Jong-Rong Wang

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

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

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

The Prediction of Startup Test With Load Rejection and Parametric Study of SBPCS for Lungmen Nuclear Power Plant

A Lungmen RETRAN-3D model has been constructed to predict the transient behaviors for startup test, furthermore verify the acceptance criteria specified in the documents of startup test procedure. This study focuses on the prediction of the startup test with Load Rejection (LR) with bypass and the parametric analysis of lead-lag time constants in pressure regulator. For the analysis of LR with bypass, the major mitigation functions, i.e., Selected Control Rods Run-In (SCRRI) and turbine bypass function, are simulated to examine whether scram is initiated during the transient or not. The analytic results show the reactor is brought to a steady state without scram. The neutron flux in the final state is around 34%, and the pressure regulator sensed maximum pressure rise is limited to a maximum of 3kPa. The result also shows that the 110% steam bypass capacity is capable to mitigate the power increase caused by the positive reactivity insertion as a result of pressure-wave-induced void collapse. For the parametric analysis of lead-lag time constants in pressure regulator, the time domain response of Steam Bypass and Pressure Control System (SBPCS) is demonstrated by a step change of pressure setpoint and different combinations of lead-lag time constants defined in pressure regulator. The results show that the responses, i.e., response time and overshooting, are minimized when the lag time constant is between 4 to 6 seconds and the lead time constant is 50% to 70% of lag time constant. The analysis result of SBPCS provides the trend as a reference for the adjustment of lead-lag time constants during the future Lungmen startup test.© 2010 ASME

Maanshan PWR Loss of Flow Transients Analysis With TRACE

TRACE model of Maanshan Nuclear Power Plant (three-loop PWR) was used to analyze Loss of Flow transient as defined in FSAR Chapter 15. The results were compared with those from RETRAN02 and LOFTRAN/THINC licensing analysis of Westinghouse Inc. Three different initiation events were involved in this analysis: Partial Loss of Flow (PLOF), Complete Loss of Flow-Under Voltage (CLOF-UV) and Complete Loss of Flow-Under Frequency (CLOF-UF). This paper compared important thermal hydraulic parameters at steady state, such as the pressure of pressurizer, cold-leg temperature, and the pressure of steam generator, etc.. It also compared system parameters under transient conditions, such as core thermal power, core flow rate, and pressure of pressurizer, etc.. It is concluded that the steady state results of TRACE calculations are in general good agreements with those from RETRAN02 and have a largest error of 3.03% in the steam generator flow. For transient condition, TRACE results are also comparable with those from LOFTRAN and RETRAN02. In summary, our studies show that Maanshan TRACE model is correct and accurate enough for future safety analysis applications.Copyright

LAPUR5.2 BWR Stability Analysis in Chinshan Nuclear Power Plant

Boiling water reactors have the unique coupling mechanisms between neutronic and two-phase flow thermal-hydraulic behaviors and may induce instability by unstable power/flow oscillations. At each core reload design, it is important to employ decay ratio for the purpose of analyzing system stability and determining its operating boundary. Making use of LAPUR5.2 and SIMULATE-3 codes, we have established a methodology to conduct such analysis. Comparisons made with vendor’s STAIF results indicated good agreements in decay ratios for Chinshan Nuclear Power Plant Unit 2 Cycle 21 reload design. This research focuses on the parametric sensitivity effect on the variation of decay ratio for different power/flow operating points. Based on the result of sensitivity studies, we presented fractional changes of decay ratios by varying certain important parameters under different power/flow points. It is concluded that density reactivity coefficient, gap conductance and recirculation loop gain on high operating power/flow points have larger fractional change of decay ratio.© 2006 ASME

Modeling Fukushima Dose Consequences with EPZDose

This paper introduces the recent developments and improvements of EPZDose and illustrates the evaluation methods of dose consequences from Fukushima Daiichi NPP accidents with available released source terms and meteorology data. EPZDose is a simulation code which can evaluate and display the dose consequences, faster than real time, in the surrounding area for radioactive materials release accidents such as possible nuclear power plant severe accidents or radioactive dirty bomb threat. With given source terms and the meteorological conditions (wind velocities, directions, Pasquill’s stability conditions, and fallout fraction), the code then evaluates and plots the distributions of whole body and thyroid doses by using Modified Gaussian Plume Model and proper dose conversion factors. We adopted this tool to evaluate the dose consequences from Fukushima Daiichi NPP. Our preliminary studies already indicated encouraging, within orders of magnitude results, compared with other publications. Further refinements of our modeling are under investigations.

Analysis of SBO ATWS for Maanshan PWR

Abstract Station blackout anticipated transient without scram (SBO ATWS) is considered as loss of off-site and on-site power but no credit for automatic reactor trip. SBO ATWS causes reactor coolant pump (RCP) trip, loss of all main feedwater pumps and turbine trip, then the reactor coolant system (RCS) pressure rises rapidly due to loss of heat removal paths. The ASME Code Level C service limit criteria of 22.06 MPa (3200 psig) is assumed to be an unacceptable plant condition in SECY-83-293. The simulation is performed by TRACE which is a thermal-hydraulic code developed by U.S. NRC. Three different AFW flows are modeled to ensure the pressures will not be beyond the criteria. RCP seal-leakage is concerned as a SBLOCA due to loss of RCP seal-cooling. Four possible leakage flows are modeled to examine the reactor core water level and temperature variation.

3RIP trip startup test simulation of TRACE/PARCS model for Lungmen ABWR under different power and flow conditions

Abstract Lungmen nuclear power plant project started long time ago, it is not yet commercially operated but Taiwan Power Company has already prepared for its startup tests. 3RIP trip startup test is one of them. Three of the 10 RIPs will be manually tripped in the test. Response of the plant for this transient will be watched and recorded to check if the test criteria are satisfied. This paper is a result of code simulation of 3RIP trip startup test of Lungmen ABWR nuclear power plant. Thermal hydraulic code TRACE coupled with neutronics code PARCS were used to build the simulation model of Lungmen nuclear power plant. Startup tests under different plant power and flow conditions are considered in this research. A sensitivity study on the impact of different pump moment of inertia has been performed. Simulation results with TRACE and PARCS shows that the acceptance criteria for this startup test can be satisfied and the impact of different pump inertia is little.