• 한국압력기기공학회 논문집
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• 제16권1호
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• pp.110-116
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• 2020

As the management plan for domestic spent nuclear fuel is delayed, the storage of the operating nuclear power plant is approaching saturation, and the Kori 1 Unit that has reached its end of operation life is preparing for the dismantling plan. The first stage of dismantling is the transfer of spent nuclear fuel stored in storage at plants. The spent fuel management process leads to temporary storage, interim storage, reprocessing and permanent disposal. In this paper, the technical issues to be considered when transporting spent fuel in this process are summarized. The spent fuels are treated as high-level radioactive waste and strictly managed according to international regulations. A series of integrity tests are performed to demonstrate that spent fuel can be safely stored for decades in a dry environment before being transferred to an intermediate storage facility. The safety of spent fuel transport container must be demonstrated under normal transport conditions and virtual accident conditions. IAEA international standards are commonly applied to the design of transport containers, licensing regulations and transport regulations worldwide. In addition, each country operates a physical protection system to reduce and respond to the threat of radioactive terrorism.

• 한국압력기기공학회 논문집
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• 제11권1호
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• pp.20-24
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• 2015

Steam generator(S/G) tubes in pressurized water reactor (PWR's) are subject to several types of degradation. This degradation includes denting, pitting, intergranular attack(IGA), intergranular stress corrosion cracking(IGSCC), fatigue, fretting and wear. Degradation can be derived from either the primary side(inside) or the secondary side(outside) of the tube. Recent issue for tube degradation in domestic steam generator is the tube end cracking on seal weld region. The seal weld region at the tube end and tube itself is regarded as a pressure boundary between the primary side and the secondary side. One of the Westinghouse Model-F S/G has experienced tube end cracking and its number of plugging approximately becomes to the operating limit up to 5% due to tube end cracking which was reported as SAI/MAI(single/multiple axial indication) or SCI/MCI(Single/multiple circumferential indication) from the results of eddy current testing. Eddy current mock-up test was carried out to determine the origin of cracking whether it is from weld zone area or parent tube. This result was helpful to analyze crack location on ECT data. Correct action on this problem was the installation of tube-end sleeve. Last year, after removing 340 installed plugs from tubes, selected 269 tubes took tube-end sleeve installation. Tube-end sleeve brought pressure boundary from parent tube to installed sleeve tube. Tube-end sleeve has the benefit of reducing outage period and increasing more revenue than replacing S/G. This paper is provided to assist interest parties in effectively understanding this issue.

• 한국압력기기공학회 논문집
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• 제13권2호
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• pp.75-83
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• 2017

In severe accident conditions of light water reactors, the loss of coolant may cause problems in integrity of zirconium fuel cladding. Under the condition of the loss of coolant, the zirconium fuel cladding can be exposed to high temperature steam and reacted with them by producing of hydrogen, which is caused by the failure in oxidation resistance of zirconium cladding materials during the loss of coolant accident scenarios. In order to avoid these problems, we develop a multi-metallic layered composite (MMLC) fuel cladding which compromises between the neutronic advantages of zirconium-based alloys and the accident-tolerance of non-zirconium-based metallic materials. Cold pilgering process is a common tube manufacturing process, which is complex material forming operation in highly non-steady state, where the materials undergo a long series of deformation resulting in both diameter and thickness reduction. During the cold pilgering process, MMLC claddings need to reduce the outside diameter and wall thickness. However, multi-layers of the tube are expected to occur different deformation processes because each layer has different mechanical properties. To improve the utilization of the pilgering process, 3-dimensional computational analyses have been made using a finite element modeling technique. We also analyze the dimensional change, strain and stress distribution at MMLC tube by considering the behavior of rolls such as stroke rate and feed rate.

• 한국압력기기공학회 논문집
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• 제13권2호
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• pp.1-18
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• 2017

세계적으로 원자력발전소의 안정적 운영 및 안전성 확보를 위해 수명기간 중 주요 기기 및 배관의 실제 운전 과도상태를 체계적으로 관리하고, 피로 손상의 정량적 평가 및 관리를 위한 체계적인 시스템이 요구되고 있는 실정이다. 이에 본 논문에서는 원자력발전소의 안전등급 1 설비에 대한 피로 평가요건을 분석하였고, 피로 감시방법 및 절차와 웹 기반으로 개발된 피로 감시 시스템인 NuFMS 개발 및 검증 내용을 기술하였다. NuFMS는 설계 시 고려한 과도상태 발생 횟수 대 비발전소의 특정 운전 시점에서의 실제 발생 횟수를 비교하여 안전 여유도의 정량적 확인이 가능하며, 누적피로사용계수 도출을 통해 정확한 피로영향 분석뿐만 아니라 손상 관리가 가능하다. 이와 같이 NuFMS의 적용을 통해 원자력발전소 기기 및 배관의 피로 건전성을 확인하고 운영 신뢰도를 향상시킬 수 있으며, 발전소의 안전성 유지 및 운영비용 절감 등의 효과를 기대할 수 있다. 따라서 향후 국내 전 원전에 NuFMS를 확대 적용할 예정이며, 이러한 기술의 해외 수출을 적극 추진 중이다.

• 한국압력기기공학회 논문집
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• 제7권3호
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• pp.61-66
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• 2011

The integrity of the SG(Steam Generator) tubes has been challenged by numerous factors such as flaws, operation, atmosphere, inherently degraded materials, loose parts and even human errors. Of the factors, loose parts(or foreign materials) on the secondary side of the tubes can bring about volumetric defects and even leakage from the primary to the secondary side in a short period of time. More serious concerns about the loose parts are their unknown influx path and rapid growth rate of the defects affected by the loose parts. Therefore it is imperative to detect and characterize the foreign materials and the defects. As a part of the measures for loose part detection, TTS(Top of Tubesheet) MRPC(Motorized Rotating Pancake Coils) ECT has been carried out especially to the restricted high probability area of the loose part. However, in the presence of loose parts in the other areas, wide range loose part detection techniques are required. In this study, loose part standard tube was presented as a way to accurately detect and characterize loose part signals. And the SG tube ECT bobbin coil and MRPC ISI(In-service Inspection) data of domestic OPR-1000 and Westinghouse Model F(W_F) were reviewed and consequently, comprehensive loose part detection technique is derived especially by applying bobbin coil signals

• 한국산업융합학회 논문집
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• 제24권6_2호
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• pp.773-779
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• 2021

CNG, which has recently been attracting attention as an alternative fuel in the transportation field to reduce emissions caused by global warming, is natural gas with abundant reserves and mainly composed of methane. Being in a gaseous state, natural gas requires the compression and liquefaction processes for transportation. Until now, general shut-off valves for liquid and gas piping have been developed in Korea, but there are few studies on shut-off valves for high pressures of about 200 bar. Currently, research on the flow analysis of valves is being actively conducted around the world. However, there are relatively many studies on large valves such as low-pressure valves or shipbuilding and marine, and the safety factor through structural analysis to check the structural integrity of the valve is checked at the design stage. Since it is necessary to have a fast response speed while minimizing pressure and speed loss due to flow change, basic research was conducted on the flow analysis of the valve to secure design data, and the numerical analysis was performed on high-pressure automatic shut-off valves applied to CNG refueling stations. After securing the basic valve shape through reverse engineering for advanced products, we compared the valve flow coefficient Cv coefficient with advanced products. As a result, it was found that the reverse engineering model was at the level of about 60%. However, we compared the Cv coefficient by modifying the reverse engineering model, and the result showed that it was improved to about 96%.

• 한국압력기기공학회 논문집
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• 제17권1호
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• pp.18-27
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• 2021

This study aims to develop an improved evaluation technology for assessing CANDU-6 safety. For this purpose, the multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout (SBO) in a CANDU-6 plant was selected as a hypothetical event scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes logic models for controlling the pressure and inventory of the primary heat transport system (PHTS) decreasing due to the u-tubes' rupture, as well as the main features of PHTS with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was successfully achieved to confirm the stable convergence of the key parameters. Until the turbine trip, the fuel channels were adequately cooled by forced circulation of coolant and supply of main feedwater. However, due to the continuous reduction of PHTS pressure and inventory, the reactor and turbine were shut down and the thermal-hydraulic behaviors between intact and broken loops got asymmetric. Furthermore, as the conditions of low-flow coolant and high void fraction in the broken loop persisted, leading to degradation of decay heat removal, it was evaluated that the peak cladding temperature (PCT) exceeded the limit criteria for ensuring nuclear fuel integrity. This study is expected to provide the technical bases to the accident management strategy for transient conditions with multiple events.

• 한국압력기기공학회 논문집
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• 제17권1호
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• pp.36-48
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• 2021

As the material aging of nuclear power plants has been progressing in domestic and overseas, crack growth becomes one of the most important issues. In this respect, the crack growth assessment has been considered an essential part of structural integrity. The crack growth assessment for nuclear power plants has been generally performed based on ASME B&PV Code, Sec. XI but the idealization of crack shape and the conservative solutions of stress intensity factor (SIF) are used. Although finite element analysis (FEA) based on iterative crack growth analysis is considered as an alternative method to simulate crack growth, there are yet no guidelines to model the crack-tip spider-web mesh for such analysis. In this study, effects of various meshing factors on FE SIF calculation are systematically examined. Based on FEA results, proper criteria for spider-web mesh in crack-tip are suggested. The validation of SIF calculation method through mapping initial stress field is investigated to consider initial residual stress on crack growth. The iterative crack-tip modeling program to simulate crack growth is developed using the proposed criteria for spider-web mesh design. The SIF results from the developed program are validated by comparing with those from technical reports of other institutes.

• 한국압력기기공학회 논문집
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• 제20권1호
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• pp.16-24
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• 2024

Denting of steam generator (SG) tube is defined as the reduction in tube diameter due to the stresses exerted by the corrosion products formed on the outer diameter surface. This phenomenon is mostly observed in the crevices between SG tube and the top-of tubesheet or tube support plate. Despite the replacement of SG tube with Alloy 690, which has better corrosion resistance than Alloy 600, the denting of SG tube still remains a potential problem that could decrease the SG integrity. Deformation of SG tube by denting phenomenon can affect the surface properties and microstructure of SG tube. In this study, the effects of plastic deformation on surface properties and microstructure of Alloy 690 thermally treated (TT) tube was investigated by using the various analysis techniques. The plastic deformation of Alloy 690 increased the surface roughness and area. Many surface defects such as ripped surface and micro-cracks were observed on the deformed Alloy 690TT specimen. Based on the electron backscatter diffraction analysis, the dislocation density of deformed SG tube increased compared to non-deformed SG tube. In addition, the effects of changes in surface properties and microstructure of SG tube on general corrosion behavior were discussed.

• 비파괴검사학회지
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• 제33권1호
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• pp.63-68
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• 2013

Some essential components in nuclear power plants are periodically inspected using non-destructive examinations, for example ultrasonic, eddy current and radiographic examinations, in order to determine their integrity. These components include nuclear power plant items such as vessels, containments, piping systems, pumps, valves, tubes and core support structure. Steam generator tubes have an important safety role because they constitute one of the primary barriers between the radioactive and non-radioactive sides of the nuclear power plant. There is potential that if a tube bursts while a plant is operating, radioactivity from the primary coolant system could escape directly to the atmosphere. Therefore, in-service inspections are critical in maintaining steam generator tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due to its high inspection speed and flaw detectability on non-magnetic tubes. However, it is not easy to analyze correctly eddy current signals because they are influenced by many factors. Therefore, the performance of eddy current data analysts for steam generator tubing should be demonstrated comprehensively. In Korea, the performance of steam generator tubing analysts has been demonstrated using the Qualified Data Analyst program. This paper describes the performance demonstration program for steam generator tubing analysts and its implementation results in Korea. The pass rate of domestic analysts for this program was 71.4%.