EC-sponsored research activities on innovative passive safety systems

On April 26th 1994, the European Union (EU) adopted via a Council Decision a EURATOM Multiannual Programme for community activities in the field of nuclear fission safety (NFS) research for the period 1994–1998. An area of work having, as an objective, to ‘explore innovative approaches’ to improve the safety of future and existing reactors, was introduced in this programme. Most of the projects selected in this area were grouped under a common cluster known as ‘INNO’ and carried out on a ‘cost-shared’ basis, i.e. contribution of the European Commission is up to 50% of the total cost. The ‘INNO’ cluster was composed of eleven projects in which 35 different organisations, representing research centres, universities, regulators, utilities and vendors from seven EU member states and Switzerland, were involved. These projects proved to be an efficient means to gain the necessary phenomenological knowledge and to solve the challenging problems, many times of generic nature, posed among others by the characteristically small driving forces of the systems studied and by the lack of really prototypical test facilities.     

Variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents variations of a certain passive safety containment for a near future BWR. It is tentatively named Mark S containment in the paper. It uses the operating dome as the upper secondary containment vessel (USCV) to where the pressure of the primary containment vessel (PCV) can be released through the upper vent pipes. One of the merits of the Mark S containment is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. Another merit is the capability to submerge the PCV and the reactor pressure vessel (RPV) above the core level by flooding water from the gravity-driven cooling system (GDCS) pool and the upper pool. The third merit is robustness against external events such as a large commercial airplane crash owing to the reinforced concrete USCV. The Mark S containment is applicable to a large reactor that generates 1830 MW electric power. The paper presents several examples of BWRs that use the Mark S containment. In those examples active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

Different variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents probable variations of passive safety boiling water reactor (BWR). In order to improve safety and economy of passive safety BWR, the authors thought of use of a kind of improved Mark III type containment. The paper presents the basic configuration of the passive safety BWR that has an improved Mark III type containment. We tentatively call this passive safety BWR advanced safer BWR⁺ (ASBWR⁺) and the containment Mark X containment in the paper. One of the merits of the Mark X containment is double containment function against fission products (FP) release. Another merit is very low peak pressure at severe accidents without active cooling systems. The third merit is coolability by natural circulation of outside air. Therefore, the Mark X containment is very suitable for passive safety BWRs. It does not need a reactor building (R/B) as the secondary containment, because it is a double containment by itself. The Mark X containment is a general concept and also useful for half-passive safety BWRs that have both active and passive safety systems. In those examples, active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

Two types of a passive safety containment for a near future BWR with active and passive safety systems

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.     

Two new passive safety systems for LWR applications

In a research activity that SIET has been conducting for years about safety systems for light water reactors (LWRs), attention has been paid to developing two passive injection systems representing an innovative solution in mitigating the consequences of loss of coolant accidents. Both systems allow the completely passive injection of cold water into a pressurised vessel. They are triggered by a low-level signal and work on the base of phenomena like natural circulation and condensation. The simplest system, Sistema Iniezione Passiva 1 (SIP-1), injects water contained in a tank into a circuit at the same pressure as the circuit. The most complex system, injection cyclic system (ICS), injects cold water, by filling cyclically a proper tank with the water stored in an atmospheric pressure pool. Thanks to the ENEA sponsorship, this activity has been conducted in three steps: the definition of the conceptual design of the systems; the application of the Relap5 code to simulate their behaviour; and the proposal of their specific applications to pressurised and boiling LWR. In this paper, both systems are presented in their structural and operating characteristics together with the main results of the code application for their simulation. Some proposals of application of SIP-1 and ICS to pressurised water reactors and boiling water reactors are also shown. The developments and reached goals of the prosecution of the research are also summarised here, together with future needs.     

Development of passive safety systems for Next Generation PWR in Japan

Our country’s energy demand is expected to increase steadily into the future. When the situation of our country, which is not rich in energy resources, is taken into account, it seems that the importance of nuclear power generation will be heightened. Based on such a background, the basic policy for nuclear power generation is ‘from light water reactors to fast breeder reactors’. However, considering that light water reactors have become common, the recent outlook for the supply and demand for uranium resources, development trends of fast breeder reactor technology, etc., the light water reactor is expected to remain dominant in our country until at least the second half of the 21st century. Therefore, five PWR utilities in Japan (Hokkaido, Kansai, Shikoku, Kyushu, and Japan Atomic Power), Mitsubishi Heavy Industries Ltd and Westinghouse Electric Corporation have jointly started researching the Next Generation PWR (N.G.P) which is expected to be the leading nuclear power plant, taking place of APWR [T. Magari, Development of Next Generation PWR in Japan, Proceedings of the 10th Pacific Basin Nuclear Conference, 1996; K. Fujimura, et al., Proceedings of the Second International Symposium on Global Environment and Nuclear Energy Systems, 1996]. In this program, construction is targeted to start from 2010 based on expected future environmental conditions. Now, the capacity of more than 1500 MWe class PWR concept is investigated and a plant concept which has innovative features of a hybrid safety systems, i.e. an optimum combination of active and passive safety systems, and horizontal steam generators for core cooling at the accidents is developed as a promising candidate. The plant concept and the results of the investigation are presented in this paper.     

Safety analysis of multiple-failure of passive safety systems in SBWR-1200 SBLOCA

The design of the simplified boiling water reactor (SBWR-1200) is characterized by utilizing fully passive safety systems. The emergency core cooling is realized by the gravity driven core cooling system, and the decay heat removal is done by the passive containment cooling system and isolation condenser system. All of the systems have multiple units and could be partially failed. The objective of this paper is to analyze the system response under the multiple malfunctions of passive safety systems in the SBWR-1200.

The chosen accident scenario is a small break loss of coolant accident with one of three gravity driven core cooling system drain lines blocked and one of three passive containment cooling system condensers disabled. An integral test has been carried out in the PUMA facility for 16 h. The facility is designed for low pressure, long term cooling operation with the multiple safety related components; therefore, it has the flexibility to demonstrate the asymmetric or multiple-failure effects with the combination of disability of safety systems. The test initial conditions at 1 MPa (150 psi) are obtained from RELAP5/MOD3.2 code simulation for the SBWR-1200 with appropriate scaling considerations.

Comparisons have been first made between the multiple-failure test and a single-failure test preformed previously. It shows that the core has been covered with liquid coolant during all of accident transient even though there is an apparent coolant inventory reduction in the multiple-failure test. The decay heat removal has no significant difference because the remaining two passive containment cooling condensers increase their cooling capacities, and even the drywell pressure is slightly lower due to the cold water injection from the suppression pool. Comparisons have also been made between the scaled-up test data and the code simulation at the prototypic level. The prototypic simulation is done by RELAP5/MOD3.2. Agreements between the code simulation and the scaled-up test data confirm the code applicability and the facility scalability for this accident scenario.     

Two passive safety systems in a 200 MW demonstration nuclear heating plant

Two new passive safety systems for a demonstration nuclear heating plant, the residual heat removal system (RHRS) and the boron injection system (BIS), are introduced in this paper. Their common characteristic is that they have no driving equipment, therefore fluid circulation depends only on gravity (in BIS) or natural circulation (in RHRS). The inherent safety and realizability of both systems are illustrated in this paper.     

Testing of the multi-application small light water reactor (MASLWR) passive safety systems

Experimental thermal hydraulic research has been conducted at Oregon State University for the purpose of assessing the performance of a new reactor design concept, the multi-application small light water reactor (MASLWR). The MASLWR is a pressurized light water reactor design with a net output of 35 MWe that uses natural circulation in both normal and transient operation. Due to its small size, portability and modularity, the MASLWR design is well suited to help fill the potential need for grid appropriate reactor designs for smaller electricity grids as may be found in developing or remote regions. The purpose of the OSU MASLWR test facility is to assess the operation of the MASLWR under normal full operating pressure and full temperature conditions and to assess the passive safety systems under transient conditions. The data generated by the testing program will be used to assess computer code calculations and to provide a better understanding of the thermal-hydraulic phenomena in the design of the MASLWR NSSS. During this testing program, four tests were conducted at the OSU MASLWR test facility. These tests included one design basis accident and one beyond design basis accident. During the performance of these tests, plant operations to include start up, normal operation and shut down evolutions were demonstrated successfully.     

Research and development on passive cooling system

Experiments are carried out to investigate the effects of the natural convection of superheated gas as well as those of the standpipes on the temperature distributions of the components and the heat removal performance in the water-cooling panel system for the MHTGR for decay heat removal, and to verify reliability of the design and evaluation methods. The numerical results of the code THANPACST2 are compared with the experimental data to verify the numerical methods and axi-symmetric model proposed, which can simulate the three-dimensional configuration of the standpipes on the upper head of the pressure vessel by using porous body cells. The experiments revealed that temperatures increased with elevation on the upper head, because the standpipes restrict radiation heat transfer to the upper cooling panel and reduce the heat transfer area on the upper head, which was superheated by natural convection of helium gas in the pressure vessel. In the presumed accident condition in which thermal radiative heat transfer is responsible for the majority of the total heat transfer, the numerical methods were able to closely duplicate the pattern of the rising temperature profile with elevation around the top of the upper head as observed in the experiments.     

非能动安注箱的设计与分析

介绍了非能动安注箱的设计与实验,并用CATHENA程序分析其特性:注入流量的峰值,高注入流量的持续时间,最低注入流量等。计算结果表明非能动安注箱设计满足主要的性能要求,CATHENA程序计算结果与实验数据基本一致,可用于概念设计与事故分析。     

IAEA activities on Mmall and Medium Reactors: Harmonization of energy demand and supply in developing countries

Small and Medium Reactors (SMRs) are attractive in developing countries because of their unique features such as: better suitability for smaller electric grids, lower investment cost, smaller components and equipment to facilitate modularization, etc. Furthermore, other factors induced by SMR implementation, such as technical transfer promotion, domestic infrastructure improvement, stabilization of energy cost, and environmental protection put SMRs into a more favorable position. From the nuclear plant suppliers, many SMR designs are available for a wide range of applications. A questionnaire study, which the IAEA conducted in 1996, confirmed that several countries are interested in SMRs and that some SMRs are already in the detailed design stage. A projection shows that the total nuclear capacity would increase in all regions that consist mainly of developing countries in the near future. For a timely and broad implementation of SMRs, information exchange and cooperation are indispensable between nuclear suppliers and buyers. The IAEA continues to play a role in encouraging and assisting development and practical application of SMRs for harmonization of energy demand and supply in developing countries.     

Progress in liquid-metal fast breeder reactor safety research and development—An overview of the issues

The safety of Liquid-Metal Fast Breeder Reactors has been the subject of massive world-wide research and development. This paper attempts to give a brief overview of the R & D directed toward an understanding of those events that could contribute to the release and transport of fission products and transuranic elements up to the boundary of the reactor secondary containment system.     

Computational methods assuring nuclear power plant structural integrity and safety: an overview of the recent activities at VTT

Numerical, simplified engineering and standardised methods are applied in the safety analyses of primary circuit components and reactor pressure vessels. The integrity assessment procedures require input relating both to the steady state and transient loading actual material properties data and precise knowledge of the size and geometry of defects. Current procedures hold extensive information regarding these aspects. It is important to verify the accuracy of the different assessment methods especially in the case of complex structures and loading. The focus of this paper is on the recent results and development of computational fracture assessment methods at VTT Manufacturing Technology. The methods include effective engineering type tools for rapid structural integrity assessments and more sophisticated finite-element based methods. An integrated PC-based program system MASI for engineering fracture analysis is described. A summary of the verification of the methods in computational benchmark analyses and against the results of large scale experiments is presented.     

Accelerator driven systems for transmutation: Fuel development, design and safety

European R&D for ADS design and fuel development is driven in the 6th FP of the EU by the EUROTRANS Programme. In EUROTRANS two ADS design routes are followed, the XT-ADS and the EFIT. The XT-ADS is designed to provide the experimental demonstration of transmutation. The EFIT, the European Facility for Industrial Transmutation, aims at a conceptual design of a full transmuter. A key R&D issue is the choice of an adequate fuel. Various fuel forms have been assessed and CERCER and CERMET fuels, specifically the matrices MgO and Mo, have finally been selected. Within EUROTRANS, the domain ‘AFTRA’ is responsible to more deeply assess the behavior of these dedicated fuels and to provide the fuel database for the EFIT. The EFIT is optimized towards: a good transmutation efficiency, high burnup, low reactivity swing, low power peaking, adequate subcriticality, reasonable beam requirements and a high safety level. In the current paper the fuels under investigation are described, including their production route and their safety limits. First core designs of CERCER and CERMET fuelled 400 MWth EFITs have been developed within AFTRA. The trends found in the design studies and first safety analyses are presented.     

国际放射防护委员会当前的活动

ICRP建议主要是面向那些负责制定防护标准的审管者和业主。ICRP现有的建议书发表于1991年,从那时起,委员会还发表了其他的建议书,防护体系越来越复杂。委员会决定简化其体系,以使更加连贯一致。委员会决定在2005年通过一套新的建议书,可以看作是对早期建议书的充实。在讨论当前的防护体系时,正在考虑许多议题。各分委员会都在提出关切的问题进行修订。新的放射防护建议书草案于2004年在马德里的IRPA-11进行了讨论。当新的建议书在2005年通过的时候正是1990年建议书发表后的第15a。