WWER reactors

In 1995, a meeting of an expert group from the Moscow Centre of WANO was held at Dukovany NPP on problems concerning the loss of natural circulation for WWER-440 reactors. The events concerning the loss of the natural circulation were discussed and reviewed in detail, the causes were analysed and the remedies were proposed to prevent the reoccurrence of similar events. 

During normal power operation, the water level in WWER-440 reactors is not monitored. In cold shutdown, a standard reactor level measurement is used based on water differential pressure in the U-tube, with the scale between 180 and 250 cm of narrow-range level. The positive reference leg of the U-tube is connected to a tube used for continual monitoring of boron concentration in reactor coolant. This standard reactor level measurement is activated prior to the depressurization of the main reactor seal flange plane and removal of the head. 

IAEA, Guidelines for Prediction of Irradiation Embrittlement of Operating WWER-440 Reactor Pressure Vessels, IAEA TECDOC 1442, International Atomic Energy Agency, Vienna, 2005. 

Similar but varying results accrue as a result of other sensitizing elements, such as nickel, manganese, silicon and phosphorus. Moreover, the effects are similar for the WWER steels, with copper and phosphorus being the dominant elements present in the WWER-440 reactors, but with nickel and manganese being the dominant elements present in the WWER-1000 reactors. 

Research work has been undertaken to determine the dependence of irradiation embrittlement on material chemistry, heat treatment and service factors (irradiation conditions, temperature, coolant chemistry, etc.) in Russian research reactors and WWER-440 reactors. The irradiations were carried out mainly on surveillance specimens irradiated in nuclear power plants in locations where the inlet temperature was 270 °C (for WWER-440). The intent was to provide carefully controlled irradiation conditions in terms of temperature and neutron spectrum. The resulting major conclusion was that a substantial body of data established an irradiation-induced increase of the brittle fracture temperature, (similar to Tnj), of the general form of (CF) (FF) ... 

For WWER reactors, a methodology for analysis of LPS has been developed as presented in the report IAEA-EBP-WWER-09. 

Degradation of natural circulation during shutdown conditions, mainly in WWER reactors, should be accounted for adequately when estimating the shutdown risk in PSA studies. Detailed computer analyses of thermal and hydraulic processes should be carried out to show whether and at what time these processes could lead to core damage. 

So, at the beginning, the fairly important part of our work consisted in learning how to use these codes and in modifying them for the hardware available at the NRI. The next stage consisted in the preparation of input data for the NPPs with WWER reactors and QA of input data decks. 

Since all presented system codes have been developed and verified for western NPPs, their verification on Czech NPPs was an important stage. This verification was performed mainly on experimental facilities with parameters similar to those of WWER reactors, and on measurements performed during start-up of the NPPs. For this verification served, especially the ATHLET, CATHARE, RELAP, MELCOR and RALOC codes. We also participated in different computations of experiments organized under the auspices of OECD (ISP), EU (PHARE projects, 5 framework projects) and national project (sponsored Ministry of Trade and Industry). 

Nowadays in eight countries 49 commercial reactors of WWER-type are in operation with total power exceeding 32 GW, including 20 reactors WWER-1000 and 29 reactors WWER-440. More than 800 reactor-years of operation without serious incidents with radioactivity release outside NPP site have demonstrated high level of WWER type reactors safety. Economic operational indices of these reactors point out the competitiveness of WWER reactors as electric power producers. 

An overview of experience gained in long term applications of HRA in PSA projects for WWER reactors... 

Abstract This chapter describes requirements for speciality WWER reactor pressure vessel materials in terms of their chemical composition and mechanical properties. The main principles of manufacturing technology for WWER pressure vessel fabrication are also discussed, including welding and cladding. 

Key words WWER reactor pressure vessel, chemical composition, mechanical properties, welding, cladding. 

B.Timofeev, M. Brumovsky, and U. von Estorff, Certification Report oflSKhlMFA/ 15Cr2MoVA Steel and its Welds for WWER Reactor Pressure Vessels, IRC 57754, Petten, Netherlands EUR 24581 EN, Luxemburg, 2010. 

Abstract This chapter describes the embrittlement processes in WWER reactor pressure vessel (RPV) materials during operation - radiation embrittlement and thermal ageing. Current trend curves for both types of WWER RPV materials are given and explained. Surveillance specimen programmes are shown, as their results are used for RPV integrity and lifetime evaluation. Finally, anneahng of the RPV is proposed as the most efficient mitigation measure. 

Characteristics of embrittlement of WWER reactor pressure vessel (RPV) materials... 

WWER reactors have these additional requirements ... 

The following measures are planned or being implemented in WWER reactors ... 

Power supply by diesel generators is provided to safety systems that are necessary to cope with design basis accidents. In WWER reactors, the scope of systems with diesel backed power supply is limited in comparison with the common international practice, and does not cover many systems that would reduce the severe accident risk by facilitating management of anticipated incidents. 

In some WWER reactors, some of the primary instrumentation use a common tap to the component to which it is connected. These parameters may be used in important control systems or in the protection system. Failure of the common tap will cause failure of all instruments connected to it and may result in actuation or non-actuation of one channel of the protection system. It is a deviation from the Russian Safety rule "General safety regulations for NPPs (OPB-88)", GAN RF, Moscow, 1989. 

INTERNATIONAL ATOMIC ENERGY AGENCY, Anticipated Transients without Scram for WWER Reactors, IAEA-EPB-WWER-12, Vienna (1999). 

IAEA-TECDOC-749 Generic initiating events for PSA for WWER reactors, (1994). 

Requirements for the Content of the Safety Analysis Report for Nuclear Plants with WWER Reactors" ... 

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