Steady-State and Transient Prediction of a 19-Tube Once-Through Steam Generator Using RELAP5/MOD1

1983 ◽  
Vol 60 (1) ◽  
pp. 143-150 ◽  
Author(s):  
Yassin A. Hassan ◽  
Charles D. Morgan
Keyword(s):  
Steady State ◽  
Steam Generator

A Numerical Model Without Truncation Error for a Steady-State Analysis of a Once-Through Steam Generator

2005 ◽  
Vol 150 (3) ◽  
pp. 315-324 ◽  
Author(s):  
Yoon Sub Sim ◽  
Eui Kwang Kim ◽  
Jae Hyuk Eoh
Keyword(s):  
Steady State ◽  
Numerical Model ◽  
Steam Generator ◽  
Truncation Error ◽  
Steady State Analysis ◽  
State Analysis

Maanshan PWR Loss of Flow Transients Analysis With TRACE

2009 ◽  
Author(s):  
Jong-Rong Wang ◽  
Hao-Tzu Lin ◽  
Wei-Chen Wang ◽  
Yi-Hsiang Cheng ◽  
Chunkuan Shih
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Nuclear Power ◽  
Thermal Power ◽  
Complete Loss ◽  
Transient Condition ◽  
Partial Loss ◽  
Core Flow ◽  
System Parameters ◽  
Trace Model

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.

Nodalization of NEK in APROS and Steady State Simulation

2014 ◽  
Author(s):  
Tadeja Polach ◽  
Klemen Debelak ◽  
Ivica Bašić ◽  
Luka Štrubelj
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Pressure Vessel ◽  
Reactor Core ◽  
Building Blocks ◽  
Reactor Pressure Vessel ◽  
Primary Circuit ◽  
Physical Parameters ◽  
Detailed Model ◽  
Reactor Pressure

A model of the primary circuit and part of the secondary circuit of the Slovenian Krško NPP – NEK was built using APROS - Advanced PROcess Simulation environment. The data used to describe the properties of the system modelled in APROS, were the data describing Krško NPP and its operational properties after the uprating and the introduction of the 18-month cycle. Basis for data collections, nodalization, structure and simplifications was NEK RELAP5\MOD3.3 Engineering Handbook and the 23rd cycle. In order to build a model describing all the important parameters, the available elements in APROS environment were used as building blocks for each system. The goal was to create a detailed model nodalization, which would give accurate results and would run on reasonable processing power. Each submodel was checked to verify that the partial results are within the allowable limits and that the description of the physical parameters is consistent with the real components. The model includes reactor pressure vessel, reactor coolant pumps and primary piping, steam generator, part of main steam, part of feedwater, pressurizer and reactor core kinetics. The regulation of pressurizer level and pressure, steam generator level and control rod is also modelled. The model consists of more than 400 thermal hydraulic volumes. The aim of building this model was a through thermal hydraulic analysis of the PWR systems present in the NPP Krško. Several simulations of the steady states at different power levels were performed. The resulting data describing the flow rates in steam generator feedwater, reactor pressure vessel, including bypass flows, heat transfer in reactor core and steam generator, thermal losses to containment, liquid level in pressurizer and steam generator, pressure drops in primary circuit and other parameters were then compared to the results of different types of calculation and to the testing data obtained from Krško NPP. The next step was to identify variations in results and determine whether they are consequence of wrong parameters, measurement deviation or numerical error. In that manner the model was verified and validated (in the sense of comparison with available system surveillance plant test results) to ensure the correct setup, initial and boundary conditions were applied in order to get reliable steady state results.

Integral Effect Test on Operational Performance of the PAFS (Passive Auxiliary Feedwater System) for a SGTR (Steam Generator Tube Rupture) Accident

2013 ◽  
Author(s):  
Yusun Park ◽  
Byoung-Uhn Bae ◽  
Seok Kim ◽  
Yun-Je Cho ◽  
Kyoung-Ho Kang
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Operational Performance ◽  
Experimental Result ◽  
Test Facility ◽  
Experimental Program ◽  
Primary System ◽  
Steam Generator Tube ◽  
Integral Effect ◽  
Effect Test

The PAFS is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus) which is intended to completely replace a conventional active auxiliary feedwater system. The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by adopting a natural convection mechanism; i.e., condensing steam in nearly-horizontal U-tubes submerged inside the PCCT (Passive Condensation Cooling Tank). With an aim of verifying the operational performance of the PAFS, the experimental program of an integral effect test is in progress at KAERI (Korea Atomic Energy Research Institute). The test facility, ATLAS-PAFS was constructed to experimentally investigate the thermal hydraulic behavior in the primary and secondary systems of the APR+ during a transient when the PAFS is actuated. Since the ATLAS-PAFS facility simulates a single train of the PAFS, the anticipated accident scenarios in the experiment include FLB (Feedwater Line Break), MSLB (Main Steam Line Break), and SGTR (Steam Generator Tube Rupture). Among them, SGTR was considered as one of the design basis accidents having a significant impact on safety in a viewpoint of radiological release. Therefore, the SGTR test was determined to be the integral effect test item in the frame of the ATLAS-PAFS experimental program. In this study, the PAFS-SGTR-HL-02 test was performed to simulate a double-ended rupture of a single U-tube in the hot side of the steam generator of the APR+. The three-level scaling methodology was taken into account to determine the test conditions of the steady-state and the transient. The pressures and temperatures of the system and the data related to the PAFS operation were collected with the measurement of the break flow. The initial steady-state conditions and the sequence of event of SGTR scenario for the APR+ were successfully simulated with the ATLAS-PAFS facility. And it was shown that the pressure and the temperature of the primary system were continuously decreased during the heat removal by the PAFS operation. The water pool in the PCCT was heated up to the saturation condition and the evaporation of the water made a decrease of the PCCT water level. It could be concluded from the present experimental result that the APR+ has the capability of coping with the hypothetical SGTR scenario with adopting the PAFS and the proper set-points of its operation.

RELAP5-3D Validation Study Using MB-2 Prototypical Steam Generator Steady-State Data

2005 ◽  
Vol 151 (3) ◽  
pp. 272-280
Author(s):  
A. Sawyer ◽  
M. Williamson ◽  
K. Zhao ◽  
A. Ruggles
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Validation Study ◽  
State Data
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