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An Italian MSE R&D programme on Nuclear Fission is funding, through ENEA, the design and testing of SPES3 facility at SIET, for IRIS reactor simulation. IRIS is a modular, medium size, advanced, integral PWR, developed by an international consortium of utilities, industries, research centres and universities. SPES3 simulates the primary, secondary and containment systems of IRIS, with 1:100 volume scale, full elevation and prototypical thermal-hydraulic conditions. The RELAP5 code was extensively used in support to the design of the facility to identify criticalities and weak points in the reactor simulation. FER, at Zagreb University, performed the IRIS reactor analyses with the RELAP5 and GOTHIC coupled codes. The comparison between IRIS and SPES3 simulation results led to a simulation-design feedback process with step-by-step modifications of the facility design, up to the final configuration. For this, a series of sensitivity cases was run to investigate specific aspects affecting the trend of the main parameters of the plant, as the containment pressure and EHRS removed power, to limit fuel clad temperature excursions during accidental transients. This paper summarizes the sensitivity analyses on the containment system that allowed to review the SPES3 facility design and confirm its capability to appropriately simulate the IRIS plant.

IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be built and operated at SIET laboratories. SPES3 simulates the primary, secondary, and containment systems of IRIS with 1 : 100 volume scale, full elevation, and prototypical thermal-hydraulic conditions. The simulation of the facility with the RELAP5 code and the execution of the tests will provide a reliable tool for data extrapolation and safety analyses of the final IRIS design. This paper summarises the main design steps of the SPES3 integral test facility, underlying choices and phases that lead to the final design.

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.

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.

The integrity of the containment will be challenged during a severe accident due to pressurization caused by the accumulation of steam and other gases and possible ignition of hydrogen and carbon monoxide. Installation of a passive filtered venting system and passive autocatalytic recombiners allows control of the pressure, radioactive releases, and concentration of flammable gases. Thermal hydraulic analysis of the containment equipped with dedicated passive safety systems after a hypothetical station blackout event is performed for a two-loop pressurized water reactor NPP with three integral severe accident codes: ASTEC, MELCOR, and MAAP. MELCOR and MAAP are two major US codes for severe accident analyses, and the ASTEC code is the European code, joint property of Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France) and Gesellschaft für Anlagen und Reaktorsicherheit (GRS, Germany). Codes’ overall characteristics, physics models, and the analysis results are compared herein. Despite considerable differences between the codes’ modelling features, the general trends of the NPP behaviour are found to be similar, although discrepancies related to simulation of the processes in the containment cavity are also observed and discussed in the paper.