Application of the Graded Approach for the Safety of NIRR-1 HEU to LEU Core Conversion

Abstract

Application of the graded approach for the Safety of NIRR-1 HEU to LEU Core Conversion. Graded Approach is a method in which the stringency of the design measures and analysis applied are commensurate with the level of risk posed by the reactor facility. The Code of Conduct on the Safety of Research Reactors and the IAEA’s Safety Standards recommend that a graded approach with regards to the risk potential be applied throughout all stages of the lifetime of a research reactor, including site evaluation, design, construction, commissioning, operation, conversion, extended shutdown and decommissioning. Factors to be considered when applying graded approach are reactor power, reactor safety characteristics, amount and enrichment of fissile and fissionable material, fuel design, type and mass of moderator, reflector and coolant, utilization of the reactor, safety design features, source term and sitting and proximity to populated areas. This presentation contains details on safety and design analysis performed for NIRR-1 core conversion from 92% Highly Enriched Uranium (HEU) fuel to 12.5% Low Enriched Uranium (LEU) fuel. The safety justifications for (NIRR-1) LEU core which are to maintain small compact core structure, excellent inherent safety features, fuel integrity is maintained under all operating conditions, Reactivity coefficient meets required limits and are comparable to the existing HEU core, increase ratio of neutron flux in the irradiation site to the core thermal power and the LEU has a better shut down margin, because it has a central control rod worth of 7.7mk, which ensures a shut down margin of 3.72mk, rather than the HEU control rod worth of 7mk and shut down margin of 3mk. It also talk about the codes and standards which are used to calculate neutronics parameters, thermal-hydraulic characteristics and radiological consequences data in the LEU core for NIRR-1: MCNP – is a soft ware packaged for the simulation of nuclear processes, WIMS-ANL – is used for the design and predictive assessment of reactor physics performance e.g. fuel transport flask, PLTEMP\ANL – is used for the calculation of Reactor Thermo-Hydraulic along fuel plates. ORIGEN – is a computer code used for calculating the decay and processing of radioactive materials.

Seminar presented by Kayode James Adedoyin