[en] An examination has been made of how the currently available computing capabilities could be used to reduce Liquid Metal Fast Reactor design, manufacturing, and construction cost. While the examination focused on computer analyses some other promising means to reduce costs were also examined. (author)

[en] The main results of the seventeen-year operation of the BN600 Nuclear Power Plant are considered. The principal backfittings of the main BN600 Power Plant equipment are presented and summarised. (author)

[en] Design of liquid metal cooled fast reactors (LMFRs) is still in evolution, and only a small number of LMFRs are in operation around the world. Specialists operating these LMFRs have gained valuable experience from incidents, failures, and other events that took place in the reactors. These unusual occurrences, lessons learned and measures undertaken to prevent recurrences are often either not reported in the literature, or reported only briefly and without sufficient detail. Hence there is a need for specialists designing and operating LMFRs to share their knowledge on unusual occurrences. Considerable experimental and theoretical knowledge in this field were collected by several Member states over the past decades. The needs in generalisation, review and documentation of fundamental knowledge in liquid metal cooled reactor technology were a major consideration in the recommendation by the International Working group on fast reactors for the IAEA to convene this Technical Committee meeting on the subject of unusual occurrences during LMFR operation and the consequences for reactor systems

[en] The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This paper summarizes some of the results of engineering and design analyses performed for the safety of KALIMER. (author)

[en] An analysis of about several unusual accidents about Emergency Residual Heat Removal System (ERHRS) of China Experiment Fast Reactor (CEFR) is presented. CEFR is a pool-type sodium-cooled fast reactor. The ERHRS of this reactor is designed in passive principle, which enhances the interior reliability of CEFR. It consists of two sets of independent channels. Each channel is comprised of decay heat exchanger (DHX), intermediate circuit, sodium-air heat exchanger (AHX) and related auxiliary system. Both DHX are located in the hot pool of the main vessel directly, which is used to cool the hot sodium. The whole set of ERHRS is completely passive except the ventilation valves of AHX. But, as a very important set of engineered safety features which is the final way to remove the heat from the reactor core, it is necessary to pay attention to all of the possibilities that may reduce this ability. Several accidents are analyzed including when the ventilation valves couldn't be opened, when only one set of ERHRS could work and so on. The calculation results show that the ERHRS can keep the reactor in a safety status. Even though it is, experiments are still necessary in the view of engineering. (author)

[en] PFBR is a 500 MWe, sodium cooled, pool type, fast breeder reactor currently under detailed design. It is essential to reduce the capital cost of PFBR in order to make it competitive with thermal reactors. Operating experience of LMFBRs provides a vital input towards simplification of the design, improving its reliability, enhancing safety and achieving overall cost reduction. This paper includes a summary of LMFBR operating experience and details the design features of PFBR as influenced by operating experience of LMFBRs. (author)

[en] The UK Prototype Fast Reactor, PFR, was designed in the 1960s and was operated at Dounreay in Scotland from 1974 to 1994. By the time it was shut down it had demonstrated the feasibility of the technology of a large sodium-cooled fast breeder reactor, and had been shown to operate safely and reliably. It had also provided an invaluable test facility for advancing the technology, particularly in developing advanced fuel and cladding materials that had achieved high burnup and neutron dose. As is usual in prototype plants the operation of PFR revealed the weak points of the original design concept. Several difficulties were encountered in the course of its operating life, all of which were successfully overcome. The purpose of this paper is to describe some of these difficulties and the steps taken to master them. In this way the benefit of the experience gained and the lessons learnt can be made available to the designers and operators of reactors of similar type. The intention is that future generations will not follow false trails in the further development of this promising technology. (author)

[en] Review of various anomalous events and abnormal operation experience gained in the process of Russian fast reactors operation is given in the paper. The main information refers to the BN-600 demonstration reactor operation. Statistical data on sodium leaks and steam generator failures are presented, and sources of these events and countermeasures taken to avoid their appearance on the operating reactors as well as related changes made in the BN-800 reactor design are considered. In the paper, some features of impurities behaviour are considered in various modes of the BN-600 reactor operation. Information is given on the impurities ingress into the circuits, on abnormal situation emerged in the process of the BN-600 reactor operation and its probable cause. Information is presented on the event related to the increased torque of the BN-600 reactor central rotating column and repair works performed. (author)

[en] In the 13 years since commissioning of the Creys-Malville nuclear power plant, exactly 100 unusual events were recorded on the French and later on the INES scales. The resultant ratio is slightly lower than the French PWR average. This is a noteworthy accomplishment, considering that the plant is a prototype, went through significant design changes, was repeatedly put to test in operating transients and, in addition, holds roughly twice the number of components as a PWR of comparable power. It may be inferred that fast reactors are not more difficult to operate than PWRs, which is also the opinion of most people having taken shifts in both types of reactors. Although Superphenix was labelled a white elephant by public opinion makers, this little known characteristic should remain part of its legacy. In this period 7 events are registered at the level 1 of the French and INES scales, owing either to misconception, material or operational failure. At the level 2 of the scales, 2 events are registered, which is admittedly quite high. The first one was the sodium leakage from and pervasive cracking of the revolving 'drum' of the fuel handling line, in retrospect the result of the choice of steel grade not fully compatible with sodium, which questions the designer's decision making process. The second level 2 event started as a massive air ingress in the primary circuit atmosphere, bringing on a pollution of the sodium up to 15,5 ppm of oxygen (although its significance in terms of corrosion was shown to be minimal). Although this event originated from a maintenance mix-up, it revealed a lack in understanding of sodium chemistry and the inadequacy of the instrumentation. The operational feed-back of the Superphenix reactor was thoroughly combed for clues to potential anomalies by a working group comprising representatives of the operator, utility, designer and R and D bodies. All the gathered information (together with experience gained from other FBRs, most notably PFR and Phenix) was then analysed in periodic project reviews validating the European Fast Reactor project, mainly in the areas specific to FBR technology (referring to Superphenix' second level 2 event, it led to the addition of a gas analyser). That feed-back process, with complementary contributions and mutual checks of designer teams of various backgrounds, allows an optimistic view of EFR's seaworthiness. (author)

[en] The PHENIX Life Extension Project groups together all the actions required to pursue operation of the reactor, particularly with a view to performing irradiation experiments in the framework of the back-end of the fuel cycle programs. As such, it comprises a series of investigations whose objective is to assess the state of the reactor after about one hundred thousand hours of operation. The following points have been particularly investigated: the materials behavior (austenitic, austeno-ferritic, and ferritic steels base metal, welds, heat affected zones) in terms of thermal aging and its effect on mechanical properties, embrittlement, sensitiveness to corrosion (in normal and incidental environment), and radiation effect on the potentially exposed structures. Furthermore, specific programs have been devoted to the assessment of thermo-mechanical response of some particular components. This concerns some types of welds with regard to fatigue or creep fatigue, some parts of large shells with regard to ratchetting and buckling, and main secondary piping. An extensive program was dedicated to the recovery of the thermo-mechanical damage undergone by the structures and its extrapolation to the future. This has led to consider in details thermo-hydraulic effects such as fluctuations in streams and bedding zones. Some intergranular cracking of welded joints had to be closely examined; this was achieved by a research work that has produced important advancements in that field. With the aim of evaluating potential defects, real progresses have been made in the knowledge of large defect's behavior in thin shells. The feedback of the examination and studies was also derived in terms of relevance of manufacturing, exploitation and monitoring conditions. It is believed that this experience will be useful for future design rules. (author)