• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2001.11a
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• pp.325-331
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• 2001

• International Journal of Precision Engineering and Manufacturing
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• v.8 no.4
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• pp.56-62
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• 2007

An analytical method to predict the post-weld deformation at the heat-affected zone (HAZ) is presented in this paper. The method was based on the assumption that the post-weld deformation is caused by external forces resulting from the inherent strain, which is defined as the irrecoverable strain after removing structural restraints and loadings. In general, the equivalent loading method can be used to analyze distortions in welding areas because it is efficient and effective. However, if additional loads are applied after welding, it is difficult to determine the final strain on a welded structure. To determine the final strain of a welded structure at the HAZ more accurately, we developed a modified equivalent loading method based on the inherent strain that incorporated hardening effects. The proposed method was applied to calculate the residual stress at the HAZ. Experiments were also conducted on welded plates to evaluate the validity of the proposed method.

• Journal of Ocean Engineering and Technology
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• v.32 no.2
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• pp.95-105
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• 2018

An FLNG (floating liquefied natural gas) or LNG FPSO (floating production, storage and offloading) unit is a notable offshore unit with the increasing demand for LNG. The liquefaction process on an FLNG unit is the most important process because it determines the economic feasibility, but would be a hazard source because of the large quantity of hydrocarbons. While a high efficiency process such as C3MR has been preferred for onshore liquefaction processes, a relatively simple process such as the SMR (single mixed refrigerant) or DMR (dual mixed refrigerant) liquefaction process has been selected for offshore units because they require a more compact size, lighter weight, and higher safety due to their space limitation for facilities and long distance from shore. It is known that an SMR has the advantages of a simple configuration, small footprint, and lower risk. However, with an increased production rate, the inherent safety of SMR needs to be evaluated because of its small train capacity. In this study, the potential explosion risks of the SMR and DMR liquefaction processes were evaluated at the conceptual design stage. The results showed that an SMR has a lower overpressure than a DMR at the same frequency, only with a small production capacity of 0.9 MTPA. With increased capacity, the overpressure of the SMR was higher than that of the DMR. The increased number of trains increased the frequency in spite of the small amount of equipment per train. This showed that the inherent risk of an SMR is not always lower than that of a DMR, and an additional risk management strategy is recommended when an SMR is selected as the concept for an FLNG liquefaction process compared to the DMR liquefaction process.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.1-12
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• 2019

One of proposed Accident Tolerant Fuel (ATF) concept is fully ceramic micro-encapsulated fuel (FCMF). FCMF using uranium mononitride (UN) has better safety aspects than $UO_2$ pellet fuel although it might not have a better neutronic performance due to the presence of matrix and high neutron-induced interaction of $^{14}N$. Before implementing UN-FCMF technology in Small-PWR, further research must be taken place to make sure the proposed design of fuel assembly has inherent safety features and maintain the fuel performance. This study focusses on the neutronic analysis of UN-FCMF based fuel assembly using Serpent and SCALE codes. It is shown in the proposed fuel assembly design has inherent safety features with respect to the fuel temperature reactivity coefficient, void reactivity coefficient, and moderator temperature reactivity coefficient. It is noted that the use of FCMF leads to a lower ratio of burnup to $^{235}U$ enrichment ratio compared to the $UO_2/Zr$ fuel.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.1022-1028
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• 2024

A neutronics study of a supercritical CO₂-cooled fast reactor core for nuclear propulsion has been performed in this work. The thermal power of the reactor core is 30 MWth and a ceramic UO₂ fuel can be used to achieve a 20-year lifetime without refueling. In order to make a compact core with inherent safety features, the drum-type reactivity control system and folding-type shutdown system are adopted. In addition, we suggest a cold shutdown system using gadolinium as a spectral shift absorber (SSA) against flooding. Although there is a penalty of U-235 enrichment for the core embedded with the cold shutdown system, it effectively mitigates the increment of reactivity at the flooding of seawater. In this study, the neutronics analyses have been performed by using the continuous energy Monte Carlo Serpent 2 code with the evaluated nuclear data file ENDF/B-VII.1 Library. The supercritical CO₂-cooled fast reactor core is characterized in view of important safety parameters such as the reactivity worth of reactivity control systems, fuel temperature coefficient (FTC), coolant temperature coefficient (CTC), and coolant temperature-density coefficient (CTDC). We can say that the suggested core has inherent safety features and enough flexibility for load-following operation.

• The Journal of Engineering Geology
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• v.24 no.4
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• pp.609-613
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• 2014

The Korean Peninsula, located at the southeastern tip of the Eurasian Plate, is known to be tectonically stable, and no critical evidence has yet been found that would override the safety design of nuclear facilities in South Korea. Because a nuclear power plant, like other major social overhead capital facilities, could cause great damage to both the environment and society through an unexpected tectonic event, even one of extremely low probability, like the Fukushima accident, a defense-in-depth safety approach is required in geological and geotechnical site safety evaluation for nuclear projects. This paper introduces the regulatory procedures that are in place to confirm nuclear site safety and site monitoring (e.g., earthquakes and groundwater) systems applied to nuclear facilities in order to reduce inherent uncertainties within the site safety review of geological and seismological issues related with a NPP project.

• Proceedings of the Safety Management and Science Conference
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• 2001.05a
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• pp.77-81
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• 2001

Burn-in is a test procedure to find and eliminate the inherent initial failure of a product during or at the final stage of production process. Software testing is the validation and verification process which is used to cut off the faults from a software. The two have the common function and objective of "debugging". This article summarizes some significant models on the optimal hardware and software burn-in time, and provides the relevant paper lists. The need for the development of the unified burn-in policy of a hardware-software system is addressed.addressed.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.268-273
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• 2016

An innovative safety concept for a light water reactor has been developed at the Korea Atomic Energy Research Institute. It is a unique concept that adopts both a fast heat transfer mechanism for a small containment and a changing mechanism of the cooling geometry to take advantage of the potential, thermal, and dynamic energies of the cold water in the containment. It can bring about rapid cooling of the containment and long-term cooling of the decay heat. By virtue of this innovative concept, nuclear fuel damage events can be prevented. The ultimate heat transfer mechanism contributes to minimization of the heat exchanger size and containment volume. A small containment can ensure the underground construction, which can use river or seawater as an ultimate heat sink. The changing mechanism of the cooling geometry simplifies several safety systems and unifies diverse functions. Simplicity of the present safety system does not require any operator actions during events or accidents. Therefore, the unique safety concept of PX can realize both economic competitiveness and inherent safety.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.162-162
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• 1998

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.06a
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• pp.73-73
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• 2009

There are many factors to evaluate nuclear fuel cycle such as safety, public acceptance, economics, etc.. Transparency, proliferation, environment issues, public acceptance and safety are essential to expansion of nuclear industry and proliferation resistance is one of key constraints in the deployment of advanced nuclear energy systems. Proliferation resistance is being considered as one of the most important factors in assessing advanced and innovative nuclear systems. IAEA defmes proliferation resistance as characteristics of nuclear energy system that impedes the diversion or undeclared production of nuclear material [1]. Barriers to proliferation is consist of intrinsic and extrinsic barriers(institutional measures). Intrinsic barriers are characterized in material barriers and technical barriers in general. Material barriers is intrinsic, or inherent, qualities of materials that reduce the inherent desirability or attractiveness of the material as an explosive. Isotopic, chemical, radiological, mass and bulk, detectability barriers are considered as material barriers attributes [2]. Proliferation resistance is examined for several nuclear fuel cycles based on previous study which is focused on the intrinsic barriers [3-4]. Pyroprocessing and DUPIC are considered as reprocessing technologies in Korea and the PWR direct disposal is considered. Comparative assessments of the proliferation attributes and merits of different fuel cycle systems will be performed and the optimal back-end fuel cycle and strategy will be proposed.