The NuScale Small Modular Reactor (SMR) is an integrated Pressurized Water Reactor (iPWR) with the coolant flow based on the natural circulation. The reactor core consists of 37 fuel assemblies similar to those used in typical PWRs, but only half of their length to generate 160MW thermal power (50 MWe). Current study involves the development of a NuScale-SMR model based on its Design Certification Application (DCA) data (from NRC) using RELAP/SCDAPSIM. The turbine trip transient (TTT) was simulated and analysed. The objective was to assess this version of the code for natural circulation system modeling capabilities and also to verify the input model against the publicly available TTT results obtained using NRELAP5. This successful benchmark confirms the reliability of the thermal hydraulic model and allows authors to use it for further safety and severe accident analyses. The reactor core channels, pressurizer, riser and downcomer pipes as well as the secondary steam generator tubes and the containment were modeled with RELAP5 components. SCDAP core and control components were used for the fuel elements in the core. The final input deck achieved the steady state with the operating conditions comparable to those reported in the DCA. RELAP/SCDAPSIM predictions are found to be satisfactory and comparable to the reference study. It confirms the code code capabilities for natural circulation system transients.

The Quench-06 experiment (ISP-45) has been used as a benchmark and training aid for Innovative Systems Software (ISS) and our users/students since it was completed in the early 2000s. The experiment was first analyzed by several international organizations using RELAP/SCDAPSIM/MOD3.2. These results were submitted to the “blind” and “open” phases of the ISP. The experiment was subsequently used for basic user training for experimental analysis by our RELAP/SCDAPSIM/MOD3.4 and MOD3.5 users. It is also used extensively in our university support and training internships. This paper describes an integrated uncertainty analysis of the QUENCH-06 electrically heated experiment, looking at the influence of uncertainties in experimental conditions and important models/correlations. The QUENCH calculations demonstrated the use of the new IUA, “Integrated Uncertainty Analysis”, option introduced into RELAP/SCDAPSIM/MOD3.4 in the summer of 2017 and MOD3.5 in the fall of 2017. The input models and results from both versions are discussed in the paper. The MOD3.4 results are based on the original input model developed for MOD3.2 and refined in the open phase of the ISP. The MOD3.5 results are based upon two base input models. The first was developed specifically to test the impact of MOD3.5 modeling improvements for the Quench electrically heated fuel rod simulator and was used in an early paper presented at this meeting in 2014. The second input model has been refined as part of the university support and training internship program and was used originally in 2016 to look at the influence of different approaches in modeling the insulated shroud used to minimize radial heat losses. The uncertainty analysis provided in this paper looks at the influence of uncertainties in (a) the parabolic equations for Zircaloy oxidation, (b) the tungsten heater element resistances, (c) the convective heat transfer coefficients, (d) the contact resistance of the heater elements, and (e) the thermal conductivity of the porous zirconia used in the shroud. The uncertainty analysis demonstrated very clearly that a bias was introduced into the 2014 MOD3.5 input model. This bias was subsequently determined to be primarily associated to the modeling of the shroud and associated radial heat losses. This bias was reduced in the 2016 version of the MOD3.5 input model and will be further refined as our training activities on the modeling of experiments continue.

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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.

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