Steady-State Simulations of a 30-Tube Once-Through Steam Generator with the RELAP5/MOD3 and RELAP5/MOD2 Computer Codes

1991 ◽  
Vol 96 (1) ◽  
pp. 123-128 ◽  
Author(s):  
Yassin A. Hassan ◽  
Parvez Salim
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Computer Codes

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

Multiphysics Analysis System for Tube Failure Accident in Steam Generator of Sodium-Cooled Fast Reactor

2013 ◽  
Author(s):  
Akihiro Uchibori ◽  
Shin Kikuchi ◽  
Akikazu Kurihara ◽  
Hirotsugu Hamada ◽  
Hiroyuki Ohshima
Keyword(s):  
Heat Transfer ◽  
Steam Generator ◽  
Chemical Reaction ◽  
Numerical Models ◽  
Stress Evaluation ◽  
Transfer Tube ◽  
Failure Propagation ◽  
Computer Codes ◽  
Analysis System ◽  
Tube Failure

Multiphysics analysis system was newly developed to evaluate possibility of failure propagation occurrence under heat transfer tube failure accident in a steam generator of sodium-cooled fast reactors. The analysis system consists of the computer codes, SERAPHIM, TACT, RELAP5, which are based on the mechanistic numerical models. The SERAPHIM code calculates the multicomponent multiphase flow involving sodium-water chemical reaction. In this study, numerical models for the chemical reaction about production of a sodium monoxide and its transport process were constructed to enable evaluation of a wastage environment. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The TACT code was integrated by the numerical models of the fluid-structure thermal coupling, the temperature and stress evaluation, the wastage evaluation and the failure judgment. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the rapidly heated tube in the present work.

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

scholarly journals Overview on Steady-state Nuclear Methods for BWR Nuclear Core Design and Analysis

2018 ◽  
Vol 35 (3) ◽  
pp. 223-227
Author(s):  
Ren-Tai Chiang
Keyword(s):  
Steady State ◽  
Cross Section ◽  
Neutron Cross Section ◽  
Verification And Validation ◽  
Boiling Water Reactors ◽  
Nuclear Core ◽  
Nuclear Methods ◽  
Computer Codes ◽  
Water Reactors ◽  
Calculation System

An overview on nuclear methods for boiling water reactors (BWR) core design and analysis is provided based on the ANS Standard 19.3. The steady-state BWR nuclear methods, composed of neutron cross section library generation method, lattice physics method and core physics method, are systematically reviewed and associated computer codes in common use for BWR core design and analysis are listed. Verification and validation, the two complementary aspects in determining the range of applicability of the calculation system, are discussed extensively. The biases and uncertainties for the predictions from the calculation system over its demonstrated range of applicability are also discussed.

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.

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