Passive Containment Filtered Vent (PCFV) was installed in Nuclear Power Plant (NPP) Krsko in 2013 as part of the safety upgrade program. It is intended for severe accident consequences prevention and mitigation by ensuring the containment integrity. In this paper, dose rates around the exhaust line of the PCFV system resulting from radioactivity release in case of a severe accident were determined in a four step methodology. The assumed severe accident scenario is a beyond design basis station blackout in NPP Krsko, which was simulated using the MELCOR code. Its results were input for the RADTRAD radiological calculations to obtain the activities released in the containment. These activities were then transformed into the gamma source intensity and spectrum using the ORIGEN-S libraries. This form of the source term is required for Monte Carlo calculations which were performed using the MCNP6.2. Two Monte Carlo calculations were performed. One for which the radiation source was modeled to emanate from the containment atmosphere and the other from the PCFV duct fluid. The main reason for the calculation was to assess limiting dose rates around PCFV duct (radiation monitor location) during actuation after severe accident. That is why the model is simple and conservative. The other task was to demonstrate that this location is not suitable for longer personnel presence in case of equipment failure during the PCFV actuation. Due to conservative assumptions, predicted dose rates are the highest expected at that location for any severe accident scenario.
NPP Krsko is introducing Emergency Control Room (ECR) as part of safety upgrades. According to 10CFR50 Appendix A, GDC 19, both main control room and emergency control room should have adequate radiation protection to permit operators to shutdown the plant and keep it in safe shutdown conditions without receiving more than 50 mSv effective whole body dose, within 30 days from accident initiation. One of the important prerequisites to achieve that is proper operation of control room HVAC. In this work we are focused to calculation of gamma doses from radioactive materials accumulated in HEPA and charcoal filters during 30 days of HVAC operation. The dose at selected points around the filter was calculated using Microshield 10.0 point kernel code. The radioactive gamma source is calculated using RADTRAD 3.03 for plant's severe accident SGTR sequence calculated with MAAP 4.0.7 code. Calculated dose rates at peak filter activity are compared against results obtained with SCALE 6.2 MAVRIC shielding sequence (Monaco Monte Carlo functional module and CADIS methodology). The reasonable agreement between point kernel and hybrid Monte Carlo results was obtained.
This work presents the results of radiation shielding calculations using modified point kernel code QAD-CGGP. The modification includes a new approach to neutron buildup factor estimations based on machine learning technique called Support vector regression (SVR). SVR neutron buildup factor models for common shielding materials are developed and built into the QAD-CGGP. The development of the models consisted of acquiring the data to be used for learning the model, optimizing the SVR parameters, and application of active learning methods for improving the learning process. The modified code is tested, and the results are compared with the MCNP6 results.
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