Effects of Tube Rupture Modeling Methods on MSSV Lift Time Following a MSGTR Event in PWR

2002 ◽  
Author(s):  
Ji Hwan Jeong ◽  
Ki Yong Choi ◽  
Keun Sun Chang
Keyword(s):  
Steam Generator ◽  
Flow Rate ◽  
Pressurized Water Reactor ◽  
Steam Generators ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Best Estimate ◽  
Main Steam ◽  
Safety Systems ◽  
Estimate System

A multiple steam generator tube rupture (MSGTR) event in APR1400, an advanced pressurized water reactor, is investigated using the best estimate thermal hydraulic system code, MARS1.4. The effects of parameters such as the number of ruptured tubes, rupture location, affected steam generator on analysis of the MSGTR event in APR1400 are taken into account. In particular, the effects of tube rupture modeling are compared. In the present study, single tube (STM) and double tube modeling (DTM) are examined for assessment on the main steam safety valve (MSSV) lift time. Nuclear steam supply system (NSSS) and several safety systems that are relevant to the APR1400 are modeled. Automatic safety systems are assumed to mitigate the MSGTR events including the reactor protection trip, reactor coolant pump trip, the pressurizer heaters, high-pressure safety injection (HPSI) pumps, and the valves for atmospheric dump, main steam safety, main steam isolation, and turbine stop and bypass. When five tubes are ruptured, the STM permits the operator response time of 2085 seconds before lifting of MSSVs. The effects of rupture location on the MSSV lift time is not significant in case of STM, while the MSSV lift time for tube-top rupture is found to be 25.3% larger than that for rupture at hog-leg side tube sheet in case of DTM. The MSSV lift time for the cases that both steam generators are affected (4C5x, 4C23x) are found to be larger than that for the single steam generator cases (4A5x, 4B5x) due to a bifurcation of the primary leak flow. The discharge coefficient of Cd is found to affect the MSSV lift time only for smaller value of Cd below 0.5. For larger values of Cd than 0.5, its effect on the leak flow rates as well as the MSSV lift time become negligible. It is found that the most dominant parameter governing the MSSV lift time is the leak flow rate. Whichever modeling method is used, it gives the similar MSSV lift time if the leak flow rate is similar, except the case of both steam generators are affected. Therefore, the system performance and the MSSV lift time of the APR1400 are strongly dependent on the break flow model used in the best estimate system code.

Age-Related Degradation of Steam Generator Internals Based on Industry Responses to Generic Letter 97-06

2002 ◽  
Author(s):  
M. Subudhi ◽  
E. J. Sullivan
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressurized Water Reactor ◽  
Long Term Effects ◽  
Steam Generators ◽  
Pressurized Water ◽  
Age Related ◽  
The U.S

This paper presents the results of an aging assessment of the nuclear power industry’s responses to NRC Generic Letter 97-06 on the degradation of steam generator internals experienced at Electricite de France (EdF) plants in France and at a United States pressurized water reactor (PWR). Westinghouse (W), Combustion Engineering (CE), and Babcock & Wilcox (B & W) steam generator models, currently in service at U.S. nuclear power plants, potentially could experience degradation similar to that found at EdF plants and the U.S. plant. The steam generators in many of the U.S. PWRs have been replaced with steam generators with improved designs and materials. These replacement steam generators have been manufactured in the U.S. and abroad. During this assessment, each of the three owners groups (W, CE, and B&W) identified for its steam generator models all the potential internal components that are vulnerable to degradation while in service. Each owners group developed inspection and monitoring guidance and recommendations for its particular steam generator models. The Nuclear Energy Institute incorporated in NEI 97-06, “Steam Generator Program Guidelines,” a requirement to monitor secondary side steam generator components if their failure could prevent the steam generator from fulfilling its intended safety-related function. Licensees indicated that they implemented or planned to implement, as appropriate for their steam generators, their owners group recommendations to address the long-term effects of the potential degradation mechanisms associated with the steam generator internals.

Control Strategy Investigation for a Multi-Purpose Modular Small Pressurized Water Reactor With Once-Through Steam Generators

2018 ◽  
Author(s):  
Qian Ma ◽  
Peiwei Sun
Keyword(s):  
Flow Rate ◽  
Control Strategy ◽  
Pressure Vessel ◽  
Control Strategies ◽  
Reactor Power ◽  
Pressurized Water Reactor ◽  
Steam Generators ◽  
Pressurized Water ◽  
Water Reactor ◽  
Load Following

A new multi-purpose modular small pressurized water reactor with once-through steam generators is being designed in China. Its key parameters are different from traditional large pressurized water reactor. There are sixteen once-through steam generators divided into two groups inside of the pressure vessel. The four coolant pumps are located on the periphery of the pressure vessel. The coolant is heated by the core and transported the heat to the secondary loop by once-through steam generators. The superheated steam is generated, and its dynamics are different from those of U-tube steam generators. The relationship between the reactor and turbine is also complicated and needs to investigate. The control strategies of traditional large pressurized water reactor cannot be applied directly to the small reactor with once-through steam generators. Therefore, it is necessary to investigate suitable control strategies of the multi-purpose modular small reactor with once-through steam generators. Three control strategies are proposed and investigated in this study: turbine-leading, reactor-leading and feedwater-leading. With the reactor-leading strategy, the reactor power is adjusted by moving the control rod. The coolant temperature follows the change of the reactor power. Feedwater flow is applied to regulate the steam pressure. The steam flow rate follows the change of the feedwater flow rate to satisfy the demand power. With the turbine-leading strategy, the steam valve is adjusted which will influence the steam flow to satisfy the demand power. The feedwater-leading control strategy is adjusting the feed water flow rate corresponding to the demand power which has been measured. And reactor power and turbine load vary with feedwater flow rate. Input-output pairings of the control systems are determined based on the different strategies and proportion-integral-derivative (PID) controllers are tuned to meet the control requirements. To evaluate the performance of control strategies, power maneuvering events including a 10%FP (Full Power) step change and a ramp change with a rate of 5%FP/min are simulated. The processes of important control parameters varying with time are compared and evaluated to obtain the suitable one. Conclusions can be drawn from the simulation analyses of the control performance. The reactor-leading control strategy is best for the base-load operation. The turbine-leading control strategy is more suitable for load-following operation. The feedwater leading control strategy can be applied to load-following operation with smooth load adjustment.

Effects of a Venturi-Type Flow Restrictor on the Thermal-Hydraulic Response of the Secondary Side of a Pressurized Water Reactor Steam Generator to a Main Steam Line Break

2016 ◽  
Vol 138 (4) ◽  
Author(s):  
Jong Chull Jo ◽  
Frederick J. Moody
Keyword(s):  
Steam Generator ◽  
Steam Line ◽  
Cfd Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Main Steam Line ◽  
Main Steam ◽  
Secondary Side ◽  
Hydraulic Response

A numerical analysis has been performed to simulate the transient thermal-hydraulic response to a main steam line break (MSLB) for the secondary side of a steam generator (SG) model equipped with a venturi-type SG outlet flow restrictor at a pressurized water reactor (PWR) plant. To investigate the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB, numerical calculation results for the SG model equipped with the flow restrictor are compared to those obtained for an SG model without the restrictor. Both analysis models contain internal structures. The present computational fluid dynamics (CFD) model has been examined by comparing to a simple analytical model. It is confirmed from the comparison that the CFD model simulates the transient response of the SG secondary to the MSLB physically plausibly and minutely. Based on the CFD analysis results for both cases with or without the restrictor, it is seen that the intensities of the steam velocity and dynamic pressure are considerably attenuated in the SG model equipped with the restrictor comparing to the case in the SG model without the restrictor.

Simulation of Small Steam Generator Tube Break (SSGTB) in a Small Pressurized Water Reactor (SPWR)

2017 ◽  
Author(s):  
Huasong Cao
Keyword(s):  
Steam Generator ◽  
System Response ◽  
Outlet Temperature ◽  
Pressurized Water Reactor ◽  
Reactor Model ◽  
Design Data ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Steam Generator Tube ◽  
Water Reactor

Lots of efforts have been made to Research & Development of Small pressurized water reactors (SPWRs). Steam generator tube break occurs due to wear and corrosion frequently in the reactor. Among the breaks, Small Steam Generator Tube Break (SSGTB) is difficult to detect. Therefore, it is necessary to investigate the features of SSGTB. A small pressurized water reactor model has been established in this paper by Relap5. The model includes reactor core, pressurizer, steam generator, main coolant pump and auxiliary safety system. The core flow, pressure of pressurizer, core outlet temperature and secondary outlet steam temperature obtained based on steady-state calculation is compared with design data to verify the model correct. SSGTB is simulated by introducing a small break in the steam generator tube. The important parameters of reactor are recorded and analyzed. The procedure of SSGTB is analyzed and the system response features are summarized.

Validation of a CFD Analysis Model for the Thermal-Hydraulic Response of PWR Steam Generator to a Steam Line Break

2016 ◽  
Author(s):  
Jong Chull Jo ◽  
Bok Ki Min ◽  
Jae Jun Jeong
Keyword(s):  
Steam Generator ◽  
Steam Line ◽  
Cfd Analysis ◽  
Analysis Method ◽  
Pressurized Water Reactor ◽  
Analysis Model ◽  
Pressurized Water ◽  
Calculation Results ◽  
Main Steam Line ◽  
Main Steam

This paper presents a validation of a computational fluid dynamics (CFD) analysis method for a numerical simulation of the transient thermal-hydraulic responses of steam generator (SG) secondary side to blowdown following a main steam line break (MSLB) at a pressurized water reactor (PWR). To do this, the CFD analysis method was applied to simulate the same blowdown situation as in an experimental work which was conducted for a simplified SG blowdown model, and the CFD calculation results were compared with the experimental results. As the result, both are in reasonably good agreement with each other. Consequently, the present CFD analysis model has been validated to be applicable for numerical simulations of the transient phase change heat transfer and flow situations in PWR SGs during blowdown.

Traceability of Thermal Power Measurments: Modified Venturi Tubes

2005 ◽  
Author(s):  
Herb Estrada ◽  
Don Augenstein ◽  
Ernie Hauser
Keyword(s):  
Thermal Power ◽  
Pressurized Water Reactor ◽  
Steam Generators ◽  
Pressurized Water ◽  
Flow Measurements ◽  
Flow Meters ◽  
Tube Type ◽  
Traceable Calibration ◽  
Flow Element

This is the second of two papers describing the traceability of nuclear feedwater flow measurements. The first considered the challenges and methodology for establishing the traceability of chordal ultrasonic flow meters. This paper considers the challenges of establishing the traceability in a measurement using a flow element of the modified venturi tube type. It specifically considers the assumptions and uncertainties associated with the extrapolation, for use in the field, of tube calibration factors measured in the laboratory. To quantify these uncertainties, the in-situ performance of four modified venturi tubes is compared with the performance of four 8-path chordal ultrasonic flowmeters. The data analyzed were collected in the feeds of four steam generators in a large pressurized water reactor plant, each feed containing one meter of each type. The meters were initially calibrated in this series arrangement in a NIST traceable calibration lab and then operated in the same arrangement in the field.

Sign in / Sign up

Export Citation Format

Share Document