WTEC panel report on European nuclear instrumentation and controls

Control and instrumentation systems might be called the 'brain' and 'senses' of a nuclear power plant. As such they become the key elements in the integrated operation of these plants. Recent developments in digital equipment have allowed a dramatic change in the design of these instrument and control (I&C) systems. New designs are evolving with cathode ray tube (CRT)-based control rooms, more automation, and better logical information for the human operators. As these new advanced systems are developed, various decisions must be made about the degree of automation and the human-to-machine interface. Different stages of the development of control automation and of advanced digital systems can be found in various countries. The purpose of this technology assessment is to make a comparative evaluation of the control and instrumentation systems that are being used for commercial nuclear power plants in Europe and the United States. This study is limited to pressurized water reactors (PWR's). Part of the evaluation includes comparisons with a previous similar study assessing Japanese technology

Developtment of a software tool for the analysis and verification of emergency operating procedures through the integrated simulation of plant and operators actions

Probabilistic safety assessment (PSA) includes operator actions as elements in the set of the considered protection performances during accident sequences. Nevertheless, its impact throughout a sequence is not usually analyzed dynamically. In this sense, it is convenient to make a more detailed analysis about its importance in the dynamics of the sequences, allowing for sensitivity studies with respect to human reliability and response times. For this reason, new developments in simulation software must be able to incorporate operator actions in conventional thermalhydraulic simulations. In this paper, we present one of these new tools, the TRETA/TIZONA–COPMA III coupled codes, which can be used for evaluating the impact in the final plant state of the execution by operators of procedures and the evaluation of the available times for the manual actions of the operators. This software tool consists of a closed-loop plant/operator simulator: a thermalhydraulic code for simulating the plant transient (TRETA for PWR NPPs and TIZONA for BWR NPPs) and the procedures processor (COPMA III) to simulate the operator actions requested by the procedures, both coupled by a data communication system which allows the information exchange (SWBus). The first pilot cases have been performed in order to analyze sequences initiated by secondary side breaks leading to loss of heat sink sequences in a PWR plant. These tests have been carried out using the real plant EOPs for COPMA-III and a PWR plant model for TRETA code. The results of these simulations are presented in this paper

Functional Modelling for Fault Diagnosis and its application for NPP.

The paper presents functional modelling and its application for diagnosis in nuclear power plants. Functional modelling is defined and its relevance for coping with the complexity of diagnosis in large scale systems like nuclear plants is explained. The diagnosis task is analyzed and it is demonstrated that the levels of abstraction in models for diagnosis must reflect plant knowledge about goals and functions which is represented in functional modelling. Multilevel flow modelling (MFM), which is a method for functional modelling, is introduced briefly and illustrated with a cooling system example. The use of MFM for reasoning about causes and consequences is explained in detail and demonstrated using the reasoning tool, the MFMSuite. MFM applications in nuclear power systems are described by two examples: a PWR; and an FBR reactor. The PWR example show how MFM can be used to model and reason about operating modes. The FBR example illustrates how the modelling development effort can be managed by proper strategies including decomposition and reuse

Review of current Severe Accident Management (SAM) approaches for Nuclear Power Plants in Europe

The Fukushima accidents highlighted that both the in-depth understanding of such sequences and the development or improvement of adequate Severe Accident Management (SAM) measures are essential in order to further increase the safety of the nuclear power plants operated in Europe. To support this effort, the CESAM (Code for European Severe Accident Management) R&D project, coordinated by GRS, started in April 2013 for 4 years in the 7th EC Framework Programme of research and development of the European Commission. It gathers 18 partners from 12 countries: IRSN, AREVA NP SAS and EDF (France), GRS, KIT, USTUTT and RUB (Germany), CIEMAT (Spain), ENEA (Italy), VUJE and IVS (Slovakia), LEI (Lithuania), NUBIKI (Hungary), INRNE (Bulgaria), JSI (Slovenia), VTT (Finland), PSI (Switzerland), BARC (India) plus the European Commission Joint Research Center (JRC). The CESAM project focuses on the improvement of the ASTEC (Accident Source Term Evaluation Code) computer code. ASTEC,, jointly developed by IRSN and GRS, is considered as the European reference code since it capitalizes knowledge from the European R&D on the domain. The project aims at its enhancement and extension for use in severe accident management (SAM) analysis of the nuclear power plants (NPP) of Generation II-III presently under operation or foreseen in near future in Europe, spent fuel pools included. In the frame of the CESAM project one of the tasks consisted in the preparation of a report providing an overview of the Severe Accident Management (SAM) approaches in European Nuclear Power Plants to serve as a basis for further ASTEC improvements. This report draws on the experience in several countries from introducing SAMGs and on substantial information that has become available within the EU “stress test”. To disseminate this information to a broader audience, the initial CESAM report has been revised to include only public available information. This work has been done with the agreement and in collaboration with all the CESAM project partners. The result of this work is presented here.JRC.F.5-Nuclear Reactor Safety Assessmen

Risk Informed Support of Decision Making in Nuclear Power Plant Emergency Zoning - Generic Framework towards Harmonising NPP Emergency Planning Practices

The report provides a systematic overview of the essential aspects of risk informed support of decision making (RIDM) in nuclear power plant (NPP) emergency zoning (EZ) as a contribution to harmonising strategic practices in the area. Owing to the state-of-the-art understanding and increased characterisation of NPP severe accidents, overall management of them should be analysed as an integrated complex process. The interrelationship of NPP emergency operating procedures, safety and risk assessments, severe accident management guidelines, and emergency off-site actions should be planned and organised to minimize the consequences of such accidents. A deterministic approach, coupled with both probabilistic safety assessment (PSA) technology and PSA results can play significant role in the development of relevant nuclear utility, regulatory and all stakeholders policies. The report describes the background, objectives and current state of a corresponding activity within JRC-IE's Analysis and Management of Nuclear Accidents (AMA) Action on probabilistic safety / risk assessment methodologies and practices for RIDM approach applied to NPP EZ. The approach is interdisciplinary, based on integration of PSA technology, severe accident phenomenology, and radiological protection.JRC.F.4-Nuclear design safet