European Pressurized Water Reactor Research Articles

Franco-German nuclear cooperation: from the ‘common product’ to the first European pressurized water reactor

Franco-German nuclear cooperation: from the ‘common product’ to the first European pressurized water reactor

The European pressurized water reactor Result of the French–German cooperation of experienced NPP suppliers and operators

In 1989 Framatome and Siemens, the two most experienced European nuclear power plant suppliers, decided to join the efforts for the development of a new reactor type for the next generation in their equally owned subsidiary Nuclear Power International (NPI). In 1992 Electricité de France and the major German utilities operating nuclear power plants merged their own development programs with that of Nuclear Power International and initiated the European Pressurized Water Reactor (EPR) project. In order to reach the two major targets of the project, the licensability in both countries, France and Germany, and the competitiveness of nuclear energy with other alternative energy sources, the design basis which had differently developed in the two countries needed to be harmonized. In parallel, the licensing authorities of both countries extended their existing cooperation in the field of a safety survey of existing nuclear power plants to the definition of safety criteria for the next generation of nuclear power plants. Through this cooperation the licensability of EPR in France and Germany will be assured. Continuously performed cost analysis show in addition that also the second target of the project, the competitiveness with alternative primary energy sources, can be achieved. Thanks to the fruitful cooperation between all parties involved, satisfactory results have been achieved not by a simple superposition of existing design features but through a careful evaluation and combination of the best available alternatives. At the end of 1997 the basic design results were compiled in a final report. Subsequently an optimization phase was launched that further improves the competitiveness of the power generation costs.

Systems organization for the European pressurized water reactor (EPR)

Systems organization for the European pressurized water reactor (EPR)

The main components of the European pressurized water reactor

The main components of the European pressurized water reactor

Emergency Core Cooling Mode of the European Pressurized Water Reactor: The Evolution from Combined Injection to Cold-Leg Injection

The European Pressurized Water Reactor’s (EPR’s) safety injection system (SIS) comprises four trains, each of them consisting of a medium head safety injection, an accumulator, and a low head safety injection (LHSI). Injection mode is into the cold legs of the main coolant line for the short term. This emergency core cooling (ECC) mode is quite different from the typical German concept with combined injection, i.e., safety injection into both the cold and the hot legs of the main coolant line at least for accumulator and LHSI. Therefore, the German Safety Authority requested justification for giving up the ECC-mode used in German pressurized water reactors, the so-called “combined injection.” Furthermore, the Reaktor-sicherheitskommission requested a comparison between cold-leg injection and combined injection in terms of ECC efficiency over all relevant accident sequences.The evolution from combined injection to cold-leg injection is described and results of comparative analysis are summarized. It is demonstrated that EPR’s SIS is a well-balanced system, which ensures high ECC efficiency and limits loads to containment over the whole accident spectrum. For the entire loss-of-coolant accident (LOCA) spectrum, ECC efficiency of EPR’s SIS is practically equivalent to ECC efficiency of a SIS of the KONVOI type with combined injection. The smaller the break, the more insignificant are differences. The ECC mode has a negligible impact on containment pressure and temperature evolution during a LOCA. Neither with combined injection nor with cold-leg injection is a containment spray system needed.

Discussion on recent concrete containment designs

After a short description of some containment designs of existing power plants in France and Germany this paper addresses, from a civil engineering point of view, current investigations being jointly performed by EDF and SIEMENS within the framework of the basic design for the future European Pressurized Water Reactor (EPR) containment. The prestressed containment will be tested at its design pressure of 6.5 bar abs and a design leak rate below 1% of the containment volume in 24 h, covering a LOCA pressure of 4.2 bar abs with 12 h grace period for spraying and a hydrogen deflagration accident. For this project the Safety Authorities require in addition the verification of leak tightness beyond the design pressure of 6.5 bar abs in order to mitigate the consequences of all possible severe accidents. Although the design of the prestressed containment is such, that without a liner the containment function will meet leak rate requirements with respect to pressures and temperatures beyond the test conditions, investigations for a composite liner glued onto the inner side of the containment wall are being performed to gain further margins. Based on pre-tests already performed at the SIEMENS laboratories and at the University of Karlsruhe the most promising materials are vinyl esters (VE) or epoxy resins reinforced by glass-fibres, so called glass-fibre-reinforced plastics (GRP). A qualification program considering the specific aspects of nuclear licensing for such materials in France and Germany is under way. Of special interest is their behaviour under high temperatures, irradiation and extensive cracking of the containment wall during postulated severe accident conditions. Intermediate results will be presented. Some information concerning the large scale mock-up being erecting under the leadership of EDF at the site of CIVAUX/France will be given. The aim of these tests is to estimate the ultimate leak rate of a containment wall without a liner for dry air (pressure test) and an air–steam mix (accident) and to test the behaviour of a composite liner up to the failure of the containment wall.

Seismic design of nuclear facilities in Germany

Seismic design forms an integral part of the overall safety concept of German nuclear power plants, despite the generally low seismic activity in Germany. Even the more active zones in Germany represent regions with low seismic hazards. The development of specific nuclear standards in Germany regarding the seismic design of nuclear power plants is outlined. The first specific nuclear standard was issued in 1975, prescribing two design earthquake levels. It was revised in 1990. Today a single design basis earthquake is considered for which the plant safety has to be demonstrated by dynamic analyses on the basis of a site-specific seismic assessment. The seismic safety of older plants that had not already been constructed according to a specific nuclear standard regarding the seismic design, have been re-evaluated recently. This resulted in backfitting measures such as the installation of bunkered emergency systems at many older plants. Furthermore, the seismic safety of all nuclear power plants is now assessed within the framework of periodic safety reviews. For future reactors like the European pressurized water reactor (EPR), the design provisions will have to ensure that external hazards will not constitute a significant part of the residual risk of the plant. A ground level acceleration value of 0.25 g is adopted for the design, which should provide ample safety margins for all relevant sites in France and Germany.

Objectives in developing the European pressurized water reactor (EPR)

In 1991, major German utilities and Electricité de France agreed to develop, together with Siemens and Framatome, the nuclear island for the next generation of nuclear power plants. This nuclear island design is based on German and French experience in the construction and operation of pressurized water reactors. The major step in the evolutionary European pressurized water reactor (EPR) design is the systematic inclusion of events beyond classic design events. The mitigation of core melt accidents by special means, such as reinforcement of the containment function, primary loop depressurization, hydrogen reduction and ultimate heat removal, is part of the design, in addition to a number of features which increase the reliability of plant operation and accident control. Nevertheless, utilities and designers are aware of the economic challenge facing the EPR by the need to compete with other nuclear power plant designs and fossil plants for electricity generation.

Large spreading of core melt for melt retention—stabilization

For future nuclear power plants possible core melt scenarios have to be taken into account in the plant layout. Different solutions for core melt retention are proposed in the literature. The core melt retention devices should as far as possible be simple in construction, to minimize the deepness of sophistication of the physical problems connected with it and to minimize unforeseen technological problems. Furthermore it should be in good compliance with normal operational needs. One very promising solution seems to be a large enough spreading of the melt with water cooling from above. This is the favoured solution for the development of the French-German European pressurized water reactor. The technical concept and the problems to be expected in the course of an anticipated core melt accident (to be overcome in an actual design) are shown. The needs for analytical and experimental support in the early conceptual phase and later on in the design stage of plant construction are discussed. Typical analytical and first experimental results for the large core melt spreading are presented.

Milestones in screen-based process control / Meilensteine der bildschirmgestützten Prozeßführung.

The paper outlines the German and the French approach to screen-based process control in nuclear power plants, both based on stepwise collection of experience. The German approach is based on the utilisation of the conceptual elements of the PRISCA information system developed by Siemens and on operational experience with screen-based process control in a conventional power plant. In the French approach, the screen-based control room for the N4 plants, designed from scratch, has undergone extensive simulator tests for validation before going into realisation. It is now used in the commissioning phase of the first N4 plants. The design of the control room for the European Pressurized Water Reactor will be based on the common experience of Siemens and Electricite de France. Its main elements are several separate operator workstations, a safety control area used as a back-up for postulated failures of the workstations, and a commonly utilisable plant overview for the operators' coordination.

The European Pressurized Water Reactor (EPR): an advanced PWR

Nuclear Power International, together with its Parent Companies Framatome and Siemens and with the support of Electricite de France and a group of 9 major German Utilities, is developing the EPR, an advanced evolutionary nuclear island. The major objective of this development is to further improve safety, reliability and competitiveness of nuclear energy in order to contribute to the Worldwide increasing need for energy.

Towards an enhanced quality in pressurized water reactor safety / Zu einer höheren Qualität der Sicherheit von Druckwasserreaktoren

An improved safety approach for future light water reactors has been derived from the reinforced defense-in-depth concept. It comprises measures of severe accident prevention and measures of mitigation of accident consequences. Severe accident phenomena will be considered already in the stage of the plant's design. The overall objective is that even in the case of a core melt accident there will be no necessity for off-site emergency measures, such as evacuation or relocation of the population from the vicinity of the damaged plant, and that large-scale land contamination can be excluded. This new approach is being followed in the development of the European Pressurized Water Reactor (APR)

Safety of future reactors in France

This document describes current thinking within the Institute for Protection and Nuclear Safety (IPSN) regarding developments desirable from the viewpoint of safety in a new generation of nuclear power plants, the construction of which could begin in France at the end of the nineties. Significant improvements must be sought, particularly as regards the containment, including the core meltdown case. The operating organization, Electricité de France (EDF), is now considering establishing the basis of a project of European dimension (European Pressurized water Reactor - EPR), in connection with German utilities and the Franco-German vendor Nuclear Power International (NPI); the first options important for safety are to be presented and discussed from September-1993.

The European Pressurized Water Reactor / Der Europäische Druckwasserreaktor.

The present state of development of the European Pressurized Water Reactor (EPR) is outlined. During the so-called harmonization phase, the French and German utilities drew up their common requirements and evaluated the reactor concept developed until then with respect to these requirements. A main result of the harmonization phase was the issue, in September 1993, of the «EPR Conceptual Safety Feature Review File» to be jointly assessed by the safety authorities in France and Germany. The safety objectives to be met by the EPR are specified in the second part of the paper, and some details of the primary and secondary side safety systems are given

14C release from a Soviet-designed pressurized water reactor nuclear power plant.

The Paks Nuclear Power Plant in Hungary runs with four pressurized water reactors, each of 440-MWe capacity. Sampling systems have been developed and used to determine the 14C of various chemical forms (14CO2, 14CO, 14CnHm) in the airborne releases. The average normalized yearly discharge rates for the time period 1988-1991 are equal to 0.77 TBq GWe-1 y-1 for hydrocarbons and 0.05 TBq GWe-1 y-1 for CO2. The contribution of 14CO was less than 0.5% of the total emission. The 14C discharge rate is estimated to be four times higher than the corresponding mean data of Western European pressurized water reactors. The calculated effective dose equivalent to individuals living in the vicinity of the power plant, due to 14C release, was 0.64 microSv in 1991 while the effective dose equivalent due to the natural 14C level was 15 microSv y-1. The long-term global impact of the 14C release in the operational period of the plant (1982-1991) was 1,270 man-Sv. The 14C excess in the environmental air has been measured since 1989 by taking biweekly samples at a distance of 1.7 km from the nuclear power plant. The long-term average of radiocarbon excess coming from the power plant was 2 mBq m-3. The local 14C deposition was followed by tree ring analysis, too. No 14C increase higher than the uncertainty of the measurement (four per thousand = 0.17 mBq m-3) was observed.