- Status
- Closed
Objectives
Part A. Development of Accident management procedures on the test facility PSB-VVER-1000 at Electrogorsk
The PSB test facility is the only existing large-scale test facility intended to address the thermal hydraulic behaviour of the coolant system of VVER-1000 plant design.
The OECD support group visited the facility and recommended the completion of the PSB construction in order to perform the experiments needed for thermal hydraulic analysis of relevant VVER-1000 transients and/or accidents and codes validation. PSB facility completion and experiments performance has been requested by Russian Regulatory Authority, Gosatomnadsor.
The specific objectives envisaged in the conception and elaboration of the present project were:
Execution of experiments in the PSB test facility for the identification/verification of thermal-hydraulic phenomena relevant to the evolution of accidents in VVER-1000/V320 reactors and, in particular, to the performance of engineered safety systems and accident management measures;
Application of thermal-hydraulic system codes such as RELAP, CATHARE and KORSAR for pre- and post-test analysis of the PSB experimental results and verification of their predictive capabilities for the simulation of VVER-1000/V320 accident management measures;
Provision of information for the development and/or modification of guidelines for the implementation of VVER-1000/V320 accident management measures.
Part B. Development of a code system for severe accident analysis in RBMK reactors.
The project was intended provide the Russian authorities with a detailed code system capable to evaluate the core behaviour during transients that could lead to extensive fuel melting and, thus, close to conditions for multiple channel tube ruptures. This code system could include Russian models that have already been validated to a certain degree. The validation of this new Russian code system should be done with an upgraded Western code system that includes all the necessary models. In specific areas of severe accident modelling some of the Western approaches would be used in the Russian code system.
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The resulting tool should be able to deal with:
Thermal, mechanical and neutronics behavior of the degraded fuel, and degradation mechanisms of the pressure tube;
Mechanical response of the graphite stacks and of the core during multiple pressure tube rupture;
Thermal hydraulics in the reactor core cavity under severe conditions;
Chemical reactions and other physical phenomena occurring in the core cavity when tube rupture;
Calculation of post accident hydrogen distribution;
Mechanistic fission product transport and retention codes.
The most important expected result of this project was an extensive review, with the new code system, of the existing safety studies related to accident scenarios that challenge the integrity of tube pressure. A formal verification matrix with 10 experiments would be established, suited to the new code system, and to be used in the subsequent qualification process.
Results
Part A. Development of Accident management procedures on the test facility PSB-VVER-1000 at Electrogorsk
The Part A of the contract dealt with the subject ‘Accident Management (AM) in VVER-1000 reactors’ and was finalized to the availability of suitable computational tools for AM optimization studies in those reactor types. This was achieved by planning and executing experiments in the PSB-VVER test facility also addressing the scaling issue and by definitely confirming the capabilities of current system codes. The understanding of the current AM technology in Russian NPP, namely in the Balakovo Unit 3 (chosen as reference NPP) and the critical appraisal of the connected methods were part of the Project activities. Contractors and Sub-contractors for the Part A were UNIPI, EREC, Kurchatov and Gidropress that constitute the technical partners for the conduct of activities.
Part B. Development of a code system for severe accident analysis in RBMK reactors.
The Part B of the contract dealt with the set-up of a chain of codes suitable for the analysis of a variety of accident conditions in RBMK. The attention was focused toward the consequences of individual fuel channel rupture and resistance of the graphite stack to the catastrophic propagation of the rupture through the overall core, addressing the MPTR (Multiple Pressure Tube Rupture) issue. Key expected products from the Project were the demonstration of suitability and the availability from the Project partners of applicable chains of computation tools. These include the codes, the input decks (i.e. the nodalisations) and the boundary and initial conditions. Reference NPP for the activity was the Smolensk Unit 3 NPP. Contractors and Sub-contractors for the Part B are UNIPI and NIKIET that constitute the technical partners for the conduct of activities.
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The demonstration of suitability for the chain of codes and their availability by the partners has been achieved from the execution of the Project. Namely, two chain of computational tools based on (primarily) Russian and Western origin codes, respectively, have been proposed and developed. The major effort was devoted to the development and the qualification of input decks or nodalisations for almost a dozen codes covering the RBMK safety technology areas thermal-hydraulics, neutron kinetics, structural mechanics, nuclear fuel behaviour, fission product generation and transport and chemistry. The Smolensk 3 was selected as reference NPP and calculations predicting accident scenarios were performed although the results should not be taken as valid for licensing or for evaluating safety margins. A roadmap for the project was established and key findings from the various steps are summarized hereafter.
Six safety technological areas were proposed and ten accident scenarios were defined accordingly for the purpose of testing the capabilities of the individual computational tools of the chain of codes. A correspondence was created between safety areas and accident scenarios.
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The capability to model the confinement systems coupled with the primary system was demonstrated. As a main finding, the need to model in detail the complex confinement system for a suitable global safety analysis was emphasized. In addition, the design margins (e.g. margin to lifting the upper core plate) available for the reactor cavity were found consistent with the consequences expected during realistic accident scenarios.
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The interaction between the individual pressure tube and the connected stack of graphite bricks was investigated by adopting an analytical approach and the finite element method. Conditions for graphite failure were characterized as well as the contribution of graphite to the strength of the coupled graphite-pressure tube. The importance of the graphite sealing annuli separating the graphite bricks from the pressure tube was emphasized from the structural mechanics point of view. A methodology was proposed for investigating the realism in the propagation of one pressure tube break to neighbouring pressure tubes.
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The activity related to the fission products behaviour was carried out in two main steps. The first step consisted in the evaluation of the source term at the level of the fuel bundle following the accident scenario of current interest: the availability of an established code technology was proven. The second step dealt with the modelling of the transport of the fission product and the trapping inside the gaps of the graphite stacks and brought to the capability of deriving the radioactivity in the various zones of the confinement and the source term toward the environment. Recommendations for developments, especially in the experimental sectors have been provided based on the capabilities of the codes and the safety needs.