Safety related valves within the nuclear power plant were governed by the Inservice Testing Program (IST) for scheduling tests or maintenances during the plant normal operation. Based on the program ...definitions, series of tests for these valves have to be performed and conducted periodically. Safety functions of safety related valves such as Motor Operated Valve (MOV), Air Operated Valve (AOV) and Check Valve (CV) under the event of Design Basis Accident (DBA) could be performed properly. IST program plays an important role for constantly maintaining these safety related valves under operable conditions; therefore safety functions such as emergency reactivity control, post-accident residue heat removal, post-accident radioactivity removal and containment isolation could be performed and satisfied during the DBA event or abnormal plant operation occurrences.
This study performed the Risk-informed evaluations for safety related valves in the IST program for BWR and PWR nuclear power plants in Taiwan. Outcome from this study, safety classifications for the safety related valves under IST program could be categorized, HSSC category valves could be identified, testing frequency relieves for the LSSC category could be suggested, systematic comparison of the safety significant classification between BWR and PWR nuclear power plants was presented and finally, due to the advantages of the RI-IST, both dose exposure for the plant staffs and outage cost reductions for the utility managements could be achieved.
A self-diagnostic monitoring system is a system that has the ability to measure various physical quantities such as temperature, pressure, or acceleration from sensors scattered over a mechanical ...system such as a power plant, in order to monitor its various states, and to make a decision about its health status. We have developed a self-diagnostic monitoring system for an air-operated valve system to be used in a nuclear power plant. In this study, we have tried to improve the self-diagnostic monitoring system to increase its reliability. We have implemented three different machine learning algorithms, i.e., logistic regression, an artificial neural network, and a support vector machine. After each algorithm performs the decision process independently, the decision-making module collects these individual decisions and makes a final decision using a majority vote scheme. With this, we performed some simulations and presented some of its results. The contribution of this study is that, by employing more robust and stable algorithms, each of the algorithms performs the recognition task more accurately. Moreover, by integrating these results and employing the majority vote scheme, we can make a definite decision, which makes the self-diagnostic monitoring system more reliable.
A pneumatic control valve is a piping element that controls the volumetric flow rate and pressure of a fluid; it is necessary to analyze the characteristics of the forces with respect to the opening ...of the valve in order to evaluate its operating performance. The forces occurring during operation are: resisting force and actuator force, where the load resistance is mostly affected by the fluid pressure difference of the valve. In this study, a force balance equation derived from the equilibrium relationship between the resisting force and the actuator force of an unbalanced globe valve is proposed, and the force balance equations are used to model the dynamic equations of a pneumatic unbalanced globe valve installed in nuclear power plants. A CFD analysis is also carried out to evaluate the pressure distribution and forces acting on the top and bottom planes of the valve plug. The results of this analysis have been verified through experimentation. This study has shown that the fluid pressure difference between the inlet and outlet of the valve, measured from the force balance equation of an unbalanced valve, should actually be examined with the fluid-pressure difference between the top and bottom side of the valve plug.PUBLICATION ABSTRACT
Implementing a risk-informed inservice testing (RI-IST) program provides a good aspect to the nuclear power plant licensee as an alternating program in the current ASME Section XI and 10 CFR 50.55a ...relevant testing programs. RI-IST concentrates testing resources on highly significant components, reduces excess testing burden, increases plant's availability, decreases dose rate on the plant's staff and also reduces cost on plant's operation and maintenance under nuclear safety expectations. Furthermore, RI-IST also gives a feature on prospective licensing change basis to a nuclear power plant's licensee.
This study will focus on safety-related and PRA-molded motor-operated valves (MOVs) and air-operated valves (AOVs) under the inservice testing program in boiling water reactor (BWR)-type nuclear power plant. As MOVs and AOVs have crucial safety functions throughout the nuclear power plant's safety systems, the steady operation and performance of MOVs and AOVs will definitely ensure that the nuclear power plant operates under safety expectations; therefore, this is the key reason to implement risk-informed evaluation for MOVs and AOVs in this study and being able to provide the safety significance classification for MOVs and AOVs under the current IST program to the plant's management. As a pilot study of RI-IST, the methodology of qualitative assessment will incorporate with probabilistic risk assessment (PRA) analyzing MOVs’ and AOVs’ safety significance within the current PRA model. The evaluating result will then classify its safety significance into a high-safety significant component (HSSC) and a low-safety significant component (LSSC) for MOVs and AOVs based on relevant regulatory criteria. With this initiating achievement, it can provide a cornerstone for further studies on the other types of valves and pumps in RI-IST program and also provide a valuable reference as proposing license change to the licensee.
Multiple failures of main steam isolation valve (MSIV) stem components have occurred at nuclear power plants. Stems are typically composed of ASM SA564 Grade 630 precipitation hardened steel, ...commonly known as 17-4 PH steel, in the H1100 condition. Several past valve stem failures have been attributed to thermal embrittlement, a condition under which the stem base material exhibits degraded mechanical properties due to microstructural changes induced by long-term exposure to a temperature range of approximately 260–316 °C (500-600 °F). Failure mechanisms of embrittled MSIV stems include sudden fracture due to loads from test actuations during plant outages as well as in-service fatigue crack growth (FCG) caused by vibrations from the steam flow. Three MSIV stems were removed from service after nine years in a pressurized water reactor (PWR) and were metallurgically and mechanically evaluated to determine the severity and extent of thermal embrittlement present. Findings from these evaluations were used to develop a condition-based projection of risk over time of the embrittlement-induced failure of the stems. It was found that the stem material removed from service did not exhibit significant metallurgical or mechanical evidence of embrittlement and that the probability of failure of the material in its current aged condition was small.
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