• Journal of Korean Society of Industrial and Systems Engineering
• /
• v.23 no.55
• /
• pp.25-31
• /
• 2000

To make a better decision about when to shutdown a nuclear power plant, we build a decision model using influence diagrams. We proceed the analysis adopting a bayesian approach. Firstly, an accident arrival rate is assumed to be known and this assumption is relaxed later. We perform our analysis on the cases of exponential time to accidents, and gamma distribution for the arrival rate. An optimal shutdown time is obtained considering the trade-off between the costs incurred by an accident due to late shutdown and the possible loss of revenues due to the early shutdown. We also derive the upper bound of the failure rate where we may operate the plant.

• Journal of the Korean Society of Safety
• /
• v.19 no.4 s.68
• /
• pp.14-19
• /
• 2004

Boiler is a hazardous equipment to have potential explosion ail the time. And not only it has malfunction at explosion. it lead to people death but also secondary accident such as explosion and fire. Therefore, this equipment should not be broken for keeping its own function. And also, high level of safety should be kept in the process of the use not to be malfunctioned. A large scale of accident due to boiler explosion can be preventive in advance. Boiler fracture is occurred by instant expansion (approximately 1700 time) from quick evaporation of rater in boiler, due to pressure decrease in boiler Emitting energy from it is tremendous and it is so dangerous because of its high temperature. Secondary explosion such as fire is also a main hazard occurring at fuel supply place. If any devices with high pressure is broken, then not only boiler vessel but also components of it are spread with high speed, causing secondary accident. This study is to analyze integrally accident cause of fire and flue tube boiler to have occurred pressure fracture actually, to show countermeasures to prevent accident loss from the fire and flue tube boiler.

• Korean Security Journal
• /
• no.5
• /
• pp.327-350
• /
• 2002

We, in Korea, have over two hundred hypermarkets and the number is continuously increasing. We drop into a hypermarket for merchandise, which is an integral part of our life style. So, we should consider safety for employees as well as customers because hundreds of thousands of people use the hypermarket every day. In addition to this consideration, the government should also be a political support relating to accidents that occur in the hypermarket because security and safety matters are important to all of us. But even now, Our security conditions do not match our ideal goal and we take countermeasures after accident or loss. This is a result of not having a security management expert coupled with a chief executive officer that has no idea about security awareness and loss prevention. In addition, we do not have specific laws to address these matters. We also lack reasonable ideas to prevent accident and loss. Now is the right time to revamp the laws and ordinances to improve the quality of civilian life. Prevention of accidents is a needed investment for all security personnel. The best solution for businesses is prevention of accidents. This will increase profits and cost-effectiveness as well as increase customer satisfaction. The company should form a security management department for comprehensive protection of assets. The goal of security management employees should be productive and effective security management. Every employee should have responsibility in mind to prevent accidents in his or her work. In addition, The company should have a systematic organization in place and regular training sessions. The most effective security management comes from cooperation of all members. In the 21st century, we pursue a high standard of living which is a result of our cooperation against any accident and loss. Sennewald says The value of security is better measure by what does not happen rather than what does.

• Journal of the Korean Society of Safety
• /
• v.27 no.5
• /
• pp.229-234
• /
• 2012

Operator's action time is evaluated from MAAP4 analysis used in conventional probabilistic safety assessment(PSA) of a nuclear power plant. MAAP4 code which was developed for severe accident analysis is too conservative to perform a realistic PSA. A best-estimate code such as RELAP5/MOD3, MARS has been used to reduce the conservatism of thermal hydraulic analysis. In this study, operator's action time of core cooling recovery operation is evaluated by using the MARS code, which its Fussell-Vessely(F-V) value was evaluated as highly important in a small break loss of coolant(SBLOCA) event and loss of component cooling water(LOCCW) event in previous PSA. The main conclusions were elicited : (1) MARS analysis provides larger time window for operator's action time than MAAP4 analysis and gives the more realistic time window in PSA (2) Sufficient operator's action time can reduce human error probability and core damage frequency in PSA.

• Journal of the Korean Society for Heat Treatment
• /
• v.18 no.2
• /
• pp.112-118
• /
• 2005

Study was focused on the effect of the cooling rate on the embrittlement behavior of Zircaloy-4 cladding simulated Loss Of Coolant Accident (LOCA) environment. Claddings were oxidized at given temperature and given time followed by various water quenching in the range of $0.6^{\circ}C$ and $100^{\circ}C$ per second. Cladding failed after water quenching above the threshold oxidation. Threshold oxidation was decreased as the cooling rate increased, which is due to the matensite structure formed during fast cooling rate.

• Nuclear Engineering and Technology
• /
• v.46 no.6
• /
• pp.803-816
• /
• 2014

Defense-in-depth is a fundamental safety principle for the design and operation of nuclear power plants. Despite its general appeal, defense-in-depth is not without its drawbacks, which include its potential for concealing the occurrence of hazardous states in a system, and more generally rendering the latter more opaque for its operators and managers, thus resulting in safety blind spots. This in turn translates into a shrinking of the time window available for operators to identify an unfolding hazardous condition or situation and intervene to abate it. To prevent this drawback from materializing, we propose in this work a novel safety principle termed "observability-in-depth". We characterize it as the set of provisions technical, operational, and organizational designed to enable the monitoring and identification of emerging hazardous conditions and accident pathogens in real-time and over different time-scales. Observability-in-depth also requires the monitoring of conditions of all safety barriers that implement defense-in-depth; and in so doing it supports sensemaking of identified hazardous conditions, and the understanding of potential accident sequences that might follow (how they can propagate). Observability-in-depth is thus an information-centric principle, and its importance in accident prevention is in the value of the information it provides and actions or safety interventions it spurs. We examine several "event reports" from the U.S. Nuclear Regulatory Commission database, which illustrate specific instances of violation of the observability-in-depth safety principle and the consequences that followed (e.g., unmonitored releases and loss of containments). We also revisit the Three Mile Island accident in light of the proposed principle, and identify causes and consequences of the lack of observability-in-depth related to this accident sequence. We illustrate both the benefits of adopting the observability-in-depth safety principle and the adverse consequences when this principle is violated or not implemented. This work constitutes a first step in the development of the observability-in-depth safety principle, and we hope this effort invites other researchers and safety professionals to further explore and develop this principle and its implementation.

• Nuclear Engineering and Technology
• /
• v.56 no.3
• /
• pp.873-879
• /
• 2024

A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.

• Journal of Korean Academy of Fundamentals of Nursing
• /
• v.10 no.3
• /
• pp.326-334
• /
• 2003

Purpose: The purpose of this study was to identify characteristics associated with fractures in elderly people in order to provide basic data for fracture preventive programs for the elderly people. Method: The participants were 84 patients over age of 65, who were admitted to the orthopedic department in a hospital in Chungnam province. Data were collected from Sep. 1, 2002 to Aug, 30, 2003 through personal interviews using a structured questionnaire. The data were analyzed using SPSSPC program. Result: Physical characteristics before the fracture included weakness or paralysis in the extremities (29.8%), need of assistance or appliances (13.1%), difficulty on balance (28.6%), visual disturbances (26.2%), hearing impairment (17.9%), speech disturbances (2.4%), urinary dysfunction (21.4%), and sleep disturbances (54.8%). The fractures occurred most frequently in winter (32.1%), between 1 pm and 6 pm (48.8%), on weekends (41.6%), in the road (58.3%) while wearing snickers (27.4%) or shoes (27.4%). The region of fractures occurred most frequently was lower extremities (57.1%), and the causes of fractures were loss of balance (31.0%) and car accident (25.0 %). A significance difference was found for type of accident, footwear at the time of the accident, place of the accident according to gender and age. Also a significance difference was found for type of accident and place of accident according to season(p<.05). Conclusion: Therefore, these results should be considered when a fracture preventive program for elderly people is designed.

• IE interfaces
• /
• v.13 no.2
• /
• pp.211-216
• /
• 2000

A methodology that determines the optimal shutdown time of a nuclear power plant is suggested. The shutdown time is decided considering the trade off between the cost of accident and the loss of profit due to the early shutdown. We adopt the bayesian approach in manipulating the model parameter that predicts the accidents. We build decision tree models and apply dynamic programming approach to decide whether to shutdown immediately or operate one more period. The branch parameters in decision trees are updated by bayesian approach. We apply real data to this model and provide the cost of accidents that guarantees the immediate shutdown.

• Nuclear Engineering and Technology
• /
• v.45 no.5
• /
• pp.589-596
• /
• 2013

In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ~2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.