A formulation has been established to estimate the error propagation in Monte-Carlo burnup calculations. The uncertainties in cross sections and the statistical errors in Monte-Carlo calculations are considered as error sources, and the error propagation of number densities of individual nuclides over a burnup period is formulated. The present formulation is applied to the burnup calculation of a simplified fast reactor core. The components of the errors in number densities due to the statistical error are up to 0.92% even when the history number is small as 10 4 • On the other hand, the components due to the cross section error are about 2~5% for the number densities of 235 U, 239 Pu, 240 Pu, 241 Pu and 242 Pu, and about 7.3% for the fission. product. Thus the contribution is mainly due to the cross section errors. The error propagation of the number densities due to the statistical errors at individual burnup steps is investigated by dividing the burnup period into two steps. The error propagation is not serious for the problem treated here because the component due to the statistical error is much smaller than that due to the cross section error.

A formulation has been established to estimate the error propagation in Monte-Carlo burnup calculations. The uncertainties in cross sections and the statistical errors in Monte-Carlo calculations are considered as error sources, and the error propagation of number densities of individual nuclides over a burnup period is formulated. The present formulation is applied to the burnup calculation of a simplified fast reactor core. The components of the errors in number densities due to the statistical error are up to 0.92% even when the history number is small as 10 4 • On the other hand, the components due to the cross section error are about 2~5% for the number densities of 235 U, 239 Pu, 240 Pu, 241 Pu and 242 Pu, and about 7.3% for the fission. product. Thus the contribution is mainly due to the cross section errors. The error propagation of the number densities due to the statistical errors at individual burnup steps is investigated by dividing the burnup period into two steps. The error propagation is not serious for the problem treated here because the component due to the statistical error is much smaller than that due to the cross section error.

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In the Fukushima Dai-ichi nuclear power station, Loss of Ultimate Heat Sink (LUHS) was caused by the great east japan earthquake and the subsequent tsunami [1]. It resulted in severe accident in three units. In that time, fuel damage in Spent Fuel Pool (SFP) were prevented by the various countermeasures such as makeup by pump truck and recovery of injection systems /cooling water system. In the past, Probabilistic Safety Assessment (PSA) has been developed with a focus on the reactor. After the accident, it has been acknowledged that SFP PSA is important to enhance the plant safety. In this study, probabilistic assessment is performed to suggest countermeasures for LUHS to SFP.

Advanced Boiling Water Reactor (ABWR) has multiple safety features to prevent core damage and mitigate the accident progression. Failures of all injection systems to the reactor (ECCS, alternative injection systems, mobile injection systems) could cause a core melt accident and the Reactor Pressure Vessel (RPV) failure. If large amount of debris particles are deposited on the structures in the containment such as access tunnel, hatches, and Vacuum Breakers (V/Bs), it may result in containment failure or Suppression Pool (S/P) bypass due to the direct heating from the corium. This is defined as Direct Debris Interaction (DDI) in this paper. The phenomenon is introduced in the containment event trees of internal events at power Level 2 Probabilistic Safety Assessment (PSA) for an ABWR and sensitivity analyses and source term analyses are performed in this paper. As a result, the contribution of DDI to total Containment Failure Frequency (CFF) is not significant for the ABWR even if the conservative DDI probability is assumed.