• Nuclear Engineering and Technology
• /
• v.37 no.4
• /
• pp.327-362
• /
• 2005

The design, construction, and operation of a light water reactor (LWR) are subject to compliance with safety criteria specified for accident situations, such as loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA). Because reactor fuel is the primary source of radioactivity and heat generation, such a criterion is established on the basis of the characteristics and performance of fuel under the specific accident condition. As such, fuel behavior under accident situations impact many aspects of fuel design and power generation, and in an indirect manner, even spent fuel storage and management. This paper provides a comprehensive review of: the history of the current LOCA criteria, results of LOCA-related investigations on conventional and new classes of fuel, and status of on-going studies on high-burnup fuel under LOCA situations. The objective of the paper is to provide a better understanding of important issues and an insight helpful to establish new LOCA criteria for modem LWR fuels.

• Nuclear Engineering and Technology
• /
• v.34 no.1
• /
• pp.68-79
• /
• 2002

The approach for evaluating the critical heat flux (CHF) margin using the departure from nucleate boiling ratio (DNBR) concept has been widely applied to PWR core design, while DNBR in this approach does not indicate appropriately the CHF margin in terms of the attainable power margin-to-CHF against a reactor core condition. The CHF power margin must be calculated by increasing power until the minimum DNBR reaches a DNBR limit. The Critical Power Ratio (CPR), defined as the ratio of the predicted CHF power to the operating power, is considered more reasonable for indicating the CHF margin and can be calculated by a CPR orrelation based on the heat balance of a test bundle. This approach yields directly the CHF power margin, but the calculated CPR must be corrected to compensate for many local effects of the actual core, which are not considered in the CHF test and analysis. In this paper, correction of the calculated CPR is made so that it may become equal to the DNB overpower margin. Exemplary calculations showed that the correction tends to be increased as power distribution is more distorted, but are not unduly large.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.14 no.2
• /
• pp.16-23
• /
• 2018

In order to investigate the ways to enhance the seismic performance of APR1400 seismic fragile equipment by direct design changes, four equipment such as Reactor Vessel Support, Integrated Head Assembly, Remote Shutdown Console, and Pressurizer are reviewed using information of the main dimensions, seismic stress evaluation results, design FRS, etc. in this paper. In addition to the direct reinforcement of equipments, the feasibility of seismic isolation for the safety related cabinet is also investigated and the actual adaption plan of a commercial spring-damper system is briefly reviewed.

• Nuclear Engineering and Technology
• /
• v.53 no.1
• /
• pp.322-336
• /
• 2021

This paper presents a numerical prediction of the transient hydraulic loads acting on the tubes and external supports of a pressurized water reactor (PWR) steam generator (SG) during blowdown following a sudden feedwater line break (FWLB). A simplified SG model was used to easily demonstrate the prediction. The blowdown discharge flow was treated as a flashing flow to realistically simulate the transient flow fields inside the SG and the connected broken feedwater pipe. The effects of the SG initial pressure or the broken feedwater pipe length on the intensities or magnitudes of transient hydraulic loads were investigated. Then predictions of the decompression pressure wave-induced impulsive pressure differential loads on SG tubes and the transient blowdown loads on SG external supports were demonstrated and the general aspects of transient responses of such transient hydraulic loads to the FWLB were discussed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.21 no.1
• /
• pp.115-147
• /
• 2023

The Fukushima accident in 2011 revealed some major flaws in traditional nuclear fuel materials under accidental conditions. Thus, the focus of research has shifted toward "accident tolerant fuel" (ATF). The aim of this approach is to develop fuel material solutions that lead to improved reactor safety. The application of protective coatings on the surface of nuclear fuel cladding has been proposed as a near-term solution within the ATF framework. Many coating materials are being developed and evaluated. In this article, an overview of different zirconium-based alloys currently in use in the nuclear industry is provided, and their performances in normal and accidental conditions are discussed. Coating materials proposed by different institutions and organizations, their performances under different conditions simulating nuclear reactor environments are reviewed. The strengths and weaknesses of these coatings are highlighted, and the challenges addressed by different studies are summarized, providing a basis for future research. Finally, technologies and methods used to synthesize thin-film coatings are outlined.

• Proceedings of the Korean Society of Propulsion Engineers Conference
• /
• 2008.03a
• /
• pp.272-274
• /
• 2008

A development of hydrazine decomposition catalyst for monopropellant thruster has been performed by Korea Aerospace Research Institute(KARI). The goal of this development is to product a catalyst showing the equivalent performance with space-proven catalysts. Catalyst production and physical/chemical analysis were conducted by Chonnam National University and the analysis result was compared with the result of other catalysts and our own specification. Using the developed prototype catalyst, short firing test was performed in a reactor to verify basic performance of catalyst. After the successful reactor test, hot firing tests were carried out in atmospheric and vacuum condition using 5N thruster to verify durability and safety of catalyst. In this paper, the catalyst development status will be presented.

• Journal of the Korea Safety Management & Science
• /
• v.15 no.3
• /
• pp.127-132
• /
• 2013

This study shows an air-purification test by the UV lamp on which TiO2 catalyst is deposited with glass fiber in the reactor chamber. This test was based on the fundamental data of air-purifier as assessing a removing ability on various contaminants such as CH3COOH, NH3, NO and SO2 as variation of the TiO2 coating, the wave of UV lamp, and the additive CaO. As a result, the highest decomposing removal ratio was shown when 5-times coated glass fiber was used. It can be due to the recombination reaction of electrons and electron-hole in the loaded CaO. Thus, the decomposing removal ratio increased as the recombination ratio decreased. In addition, it was confirmed that the decomposing removal ratio lowered when CaO was considerably deposited because it hided the lamp of OH-1 radical.

• IE interfaces
• /
• v.15 no.2
• /
• pp.114-125
• /
• 2002

The objective of this paper is to describe a systematic accident scenario analysis method(SASA) adept at creating accident scenarios for the design of safer products. This approach was inspired by the Quality Function Deployment(QFD) method, which is conventionally used in quality management. In this study, the QFD provides a formal and systematic scheme to devise accident scenarios while maintaining objectivity. SASA consists of three key stages to be broken down into a series of consecutive steps:(1) developing an accident analysis tableau,(2) devising the accident scenarios using the accident analysis tableau,(3) performing a feasibility test, a clustering process and a patterning process, and finally(4) performing quantitative evaluation of each accident scenario. The SASA was applied to a case study of child safety seats. The accident analysis tableau devised 2828(maximum) accident scenarios from all possible relationships between the hazard factors and situation characteristics. Among them, 270 scenarios were devised through the feasibility test and the clustering process. The patterning process reduced them to 29 patterns representative of all accident scenarios. Based on an intensive analysis of the accident patterns, design guidelines for a safer child safety seat were recommended. The implications of the study on the child safety seat case were then discussed.

• Nuclear Engineering and Technology
• /
• v.28 no.5
• /
• pp.440-451
• /
• 1996

With emphasis on safety, this study addresses for better design condition for the cooling system in a wet-type interim spent fuel storage facility, using a probabilistic safety assessment method. To incorporate the design renovation into the design phase, a simple approach is proposed. By taking the cooling system of a reference design, a fault tree analysis was performed to identify the weak point of the considered system, and then basic factors for design renovation were defined. A total of 21 design alternatives were selected through the combination of the basic factors. Finally, the optimum design alternative for the cooling system is derived by means of the cost and effect analysis based on the estimated cost, system reliability and assumed probabilistic safety criteria. With the assumption that the failure frequency of at-reactor spent fuel cooling system compiles with probabilistic safety criteria for the interim spent fuel cooling system, it was shown that the optimum alternative should have l00% cooling loop redundancy with one pump per cooling loop and a cleanup system installed separately from the main loop. Furthermore, it also should be classified into safety system. The result of this study can be used as a useful basis to identify factors of safety concern and to establish design requirements in the future. The method also can be applied for other nuclear facilities.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.29 no.3
• /
• pp.256-263
• /
• 2009

After several PWSCCs were found in Bugey(France), Ringhals(Sweden), Tihange(Belgium), Oconee, Arkansas, Crystal Fever, Davis-Basse, VC Summer(U.S.A.), Thuruga(Japan), USNRC and PNNL started the research on PWSCC, that is, the PINC project. USNRC required KINS to participate in the PINC project in May 2005. KINS organized the Korean consortium at March 2006 and Pre-RRT for RVHP were performed for the preparation of PINC RRT. Through these preliminary RRT, Korea NDE teams can learn and develop the detection and sizing technique for RVHP dissimilar metal weld. These techniques are now being prepared in Korea and need to be utilized for the In-service inspection of the RVHP and BMI of Korea Nuclear Power Plants. PINC RRT mock-ups will be helpful to training.