• 제목/요약/키워드: Spent fuel rod

검색결과 81건 처리시간 0.023초

사용후핵연료 저장용기의 정상 및 비정상조건에 대한 열해석 (Thermal Analysis of a Spent Fuel Storage Cask under Normal and Off-Normal Conditions)

  • Ju-Chan Lee;Kyung-Sik Bang;Ki-Seog Seo;Ho-Dong Kim;Byung-Il Choi;Heung-Young Lee
    • 방사성폐기물학회지
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    • 제2권1호
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    • pp.13-22
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    • 2004
  • This study presents the thermal analyses of a spent fuel dry storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 15 $^{\circ}C$ under the normal condition. The off-normal condition has an environmental temperature of 38 $^{\circ}C$. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Two of the four air inlet ducts are assumed to be completely blocked. The significant thermal design feature of the storage cask is the air flow path used to remove the decay heat from the spent fuel. Natural circulation of the air inside the cask allows the concrete and fuel cladding temperatures to be maintained below the allowable values. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. The maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal and off-normal conditions.

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Realistic thermal analysis of the CANDU spent fuel dry storage canister

  • Tae Gang Lee;Taehyeon Kim;Taehyung Na;Byongjo Yun;Jae Jun Jeong
    • Nuclear Engineering and Technology
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    • 제55권12호
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    • pp.4597-4606
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    • 2023
  • Thermal analysis of the CANDU spent fuel dry storage canister is very important to ensure the integrity of the spent fuel. The analyses have been conducted using a conservative approach, with a particular focus on the peak cladding temperature (PCT) of the fuel rods in the canister. In this study, we have performed a realistic thermal analysis using a computational fluid dynamics (CFD) code. The canister contains 9 fuel bundle baskets. A detailed analysis of even a single basket requires significant computational resources. To overcome this challenge, we replaced each basket with an equivalent heat conductor (EHC), of which effective thermal conductivity (ETC) is developed from the results of detailed CFD calculations of a fuel bundle basket. Then, we investigated the effects of some conservative models, ultimately aiming at a realistic analysis. The results revealed: (i) The influence of convective heat transfer in the basket cannot be ignored, but it's less significant than expected. (ii) Modeling of the lifting rod is crucial, as it plays a decisive role in axial heat transfer at the center of the canister and significantly reduces the PCT. (iii) Convection within the canister is very important, as it not only reduces the PCT but also shifts its location upwards.

산화속도 및 회수율 향상을 위한 고효율 장치 핵심 메커니즘 설계 (Design on Main Mechanism of High Throughput Device for Enhancement of Oxidation and Recover Rate)

  • 김영환;박병석;정재후;윤지섭;황정식
    • 한국소성가공학회:학술대회논문집
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    • 한국소성가공학회 2008년도 춘계학술대회 논문집
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    • pp.473-476
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    • 2008
  • In this paper, we designed the main mechanism of high throughput device for rod-cuts of spent fuel. For this, we analyzed the mechanical methods(slitting, ball mill, roller straightening) and chemical methods(muffle furnace, rotary kiln). As the results, methods of ball drop and rotary drum for concepts design were selected in the analysis step. For enhancement of oxidation rate, we devised the blades on the reactor with mesh type. Also, for enhancement of decladding rate, we designed ball size and rotation reactor with mesh type and devised the vacuum system for fission products. Mechanisms of oxidation and recovery can simultaneously handle the rod-cuts of spent fuel and independently recover. The results of mechanism design can be used for scale-up of high throughput device.

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가압경수로 사용후핵연료 이용확대 방안연구 (A Scheme of Better Utilization of PWR Spent Fuels)

  • Chung, B.J.;Kang, C.S.
    • Nuclear Engineering and Technology
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    • 제23권2호
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    • pp.165-173
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    • 1991
  • 가압경수로의 사용후핵연료를 CANDU 원자로에 재순환시키는, 이른바, 탄뎀 핵연료주기가 본 연구에서 다루어졌다. 이러한 방식으로 가압경수로의 사용후핵연료를 활용하는 것은 우라늄자원의 이용을 개선시킬뿐만 아니라 사용후핵연료 저장능력의 부족도 다소 해결할 수 있을 것이다. 핵연료를 재순환 시키는데 있어서는 CANDU 원자로의 수정을 최소화하는 방향으로 연구가 진행되었으며 본 연구에서는 9종의 핵연료가 고려되었다. 탄뎀 핵연료는 크게 핵연료재가공과 노심재구성의 두 분야로 나뉘어지는데, 핵연료 재가공의 경우, 가압경수로의 사용후핵연료는 처리되고 현재의 37 봉형 격자구조인 핵연료 다발에 맞도록 다시 성형가공되며 노심재구성의 경우, 가압경수로 사용후핵연료는 단지 격자 구조를 해체하고 CANDU의 격자길이에 맞춰 재구성만 된다. 각 탄뎀 핵연료 옵션에 대하여, 허용연소도와 출력분포를 계산하기 위해 노심연소계산이 수행되었다. 또한 경제성에 대한 접근으로 각 핵연료 옵션에 대한 핵주기비가 계산되었다. 그 결과 본 연구에서 다루어진 대부분의 탄뎀 핵연료 옵션이 경제성이 있었을 뿐만 아니라 기술적인 타당성이 있었다.

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펠릿과 헐의 분리 연구를 위한 슬리팅 장치 개발 (Development of the slitting device on separation study of pellet and hull)

  • 정재후;윤지섭;홍동희;김영환;진재현;박기용
    • 한국소성가공학회:학술대회논문집
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    • 한국소성가공학회 2003년도 춘계학술대회논문집
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    • pp.236-239
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    • 2003
  • The spent fuel slitting device is an equipment developed in order to feed UO$_2$pellet to the dry pulverizing/mixing device. In this study, we have compared and analyzed the handling method of the slitting and that of the pellet and hull, processing time, separating time for 20kgHM, the number of blades, on the existing slitting device using in DUPIC, and spent fuel management technology research and test facility. Also, we have compared and analyzed about an advantage and weak point, designing and producing, processing, establishment, operation, maintenance about the vertical and horizontal slitting device. Based on these results, we have developed the vertical slitting device. By using the results, we have enhanced the slitting processing time(over 40%)in comparison with DUPIC device, and it will is effectively applied to available data for designing and producing of the hot test facility.

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RC 기술의 자동화 장치 개발에 관한 연구(I) -장치 설계 및 제작- (A Study on the Automation Equipment Development of RC Technology(I) -Equipment Design and Manufacture-)

  • 김기준;김상진;송자윤
    • 대한전기학회:학술대회논문집
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    • 대한전기학회 1994년도 추계학술대회 논문집 학회본부
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    • pp.311-313
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    • 1994
  • RC(Rod Consolidation) technology in spent nuclear fuel management is an essential method and requires remote operation due to radiation exposure. Its technology may provide an effective means to double the storage capacity of spent fuel storage space. So development of this technology will provide a valuable contribution to establishing economical as well as technological basis for future spent fuel management.

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Dry storage of spent nuclear fuel and high active waste in Germany-Current situation and technical aspects on inventories integrity for a prolonged storage time

  • Spykman, Gerold
    • Nuclear Engineering and Technology
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    • 제50권2호
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    • pp.313-317
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    • 2018
  • Licenses for the storage of spent nuclear fuel (SNF) and vitrified highly active waste in casks under dry conditions are limited to 40 years and have to be renewed for prolonged storage periods. If such a license renewal has to be expected since as in accordance with the new site selection procedure a final repository for spent fuel in Germany will not be available before the year 2050. For transport and possible unloading and loading in new casks for final storage, the integrity and the maintenance of the geometry of the cask's inventory is essential because the SNF rod cladding and the cladding of the vitrified highly active waste are stipulated as a barrier in the storage concept. For SNF, the cladding integrity is ensured currently by limiting the hoop stress and hoop strain as well as the maximum temperature to certain values for a 40-year storage period. For a prolonged storage period, other cladding degradation mechanisms such as inner and outer oxide layer formation, hydrogen pick up, irradiation damages in cladding material crystal structure, helium production from alpha decay, and long-term fission gas release may become leading effects driving degradation mechanisms that have to be discussed.

Uncertainty quantification in decay heat calculation of spent nuclear fuel by STREAM/RAST-K

  • Jang, Jaerim;Kong, Chidong;Ebiwonjumi, Bamidele;Cherezov, Alexey;Jo, Yunki;Lee, Deokjung
    • Nuclear Engineering and Technology
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    • 제53권9호
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    • pp.2803-2815
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    • 2021
  • This paper addresses the uncertainty quantification and sensitivity analysis of a depleted light-water fuel assembly of the Turkey Point-3 benchmark. The uncertainty of the fuel assembly decay heat and isotopic densities is quantified with respect to three different groups of diverse parameters: nuclear data, assembly design, and reactor core operation. The uncertainty propagation is conducted using a two-step analysis code system comprising the lattice code STREAM, nodal code RAST-K, and spent nuclear fuel module SNF through the random sampling of microscopic cross-sections, fuel rod sizes, number densities, reactor core total power, and temperature distributions. Overall, the statistical analysis of the calculated samples demonstrates that the decay heat uncertainty decreases with the cooling time. The nuclear data and assembly design parameters are proven to be the largest contributors to the decay heat uncertainty, whereas the reactor core power and inlet coolant temperature have a minor effect. The majority of the decay heat uncertainties are delivered by a small number of isotopes such as 241Am, 137Ba, 244Cm, 238Pu, and 90Y.