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Because the Taiwan domestic boiling water reactor nuclear power plant is under shutdown and going to be decommissioned, the development of radioactive waste container used for containing the radioactive waste from nuclear facility decommissioning and dismantling is a critical issue in recent years. The main purpose of this study is to design a metal container which can provide the highest storage efficiency for the low level radioactive waste, and ensure the structure meeting the safety requirements of related regulations. This paper first introduces the related regulations for the container in Taiwan. To determine the optimal storage efficiency of the container, the shielding analyses were performed by varying dimensions of 3 cubic meter containers to select the best one, and conduct the structural analyses. Then, a series of structural analyses for operational conditions such as static standstill analysis, lifting test and stacking test with full load were performed, as well as the drop impact analysis, to ensure the container structure meeting the safety requirements of regulations. Present work can provide design information of the development of radioactive waste container for decommissioned nuclear facility in Taiwan.

The pressure-temperature limits (P-T limits) of embrittled reactor pressure vessels (RPVs) should be determined for normal heat-up and cool-down operations based on fracture mechanics to ensure the structural integrity. In general, the P-T limits are mainly determined by the fracture toughness of beltline region material which suffers the highest level of neutron embrittlement. However, some other components such as nozzles which suffer insignificant neutron embrittlement may restrict the P-T limits as a result of higher stress concentration from their structural discontinuities. Therefore, beltline region material with the highest reference temperature as well as nozzles that have structural discontinuities need to be considered when developing the P-T limits of RPV. In this paper, the P-T limit calculations for a Taiwan domestic pressurized water reactor (PWR) vessel were performed considering both beltline shell and extended beltline nozzles according to the nonmandatory Appendix G to ASME Code Section XI. The three-dimensional finite element analyses were conducted to obtain the pressure and thermal stress distributions of extended beltline nozzles for P-T limits calculation. It is found that the beltline region still dominates the cool-down P-T limit, but the extended beltline region will partially control the heat-up limit. Further, the reference temperature ranges of extended beltline nozzles that may influence the P-T limits of RPV were also investigated. Present study is helpful for safe operation and also provides a reference for regulation of PWR plants in Taiwan.

To ensure the structural integrity of the embrittled reactor pressure vessels (RPVs) during startup or shutdown operation, the pressure-temperature (P-T) limits are mainly determined by the fracture toughness of beltline region material with the highest level of neutron embrittlement. However, other vessel parts such as nozzles with structural discontinuities may affect the limits due to the higher stress concentration, even though the neutron embrittlement is insignificant. Therefore, not only beltline material with the highest reference temperature, but also other components with structural discontinuities have to be considered for the development of P-T limits of RPV. In the paper, the pressure-temperature operational limits of a Taiwan domestic pressurized water reactor (PWR) pressure vessel considering beltline and extended beltline regions are established per the procedure of ASME Code Section XI-Appendix G. The three-dimensional finite element models of PWR inlet and outlet nozzles above the beltline region are also built to analyze the pressure and thermal stress distributions for P-T limits calculation. The analysis results indicate that the cool-down P-T limit of the domestic PWR vessel is still dominated by the beltline region, but the heat-up limit is partially controlled by the extended beltline region. On the other hand, the relations of reference temperature between nozzles and beltline region on the P-T limits are also discussed. Present work could be a reference for the regulatory body and is also helpful for safe operation of PWRs in Taiwan.

In recent years, the probabilistic fracture mechanics (PFM) approach has been widely applied to estimate the fracture risk of nuclear power plant piping systems. In the paper, the probabilistic fracture mechanics code, PRO-LOCA, developed by the Probabilistic Analysis as a Regulatory Tool for Risk Informed Decision Guidance (PARTRIDGE) project, is employed to practically evaluate the fracture probability of the recirculation piping system welds in a Taiwan domestic boiling water reactor (BWR) nuclear power plant. To begin with, the models based on the real situation of the recirculation piping welds are built. Then, the probabilities of through-wall cracking, leak with different rates, and rupture on the welds considering both in-service inspection and leak detection are analyzed. Meanwhile, the effects of probability of detection curves of ISI on the piping are simulated. Further, the efficiencies of performing the induction heating stress improvement and weld overlay are also studied and discussed. The present work could provide a reference of operation, inspection and maintenance for BWR plants in Taiwan.

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