We investigate the effect of limiting the number of retransmission trials on the stability of slotted ALOHA systems with no capture effect. Using an argument of the catastrophe theory, we prove that a slotted ALOHA system is mono-stable, if the number of retransmission trials is limited to, at most, eight. We also show how the bistable region is enlarged, as the number of retransmission trials grows over eight.

Recently, digital instrumentation and control systems have been increasingly installed for important safety functions in nuclear power plants such as the reactor protection system (RPS) and the actuation system of the engineered safety features. Since digital devices consist of not only electronic hardware but also software that can control microprocessors, the functions specific to digital equipment such as self-diagnostic functions have been becoming available. These functions were not realized with conventional electric components. On the other hand, it has been found that it is difficult to model the digital equipment reliability in probabilistic risk assessment (PRA) using conventional fault tree analysis technique. OECD/NEA CSNI Working Group of Risk Assessment (WGRisk) set up the task group DIGREL to develop the basis of reliability analysis method of the digital safety system and is now discussing about several issues including quantitative dynamic modeling. This paper shows that, taking account of the relationship among the RPS failures, demand after the initiating event, detection of the RPS fault by self-diagnostic or surveillance tests, repair of the RPS components and plant shutdown operation by the plant operators as a stochastic process, the anticipated transient without scram (ATWS) event can be modeled by the event logic fault tree and Markov state-transition diagrams assuming the hypothetical 1-out-of-2 digital RPS.

After the severe accident in Fukushima Daiichi Nuclear Power Station, safety improvement and enhancement have been installed. In midterm and long term, continuous efforts to improve and enhance safety are required, and technical basis and fundamentals are needed to achieve them.Probabilistic Risk Assessment for seismic event (seismic PRA) is an effective measure to consider the countermeasures and improvement plans to secure the further safety of nuclear power plants regarding to seismic risk for the earthquake exceeding the design basis earthquake ground motion. However, the application of seismic PRA has not been utilized sufficiently so far. One of the reasons is that there is not enough agreement among stakeholders regarding to the evaluation methodology and consideration of uncertainty for decision-making.This study proposes the mathematic framework to treat the uncertainty properly related to the evaluation of core damage frequency (CDF) induced by earthquake, the methodology to evaluate the fragility utilizing expert knowledge, the probabilistic model to cope with the aleatory uncertainty as well as the development of analysing code including these considerations for the improvement of the reliability of the methodology and enhancement of utilization of the products of seismic PRA. This paper presents current status and some results from scoping calculations.

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