T. P. Speis
U.S. Nuclear Regulatory Commission, Washington, DC 20555

In October 1987, the U.S. Nuclear Regulatory Commission proposed that a joint international cooperative program be formed that would be sponsored by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (NEA-OECD) to conduct further investigations of the potential damage to the TMI-2 reactor vessel lower head during the accident. This proposal was accepted by NEA-OECD and the project was initiated in 1988. The program was completed five years later. This paper discusses the results from this program and the related conclusions. It specifically addresses the findings regarding the extent and type of damage to the vessel lower head and the margin of structural integrity that remained in the vessel during the accident; the conditions for potential failure (which apparently were never reached); enhanced cooling mechanisms of the debris and lower head not currently accounted in severe accident analyses; and insights for severe accident management strategies. Finally, the results/conclusions from the program have pointed to a number of areas where further studies, including appropriate experiments, can add to our understanding of the conditions under which the reactor vessel can be cooled and its integrity maintained during a severe accident. Such efforts are already underway and they will also be addressed in this paper, including their specific goals and significance.


V. Sidorenko, V. Voznesenski*, N. Fil**, E. Tsygankov*
Ministry of Russian Federation on Atomic Energy, Moscow, Russia
*Russian Research Center Kurchatov Institute, Moscow, Russia
**EDO "Gidropress", Podolsk, Moscow Region, Russia

During 30 years passed since the startup of the first WWER reactor in Novovoronezh, 55 WWERs have been put into operation at 17 NPPs sited on the ex-USSR territory, in the East European countries and Finland. As of 01.01.95, the WWERs of different designs have the total reactor years exceeding 600. Beginning from the first WWER a great importance was attached to elimination of phenomena and events that could cause a severe accident of aggravate development of an accident making it eventually a severe one. The first NPPs with the WWERs were built in the regions with the least possible external impacts (in non-seismic regions, far away from the aviation air corridors, harmful production etc.). In designing the WWERs the materials and structures resistant to rapid destruction were used. Potentially harmful phenomena such as reactor vessel embrittlement, initiation of cracks in the SG headers etc., were under continuous observation during the operation. As the experience showed, this permitted timely measures to be taken against propagation of failures in severe accident. The WWER equipment characteristics (secondary, water inventory in the SG, pressurizer capacity etc.) were chosen so that the accident would slowly develop, and the operational personnel could take effective measures for its mitigation. An example of successful employment of the WWER capabilities by the personnel is the minimization of the consequences of a fire at the Armenian NPP resulting in complete loss of the SG feed water. A serious impetus to redouble efforts at investigating severe accidents was the Three Mile Island and Chernobly accidents. The normative documentation adopted in Russia in the late 80s stipulates the proof of the low probability of severe accidents in the designs. For operating NPPs special guides and instructions, determining the personnel's actions during the beyond-design accidents, were worked out. The measures for increasing their safety stipulate the modernization of the old NPPs and introduction of new systems and equipment for of the beyond-design accidents management. The efficiency of these measures is substantiated by the use of the PSA methods. The designs of new power units with LWR envisage special measures both for the elimination of severe accidents and mitigation of their consequences. In the design of the NPP with WWER-640, which continue the traditional WWER line, passive systems are used for the long-term core cooling down during the accident. In spite of the low core melting probability measures are taken to keep the melt within the reactor vessel or at least inside the containment. The WPBER-600 design extends the line of safety enhancement, which has been realized in the design of the Soviet nuclear district heating plant. The integral layout of the NSSS equipment, low heat loss, second (guard) vessel make the core melting practically impossible. However, just as in the WWER-640 design, the containment is provided with an additional facility for trapping the melted core. As one of the measures for the most reliable protection of the population against the consequences of a severe accident, the underground NPP siting is considered by the Russian specialists. The Russian investigations of processes occurring in the severe accidents and measures for their management are currently closely connected with the international efforts in this direction. One of the international severe accident projects directed to the confirmation of the possibility of melt retention inside the reactor vessel (RASPLAV) is being carried out on the basis of the experimental complex now under construction at the RRC "Kurchatov Institute". The close connection of the new engineering approaches undertaken for LWR safety enhancement in each nuclear country with the international program of investigations permits the desired results to be reached in an optimal way.


E. O. Adamov, Yu. M. Cherkashov, Yu. V. Mironov and Yu. M. Nikitin
Reactor and Development Institute of Power Engineering, Moscow, Russia

E. B. Burkalov and N. E. Kukharkin
Reactor Scientific Center "Kurchatov Institute", Moscow, Russia

Key features that affect the progression of severe accidents in RBMK reactors are described. Estimates of plant responses are presented for severe accidents initiated by a positive reactivity insertion, a loss of electrical power, and a loss of heat transport circuit integrity. Existing and potential measures to mitigate the severe accidents are discussed.

It is noted that Probabilistic Risc Assessment techniques and criteria established for Western Light Water Reactors may not be appropriate for the channel-type RBMK reactors. Severe accident issues in RBMK are related to developing preventive and mitigation measures which make effective use of the operating experience as well as the knowledge gained from international programs.


R. A. Brown, C. Blahnik and J. P. Karger
Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario, Canada

In 1992 at the Munich G-7 Summit Meeting the safety of Soviet-designed reactors in the Former Soviet Union (FSU) countries was raised as a priority issue and the G-7 governments established both Bi-lateral and Multi-lateral programmes of assistance to these countries. A Nuclear Safety Account (NSA) was established by 13 participating governments to fund a series of grants to selected nuclear power plants, the fund being administered on their behalf by the European Bank for Reconstruction and Development. As a member of the G-7 Canada contributed to this multinational safety fund.

The NSA funds both technical improvements at selected sites as well as the project management teams needed to develop specifications, contracts and review bids for these safety improvements. AECL has concentrated its efforts on RBMK reactors because of its expertise in the design and construction of CANDU channel type reactors which share certain similarities with the RBMK. Specifically it is participating in two programmes funded by EBRD for the Ignalina NPP (INPP) in Lithuania; one related to the design of a diverse shutdown system and associated activation sensors and logic, the second related to the preparation of a western styled safety analysis in conjunction with Vattenfall from Sweden and Stone & Webster Engineering Corporation from the United States, the RBMK Chief Design organisation and INPP staff. The objective of this latter study is to apply western type tools and methodology to an extensive analysis of both the station operation, its safety management practices and the accident analysis. The work is currently underway and the paper will describe the process and progress to date.

In 1992 the Canadian Government announced the $30M CDN Canadian Nuclear Safety Initiative aimed at providing regulatory, design, operational and utility support for RBMK reactors in Russia and Lithuania. As part of this programme AECL developed a Nuclear Safety and Engineering Programme (NSEP) which has four parts.

  1. An operating team initially residing at one RBMK station and the rotating to others, aimed at introducing western operating practices and safety culture to NPPs with RBMK reactors.
  2. An engineering and safety team located in Moscow to work with with RBMK design institutes to address some short term safety design improvements and introduce Canadian expertise in the area of fuel channel inspection techniques.
  3. A programme which will be carried out both in Canada and Russia.
  4. An operations staff exchange programme.
As part of the CNSI Canada participated in the 8 nation International RBMK Safety Review which undertook an extensive review of the design and operation of an RBMK based primarily upon Unit 3 at Smolensk. The results of this review were reported in June 1994 and the paper will discuss highlights in various areas with particular emphasis on the accident analysis.


G. M. Frescura

R. J. Barrett

This paper summarizes the results and conclusions of a study conducted by the OECD/NEA on the approach taken by Regulatory Organizations to deal with severe accident issues. The information for the study was collected through a questionnaire distributed to the OECD/NEA Member countries. This information was then discussed in depth by a committee consisting of senior representatives of Regulatory Organizations. The paper shows that most countries have taken substantive steps to reduce the risk resulting from postulated severe accidents via measures that include prevention, mitigation and accident management. Practical considerations have resulted in most countries placing greater emphasis on prevention of severe accidents and implementation of accident management. However, design modifications have been done or are planned in a number of countries. For future reactors, most OECD/NEA countries require specific consideration of severe accidents in the design process, however severe accidents are not included in the design basis set.

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