Chunkuan Shih

Application of TRACE and FRAPTRAN in the Spent Fuel Pool of Chinshan Nuclear Power Plant

In the nuclear power plant (NPP) safety, the safety analysis of the NPP is very important work. In Fukushima NPP event, due to the earthquake and tsunami, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. In this study, the safety analysis of the Chinshan NPP spent fuel pool was performed by using TRACE and FRAPTRAN, which also assumed the cooling system of the spent fuel pool failed. There are two cases considered in this study. Case 1 is the no fire water injection in the spent fuel pool. Case 2 is the fire water injection while the water level of the spent fuel pool uncover the length of fuel rods over 1/3 full length. The analysis results of the case 1 show that the failure of cladding occurs in about 3.6 day. However, the results of case 2 indicate that the integrity of cladding is kept after the fire water injection.

Film boiling and vapor explosions from small spheres

Experimental and analytical studies of free convection film boiling around small spheres are reported. The relation of film boiling to possible vapor explosions is discussed. The system simulates t...

Benchmark Calculations of Pressurizer Model for Maanshan Nuclear Power Plant Using TRACE Code

Pressurizer plays an important role in controlling the pressure of the primary coolant system in pressurized water reactor (PWR) power plants. An accurate modelling of the pressurizer is needed to determine the pressure histories of the primary coolant system, and thus to successfully simulate overall PWR power plant behavior during transients. The purpose of this study is to develop a pressurizer model, and to assess its pressure transients using the TRACE code version 5.0. The benchmark of the pressurizer model was performed by comparing the simulation results with those from the tests at the Maanshan nuclear power plant. Four start-up tests of the Maanshan nuclear power plant are collected and simulated: 1) turbine trip test from 100% power; 2) large-load reduction at 100% power; 3) net-load trip at 100% power; and 4) net-load trip at 50% power. The simulation results are in reasonable agreement with the start-up tests, and thus the pressurizer model built in this study is successfully verified and validated.Copyright

Evaluations and modifications of the EPRI-1 correlation on PWR critical heat flux predictions under normal and abnormal fuel conditions

The performance of five critical heat flux (CHF) correlations with the COBRA IIIC/MIT-1 code was evaluated. These correlations were evaluated against a data group comprised of 2943 axial nonuniform...

Application of LAPUR6 to Lungmen ABWR Stability Analysis

Lungmen nuclear power plant (NPP) is the first ABWR (Advanced Boiling Water Reactor) in Taiwan and still under construction. It has two identical units with 3,926 MWt rated thermal power each and 52.2×106 kg/hr rated core flow. The core has 872 bundles of GE14 fuel, and the steam flow is 7.637×106 kg/hr at rated power. According to the chapter 4 of Lungmen NPP FSAR (Final Safety Analysis Report), the design features of Lungmen NPP improve the core stability performance and assure that it is more stable than the current BWR (Boiling Water Reactor) NPP in the normal operating regions. In this research, the LAPUR6 stability analysis of Lungmen NPP is performed in order to versify the design features of Lungmen NPP which causes the more stable than the current BWR NPPs. The analysis results of LAPUR6 indicate that the design features of Lungmen NPP can improve the core stability performance effectively and result in the more stable than the current BWR NPPs.

Digital Instrumentation and Control Failure Events Derivation and Analysis by Frame-Based Technique

A frame-based technique, including physical frame, logical frame, and cognitive frame, was adopted to perform digital IC and (2) postulated ABWR digital IC and then can take early corrective actions to avoid the system hazard. This paper also discusses the advantage of Simulation-based method, which can investigate more in-depth dynamic behavior of digital I&C system than other approaches. Some unanticipated interactions can be observed by this method.Copyright

Assessment of TRACE CCFL Model with SBLOCA Experiment of IIST Facility

In this paper, the TRACE model for IIST facility is developed and verified with the Small Break loss of coolant accident (SBLOCA) experiment of IIST (Institute of Nuclear Energy Research Integral System Test) facility. By using the Wallis and Kutateladze correlations of countercurrent flow limitation (CCFL) model, the TRACE analyses results, such as break flow rate, primary pressure, and the temperature of cold-leg and hot-leg, are consistent with the IIST data. The results show the Kutateladze correlation of CCFL model can well predict the SBLOCA behavior and present good agreement with IIST experiment data in this paper. Besides, the sensitivity study results of Kutateladze correlation in CCFL model are verified and compared with the IIST data.

A STUDY OF STEAM-WATER COUNTERCURRENT FLOW MODEL IN TRACE CODE

This paper studied the countercurrent flow model in the TRACE code version 5.0. Steam and water are chosen as the working fluids that flow counter-currently in a circular pipe. Three types of the countercurrent flow models, the Wallis, the Kutateladze and the Bankoff correlations, are investigated. A single pipe model was built for the studies of the Wallis and the Kutateladze correlations, and the variable in the calculation model is the pipe diameter. A perforated plate model was built to study the Bankoff correlation, and the variables include the pipe diameter, the hole diameter, the number of holes and the plate thickness. The hydraulic diameter of the pipe varies from 2.5–200 mm for validating both the Wallis and the Kutateladze correlations. While validating the Bankoff correlation, the hydraulic diameter of the pipe is of 50 and 200 mm, and the plate thickness changes as 10 and 40 mm. Through this study, we validate the countercurrent flow model in the TRACE code, and provides comments on the application ranges of these three correlations.Copyright

The ultimate response guideline simulation and analysis by using TRACE for Lungmen ABWR nuclear power plant

Abstract In this research, the TRACE/SNAP model of Lungmen ABWR nuclear power plant (NPP) has been established for the simulation and analysis of ultimate response guideline (URG). The main actions of URG are depressurization and low pressure water injection of reactor and containment venting. This research focuses to assess the URG utility of Lungmen NPP under Fukushima-like conditions. This study consists of three steps. The first step is the establishment of Lungmen NPP TRACE/SNAP model. In order to evaluate the system response of TRACE/SNAP model, FSAR data (MSIV closure and loss of feedwater flow transient) were used to compare with the results of TRACE. The second step is the URG simulation and analysis under Fukushima-like conditions by using Lungmen NPP TRACE/SNAP model. In this step, the no URG case was also performed in order to evaluate the URG effectiveness of Lungmen NPP. In order to confirm the mechanical property and integrity of fuel rods, the final step is FRAPTRAN analysis. According to TRACE analysis results, the URG can keep the peak cladding temperature (PCT) below the criteria 1 088.7 K under Fukushima-like conditions which indicates that Lungmen NPP can be controlled in a safe situation. Nevertheless, if Lungmen NPP does not perform the URG under Fukushima-like conditions, the water level may drop lower than TAF after 1 100 s which means a safety issue about the fuel rods may be generated. The analysis results of FRAPTRAN also indicate the integrity of fuel rods cannot be kept under the above conditions.

Best Estimate Analysis of Maanshan PWR LBLOCA by TRACE Coupling With DAKOTA

This research is focused on the Large Break Loss of Coolant Accident (LBLOCA) analysis of the Maanshan power plant by TRACE-DAKOTA code. In the acceptance criteria for Loss of Coolant Accidents (LOCAs), there are two accepted analysis methods: conservative methodology and best estimate methodology. Compared with conservative methodology, the best estimate and realistic input data with uncertainties to quantify the limiting values i.e., Peak Cladding Temperature (PCT) for LOCAs analysis. By the conservative methodology, the PCTCM (PCT calculated by conservative methodology) of Maanshan power plant LBLOCA calculated is 1422K. On the other hand, there are six key parameters taken into account in the uncertainty analysis in this study. In PCT95/95 (PCT of 95/95 confidence level and probability) calculation, the PCT95/95 is 1369K lower than the PCTCM (1422K). In addition, the partial rank correlation coefficients between input parameters and PCT indicate that accumulator pressure is the most sensitive parameter in this study.Copyright