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 sensitivity study of physical models using in RELAP5 code based on FEBA experimental data

In the thermal-hydraulic safety analysis, simulation results using thermal-hydraulic codes depend mainly on modeling the physical phenomena built-in the codes. These models are the equations, and empirical formulas developed based on matching them to experimental data or based on the assumptions, simplifications for solving theoretical equations. Therefore, it is recommended that these physical models need to take into account the uncertainty they cause. The sensitivity study is performed to investigate the influence of physical models on the calculation results during the reflood phase after the loss of coolant accident. It is allowable to choose the physical models that have the most significant influence on the calculation results. This study conducted a sensitivity analysis of physical models in RELAP5 code based on experimental data measured on the FEBA test facility. Sixteen physical models have been selected for sensitivity analysis to find the most important models that influence the calculation results. Based on two criteria, the maximum cladding temperature and the quench time, the sensitivity analysis results show that four physical models significantly impact the calculation result. Four chosen physical models are considered further in the next step of their uncertainty evaluation.

Comparative Thermal Analyses Between Theoretical Mode and RELAP5 Code Simulation for OTSG of a Small PWR

Due to the low nuclear safety risk, low initial investment cost and short construction period, integrated small nuclear reactors have received wide attention from all over the world. As advanced new type of nuclear reactors, the technologies of integrated small nuclear reactors are in the process of exploration and development. Steam generators are used as the heat transfer system for energy exchange between the primary and secondary circuit in reactors, and their heat transfer analysis is very important for reactor design and development. Due to the simple structure, strong heat exchange capacity and timely load following, Once-Through Steam Generators (OTSGs) are the mainly used steam generators in the design of integrated small nuclear reactors. RELAP5/MOD4.0 is a commercial software developed by Innovative System Software, LLC for transient analysis of light water reactors (LWR). After years of development and improvement, RELAP5 has been a basic tool for analysis and calculation of various simulators of nuclear power plants. However, when RELAP5 models steam generators, only structural models related to straight pipes can be established, which is very inconvenient for the heat transfer research of Once-Through Steam Generators. In this paper, Once-Through Steam Generators with specified structural parameters are taken as the research object. The heat transfer calculation is performed on the simplified inclined tube models by RELAP5 code and the theoretical calculation of the spiral tube heat transfer models is also carried out. Comparing the steam outlet temperature on the primary and secondary sides, the heat exchange power, the average heat transfer coefficient and the tube length of different heat exchange zones under given primary and secondary side inlet fluid conditions, it is confirmed that the RELAP5 heat transfer calculation is verified for simplifying Once-Through Steam Generators with inclined tube models.

Validation of RELAP5/MOD3.4 for Flashing-Induced Instabilities in a Natural Circulation Loop

This paper aims to validate the effectiveness of the widely used Relap5 code in simulating two-phase natural circulation, and its capability to predict flashing-induced instabilities. The RELAP5 code is validated against experimental data from the NMR test facility, which was designed to investigate the flow instability for a BWR-type novel modular reactor (NMR). The simulations by RELAP5/MOD3.4 code had been performed under various conditions by changing system pressure, core inlet subcooling, core inlet flow resistance, and core heat power etc. The flow stability for a certain operating condition could be determined from the time trace profile of the loop natural circulation flow rate. The results showed that the simulated mass flow rate increased with increasing core inlet temperature, reproducing the experimental trend. And the maximum error between the experimental data and the calculated results is within 10%. The predicted natural circulation dimensionless numbers, the phase change number and inlet subcooling number, also had a good agreement with the experimental data. In general, the RELAP5 code is able to simulate flashing-induced instability and density wave oscillations, which occurred in the natural circulation test facility at low pressures. However, flashing tends to be suppressed at a higher pressure (400kPa). And the enlargement of core inlet resistance coefficient can also have a positive impact on natural circulation system stability.