Conference Agenda

Severe accident analyses in nuclear power plants are highly complex and models applied in simulation codes are often derived based on limited amount of available experimental data. Evaluating the accuracy and uncertainty of such models provides valuable information for safety analyses, as well as further development needs and helps the priorisation of experimental resources. Consequently, there has been a growing interest in the recent years in uncertainty quantification and BEPU analyses in the context of severe accident analyses.

This paper investigates a postulated cold leg break combined with a station blackout scenario in a generic German PWR using the AC2 code package, developed by GRS. Both the reactor cooling circuit and containment are modelled in detail and the scenario is analysed starting from normal operational conditions up to the evaluation of the source term into the environment.

Based on the best estimate simulation an uncertainty and sensitivity analysis is performed. The selected uncertain input parameters are propagated through the model and the 95/95 uncertainty ranges are determined based on Wilks` findings. Furthermore, the Spearman Rank Correlation Coefficient is derived as sensitivity index. This allows to characterize both the variation in model responses and identify important uncertain parameters.

In 2019 was launched the Horizon-2020 MUSA project with the aim of reviewing the uncertainty sources as well as defining Uncertainty Quantification methodologies for assessing Severe Accidents (SA) scenarios. Energy Software Ltd. (ENSO) contributed to the Working Package 5 “reactor applications” with the simulation of a long-term Station Black Out (SBO) occurring in a Gen-II PWR. The Uncertainty Analysis (UA) was carried out with RELAP/SCDAPSIM/MOD3.4 (RS3.4) and its IUA2.0 module, using the Wald multi-variable form of the Wilks’ equation. A group of 20 input parameters and 19 Figures of Merit (FOM) were selected to assess the uncertainty propagated to the Source Term (ST) released to the containment in an in-vessel simulation. To improve the ST capabilities of the code, a Fission Product Transport model was implemented into RS3.4. The relevant outcome of ENSO contribution was the estimation of tolerance regions for the Fission Products, Gases and Debris materials released at the vessel failure. Such outputs can be then used as initial conditions and/or probability density functions (PDF) for ex-vessel calculations performed with containment codes. The sensitivity analysis was conducted for the original sample used in the UA, and for an increased sample size, to evaluate the consistency of the correlation coefficients and resulting PDFs. The results showed rather large standard deviations for some of the output parameters because of cliff edge phenomena in the material slumped to the lower plenum. Such results suggested the use of one-sided tolerance limits to set the initial conditions for posterior containment UA.

In 2019, the International Atomic Energy Agency (IAEA) launched the five-year Cooperative Research Project (CRP) I31033 to advance the understanding and characterization of sources of uncertainty and investigate their effects on the key figures-of-merit (FOMs) of the severe accident code predictions in water-cooled reactors (WCRs). Energy Software Ltd. (ENSO) contributed with an assessment of the uncertainty propagation in a long-term SBO scenario postulated for a CAREM-like integral PWR (iPWR). The study aimed at demonstrating the RELAP/SCDAPSIM/MOD3.5 (RS3.5) capability to carry out a BEPU calculation of a Severe Accident scenario in a single sequence from operational conditions to Reactor Pressure Vessel (RPV) creep rupture. The uncertainty analysis was conducted with the IUA2.0 module integrated into RS3.5 code, using the input-propagation methodology with a statistical description of the uncertainty proposed by Wilks. A group of 20 input parameters and 10 Figures of Merit (FOM) were selected for the assessment. The input parameters included boundary and initial conditions, material properties and code correlations, and the FOMs were related to the time of the main events and the fission product releases. To support the results, the Pearson, Spearman and Kendall correlation coefficients were analyzed for the selected input parameters and FOMs by using scalar values tables of the significance level. The relevant conclusions of the assessment are first, the importance of using the relative time for FOMs during core damage progression, and second, that including the Kendall formulation is advisable because it seems less dependent to singular data.

In an effort to address the inherent uncertainties in severe accident codes used in nuclear accident analysis, Energy Software S.L. (ENSO) has embarked on an innovative project aimed at developing a stepwise methodology for analyzing these uncertainties. The results obtained from two international projects, IAEA CRP I31033 and HORIZON-2020 MUSA, have provided ENSO with a solid foundation to identify and quantify the sources of uncertainty in severe accident analyses. However, they also revealed significant limitations, such as excessive computation time, multiple simulation errors, and truncation effects.

The proposed approach is stepwise, applying the Wilks/Wald method in two consecutive phases: the first linked to the "in-vessel" phase, where the accident prior to vessel failure is analyzed, and the second focused on the "ex-vessel" phase, examining subsequent events in the containment. With this methodology, ENSO plans to develop a model of a Generation II four-loop Westinghouse PWR reactor to simulate the "ex-vessel" phase of a low-pressure station blackout (SBO) accident using the MELCOR code. For the "in-vessel" phase, a previously created model with RELAP/SCDAPSIM will be used. The project also encompasses the development of pre- and post-processing tools with Python for uncertainty analysis with MELCOR, and the selection of input parameters based on probability distributions to apply the Wilks/Wald methodology.

Implementing the stepwise methodology will allow the identification and quantification of the sources of uncertainty at different stages of the accident, providing critical information for decision-making in emergency situations and for designing future research in this field.

In the process of severe accidents of the pressurized water reactor (PWR), the high temperature on the primary side of the steam generator and the high pressure difference between the primary and secondary sides can pose a high risk of creep rupture for the heat transfer tubes. Once the heat transfer tubes rupture, they lead to a bypass of the containment shell, causing radioactive materials to be directly released into the environment, and creating significant safety issues. The present study uses software to investigate the changes in parameters on the primary and secondary sides of the reactor under severe accident conditions, and calculates the rupture probability of the steam generator heat transfer tubes based on these parameter changes. MELCOR is used to simulate the station black out event (SBO), which has great impact on the heat transfer tubes. The main objective of the study is to clarify the changes in pressure and temperature on both sides of the heat transfer tubes under this accident condition, and to calculate the creep failure probability. The influence of microscopic and large cracks on the failure time of the heat transfer tubes is calculated. Considering that the surge line has the same failure risk, this study also shows the probability of the heat transfer tubes failing before the surge line.