The REPAS Study: Reliability Evaluation of Passive Safety Systems

The strategy of approach to the problem moves from the consideration that a passive system should be theoretically more reliable than an active one. In fact it does not need any external input or energy to operate and it relies only upon natural physical laws (e.g. gravity, natural circulation, internally stored energy, etc.) and/or “intelligent” use of the energy inherently available in the system (e.g. chemical reaction, decay heat, etc.). Nevertheless the passive system may fail its mission not only as a consequence of classical mechanical failure of components, but also for deviation from the expected behaviour, due to physical phenomena mainly related to thermalhydraulics or due to different boundary and initial conditions. The main sources of physical failure are identified and a probability of occurrence is assigned. The reliability analysis is performed on a passive system which operates in two-phase, natural circulation. The selected system is a loop including a heat source and a heat sink where the condensation occurs. The system behavior under different configurations has been simulated via best-estimate code (Relap5 mod3.2). The results are shown and can be treated in such a way to give qualitative and quantitative information on the system reliability. Main routes of development of the methodology are also depicted.

Development of New Constraction Method for LSBWR

LSBWR (Long operating cycle Simplified BWR) is a modular, direct cycle, light water cooled, and small power (100–300MWe) reactor. The design considers requirements from foreign utilities as well as from Japanese. LSBWR is currently being developed by Toshiba Corporation and Tokyo Institute of Technology. Major characteristics of the LSBWR are: 1) Long operating cycle (target: over 15 years), 2) Simplified systems and building, 3) Factory fabrication in module. From the perspective of economic improvement of nuclear power plant, it is needed to shorten the plant construction period and to reduce building volume. In designing LSBWR building, a new building structure, where the hull structure of a ship is applied to floors and walls of LSBWR has been studied. Since the hull structure is manufactured at a shipyard, building module that includes plant equipment becomes possible. The application of the hull structure, which can make large modules at a shipyard, is an effective solution to the lack of laborer and economic improvement. LSBWR is a small size BWR, turbine is smaller size and lighter weight than medium or larger size plant. Then, it has been studied to install a reactor and a turbine in the same building for decreasing building volume. From the view point of standardization, whole building is supported by three dimensional seismic isolation mechanism.

Studies on Materials for Heavy-Liquid-Metal-Cooled Reactors in Japan

Recent studies on materials for the development of lead-bismuth (Pb-Bi)-cooled fast reactors (FR) and accelerator-driven sub-critical systems (ADS) in Japan are reported. The measurement of the neutron cross section of Bi to produce 210Po, the removal experiment of Po contamination and steel corrosion test in Pb-Bi flow were performed in Tokyo Institute of Technology. A target material corrosion test was performed in the project of Transmutation Experimental Facility for ADS in Japan Atomic Energy Research Institute (JAERI). Steel corrosion test was started in Mitsui Engineering & Shipbuilding Co., LTD (MES). The feasibility study for FR cycle performed in Japan Nuclear Cycle Institute (JNC) are described.

Hydrogen Mixing Analyses for a VVER Containment

Hydrogen combustion may represent a threat to containment integrity in a VVER-440/213 plant owing to the combination of high pressure and high temperature. A study has been carried out using the GASFLOW 2.1 three-dimensional CFD code to evaluate the hydrogen distribution in the containment during a beyond design basis accident. The VVER-440/213 containment input model consists of two 3D blocks connected via one-dimensional (1D) ducts. One 3D block contains the reactor building and the accident localization tower with the suppression pools. Another 3D block models the air traps. 1D ducts represent the check valves connecting the accident localization tower with the air traps. The VVER pressure suppression system, called “bubbler condenser,” was modeled as a distributed heat sink with water thermodynamic properties. This model accounts for the energy balance. However, it is not currently possible to model dynamic phenomena associated with the water pools (e.g., vent clearing, level change). The GASFLOW 2.1 calculation gave detailed results for the spatial distribution of thermal-hydraulic parameters and gas concentrations. The range and trend of the parameters are reasonable and valuable. There are particularly interesting circulation patterns around the steam generators, in the bubbler tower and other primary system compartments. In case of the bubbler tower, concentration and temperature contour plots show an inhomogeneous distribution along the height and width, changing during the accident. Hydrogen concentrations also vary within primary system compartments displaying lower as well as higher (up to 13–20% and higher) values in some nodes. Prediction of such concentration distributions was not previously possible with lumped parameter codes. GASFLOW 2.1 calculations were compared with CONTAIN 1.2 (lumped parameter code) results. Apart from the qualitatively similar trends, there are, for the time being, quantitative differences between the results concerning, for example, pressure histories, or the total amount of steam available in the containment. The results confirm the importance of detailed modeling of the containment, as well as of the bubbler condenser and sump water pools. The study showed that modeling of hydrogen distribution in the VVER-440/213 containment was possible using the GASFLOW 2.1 code with reasonable results and remarkable physical insights.

U.S. Department of Energy Sponsored In-Depth Safety Assessments of VVER and RBMK Reactors

Throughout the 1990s the National Nuclear Security Administration of the U.S. Department of Energy has worked to build capability in countries of the former Soviet Union to assess the safety of their VVER and RBMK reactors. Through this Plant Safety Evaluation Program, deterministic and probabilistic analyses have been used to provide a documented plant risk profile to support safe plant operation and to set priorities for safety upgrades. Work has been sponsored at fourteen nuclear power plant sites in eight countries. The Plant Safety Evaluation Program has resulted in immediate and long-term safety benefits for the Soviet-designed nuclear plants.

RELAP5/MOD3.2 Investigation of a VVER-1000 MCP Switching on Problem

This paper provides a discussion of various RELAP5 parameters calculated for the investigation of the nuclear power reactor parameter behavior in case of switching on one main coolant pump (MCP) when the other three MCPs are in operation. The reference power plant for this analysis is Unit 6 at the Kozloduy Nuclear Power Plant (NPP) site. Operational data from Kozloduy NPP have been used for the purpose of assessing how the RELAP5 model compares against plant data. During the plant-commissioning phase at Kozloduy NPP Unit 6 a number of experiments have been performed. One of them is switching on MCP when the other three MCPs are in operation. The event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which leads to insertion of positive reactivity due to the modeled feedback mechanisms. This investigation has been conducted by Bulgarian and Russian specialists on the stage when the reactor power was at 75% of the nominal level. The purpose of the experiment was the complete testing of reliability of all power plant equipment, testing the reliability of the main regulators and defining a jump of the neutron reactor power in case of switching on of one main coolant pump. The Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences (INRNE-BAS), Sofia, and Kozloduy NPP have been developing a RELAP5/MOD3.2 model for Kozloduy NPP VVER-1000 for investigation of operational occurrences, abnormal events, and design basis scenarios. This investigation is a process that compares the analytical results obtained by the RELAP5 computer model of the VVER-1000 against the experimental transient data received from the Kozloduy NPP Unit 6. The comparisons between the RELAP5 results and the test data indicate good agreement. This report was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.

Design of Compact and Full-Size Simplified Boiling Water Reactors With Fully Passive Safety Systems

Scientific designs of two next-generation simplified boiling water reactors (SBWRs) namely, a compact modular 200 MWe SBWR and a full-size 1200-MWe SBWR have been developed. The design involved identification of principal design criteria dictated by the safe operation of the reactor, identification of coolant requirements, and the design of the engineered safety and emergency cooling systems based on passive systems. A detailed scaling analysis was performed. The results of the scaling study were used in the performance of the integral tests and data analysis. The scaling analysis identified key thermalhydraulics parameters that govern flow phenomena in SBWRs. The analysis was based on the three-level scaling approach.

Hydrogen Recombiner: Theoretical Parametric Study

There is not yet published data concerning a complete overview of the behaviour of a SIEMENS recombiner versus the thermal hydrualic conditions and the geometry of the catalytic plates. This paper reports on a numerical behaviour of the recombiner depending on several gas parameters as the total pressure, and the hydrogen concentration, as well as geometrical parameters of the catalytic elements as the height and the spacing. We use a theoretical model validated by using the KALI experiments. In this model (Avakian, 1999), the reaction rate is diffusion-controlled, i.e. the contribution of surface kinetics to the total rate of reaction is neglected. We demonstrate a quasi-linear behaviour of the recombination rate vs. the total pressure and the hydrogen concentration. We display the benefit in using smaller catalytic plates instead of taller plates and we give an idea of the influence of the spacing between the catalytic plates.

Integrated Risk-Informed Asset Management for Commercial Nuclear Power Stations

The objective of this paper is to provide electric utilities with a concept for preparing and implementing integrated risk-informed asset management (RIAM) programs for power stations and generating companies. RIAM is a process by which analysts review historical performance and develop predictive logic models and data analyses to predict critical decision support figures-of-merit (or metrics) for generating station managers and electric utility company executives. These metrics include, but are not limited to, the following: profitability, projected revenue, projected costs, asset value, safety (catastrophic facility damage frequency and consequences, etc.), power production availability (capacity factor, etc.), efficiency (heat rate), and others. RIAM applies probabilistic safety assessment (PSA) techniques and generates predictions probabilistically so that metrics information can be supplied to managers in terms of probability distributions as well as point estimates. This enables the managers to apply the concept of “confidence levels” in their critical decision-making processes.

Main Results of Na-K Alloy Boiling Investigation

Boiling experiments on eutectic sodium-potassium alloy in the model of fast reactor subassembly under conditions of low-velocity circulation carried out at the IPPE call for further investigations into numerical modeling of the process. The paper presents analysis of pin bundle liquid metal boiling, stages of the process, its characteristics (wall temperature, coolant temperature, flow rate. pressure void fraction and others), that allowed the pattern map to be drawn. The problem of conversion of the data gained in Na-K mock-up experiments to in-pile sodium reactor operating conditions is analyzed here, as well as thermodynamic similarity of liquid metal coolants and eutectic Na-K alloy. Data on bundle boiling in Na-K are presented in comparison with those in different liquid metals. Analysis of data on liquid metal heat transfer in cases of pool boiling, boiling in tubes, in slots, and in pin bundles, as well as data on critical heat flux in tubes was performed and discussed in the paper. The relationship for calculation of critical heat flux in liquid metal derived by the authors is presented. Results of numerical modeling of liquid metal boiling heat transfer during accident cooling of reactor core applied to experimental conditions of going from forced to natural circulation are presented, too.