• 방사성폐기물학회지
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• 제21권2호
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• pp.193-204
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• 2023

To ensure radiological safety margin in the transport and storage of spent nuclear fuel, it is crucial to perform source term and shielding analyses in advance from the perspective of conservation. When performing source term analysis on UO₂ fuel, which is mostly used in commercial nuclear power plants, uranium and oxygen are basically considered to be the initial materials of the new fuel. However, the presence of impurities in the fuel and structural materials of the fuel assembly may influence the source term and shielding analyses. The impurities could be radioactive materials or the stable materials that are activated by irradiation during reactor power operation. As measuring the impurity concentration levels in the fuel and structural materials can be challenging, publicly available information on impurity concentration levels is used as a reference in this evaluation. To assess the effect of impurities, the results of the source term and shielding analyses were compared depending on whether the assumed impurity concentration is considered. For the shielding analysis, generic cask design data developed by KEPCO-E&C was utilized.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4030-4048
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• 2022

In this study, road transportation tests were conducted with surrogate fuel assemblies under normal conditions of transport to evaluate the vibration and shock load characteristics of spent nuclear fuel (SNF). The overall test data analysis was conducted based on the measured acceleration and strain data obtained from the speed bump, lane-change, deceleration, obstacle avoidance, and circular tests. Furthermore, representative shock response spectrums and power spectral densities of each test mode were acquired. Amplification or attenuation characteristics were investigated according to the load transfer path. The load attenuated significantly as it transferred from the trailer to the cask. By contrast, the load amplified as it transferred from the cask to the surrogate SNF assembly. The fuel loading location on the cask disk assembly did not exhibit a significant influence on the strain measured from the fuel rods. The principal strain was in the vertical direction, and relatively large strain values were obtained in spans with large spacing between spacer grids. The influence of the lateral location of fuel rods was also investigated. The fuel rods located at the side exhibited relatively large strain values than those located at the center. Based on the strain data obtained from the test results, a hypothetical road transportation scenario was established. A fatigue evaluation of the SNF rod was performed based on this scenario. The evaluation results indicate that no fatigue damage occurred on the fuel rods.

• Nuclear Engineering and Technology
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• 제41권4호
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• pp.493-520
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• 2009

During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained $UO_2$ pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

• Advances in Energy Research
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• 제5권2호
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• pp.91-105
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• 2017

Thermal hydraulic (TH) analysis of nuclear power reactors is utmost important. In this way, the numerical codes that preparing TH data in reactor core are essential. In this paper, a subchannel analysis of a Russian pressurized water reactor (WWER1000) core with enhanced numerical code is carried out. For this, in fluid domain, the mass, axial and lateral momentum and energy conservation equations for desired control volume are solved, numerically. In the solid domain, the cylindrical heat transfer equation for calculation of radial temperature profile in fuel, gap and clad with finite difference and finite element solvers are considered. The dependence of material properties to fuel burnup with Calza-Bini fuel-gap model is implemented. This model is coupled with Isotope Generation and Depletion Code (ORIGEN2.1). The possibility of central hole consideration in fuel pellet is another advantage of this work. In addition, subchannel to subchannel and subchannel to rod connection data in hexagonal fuel assembly geometry could be prepared, automatically. For a demonstration of code capability, the steady state TH analysis of a the WWER1000 core is compromised with Thermal-hydraulic analysis code (COBRA-EN). By thermal hydraulic parameters averaging Fuel Assembly-to-Fuel Assembly method, the one sixth (symmetry) of the Boushehr Nuclear Power Plant (BNPP) core with regular subchannels are modeled. Comparison between the results of the work and COBRA-EN demonstrates some advantages of the presented code. Using the code the thermal modeling of the fuel rods with considering the fission gas generation would be possible. In addition, this code is compatible with neutronic codes for coupling. This method is faster and more accurate for symmetrical simulation of the core with acceptable results.

• 한국압력기기공학회 논문집
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• 제18권2호
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• pp.43-49
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• 2022

The purpose of this study is to evaluate the radiation effect on damage when the external shield of the spent nuclear fuel transport cask is damaged due to impact as the cause of an unexpected accident. The neutron and gamma-ray intensities and spectra are calculated using the ORIGEN-Arp module in the SCALE 6.2.4 code package⁽¹⁾ and then using MCNP6.2⁽²⁾ code calculate the dose rate. In order to evaluate the radiation dose according to the size of damage caused by external impact, various sized holes of 0.3~13.7% are assumed in the outer shield of the cask to evaluate the sensitivity to the dose. In the case of radiation source leakage, damage to the nuclear fuel assembly is assumed to be up to 6% based on overseas test cases. When only the outer shield is damaged, the maximum surface dose is calculated as 3.12E+03 mSv/hr. However, if the radiation source is leaked due to damage to the nuclear fuel assembly, it becomes 7.00E+05 mSv/hr which is about 200 times greater than the former case.

• 항공우주시스템공학회지
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• 제12권3호
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• pp.1-8
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• 2018

항공기에서 사용되는 연료펌프를 민군겸용구성품개발사업을 통하여 국내 최초로, 국내 독자기술로 개발하였다. 연료승압펌프의 모터 조립체에 있어서 모터는 효율성, 내구성, 내폭발성 등의 관점에서 DC 브러시 모터보다 특성이 우수한 BLDC 형식을 채택하였다. Maxwell 방정식과 환경조건을 활용하고 자속밀도를 통하여 전자기해석을 수행하였다. 설계를 검증하기 위하여 유한요소법을 이용하여 부하에 따른 모터성능을 해석하였고 모터 드라이브와 EMI 필터를 설계하여 모터 조립체가 개발되었다. 연료승압펌프 모터 조립체의 성능시험 결과는 해석결과와 매우 일치하는 것으로 확인되었다.

• 센서학회지
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• 제23권4호
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• pp.245-250
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• 2014

When the charged particle travels in transparent medium with a velocity greater than that of light in the same medium, the electromagnetic field close to the particle polarizes the medium along its path, and then the electrons in the atoms follow the waveform of the pulse which is called as Cerenkov light or radiation. This type of radiation can be easily observed in a spent fuel storage pit. In optical fibers, the Cerenkov light also can be generated due to their dielectric components. Accordingly, the radiation-induced light signals can be obtained using optical fibers without any scintillating material. In this study, to measure the intensities of Cerenkov radiation induced by gamma-rays, we have fabricated the fiber-optic Cerenkov radiation sensor system using silica optical fibers, plastic optical fibers, multi-anode photomultiplier tubes, simulated spent fuel assembly and a scanning system. To characterize the Cerenkov radiation generated in optical fibers, the intensities of Cerenkov radiation generated in the silica and plastic optical fibers were measured. Also, we measured the longitudinal distribution of gamma rays emitted from the Ir-192 isotope by using the fiber-optic Cerenkov radiation sensor system and simulated spent fuel assembly.

• 대한기계학회논문집B
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• 제40권5호
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• pp.329-337
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• 2016

연료집합체의 지지격자에 설치된 혼합날개는 난류 강화 기구로서 부수로 내부에서 선회류 또는 연료봉 간극사이에서 횡류를 발생시켜 대류열전달을 증진시키는 역할을 한다. 따라서 혼합날개의 기하학적인 형상 및 배열 형태는 혼합날개의 성능을 결정하는 중요한 인자이다. 본 연구에서는 OECD/NEA의 벤치마크 계산에서 활용된 분할 형태의 혼합날개가 장착된 $5{\times}5$ 연료집합체 내부에서의 유동분포 특성을 파악하기 위해 상용 전산유체역학 소프트웨어인 ANSYS CFX R.14를 사용하여 계산을 수행하였고, 계산결과를 MATiS-H 시험장치의 측정값과 비교하였다. 또한 분할 형태의 혼합날개 형상이 연료집합체 내부유동 형태에 미치는 영향에 대해 설명하였다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.83-89
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• 2003

조사후 시험시설내에는 사용후 핵연료 집합체의 취급을 위하여 감온, 감압 공정이 있다. 이 공정에는 3가지 공정으로 분류하는데 첫째, 사용후핵연료집합체 캐스크를 제염하기 위한 제염시키는 공정, 둘째, 사용후핵연료집합체 내의 붕괴열에 의해 온도, 압력이 상승된 폐액을 감온, 감압 시키기 위한 냉각 공정 셋째, 사용후핵연료 피폭관 결함에 의해 발생되어 캐스크 내에 존재하는 불용성 입자를 여과기를 통해 여과하는 공정으로 되어 있다. 본 보고서에서는 감온, 감압 공정과 관련하여 현재까지 수행된 기술검토와 사용후핵연료집합체에 의한 감온, 감압의 실용적 이론에 관해 고찰하였고 또한 각종 시험을 통한 시운전 내용과 실제 원자력발전소로부터 수송해온 사용후핵연료집합체 J-44, K-23 대한 감온, 감압 결과들을 상세히 기술하였다. 본 보고서는 향후 지속적인 가동과 도출되지 않은 문제점 등을 계속 보완하여, 원만하고 안전한 정상조업을 수행하는데 효과적으로 이용될 수 있을 것으로 본다.

• Nuclear Engineering and Technology
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• 제43권2호
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• pp.149-158
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• 2011

Metal waste generated from the pyroprocessing of 10 MtU of spent fuel was classified by comparing the specific activity of a relevant radionuclide with the limit value of the specific activity specified in the Korean acceptance criteria for a lowand intermediate-level waste repository. A Korean Optimized Fuel Assembly design with a 17${\times}$17 array, an initial enrichment of 4.5 weight-percent, discharge burn-up of 55 GWD/MtU, and a 10-year cooling time was considered. Initially, the mass and volume of each structural component of the assembly were calculated in detail, and a source term analysis was subsequently performed using ORIGEN-S for these components. An activation cross-section library generated by the KENO-VI/ORIGEN-S module was utilized for top-end and bottom-end pieces. As a result, an Inconel grid plate, a SUS plenum spring, a SUS guide tube subpart, SUS top-end and bottom-end pieces, and an Inconel top-end leaf spring were determined to be unacceptable for the Gyeongju low- and intermediate-level waste repository, as these waste products exceeded the acceptance criteria. In contrast, a Zircaloy grid plate and guide tube can be placed in the Gyeongju repository. Non-contaminated Zircaloy cladding occupying 76% of the metal waste was found to have a lower level of specific activity than the limit value. However, Zircaloy cladding contaminated by fission products and actinides during the decladding process of pyroprocessing was revealed to have 52 and 2 times higher specific activity levels than the limit values for alpha and $^{90}Sr$, respectively. Finally, it was found that 88.7% of the metal waste from the 17${\times}$17 Korean Optimized Fuel Assembly design should be disposed of in a deep geological repository. Therefore, it can be summarized that separation technology with a higher decontamination factor for transuranics and strontium should be developed for the efficient management of metal waste resulting from pyroprocessing.