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• Activation analysis was performed according to impurities in the reactor structure. • OPR-1000 was modeled with MCNP, and activation analysis through ORIGEN-S was performed using the calculated neutron flux. • It was calculated that Co increases the total specific activity the most in metal, and Li increases the most in concrete. Various kinds of radioactive wastes are expected to be generated resulting from the decommissioning of nuclear power plants. For the safe management of decommissioning wastes, the preliminary classification of radioactive wastes such as very low level, low level, intermediate level, and regulatory clearance has to be performed. In this regard, we intend to conduct an activation analysis and radioactive waste level estimation using a computational simulation code for the structure materials of an OPR-1000, a standard nuclear power plant in South Korea. In addition, the structure material that constitutes the reactor vessel, reactor vessel internals, and bio-shield concrete of the nuclear power plant contains small amounts of unintended impurities. The results of the activation analysis may vary depending on the presence of those impurities. In this study, a detail modeling of nuclear reactor and the activation analysis are carried out combined the MCNP code with the ORIGEN-S code. Based on this methodology, the sensitivity test is done for impurities of structure materials. From the result of the specific activity of activated structure material, the cobalt impurity is found to be the most important contributor in the level determination of radioactive wastes. However, the bio-shield concrete has the highest specific activity when Li impurity is contained. From this study, some impurities affect radioactive waste level, thus it is recommended to consider structure material impurities for decommissioning the nuclear power plants. [ABSTRACT FROM AUTHOR]

• Risk-informed approach for safety improvement for research reactor. • Improvement of the safety level of domestic research reactor. • Optimization of reactor operation with regard to safety by conducting PSA. • Derivation of insights such as design and procedure improvements from sensitivity analysis. This paper describes the effort to improve the safety level of a domestic research reactor and optimize its operation with regard to safety by conducting a probabilistic safety assessment (PSA) under full-power operating conditions. The PSA was undertaken to assess the level of safety for an operating research reactor in Korea, evaluate whether it is probabilistically safe to operate and reliable to use, and obtain insights about the procedure and design improvements for achieving increased safe operation. The scope of the PSA reported here is a Level 1 PSA, which addresses the risks associated with core damage. The technical objectives of this study were to identify the accident sequences causing core damage and to derive insights, such as design and procedure improvements, from the dominant accident sequences by conducting sensitivity analyses. The event tree (ET)/fault tree (FT) linking approach was used as the basic method for accident sequence quantification for the PSA. The accident sequence models followed the small ET/large FT approach. The total core damage frequency (CDF) was estimated as the sum of the frequencies of the individual ET sequences causing core damage. A truncation value of 1.0E−15/yr was applied for all the FTs or accident sequences. The PSA results indicated a point estimate of 4.55E−06/yr for the overall CDF, attributable to the internal initiating events for the Korea research reactor. Based on the dominant accident sequences from the PSA, eight types of sensitivity analyses were performed, and relevant insights for improved safe operation were derived. When the insights from these sensitivity analyses were applied to the reactor design and the operating procedure, the risk was reduced by approximately ten times compared to that before, and its safety was significantly improved. The PSA methodology is very effective in improving reactor safety in the full-power operating phase, and, in particular, it is a highly suitable approach to determine the deficiencies of a reactor at full power and improve the reactor safety by overcoming them. [ABSTRACT FROM AUTHOR]