Gateway for:

Member Countries

Verification of Reactor Data Bases (R)

#0116


Development of methodical and calculation technology verification of nuclear data bases used in the calculation of neutron-physical characteristics and in analysis of nuclear safety of reactor facilities and nuclear conversion technological processes.

Tech Area / Field

  • FIR-MOD/Modelling/Fission Reactors

Status
8 Project completed

Registration date
04.11.1993

Completion date
15.03.2001

Senior Project Manager
Tocheny L V

Leading Institute
Federal State Unitary Enterprise Research and Development Institute of Power Engineering named after N.A.Dollezhal, Russia, Moscow

Supporting institutes

  • Kurchatov Research Center, Russia, Moscow\nFEI (IPPE), Russia, Kaluga reg., Obninsk\nVNIITF, Russia, Chelyabinsk reg., Snezhinsk

Collaborators

  • Belgonucléaire, Belgium, Brussels\nBundesamt für Strahlenschutz (BfS - S), Germany, Berlin\nGRS mbH Foeschungsgelande Hauptgelande, Germany, Garching\nSiemens AG / Power Generation (KWU), Germany, Erlangen\nFRAMATOME, France, Paris La Défense\nCEA / DCC / CEN Valrho, France, Marcoule\nECN Petten, The Netherlands, Petten

Project summary

The purposes of the proposed project are:

- verification of nuclear data libraries and codes used in calculations of reactors with Plutonium-including inventory, in particular, the reactors of RBMK type, primarily, in calculations validating their safety;

- development of methods of data condensation and the analysis of errors possible on transition from evaluated nuclear data files to group data libraries;

- planning of benchmark-type experiments for assemblies containing mixed U-Pu fuel (for example, of equilibrium refueling regime vector of RBMK-type reactors) with the use of SRITP potential as well as that of all organizations involved in the present project;

- integration of the results in the field of experimental and calculation verification of reactor codes and data libraries to the international scientific community.

The project implies:

- development of the system of numerical tests on the basis of pointwise or multigroup calculation to perform verification of group data libraries: the tests would be oriented on U-C systems with U, U-Pu inventories, in particular, on ROMB, UG assemblies;

- performance of calculation investigations with the implementation of precision codes and experimental data;

- verification of nuclear data used and the analysis of its uncertainties, the analysis of errors for major approximations used in reactor calculations;

- development of the codes allowing for the acknowledged methods of correction of nuclear data, in particular, with the use of generalized least-square method;

- planning, on the basis of the results obtained, of benchmark-type experiments for ROME critical facility with assemblies containing mixed U-Pu fuel, in particular, with assemblies of RBMK (MKR) type.

The studies would incorporate both well-acknowledged methods and codes for transport equation solution (Monte-Carlo method, method of discrete ordinates), for the analysis of statistic and systematic errors, sensitivity coefficients, etc., and the original ones (the solution of transport equation in irregular 3D-geometry by the method of characteristics), that would provide high scientific value and validity of the results obtained.

THE EXPECTED RESULTS OF THE PROJECT

1. The library of numerical tests presented in the format allowing for verification of both existing and new codes, methods, algorithms.

2. The files containing the results of benchmark-type experiments, including the ones obtained on ROMB critical facility, in the format allowing for verification of the implemented nuclear data, codes and algorithms.

3. The compiled library of files of evaluated nuclear data in ENDF/B format.

4. The verified group library of problem-oriented constants and corresponding covariation matrices for thermal reactors of channel type. Numerical tests for library verification.

5. Alternative systems of group constants, obtained on the basis of implementation o of different models of the problem of space-energy neutron distribution simulation.

6. The algorithm and code modules for calculation of the errors of reactor functions allowing for sensitivity distributions obtained in linear perturbation theory approximation.

7. The algorithm and the code for the correction of group constants on the results of experiments for thermal reactors of channel type.

8. The program of benchmark experiments on the base of ROMB critical facility with fuel assemblies of RBMK type with U-Pu fuel.


Back