• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.20-24
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• 2013

Reactor Vessel Internals(RVIs), which are installed within the reactor pressure vessel and support the fuel assembly, take responsibility for safety of reactor core. In operating Nuclear Power Plants(NPPs), the RVIs have been exposed to severe conditions such as neutron irradiation, high temperature, high pressure, and high velocity of coolant flow and have degraded by materials aging with long-term operation. Therefore, the effective aging management plan and the appropriate regulatory requirements are necessary to maintain the integrity of RVIs. The purpose of this paper is to provide a review guide for Periodic Safety Review(PSR) of RVIs in presurized water reactor. The review guide is developed based on the revised review guides and reports established from IAEA and USNRC, and the analysis results of design characteristics, aging mechanisms, and operating experiences of RVIs in domestic and international NPPs. Consequently, the developed review guide for PSR of RVIs is expected to contribute an overall strategy and standard for the PSR of RVIs.

• Nuclear Engineering and Technology
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• 제38권3호
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• pp.301-310
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• 2006

Destructive methods were used for the burnup determination of U-Zr alloy nuclear fuel irradiated in the High-flux Advanced Neutron Application Reactor (HANARO) at KAERI. The dissolution rate of unirradiated U-Zr alloy fuel in $HNO_3$/HF mixtures was investigated for the experimental conditions of a different temperature, and initial concentrations of HF and $HNO_3$. The irradiated U-Zr alloy fuel specimen was dissolved in a mixed acid condition of 3 M HNO3 and 1 M HF at $90^{\circ}C$ for 8 hours under reflux. The total burnup was determined from measurement of the Nd isotope burnup monitors. The method includes U, Pu, $^{148}Nd,\;^P{145}Nd+^{146}Nd,\;^{144}Nd+^{143}Nd$ and total Nd isotopes determination by the isotope dilution mass spectrometric method (IDMS) using triple spikes $(^{233}U,\;^{242}Pu\;and\;^{150}Nd)$. The effective fission yield was calculated from the weighted fission yields averaged over the irradiation period. The results are compared with that obtained by the destructive -spectrometric measurement of the $^{137}Cs$ monitor.

• 비파괴검사학회지
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• 제17권1호
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• pp.31-40
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• 1997

원자로 압력용기 재료의 중성자 조사 취화 문제는 원자력발전소의 안전성 및 수명 관리에 가장 중대 한 영향을 미친다. 재료의 조사 취화를 평가하기 위하여 수행하고 있는 충격 및 인장시험 같은 파괴적 시험 결과는 석출물 크기나 분포, 전위 밀도 등, 재료 자체의 조직학적 특성에 좌우되므로 한정된 시편을 이용한 평가에는 많은 불확실성이 존재하게 된다. 따라서 이와 같은 문제점을 해결하기 위하여 비파괴기술을 이용한 조사 취화 평가에 대한 많은 연구가 진행되고 있다. 현재 원자로 압력용기 재료의 조사 취화에 따른 미세 조직 변화를 분석하기 위하여 응용되고 있는 비파괴기술로는 전기, 자기, 전자기, 초음파 및 경도측정법 등이 있으나 비파괴피험 결과와 미세조직의 변화, 기계적 성질 및 취화 정도 등과의 상관 관계를 정립해야만 기존 파괴적 시험의 대체가 가능하게 된다. 따라서 현재까지 수행되고 있는 여러 비파괴기술을 이용한 조사 취화 평가 연구결과를 비교 분석하여 보다 실현 가능성 있는 비파괴기술을 검토하였다.

• 전기학회논문지
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• 제61권12호
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• pp.1948-1952
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• 2012

The paper presents modal testing and analysis in order to obtain the dynamic characteristics of a middle spacer grids of a nuclear fuel rod. A spacer grid is one of the important structural elements supporting nuclear fuel rods. Such a fuel rod can be oscillated by its thermal expansion, neutron irradiation and etc. due to cooling water flow under the operation of a nuclear power plant. When the fuel rod vibrates, fretting wear due to repeated friction motion between the fuel rods and spacer grids can be occurred, and so the fuel rod is damaged. In this paper, through modal analysis and testing, natural frequencies and modes of a middle spacer grid were calculated, and the following conclusions were obtained. Firstly the numerical first-seven natural frequencies for spacer grids of a fuel rod having complicated structures have a small difference within 3.8% with experimental natural frequencies, and so the suitability of simulation results was verified. Secondly, experimental mode shapes for a middle spacer grid of a nuclear fuel rod were verified by obtaining lower non-diagonal terms through MAC(Modal Assurance Criteria), and were confirmed by the simulation modes.

• 한국전산유체공학회지
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• 제16권3호
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• pp.95-103
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• 2011

In the block type VHTR core, there are inevitable gaps among core blocks for the installation and refueling of the fuel blocks. These gaps are called bypass gap and the bypass flow is defined as a coolant flows through the bypass gap. Distribution of core bypass flow varies according to the reactor operation since the graphite core blocks are deformed by the fast neutron irradiation and thermal expansion. Furthermore, the cross-flow through an interfacial gap between the stacked blocks causes flow mixing between the coolant holes and bypass gap, so that complicated flow distribution occurs in the core. Since the bypass flow affects core thermal margin and reactor efficiency, accurate prediction and evaluation of the core bypass flow are very important. In this regard, experimental and computational studies were carried out to evaluate the core bypass flow distribution. A multi-block experimental apparatus was constructed to measure flow and pressure distribution. Multi-block effect such as cross flow phenomenon was investigated in the experiment. The experimental data were used to validate a CFD model foranalysis of bypass flow characteristics in detail.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.899-906
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• 2018

The irradiation performance of $U_3Si$ dispersion fuel in an Al matrix, $U_3Si-Al$, under the Hi-Flux Advanced Neutron Application Reactor (HANARO) design full-power condition of 30 MW was tested for full-power qualification of the fuel. A test assembly was fabricated containing 18 fuel rods made with atomized $U_3Si$ powder manufactured at the Korea Atomic Energy Research Institute. The test assembly was irradiated for 188 full-power operation days in the HANARO subject to the normal fuel-loading scheme and achieved about 60 at% U-235 average burnup and 75 at% U-235 peak burnup. The maximum linear power of the test assembly was 98 kW/m. Nondestructive and destructive postirradiation examinations were conducted. The measured postirradiation examination data were compared with data from previous irradiations and the design criteria required for HANARO fuel. Consequently, it was concluded that in-pile performance was acceptable and fuel integrity was maintained, and the behavior satisfied the fuel design requirements.

• 방사선산업학회지
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• 제9권4호
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• pp.167-170
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• 2015

Radioisotope (Nuclear) battery using $^{63}Ni$ was prepared as beta cell. The electroplated $^{63}Ni$ on Ni foil is fabricated, and beta cell and photovoltaic hybrid battery was designed to use at both day and night in space project. A Ni-plating solution is prepared by dissolving metal particles including $^{62}Ni$ and $^{63}Ni$ from neutron irradiation of ($n,{\gamma}$). Electroplating solution of a chloride bath consists on nickel ions in HCl, $H_3BO_3$, and KOH. The deposition was carried out at current density of $10mA\;cm^{-2}$. The prepared beta source was attached on a PN junction and measured I-V properties. The power output at activity of 0.07 mCi and 0.45 mCi were 0.55 pW and 2.69 nW, respectively.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.715-731
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• 2021

Uncertainties are calculated for pressurized water reactor (PWR) spent nuclear fuel (SNF) characteristics. The deterministic code STREAM is currently being used as an SNF analysis tool to obtain isotopic inventory, radioactivity, decay heat, neutron and gamma source strengths. The SNF analysis capability of STREAM was recently validated. However, the uncertainty analysis is yet to be conducted. To estimate the uncertainty due to nuclear data, STREAM is used to perturb nuclear cross section (XS) and resonance integral (RI) libraries produced by NJOY99. The perturbation of XS and RI involves the stochastic sampling of ENDF/B-VII.1 covariance data. To estimate the uncertainty due to modeling parameters (fuel design and irradiation history), surrogate models are built based on polynomial chaos expansion (PCE) and variance-based sensitivity indices (i.e., Sobol' indices) are employed to perform global sensitivity analysis (GSA). The calculation results indicate that uncertainty of SNF due to modeling parameters are also very important and as a result can contribute significantly to the difference of uncertainties due to nuclear data and modeling parameters. In addition, the surrogate model offers a computationally efficient approach with significantly reduced computation time, to accurately evaluate uncertainties of SNF integral characteristics.

• Journal of Radiation Protection and Research
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• 제48권1호
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• pp.52-57
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• 2023

Background: Korea Atomic Energy Research Institute (KAERI) operates several nuclear research facilities licensed by Nuclear Safety and Security Commission (NSSC). The emergency preparedness requirements, GSR Part 7, by International Atomic Energy Agency (IAEA) request protection strategy based on the hazard assessment that is not applied in Korea. Materials and Methods: In developing the protection strategy, it is important to consider an accident scenario and its consequence. KAERI has tried the hazard assessment based on a hypothesis accident scenario for the major nuclear facilities. During the assessment, the safety analysis report of the related facilities was reviewed, the simulation using MELCOR, MACCS2 code was implemented based on a considered accident scenario of each facility, and the international guidance was considered. Results and Discussion: The results of the optimized protective actions were 300 m evacuation and 800 m sheltering for the High-Flux Advanced Neutron Application Reactor (HANARO), the evacuation to radius 50 m, the sheltering 400 m for post-irradiation examination facility (PIEF), 100 m evacuation or sheltering for HANARO fuel fabrication plant (HFFP) facility. Conclusion: The results of the optimized protective actions and its distances for the KAERI facilities for the maximum postulated accidents were considered in establishing the emergency plan and procedures and implementing an emergency exercise for the KAERI facilities.

• 치과방사선
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• 제23권2호
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• pp.229-250
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• 1993

This study was performed to investigate the morphological and structural changes of bone tissues and the effects of irradiation on the mandibular bodies of rats which were fed low calcium diets. In order to carry out this experiment, 160 seven-week old Sprague-Dawley strain rats weighing about 150 gm were selected and equally divided into one normal diet group of 80 rats and one low calcium diet group with the remainder. These groups were then subdivided into two groups, 40 were assigned rats for each subdivided group, exposed to radiation. The Group 1 was composed of forty non-irradiated rats with normal diet, Group 2 of forty irradiated rats with normal diet, Group 3 forty non-irradiated rats with low calcium diet, and Group 4 forty irradiated rats with low calcium diet. The two irradiation groups received a single dose of 20 Gy on the jaw area only and irradiated with a cobalt-50 teletherapy unit. The rats with normal and low calcium diet groups were serially terminated by ten on the 3rd, the 7th, the 14th, and the 21st day after irradiation. After termination, both sides of the dead rats mandible were removed and fixed with 10% neutral formalin. The bone density of mandibular body was measured by use of bone mineral densitometer(Model DPX -alpha, Lunar Corp., U.SA). Triga Mark ill nuclear reactor in Korea Atomic Research Institute was used for neutron activation and then calcium contents of mandibular body were measured by using a 4096 multichannel analyzer (EG and G ORTEC 919 MCA, U.SA). Also the mandibular body was radiographed with a soft X-ray apparatus(Hitex Co., Ltd., Japan). Thereafter, the obtained microradiograms were observed by a light microscope and were used for the morphometric analysis using a image analyzer(Leco 2001 System, Leco Co., Canada). The morphometric analysis was performed for parameters such as the total area, the bone area, the inner and outer perimeters of the bone. The obtained results were as follows: 1. In the morphometric analysis, total area and outer perimeter of the mandibular bodies of Group 3 were a little smaller than that of Group 1. The mean bone width and bone area were much smaller than that of Group 1 and the inner perimeter of Group 3 was much longer than that of Group 1. The total area and outer perimeter of Group 2 and Group 4 showed little difference. The mean bone width and bone area of Group 4 were smaller than that of Group 2 and the inner perimeter of Group 4 was longer than that of Group 2. 2. The remarkable decreases of the number and thickness of trabeculae and also the resorption of endosteal surface of cortical bone could be seen in the microradiogram of Group 3, Group 4 since the 3rd day of experiment. On the 21st day of experiment, the above findings could be more clearly seen in Group 4 than in Group 3. 3. The bone mineral density of Group 3 was lesser than that of Group 1 and the bone mineral density of Group 4 was lesser than that of Group 2 on the 7th, 14th, 21st days. The irradiation caused the bone mineral density to be decreased regardless of diet. In the case of Groups with low calcium diet, the bone mineral density was much decreased on the 21st day than on the 3rd day of experiment. 4. The calcium content in mandible of Group 3 was smaller than that of Group 1 throughout the experiment. roup 4 showed the least amount of calcium content. The irradiation caused the calcium content to be decreased regardless of diet. In the case of Groups with low calcium diet, the calcium content was much decreased on the 21st day than on the 3rd day of experiment. In conclusion, the present study demonstrated that morphological changs and decrease of bone mass due to resorption of bone by low calcium diet, and that the resorption of bone could be found in the spongeous bone and endosteal surface of cortical bone. So the problem of resorption of bone must be considered when the old and the postmenopausal women are taken radiotherapy because the irradiation seems to be accelerated the resorption of osteoporotic bone.