Jaejoo Ha

A fast BDD algorithm for large coherent fault trees analysis

Abstract Although a binary decision diagram (BDD) algorithm has been tried to solve large fault trees until quite recently, they are not efficiently solved in a short time since the size of a BDD structure exponentially increases according to the number of variables. Furthermore, the truncation of If–Then–Else (ITE) connectives by the probability or size limit and the subsuming to delete subsets could not be directly applied to the intermediate BDD structure under construction. This is the motivation for this work. This paper presents an efficient BDD algorithm for large coherent systems (coherent BDD algorithm) by which the truncation and subsuming could be performed in the progress of the construction of the BDD structure. A set of new formulae developed in this study for AND or OR operation between two ITE connectives of a coherent system makes it possible to delete subsets and truncate ITE connectives with a probability or size limit in the intermediate BDD structure under construction. By means of the truncation and subsuming in every step of the calculation, large fault trees for coherent systems (coherent fault trees) are efficiently solved in a short time using less memory. Furthermore, the coherent BDD algorithm from the aspect of the size of a BDD structure is much less sensitive to variable ordering than the conventional BDD algorithm.

A new method to evaluate alternate AC power source effects in multi-unit nuclear power plants

Abstract In order to evaluate accurately a station blackout (SBO) event frequency of a multi-unit nuclear power plant that has a shared alternate AC (AAC) power source, an approach has been developed which accommodates the complex inter-unit behavior of the shared AAC power source under multi-unit loss of offsite power conditions. The SBO frequency at a target unit of probabilistic safety assessment could be underestimated if the inter-unit dependency of the shared AAC power source is not properly modeled. The approach is illustrated for two cases, 2 units and 4 units at a single site, and generalized for a multi-unit site. Furthermore, the SBO frequency of the first unit of the 2-unit site is quantified. The methodology suggested in the present paper is believed to be very useful in evaluating the SBO frequency and the core damage frequency resulting from the SBO event. This approach is also applicable to the probabilistic evaluation of the other shared systems in a multi-unit nuclear power plant.

Development of measures to estimate truncation error in fault tree analysis

Abstract The fault tree quantification uncertainty from the truncation error has been of great concern for the reliability evaluation of large fault trees in the probabilistic safety analysis (PSA) of nuclear plants. The truncation limit is used to truncate cut sets of the gates when quantifying the fault trees. This paper presents measures to estimate the probability of the truncated cut sets, that is, the amount of truncation error. The functions to calculate the measures are programmed into the new fault tree quantifier FTREX (Fault Tree Reliability Evaluation eXpert) and a Benchmark test was performed to demonstrate the efficiency of the measures. The measures presented in this study are calculated by a single quantification of the fault tree with the assigned truncation limit. As demonstrated in the Benchmark test, lower bound of truncated probability (LBTP) and approximate truncation probability (ATP) are efficient estimators of the truncated probability. The truncation limit could be determined or validated by suppressing the measures to be less than the assigned upper limit. The truncation limit should be lowered until the truncation error is less than the assigned upper limit. Thus, the measures could be used as an acceptability of the fault tree quantification results. Furthermore, the developed measures are easily implemented into the existing fault tree solvers by adding a few subroutines to the source code.

Analysis of operators' performance under emergencies using a training simulator of the nuclear power plant

Abstract It is well known that there are many factors that affect the reliability of nuclear power plants (NPPs). Among them, human reliability has been considered one of the most important factors. Thus, not only in order to quantify human reliability but also to identify main causes that can degrade human reliability, various kinds of human reliability analysis (HRA) methods have been suggested and utilized in many countries. However, to perform HRA more appropriately, it is necessary to collect plant-specific or domain-specific human performance data: especially for emergencies: because they can be used to generate requisite information for HRA. From this point of view, simulator studies under emergencies may be considered important sources for obtaining human performance data. In this study, the performance data of operating crews in coping with emergencies of the reference NPP have been collected and analyzed to develop human performance database (HPDB). Since the number of collected records is 112, it can be said that extracted/analyzed results included in HPDB are statistically meaningful. Therefore, HPDB can be used not only for HRA input data but also for multiple purposes such as improving emergency operating procedures and developing advanced HRA methods.

POSSIBILITIES AND LIMITATIONS OF APPLYING SOFTWARE RELIABILITY GROWTH MODELS TO SAFETY- CRITICAL SOFTWARE

It is generally known that software reliability growth models such as the Jelinski-Moranda model and the Goel-Okumotos non-homogeneous Poisson process (NHPP) model cannot be applied to safety-critical software due to a lack of software failure data. In this paper, by applying two of the most widely known software reliability growth models to sample software failure data, we demonstrate the possibility of using the software reliability growth models to prove the high reliability of safety-critical software. The high sensitivity of a piece of softwares reliability to software failure data, as well as a lack of sufficient software failure data, is also identified as a possible limitation when applying the software reliability growth models to safety-critical software.

OPPORTUNITIES AND CHALLENGES OF NEUTRON SCIENCE AND TECHNOLOGY IN KOREA

Neutron science and technology, the utilization of neutron beams for a wide variety of scientific and engineering research ranging from materials and life science to industrial applications, has been one of the key elements of modern science and technology. Currently, the neutron science and technology in Korea is in rapid growth with the operation of the 30 MW High-flux Advanced Neutron Application Reactor (HANARO) at the Korea Atomic Energy Research Institute, which is one of the most powerful nuclear research reactors in the world. Furthermore, a state of the art HANARO cold neutron research facility, which will open a new era for the neutron science and technology in Korea, is expected to become available in 2010. In this paper, the progress of neutron science and technology in Korea is reviewed and its unprecedented new opportunities and challenges in coming years are presented.

SAFETY ASSESSMENT OF KOREAN NUCLEAR FACILITIES: CURRENT STATUS AND FUTURE

This paper introduces the development of safety assessment technology in Korea, focusing on the activities of the Korea Atomic Energy Research Institute in the areas of system thermal hydraulics, severe accidents and probabilistic safety assessment. In the 1970s and 1980s, safety analysis codes and methodologies were introduced from the United States, France, Canada and other developed countries along with technology related to the construction and operation of nuclear power plants. The main focus was on understanding and utilizing computer codes that were sourced from abroad up to the early 1990s, when efforts to develop domestic safety analysis codes and methodologies became active. Remarkable achievements have been made over the last 15 years in the development and application of safety analysis technologies. In addition, significant experimental work has been performed to verify the safety characteristics of reactors and fuels as well as to support the development and validation of analysis methods.

Development of an Efficient BDD Algorithm to Solve Large Fault Trees

This paper presents an efficient BDD (Binary Decision Diagram) algorithm for large fault trees by which subsuming and truncation could be performed in the process of the construction of the BDD structure. That results in a fast computation of the minimal cut sets (MCSs) and a smaller memory usage.

Development of severe accident management advisory and training simulator (SAMAT)

Abstract The most operator support systems including the training simulator have been developed to assist the operator and they cover from normal operation to emergency operation. For the severe accident, the overall architecture for severe accident management is being developed in some developed countries according to the development of severe accident management guidelines which are the skeleton of severe accident management architecture. In Korea, the severe accident management guideline for KSNP was recently developed and it is expected to be a central axis of logical flow for severe accident management. There are a lot of uncertainties in the severe accident phenomena and scenarios and one of the major issues for developing a operator support system for a severe accident is the reduction of these uncertainties. In this paper, the severe accident management advisory system with training simulator, SAMAT, is developed as all available information for a severe accident are re-organized and provided to the management staff in order to reduce the uncertainties. The developed system includes the graphical display for plant and equipment status, the previous research results by knowledge-base technique, and the expected plant behavior using the severe accident training simulator. The plant model used in this paper is oriented to severe accident phenomena and thus can simulate the plant behavior for a severe accident. Therefore, the developed system may make a central role of the information source for decision-making for a severe accident management, and will be used as the training simulator for severe accident management.

An Approach to Estimate SBO Risks in Multi-unit Nuclear Power Plants with a Shared Alternate AC Power Source

An appropriate method to evaluate accurately the amounts of risks, core damage frequencies and site risks, resulting from a station blackout (SBO) event of a multi-unit site that has a shared alternate AC (AAC) power source has been developed. It accommodates the complex inter-unit behaviour of the shared AAC power source under a multi-unit loss of offsite power (LOOP) conditions.