• Nuclear Engineering and Technology
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• 제40권6호
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• pp.511-516
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• 2008

Korea Hydro & Nuclear Power Co., Ltd. (KHNP) developed new IP-2 packages to transport low- and intermediate-level radioactive waste (LILW) steel drums from nuclear power plants to a disposal facility in accordance with IAEA and Korean transport regulations of radioactive material. Radiation shielding evaluation of the packages was carried out to demonstrate compliance with the regulatory requirements for IP-2 packages of radioactive material. Dose rate limits of LILW drums contained in the packages were determined.

• 한국국방경영분석학회지
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• 제18권1호
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• pp.1-14
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• 1992

As North Korea signed 'the Korea Peninsula Non-Nuclearization Joint Declaration' at the end of last year as well as full-scope safeguards agreement with the IAEA in Jan.30 1992, her nuclear activity was incorporated into the international monitoring system and opportunities were arranged to .obstruct her nuclear weapon development and nuclear material diversion, which have been international issues up to recent years. However, achieving goals of the Joint Declaration and safeguards agreement should presuppose North Korea's sincerity toward the abandonment of nuclear weapon development. In this study, first of all, her nuclear policy, current situation of nuclear development and the capability of nuclear development are analyzed. Also, based on the analysis. attempts have been made to find methods of effective performance of the South-North Korea mutual nuclear inspection and the suggested methods are as follows; 1) Analysis of the limits of IAEA inspection and suggestion of its supplementary strategies 2) Securing and training professional inspectors for the South-North mutual inspection 3) Establishment of the verification technology to detect nuclear material diversion.

• 방사성폐기물학회지
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• 제20권1호
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• pp.23-32
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• 2022

Pyroprocessing is a promising technology for managing spent nuclear fuel. The nuclear material accounting of feed material is a challenging issue in safeguarding pyroprocessing facilities. The input material in pyroprocessing is in a solid-state, unlike the solution state in an input accountability tank used in conventional wet-type reprocessing. To reduce the uncertainty of the input material accounting, a double-stage homogenization process is proposed in considering the process throughput, remote controllability, and remote maintenance of an engineering-scale pyroprocessing facility. This study tests two types of mixing equipment in the proposed double-stage homogenization process using surrogate materials. The expected heterogeneity and accounting uncertainty of Pu are calculated based on the surrogate test results. The heterogeneity of Pu was 0.584% obtained from Pressurized Water Reactor (PWR) spent fuel of 59 WGd/tU when the relative standard deviation of the mass ratio, tested from the surrogate powder, is 1%. The uncertainty of the Pu accounting can be lower than 1% when the uncertainty of the spent fuel mass charged into the first mixers is 2%, and the uncertainty of the first sampling mass is 5%.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.791-798
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• 2012

Since standardized fracture test specimens cannot be easily extracted from in-service components, several alternative fracture toughness test methods have been proposed to characterize the deformation and fracture resistance of materials. One of the more promising alternatives is the local approach employing the SP(Small Punch) testing technique. However, this process has several limitations such as a lack of anisotropic yield potential and tediousness in the damage parameter calibration process. The present paper investigates estimation of ductile fracture resistance(J-R) curve by FE(Finite Element) analyses using an anisotropic damage model and enhanced calibration procedure. In this context, specific tensile tests to quantify plastic strain ratios were carried out and SP test data were obtained from the previous research. Also, damage parameters constituting the Gurson-Tvergaard-Needleman model in conjunction with Hill's 48 yield criterion were calibrated for a typical nuclear reactor material through a genetic algorithm. Finally, the J-R curve of a standard compact tension specimen was predicted by further detailed FE analyses employing the calibrated damage parameters. It showed a lower fracture resistance of the specimen material than that based on the isotropic yield criterion. Therefore, a more realistic J-R curve of a reactor material can be obtained effectively from the proposed methodology by taking into account a reduced load-carrying capacity due to anisotropy.

• 대한방사선기술학회지:방사선기술과학
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• 제42권1호
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• pp.53-59
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• 2019

In the field of nuclear medicine, the various static phantoms of international standards are used to assess the performance of the nuclear medicine equipment. However, we only reproduced a fixed situation in spite of the movement of the cardiac, and the demands for dynamic situations have been continuously raised. More research is necessary to address these challenges. This study used flexible materials to design the dynamic cardiac phantom, taking into account the various clinical situations. It also intended to reproduce the images through dynamic cardiac flow to confirm the usefulness of the proposed technique. The frame of dynamic cardiac phantom was produced based on the international standard phantom. A nuclear medicine dynamic cardiac phantom was produced rubber material and silicone implemented by 3D printing technique to reproduce endocardium and epicardium movement. Therefore we compared and evaluated the image of a cardiac phantom made of rubber material and a cardiac phantom made of silicone material by 3D printing technique. According to the results of this study, the analysis of the Summed Rest Score(SRS) showed abnormalities in the image of a cardiac phantom made of rubber material at 10, 20, and 30 stroke rates, but the image of a cardiac phantom made of silicone material by 3D printing technique showed normal levels. And the analysis of the Total Perfusion Deficit(TPD) showed that TPD in the image of a cardiac phantom made of rubber material was higher than that of the image of a cardiac phantom made of silicone material by 3D printing technique at 10, 20, and 30 stroke rates. The potential for clinical application of the proposed method was confirmed in the dynamic cardiac phantom implemented with 3D printing technique. It is believed that the objective information secures the reliability of inspection equipment and it contributes to improve the diagnostic value of nuclear medicine.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.364.2-364
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• 2002

• 대한기계학회논문집A
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• 제20권2호
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• pp.575-584
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• 1996

The objective of this paper is to develop an expert system for nuclear piping integrity. This paper describes the selection methodology of integrity evalution method and the inference of material properties. To select the integrity evaluation method, the weight factor for respective material properties was obtained by the sensitivity analysis of the effect of material properties on integrity evaluation method. Subsequently the possession ratio for respective integrity evaluation method was computed, and the most appropriate integrity evaluation method for given input information is selected. In the material properties inference, stress-strain curves and J-R curves were predicted from tensile properties such as yield strength and tensile strength.

• Nuclear Engineering and Technology
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• 제46권1호
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• pp.101-108
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• 2014

About 40% of reactors in the world are being operated beyond design life or are approaching the end of their life cycle. During long-term operation, various degradation mechanisms occur. Fatigue caused by alternating operational stresses in terms of temperature or pressure change is an important damage mechanism in continued operation of nuclear power plants. To monitor the fatigue damage of components, Fatigue Monitoring System (FMS) has been installed. Most FMSs have used Green's Function Approach (GFA) to calculate the thermal stresses rapidly. However, if temperature-dependent material properties are used in a detailed FEM, there is a maximum peak stress discrepancy between a conventional GFA and a detailed FEM because constant material properties are used in a conventional method. Therefore, if a conventional method is used in the fatigue evaluation, thermal stresses for various operating cycles may be calculated incorrectly and it may lead to an unreliable estimation. So, in this paper, the modified GFA which can consider temperature-dependent material properties is proposed by using an artificial neural network and weight factor. To verify the proposed method, thermal stresses by the new method are compared with those by FEM. Finally, pros and cons of the new method as well as technical findings from the assessment are discussed.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2298-2304
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• 2023

Titanium alloys are expected to become one of the candidate materials for nuclear-powered spacecraft due to their excellent overall performance. Nevertheless, atomistic mechanisms of the defect accumulation and evolution of the materials due to long-term exposure to irradiation remain scarcely understood by far. Here we investigate the heavy irradiation damage in a-titanium with a dose as high as 4.0 canonical displacements per atom (cDPA) using atomistic simulations of Frenkel pair accumulation. Results show that the content of surviving defects increases sharply before 0.04 cDPA and then decreases slowly to stabilize, exhibiting a strong correlation with the system energy. Under the current simulation conditions, the defect clustering fraction may be not directly dependent on the irradiation dose. Compared to vacancies, interstitials are more likely to form clusters, which may further cause the formation of 1/3<1210> interstitial-type dislocation loops extended along the (1010) plane. This study provides an important insight into the understanding of the irradiation damage behaviors for titanium.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4266-4273
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• 2023

In a severe reactor accident, a crust will form on the surface of the molten material during the core melting process. The crust will have a contact melting with the internal components of the reactor. In this paper, the contact melting process of the molten material on the austenitic stainless steel plate bundles is studied. The contact melting model of parabolic molten material on the plate bundles is proposed, and the rule and main effect factors of the contact melting are analyzed. The results show that the melting velocity is proportional to the slope of the paraboloid, the heat flux and the distance between two plates D. The influence of melt gravity and the plate width on melting velocity is negligible. The thickness of the molten liquid film is proportional to the heat flux and plate width, and it is inversely proportional to the gravity. With the increase of D, the liquid film thickness decreases at first and then increases gradually. The liquid film thickness has a minimum against D. When the width of the plate is small, the width of the plate is the main factor affecting the thickness of the liquid film. The parameters are coupled with each other. In a severe reactor accident, the wider internal components of reactor, which can increase the thickness of the melting liquid film and reduce the net input heat flux from the molten material to the components, are the effective measures to delay the melting process.