Overview of the OECD–Halden reactor project

The OECD Halden Reactor Project is an international network dedicated to enhanced safety and reliability of nuclear power plants. The Project operates under the auspices of the OECD Nuclear Energy Agency and aims at addressing and resolving issues relevant to safety as they emerge in the nuclear community. This paper gives a concise presentation of the Project goals and of its technical infrastructure. The paper contains also a brief overview of results from the programme carried out in the time period 1997–1999 and of the main issues contemplated for the 3-year programme period 2000–2002.     

Control of the energy distribution and safety of a VVÉR-1000 reactor when operating in the variable mode

Conclusions The main operations in the proposed algorithm for, controlling the distribution of energy under variable-load conditions are carried out at a reduced power level. When the power falls, all the control rods are found in the rated position. We show that the main problems in controlling the energy distribution can be solved within the framework of a single computer procedure based on the quadratic programming method. When the proposed algorithm is set up in the computer, together with a realization of the optimum control actions, found from a digital model, we can expect to achieve an increase in efficiency in the work of the operating staff, and as a result, an increased level of safety in the operation of nuclear power stations.Translated from Atomnaya Énergiya, Vol. 56, No. 2, pp. 67–71, February, 1984.     

A preliminary design study for the HYPER system

In order to transmute the long-lived radioactive nuclides such as transuranics (TRU), Tc-99, and I-129 in LWR spent fuel, a preliminary conceptual design study has been performed for an accelerator driven subcritical reactor system, called HYPER (HYbrid Power Extraction Reactor). The core has a hybrid neutron energy spectrum which includes fast and thermal neutrons for the transmutation of TRU and fission products, respectively. TRU are loaded into the HYPER core in a TRU–Zr metal form because a metal type fuel has very good compatibility with the pyro-chemical process which retains the self-protection of transuranics at all times. On the other hand, Tc-99 and I-129 are loaded as pure technetium metal and sodium iodide, respectively. Pb–Bi is chosen as a primary coolant because Pb–Bi can provide a good spallation target and produce a very hard neutron energy spectrum. As results, the HYPER system does not need any independent spallation target system. 9Cr–2WVTa is used as a window material because this advanced ferritic/martensitic steel is known to have a good performance in the highly corrosive and radiative environment. The support ratios of the HYPER system are about 4–5 for TRU, Tc-99, and I-129. Therefore, a radiologically clean nuclear power, i.e. zero net production of TRU, Tc-99 and I-129 can be achieved by combining 4–5 LWRs with one HYPER system. In addition, the HYPER system, having good proliferation resistance and high nuclear waste transmutation capability, is believed to provide a breakthrough to the spent fuel problems the nuclear industry is facing with.     

Nuclear-Hydrogen Power

Methods for obtaining hydrogen and using hydrogen in power engineering, transportation, and industry, and methods for handling hydrogen (storage and safety) are examined.The concept of nuclear-hydrogen power – using the energy generated by nuclear reactors to produce hydrogen and using this hydrogen in power engineering and industry – is presented. The development of nuclear-hydrogen power will contribute to global energy security and decrease the demand for fuels which affect climate change on our planet.The technologies needed for nuclear-hydrogen power to become a reality – high-temperature nuclear reactors, apparatus for the efficient production of hydrogen from water, hydrogen fuel cells, chemothermal converters, and hydrogen storage and shipment technology – are analyzed.     

Assessment the safety performance of nuclear power plants using Global Safety Index (GSI)

The safety performance of the nuclear power plant is a very important factor enhancing the nuclear energy option. It is vague to evaluate the nuclear power plant performance but it can be measured through measuring the safety performance of the plant.In this work, the safety of nuclear power plants is assessed by developing a “Global Safety Index” (GSI).The GSI is developed by introducing three indicators: probability of accident occurrence, performance of safety system in case of an accident occurrence (during an accident), and the consequences of the accident.The GSI is developed by tracking the performance of the safety system during a design basis accident such as loss of coolant accident (LOCA). This is done by using the PCTran simulation code in simulation a PWR LOCA and introducing four indicators: the sensation time, the response time, and the recovery time together with Core Damage Frequency (CDF). Then Fuzzy Inference System is used for obtaining the GSI.The GSI is also evaluated for the advanced types for nuclear power plants, such as AP1000, and a comparison is made between the GSI evaluated for both conventional and advanced types.     

Main achievements of FP-4 research in reactor safety

This summary paper deals with the strategy, the organisation and main achievements of the 67 multi-partner projects cosponsored by the European Union (EU) as ‘indirect actions’ (shared-cost and concerted actions) and co-ordinated under seven clusters, each devoted to one key safety issue in nuclear reactor safety, as they were presented at FISA-99. The fundamental safety objective for nuclear power plants (NPPs) consists in protecting the public and the environment from the harmful effects resulting from ionising radiations. The 4th Euratom framework programme 1994–1998 (FP-4) has been carrying out research with this objective both through ‘indirect actions’ and through ‘direct actions’ in co-operation with the Joint Research Centre of the European Commission (EC). The total cost of this research programme was 62.8 million, out of which 34.2 million was contributed by the EU budget. Besides technological safety requirements, socio-economic aspects are becoming increasingly important due to the level of public and political acceptance and to the economic pressure of deregulated electricity markets. It is shown that research conducted in the Euratom framework may contribute to meet these requirements, thereby maintaining nuclear power as a competitive and sustainable option for the energy policy of the European Union.     

Use of safety valves in the first loop of a nuclear power plant with a water-moderated-water-cooled power reactor

Conclusions It is not useful to transfer the requirements of the supervising committees on the safety of the usual vessels and high-pressure systems against their exceeding the allowable value of the pressure in nuclear installations with the conditions of radioactivity in the first loop of nuclear power plants, especially those onboard a vessel, with such a system as the active zone, which does not permit exposure and prolonged liberation of heat. Protection against an increase in pressure can be more effectively provided by reliable systems for shutting down the reactor, removing heat, and others, which is indicated by the comparative experience in using nuclear-powered icebreakers and some nuclear power plants abroad.Translated from Atomnaya Énergiya, Vol 50, No. 5, pp. 308–310, May, 1981.     

The use of nuclear reactors for high-temperature power-technological industrial processes

Conclusions Arising from the expected scales of development of high-temperature and power-capacity industrial establishments, and also from the reserves and disposition of fuel resources, it can be confirmed that the use of nuclear fuel for these factories can be completely justified. In this case, the actual economic expediency will depend on a sufficient level of technical reliability of the high-temperature complexes and nuclear reactors. The steam catalytic conversion of methane — which is one of the main links of the technological cycle for the large-scale production plants in the chemical industry for obtaining ammonia, methanol, higher alcohols, etc. — is in the maximum state of preparedness for achievement, based on the utilization of high-temperature heat from nuclear reactors. The conversion of methane for the chemical industry may become the technical basis for the introduction of nuclear reactors into ferrous metallurgical processes, the conversion of energy carriers and the production of hydrogen. The main technical problems lie in the field of transfer of the high-temperature heat from the core to the working space for carrying out the technological process, while observing the conditions of radiation safety of the production plant, the personnel and the environment. The search for an effective solution of the problem will be accomplished advantageously both in the direction of development of heat-transfer technical methods — which should allow gas-cooled nuclear reactors to be used for industrial technology — and also in the direction of development of new types of high-temperature nuclear reactors, directly oriented toward their introduction into industrial technology, e.g., based on the removal and transfer of heat from the core by radiative heat exchange, by means of a large-celled solid coolant.A preliminary estimate has shown that the incorporation of high-temperature nuclear reactors in industrial technology can be accompanied by a positive economic effect, if the specific capital investments in them do not exceed the specific capital investments in the power reactors of nuclear power stations by more than a factor of 1.5–2.0.Translated from Atomnaya Énergiya, Vol. 43, No. 6, pp. 432–437, December, 1977.     

Radiation Characteristics of Fuel and Wastes in the Uranium–Plutonium and Thorium–Uranium Fuel Cycle

The radiation characteristics of fuel cycles of various reactors – replacement candidates in the future nuclear power – are compared. Proceeding from the basic requirements (safety, fuel supply, and nonproliferation of fissioning materials), inherently safe fast reactors of the BREST type can be used as the basis for large-scale nuclear power. Thermal reactors, which can burn enriched uranium, thorium–uranium fuel, or mixed uranium–plutonium fuel with makeup with fissioning materials from fast reactors, will operate for a long time simultaneously with fast reactors in the future nuclear power. VVÉR-1000 and CANDU reactors are examined as representatives of thermal reactors; for each of these reactors the operation in various variants of the fuel cycle is simulated. It is shown that with respect to radiation characteristics of the fuel and wastes the thorium–uranium fuel cycle has no great advantages over the uranium–plutonium cycle.     

Seismic safety of nuclear power plants in eastern Europe

This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety review and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular, WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on "Benchmark study for the seismic analysis and testing of WWER type nuclear power plants". These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. The main conclusion of this paper is that even though there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems.     

Ecological problems in the development of nuclear power

Conclusions The initial and middle stages of the nuclear fuel cycle, i.e., mining and reprocessing of ore, uranium enrichment, production of fuel elements, and the normal operation of a nuclear power plant, do not cause any serious danger to the environment. Comparisons show that the negative-effect coal-fired HEP is much greater.The probability for accidents involving the emission of a large quantity of radionuclides in modern nuclear power plants equipped with tested safety systems is significantly lower than the accident probability in other areas of industry. This conclusion is valid, however, if safety requirements, starting with the nuclear power plant, are satisfiedunscrupulously, if the strictest technological discipline, making sure that all the components have sufficient reliability, is followed, and if constant efforts are made to train personnel.It is as yet impossible to evaluate quantitatively the environmental effects of reprocessing plants.Czechoslovakian Technical University, Prague. Translated from Atomnaya Énergiya, Vol. 49, No. 6, pp. 352–357, December, 1980.     

Construction management of Indian pressurized heavy water reactors

Pandit Jawaharlal Nehru and Dr. Homi J. Bhabha, the visionary architects of Science and Technology of modern India foresaw the imperative need to establish a firm base for indigenous research and development in the field of nuclear electricity generation. The initial phase has primarily focused on the technology development in a systematic and structured manner, which has resulted in establishment of strong engineering, manufacturing and construction base.The nuclear power program started with the setting up of two units of boiling light water type reactors in 1969 for speedy establishment of nuclear technology, safety culture, and development of operation and maintenance manpower. The main aim at that stage was to demonstrate (to ourselves, and indeed to the rest of the world) that India, inspite of being a developing country, with limited industrial infrastructure and low capacity power grids, could successfully assimilate the high technology involved in the safe and economical operation of nuclear power reactors. The selection of a BWR was in contrast to the pressurized heavy water reactors (PHWR), which was identified as the flagship for the first stage of India's nuclear power program. The long-term program in three stages utilizes large reserves of thorium in the monazite sands of Kerala beaches in the third stage with first stage comprising of series of PHWR type plants with a base of 10,000 MW. India has at present 14 reactors in operation 12 of these being of PHWR type.The performance of operating units of 2720 MW has improved significantly with an overall capacity factor of about 90% in recent times.The construction work on eight reactor units with installed capacity of 3960 MW (two PHWRs of 540 MW each, four PHWRs of 220 MW each and two VVERs of 1000 MW each) is proceeding on a rapid pace with project schedules of less than 5 years from first pour of concrete. This is being achieved through advanced construction technology and management. Present efforts are focused on further reduction of gestation period. This is in contrast to construction period of 7–14 years in the earlier projects with labour intensive construction methods, learning period and indigenisation. The schedule and cost are interrelated and ultimately determine the viability and competitive edge of a project. With rich experience of over 30 years of operation and construction management it is well established that setting up of nuclear power projects in India in 4–5 years is quite feasible because of tremendous developments in construction technology; mechanization, parallel civil works and equipment erection, computerized project monitoring and accounting systems. By considering the best achieved times for the critical path activities of previous and ongoing projects, even a 4-year schedule is achievable. For nuclear power to be competitive it is essential that the gestation period is reduced and the capacity utilization enhanced. Both of these are the goals of the Indian nuclear power program. Presently the overnight cost per kW installed capacity is in the range of US$ 1100–1300 with levellised tariff of 5 c/kWh.     

Automated Control and Safety of Nuclear Power Plants

The conceptual problems of developing and modernizing an automated system for the control of a technological process for nuclear power plants, which are under construction, or are currently operating, are discussed. Attention is focused on the human factor in the control of a nuclear power plant. The prospects for solving problems such as the optimal – from the safety standpoint – separation of control functions between man and machine, choice of means of automation adequate for the importance of the functions being performed with respect to safety, determination of the concept of the operator's job, and the systems for supporting the operator's work, and development of a modern man–machine interface, are examined.The solution of these problems is studied on the basis of a homocentric model of an automated system for controlling a technological process. The model uses extensively subsystems for supporting the work performed by the operator. It is noted that a supervisor role for the operator is preferable, since it provides the best psychological work environment. 11 references.     

Formation, characterisation and cooling of debris: Scenario discussion with emphasis on TMI-2

Regarding safety improvements for existing nuclear power plants, the TMI-2 accident is interesting because of the present commercial dominance of light water reactors (LWR). This accident demonstrated that the nuclear safety philosophy evolved over the years has to cover accident sequences involving massive core melt progression in order to develop reliable mitigation strategies for both, existing and advanced reactors. Although the TMI-2 core was reflooded, the results also appear applicable to the general melt progression phenomenology of most unrecovered (unreflooded) blocked core accident scenarios. Nevertheless, a large range in the initial conditions of core melt progression provides significant uncertainties in assessing the integrity of the lower head, the containment in severe reactor accidents, and the consequences of recovery actions in accident management, as well as core reflooding in particular. The probability of success of reflooding as an accident management strategy – in-vessel reflooding to terminate the accident and ex-vessel flooding to prevent reactor vessel melt-through – has to be assessed and discussed in detail.     

Nuclear power development in market conditions with use of multi-purpose modular fast reactors SVBR-75/100

This article presents an innovative nuclear power technology, based on the use of modular type fast-neutron reactors SVBR-75/100 having heavy liquid-metal coolant, i.e. eutectic lead–bismuth alloy, which was mastered in Russia for the nuclear submarines’ reactors. Reactor SVBR-75/100 possesses inherent self-protection and passive safety properties that allow excluding of many safety systems necessary for traditional type reactors. Use of this nuclear power technology makes it possible to eliminate conflicting requirements among safety needs and economic factors, which is particularly found in traditional reactors, to increase considerably the investment attractiveness of nuclear power based on the use of fast-neutron reactors for the near future, when the cost of natural uranium is low and to assure development of nuclear power in market conditions. On the basis of the factory-fabricated “standard” reactor modules, it is possible to construct the nuclear power plants with different power and purposes. Without changing the design, it is possible for reactor SVBR-75/100 to use different kinds of fuel and operate in different fuel cycles with meeting the safety requirements.     

Inherently safe containments for nuclear power plants

Nuclear energy cannot be avoided in the near future. To regain public acceptance the safety of nuclear power plants has to be increased. Consequently, feasibility studies have been carried out for a containment proposal for future pressurized water reactors which will keep people unharmed even in the case of severe nuclear accidents under the assumption “all that can go wrong, will go wrong”. The main features of the design concept are a core melt cooling and retention device, a passively acting cooling system to remove the decay heat and a double-wall containment which is able to withstand high static and dynamic internal pressures due to hydrogen detonation. Internal structures are designed to resist extreme loadings resulting from various accident scenarios including in-vessel steam explosion and vessel failure under high system pressure.     

Determination of the highest possible capital costs for the construction of a nuclear power plant based on data from construction abroad

In many countries, different points of view concerning the advantages of one or another technology for generating electricity are now encountered in the course of developing a strategy for increasing power production. The advantages and drawbacks of coal- and gas-burning and other types of power plants are discussed. Analysis of the possible development of nuclear power is given prominent place in the discussions. China, Finland, and India have now assigned an important role for the construction of new nuclear power plants in their energy programs. In other countries, discussions are continuing. In the present paper, the data on the construction of power plants abroad are used as a basis to extrapolate the basic factors influencing the profitability and competitiveness of various designs and to determine the role of Russian nuclear power in the competition on foreign markets.Translated from Atomnaya Énergiya, Vol. 97, No. 5, pp. 338–345, November, 2004.     

核电厂安全级仪表在线监测系统技术研究

目前国内核电厂主要采取定期校准的方式对安全级仪表漂移进行管理,但该方法过于保守且经济性差。基于此,本文对安全级仪表在线监测系统技术进行了研究,首先对安全级仪表实际漂移数据进行了分析,明确了核电厂安全级仪表漂移的主要类型,证明了对安全级仪表开展在线监测的可行性。其次,通过对相关法规及标准的分析和研究,明确了核电厂安全级仪表在线监测技术的基本要求。最后,开展了在线监测系统技术的数据分析研究,对冗余仪表提出了等价平均算法,对非冗余仪表算法进行了分析并对多元状态估计模型（MSET）方法开展了基于电厂实际数据的建模验证,证明了该方法在核电厂应用的可行性。      

Construction materials for reactor buildings for decommissioning and recycling

A conceptual fluid–steel structure was studied to investigate the seismic characteristics of its use in the reactor building of nuclear power plants. The results of the earthquake response analysis of the conceptual fluid–steel structure showed that the structure had the same seismic safety ability as conventional reactor buildings. Applying the fluid–steel structure to a rector building, results in the following advantages: more elastic and light weight building materials, reducing the decommissioning wastes; the ability to recycle the structure materials because the fluid in the steel structure can be discharged the steel can be reused easily; the fluid in the steel structure has the possibility of reducing the seismic response of the structure by the sloshing damper effect. Further study is encouraged by this results.     

Japanese regulatory practices on licensing, inspection, operational supervision and related standards

Japanese view on the safety of nuclear power plants is based on the concept that the primary responsibility for securing safety lies on electric power companies, installers of reactors.Under this concept, the Ministry of International Trade and Industry (MITI), in the course of designing and construction, has been performed an examination of the basic design and the detailed design of nuclear power plants, and in each stage of construction, a pre-operational inspection process. In addition, MITI, in operating stage, has been made throughgoing investigations on the causes of troubles and incidents as well as accidents that may affect operation, forcing utilities to take measures to prevent recurrence, and implementing safety regulation based on the “preventive maintenance” including elaborate checkings and overhaulings at the periodical inspections conducted for a period of three to four months after every 12-month operation cycle under the laws and regulations.This paper discusses the current status of nuclear power development in Japan, safety regulatory systems, views on safety and future prospects of securing safety.