LOCA analysis of BWR-4/Mark-I nuclear power plant with TRACE

Kerntechnik ◽  
2021 ◽  
Vol 86 (2) ◽  
pp. 128-142
Author(s):  
J.-J. Huang ◽  
S.-W. Chen ◽  
J.-R. Wang ◽  
C. Shih ◽  
H.-T. Lin
Keyword(s):  
Nuclear Power ◽  
Coupled Model ◽  
Short Term ◽  
Containment System ◽  
Loss Of Coolant Accident ◽  
Main Steam Line ◽  
Main Steam ◽  
Temperature Responses ◽  
Term Analysis

Abstract This study established an RCS-Containment coupled model that integrates the reactor coolant system (RCS) and the containment system by using the TRACE code. The coupled model was used in both short-term and long-term loss of coolant accident (LOCA) analyses. Besides, the RELAP5/CONTAN model that only contains the containment system was also developed for comparison. For short-term analysis, three kinds of LOCA scenarios were investigated: the recirculation line break (RCLB), the main steam line break (MSLB), and the feedwater line break (FWLB). For long-term analysis, the dry-well and suppression pool temperature responses of the RCLB were studied. The analysis results of RELAP5/CONTAN and TRACE models are benchmarked with those of FSAR and RELAP5/GOTHIC models, and it appears that the results of the above four models are consistent in general trends.

Environmental Qualification of an Actuator Torque Switch

2006 ◽  
Author(s):  
A. M. Chan ◽  
S. L. Barreca ◽  
T. Kostela
Keyword(s):  
Nuclear Power ◽  
Normal Operation ◽  
Normal Contact ◽  
Electric Actuator ◽  
Loss Of Coolant Accident ◽  
Accident Simulation ◽  
Qualification Testing ◽  
Service Conditions ◽  
Main Steam Line ◽  
Main Steam

Environmental qualification testing was performed on a modified Limitorque torque switch for the torque switch safety functions in the Limitorque type SMB actuators located inside and outside containment in a nuclear power plant. The torque switch specimen was installed in a Limitorque SMB-1 electric actuator mounted on an 8” Velan gate valve and operated with a customized programmable logic controller to allow normal torque switch behaviour to be observed. The present paper describes the qualification testing performed. The modified torque switch was aged to a 30-year service life at the normal service conditions for both inside and outside containment. Aging included radiation, thermal and cycle aging. A seismic test and then a combined Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB) steam accident simulation were followed. After each stage of aging, functional tests were done to confirm normal insulation resistance, normal contact resistance and normal operation.

VVER-440 Containment Thermal Hydraulic Analyses With MELCOR and CONTAIN Codes

2002 ◽  
Author(s):  
Gregory Gromov ◽  
Igor Lola ◽  
Stanislav Sholomitsky ◽  
Dmitry Gumenyuk ◽  
Valery Shikhabutinov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
Design Basis Accident ◽  
Sar Model ◽  
Release Data ◽  
Main Steam Line ◽  
Safety Features ◽  
Main Steam ◽  
Containment Pressure

In support of the analyses for the Rivne Nuclear Power Plant (RNPP) VVER-440/213 (Ukraine) Safety Analysis Report (SAR), detailed MELCOR and CONTAIN models of the containment were developed. The RNPP containment features a bubble condenser tower with air locks and active and passive spray systems. Code input models were developed to accurately represent the containment volumes, room interconnections, structural masses, and the engineering safety features. Although MELCOR 1.8.3 [1] was the primary tool for the SAR containment analysis, comparison calculations were performed using CONTAIN Version 1.12 [2]. Consequently, both the response of the VVER-440 containment to limiting design conditions as well as a comparison of the two codes is presented. In the context of SAR requirements, the present application was performed for design basis accidents with conservative assumptions to compare the containment temperature and pressure with design criteria. The peak containment pressure and temperature were evaluated using the most intensive release of the primary and secondary coolant into the hermetic compartments, in particular, for the large break loss of coolant accident and main steam line break. Conservative coolant release data were evaluated using the RELAP5/Mod3.2 SAR model. The selection of the accident scenario, initial and boundary conditions, and the major results are presented. The results of the analyses will be included in the design basis accident analysis chapter of the RNPP SAR.

Strength Performance of Steel Catenary Riser Using Short Term and Long Term Analysis Methodologies

2016 ◽  
Author(s):  
Feng Wang ◽  
Roger Burke ◽  
Anil Sablok ◽  
Kristoffer H. Aronsen ◽  
Oddgeir Dalane
Keyword(s):  
Bending Moment ◽  
Current Profile ◽  
Short Term ◽  
Sea State ◽  
Strength Performance ◽  
Steel Catenary Riser ◽  
Analysis Methodology ◽  
Riser Design ◽  
Term Analysis

Strength performance of a steel catenary riser tied back to a Spar is presented based on long term and short term analysis methodologies. The focus of the study is on response in the riser touch down zone, which is found to be the critical region based on short term analysis results. Short term riser response in design storms is computed based on multiple realizations of computed vessel motions with various return periods. Long term riser response is based on vessel motions for a set of 45,000 sea states, each lasting three hours. The metocean criteria for each sea state is computed based on fifty six years of hindcast wind and wave data. A randomly selected current profile is used in the long term riser analysis for each sea state. Weibull fitting is used to compute the extreme riser response from the response of the 45,000 sea states. Long term analysis results in the touch down zone, including maximum bending moment, minimum effective tension, and maximum utilization using DNV-OS-F201, are compared against those from the short term analysis. The comparison indicates that the short term analysis methodology normally followed in riser design is conservative compared to the more accurate, but computationally more expensive, long term analysis methods. The study also investigates the important role that current plays in the strength performance of the riser in the touch down zone.

scholarly journals Analysis of Containment Pressure and Temperature Changes Following Loss of Coolant Accident (LOCA)

2015 ◽  
Vol 5 (4) ◽  
pp. 1-8
Author(s):  
Van Thai Nguyen ◽  
Ngoc Dung Kieu
Keyword(s):  
High Pressure ◽  
Cooling System ◽  
Primary System ◽  
Similar Work ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Pressure Water ◽  
Main Steam Line ◽  
Main Steam ◽  
Work Done

This paper present a preliminary thermal-hydraulics analysis of AP1000 containment following loss of coolant accident events such as double-end cold line break (DECLB) or main steam line break (MSLB) using MELCOR code. A break of this type will produce a rapid depressurization of the reactor pressure vessel (primary system) and release initially high pressure water into the containment followed by a much smaller release of highly superheated steam. The high pressure liquid water will flash and rapidly pressurize the containment building. The performance of passive containment cooling system for steam removal by condensation on large steel containment structure is a major contributing process, controlling the pressure and temperature maximum reached during the accident event. The results are analyzed, discussed and compared with the similar work done by Sandia National Laboratories.

The Need for Strategic Planning for Nuclear Power Plant I&C Upgrades

2002 ◽  
Author(s):  
J. Douglas Hill ◽  
Paul Moore
Keyword(s):  
Strategic Planning ◽  
Control Systems ◽  
Digital Technology ◽  
Nuclear Power ◽  
Power Plants ◽  
Short Term ◽  
Nuclear Plants ◽  
And Control ◽  
Plant Control

Nuclear power plants rely on Instrumentation and Control (I&C) systems for control, monitoring and protection of the plant. The original, analog designs used in most nuclear plants have become or soon will be obsolete, forcing plants to turn to digital technology. Many factors affect the design of replacement equipment, including long-term and short-term economics, regulatory issues, and the way the plant operates on a day-to-day basis. The first step to all modernization projects should involve strategic planning, to ensure that the overall long and short-term goals of the plant are met. Strategic planning starts with a thorough evaluation of the existing plant control systems, the available options, and the benefits and consequences of these options.

Suppression Methods of Acoustic Noise Generated in Main Steam Stop Valve

2018 ◽  
Author(s):  
Shiro Takahashi ◽  
Eiji Ozaki ◽  
Atsuyuki Minenaga
Keyword(s):  
Pressure Loss ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Acoustic Noise ◽  
Acoustic Resonance ◽  
Stop Valve ◽  
Cfd Analyses ◽  
Main Steam Line ◽  
Main Steam

The main steam stop valve (MSSV) is installed in the main steam line in thermal and nuclear power plants. The MSSV is a safety valve that instantaneously shuts off the steam flowing into the steam turbine in an emergency. However, as high-speed steam flow goes through the MSSV during even the rated operation, acoustic sound or noise is generated in the MSSV. Moreover, there is a possibility that flow-induced acoustic resonance occurs in the MSSV. Flow-induced acoustic resonance must be suppressed to decrease the sound noise. Reducing the pressure loss of the MSSV is also an important issue that cannot be neglected with respect to the plant thermal efficiency. Therefore, we have developed the MSSV which can suppress the flow-induced acoustic resonance and its pressure loss. To develop this MSSV, we conducted scale air tests and computational fluid dynamics (CFD) analyses that are described in this paper. Mach and Strouhal number of the test conditions were the same as those of an actual plant. Reynolds number was sufficiently large to obtain the developed turbulent flow. An unsteady compressible CFD analysis was also conducted using large eddy simulation as a turbulence model. We developed new tilted triangular tabs and installed them in the MSSV to suppress the intense vortex generation and pressure loss. As a result, the sound noise due to the flow-induced acoustic resonance was completely attenuated and pressure loss was reduced compared to the case using the current tilted tabs. CFD results also showed that the tilted triangular tabs could suppress the generation of intense vortexes and the flow-induced acoustic resonance.

Investigation of Blast Wave Effects on Containment Wall and Steam Generator

2018 ◽  
Author(s):  
Tae Jin Kim ◽  
Yoon-Suk Chang
Keyword(s):  
Blast Wave ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Criteria ◽  
High Energy ◽  
Element Analysis ◽  
Secondary Damage ◽  
Main Steam Line ◽  
Main Steam

When a sudden rupture occurs in high energy lines such as MSL (Main Steam Line) and safety injection line of nuclear power plants, ejection of inner fluid with high temperature and pressure causes blast wave, and may lead to secondary damage of adjacent major components and/or structures. The objective of this study is to assess integrity of containment wall and steam generator due to the blast wave under a postulated high energy line break condition at the MSL piping. In this context, a preliminary analysis was conducted to examine the blast wave simulation using coupled Eulerian-Lagrangian technique. Subsequently, a finite element analysis was carried out to assess integrity of the structures. As typical results, strain and stress values were calculated at the containment wall and steam generator, which did not exceed their failure criteria.

The Development of a Generic Qualified Digital Platform

2010 ◽  
Author(s):  
Victor S. S. Shyu ◽  
Ming-Huei Chen
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Systems Design ◽  
Nuclear Industry ◽  
Regulatory Body ◽  
Short Term ◽  
Digital Platform ◽  
And Control ◽  
Nuclear Instrumentation

The nuclear industry and research institutes in Taiwan are conducting a joint effort project to establish a self-reliant nuclear Instrumentation and Control (I&C) system design and fabrication capabilities in Taiwan. The purposes of this project, as called Taiwan’s Nuclear I&C System (TaiNICS), are planned to support digital upgrade of the existing nuclear power plants and the new nuclear installations in Taiwan. The project will be a long term pursuit of several task branches, including establishment of a generic qualified digital platform, qualification and certification processes, nuclear I&C systems design, safety analyses for software common cause failure, licensing, and collaboration. The short term goal of this project is to submit the License Topical Report (LTR) of a generic digital platform for the review of Taiwan’s regulatory body in 2013.

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