Validation Results of Scale and MCNP Calculation Codes on Data of Benchmark Critical Experiments at SF-9 Installation

V. I. Borysenko1,2, V. V. Goranchuk1, Yu. F. Piontkovskyi2,1,I. O. Titimets2

1Institute for Safety Problems of Nuclear Power Plants, NAS of Ukraine, Lysogirska str., 12, Kyiv, 03028, Ukraine
2Nuclear Physics Department, Taras Shevchenko National University, Prospect Glushkova, 4, building 1, Kyiv, 03022, Ukraine



The article presents the results of calculations of benchmark critical experiments in the codes SCALE and MCNP. Experiments were performed on SF-9 installation, which was operated by Russian Research Centre “Kurchatov Institute” in 1966-1987. The SF-9 is designed to conduct criticality studies, depending on the number of fuel rods (type VVER-1000) and the level of coolant (water) in the installation tank. Such research is obligatory when substantiating the possibility of using the calculation codes selected for the analysis of nuclear safety of technological operations on the transfer, transportation and storage of fresh nuclear fuel (the storage site of nuclear fuel on the power unit), as well as spent nuclear fuel (transfer casks, repositories of spent nuclear fuel) reactors of various types. For example, when substantiating the nuclear safety of spent fuel storage systems, it must be confirmed that the maximum value of the effective neutron multiplication factor kef is lower than the established normative limit of 0.95 for normal operation conditions, in case of violations of normal operation and in case of design accident.

A series of 12 experiments on the SF-9 installation investigating the neutron-physics parameters of VVER-type uranium-water lattices was carried out in 1973. Near-cylindrical cores with a regular hexagonal lattice at a pitch of 12.7 mm and 3.5 wt.% 235U fuel enrichment were build. Criticality was reached by varying the moderator height. The number of fuel rods varied from 691 to 1897 and the critical moderator height varied from 30.68 cm to 117.89 cm above the bottom surface of the fissile column.

The article presents the results of calculations of krf for these 12 critical experiments, which allow us to draw conclusions regarding the error of the value determination of kef , of SCALE and MCNP codes, and to select a calculation code for the analysis of nuclear safety of nuclear fuel storage systems, including spent nuclear fuel, on the base of validation results.

Keywords: nuclear safety, critical experiments, fuel rod, spent nuclear fuel, effective neutron multiplication factor.


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