European Pressurized Reactor Research Articles

Cost of Electricity from the Jaitapur Nuclear Power Plant

The Indian Government has announced that it plans to purchase six European Pressurized Reactors (EPRs) for Jaitapur from the French company, Areva. No EPR is in commercial operation anywhere else in the world. Estimates of costs from plants under construction in Finland and France suggest that each unit may cost as much as Rs. 60,000 crores; at this price, six units will cost Rs. 3.6 lakh crores. We show that the expected starting tariff from these reactors, without including transmission and distribution costs, is likely to be around Rs. 14 per unit (kWh) of electricity. We point out that existing revenue model used by the Government already involves a large loss for the taxpayer. The Government may seek to make the tariff from Jaitapur competitive by increasing the scope and nature of these handouts.

Looking back to see the future: building nuclear power plants in Europe

The so-called ‘nuclear renaissance’ in Europe is promulgated by the execution of two large engineering projects involving the construction of two European Pressurized Reactors (EPRs) in Flamanville, France and Olkiluoto in Finland. As both projects have faced budget overruns and delays, this paper analyses their governance and history to derive lessons useful for the construction of future projects. Analysis indicates that the reasons for these poor outcomes are: overoptimistic estimations, first-of-a-kind (FOAK) issues and undervaluation of regulation requirements. These pitfalls have the potential to impact on many other engineering construction projects and highlight fruitful areas of further research into project performance.

Life cycle assessment of the European pressurized reactor and the influence of different fuel cycle strategies

This article presents a life cycle assessment (LCA) of electricity generated by the European pressurized reactor (EPR), a Generation III (Gen III) nuclear reactor concept. The focus of the LCA is on the potential locations in Switzerland and France in the year 2030. Environmental burdens from the EPR are also compared to those from current pressurized water reactor (PWR) technology in Switzerland. The entire fuel cycle is analysed and includes improved modelling of natural uranium mining. The indicators employed in the impact assessment characterize a broad range of burdens to the environment and human health at both the global and the regional/local level and includes an analysis on the basis of external costs. The LCA shows a reduction of greenhouse gas emissions by around just 7 per cent for the EPR in Switzerland in 2030 compared to the current PWR cycle emitting just above 5 g CO2-eq/kWh. Due to differences in cooling strategies, which cause slightly different output capacities, the French EPR is expected to achieve marginally better reductions of almost 16 per cent. The processing of natural uranium to yellowcake was shown to be the stage in the nuclear fuel cycle having the highest burdens and potential impacts across a majority of indicators, accounting for between 25 per cent and 60 per cent of burdens and potential impacts. External costs of normal operation are shown to be around just 2 per cent of estimated internal (production) costs and similar to those of hydro and offshore wind. The potential impacts due to nuclear-related accidents were not considered.

The contribution of ergonomics to risk analysis in the design process: the case of a future control room

The aim of this paper is to highlight how ergonomics contributes to risk analysis and risk management in a design project for a new reactor, the French EPR (European Pressurized Reactor). An iterative ergonomics design process has been conducted over the last 10 years through a Human Factors Engineering program at the French energy company EDF. A parallel has been drawn between a risk management process and this ergonomics process based on International Organization for Standardization (ISO) publications such as ISO/IEC Guide 73, ISO Guide 51, etc. The contribution of ergonomics to risk analysis is illustrated by an example: the automatic diagnosis, which is a very important technical device for safety. Five main types of risk have been identified via ergonomic analyses during the different design stages. Counter-measures have been implemented and their efficacy tested within the scope of new campaigns to assess human factors. However, the management of risks in such a design project requires the participation both of the different design entities involved in the project and of other experts in aspects of risk management, such as human reliability. The organization of collaborative participation remains a challenge to be addressed.

An Analysis of Reflector Homogenization Techniques for Full Core Diffusion Calculations

Today’s reactor core calculations are done in diffusion with a few coarse groups and require the homogenization of the core assemblies as well as a correct representation of the reflector. In industrial applications a homogeneous reflector is often used with cross sections obtained from transport calculations and adjusted to fit in-core measurements. However, the need for better precision in the core diffusion calculations and the emergence of new reflector concepts, such as for the European Pressurized Reactor (EPR), require an increase in the number of coarse groups for novel loading patterns and a rethinking of how to define the equivalent reflector. In this work we analyze and extend current techniques for the reflector homogenization for core calculations. Following the adopted industrial methodology, we have perfected a technique for the determination of an equivalent homogenous reflector by implementing a Particle Swarm Optimization Algorithm and showed its limitations through the analysis of an academic slab reactor model and of a realistic two-dimensional representation of the EPR. We have compared the precision of the resulting core calculations to transport reference calculations as well as to diffusion calculations using a multigroup albedo boundary condition. We have also explored the use of current-preserving flux discontinuity coefficients at the core-reflector interface in conjunction with an equivalent reflector.

PC-based simulator for education in advanced nuclear power plant construction

The PC-based reactor simulation software PCTRAN was recently expanded to cover light water advanced reactors. The plant models includes Generation III+ advanced Pressurised Water Reactor (PWR) and Boiling Water Reactor (BWR). The evolutionary designs are Areva European Pressurized Reactor (EPR) and GE Advanced Boiling Water Reactor (ABWR), which includes cooling path redundancy and devices for severe accident mitigation. One step further are the passive-cooled Westinghouse AP1000 and GE Economical and Simplified BWR (ESBWR). Combined with PCTRAN's models of an experimental pool reactor and conventional PWR and BWR with server accident and dose dispersion capability, they form a complete series for education in nuclear technology. The PC-based simulation system helps to train the first generation of technical staff for countries entering the nuclear era.

The ASTEC Integral Code for Severe Accident Simulation

For several years the French Institut de Radioprotection et de Sûreté Nucléire (IRSN) and the German Gesellschaft für Anlagen und Reaktorsicherheit (GRS) mbH have been jointly developing a system of calculation codes—the integral Accident Source Term Evaluation Code (ASTEC)—to simulate the complete scenario of a hypothetical severe accident in a nuclear light water reactor, from the initial event until the possible radiological release of fission products out of the containment, i.e., the source term. ASTEC has progressively reached a larger European dimension through projects of the European Commission Framework Programme. In particular, in the frame of the European Severe Accident Research NETwork of Excellence (SARNET), jointly executed research activities were performed with the ultimate objectives of providing physical models for integration into ASTEC and making the code the European reference. This effort will go on in the frame of the SARNET2 next network. The ASTEC models are today at the state of the art, except for reflooding of a degraded core. Many applications have been performed by IRSN for significant safety studies, including the probabilistic safety analysis level 2 on a French pressurized water reactor. The first version V2.0 of the new ASTEC series, released in spring 2009, will allow simulation of the European Pressurized Reactor (EPR) and will include advanced core degradation models. Then, ASTEC will remain the repository of knowledge gained from international research and development. Other long-term objectives are on one hand extension of the scope of application to boiling water reactors and CANada Deuterium Uranium (CANDU) reactors, to accidents in the ITER Fusion facility, and to Very High Temperature Reactor (VHTR) Generation IV reactors, and on the other hand to the use for emergency response tools and for severe accident simulators.

Hydrogen production by high temperature electrolysis coupled with an EPR, SFR or HTR: techno-economic study and coupling possibilities

Hydrogen production by high temperature electrolysis coupled with three nuclear reactors (the European pressurised reactor, the sodium-cooled fast reactor and the very high temperature reactor) was studied in terms of perspectives and hydrogen production costs. Firstly, we present the features of producing water steam by using the three nuclear reactors. Secondly, we present the hydrogen production cost for the HTE process coupled with each type of nuclear reactor. These costs are optimal values of the hydrogen production cost for the mentioned couplings and they were estimated by using a genetic algorithm procedure. High potentiality for these HTE couplings was assessed and contrary to steam source temperatures, the electricity price appeared to be a key parameter for low hydrogen production costs.

The TONUS CFD code for hydrogen risk analysis: Physical models, numerical schemes and validation matrix

The French Atomic Energy Commission (CEA) and the Institute for Radiological Protection and Nuclear Safety (IRSN) are developing a hydrogen risk analysis code, called the TONUS code, which incorporates both lumped-parameter (LP) and computational fluid dynamics (CFD) formulations. In this paper we present the governing equations, numerical strategy and schemes used for the CFD approach. Several benchmark exercises based on experimental results obtained on large-scale facilities, such as MISTRA, TOSQAN and RUT, are presented. They have been used as verification and assessment procedures for modelling and numerical approaches of the code. Specific emphasis is given to the sensitivity analysis of the computed results with respect to numerical and physical parameters. The powerful Design-Of-Experiments technique for sensitivity analysis is successfully applied to the ISP47 MISTRA test case. The TONUS CFD code is presently used to support the hydrogen risk assessment for the European Pressurized Reactor (EPR) plant and to investigate the impact of the two-room concept on hydrogen distribution in the EPR containment.

CCF analysis for new reactor designs

Abstract Common Cause Failures (CCF) have been an issue of safety consideration in nuclear safety since a long time. For the design of the European Pressurised Reactor (EPR) probabilistic safety goals in conjunction with the limit set by CCF on failure probability that can be achieved by redundancy led to the requirement to fulfil vital safety functions by functionally diverse systems. The probabilistic safety analysis (PSA) which was performed for EPR takes into account CCFs which are modelled using the beta-factor model of the European Utility Requirements (EUR). The PSA shows that the probabilistic safety goal for core damage frequency – CDF < 10−5/year – is met with sufficient margin. In addition to the modelling of CCF in the PSA a deterministic CCF analysis is being performed. It is a safety function based approach on component level analysing frontline and backup systems with respect to common components or type identical components. It ensures that diverse safety functions are fulfilled by systems being sufficiently diverse including all necessary support systems. Components are identified where further diversification or additional defences against CCF are required.

The severe accident mitigation concept and the design measures for core melt retention of the European Pressurized Reactor (EPR)

For the mitigation of severe accidents, the European Pressurized Water Reactor (EPR) has adopted and improved the defense-in-depth approaches of its predecessors, the French “N4” and the German “Konvoi” plants. Beyond the corresponding evolutionary changes, the EPR includes a new, 4th level of defense-in-depth that is aimed at limiting the consequences of a postulated severe accident with core melting. It involves a strengthening of the confinement function and the avoidance of large early releases. The latter requires the prevention of scenarios and events that can result in high loads on the containment, e.g., a failure of the Reactor Pressure Vessel (RPV) at high internal pressure. This is achieved by dedicated design measures. The paper gives an short overview of the general concept and the strategies for: primary circuit depressurization, H 2 mitigation and the avoidance of energetic Fuel Coolant Interactions (FCIs). It then describes, in detail, the conceptual solution for the stabilization and long-term cooling of the molten core. The EPR melt retention strategy supports itself on the use of an ex-vessel core catcher located in a compartment lateral to the pit. The related spatial and functional separation isolates the core catcher from the various loads during RPV failure and, at the same time, avoids risks resulting from an unintended initiation of the system during power operation. Within the core catcher, the melt will be passively flooded with water from the Internal Refueling Water Storage Tank (IRWST). Due to the effective cooling of the melt from all sides a stable state will be reached within hours and complete solidification of the melt is achieved after a few days. The core catcher can optionally be supplied by the Containment Heat Removal System (CHRS). In this active mode of operation, the water levels inside spreading compartment and reactor pit rise and the pools become subcooled, so further steaming is avoided. This results in a depressurization of the containment in the long-term.

Thermomechanical Stability of a Zirconia-Based Refractory Material

Results for ZrO2-based refractory materials tested for contact with molten iron, thermal cycling, and single-time thermal impact in an electric furnace are presented. The test results are compared with calculated thermal stresses for the protective layer of an EPR (European Pressurized Reactor) core melt catcher simulating the spill of the molten corium. Potential damage of the refractory material associated with the development of thermal stresses exceeding the critical stress value is discussed.

ICONE11-36196 Severe accident mitigation and core melt retention in the European Pressurized Reactor (EPR)

ICONE11-36196 Severe accident mitigation and core melt retention in the European Pressurized Reactor (EPR)