Reactor pressure vessel RPV steels

Eight samples of un-irradiated reactor pressure vessel (RPV) steel, two each from Trawsfynydd (TRA), Dungeness A (DNA), Sizewell A (SXA) and Bradwell (BWA) reactors were analysed. ICP-MS analysis was carried out using a high resolution magnetic sector instrument. Despite the sensitivity of this method, i.e. lower limit of detection (LLD) of around 8 pg g for procedural blanks, it failed to detect Li and achieved a detection limit of 80 ng g, which was well above the level of interest. However, the results were consistent and did show that the Li concentration was well below that found from the earlier analytical attempts (ICP-OES) and below the levels conservatively assumed in the waste inventory assessments. 

It is widely known that radiation embrittlement behaviour of reactor pressure vessel (RPV) steels depends on various parameters such as material composition, neutron flux and irradiation temperature. Sound understanding and modelling of embrittlement mechanisms require systematic knowledge of effects of individual parameters and their synthesis on microstructural development and then mechanical properties. Most such knowledge has been obtained from single-parameter experiments using test reactor irradiation. This is because test reactor irradiation allows researchers to obtain mechanical property data together with microstructural data on materials with well-controlled chemical compositions under well-controlled irradiation conditions such as flux and temperature. Surveillance data in commercial power reactors are non-systematic in this context and relevant microstructural data are very scarce. 

In the first part of this chapter, we provide an overview of the radiation damage processes that control the form of the irradiation-induced microstructure under the radiation conditions reactor pressure vessel (RPV) steels experience in service. We demonstrate that the irradiation damage in RPV steels can broadly be classified as matrix damage (for instance voids, interstitial clusters, dislocation loops, complexes trapped at sinks such as... 

Application to reactor pressure vessel (RPV) steels general microstructure... 

Steel embrittlement is a complex process that depends on various factors (thermal and radiation treatment, chemical compositions, conditions during preparation, ageing, etc.). The properties of reactor pressure vessel (RPV) steels and the influences of thermal and neutron treatments on these properties are routinely investigated by macroscopic methods such as Charpy V-notch and tensile tests. A number of semiempirical laws based on macroscopic data have been established, but these laws are unfortunately not completely consistent with all data and do not provide the desired accuracy [172-178]. Therefore, many additional test methods have been developed to unravel the complex microscopic mechanisms responsible for RPV steel embrittlement. 

Figure 7.14 Number density of radiation-induced damage as a function of dose obtained using SANS, APT, TEM, and PAS in ferritic alloys and a reactor pressure vessel (RPV) steel [69].

The brittle fracture temperature of steel is a temperature below which its ductility has decreased so that brittle fracture of material is possible. Neutron flux irradiation of the reactor core causes damage in reactor pressure vessel (RPV) wall material and its brittle fracture temperature decreases. If the RPV is cooled below brittle fracture temperature, there is a danger of brittle fracture if there is an initial crack in the RPV wall material. The phenomenon is called pressurised thermal shock (PTS) and its worst consequence is a catastrophic failure of the RPV. PTS is more relevant for PWRs than for BWRs because PWRs generally have a narrower water gap between the reactor core and the RPV wall than BWRs. 

The major difficulty to limit the higher thermal efficiency of the indirect cycle is the lower core inlet temperature. For the direct cycle, the cold gas leaving the precooler can be extracted to cool the reactor pressure vessel (RPV) and the other steel structures. Therefore the core inlet temperature can be as higher as 500- 600°C. For the indirect cycles proposed before, such as MGR-GTI proposed by Yan and Lidskyl, the core inlet temperature is kept as lower as 310°C in order to cool RPV The plant busbar efficiency of MGR-GTI is about 42.1%. In order to achieve higher power plant efficiency, it seems special designs of RPV cooling should be provided for the indirect gas turbine cycle. 

PWR 1965 Abrosimov Fjord a. Reactor compartment (RC) huli b. Miid steel bulkheads c. Reactor pressure vessel (RPV) a. 25% corroded sound b. Unsound c. Little corroded - sound a. 40% -weakened b. Unsound c. 5% corroded sound RC bulkheads now (1996) believed ineffective as containment barrier. It may not be possible to use hull for recovery purposes. RPV remains intact. 

The spherical fuel particle measuring about 1 mm in diameter consists of an inner nuclear kernel coated in successive layers of carbon and ceramics. Thousands of the particles are packed in graphite matrix into a spherical pebble of roughly tennis ball size or a cylindrical compact about the size of man s thumb. A pebble bed core contains a large number of fuel pebbles (for example, 27,000 in the HTR-10 core), and the helium coolant flows in the void volume formed in the pile of the pebbles. On the other hand, a prismatic core contains many hexagonal graphite blocks (150 in the HTTR core) in which the fuel compacts are embedded and the hehum coolant flows in the channels provided in the block. Both cores are surrounded by graphite reflector and enclosed in steel pressure vessel. Reactivity control rods (RCRs) are inserted from above the reactor pressure vessel (RPV). 

The reactor core and the SG are housed in two steel pressure vessels that are connected by a connecting vessel. Inside of the connecting vessel, the hot gas duct is designed. All of the pressure retaining components, which comprise the primary pressure boundary, are in touch with the cold helium of the reactor inlet temperature. The primary pressure boundary consists of the reactor pressure vessel (RPV), the SG pressure vessel (SGPV), and the hot gas duct pressure vessel (HDPV), which all are housed in a concrete shielding cavity as shown in Fig. 14.8. 

Trend of copper content in Japanese RPV steels. (Reprinted, with permission, from Industry Practice for the Neutron Irradiation Embrittlement of Reactor Pressure Vessels in Japan, by Norimichi Yamashita, Masanobu Iwasaki, Koji Dozaki, Naoki Soneda, Journal of Engineering for Gas Turbines and Power, Voiume 132(10), 2010, copyright ASME, Two Park Avenue, NY 10016-5990). 

A. Amyev, A. Krynkov, M. Sokolov, Recovery of Transition Temperature of VVER RPV by Annealing, Steele L. E. (ed.). Radiation Embrittlement of Nuclear Reactor Pressure Vessel Steels An International Review (Fourth Volume), ASTM STP1170, American Society for Testing and Materials, Philadelphia, PA, pp. 369-379,1993. 

Microstructural Characterization of Reactor Pressure Vessel Steels Post-Irradiation Annealing Experiments Joint EPRI-CRIEPI RPV Embrittlement Studies (1999-2004), Electric Power Research Institute, Palo Alto, CA, 2004, 1003531, and CRIEPI, Tokyo, Japan, Q980401. 

T.J. Williams, K. Wilford, G.R. Odette and T. Yamamoto, A new model of irradiation hardening in low copper RPV steels from stable matrix damage, IAEA Technical Meeting on Irradiation Embrittlement and Life Management of Reactor Pressure Vessels in Nuclear Power Plants, Znojmo, Czech Repnblic, 18-22, October, 2010. 

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