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«Item 7b Severe Accidents Related Issues Preliminary Monitoring Report Report to the Federal Ministry of Agriculture, Forestry, Environment and Water ...»

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The review of regulatory approaches included the position of US Nuclear Regulatory Commission (US NRC) on PWRs with large dry containments of US design, of licensing authorities within the EU such as the French IRSN, the German Reaktorsicherheits-kommission (RSK) and of Western European technical support organizations (TSO) such as FrenchGerman consortium GRS/Riskaudit.

The US NRC for their ruling has determined that dry containments offer such large safety margins that for the existing US plants which implement WOG Westinghouse SAMGs, the hydrogen combustion and direct containment heating issues can be considered resolved.

Licensing authorities within the EU apply various approaches in their respective countries.

Detailed findings are presented in Annex A to this report.

The analysis of the actual situation in Temelín NPP showed that the plant is provided with a large dry containment, with comparable features to those in use with US pressurised water reactors (PWRs), but differing in some geometrical aspects, in particular in the shape of steam generator boxes, which are horizontal and not vertical as in the PWRs. This can result in slower dispersal or propagation of hydrogen released during not only the in-vessel accident phase and in possible increase of its local concentration to the values higher than in typical PWR containments. On the other hand, compared to the US plants, Temelín has the advantage of the installed hydrogen recombiners, which deplete hydrogen by combining it with oxygen to water over long-term operation. Detailed findings are presented in the report and summarized in Section 4.3.

ETE Road Map - Preliminary Monitoring Report – Item 7b: Severe Accidents Related Issues 5

III.4 Code calculations for assessment of selected sequences

The initial activity in this step was identification and selection of accident sequences leading to an immediate thread to the integrity of the containment either due to Core Concrete Interaction failure or due to Hydrogen formation leading to a detonable gas mixture within the containment eventually resulting in detonation pressure bulid up. Both these cases were analysed using MELCOR, a computer code suitable to analyse to the detail required severe accidents and determine consequences. The MELCOR analyses have been done within both the horizontal segment and the vertical segment of PN7. The WWER 1000 MELCOR input deck available for Kozloduy NPP (KNPP) plant has been modified taking into consideration specifics of Temelín insofar as they were known. The results have been assessed to verify which of the problem issues need to be specifically addressed for the development of SAMGs.

The calculations made in PN7 covered more than twelve scenarios with some variants aimed at checking sensitivity of results to the assumptions adopted in calculations. Several points in which no sufficient data were available to judge Temelín statements have been identified, but generally the agreement of PN7 calculations results with those of TACIS programme for WWER 1000 NPPs and with Czech results for Temelín was reasonable.

As it was recognized that besides general review there is a need for in-depth analysis of the topics connected with hydrogen hazards, the problems of hydrogen generation and transient local distributions were addressed in MELCOR analyses of three scenarios plus a special 3dimensional GASFLOW analysis for a specific scenario, providing insights into nonuniformities of hydrogen distribution during the phase of most intensive hydrogen release and the related hazards.

III.5 Identification of relevant steps in the development of SAMGs

With consideration of the severe accident sequences, their outcomes, probabilities, and possible mitigation measures, the steps relevant to the development of SAMGs have been established, including specifics to be addressed in Severe Accident Management Guidelines (SAMGs).

Based on experience of the project team in evaluation and validation of SAMGs, the specifics to be investigated in this area have been identified.

The issues have been listed, which need to be addressed in relation to the adaptation of SAMGs and training of plant staff in the use of SAMGs.

The Severe Accident Management requirements have been analysed as indicated by international practice and formulated in the Westinghouse Owners Group (WOG) SAMGs. The available information on the Temelín approach and the SAMG development status was reviewed. The result showed that the approach followed by Temelín corresponds to good international practice and after completion - expected by the end of 2004 - the SAMGs in Temelín NPP should be equivalent to those in other plants using the WOG approach. Some minor points needing further monitoring have been found as identified below.

6 ETE Road Map - Preliminary Monitoring Report – Item 7b: Severe Accidents Related Issues

III.6 Verifiable line items

The objective of this task was to break down the overall subject into the line items, which could then be verified for completeness and compliance with the accepted international practice. This task was the “road map” for the whole project. On the basis of all the analyses, which are discussed above, the project team has identified all necessary elements, which are of interest in developing and implementing an acceptable severe accident management program including its verification. The list of Verifiable Line Items (VLIs) covering more than 240 questions in 40 topical areas has been developed, covering both SAMG development and accident sequences in the plant. It was the basis for consolidation of the information achieved during the joint workshop with representatives of the Safety Authority and the Temelín NPP operator, which took place in June 2003 in Prague.





III.7 Specialists Workshop

The preparatory activities for the workshop included the development of briefing material and a briefing session for the Austrian delegation, proposing experts to participate in the workshop as well as participation in the workshop. The compliance and differences with the state of the art practices have been identified and commented on for their safety significance.

A list of documents was prepared, the Specific Information Request, that considered to contain that kind of information required to provide profound answers to the VLIs.

In the Workshop the Czech side presented a set of twelve papers, which together with the discussion sessions made it possible to determine the answers to most VLIs, as shown in the report. Some methodological aspects have been left unanswered due to the limitations of time available and the complex nature of the phenomena involved in severe accidents analysis. Nevertheless, the information accumulated in the preparation of the workshop and obtained during the workshop is sufficient to formulate a coarse picture of the Temelín NPPs preparation to cope with severe accidents, (given the fact that Sections 3.1.4, 3.1.5, 3.2.3, 3.2.4, 3.3.1, 3.3.2, 3.3.3, 3.3.4, 3.3.5, 3.3.6, 3.4.3, 3.6.2, and 3.6.5 of the main report discuss areas where the information presented was evaluated as insufficient; in addition, see Section

1.4 of the main report for an explanation of the assessment framework). In some cases the Austrian delegation requested further written information and the Czech side provided it soon after the workshop. This information was subsequently used to repeat a selection of the calculations performed within PN7 project with updated characteristics of basemat concrete and hydrogen recombiners in the Temelín NPP. The final conclusions in the report are based on those updated calculations.

IV. Main findings IV.1 Regulatory approach and practice The Czech Nuclear Regulatory Authority (SUJB) has required the plant to prepare and accomplish a program to deal with BDBAs, including estimation of plant vulnerabilities, proposed accident management procedures and the schedule of their implementation. The targets set by SUJB for severe core damage frequency and for large off-site releases are to underscore 10-4 and 10-5 per reactor year, respectively, which is consistent with the INSAG targets for existing NPPs.

The responsibility for development of SAMGs is left to the utility. The regulatory body defines acceptance criteria and provides guidance to Temelín NPP, leaving enough flexibility for potential candidate actions to address specific challenges.

ETE Road Map - Preliminary Monitoring Report – Item 7b: Severe Accidents Related Issues 7

IV.2 Temelín programme of severe accident management

The development and implementation of Temelín SAM programme has not been finalized, however, it is well advanced.

The overall concept and approach to development/implementation of SAMG package was found to reflect the current good practice in the SAM area. The selection of plant specific SAM strategies has been based on the well-established generic approach developed by Westinghouse Owners Group. These generic strategies have been adapted to Temelín plant conditions based on a systematic process that reflects the current state-of-the-art in this area.

The programme is supported by severe accident analysis and plant specific probabilistic safety assessment (PSA). However, there were some instances when the existing results of SA analysis were not properly incorporated into the PSA. It should be noted that also some SAM strategies, apparently the most recent, are not well supported by severe accident (SA) analysis. The interface between the PSA team and thermal hydraulic analysis team needs improvement.

The calculation tools used for SA analysis are similar to those used worldwide for the purpose of SAM and the team that has been responsible for calculations is qualified. The existing analyses provide a reasonable basis for understanding plant specific vulnerabilities to severe accidents and the identification of AM strategies. Some of the existing analyses are old and do not necessarily reflect the current plant status and state-of-the-art in the area of SA codes, modelling and simulation, in particular with respect to hydrogen distribution within the containment and with respect to the discharge of molten core material in case of a pressure vessel defect. The plant is planning to improve these analyses using more current codes and improved modelling concepts.

The PSA study includes Level 1 and 2. The first version of PSA has been reviewed during an IAEA mission and the resulting recommendations are reported to be incorporated into the upgraded study. However, the upgraded PSA is still not finalized. Generally, the 1996 PSA study was developed in compliance with the current state-of-the-art, and the updated analysis was intended to address IAEA comments and the as-built design of the plant. The PN7 team has observed some deficiencies, but they are not expected to have significant impact on the final conclusions with regard to SAM strategies. The existing results have been used in the development of SAMG strategies and setting up priorities in the execution of strategies.

Westinghouse in close co-operation with plant staff has developed a plant specific SAMG package. The contents, structure, and format of plant specific SAMG, which were shown at the Workshop, have been found to reflect the current state-of-the-art practice. This package is currently under internal review and translation into Czech language.

Organizational arrangements related to SAMG have not been finalized yet. Although the upgraded ERP Emergency Response Plan has been submitted to SUJB for approval, the updated version of the Emergency Operating Procedures including transition points to SAMGs need to be developed and implemented. Some concerns can be raised in the definition of responsibilities/authorities for determination and approval of an intentional release of radioactive material during a SA. The staffing of SAMG Evaluation Group within the Technical Support Centre is another issue that is not clear enough..It is recommended that the Austrian Government addresses these aspects in the ongoing joint monitoring process on technical level.

The plant has properly considered all further steps of SAMG implementation including validation and training and plans for their execution are being developed. Based on the available knowledge all the related plant arrangements are considered adequate. Little is known also about the training and refreshing courses of SAM staff and the related schedules for implementation. Therefore, it would be welcome if the related activities would be subjected to further joint monitoring in the framework of the pertinent bilateral Agreement between Austria 8 ETE Road Map - Preliminary Monitoring Report – Item 7b: Severe Accidents Related Issues and the Czech Republic. It should be noted that proper evaluation of the SAMG package including the supporting analyses would require detailed investigations that involve specialized expertise and considerable effort. Such an evaluation was beyond the scope of PN 7 project.

Therefore, it would be very desirable to have detailed aspects of SAM development and implementation addressed by qualified independent external reviewers. It is known that the plant management and SUJB seriously consider having an independent review of SAM (i.e.

a IAEA RAMP mission).

IV.3 Technical measures available in Temelín for SA management.

One of the main areas of hazards due to severe accidents is that of primary to secondary circuit leakages, since such leakages involve loss of coolant accidents with the leak point situated outside the containment. In case of such an accident all four barriers preventing radioactivity release to the environment can be lost simultaneously. Both contemporary regulatory guidance and industrial practice stress the necessity to avoid large PRISE events. In Temelín the hazards involved in primary to secondary leakage (PRISE) accidents are well recognized, the appropriate strategies developed and the technical means are provided to cope with PRISE events.

Another potential hazard is connected with long term complete loss of electric power, both from outside sources and from emergency diesel generators installed at the NPP (station blackout). In such a case the means of heat removal from the reactor are lost, except for gradual evaporation of water, first in the secondary, then in the primary coolant circuit. If this situation persists for several hours, the coolant in the core evaporates, the core dries out, and will be damaged.



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