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.

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.

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.

Large Break LOCA Safety Injection Sensitivity for a CE/ABB System 80+ PWR

A WCOBRA/TRAC model of an evolutionary pressurized water reactor with direct vessel injection was constructed using publicly available information and a hypothetical double-ended guillotine break of a cold leg pipe was simulated. The model is an approximation of a ABB/Combustion Engineering System 80+ pressurized water reactor (PWR). WCOBRA/TRAC is the thermal-hydraulics code approved by the U.S. Nuclear Regulatory Commission for use in realistic large break LOCA analyses of Westinghouse 3- and 4-loop PWRs and the AP600 passive design. The AP600 design uses direct vessel injection, and the applicability of WCOBRA/TRAC to such designs is supported by comparisons to appropriate test data. This study extends the application of WCOBRA/TRAC to the investigation of the predicted behavior of direct vessel injection in an evolutionary design. A series of large break LOCA simulations were performed assuming a core power of 3914 MWt, and Technical Specification limits of 2.5 on total peaking factor and 1.7 on hot channel enthalpy rise factor. Two cladding temperature peaks were predicted during reflood, one following bottom of core recovery and a second caused by saturated boiling of water in the downcomer. This behavior is consistent with prior WCOBRA/TRAC calculations for some Westinghouse PWRs. The simulation results are described, and the sensitivity to failure assumptions for the safety injection system is presented.

Westinghouse AP1000 Advanced Plant Simplification Results, Measures, and Benefits

A clear distinction between advanced plants, such as the Westinghouse AP1000 and AP600, and evolutionary plants is the policy in the latter to use current systems’ and buildings’ configurations. This approach does not promote simplification or streamlining, especially in the mechanical systems of the plant. The most significant simplification in evolutionary designs has arguably been in the plant electronics where compact digital components and multiplexing have led to improvements, especially in the areas of information display, installation, and testing. The Westinghouse advanced, passive plants take a different approach. Their design engineers presume that if regulatory requirements can be satisfied by using passive systems, then active plant systems that are only designed to meet plant control functions and not burdened with meeting a safety pedigree can be implemented. This separation of safety and control allows the plant designer to focus on systems’ optimization and reliability by reducing complexity and its associated cost. This design policy has led Westinghouse to the AP600 and AP1000 plant configurations, both of which incorporate significant improvements in areas of plant simplification and enhanced safety.