Protection of Drain Pumps Against Transient Cavitation

1976 ◽  
Vol 98 (3) ◽  
pp. 401-410
Author(s):  
G. S. Liao
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Injection System ◽  
Design Parameters ◽  
Low Profile ◽  
Tank System ◽  
Pros And Cons ◽  
Protective Methods ◽  
Tank Pressure

It has been known that drain pumps in nuclear power plants may suffer cavitation under transient turbine load reductions [1]. Although increasing the size and height of the drain tank can prevent it, such provisions are often not practical, and in some instances even not possible. Some components, such as moisture separators, normally drain through drain receivers into the surface heater or drain tank. Since the drains are at the saturation condition, the drainage is generally accomplished by gravity. This necessitates locating the moisture separators considerably above the drain tank. With a limited physical elevation of the moisture separators relative to the main turbine in connection with a low profile of nuclear power plants, it is impractical to raise the heater-drain tank system somewhat similar to the deaerator in fossil power plants. This paper explores some conceptual protective methods, and briefly discusses their pros and cons as applied to both drain pumping forward and backward systems. The method of quantitative determination of design parameters required for each protective method is either referred to or derived. Based on simplicity, economy, and reliability, this paper concludes that the drain tank pressure decay control system appears to be the most promising protective method for the drain pumping forward system, whereas either the continuous feedwater injection system or the continuous drain subcooling system is the optimum method for the drain pumping backward system.

Strategies for Improving Organizational Communication: Case Studies of Nuclear Power Plants Revisited

1985 ◽  
Vol 29 (4) ◽  
pp. 375-379
Author(s):  
Marjorie B. Bauman ◽  
Margery Davidson Boulette ◽  
Harold P. Van Cott
Keyword(s):  
Organizational Communication ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Intervention Strategy ◽  
Weak Links ◽  
Alternative Programs ◽  
Survey Results ◽  
Pros And Cons ◽  
Maintenance Work

This EPRI-sponsored study reviewed the organizational communications used by nuclear power plants (NPPs) and identified weak links in the chain of coordination and information processing required to effectively perform corrective and preventive maintenance in the plants. Preliminary survey results from four NPPs showed that many communication areas deserve special attention in order to improve the work request process and decrease the time delays involved in the performance of maintenance work. This study evaluates two alternative programs designed to improve the effectiveness of the work request process. One approach involves evaluating an automated work request system as a way of improving interdepartmental communication and job performance as they relate to the implementation of maintenance work requests. Another approach assesses the effectiveness of interdepartmental meetings for supervisors as a method for improving organizational communication. Results of this longitudinal study are reported. Pros and cons of each intervention strategy are also discussed.

Innovative seismic resistant structure of shield building with base isolation and tuned-mass-damping for AP1000 nuclear power plants

2019 ◽  
Vol 36 (4) ◽  
pp. 1238-1257 ◽  
Author(s):  
Gangling Hou ◽  
Meng Li ◽  
Sun Hai ◽  
Tianshu Song ◽  
Lingshu Wu ◽  
...  
Keyword(s):  
Nuclear Power ◽  
High Performance ◽  
Seismic Isolation ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Base Isolation ◽  
Risk Mitigation ◽  
Design Parameters ◽  
Seismic Safety ◽  
Content Type

Purpose Seismic isolation, as an effective risk mitigation strategy of building/bridge structures, is incorporated into AP1000 nuclear power plants (NPPs) to alleviate the seismic damage that may occur to traditional structures of NPPs during their service. This is to promote the passive safety concept in the structural design of AP1000 NPPs against earthquakes. Design/methodology/approach In conjunction with seismic isolation, tuned-mass-damping (TMD) is integrated into the seismic resistance system of AP1000 NPPs to satisfy the multi-functional purposes. The proposed base-isolation-tuned-mass-damper (BIS-TMD) is studied by comparing the seismic performance of NPPs with four different design configurations (i.e. without BIS, BIS, BIS-TMD and TMD) with the design parameters of the TMD subsystem optimized. Findings Such a new seismic protection system (BIS-TMD) is proved to be promising because the advantages of BIS and TMD can be fully used. The benefits of the new structure include effective energy dissipation (i.e. wide vibration absorption band and a stable damping effect), which results in the high performance of NPPs subject to earthquakes with various intensity levels and spectra features. Originality/value Parametric studies are performed to demonstrate the seismic robustness (e.g. consistent performance against the changing mass of the water in the gravity liquid tank and mechanical properties) which further ensures that seismic safety requirements of NPPs can be satisfied through the use of BIS-TMD.

Design and Baseline Testing of Motor Operated Valves (MOVs) for Generation 3+ Nuclear Power Generating Plants

2012 ◽  
Author(s):  
Matthew E. Hobbs ◽  
Richard J. Gradle ◽  
Floyd A. Bensinger
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Lessons Learned ◽  
Baseline Data ◽  
Design Parameters ◽  
Production Testing ◽  
Class 1 ◽  
Plant Operations ◽  
Thrust And Torque

Flowserve is supplying motor operated valves (MOVs) to Generation 3+ nuclear power plants. These valves have been custom designed to meet the design and qualification criteria for ASME Section III, Class 1 nuclear service. To support plant operations, the valve designs benefit from the lessons learned from US NRC Generic Letter 89-10, “Safety-Related Motor-Operated Valve Testing and Surveillance”, the subsequent MOV testing programs, the Joint Owners Group (JOG) MOV Periodic Verification Program and ASME QME-1 functional qualifications. As a result, some new MOV designs are larger and heavier than typical valves of corresponding size and pressure class supplied to existing nuclear power plants. During the valve functional testing portion of valve manufacturer’s production testing, each MOV assembly is instrumented to record stem thrust and torque, position, motor operator voltage and current draw, and limit switch and torque switch function. The data are digitally recorded for further review and acceptance. The baseline data will allow the end user to confirm proper MOV installation and setup. The baseline data can also be compared to future in-situ test data to evaluate potential performance degradation during the nuclear power plant operation. This paper discusses: • MOV design parameters and design features • The production testing to establish and record MOV baseline functional data and • ASME QME-1-2007 qualification of these MOVs

A Study on Thermal Stress Evaluation of Safety Injection System for NPPs Considering Leakage Effect

2011 ◽  
Author(s):  
Hee-Dong Sung ◽  
Sun-Hye Kim ◽  
Ik-Joong Kim ◽  
Young-Jin Kim ◽  
Jeong-Soon Park ◽  
...  
Keyword(s):  
Thermal Stress ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Injection System ◽  
Stress Evaluation ◽  
Branch Line ◽  
Leakage Model ◽  
Cfd Analyses

Several piping failures caused by thermal stratification have been reported in some nuclear power plants since the early 1980s. However, this kind of thermal effect was not considered when the old vintage nuclear power plants were designed. Thermal stratification is usually generated by turbulent penetration from the RCS to branch line or leakage through damaged part of valve in branch line. In this paper, using the CFD analysis, characteristics of thermal stratification in a safety injection system of PWR plant were investigated and thermal stress evaluation was also conducted. First, CFD analyses were carried out on in-leakage model and out-leakage model according to operating condition. The case of out-leakage, the thermal stratification based on temperature distribution was generated a little at the rear of 1st valve. In contrast, significant thermal stratification was generated in front of 1st valve in in-leakage model because the effect of rapid flow velocity from RCS.

Development of a Water Injection System Fusible Plug for Severe Accidents at Nuclear Power Plants

2020 ◽  
Author(s):  
Sei Hirano ◽  
Daisuke Hirasawa ◽  
Yoshihisa Kiyotoki ◽  
Keisuke Sakemura ◽  
Keiji Sasaki ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Reliability ◽  
Water Injection ◽  
Severe Accident ◽  
Injection System ◽  
Normal Operation ◽  
Autonomous Operation ◽  
Performance Requirements

Abstract Background: When terminal stage of Severe Accident (SA) with no coolant injection at a nuclear power plant, the equipment that has cooled and solidified through water injection to a molten core that has ex-vessel and fallen outside of the pressure vessel will then be required to operate autonomously by heat detection, without external signals or power (e.g. electricity, air). The fusible plug operation is triggered by fusible alloy which receives heat from molten core and will melt. Because the fusible plug is also the boundary of Suppression Pool (S/P), high reliability is required for sealing performance. It is for that reason that Hitachi GE Nuclear Energy Ltd. (Hitachi-GE) has developed a fusible plug to serve as a device necessary to operate this system. Features of the Fusible Plug: The autonomous operation of the fusible plug is triggered by the melting of a fusible alloy, which is part of the fusible plug. However, the fusible alloy has a remarkably low mechanical strength and therefore is not suitable as a strength member. As such, it is necessary to ensure reliable plug sealing without applying a load to the fusible alloy so as to prevent the fusible plug from malfunctioning during normal operation. Therefore, to reduce the load to be applied to the fusible alloy, Hitachi-GE has developed a fusible plug structure that operates autonomously by detecting the ambient temperature without using the fusible alloy as a strength member. We have performed a verification test using this fusible plug and confirmed that it satisfies the predetermined performance requirements. Future Actions: Hitachi-GE is holding discussions on using the fusible plug at nuclear power plants in Japan. In the future, we plan to expand to the overseas.

scholarly journals Some Characteristics of Hypothetic TVSA with Fuel Enriched beyond 5% for VVER1000

2016 ◽  
pp. 22-26
Author(s):  
Ye. Bilodid ◽  
Yu. Kovbasenko
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Management Systems ◽  
Design Parameters ◽  
Spent Fuel ◽  
Fuel Management ◽  
Severe Accidents ◽  
Fuel Assemblies

The paper presents comparison of regular TVSA with average enrichment of 4,386% and hypothetical TVSA with enrichment of 10% based on design parameters and materials of TVSA fuel assemblies produced by TVEL (Russia), which today are widely used at nuclear power plants in Ukraine. It is shown that implementation of new fuel assemblies will result in improved use of fuel and increase of installed capability factor. At the same time, fresh and spent fuel management systems shall be modernized to meet relevant nuclear safety criteria. The paper analyzes possible criticality initiation at different stages of severe accidents related to core melt and using fuel with higher enrichment.

Nuclear power pros and cons: A comparative analysis of radioactive emissions from nuclear power plants and thermal power plants

2012 ◽  
Vol 67 (1) ◽  
pp. 120-127 ◽  
Author(s):  
V. A. Gordienko ◽  
S. N. Brykin ◽  
R. E. Kuzin ◽  
I. S. Serebryakov ◽  
M. V. Starkov ◽  
...  
Keyword(s):  
Comparative Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Thermal Power ◽  
Thermal Power Plants ◽  
Pros And Cons

Experiments and Analytical Simulation Work on an Innovative Steam-Injector-Driven Passive Core Injection Cooling System

2007 ◽  
Author(s):  
Michitsugu Mori ◽  
Tadashi Narabayashi ◽  
Shuichi Ohmori ◽  
Fumitoshi Watanabe
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Pressures ◽  
Cooling System ◽  
Steam Pressure ◽  
Injection System ◽  
The Core ◽  
Coolant Injection ◽  
Core Cooling

A Steam Injector (SI) is a simple, compact, passive pump which also functions as a high-performance direct-contact compact heater. We are developing this innovative concept by applying the SI system to core injection systems in Emergency Core Cooling Systems (ECCS) to further improve the safety of nuclear power plants. Passive ECCS in nuclear power plants would be inherently very safe and would prevent serious accidents by keeping the core covered with water (Severe Accident-Free Concept). The Passive Core Coolant Injection System driven by a high-efficiency SI is one that, in an accident such as a loss of coolant accident (LOCA), attains a higher discharge pressure than the supply steam pressure used to inject water into the reactor by operating the SI using water stored in the pool as the water supply source and steam contained in the reactor as the source of pressurization energy. The passive SI equipment would replace large, rotating machines such as pumps and motors, so eliminating the possibility of such equipment failing. In this Si-driven Passive Core Coolant Injection System (SI-PCIS), redundancy will be provided to ensure that the water and steam supply valves to the SI open reliably, and when the valves open, the SI will automatically start to inject water into the core to keep the core covered with water. The SI used in SI-PCIS works for a range of steam pressure conditions, from atmosphere pressure through to high pressures, as confirmed by analytical simulations which were done based on comprehensive experimental data obtained using reduced scale SI. We did further simulations and evaluations of the core cooling and coolant injection performance of SI-PCIS in BWR using RETRAN-3D code, developed using EPRI and other utilities, for the case of small LOCA. Reactors equipped with passive safety systems — the gravity-driven core cooling/injection system (GDCS) and depressurization valves (DPV) — would inevitably end up having large LOCA, even if they are initially small LOCA, as depressurization valves are forcibly opened in order to inject coolant from the GDCS pool to the GDCS water head at up to ∼0.2MPa. On the other hand, our simulation demonstrated that SI-PCIS could prevent large LOCA occurring in reactors by having by coolant discharged into the core in the event of small LOCA or when DPV unexpectedly open.

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