Fuel Assemblies under Severe Accident
Study of Fuel Assemblies under Severe Accident Top Quenching Conditions in the PARAMETER-SF Test Series
Tech Area / Field
- FIR-ENG/Reactor Engineering and NPP/Fission Reactors
- FIR-EXP/Experiments/Fission Reactors
- FIR-MAT/Materials/Fission Reactors
- FIR-MOD/Modelling/Fission Reactors
8 Project completed
Senior Project Manager
Tocheny L V
NPO Lutch, Russia, Moscow reg., Podolsk
- Nuclear Safety Institute, Russia, Moscow\nOKB Gidropress, Russia, Moscow reg., Podolsk
- Hungarian Academy of Sciences / KFKI Atomic Energy Research Institute, Hungary, Budapest\nForschungszentrum Karlsruhe GmbH, Germany, Karlsruhe\nEDF - Electricitè de France, France, Clamart\nEuropean Commission / Joint Research Center / Institute for Transuranium Elements, Germany, Karlsruhe\nCEA / DEN / Direction du Soutien Nucleaire Industriel, France, Gif-sur-Yvette Cedex\nPaul Scherrer Institut, Switzerland, Villigen\nGesellschaft für Anlagen und Reaktorsicherheit mbH, Germany, Köln\nInstitute for Nuclear Research and Nuclear Energy, Bulgaria, Sofia
Project summaryIn terms of consequences the most serious accident at NPP with WWER (PWR) is beyond design basis accident (BDBA) with loss of coolant (LOCA) – severe accident (SA), that could lead to core melting, damage of reactor plant (RP) vessel, release of hydrogen, release of fission radioactive products into the containment and the environment.
The main method of SA analysis is the numerical modeling using computer codes. The complexity, consistency of the physical processes and phenomena, accompanying the accident scenario, including the stage of temperature escalation and the core reflooding, require the comprehensive verification of the calculated models. Moreover, it is necessary to have a good representation of the core behavior control in the course of accident and of possible techniques of its cooling down to make the justified solutions on accident management and bringing the reactor into a safe state. In this respect, the study of the initial stage of BDBA with the investigation of a possibility of cooling down the overheated core as a possible way of accident management is of particular importance. Detailed data on fuel rods behavior under the conditions of reflooding are required both at the stage of elaboration of symptom-oriented EOP and SAMG, and during verification of the proposed procedures. Obtaining such information requires, in its turn, the detailed experimental studies of fuel rods assemblies, including studies at ex-reactor test facilities.
In general cases the reflooding of WWER core can be performed by water injection from “bottom” – into the space under the core, from “top” – into the space above the core, and in a combined way – “from top and bottom”.
In the considered proposal, that logically proceeds from a number of experiments performed earlier at СОRA and PARAMETER facilities, a task is formulated on studying the behaviour of the fuel assembly (FA) under the severe accident conditions with top flooding as the less studied.
The analysis of results of PARAMETER-SF1 test, performed at the test facility PARAMETER within the framework of ISTC project #3194 using the 19-fuel rod model FA of WWER, completed with standard structural materials of WWER-1000, under simulated conditions of severe accident, including the stage of low-rate top flooding with water, allowed to identify the following typical processes:
- During top quenching it was observed that there was quick cooling (during 3-5 s) of the model assembly upper part, as well as water blockage in the assembly middle part, fast assembly heating-up, damage of its middle part with water bypassing and cooling of the assembly lower part in 400-600 s while moving the cooling front from bottom to top;
- In the course of the assembly degradation the oxidation and melting of fuel rod claddings occurred, as well as melting flowing down to the assembly middle part, accompanied with its heating up and fuel rod cladding damage in this area, solidification of melting (U, Zr, O) in the assembly cold part. No regions of considerable damage of fuel pellets (debris) were observed that is caused by presence of the frame of intact heaters of fuel rod simulators.
The obtained results showed the necessity of more detailed study of the effects observed. Due to this fact the objective of the proposed Project is the study of behaviour of two 19-fuel rod model FAs of WWER-1000 reactor completed with standard reactor structural materials, and namely, with fuel rod claddings of alloy Zr+1%Nb, fuel pellets of uranium dioxide, spacing grids and shroud of alloy Zr+1%Nb, under the conditions of the initial stage of severe accident with top quenching.
The project is executed by three organizations:
- FSUE EDO “GIDROPRESS” – the leading organization of Federal Atomic Energy Agency on NPP safety justification;
- IBRAE RAS – the leading institute of Russian Academy of Sciences engaged in safety analyses of nuclear power objects;
- FSUE SRI SIA “LUCH” – the organization possessing a unique complex of test facility PARAMETER, with experience in performing the ex-reactor tests of model FAs of WWER-1000 followed by material studies.
The following results are expected in the course of the project execution:
- obtaining and systematization of information on FA behaviour under the conditions of the initial stage of sever accident with top quenching;
- study of moving the cooling front of the assembly heated up to temperature 1800 oС WWER under top flooding;
- broadening the database on verification of severe accident computer codes (SOCRAT/В1, ATHLET, ICARE-CAТHARE, etc.).
The obtained results can be used for safety justification of WWER and PWR type reactors, as well as for elaboration of methods and means of the control and protection system, capable to fulfill their functions under accident conditions.
The scope of activities within 24 months includes preparation and performing the two ex-reactor experiments (PARAMETER-SF test series) at the test facility PARAMETER in FSUE SRI SIA “LUCH” – the tests of two model FAs of WWER-1000 under severe accident continuations with top quenching that is a continuation of studies of FA behaviour in experiments PARAMETER-SF1 and SF2 under ISTC project #3194:
- PARAMETER-SF3 – tests of the model FA with 18 heated and central unheated fuel rods under the conditions of severe accident with top quenching at the assembly temperature less than 1600oС;
- PARAMETER-SF4 – tests of the model FA with 16 heated fuel rods and 3 passive fuel rods, located in the second row, with top quenching at temperature 1800oС.
Besides, the proposed complex of activities will include the post-test material analysis of the assembly, tested in the experiment SF2 under the conditions of the initial stage of severe accident with top and bottom flooding.
The proposed methodological approach to the Project execution is provided by:
- elaboration of a realistic scenario of experiments on the basis of numerical simulation of the initial stage of severe accident at RP of WWER-1000 using computer codes TECH-M and MELCOR;
- calculated simulation of the experiments with the use of computer code package SOCRAT/В1;
- completing the fuel assembly simulator with the standard structural materials of fuel rods and FA of WWER-1000 (fuel rod claddings of alloy Zr+1%Nb, fuel pellets of uranium dioxide, spacing grids and shrouds of alloy Zr+1%Nb);
- simulation of residual power in fuel rods with indirect heating up the fuel rod by the internal electrical heater.
The proposed Project meets the objectives and tasks of ISTC because its execution will involve scientists and specialists of the defense industry, and the results of the Project execution will contribute to the increase in reliability and safety of nuclear power reactors both under operation and those under design.
The joint work with foreign collaborators from FZK, GRS, EdF, IRSN, CEA, which is being fruitfully fulfilled within the framework of Project #3194, will be continued during execution of the Project. The continuation of the tests of PARAMETER-SF series was approved by the participants of the 10-th meeting of СEG-SAM, held on September 4-8, 2006, in Kurchatov, Kazakhstan Republic, and of the 12-th Symposium QUENCH, held on October 23-29, 2006, in Karlsruhe, Germany.
The International Science and Technology Center (ISTC) is an intergovernmental organization connecting scientists from Kazakhstan, Armenia, Tajikistan, Kyrgyzstan, and Georgia with their peers and research organizations in the EU, Japan, Republic of Korea, Norway and the United States.
ISTC facilitates international science projects and assists the global scientific and business community to source and engage with CIS and Georgian institutes that develop or possess an excellence of scientific know-how.