Effects on Core Damage Frequency to Sensitivity of Instrumentation and Control Reliability

Author(s):  
John T. Kitzmiller ◽  
Richard G. Anderson ◽  
Laura L. Genutis ◽  
Dennis M. Popp

Core Damage Frequency (CDF) and Large Release Frequency (LRF) can be greatly affected by the application of Instrumentation and Control (I&C) within the control and safety systems utilized in the operations of a nuclear power plant. It is important that the modeling of these systems be an accurate representation of the I&C design to capture the effects of the I&C on the operations of the nuclear power plant as far as control and safety that are needed for the effective and safe operation of power generation. Various sensitivities can be performed on I&C modeling in the AP1000 Probabilistic Risk Assessment (PRA) to determine the impact on CDF. These sensitivities can be in the areas of base reliability values (probabilities), common cause failures (CCF) of software as well as common cause (CC) factors of computer modules. Insights can be found as to the effects of these CCFs by varying the modeled probabilities in the most conservative direction to determine how the model will respond. The model response, i.e., CDF and cutset importance orders will provide insights as to how sensitive I&C modeling is to CCF and how sensitive the rest of the PRA model is to I&C. Increasing failure probability of the Protection and Safety Monitoring System (PMS) and Plant Control System (PLS), and Diverse Actuation System (DAS) to explore the minimum reliability that would support favorable CDF and LRF values, as well as, the effect of total failure of the PMS (no credit taken for PMS in core damage sequences) and PLS (no credit taken for PLS in core damage sequences) on CDF values are explored. Sensitivity analyses show that the CDF increases if no credit is taken for operator actions. For AP1000, this sensitivity study indicated a decrease in dependence on operator actions over conventional nuclear plants. This conclusion is likely a result of the increased reliability of the PMS to automatically actuate the given systems and components. For this reason, the most important system is PMS, in the case where no credit is taken for PMS in core damage sequences.

Author(s):  
Deucksoo Lee ◽  
Dong-Su Kim ◽  
Young-Taik Lee ◽  
O-Keol Kwon ◽  
Jung-Cha Kim

Ulchin nuclear power plant units 5&6 (UCN 5&6), which started excavation on January 1999, are two loop pressurized water reactors (PWR) with the capacity of 1000 MWe, and planned to start commercial operation on June, 2004 and June, 2005, respectively. The reactor coolant system of the UCN 5&6 consist of a reactor vessel, internals, and two steam generators, four reactor coolant pumps, a pressurizer and primary piping. Based on the system design of the first Korean Standard Nuclear Power Plant (KSNP), UCN 5&6 is designed to provide improvements in safety, reliability and cost by applying both advanced proven technology and experiences gained from the construction and operation of the previous KSNP plants. The result of the preliminary probabilistic safety assessment study for UCN 5&6 shows that the core damage frequency is lowered significantly. Several design improvement items have been adopted to the system design and contributed to lower the core damage frequency value. Among the design improvements, digital PPS and digital ESFAS are the key to the UCN 5&6 design. Furthermore, digitization of the Plant Protection System (PPS) and Engineered Safety Feature Actuation System (ESFAS) for the PWR is the first case in the PWR construction history. The Korean regulatory body reviewed the design concept of the digital PPS and digital ESFAS, and evaluated to be acceptable for the plant safety. Also, in-depth review on the detail design of the digital PPS/ESFAS and the special evaluation/audit for the software design process are underway to secure the software quality. The safety of the UCN 5&6 design has been evaluated through a two-year review on the preliminary safety analysis report. As a result, the construction permit was issued on May 17, 1999 by the government. In this paper, design characteristics of UCN 5&6 are discussed focussed on design improvements comparing with KSNR. And, some of the safety analysis results are presented as well as licensing status.


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