G. B. Davydova

Induced Radioactivity and Waste Classification of Reactor Zone Components of the Chernobyl Nuclear Power Plant Unit 1 After Final Shutdown

Abstract The dismantlement of the reactor core materials and surrounding structural components is a major technical concern for those planning closure and decontamination and decommissioning of the Chernobyl Nuclear Power Plant (NPP). Specific issues include when and how dismantlement should be accomplished and what the radwaste classification of the dismantled system would be at the time it is disassembled. Whereas radiation levels and residual radiological characteristics of the majority of the plant systems are directly measured using standard radiation survey and radiochemical analysis techniques, actual measurements of reactor zone materials are not practical due to high radiation levels and inaccessibility. For these reasons, neutron transport analysis was used to estimate induced radioactivity and radiation levels in the Chernobyl NPP Unit 1 reactor core materials and structures. Analysis results suggest that the optimum period of safe storage is 90 to 100 yr for the Unit 1 reactor. For all of the reactor components except the fuel channel pipes (or pressure tubes), this will provide sufficient decay time to allow unlimited worker access during dismantlement, minimize the need for expensive remote dismantlement, and allow for the dismantled reactor components to be classified as low- or medium-level radioactive waste. The fuel channel pipes will remain classified as high-activity waste requiring remote dismantlement for hundreds of years due to the high concentration of induced 63Ni in the Zircaloy pipes.

Computational Estimates of the Radiation Characteristics of Irradiated Graphite after Final Shutdown of a Nuclear Power Plant with an RBMK Reactor

A method of calculating the radiation characteristics of irradiated graphite masonry of an RBMK reactor is described. The MCNP computer code is used to determine the spatial distribution of the neutron flux density in the interior volume of the graphite, the CHAIN code is used to determine the isotopic composition and the radition characteristics of the irradiated graphite on the basis of the MCNP fluxes.The results of the calculation of the radiation characteristics of graphite from the reactors in the Nos. 2 and 3 units of the Leningrad nuclear power plant and the No. 1 unit of the Chernobyl nuclear power plant are presented and the contribution made by the accident to the flow of fuel mass into the masonry is estimated.

Validation of MCNP for RBMK criticality calculations

The Los Alamos National Laboratory Monte Carlo MCNP code is applied to critical experiments performed at the RBMK critical facility of the Russian Research Center Kurchatov Institute, Moscow. The validation investigations are completed by whole-core criticality calculations of experiments at the Smolensk Unit 3 nuclear power plant as part of the start-up procedure. The geometric model exploits the powerful capabilities of MCNP by precise representation of the fuel and different types of nonfuel channels, which add up to a detailed model of the critical facility and the RBMK core. Continuous-energy cross-section tables are taken from the ENDF/B-IV and ENDF/B-VI libraries. As the most important uncertainty inherent to the experimental setup, the concentration of impurity isotopes in the graphite moderator is identified. Within the resulting error limits, the k{sub eff} and the void effect are well reproduced with both cross-section libraries.

The Monte Carlo codes MCNP and MCU for RBMK criticality calculations

Abstract The modern Monte Carlo codes MCNP and MCU have been established as important tools to determine the neutronic behavior of reactor cores. For a comparison of their capabilities, detailed representations of seven critical experiments performed at the Russian Research Center ‘Kurchatov Institute’ were developed, identical in geometry and material composition for each code. Despite the different philosophy of code development, especially in the process of generating cross-section tables, the calculated isotopic reaction rates and the flux distributions are in excellent agreement. Effective multiplication factor and void reactivity effect agree well with experiment. Additional uniform lattice calculations confirm the equivalent potential of MCNP and MCU, but exhibit significant differences to results achieved with transport codes like WIMS-D4.