Ex-Vessel Source Term ANalysis (EVAN)
Source Term Assessment at Ex-vessel Stage of Severe Accident
Tech Area / Field
- ENV-MRA/Modelling and Risk Assessment/Environment
- FIR-EXP/Experiments/Fission Reactors
- FIR-MOD/Modelling/Fission Reactors
8 Project completed
Senior Project Manager
Tocheny L V
FSUE "Saint Peterburg Research and Design Institute ATOMENERGOPROEKT", Russia, St Petersburg
- All-Russian Research and Designing Institute of Complex Energetic Technology, Russia, St Petersburg\nInstitute of Safe Atomic Power Engineering Development, Russia, Moscow\nJoint Stock Company «I.I.Polzunov Scientific and Design Association on Research and Design of Power Equipment», Russia, St Petersburg\nResearch Institute of Technology, Russia, Leningrad reg., Sosnovy Bor
- CEA / DEN, France, Gif-sur-Yvette Cedex\nEuropean Commission / Joint Research Center / Institute for Transuranium Elements, Germany, Karlsruhe\nPaul Scherrer Institut, Switzerland, Villigen\nCEA Commissariat a l'Energia Atomique, Cadarache, France, Saint-Paul-lez-Durance\nIRSN - Institut de Radioprotection et de Sûreté Nucléaire, France, Fontenay aux Roses\nCiemat, Spain, Madrid\nGesellschaft für Anlagen und Reaktorsicherheit mbH, Germany, Köln\nVTT Processes, Finland, Espoo
Project summaryThe “Ex-vessel Source Term Analysis” (EVAN) ISTC project includes theoretical and experimental research of the processes affecting the late phase fission product release to the PWR containment atmosphere at the ex-vessel stage of the hypothetical severe accident with core meltdown.
This stage is characterized by corium relocation outside the reactor pressure vessel into the core catcher/reactor cavity and a need for long-term heat removal from the containment (with possible loss of integrity, increased leak rate, and internal overpressure). At this stage, the fission products are released to the containment atmosphere (and thus made available for release to the environment) from the core melt located in the reactor cavity/lower containment compartment, along with various secondary sources like contaminated solution in the containment sump and FP deposited at the surfaces of process equipment and building structures.
Ex-vessel core melt fission products release is affected by design features and the accident management strategy. Partly, the oxidation degree of the melt is important for the low-volatile oxidizing FP release, especially Ru, Ba, and Mo, while water supply onto the melt can effectively reduce FP transport from the melt to the atmosphere but generate a secondary source of soluble FP (Cs, I, Ru) from contaminated solution boiling at the melt surface.
Fission products deposited at the in-vessel stage in the primary circuit (including heat transfer surfaces of steam generators) can be resuspended/revaporized and be an important source in a long-term perspective. Investigation of aerosol transport processes (deposition, resuspension, and revaporization) is of practical interest for characteristic primary circuit geometries (vertical, horizontal sections, tube bends and so on).
For various severe accident management strategies, containment sump solution is often used for long-term heat removal from the melt and from the containment. Investigation on how different chemical species in the sump solution (boric acid, Fe oxides and organic forms) affect concentrations and partitioning of the iodine species is very important for radioiodine source term predictions.
The project features possibility to obtain data directly applicable to existing VVER-type designs, and data allowing comparison with PWR analogues. Necessary quality of the project is achieved through the combination of intellectual input from competent Russian scientific sub-teams having experience both in EU and ISTC research projects, using unique experimental setups and up-to-date computer codes, elaborated ISTC organizational framework, and constant interaction with foreign collaborators interested in the project outcome.
Four main work packages are included in the project:
- WP1/Task 1. Analysis of results of severe accident scenarios calculations for various NPP with PWR/VVER. Participants: SPAEP, IBRAE. The goal of the analysis is to determine the representative (envelope) boundary ranges for the fluid parameters in the reactor plant and containment, for the core melt parameters, FP aerosol characteristics, boundary conditions at the surfaces of structures and equipment, sump solution chemical composition, dose rate level, and other parameters necessary for specification of the experimental conditions for WP2-4. Further on, the review of the present modeling capabilities for up-to-date codes (RATEG/SVECHA, ICARE2, ASTEC) is to be performed, and applicability of the obtained experimental results for computer models validation is to be justified.
- WP2: Theoretical and experimental investigations of fission product release from the molten pool/core catcher.
Task 2. Experimental investigations of fission product release from the molten pool/core catcher. Participant: NITI. Tests for model corium compositions are carried out within the approved test matrix, and the following is to be determined: low volatile fission product release from the molten pool during its transition from sub-oxidized to fully-oxidized state and fission product release from the molten pool with water supply onto the melt surface (for pure water and water contaminated with FP species).
Task 3. Theoretical and numerical modeling of fission product release from the molten pool/core catcher. Participant: IBRAE. Computer codes RATEG/SVECHA, appropriate models of ICARE2 (IRSN, France) and, if collaborators provide such an opportunity, RELOS are to be used for numerical estimations.
- WP3. Theoretical-experimental investigations of deposition, transport and revaporization of aerosols in the primary circuit pipes.
Task 4. Experimental investigations of aerosols transport in the primary circuit pipes. Participants: CKTI. Tests for different aerosol types are carried out within the approved test matrix. The experimental programme would take a typical aerosol and examine their deposition and resuspension behavior in horizontal or vertical tubes for the specified Re ranges.
Task 5. Theoretical and numerical modeling of deposition, transport and revaporization of aerosols in the primary circuit pipes. Participants: IBRAE, SPAEP. Aerosol transport calculations are performed with use of the reference models implemented in RATEG/SVECHA code. Cross-validation of results obtained by means of CFD-codes is carried out (SPAEP).
- WP4. Assessment of containment media parameters impact on content and proportioning of the volatile iodine species in the containment atmosphere.
Task 6. Experimental investigations of containment parameters impact on content and proportioning of the volatile iodine species. Participants: VNIPIET (in collaboration with NITI, department 5). Experiments on the effect of impurity complex, coming out into the containment water medium under emergency conditions, on the content of I2 in the solution and volatile iodine species in the gas phase would be performed within the approved test matrix.
Task 7. Theoretical and numerical modeling of containment parameters impact on content and proportioning of the volatile iodine species. Participants: VNIPIET, SPAEP. VNIPIET adapts the iodine species behavior model developed earlier for the containment atmosphere under emergency conditions to the conditions of the experimental tests (task 6) in order to compare calculation results with the experimental ones. VNIPIET performs numerical modeling of the experimental test modes, develops the model for accident environmental radioiodine source term assessment, the algorithm and computer programme for numerical description of the experimental tests. SPAEP analyses results and adapts the iodine species behavior model and computer code to accident environmental radioiodine source term assessment.
The scope of the project may be extended by additional adaptation and validation of computer models, continuation of experimental programmes and radiological consequences calculation for selected severe accident scenarios.
The EVAN project is an applied research. Theoretical and experimental research results for assessment of consequences of fission product release to the PWR containment atmosphere at the late stage of severe accidents can be used for safety assessment of both existing and new NPP designs, including probabilistic safety analyses of 2 and 3 level, for development of severe accident management strategies and for emergency planning analysis. The project features possibility to obtain data directly applicable to new designs of NPP with VVER-type reactors, providing severe accident management measures.
Supporting letters have already been received from collaborators of VTT, GRS, and ITU; if necessary, in addition supporting letters from PSI, CEA and IRSN may be presented. The foreign collaborators are expected to partly provide material for analysis within the WP1/task 1, to participate in developing the text matrices, and in discussions on modeling issues for WP2-4. Also joint publications can be planned, proposals for modifications of models and codes, and for future cooperation.
The project is implemented with the use of up-to-date computer codes and unique experimental setups applied by organizations-participants. Approved theoretical and experimental base techniques are used.
The project promotes ISTC goals to let Russian weapon scientists and specialists to apply their knowledge to peaceful applications, to facilitate their integration to the global NPP accident analysis community, and also to support the applied research in the field of environment protection, energy generation, and nuclear safety. The project is also important for nuclear power engineering and complies with the European Framework Programmes goals. The proposed work packages for the project were approved by EU-ISTC CEG-SAM and SARNET experts.