Containments design for the advanced and simplified boiling water reactor standard plants

The Advanced Boiling Water Reactor (ABWR) design is based on construction and operating experience of nuclear power plants in Japan, United States, and Europe. To optimize the plant arrangement of the Advanced Boiling Water Reactor (ABWR) and to verify the structural feasibility to carry design loads a study was conducted. To arrive at an optimized plant arrangement with a minimum size reactor building (RB), a circular cylindrical reinforced concrete containment vessel (RCCV) with a flat top slab and a monolithically connected diaphragm slab has been selected.The Simplified Boiling Water Reactor (SBWR) is being developed as a standardized 600 MWe Advanced Light Water Reactor. The design concept of the SBWR is based on simplicity and passive features to enhance safety and reliability, improve performance and increase economic viability. Due to the use of passive containment cooling, SBWR has features that are different from those of existing designs.The objectives of the study for the ABWR containment and RB are to perform a structural analysis of the containment and RB and to evaluate the structure for conformance to the U.S. NRC requirements. The main objective of the studies for the SBWR is to demonstrate the structural design feasibility of the containment for the pressure and the temperature response associated with the passive systems adopted for the SBWR.     

Optimization of material and production to develop fluoroelastomer inflatable seals for sodium cooled fast breeder reactor

The feasibility of producing thin-walled fluoroelastomer profiles under continuous, atmospheric-pressure vulcanization conditions in air has been demonstrated by successful manufacture of ∼2 m diameter test inflatable seals for the 500 MWe, Prototype Fast Breeder Reactor (PFBR) using a 50/50 blend formulation of Viton^® GBL-200S/600S based on advanced polymer architecture (APA). A commercial cold feed screw extruder with 90 mm diameter screw was used along with continuous cure by microwave (2.45 GHz) and hot air heating (190 °C) at a line speed of 1 m/min to produce the seals. The blend formulation promises significant improvement in the performance and safety of the seals. This article depicts the relevant characteristics of the original inflatable seal compound that was used as reference to achieve the objectives through synchronized optimization of material and production technologies. The production trials are outlined and the blend formulation used with minor factory modifications to produce the test seals is reported. Progressive refinements of the original, Viton^® A-401C based compound to the blend formulation is presented along with an assessment of potential performance gains. Possible uses of the reported formulation and production technique for other large diameter, high temperature seals of PFBR are indicated along with the envisaged activities en-route the production of perfected reactor inflatable seals.     

Performance of containment penetrations under severe accident loadings

The paper provides a summary of efforts to date to better understand the leakage behavior of containment penetrations when subjected to severe accident conditions. The research activities discussed herein are a part of the Containment Integrity Programs, which are managed by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. Past containment penetration research topics, which are briefly described, include testing of typical compression seals and gaskets, electrical penetration assemblies, and a personnel airlock, as well as an investigation of leakage due to ovalization of penetration sleeves. The primary focus of the paper is on recent or ongoing research programs on the behavior of inflatable seals, bellows, and of pressure unseating equipment hatches.     

Insights into the behavior of LWR steel containment buildings during severe accidents

Investigations into the performance of steel containment subject to pressure and temperature greater than their design basis loads are discussed. The timing, mechanism, and location of a containment failure, i.e., release of radioactive material, have an important impact on the consequences of a severe accident. We review the results of experiments on steel containment models pressurized to failure, on aged and unaged seals subjected to elevated temperature and pressure, and on electrical penetration assemblies tested for leakage. Based on the results, the important features and details of analytical methods that can be used to predict containment performance are identified. Finally, we speculate on the performance of steel containments in severe accident conditions.     

A generic approach for steel containment vessel success criteria for severe accident loads

Safety has been defined as the foremost design criterion for the Heavy Water New Production Reactor (NPR-HWR) by the U.S. Department of Energy (DOE), Office of New Production Reactors (NP). The DOE-NP issued the Deterministic Severe Accident Criteria (DSAC) concept to guide the design of the NPR-HWR containment for resistance to severe accidents. The DSAC concept provides for a generic approach for containment vessel success criteria to predict the threshold of containment failure under severe accident loads. This concept consists of two parts: (1) Problem Statements and (2) Success Criteria. This paper is limited to a discussion of a generic approach for steel containment vessel success criteria. These criteria define acceptable containment response measures and limits for each problem statement. The criteria are based on the “best estimate” of failure with no conservatism. Rather, conservatism, if required, is to be provided in the problem statements prepared by the designer and/or the regulatory authorities. The success criteria are presented on a multi-tiered basis for static pressure and temperature loadings, dynamic loadings, and missiles that may impact the containment. Within the static pressure and temperature loadings and the dynamic loadings, the criteria are separated into elastic analysis success criteria and inelastic analysis success criteria. Each of these areas, in turn, defines limits on either the stress or strain measures as well as on measures for buckling and displacements. The rationale upon which these criteria are based is contained in referenced documents. Rigorous validation of the criteria by comparison with results from analytical or experimental programs and application of the criteria to a containment design remain as future tasks.     

Leakage potential of LWR containment penetrations under severe accident conditions

This paper is an overview of a Sandia National Laboratories, Albuquerque (SNLA) study of the performance of mechanical penetrations in light-water reactor (LWR) containment buildings that are subjected to severe accident environments. The study is concerned with modes of failure as well as the magnitude of leakage. The following tests have been completed, are under way, or are planned: (a) seals and gaskets have been tested to register the effects of radiation aging, thermal aging, seal geometry, and squeeze on seal and gasket materials in severe accident environments; (b) the performance of a full-scale airlock will be evaluated at severe accident temperature and pressure levels; (c) personnel airlock and equipment hatch tests were made on a model of a steel containment building; and (d) tests of mechanical penetrations are planned as part of a test on a model of a reinforced concrete building. This program is part of an overall US Nuclear Regulatory Commission (USNRC) effort to evaluate the integrity of LWR containment buildings.     

Temperature Measuring System of the In-core Components for Chinese 10 MW High Temperature Gas-cooled Reactor

The 10 MW High Temperature Gas-cooled Reactor (HTR-10) constructed at the Institute of Nuclear Energy Technology (INET), Tsinghua University in China reached its first criticality by the end of 2000. The temperature measuring system of the in-core components is described in this paper. This system consists of the thermocouple penetration assembly of the reactor pressure vessel (TPARPV), the thermocouple penetration assembly of the reactor containment (TPARC) and the distributed computer-based data acquisition and processing system (DCS). Some new techniques were developed and applied, such as the thermocouple penetration technology under the high temperature and high-pressure environment and the laser welding technique. The TPARPV is the key measurement device and is described in detail. The general behavior of the TPARPV and TPARC was confirmed under HTR-10 operating conditions. The helium leakage rate of the TPARPV is 1×10^-7Pa-m³/s while the helium leakage rate of the TPARC is less than 1×10^-2 Pa-m³/s. The insulation resistance of the sheathed thermocouple is more then 10⁹Ω. The temperature measurement error of the system is 2.3°C. The results of testing and field inspection and operation demonstrate that the design of the temperature measuring system is reasonable and reliable and that the performance of the system satisfies the design requirements of the HTR-10. These new techniques used in the temperature measuring system can be applied not only to other high temperature gas-cooled reactors but to various reactor types as well.     

秦山核电厂安全壳系统B、C类密封性试验

叙述了秦山核电厂安全壳系统B、C类密封性能试验概况,主要包括试验范围、泄漏率分配、试验结果和总体评价等。     

Experimental Testing on a Simulated Flask Lid Seal System to Confirm Containment Measuring Procedures

Abstract

British Nuclear Fuels Ltd (BNFL) operates a large number of package types for the transport of irradiated nuclear fuels from customers' utilities to the Sellafield reprocessing facility in Cumbria. All fuel loading operations are carried out under water and consequently package lid sealing systems are saturated. All BNFL package types employ a double seal system on the lid which must be tested before despatch, to confirm containment integrity. The normal test procedure involves drying the interspace between the seals with compressed air before applying a gas pressure and measuring the pressure drop; the reliability of this procedues depends upon the seals being dry. In order to demonstrate the reliability of BNFL's containment testing methods, and to develop operational procedures that ensure acceptable dryness is achieved, an experimental test rig was designed and manufactured. Closely based on a typical package lid seal arrangemens, the test rig allowed leakage paths to be introduced by fine wires fitted across the seal faces. BNFL conducted a series of tests to investigate how the measured leak rate was influenced by the presence of water. Existing drying procedures were evaluated, and shown not to be fully effective in removing all moisture. New drying procedures were subsequently developed, which are totally efficient in drying the test inter-space and ensure that accurate containment measurements can be undertaken.     

Annulus pressurization load evaluation after high energy line break

In Light Water Reactor design, it is required by the US Nuclear Regulatory Commission (NRC) that a conservative method must be used to estimate the loads on the Reactor Pressure Vessel (RPV) and components in the analysis of the loss of coolant accident (LOCA). As part of the safety design of the containment for light water reactors, jet forces resulting from the postulated rupture of high-pressure piping are accounted for in the design of internal structures. Traditionally, the jet force was estimated using ANS 58.2. However, it does not address the compressibility, reflection, and feedback amplification of the pressure forces as directed by NRC. In this paper, a numerical method is developed to estimate the annulus pressurization caused by the high-energy line break in the Economic Simplified Boiling Reactor (ESBWR). The pressure loading time history caused by the blast wave was predicted by the method. From the prediction, the maximum pressure loading was obtained. The method and the results have been submitted to the NRC as the supporting documents for the ESBWR certification.     

Optimization of Hot Gas Chamber in High Temperature Reactor

The flow field in the hot gas chamber of the High Temperature Gas Cooled Reactor (HTGCR) was studied with the Computational Fluid Dynamics (CFD) program CFX5. On the basis of the experimental studies, the velocity field, pressure field and temperature field in the hot gas chamber and hot gas duct were obtained, and the simulation's accuracy and reliability were validated by comparison with the results of previous experiments. Two other design schemes of the hot gas chamber were calculated in order to determine which hot gas chamber would be optimal for minimizing temperature differences at the inlet of the heat exchanging components. The results indicated that there was much highly turbulent twisting flow in the hot gas chamber, which was responsible for the excellent temperature mixing effect of the hot gas chamber. But the flow in the rib region was calm, and this fact hindered the heat transfer between the hot and cold gas. The temperature mixing coefficient increases with the increase of the hot gas duct's distance from the hot gas chamber. The hot gas chamber without ribs was more beneficial to the heat transfer between air flows with different temperatures, so the hot gas chamber without ribs was indicated as the optimal design.     

A fundamental approach to specify thermal and pressure loadings on containment buildings of sodium cooled fast reactors during a core disruptive accident

Reactor Containment Building (RCB) is the ultimate barrier to the environment against activity release in any nuclear power plant. It has to be designed to withstand both positive and negative pressures that are credible. Core Disruptive Accident (CDA) is an important event that specifies the design basis for RCB in sodium cooled fast reactors. In this paper, a fundamental approach towards quantification of thermal and pressure loadings on RCB during a CDA, has been described. Mathematical models have been derived from fundamental conservation principles towards determination of sodium release during a CDA, subsequent sodium fire inside RCB, building up of positive and negative pressures inside RCB, potential of in-vessel sodium fire due to failed seals and temperature evolution in RCB walls during extended period of containment isolation. Various heating sources for RCB air and RCB wall and their potential have been identified. Scaling laws for conducting CDA experiments in small-scale water models by chemical explosives and the rule for extrapolation of water leak to quantify sodium leak in reactor are proposed. Validation of the proposed models and experimental simulation rules has been demonstrated by applying them to Indian prototype fast breeder reactor. Finally, it is demonstrated that in-vessel sodium fire potential is very weak and no special containment cooling system is essential.     

核电厂安全壳首次整体在役试验方案优化

安全壳整体试验是压水堆核电机组一项特大型、高风险、高难度的试验,通过模拟设计基准事故工况下安全壳内的峰值压力,在事故峰值压力平台下,进行安全壳整体泄漏率测量及各压力平台安全壳结构试验,以验证其密封和结构性能。安全壳整体试验是国家核安全局监管的一个重要见证点,试验结果直接决定是否能够启动反应堆发电。301大修安全壳整体试验是3号机组首次在役试验,本次试验汲取了秦山第二核电厂以往6次安全壳整体试验的经验和其他电厂的反馈,试验方案更加科学,试验的组织管理更为规范。文章对301大修安全壳整体试验的经验进行了论述和总结,希望对电厂以后的安全壳整体试验提供参考。     

An analysis of containment failure by a steam explosion following a postulated core meltdown in a light water reactor

The Reactor Safety Study (WASH-1400) assessed the probability of containment failure via a steam explosion during a postulated core meltdown accident to be 10⁻². Large uncertainties were attached to this probability and research has continued to reduce the uncertainty.In this paper, we discuss the possible consequences of a steam explosion for a specific reactor system (Zion Nuclear Station—Pressurized Water Reactor). It is our opinion, based on the analysis performed, that generation of large mass missiles by the explosion is unlikely, while small mass missiles, although more likely would not pose a threat to the containment. We do not mean to imply that steam explosions can be disregarded during a postulated meltdown accident, but rather that emphasis should now be placed on how the explosion affects the overall core meltdown accident instead of causing a direct failure.     

European supercritical water cooled reactor

The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 °C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers in the technologies of light water reactors. More than 20 bachelor or master theses and more than 10 doctoral theses on HPLWR technologies have been submitted at partner organizations of this consortium since the start of this project.     

Containment behaviour in the event of core melt with gaseous and aerosol releases (CONGA)

The CONGA project concentrated on theoretical and experimental studies investigating the behaviour of advanced light water reactor containments containing passive containment heat removal systems and catalytic recombiners expected to be effectively operational during a hypothetical severe accident involving large quantities of aerosol particles and noncondensable gases. The central point of interest was the investigation of the effect of aerosol deposition on the condensation heat transfer of specially designed finned-type heat exchangers (HX) as well as the recombination efficiency of catalytic recombiners. A conceptual double-wall Italian PWR design and a SWR1000 design from Siemens were considered specifically as the reference Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) designs. An assessment of selected accident scenarios was performed in order to define the range of boundary conditions necessary to perform the experimental studies of the other work packages. Experimental investigations indicated that aerosol deposition accounted for up to 37% loss in the heat removal capacity of the two-tube-layer BWR HX units. However, no significant heat transfer degradation could be observed for the PWR HX units. These results can be attributed to the important differences in the designs and operating conditions of the two units. The tests to study the effect of hydrogen (simulated by helium) on the heat transfer rate for heat exchanger units designed for BWR and PWR applications indicated a degradation less than 30% under various conditions. This was found to be acceptable within the over capacity designed for the heat exchangers or containment characteristics. The tests performed to study the long-term aerosol behaviour in the pressure suppression chamber of the current operating BWRs indicated that the water pool scrubs the aerosol particles effectively and reduces the ultimate aerosol load expected on the off-gas system. The efficiency of the catalytic recombiner system designed by Siemens for the off-gas system was found to be insensitive to the aerosol deposited in the recombiner. A computation code, HTCFOUL, was developed to predict the heat transfer rate of a finned-type heat exchanger subjected to a steam–noncondensable gas mixture containing airborne aerosol particles. The model predicts the non-aerosol part of two tests within a variation of 26% and the aerosol part within 32%.     

中国一体化反应堆核电厂创新安全壳设计研究

中国一体化反应堆核电厂(CIP)是中国核反应堆系统设计技术国家重点实验室正在开发的新一代革新型、完全一体化的压水堆,其电功率约为300 MW.CIP采用堆内一体化布置,反应堆冷却剂系统设备以及控制棒驱动机构全部布置在反应堆压力容器内.这种一体化设计消除了传统的冷却剂回路管道,消除了大LOCA事故,具有更高的安全性.本文介绍了CIP安全壳系统方案选择、安全壳设计、安全壳设计压力的确定以及安全壳结构的计算分析.     

高温气冷堆热气联箱内部流场数值模拟

以高温气冷堆热气联箱为研究对象，在实验研究基础上，采用流体力学计算程序CFX5对热气联箱和热气导管内部流场进行数值模拟，以获得热气联箱和热气导管内的速度场、压力场和温度场，为高温气冷堆热气联箱的设计和实验研究提供参考。数值计算结果表明：热气联箱内气流发生剧烈搅混，加速了不同温度气流间的热传递，有利于高温和低温气流间的温度混合，存在肋片的区域未发生剧烈的气流搅混，不利于气流间的热传递；热气导管内温度混合率随其长度的增加逐渐增大，当热气导管长度为2.5m以上时，温度混合率达到99％以上。     

Small Gas Leakage Rate Measuring and Monitoring System for Spent Fuel Transport/Storage Cask

A containment function of transport and/or storage casks of radioactive materials is essential to prevent the materials from being released excessively into the environment. It is not practical for containment tests to measure directly the radioactivity release so that gas volumetric leakage rates are usually assessed and gas pressure decrease or increase method is usually applied. As gas flow model for evaluation, the ISO standards has deleted the concept of choked flow which is adopted by ANSI N14.5. Provided that the choked flow is not adopted to the leakage rate evaluation, the criteria of the test should be severer, and a new leakage rate measuring system with high accuracy and reasonable measuring time is required. Transport casks are often inspected in a temporary cask-storage facility where simultaneous measurement of many casks is required. In a storage cask system, multiple casks are monitored on their containment function during a storage period, and the method for simultaneous monitoring at many points for long term is required. In this study, two kinds of small gas leakage rate measuring systems are developed. One is to measure gas leakage rates directly and is called “flow measuring system”, which can measure gas leakage rate of 10^?4 to 10^?2 cm³/s with high accuracy and short measuring time. The other is to measure the pressure decreasing rate and is called “pressure decreasing rate measuring system”, which can monitor the pressure change at many points simultaneously.