Assessment of the potential consequences of a large primary to secondary leakage accident – Final Report

The present paper discusses one of the IAEA’s Coordinated Research Projects (CRPs). The CRP was started in 2003 to evaluate complex phenomena of primary to secondary leakage (PRISE) accidents for WWER-440 reactors. The first Research Coordination Meeting (RCM), held in March 2003, identified the possible consequences of PRISE accidents (radioactive release to the atmosphere, pressurized thermal shock, boron dilution, loss of integrity of secondary systems and severe accidents) and designated six task groups to evaluate these, as well as uncertainties associated with PRISE analyses. The second RCM, held in March 2004, discussed the preliminary results of each task group and addressed the main safety concerns related to PRISE phenomena as well as providing recommendations on modelling for PRISE analyses and on operator actions. The third RCM, held in March 2005, discussed the results of the work performed in 2004. The CRP was concluded in 2005. Publication of the final results of the CRP is planned as an IAEA TECDOC. The paper provides a review of the final results of the project

Safety Analysis of WWER-440 Nuclear Power Plants: Potential Consequences of a Large Primary to Secondary Systam Leakage Accident

In 1990 the IAEA initiated a programme to assist the countries of central and eastern Europe and the former Soviet Union in evaluating the safety of their first generation WWER-440/230 nuclear power plants. The main objectives of the programme were to identify major design and operational safety issues; to establish international consensus on priorities for safety improvements; and to provide assistance in the review of the completeness and adequacy of safety improvement programmes. The scope of the programme was extended in 1992 to include RBMK, WWER-440/213 and WWER-1000 plants in operation and under construction. The programme is complemented by national and regional technical cooperation projects. The programme is pursued by means of plant specific safety review missions to assess the adequacy of design and operational practices; Assessment of Safety Significant Events Team (ASSET) reviews of operational performance; reviews of plant design, including seismic safety studies; and topical meetings on generic safety issues. Other components are follow-up safety missions to nuclear plants to check the status of implementation of IAEA recommendations; assessments of safety improvements implemented or proposed; peer reviews of safety studies, and training workshops. The IAEA is also maintaining a database on the technical safety issues identified for each plant and the status of implementation of safety improvements. An additional important element is the provision of assistance by the IAEA to strengthen regulatory authorities. The programme implementation depends on voluntary extrabudgetary contributions from IAEA Member States and on financial support from the IAEA Regular Budget and the Technical Cooperation Fund. For the extrabudgetary part, a Steering Committee provides co-ordination and guidance to the IAEA on technical matters and serves as a forum for exchange of information with the European Commission and with other international and financial organizations. The general scope and results of the programme are reviewed at relevant Technical Cooperation and advisory group meetings. Guidance relevant to water moderated, water cooled WWER-type reactors and graphite moderated, boiling water RBMK type reactors has been developed within the IAEA’s Extrabudgetary Programme on the Safety of WWER-and RBMK nuclear power plants. To a certain extent, accident analysis is also covered in several publications of the IAEA Safety Standards series, for example in the Safety Requirements on Safety of Nuclear Power Plants: Design (NS-R-1) and in the Safety Guide on Safety Assessment and Verification for Nuclear Power Plants (NS-G-1.2). Consistent with these publications, the IAEA also developed a Safety Report on Accident Analysis for Nuclear Power Plants (Safety Reports Series No. 23). A Coordinated Research Project (CRP) on Assessment of the Interfaces between Neutronic, Thermohydraulic, Structural and Radiological Aspects in Accident Analysis was implemented from 2003 to 2005 to comprehensively evaluate a complex accident scenario within the framework of the IAEA subprogramme on Development of Safety Assessment Methods and Tools. Twelve organizations from eight Member States participated in the CRP. This report provides a summary of the CRP. Results, recommendations and conclusions resulting from the IAEA programme are intended only to assist national decision makers who have the sole responsibility for the regulation and safe operation of their nuclear power plants. Moreover, they do not replace a comprehensive safety assessment which needs to be performed in the framework of the national licensing process

Thematic network for a Phebus FPT1 international standard problem (THENPHEBISP)

The THENPHEBISP 2-year thematic network started in December 2001, and was concerned with OECD/CSNI International Standard Problem 46, itself based on the Phebus FPT1 core degradation/source term experiment. The aim was to assess the capability of computer codes to model in an integrated way the physical processes taking place during a severe accident in a pressurised water reactor, from the initial stages of core degradation, the fission product transport through the primary circuit and the behaviour of the released fission products in the containment. ISP-46, coordinated by IRSN/DRS Cadarache, attracted 33 participating organisations, from 23 countries and international bodies, who submitted 47 base case calculations and 21 best-estimate calculations, using 15 different codes

Thematic Network for a Phebus FPT1 International Standard Problem (THENPHEBISP)

The THENPHEBISP 2-year thematic network was concerned with the OECD/CSNI International Standard Problem No. 46, iteself based on Phebus FPT1 core degradation/source term experiment. The main aim was to assess the capability of computer codes to model in an integrated way the physical processes taking place during a severe accident in pressurized water reactor from the initital stages of core degradation, the fission product transport through the circuit and the behaviour of aerosol in the containment. 33 organisations from 23 countries and international bodies participated on the ISP.JRC.F.4-Nuclear design safet