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Fuel Assemblies under Severe Accident Conditions


Study of Fuel Assemblies with Boron Carbide Absorber Rods under Severe Accident Conditions in the PARAMETER-SF Tests 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

3 Approved without Funding

Registration date

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

Supporting institutes

  • Nuclear Safety Institute, Russia, Moscow\nOKB Gidropress, Russia, Moscow reg., Podolsk


  • Forschungszentrum Karlsruhe GmbH, Germany, Karlsruhe\nEuropean Commission / Joint Research Center / Institute for Transuranium Elements, Germany, Karlsruhe\nGesellschaft für Anlagen und Reaktorsicherheit mbH, Germany, Köln\nHungarian Academy of Sciences / KFKI Atomic Energy Research Institute, Hungary, Budapest

Project summary

Nowadays a serious risk appears all over the world concerning availability of power resources, safe production of energy, changes in climate and quality of the air, therefore there is an increase in the role of nuclear power it could play in the future power systems and energy supply. However the long-term prospects of nuclear power utilization should be considered in general context of assurance of safety, economic competitiveness and risks of proliferation.

Safety assurance is one of the main tasks in designing, construction, operation and decommissioning of NPP. The key role in NPP safety problem is devoted to reliability of nuclear reactor control features. The considerable reliability improvement of control features can be gained by usage of radiation-resistant and high efficient absorber materials.

For successful and efficient usage of such materials and structures the calculation and experimental verification of their availability is necessary not only for the operation conditions but also for the conditions of postulated design basis accidents (DBA) and initial stages of beyond design basis accidents (BDBA), and the studies are required for the effect of structural materials of absorber rods (AR) on the core behavior at the initial stage of the severe accident with partial or complete melting of the core components.

In the analysis of accident a special attention should be paid to the interference of fuel rods and absorber rods within the fuel assembly (FA). Deformation of fuel rod claddings could lead to deformation and blockage of the guiding tube and the AR cladding itself. Melting and relocation of AR components can cause the additional core overheating both due to reaching the state of local criticality and power spike, and due to blockage with melt of flow area of some sells of spacing grids and worsening of the fuel rods cooling conditions.

For detailed study of the core structural components interaction processes the complex bench studies are required for the behaviour of absorber rods in a set of FA, including fuel rods, spacer grids and guiding channels under the conditions simulating severe accident at VVER reactor plant.

The PARAMETER test facility complex in SRI SIA “LUCH” is the most appropriate for such kind of studies that allows to conduct the required scope of experimental and material studies with the calculation and methodical support of studies made by specialists of IBRAE RAS, OKB “GIDROPRESS”, VNIINM, RRC “Kurchatov Institute”, SRI SIA “LUCH” and JSC “MZP”.

The studies of AR behaviour under accidents shall include the following:

  • study of AR thermomechanical behaviour under tests of the model FAs using different scenarios of DBA and BDBA;
  • post-test material studies of FA and AR to determine the following:
    • the degree of cladding oxidation over AR length depending on temperature;
    • the degree of metal melting of the cladding, guiding tube and AR materials;
    • the composition of the solidified mixtures after flowing down of melt and formation of solid layer of melted structural materials (corium) that can block the coolant cross section;
    • the character of cooling of the model assembly with AR under top flooding.
On the basis of the analysis of indications of measured temperature and using the post-test material studies it is possible to identify the processes and temperature regimes that caused damage and melting of structural components.

The objective of the proposed Project is the study of behaviour of two 18-rods model FAs of VVER-1000 with the guiding tube and the central AR, completed with standard reactor materials (structural materials, fuel pellets and boron carbide absorber rods) under the conditions of the initial stage of severe accident with top flooding.

The studies planned in new Project present a continuation of PARAMETER-SF test series at the PARAMETER test facility started under ISTC Projects # 3194 and # 3690.

The Project will be jointly implemented by the leading organizations of the State Atomic Energy Corporation “ROSATOM” and Russian Academy of Sciences:

  • FSUE SRI SIA “LUCH” – performing the experiments, post-test calculations and material study;
  • IBRAE RAS – making scenarios of experiments, pre-test and post-test calculations;
  • OKB “GIDROPRESS” – making scenarios of experiments, analysis of the experiment model, pre-test and post-test calculations.

The following results are expected during implementation of the Project:
  • obtaining and systematization of information on behaviour of the model FA with boron carbide AR under the conditions of severe accident with top flooding;
  • study of the degree of the cladding oxidation over AR length depending on temperature and of the degree of metal melting of the cladding, guiding tube and AR materials;
  • determination of the composition of solidified mixtures after flowing down of melt;
  • extension of database for verification of severe accident codes (SOCRAT/B1, ATHLET, ICARE-CATHARE, etc.).

The obtained results can be used for safety justification of VVER and PWR type reactors.

The scope of the work for 24 months includes preparing and conducting two experiments at PARAMETER test facility on studying two model FAs of VVER-1000 with 18 heated fuel rods and the central boron carbide AR:

  1. heating-up of the model assembly in steam-argon flow to maximum temperature of fuel rods before the beginning of flooding ~1250oC (PARAMETER-SF5 experiment);
  2. heating-up of the model assembly in steam-argon flow to maximum temperature of fuel rods before the beginning of flooding ~1450oC (PARAMETER-SF6).

In both experiments the top flooding water flow rate is 40g/s.

Following the experiments SF5 and SF6 the post-test material studies of model assemblies will be carried out.

The proposed methodological approach to implementation of the Project is provided by:

  • making of a realistic scenario of experiments based on safety justification of VVER using computer codes TECH-M and KORSAR;
  • calculational modeling of experiments using the certified computer code package SOCRAT/B1;
  • completing the fuel assembly simulator with the standard structural materials of fuel rods and FAs of VVER-1000 (fuel rod claddings of alloy Zr+1%Nb, fuel pellets of uranium dioxide, spacing grids and shell of alloy Zr+1%Nb, boron carbide AR).

The proposed Project meets ISTC objectives and tasks since the defense industry scientists and specialists will be involved in its implementation, and the final results will contribute to improvement of reliability and safety of nuclear power reactors both under operation, and those being designed and constructed.


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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