A comprehensive uncertainty analysis in the event of a severe accident in a two-loop pressurized water reactor is conducted using an uncertainty package integrated in the ASYST code. The plant model is based on the nuclear power plant (NPP) Krško, a Westinghouse-type power plant. The station blackout scenario with a small break loss of coolant accident is analyzed, and all processes of the in-vessel phase are covered. A best estimate plus uncertainty (BEPU) methodology with probabilistic propagation of input uncertainty is used. The uncertain parameters are selected based on their impact on the safety criteria, the operation of the NPP safety systems and to describe uncertainties in the initial and boundary conditions. The number of required calculations is determined by the Wilks formula from the desired percentile and confidence level, and the values of the uncertain parameters are randomly sampled according to appropriate distribution functions. Results showing the thermal hydraulic behaviour of the primary system and the propagation of core degradation are presented for 124 successful calculations, which is the minimum number of required calculations to estimate a 95/95 tolerance limit at the 3rd order of the Wilks formula application. A statistical analysis of the dispersion of results is performed afterwards. Calculation of the influence measures shows a strong correlation between the decay heat and the representative output quantities, which are the mass of hydrogen produced during the oxidation and the height of molten material in the lower head. As the decay heat increases, an increase in the production of hydrogen and the amount of molten material is clearly observed. The correlation is weak for other input uncertain parameters representing physical phenomena, initial and boundary conditions. The influence of the order of the Wilks formula is investigated and it is found that increasing the number of calculations does not significantly change the bounding values or the distribution of results for this particular application.

Electromagnetic compatibility (EMC) in nuclear power plants today mostly relies on the qualification tests of the new equipment and adhering to some good installation practices. Diversity of the electromagnetic environment and different susceptibility of the plant equipment calls for a systematic classification of the EMC zones in a nuclear power plant. The paper proposes a methodology that uses a combination of the qualification tests, in situ and bench immunity tests, site survey measurements, operational experience, and numerical calculations to divide a nuclear power plant into a reasonable number of EMC zones. This would primarily help to have a better overview of the current EMC level in the plant and to unify emission and susceptibility requirements for the new equipment. In this paper, special attention is given to the preparation and performance of the in situ tests, which present the most challenging step of the methodology. In addition, the paper proposes some of the possible applications of the numerical calculations and addresses their challenges and limitations. The novel classification methodology, inspired by the equipment qualification program, is illustrated with examples from Krško Nuclear Power Plant, which recently performed a comprehensive EMC assessment.

In this paper, shielding analysis was performed to determine neutron and gamma dose rates around the transfer cask HI-TRAC VW (Holtec International Transfer Cask Variable Weight) loaded with spent fuel assemblies (SFA) from nuclear power plant (NPP) Krsko spent fuel dry storage (SFDS) campaign one. The HI-TRAC VW is a multilayered cylindrical vessel designed to accept a multipurpose canister (MPC) during loading, unloading, and transfer to dry storage building. The MPC can contain up to 37 spent fuel assemblies. The analysis was divided into two steps. The first step was the source term generation using origen-s module of the scale code package. The source was calculated based on the operating history of spent fuel assemblies currently located in the NPP Krsko spent fuel pool. The obtained particle intensities and source spectra of the SFA were used in the second step to calculate the dose rates around the transfer cask. A comprehensive hybrid shielding analysis included the calculation of dose rates resulting from fuel neutrons and gammas, neutron-induced gammas (n-γ reaction), and hardware activation gammas under normal conditions and during accident scenario. To obtain the dose rates within the acceptable uncertainties, forward weighted consistent adjoint driven importance sampling (FW-CADIS) variance reduction scheme, as implemented in advantg code, was adopted for accelerating final mcnp6.2 calculations. The dose rates around HI-TRAC VW cask were calculated using mcnp6.2 code for all 16 cask loadings belonging to campaign one in order to illustrate the impact of fuel assembly selection schemes proposed by company responsible for project realization - Holtec International (HI).

Passive containment filtered vent (PCFV) was installed in Nuclear Power Plant Krsko (NEK) in 2013 as part of the safety upgrade program. It is intended for severe accident consequences prevention and mitigation by ensuring the containment integrity. When the pressure in the containment reaches limiting value, containment atmosphere is released into the environment through the PCFV system exhaust line. But, before release in the environment, the containment atmosphere passes through five aerosol filters in the containment and one iodine filter in the auxiliary building (AB) to reduce isotopic activity. 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 accident station blackout (SBO) in NEK, which was simulated using the MELCOR code. Its results were input for the radionuclide transport and removal and dose estimation (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.

Containment is the last barrier for release of radioactive materials in the case of an accident in the nuclear power plant (NPP). Its overall integrity is tested during a containment integrated leak rate test (CILRT) at the design pressure, at regular intervals. Due to applied risk based licensing, the test intervals can be increased up to once in 10 years and beyond. Taking that into account it is important to prepare the test properly and to use obtained results to assess the real status of the containment. The test can be used to verify existing containment calculation models. There is a potential benefit of verified computer models usage for the explanation of some test results, too. NPP Krško has performed CILRT during the plant outage in 2016. The paper presents a comparison between measured data and results calculated using a multivolume GOTHIC (Generation Of Thermal Hydraulic Information For Containment) model. The test scenario was reproduced using limited available data up to the end of the pressurization phase. The depressurization phase is calculated by the code and measured leakage rate is implemented in the model. Taking into account the necessary adjustments in the model, overall prediction of the measured results (in terms of pressure, temperature and humidity) is very good. In the last phase of the test some non-physical behavior is noticed (without influence on overall test results), probably caused by the combination of air redistribution within the containment and influence of heat transfer to plant systems that were in the operation during the test. GOTHIC model was used to check sensitivity of the predicted pressure (leak rate) to different heat inputs and to investigate the influence that operation of only one reactor containment fan cooler (RCFC) train during pressurization can have on the mixing of air within the containment. In addition, the influence of currently used weighting factors (weighting of measured temperature, relative humidity and pressure data) on the used test methodology is investigated. The possible non-conservative direction of the influence (currently used weighting factors are giving lower leakage rate) was demonstrated and a new set of weighting factors is proposed too.