The nuclear criticality analyses of the spent fuel pool under the postulated conditions of loss of spent fuel pool water and loss of neutron absorbers in the spent fuel racks, for Taipower’s Chinshan Nuclear Power Plant, were performed primarily using the Monte Carlo program MCNP5 in association with the deterministic neutron transport code CASMO-4. The results of these analyses can be used to help understand the impact of these beyond-design-basis accidents to the nuclear criticality, as well as facilitate nuclear utilities and regulatory bodies to develop the safety measures and regulations needed to prevent the criticality accidents.

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