The VVER-440/213 reactor pressure vessel

The design of the VVER-440 RPV is rather specific: the relatively small RPV diameter has to allow its transportation on rails. As a consequence of its limited diameter, the water gap between the RPV and the core is small, so the fast neutron flux (E>0.5 MeV) on the RPV is rather high at 1015 m-2s-1 and the RPV base material should therefore be more resistant to irradia­tion embrittlement. The RPV is assembled from forged rings without lon­gitudinal welds. The coolant from the low-pressure emergency core cooling systems and hydro-accumulators is directly injected into the RPV, and from the high-pressure system into the cold leg of the loops. The inlet and outlet nozzles of the loops are separated on different levels. The penetrations for the instrumentation for core control are on the RPV head.

The ferritic steel reactor pressure vessel is clad internally with austen­itic stainless steel. The RPVs are made from low alloy steel (15Cr2MVA; at Loviisa NPP 12Cr2MFA) and the circumferential submerged arc weld­ing was made using Sv-10CrMoVTi wire. The RPV was covered inter­nally by a welded clad of two stainless steel layers. The inner layer is a non-stabilized stainless steel (Sv-07Cr25Ni13, similar to AISI 309) and that, when in contact with the coolant, is a niobium stabilized stainless steel (Sv-08Cr19Ni10Mn2Nb; Sv-07Cr19Ni10Nb at Loviisa; both equivalent to

AISI 347). Components of the primary circuit in contact with the primary coolant, other than the RPV, are also made of austenitic stainless steel, that is the piping of the primary loop, the main circulating pumps, gate valves and the emergency and auxiliary systems pipework.

From the point of view of longer-term operation, the main deficiency of VVER-440/230 was the high irradiation exposure of the reactor pressure vessel wall by fast neutrons, and the relatively quick embrittlement of the RPV material. The issue had been aggravated by the lack of a proper RPV surveillance programme at these plants. Several attempts have been made to assess the embrittlement of the base and weld material of those RPVs. For the first generation RPVs, essential data for RPV materials were absent, for example transition temperature, concentration of copper and phosphorus; the archive metal of the RPVs was not available. The phosphorus and cop­per contents in the welds of VVER-440/230 are in the range 0.030-0.048% and 0.10-0.18%, respectively. In the case of VVER-440/213, the same con­centrations are in the range 0.010-0.028% for P and 0.03-0.18% for Cu (Brumovsky et al. 2005; Vasiliev and Kopiev, 2007).

Reactor pressure vessel surveillance programmes became obligatory in all VVER plants that had been commissioned after Units 1 and 2 at Loviisa. Proper RPV surveillance programmes have been implemented at VVER-440/213 plants outside of the former Soviet Union from the com­mencement of plant operation.

An ‘ Extended Surveillance Specimen Programme ’ was prepared with the objective of validating the results of the standard programme (Kupca, 2006). It aimed to increase the accuracy of the neutron fluence measure­ment, make a substantial improvement in the determination of the actual temperature of irradiation, fix the orientation of RPV samples to the centre of the reactor core, minimize the differences in neutron dose between the Charpy-V notch and crack-opening-displacement specimens and evaluate any dose-rate effects. For Units 1 and 2 of the Mochovce NPP, a completely new surveillance programme was prepared, based on the philosophy that the results of the programme must be available during the entire service life of the NPP. The new, advanced surveillance programme deals with the irradiation embrittlement of both the weld area heat affected zone and the austenitic stainless steel cladding of the RPV, which were not previously evaluated in surveillance programmes.

Several measures were implemented for the resolution of the RPV embrittlement issue: [11]

• Decreasing the stressors, for example, heating up the water in the emer­gency core cooling system (ECCS) to lessen thermal shock in a pressur­ized thermal shock (PTS) situation; steam-line isolation; system solu­tions interlocks.

• Introduction of volumetric non-destructive testing for in-service inspection.

Annealing of RPV has been implemented at Loviisa NPP and Kola NPP (also at the shut down plant Bochunice V1). Annealing in the case of the VVER-440 reactor vessel weld was performed at a temperature of 475±15°C and the holding time was 150 hours. Assessment of annealing effectiveness (level of properties recovering after annealing), determination of re-irradiation re-embrittlement rates after annealing, and the behaviour of VVER-440 weld materials, showed the real possibility of recovering RPV toughness properties of irradiated VVER-440 RPV materials. Measures were also taken to improve the knowledge of the vessel material by ves­sel sampling. A more detailed description of the RPV neutron irradiation embrittlement issue is provided by Erak et al. (2007), for example. Based on the results of the VVER-440/213 plants, annealing of the RPV has been implemented at Rivne NPP.

In order to determine the time limit of operation of the RPV, it is neces­sary to consider and analyse the neutron irradiation damage, thermal age­ing and low-cycle fatigue in decreasing the fracture toughness of the RPV materials.

Pressurized thermal shock (PTS) is the most critical lifetime lim­iting event for the RPV. Since the PTS screening requirement (pressure-temperature-loading limits) is the lifetime limiting process for the RPV of VVERs, the methodology of PTS evaluation has to be established in the national regulations. This will take into account the applicable best practices, features of the RPV and the thermal-hydraulic peculiarities of the VVERs. The assumptions of renewed PTS analyses have been confirmed with mixing tests. International research projects supported the effort of VVER plants in the evaluation of PTS for the RPV (IAEA, 2005).

The results of the PTS calculations, based on the analysis of postulated embedded flaws, endorse the possibility of 50 years of operation for all of the units, without annealing of the 5/6 welds. At the Paks NPP, the assump­tion of the embedded postulated crack (under-cladding semi elliptical type) was justified by the results of qualified in-service inspections, which followed the procedure of European Network for Inspection Qualification (ENIQ). Two types of inspection were applied to the full cladding area: (1) ultra­sonic inspections from the inner surface and (2) Eddy current inspection, overlapping the first 5 mm thickness of the RPV inner-wall area. There is

no generic need for the heating up of the emergency core cooling water. It was introduced as an example at Rivne NPP in Ukraine, however it seems unnecessary at Paks NPP in Hungary.

The critical locations when considering fatigue are the welds of the inner tubes of the control rod drive nozzles.