Reactor Pressure Vessel

1 VVER Reactor Types, Designs, Listing and Materials 

The abbreviation VVER is used in this section, although WWER (water-water power reactor) is also used according to IAEA convention. The use of VVER is also recommended by international transliteration standards, and is used by OKB 

Gidropress, who designed the VVER NPP for Russia and other countries that used to belong to the Community for Mutual Economic Assistance (CMEA). 

The materials used in VVER pressure vessels vary with the particular version of the reactor pressure vessel considered. The vessels are all manufactured at three plants Izhora near St. Petersburg, Atommash on the Volga, and the Skoda plant in Plzen in the Czech Republic. The VVER production rate has decreased dramatically in the last few decades. Background information on the different types of VVER is given below and in Table 4.4 [35]. 

The first two VVERs were the 70MWe unit at Rheinsberg and the 210 commissioned in 1963 at Novovoronezh. These were followed by a second prototype (365 MWe) that became operational in 1969. From these prototypes came a standard 440 MWe nuclear power plant that was designated the VVER-440-V230. Usually built in modules with twin units, all such plants have six loops, isolation valves on each loop, horizontal steam generators, and all of them use 220 MWe steam turbines. 

Although the primary coolant system of the Super LWRs and Super FRs is a direct-cycle which is similar to that of BWRs, the RPV design is quite similar to that of PWRs. The major functions of the Super LWR RPV assembly are the following. 

Serve as part of the reactor coolant pressure boundary 

Provide a flow path for circulation of coolant past the fuel 

The RPV of a Super FR is illustrated in Fig. 3.5 [3]. The RPV of the Super LWR is similar to that of the Super FR. Like a typical RPV in PWRs, it has no major penetrations through the lower head. Control rods are inserted from the top of the core because there are no steam separators or dryers. All the internal walls are cooled by the inlet coolant. Only the outlet nozzles are exposed to the hot outlet coolant consequently, they may require a thermal sleeve. This limited exposure avoids thermal creep of the structural steel at the elevated temperature so the conventional steel of the PWR vessel can be used. There are only two inlet and outlet nozzles because the coolant or steam flow rate per electric power is smaller than that of LWRs. 

Further investigations on the Super LWR RPV should be carried out to achieve the following. 

---

Nuclear Boiler Assembly. This assembly consists of the equipment and instrumentation necessary to produce, contain, and control the steam required by the turbine-generator. The principal components of the nuclear boiler are (1) reactor vessel and internals—reactor pressure vessel, jet pumps for reactor water circulation, steam separators and dryers, and core support structure (2) reactor water recirculation system—pumps, valves, and piping used in providing and controlling core flow (3) main steam lines—main steam safety and relief valves, piping, and pipe supports from reactor pressure vessel up to and including the isolation valves outside of the primary containment barrier (4) control rod drive system—control rods, control rod drive mechanisms and hydraulic system for insertion and withdrawal of the control rods and (5) nuclear fuel and in-core instrumentation,... 

Figure 3 shows the nodalisation of the HTTR-IS system model. The reactor consists of the internal flow path (P2), permanent reflector blocks (HS25), upper plenum (B4), reactor pressure vessel (RPV) (HS30), vessel cooling system, reactor core bypass flow (P10), lower plenum (B12) and reactor core. The... 

Reactor Pressure vessel for heating and mixing chemicals and changing their chemical composition. 

Lebedev, V. T., Torok, Gy., Didenko, V. I., Lapin, A. N., Petrov, V. A. and Margolin, B. Z (2004) Investigation of nanostructure of reactor pressure vessel steel with different degree of embrittlement, Physica B 350, e471-e474. 

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. 

As noted above, the ECP is the key parameter in describing the susceptibility of reactor coolant components to corrosion damage. As experience has shown, the direct measurement of ECP in reactor coolant circuits has proved to be very difficult, notwithstanding the monumental efforts of Indig et al. at the General Electric Company-see [3], for example. The major challenge in in-reactor ECP measurements has been to devise a reference electrode that can withstand the harsh environmental conditions that exist within a reactor pressure vessel (RPV). 

Decay heat in fuel elements is assumed to be dissipated by means of heat conduction and radiation to the outside of the reactor pressure vessel, and then taken away to the ultimate heat sink by water cooling panels on the surface of the primary concrete cell. Therefore, no coolant flow through the reactor core would be necessary for the decay heat removal in loss of coolant flow or loss of pressure accidents. The maximum temperature of fuel in accidents shall be limited to 1 bOO C. 

Reactor pressure vessel Core inlet temperature Core outlet temperature Coolant inlet pressure Coolant flow Core power density Average fuel bumup Refuelii interval Gas turbine cycle type... 

Conventional steel reactor pressure vessel construction. 

Popp, K., Brauer, G., and Leonhardt, W.-D. et al., (1989) in Radiation Embrittlement of Nuclear Reactor Pressure Vessel Steels An International Review, ASTMSTP1011, ed. By L.E. Steele (American Society fro Testing and Materials, Philadelphia)... 

Office of Nuclear Regulatory Research, U S. Nuclear Regulatory Commission, Washington NRC (1988) Improved Embrittlement Correlations for Reactor Pressure Vessel Steels , p. 569, Washington DC. 

Odette, G. R. (1998) in M. Davies (Ed ), Neutron Irradiation Effects in Reactor Pressure Vessel Steels and Weldments, lAEA-JWG-LMNPP-98/3, Vienna, 438... 

Avrami and Voreck (Ref 192) reported on a transient radiation test in which a group of nine expls and proplnts in expln-proof irradiation capsules were subjected to a transient burst of energy of about 1 millisecond duration which resulted from fission of about 10 atoms of. One set of nine capsules was attached to the KIWI reactor pressure vessel and another group on the support structure as part of the KIWI-B4-E excursion expt conducted by Los Alamos Scientific Laboratory at the Nevada Test Site. The materials selected were TATB, DATB and TACOT as the secondary expls, HNS as the booster expl, Pb styphnate as the primary expl, BlkPdr as the igniter, and three composite proplnts. Each capsule contained 3.6 grams of the selected material... 

The principal barriers against fission product release into the environment are the high quality TRISO fuel, the reactor pressure vessel, and the reactor building. The calculation of the fission product release during normal operation of the reactor (Fig. 3-5) which determines the contamination of the primary circuit and thus the source term in case of a depressurization or a water ingress accident has again identified silver to be the nuclide with the largest release fraction. 

Thermal stresses are a major concern in reactor systems due to the magnitude of the stresses involved. With rapid heating (or cooling) of a thick-walled vessel such as the reactor pressure vessel, one part of the wall may try to expand (or contract) while the adjacent section, which has not yet been exposed to the temperature change, tries to restrain it. Thus, both sections are under stress. Figure 1 illustrates what takes place. 

Locations in the reactor system, in addition to the reactor pressure vessel, that are primary concerns for thermal shock include the pressurizer spray line and the purification system. 

DESCRIBE the two changes made to reactor pressure vessels to decrease NDT. 

---