Conference Agenda

Critical heat flux (CHF) can potentially cause catastrophic incidents in numerous thermal industries. At low-pressure conditions, due to high surface tension, the vapour bubbles grow in bigger sizes compared to the high-pressure conditions and may locally accumulate on the heated wall. Due to this, a local dry patch is formed on the heated wall causing a sharp rise in the wall temperature which is referred as DNB-type CHF. Therefore, CHF occurrence is the most crucial factor for ensuring the safe operation of thermal systems that experience coolant phase change. The present work predicts the subcooled flow boiling characteristics and CHF under low-pressure conditions in hexagonal fuel sub-assembly with plate-type spacer. In fuel assembly, spacer grids support fuel rods, reduce flow-induced vibrations, and increase coolant mixing. A WHFP model is employed with the EMF model to simulate low-pressure conditions. The Tolubinsky and Kostanchuk correlation for bubble departure diameter is modified to incorporate the bigger vapor bubble sizes that occur in low-pressure conditions. The current methodology demonstrates strong consistency when validated against the experimental data available for low-pressure conditions. The numerical analysis is further extended to investigate the influence of the spacer on subcooled flow boiling characteristics and the occurrence of CHF. The results show that the spacer acts as blockage, resulting in an increased pressure drop in the spacer's region and inducing a secondary flow within subchannels. In fuel sub-assembly with spacer, coolant velocity was found to be maximum at the spacer's position.

Accurate prediction of two-phase parameters in pressurized water reactors (PWRs) is crucial for the safety analysis of nuclear reactor cores. The refined subchannel model can enhance the spatial resolution of traditional subchannel codes by a factor of four. The ATHAS-H subchannel code, based on the refined subchannel model, has already completed the development of a single-phase flow and heat transfer calculation model. This study represents a continuation of previous work, developing a two-phase flow and heat transfer model for ATHAS-H based on a homogeneous flow model with slip ratio. The code was validated using experimental data from two different types of mixing grid crossflow experiments, the CE5×5 subcooled boiling experiment, and the PSBT bundle void fraction experiment. The validation included directional crossflow, subchannel outlet temperature, rod wall temperature, and void fraction. The results indicate that the ATHAS-H calculations are in good agreement with the experimental data. ATHAS-H can accurately reflect the non-uniformity of local parameters within subchannels caused by mixing grids and non-uniform power distribution in the rod bundle. This study demonstrates the advantages of high-precision subchannel code ATHAS-H in improving the accuracy of two-phase parameter predictions in PWRs. This capability lays a solid foundation for high-precision analysis of critical heat flux (CHF).

In LOCA, there are various uncertain variables that must be considered. These uncertainties determination have been developed by using simple random sampling method. Here, fuel burn-up must be considered and also thermal conductivity and random variables must be considered to derive staistic evaluation results. In this study, the developed SUPER code was used to perform SRS evaluation considering LODA optimization and uncertainty due to fuel burn-up. While considering how to apply the fuel burn up effect and thermal conductivity using the FRAPCON correlation, we introduce a method of fully automated evaluation using SUPER code for uncertainty calculation and optimization.

In this study, variables that should be considered fuel burn up condition, thermal conductivity, and uncertainty were selected to compare and review the PCT evaluation according to fuel burn up and the uncertainty distribution according to fuel burn up, and the most conservative evaluation results were derived. The evaluation results confirmed that the thermal conductivity and SRS statistical distribution results were limited aroung fuel burn up 30 MWD/kgU.In this study, 124 SRS(Simple Random Sampling) calculation is carred out by SUPER code.

However, the different burn up 7 cases between 0 MWD/kgU and 60 MWD/kgU are estimated.

Throught the sensitivity study, some conclusions are introduced as below:

1) Under various fuel burn up condition, in each case, 124 SRS calculations are carried out and PCT statistical distribution and 95/95 percent/accuracy results are introduced.

2) In high burn up conditions, PCT results are decreased by FQ burn down.

It is crucial to understand the behaviour of non-condensable gases (NCGs) in light water reactors (LWR) coolant systems, as their presence could exacerbate accidents and transients by interfering with heat transfer and flow paths, especially during long-term post-accident reactor cooling. The potential sources of NCGs in the reactor coolant system have been thoroughly investigated. However, despite the significant role that NCGs play in the reactor coolant system, there is a relative scarcity of published works addressing the details of the release and dissolution of NCGs. This paper presents previous research efforts on the release and dissolution of NCGs, covering experiments and modelling.

The release and dissolution of NCGs is an intricate phenomenon. When simulating the dissolution and release of NCGs, it is crucial to consider various physical aspects. These include the transport equation for dissolved gas content, release and dissolution rates, conservation equations for the gas phase, and the equations of state for a mixture of two components, where one component is water that may exist in liquid and vapour forms. This paper discusses the gaps in modelling NCG release and dissolution.

Additionally, the paper provides insights into the ongoing NCGDENSE project of LUT University of Finland which is funded by SAFER2028 (National Nuclear Safety and Waste Management Research Programme 2023-2028), which focuses on studying the release and dissolution of NCGs through analytical, experimental, and numerical methods. A separate-effect-test (SET) facility will be constructed to serve as a platform for direct NCG release and dissolution measurements.

A Loss of Coolant Accident (LOCA) might occur if there is a rupture in any of the piping systems linked to the reactor vessel. This rupture may lead to the continuous and uncontrolled loss of reactor coolant into containment. In the absence of an adequate emergency cooling water source, the subsequent increase in fuel temperature could cause damage to the fuel and the release of fission products. The use of best estimate codes and methodologies for simulating LOCA can offer detailed insights into the actual plant response, essential for evaluating the effectiveness of an emergency core cooling system. The TRACE thermal-hydraulics code was specifically developed to simulate transient scenarios in LWRs, including LOCA. This code was applied to simulate the full spectrum LOCA in several locations for BWR-6. The analysis was done using the actual plant configuration and operating conditions available from the core follow simulator. An advanced hot channel methodology was specially developed for these simulations. LOCA analysis for a specific BWR-6 proves the plant compliance to the applicable safety criteria and confirm the TRACE BWR-6 LOCA methodology applicability for the full spectrum LOCA analysis. In addition, this study helps to understand better the LOCA phenomenology as well as the plant response to LOCA transient.

When exposed to extreme conditions accompanied by loss of coolant accident (LOCA), the fuel rods experience swelling or, in severe cases, burst of fuel clad, in accordance with the heat up due to the loss of cooling performance during the accident. In such extreme conditions, the multiple deformation of fuel rods impairs coolable geometry, imposing further degradation of cooling performance with flow blockage. Nevertheless, the system code analysis has conventionally focused on the behavior of single hot pin, by which the details of its surroundings were lumped as averaged assembly-scale conditions. Thus, using the conventional modeling scheme, it is difficult to consider the effect of flow blockage accompanied by the deformation of individual fuel rods surrounding the hot pin of interest. Therefore, in this study, LBLOCA analysis has been performed on APR1400 plant using different modeling scheme for the reactor core by additionally modeling the individual fuel rods surrounding the hot pin in the subchannel-scale level. The effect of flow restriction with multiple deformation of fuel rods has been evaluated, using the thermal-hydraulic system code, MARS-KS. As a result, the clad expansion resulted in about 14% volume reduction in maximum within the subchannel where the hot pin was located. Despite of small deformation as such, the PCT of hot pin increased about 36K during reflood.