• Nuclear Engineering and Technology
• /
• v.56 no.6
• /
• pp.2375-2387
• /
• 2024

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

• Journal of the Society of Disaster Information
• /
• v.4 no.2
• /
• pp.124-136
• /
• 2008

The systematic information management of chemical accidents has been required as a tool for the policy making, system improvement and release of information concerning accident prevention. However, there is not yet a systematic chemical accidents tracking system in Korea, which make confusion among the related government agencies and the parties to accidents that the related statistics are different from each others. In this study, We developed the Chemical Accident Tracking System (CATS) using chemical accident classification which was made up of 12 upper classes, 70 middle classes, 272 lower classes. The CATS is mainly consist list up module, reporting module, searching and statistic module, etc. The CATS is expected to be applied to the information tracking and database system for chemical accidents and improve its manageability.

• Journal of the Korean Society of Safety
• /
• v.10 no.2
• /
• pp.97-104
• /
• 1995

To evaluate readily the effect of unconfined vapor cloud explosion(UVCE) having high possibility of accident and risk in chemical industries, the expert system of UVCE was developed and its applicability on a real accident was analyzed. We found that the hazard of UVCE could be well evaluated from the TNT equivalency model and the empirical loss data produced by overpressure for chemical facilities. By using the developed expert system, the size of vapor cloud, the quantity of vaporization, the released energy, the overpressure range from explosion point, and the impact damage of each installation could be estimated respectively. Also, probable maximum loss and catastrophic loss potential for real accident( cyclohexane release in Flixborough Nypro company) were estimated and compared with damages of the accident. As a result, the developed expert system could be well applicable to real accident.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.21 no.2
• /
• pp.247-253
• /
• 2023

According to NSSC Notice No. 2021-10, safety analysis needs to be introduced in the decommissioning plan. Public and occupational dose analyses should be conducted, specifically for unexpected radiological accidents. Herein, based on the risk matrix and analytic hierarchy process, the method of selecting accident scenarios during the decommissioning of nuclear power plants has been proposed. During decommissioning, the generated spent resin exhibits relatively higher activity than other generated wastes. When accidents occur, the release fraction varies depending on the conditioning method of radioactive waste and type of radioactive nuclides or accidents. Occupational dose analyses for 2 (fire and drop) among 11 accident scenarios have been performed. The radiation doses of the additional exposures caused by the fire and drop accidents are 1.67 and 4.77 mSv, respectively.

• Nuclear Engineering and Technology
• /
• v.52 no.8
• /
• pp.1638-1649
• /
• 2020

In the case of a severe accident in a Light Water Reactor, the issue of late release of fission products, from the primary circuit surfaces is of particular concern due to the direct impact on the source term. CsI is the main iodine compound present in the primary circuit and can be deposited as particles or condensed species. Its chemistry can be affected by the presence of molybdenum, and can lead to the formation of gaseous iodine. The present work studied chemical reactions on the surfaces involving gaseous iodine release. CsI and MoO₃ were used to highlight the effects of carrier gas composition and oxygen partial pressure on the reactions. The results revealed a noticeable effect of the presence of molybdenum on the formation of gaseous iodine, mainly identified as molecular iodine. In addition, the oxygen partial pressure prevailing in the studied conditions was an influential parameter in the reaction.

• Korean Chemical Engineering Research
• /
• v.54 no.6
• /
• pp.781-785
• /
• 2016

Chloric acid is a toxic chemical and the risk of facility handling chloric acid was assessed from the list of accident scenario to provide countermeasure to keep the vicinity safe. Accident scenarios were listed by using MSDS and process safety information. The scenarios having effect to the off-site were selected and assessed further according to guideline provided by Korea government. Worst case and alternative scenarios including other interested scenarios were evaluated using ALOHA. Each evaluated scenario was assessed further considering countermeasures. The results showed that the facility handling chloric acid is safe enough and needed no further protections at the moment.

• Nuclear Engineering and Technology
• /
• v.34 no.2
• /
• pp.117-131
• /
• 2002

Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of the work to demonstrate the inherent and ultimate safety of conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 MWe pool- type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method and associated computer program, SCHAMBETA, was developed using a modified Bethe-Tait method to simulate the kinetics and thermodynamic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of the energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the SCHAMBETA code for various reactivity insertion rates up to 100 S/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies were also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters.

• Nuclear Engineering and Technology
• /
• v.49 no.2
• /
• pp.426-433
• /
• 2017

This paper describes the development process, the innovative techniques used and insights gained from the latest integrated, full scope, multistate Level 2 PSA analysis conducted at the Leibstadt Nuclear Power Plant (KKL), Switzerland. KKL is a modern single-unit General Electric Boiling Water Reactor (BWR/6) with Mark III Containment, and a power output of $3600MW_{th}/1200MW_e$, the highest among the five operating reactors in Switzerland. A Level 2 Probabilistic Safety Assessment (PSA) analyses accident phenomena in nuclear power plants, identifies ways in which radioactive releases from plants can occur and estimates release pathways, magnitude and frequency. This paper attempts to give an overview of the advanced modeling techniques that have been developed and implemented for the recent KKL Level 2 PSA update, with the aim of systematizing the analysis and modeling processes, as well as complying with the relatively prescriptive Swiss requirements for PSA. The analysis provides significant insights into the absolute and relative importances of risk contributors and accident prevention and mitigation measures. Thanks to several newly developed techniques and an integrated approach, the KKL Level 2 PSA report exhibits a high degree of reviewability and maintainability, and transparently highlights the most important risk contributors to Large Early Release Frequency (LERF) with respect to initiating events, components, operator actions or seismic component failure probabilities (fragilities).

• Journal of Radiation Protection and Research
• /
• v.28 no.4
• /
• pp.291-298
• /
• 2003

The KN-12 transport cask has been designed to transport 12 PWR spent nuclear fuel assemblies and to comply with the regulatory requirements for a Type B(U) package. The containment boundary of the cask is defined by a cask body, a cask lid, lid bolts with nuts, O-ring seals and a bolted closure lid. The containment vessel for the cask consists of a forged thick-walled carbon steel cylindrical body with an integrally-welded carbon steel bottom and is closed by a lid made of stainless steel, which is fastened to the cask body by lid bolts with nuts and sealed by double elastomer O-rings. In the cask lid an opening is closed by a plug with an O-ring seal and covered by the bolted closure lid sealed with an O-ring. The cask must maintain a radioactivity release rate of not more than the regulatory limit for normal transport conditions and for hypothetical accident conditions, as required by the related regulations. The containment requirements of the cask are satisfied by maintaining a maximum air reference leak rate of $2.7{\times}10^{-4}ref.cm^3s^{-1}$ or a helium leak rate of $3.3{\times}10^{-4}cm^3s^{-1}$ for normal transport conditions and for hypothetical accident conditions.

• Journal of the Korean Institute of Gas
• /
• v.7 no.1 s.18
• /
• pp.10-18
• /
• 2003

Consequence model has been suggested to evaluate consequence of city gas accident considering actual situation. Through event tree analysis(ETA), probable accidents were summarized and listed. Then release rate was calculated both sonic and subsonic conditions. Among city gas accidents, fire damage is the dominant one and mainly discussed; fatality, burning injury, and damage to building were estimated using the consequence model suggested. With an appropriate conditions, calculating total cost by accident was suggested.