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The warm pre-stress (WPS) of a flawed structure occurs when it is pre-loaded at high temperature in the ductile domain then cooled and loaded up to fracture in the brittle to ductile transition temperature domain. This load history is a feature of RPV accidental transients of LOCA type. Numerous tests on non irradiated specimens and structures have shown the favourable effect of WPS on fracture behaviour. Theorical knowledge let expect that the WPS effect occurs by the same way on irradiated material, but experimental approach had to be completed in such conditions. The experimental program presented in the present article consists in fracture toughness tests under WPS loading conditions performed on two RPV steels irradiated up to a fluence of 6,5.1019 n/cm2. The CT12.5 specimens used for these tests had been irradiated in the capsules of the pressure vessel surveillance program of two french reactors. Different types of WPS load history have been applied to cover typical accidental transients. All the results obtained confirmed for an irradiated steel the two assumptions generally made about the WPS effect: no fracture occurred during the cooling step of the loading even at high load level and the mean fracture toughness value is higher than that measured with conventional mono-temperature tests.

The EPR (initially European Pressurized water Reactor then Evolutionary Power Reactor and considered today as only a trademark) reactor pressure vessel (RPV) has been designed taking into account the requirement of an end-of-life (EOL) RTNDT lower than or equal to 30°C after 60 years of operation. The maximum acceptable fluence at the RPV inner wall, calculated with RCC-M formula and emerging from that requirement, is 3.3 × 1019n/cm2 (E > 1 MeV). To not exceed this fluence level, a heavy reflector has been introduced together with an increased water gap in the reactor downcomer. According to neutron calculations, the expected actual EOL fluence on EPR RPV in the presence of the heavy reflector is between 1 and 2.25 × 1019n/cm2 (E > 1 MeV) depending on the fuel management route. The irradiation surveillance program consists of positioning representative specimens of core region materials in the capsules attached to the outer face of the RPV core barrel. In former reactors equipped with a conventional core baffling, the neutron energy spectrum affecting the surveillance capsules is quite similar to the one affecting the RPV inner wall. In EPR, the heavy reflector presence distorts the neutron spectrum at the capsules’ location and, consequently, the EPR surveillance specimens used to monitor progress of the embrittlement in service would be subjected to a significantly different neutron energy spectrum from that on the RPV inner wall. The aim of this paper is to describe the approach followed by AREVA to limit this neutron spectrum effect and to design an appropriate capsule withdrawal schedule. This approach includes: (1) mitigation, with a new capsule basket design introducing an additional water gap between the reflector and the specimens, and (2) assessment of the embrittlement, not only with the conventional fast neutron fluence (E > 1 MeV), but also with the dose per atom as an additional dose damage parameter.

A significant extensive Research & Development work is conducted by Electricite´ de France (EDF) related to the structural integrity re-assessment of the French 900 and 1300 MWe reactor pressure vessels in order to increase their lifetime. Within the framework of this programme, numerous developments have been implemented or are in progress related to the methodology to assess flaws during a pressurized thermal shock (PTS) event. The paper contains three aspects: a short description of the specific French approach for RPV PTS assessment, a presentation of recent improvements on thermalhydraulic, materials and mechanical aspects, and finally an overview of the present R&D programme on thermalhydraulic, materials and mechanical aspects. Regarding the last aspect on present R&D programme, several projects in progress will be shortly described. This overview includes the redefinition of some significant thermalhydraulic transients based on some new three-dimensional CFD computations (focused at the present time on small break LOCA transient), the assessment of vessel materials properties, and the improvement of the RPV PTS structural integrity assessment including several themes such as warm pre-stress (WPS), crack arrest, constraint effect ....

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