NPP Design Research Articles

RELIABILITY ANALYSIS OF DIGITAL SYSTEMS IN A PROBABILISTIC RISK ANALYSIS FOR NUCLEAR POWER PLANTS

To assess the risk of nuclear power plant operation and to determine the risk impact of digital systems, there is a need to quantitatively assess the reliability of the digital systems in a justifiable manner. The Probabilistic Risk Analysis (PRA) is a tool which can reveal shortcomings of the NPP design in general and PRA analysts have not had sufficient guiding principles in modelling particular digital components malfunctions. Currently digital I&C systems are mostly analyzed simply and conventionally in PRA, based on failure mode and effects analysis and fault tree modelling. More dynamic approaches are still in the trial stage and can be difficult to apply in full scale PRA-models. As basic events CPU failures, application software failures and common cause failures (CCF) between identical components are modelled.The primary goal is to model dependencies. However, it is not clear which failure modes or system parts CCF:s should be postulated for. A clear distinction can be made between the treatment of protection and control systems. There is a general consensus that protection systems shall be included in PRA, while control systems can be treated in a limited manner. OECD/NEA CSNI Working Group on Risk Assessment (WGRisk) has set up a task group, called DIGREL, to develop taxonomy of failure modes of digital components for the purposes of PRA. The taxonomy is aimed to be the basis of future modelling and quantification efforts. It will also help to define a structure for data collection and to review PRA studies.

Irradiation embrittlement of reactor pressure vessel steel at very high neutron fluence

For the prediction of radiation embrittlement of RPV materials beyond the NPP design time the analysis of research data and extended surveillance data up to a fluence ∼23×1020cm−2 (E>0.5MeV) has been carried out. The experimental data used for the analysis are extracted from the International Database of RPV materials. Key irradiation embrittlement mechanisms, direct matrix damage, precipitation and element segregation have been considered. The essential part of the analysis concerns the assessment of irradiation embrittlement of WWER-440 steel irradiated with very high neutron fluence. The analysis of several surveillance sets irradiated at a fluence up to 23×1020cm−2 (E>0.5MeV) has been performed. The effect of the main influencing chemical elements phosphorus and copper has been verified up to a fluence of 4.6×1020cm−2 (E>0.5MeV). The data are indicating good radiation stability, in terms of the Charpy transition temperature shift and yield strength increase for steels with relatively low concentrations of copper and phosphorus. The linear dependence between ΔTk and ΔRp0.2 can be an evidence of strengthening mechanisms of irradiation embrittlement and absence of non-hardening embrittlement even at very high neutron fluence.

Trace Code Validation for BWR Spray Cooling Injection and CCFL Condition Based on GÖTA Facility Experiments

Best estimate codes have been used in the past thirty years for the design, licensing, and safety of NPP. Nevertheless, large efforts are necessary for the qualification and the assessment of such codes. The aim of this work is to study the main phenomena involved in the emergency spray cooling injection in a Swedish-designed BWR. For this purpose, data from the Swedish separate effect test facility GÖTA have been simulated using TRACE version 5.0 Patch 2. Furthermore, uncertainty calculations have been performed with the propagation of input errors method, and the identification of the input parameters that mostly influence the peak cladding temperature has been performed.

Contribution of European Nuclear Licensees toward Harmonisation of Regulations for Nuclear Installations

Liberalisation of the European energy market has led to the deregulation of electricity generation and supply. More and more nuclear industry and utilities evolve in an international framework. Diversity of national regulations can seriously distort competition. Undoubtedly, harmonizing regulations is the best way of ensuring that the industry can evolve within a stable legal framework. Consequently, European nuclear license holders supported the work of the Western European Nuclear Regulators Association (WENRA) on the harmonization of European safety standards for existing nuclear power plants, as well as for waste and decommissioning.This support led to the creation, within FORATOM, of the ENISS (European Nuclear Installations Safety Standards) Initiative, in May 2005, in Brussels. The principal mission of ENISS is to bring together decisionmakers, operators and specialists from the nuclear industry with national regulators in order to identify and possibly agree upon the scope and substance of harmonized safety standards.ENISS currently represents the nuclear utilities and operating companies from 16 European countries with nuclear power programme. ENISS above all provides the nuclear industry with the platform that it needs to express its views, provides expert input and interacts fully with regulators throughout the harmonization process. ENISS first task has been to present a common industry position with regards to the Safety Reference Levels that WENRA has proposed. By engaging in a constructive debate with WENRA and playing a dynamic role in the process, ENISS also defends the industry’s interests in a proactive way.Another task of ENISS is to strengthen the industry influence in the revision work of the IAEA Safety Standards. In February 2007, the IAEA and ENISS launched their cooperation agreement. ENISS as an NGO is actively involved in the IAEA safety standards revision process in providing comments on draft safety standards, in particular as regards those where the industry has particular competence or interest. Those areas are NPP design and operation, management systems, waste management/treatment, decommissioning and radiation protection with respect to nuclear safety. ENISS send also experts to participate in IAEA drafting groups and attend the meetings of the IAEA Safety Standards Committees (NUSSC, WASSC, and RASSC) as an observer.At the European level, ENISS is also monitoring the work of the European Commission on regulatory issues in the area of nuclear safety, waste management, decommissioning and radiation protection and participate in ENEF groups.

Application of fractional scaling analysis (FSA) to loss of coolant accidents (LOCA): Methodology development

A quantitative methodology is developed to (a) scale time-dependent evolution processes involving an aggregate of interacting modules and processes (such as a NPP) and (b) integrate and organize information and data of interest to NPP design and safety analyses. The methodology is based on two concepts: fractional scaling and hierarchy. Fractional scaling is used to provide a synthesis of experimental data to generate quantitative criteria for assessing the effects of various design and operating parameters on thermo-hydraulic processes in a NPP. The synthesis via fractional scaling is carried out at three hierarchical levels: process, component and system. The methodology is demonstrated by applying it to a LOCA. The fractional scaling analysis (FSA) identifies dominant processes, ranks them quantitatively according to their importance and provides thereby an objective basis for establishing phenomena identification and ranking tables (PIRT) as well as a basis for conducting uncertainty analyses. The paper also discusses the benefits to be realized by applying the methodology to presently operating NPP as well as to future design of NPP.

Safety investigation of team performance in accidents

The paper presents the capacities of the performance evaluation of teamwork (PET) method. Its practicability and efficiency are illustrated by retrospective human reliability analyse of the famous nuclear and maritime accidents. A quantitative assessment of operators’ performance on the base of thermo-hydraulic (T/H) calculations and full-scope simulator data for set of NPP design basic accidents with WWER is demonstrated. The last data are obtained on the ‘WWER-1000’ full-scope simulator of Kozloduy NPP during the regular practical training of the operators’ teams. An outlook on the “evaluation system of main control room (MCR) operators’ reliability” project, based on simulator data of operators’ training is given.

Review of revised japanese seismic guidelines for npp design

From new knowledge in recently aseismic design studies, some of the conventional design methods we understood to be conservative, and review for guideline of private sector and preparation for by-laws of administrative office thereby have been promoted since 1984. This paper presents the state-of-arts and trends of current aseismic design methods and the guideline and the by-laws thereof for Japanese NPP, including items on determination of the design basis earthquake, investigation of ground condition and seismic stability, treatment of soil—structure interaction, limitation for uplift of building basement, evaluation of non-linear characteristics and response, and provision of design basis damping.

Principles of earthquake design for nuclear power plants in the FRG — A critical review and a realistic approach

The principles of earthquake design for NPP in the FRG are recorded in the German nuclear safety standard KTA 2201.1 “Design of NPP's against Seismic Events, Part 1: Basic Principles”, prepared in 1971 and 1972 and issued in 1975. The regulations of KTA 2201.1 are based on the US regulation App. A CFR 100 “Seismic and Geologic Siting Criteria for NPP's”. The principles are therefore basically the same in the States as in the FRG. This results at least in an overestimation of the earthquake risk in Germany. One of the important reasons for unrealistic design procedures and therefore uneconomic soft- and hardware consequences is bound to the requirement for a double earthquake design, namely the design against the Safe Shutdown Earthquake (SSE) and the Operating Basis Earthquake (OBE). Because of the identical physical consequences of these two earthquakes, at least for most of the known plant sites, where fatigue due to the smaller but more frequent eartquake (OBE) has no technical relevance, there is no reason for an OBE. Therefore the present regulatory standpoint should be revised. On the basis of this knowledge a group of German experts has drafted a new proposal for KTA 2201.1, including the following basic modifications: • - calculations for the “Design Earthquake” will be deleted • - design for a single maximum earthquake, the “Safety Earthquake” • - definition of a smaller earthquake, with the requirement to inspect and if necessary to shut down the plant in the case that such a defined earthquake occurs. • - possibility of using site-specific seismic input data instead of standard response spectra. • - no need for any calculations in the case of site intensity I ⩽ 6.