STUDY ON THERMAL CHARACTERISTICS OF U-SHAPED HEAT PIPE

The latest accident in Japan's nuclear power station became a valuable experience to start engaging passive cooling systems (PCS) more aggressively to improve safety aspects in nuclear power reactors being studied in Indonesia. This investigation is related to the U-shaped heat pipe (UHP) research as PCS of water in the cooling tank (CT). The objective of this research is to study the thermal characteristics of UHP as PCS in the CT. The experiment on small-scale UHP and simulation with RELAP5 code has been conducted to understand the performance of UHP. The experiment results of the small-scale UHP model will be used as a basic understanding of simulating and designing a UHP with big scaling. The study result showed the highest thermal performance of UHP was obtained when it operated on the higher temperature of heat load and higher air cooling velocity. The more UHPs inserted into the cooling pool, the more heat that can be discharged into the environment. This result also shows promising use of UHP for CT PCS. The use of UHP as PCS can enhance the safety aspect of the nuclear reactor, especially in station blackout event.

Evaluation of the Potential for Containment Bypass due to Steam Generator Tube Rupture in VVER-1000/V320 Reactor during Extended SBO sequence using SCDAP/RELAP5 code

A severe accident-induced of a Steam Generator (SG) tube releases radioactivity from the Reactor Coolant System (RCS) into the SG secondary coolant system from where it may escape to the environment through the pressure relief valves and an environmental release in this manner is called “Containment Bypass”. This study aims to evaluate the potential for “Containment Bypass” in VVER/V320 reactor during extended Station Blackout (SBO) scenarios that challenge the tubes by primarily involving a natural circulation of superheated steam inside the piping loop and then induce creep rupture tube failure. Assessments are made of SCDAP/RELAP5 code capabilities for predicting the plant behavior during an SBO event and estimates are made of the uncertainties associated with the SCDAP/RELAP5 predictions for key fluid and components condition and for the SG tube failure margins.

A pre-test analysis of ATLAS SBO with RCP seal leakage using MARS code

The accident in Fukushima Daiichi nuclear power plants shows the important of developing coping strategies for extended station blackout (SBO) scenarios of the nuclear power plants (NPPs). Many NPPs in United State of America are applying FLEX approach as main coping strategies for extended station blackout (SBO) scenarios. In FLEX strategies, outside water injection to reactor cooling system (RCS) and steam generators (SGs) is considered as an effective method to remove residual heat and maintain the inventory of the systems during the accident. This study presents a pretest calculation using MARS code for the Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) SBO experiment with RCP seal leakage scenario. In the calculation, the turbinedriven auxiliary feed water pumps (TDAFPs) are firstly used after SBO initiation. Then, the outside cooling water injection method is used for long term cooling. In order to minimize operator actions and satisfy requirements of APR1400 emergency operation procedure (EOP), the SGs Atmospheric Dump Valve (ADV) opening ratio, auxiliary feed water (AFW) and outside cooling water injection flow rates were investigated to have suitable values. The analysis results would be useful for performing the experiment to verify the APR 1400 extended SBO optimum mitigation strategy using outside cooling water injection.

Analysis of ESFR Decay Heat Removal Systems in Protected Station Blackout

As part of the ESFR-SMART project, safety assessments are being conducted on the updated European Sodium Fast Reactor (ESFR) design. An important part of the study is the evaluation of the reactor's behavior within hypothetical accidental conditions to assess and ensure that the accident would not lead to unexpected and disastrous events. In the current paper, the analyzed accidental scenario is the so called Protected Station Blackout (PSBO), where the offsite power is lost for the power plant, simulated by using the TRACE and SIM-SFR system codes. The assessment started from the simulation of the reactor behavior without the decay heat removal systems (DHRS). Following this, calculations of multiple DHRS arrangements have been performed to evaluate the individual and combined efficiency of the systems. Where it was possible, the results from the two system codes have been compared to show the consistency of the separate calculations. Through this study, the design of the DHRSs proposed at the beginning of the project have been investigated, and certain recommendations have been made for further improvement of the DHRS systems performance.

Wall Materials Effects on Sheltered Indoor Doses From an SMR Hypothetical Accident Release

Small modular reactors (SMRs) are expected as a suitable candidate to fulfill energy needs in the future. The regulation of the emergency planning zone (EPZ) has been a controversial issue. The possibility of smaller EPZs because of their small core size and passive safety functions has still under discussion. The major emergency responses to radiological incidents in the early phase are evacuation from the area and sheltering-in-place within a building. Comparison between the dose incurred during evacuation and that with sheltering-in-place is necessary to consider the proper protective actions. This study focuses on effect of wall materials on indoor doses for sheltered population from small modular reactor severe accident. The source term came from loss of coolant accident or station blackout, and the time change of air concentration and the ground deposition data was calculated with RASCAL, a software developed by NRC to provide dose projection around the plant. Then general one-story and two-story houses were set up, and 6 wall materials were selected for calculating indoor doses. Cloudshine and groundshine were calculated with Monte Carlo methods, and the shielding function of each house was evaluated by comparing the indoor dose with outdoor dose. The result will be a basis for calculating the radiological dose for sheltered cases in case of nuclear emergency for SMR, which will be valuable to have a more effective emergency planning.