°C) characteristic of the HFIR pressure vessel. These measured increases were similar to those previously obtained following neutron irradiation, despite the highly disparate nature of the damage generated during electron and neutron irradiations.

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.

4 - Nuclear power plants: types, components and material requirements

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 thermal hydraulics simulation of the reactor coolant system for a typical three-loop pressurized water reactor was conducted based on numerical solution of Reynolds-averaged Navier-Stokes equations by commercial CFD software. This study aims to obtain the three-dimensional, global and localized flow features of the reactor coolant system. The completed model of the reactor coolant system is built including reactor pressure vessel and internals, core, steam generator, primary pump and linking pipe. The flow and the temperature have been investigated under normal steady operating condition with the full core thermal power and unbalanced operating condition with the failure of any primary pump during cold shutdown state. The local thermal hydraulic features of the reactor pressure vessel head dome, the thermal stratification in the reactor pressure vessel upper plenum and the hot legs, and the swirling flow of the primary pump are characterized to give reference to the reactor safety operation. This analysis practice provides an effective evaluation for the three-dimensional thermal hydraulics phenomena, and these encouraging results allow performing the comprehensive analysis for the reactor coolant system.

Magnetic evaluation of hardening of thermal aged Cu-rich RPV steel

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.

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 primary goal of the Heavy-Section Steel Irradiation (HSSI) Program is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure vessel (RPV) integrity. The program includes studies of the effects of irradiation on the degradation of mechanical and fracture properties of vessel materials augmented by enhanced examinations and modeling of the accompanying microstructural changes. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and post-irradiation mitigation are being examined on a wide range of fracture properties. This program will also maintain and upgrade computerized databases, calculational procedures, and standards relating to RPV fluence-spectra determinations and embrittlement assessments. Results from the HSSI studies will be incorporated into codes and standards directly applicable to resolving major regulatory issues that involve RPV irradiation embrittlement such as pressurized-thermal shock, operating pressure-temperature limits, low-temperature overpressurization, and the specialized problems associated with low upper-shelf welds. Six technical tasks and one for program management are now contained in the HSSI Program.

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.

Changes in microstructure and mechanical properties of 1Cr-0.5Mo and 2.25Cr-1Mo steels during long-term service have been investigated. The study includes inspection of a 1Cr-0.5Mo steel reactor vessel which operated for 20 yr and tests on specimens that had been exposed to service environments. The reactor vessel, exposed at 490 to 530°C for 170,000 hr (about 20yr), showed appreciable decrease in the room temperature yield strength, impact toughness and creep rupture strength compared with the original properties. The service exposure caused changes in carbide morphology and species, forming M2C and M7C3 carbides and transferring significant amounts of Cr and Mo from the matrix to carbides; only 32 percent of the total Mo and 72 percent of the total Cr remained in the matrix. The microstructure of 2.25Cr-1Mo steel showed higher stability than that of 1Cr-0.5Mo steel, although a similar transfer of Mo and Cr from the matrix to carbides took place. The Mo and Cr contents remaining in the 2.25Cr-1Mo steel matrix after a 2-yr exposure were only 25 and 74 percent, respectively, of the total quantities in the steel. The partitioning of Mo and Cr to carbides could increase the stability of carbides and consequently reduce the carbon content in the matrix. The mechanical properties are influenced by both the change in the composition of the matrix and the change in carbide morphology. The metallographic examination provides useful information, although qualitative at this stage, on the degree of deterioration of materials.

Applicability of LR-0 mock-up results to VVER-1000 reactor pressure vessel issues