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Fuel Assembly Under Severe Accident Conditions


Fuel Assembly Tests under Severe Accident Conditions

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

Registration date

Completion date

Senior Project Manager
Tocheny L V

Leading Institute
NPO Lutch, Russia, Moscow reg., Podolsk

Supporting institutes

  • OKB Gidropress, Russia, Moscow reg., Podolsk\nInstitute of Safe Atomic Power Engineering Development, Russia, Moscow


  • Forschungszentrum Karlsruhe GmbH, Germany, Karlsruhe\nIRSN - Institut de Radioprotection et de Sûreté Nucléaire, France, Fontenay aux Roses\nEuropean Commission / Joint Research Center / Institute for Transuranium Elements, Germany, Karlsruhe\nGesellschaft für Anlagen und Reaktorsicherheit mbH, Germany, Köln\nEDF, France, Paris\nCEA / DEN / Laboratoire de Physicochimie et Thermohydraulique Multiphasique, France, Grenoble

Project summary

Safety is one of the main tasks of the design, construction, operation and decommissioning of NPPs. One of the main aspects of safety justification of NPPs with VVER (PWR) reactors is the analysis of the most severe loss of coolant accidents (LOCA), which are accidents which could lead to severe damages, melting of the core, failure of the reactor pressure vessel, hydrogen release, release of radioactive fission products and core melt into the containment.

While analyzing the severe stage of an accident one is to have information on the core conditions and its cooldown possibilities to chose appropriate actions to prevent further accident development and provide the reactor safety.

In this regard, the targeted and systematic work is required to study properties of the core structural materials and their degrading margins during the accident to determine their safe operation limits at which the materials retain their capability to withstand high temperatures without losing main performance characteristics, as well as to study thermo-mechanical behavior of the core structural components under from top and bottom flooding conditions and how the core conditions affect rate, ways and methods of cooling.

Experiments, which are underway in Russia and other countries, do not give sufficient answers to the posed questions both for interaction of core structural materials because of the use of fuel imitators and for thermal-mechanical behavior of the core structural components under severe accident conditions as well as under combined flooding from top and bottom.

In this connection, the objective of the proposed Project is experimental investigation of fuel rod VVER-1000 assemblies (made of standard structural materials used for VVER-1000 - Zr+1%Nb-alloy fuel cladding, uranium dioxide fuel pellets and guiding tubes made of Zr+1%Nb alloy) behavior under simulated conditions of a severe accident including the stage of low rate flooding from top or high rate flooding from top and bottom.

The basis for the Project implementation is the 6-th meeting of the Contact Expert Group on Severe Accidents Management (CEG-SAM) held in Dimitrovgrad (Russia) September 14-17, 2004, as per the revised ISTC Project Proposal № 1134 and recommendations of the 10-th working meeting on the international program QUENCH (10th International QUENCH Workshop, FZK Karlsruhe, October 26-28, 2004).

The Project is implemented jointly by the following organizations pertaining to the Federal Agency of Atomic Energy:

· FSUE EDO “GIDROPRESS” – selection and justification of an experimental scenario for simulated conditions of a severe accident at VVER-1000 reactor;
· IBRAE RAS – development of a computational model of PARAMETER facility including fuel rod assembly and numerical assessment of test modes and parameters;
· FSUE SRI SIA “LUCH”– rig experiments, analysis of experimental results and material studies with involvement of experts from FSUE VNIINM named after A.A.Bochvar, SSC RF-IPPE named after A. I. Leipunsky, ESC of MAE.

The main tasks of the proposed Project are:

· Studies of thermal-mechanical and corrosion behavior of VVER fuel rod assemblies in simulated conditions of a severe accident development stages and determining their damage parameters.
· Studies of thermal-mechanical behavior of structural components of VVER fuel rod assemblies (fuel rod cladding, fuel pellets, guiding tube, spacing grids) under flooding from top/top and bottom of the lead assembly superheated up to 2000оC.
· Studies of the VVER fuel rod assemblies in condition of high rate flooding form top and simultaneous flooding from top and bottom.
· Determining an oxidation degree of the VVER fuel rod assembly structural components.
· Studies of the interaction and structural-phase changes in the VVER fuel rod assembly materials (fuel cladding, fuel pellets).
· Studies of hydrogen release rates under severe accident conditions including stage of bundle flooding.

The set tasks will result in:

· appropriate physical and chemical models development and verification of computer codes designed for severe accidents analyzes at VVER (PWR) reactors;
· obtaining and systematizing valuable information on thermal-mechanical and corrosion behavior of fuel assemblies under the severe accident conditions, which can be used to improve safety of new generation reactor designs and upgrading of existing reactors;
· comparative analysis of the fuel assemblies’ behavior under the simulated conditions of a severe accident, in cases of application of uranium dioxide fuel pellets and in case of zirconium dioxide pellets.

The developed in course of project new methods and related equipment necessary for the analysis of the VVER (PWR) core components’ behavior under a severe accident can be useful for detailed analyzes of NPP with VVER accident sequences, justification of accident mitigations and making safety ensurance decisions in terms of methodology and technology.

The scope of work includes preparation and conduct of two out-of-pile experiments at the rig PARAMETER FSUE SRI SIA “LUCH” during 24 months: tests of two 19-rod fuel assemblies of VVER-1000 under simulated conditions of a severe accident at VVER-1000 reactor, including the stage of flooding (for the first experiment: the low rate flooding from the top; for the second experiment: the high rate combined flooding from the top and bottom).

The modeling nature of experiments is ensured through:

· the use of standard structural materials of VVER-1000 fuel rods and assemblies (fuel rod cladding and pellets, spacing grids, guiding tubes);
· the modeling of residual power density in fuel rods using indirect fuel rod heating by the inner electrical heater;
· the development of a fuel assembly computer model and numerical pre-test analysis;
· the modeling of a temperature-related accident development scenario at VVER-1000 reactor.

The engineering approach and methodology of the Project implementation include the following main stages:

1) Pre-test operations:
· selection and justification of the experiment scenario;
· setting the main tasks of the experiment;
· development of a computational model of a VVER-1000 19-rod fuel assembly and justification of the modeling nature of the experiment;
· pre-test calculations of modes and main parameters of the foreseen fuel assembly tests using RATEG/SVECHA code;
· development of basic technical requirements for the experiment conduct;
· drafting and approval of the experiment program;
· selection of components, assembling, installation and preparation of the fuel assembly and rig process systems.
2) Experiment conduct.

The following is monitored during the experiment:

· cyclogram of the experiment scenarios;
· cladding temperature and internal fuel rod pressure;
· main thermal-hydraulic parameters of the coolant inside the fuel assembly;
· main electric parameters of the fuel assembly;
· change in hydrogen concentration and quantity at the fuel assembly outlet;
· chemical composition of water in the steam generating systems and emergency water injection system;
· chemical composition of the condensate in the high-temperature heat exchanger.
3) Post-test operations:
· video-filming and taking pictures of the dummy fuel assembly after tests;
· processing of the recorders’ data;
· material studies of specimens:
  • metallographic analysis;
  • X-ray structure and X-ray phase studies;
  • studies with electron microscope;
  • micro-X-ray spectral analysis;
  • drafting of reports.

The proposed Project meets the ISTC goals and objectives and will allow:
· to convert for peaceful activities a part of the Russian specialists who formerly were involved in weapons programs development;
· to use the obtained results for justification of safety of VVER (PWR) reactors as well as for development of methods and equipment of the control and protection system which are capable of performing their functions under accident conditions;
· to improve reliability and safety of nuclear power reactors both existing and under design and development.

The Project is implemented jointly with foreign Collaborators in frames of:

· information exchange and joint review of science and technology reports;
· joint workshops, meetings and consultations;
· verification of results using independent methods and/or equipment;
· shared use of test materials and specimens;
· joint verification of results obtained through the Project;
· consultations on the intellectual property rights in case of a joint invention.


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.

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