Source Term Estimation under the SBLOCA-Induced Severe Accident Condition in the SMART

The SMART is a system-integrated modular reactor in which a nuclear steam supply system with a thermal power of 365 MW is contained inside of the reactor vessel. Although the probability is very low, the reactor core can be damaged during a small break loss-of-coolant accident when both the passive safety injection system and the passive residual heat removal system are completely unavailable. In this work, a total of five cases were analyzed considering the reactor vessel condition and the availability of the radioactivity removal tanks and the ancillary containment spray system as containment condition variables using MELCOR code. It was estimated that there is no containment failure based on pressure, hydrogen mole fraction, and ablation depth, so that the release fractions of the 12 classes of fission products in MELCOR were evaluated considering design leak only for all cases. The overall source term of the case in which the integrity of the reactor vessel is maintained by the early initiation of the cavity flooding system was similar to that of the reactor vessel failure case. While the release fraction of cesium to the environment was analyzed to increase when there is no water in the radioactivity removal tanks, the fraction is small enough at which the radioactivity of the released cesium-137 remains well below 100 TBq, a regulatory limit. Moreover, it was found that the source term can be cut in half if the ancillary containment spray system is available. The results of this study verify the safety performance of the SMART under the small break loss-of-coolant severe accident condition with respect to the source term of interest.

Analysis on passive residual heat removal system with heat pipes for longterm decay heat removal of small lead-based reactor

.A lead-based reactor with employing heat pipes as passive residual heat removal system (PRHRS) for longterm decay heat removal was designed. Three-dimensional computational fluid dynamics (CFD) software FLUENT was adopted to simulate the thermal-hydraulic characteristics of the PRHRS under Station-Black-Out (SBO) accident condition. The results showed that heat in the core could be removed smoothly by the PRHRS, and the core temperature difference is less than 20 K.

Sensitivity and Uncertainty Analysis of the Maximum Fuel Temperature under Accident Condition of HTR-PM

The maximum fuel temperature under accident condition is the most important parameter of inherently safe characteristics of HTR-PM, and the DLOFC accident may lead to a peak accident fuel temperature. And there are a variety of uncertainty sources in the maximum fuel temperature calculations, and thus the contributions of these uncertainty sources to the final calculated maximum fuel temperature should be quantified to check whether the peak value exceed the technological limit of 1620°C or not. Eight uncertainty input parameters are selected for inclusion in this uncertainty study, and their associated 2 standard deviation uncertainties and probability density functions are specified. Then, the DLOFC thermal analyses and uncertainty analysis are performed with the home-developed ATHENA and CUSA. The numerical results indicate that the pebble-bed effective conductivity and the decay heat contribute the most of the uncertainty in the DLOFC maximum fuel temperature while this peak fuel temperature is most sensitive to the initial reactor power and the decay heat. In short, uncertainties in these selected eight parameters lead to the two standard deviation (2σ) uncertainty of ±77.6°C (or 5.2%) around the mean value of 1493°C for the maximum fuel temperature under DLOFC accident of HTR-PM. At the same time, the LHS-SVDC method of CUSA is recommended to propagate uncertainties in inputs and 100–200 model simulations seem to be sufficient to get an uncertainty prediction with full confidence.