VOID-REACTIVITY EFFECTS - RBMK

TACIS Design Safety projects were service or consultancy type projects that were able to address research, design and engineering issues. They were usually implemented in such a way that an EU research, design or engineering organisation worked in close collaboration with Beneficiary country counterpart organisations on a specific technical problem.
The design safety projects therefore achieved a dual purpose of providing the necessary technical resources to solve the issue of concern whilst at the same time transferring know-how and developing capabilities within the Beneficiary country organisations wherever a lack had been identified. This dual purpose also helped to provide the sustainability of the project benefits that was a major objective of the TACIS programme.

The RBMK reactor is a boiling light water, graphite moderated reactor of very large power (3000MWt) designed and operated in the former Soviet Union. After the Chernobyl accident, the RMBK safety was under the scrutiny of the internal as well as the international nuclear society.
One of the major safety issues of the RBMK reactor is a positive void coefficient, which makes the reactor unstable especially in low power operational states. This was also one of the major contributors to the Chernobyl accident of April 1986.
In 1991, when the TACIS nuclear safety programme formally started, the European Union had very little information about the design and operational characteristics of the RBMK reactors. In order to learn more about RBMK specific design features, and in particular, perform an independent analysis of the RBMK thermal hydraulic and neutron kinetic behavior, the European Commission launched several project in this direction.
A PSB RBMK integral test facility has been constructed in Electrogorsk Research and Engineering Center (EREC, Moscow Region) as part of the effort to experimentally validate T/H codes. The PSB RBMK test facility is a large-scale integral testbed which models one loop of the multiple forced circulation circuit (MFCC) of the RBMK-1000 reactor.
A number of scaled experiments were performed in order to simulate anticipated accidents and transients in RBMK at the large-scale integral test facility PSB RBMK. Obtained experimental data are used for validation of basic thermal hydraulic codes currently in use in the Russian Federation for RBMK safety analysis. Results of performed activities will give the basis for certification of the codes in the Russian Federation Regulatory Authorities – Rostechnadzor (former GOSATOMNADZOR).
The project "RBMK void reactivity effect calculation" belongs to a group of projects aimed at enhancing nuclear safety of RBMK reactors through validation and precertification of basic thermal hydraulic/neutronic codes used in the Russian Federation.
The objective of RBMK void reactivity effect calculation project was to provide for reliable quantification of void –induced reactivity effects in current RBMK reactor core designs and thus to assist the regulatory authority in the Commonwealth of Independent Sates (CIS) in evaluating related safety features of RBMK reactors.

The European Commission (EC) Consultant chosen for this project was University Bremen, that worked closely with the Russian counterpart - the Kurchatov Institute. The RBMK void reactivity effect calculation provided for a quantification of void induced reactivity effects in current RBMK reactor core designs; obtained calculation results were used in evaluating safety features of RBMK reactors.
In order to compare the results obtained with Russian and EU methods and thus increase the calculation reliability, advanced calculation tools were used to investigate the RBMK core neutronics behavior. Numerous calculations performed in the course of the project were performed independently from the established Russian computational tools.
In the past, Russian institutions mainly used the British WIMS-D code and the Russian STEPAN code to calculate the core neutronics behavior of RBMK reactors. The EC Consultant decided to use the advanced three-dimensional Monte Carlo code MCNP in order to establish a basis for an independent analysis of the RBMK void reactivity effects. The code version 4.2 Monte Carlo N-Particle code (MCNP) distributed by OECD/NEA data bank, and MCNP-4A received directly from Los Alamos, were implemented and successfully tested by the team in Bremen and Moscow. In addition, the MCNP compatible LANL software package SABRINA for plotting of the geometrical material configuration and particle tracking was implemented in Bremen.
Large efforts were made for the experimental verification of the MCNP code and the utilized neutron cross-section libraries. Test experiment performed at the RBMK critical facility in the Russian research Center Kurchatov institute (RRC) and whole reactor experiments performed at Smolensk NPP, Unit 3 during the start-up were used for the verification of calculation results. Although the MCNP results agree rather well with the experiments, there are still uncertainties concerning modeling assumptions of the reactor which were not yet quantified satisfactorily.
Calculation results using different computational codes and their comparison showed large differences in the void effect simulations. The influence of different neutron cross section libraries and computing codes on the results was significant. Because of these large differences, it was difficult to quantify the influence of non-fuel channels on the void reactivity effect. The uncertainty contributions originating from the cross-section libraries, from the numerical approaches and from differences in the geometrical modeling need to be quantified by future investigations.
The EC Consultant made several recommendations, e.g. to define RBMK core physics related benchmarks based on RRC Critical Test Facility experiments for establishing a more standardized verification in the future.