• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 추계학술논문요약집
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• pp.104-104
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• 2018

Kori Unit 1 is the first permanent shutdown nuclear power plant in Korea and it is on June 18th, 2017. Spent fuel assemblies began to be discharged from the reactor core to the spent fuel pool(SFP) within one week after shutdown of Kori unit 1 and the campaign was completed on June 27th, 2017. The total number of spent nuclear fuel assemblies in SFP of Kori Unit-1 is 485 and their discharging date is different respectively. So, decay heat was evaluated considering the actual enrichment, operation history and cooling time of the spent fuel assemblies stored in SFP of the Kori Unit-1. The code used in the evaluation is the ORIGEN-based CAREPOOL system developed by KHNP. Decay heat calculation of PWR fuel is based on ANSI/ANS 5.1-2005, "Decay heat power in light water reactors" and ISO-10645, "Nuclear energy - Light water reactors - Calculation of the decay heat power in nuclear fuels. Also, we considered the contribution of fission products, actinide nuclides, neutron capture and radioactive material in decay heat calculation. CAREPOOL system calculates the individual and total decay heat of all of the spent fuel assemblies in SFP of Kori Unit-1. As a result, the total decay heat generated in SFP on June 28th, 2017 when the spent fuel assemblies were discharged from the reactor core, is estimated to be about 4,185.8 kw and to be about 609.5 kw on September 1st, 2018. It was also estimated that 119.6 kw is generated in 2050 when it is 32 years after the permanent shutdown. Figure 1 shows the trend of total decay heat in SFP of Kori Unit-1.

• Corrosion Science and Technology
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• 제18권5호
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• pp.175-181
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• 2019

A failed spent fuel rod with 53,000 MWd/tU from a nuclear power plant was characterized, and the fission products and oxygen layer in the pellet-clad mechanical interaction region were observed using an EPMA (Electron Probe Micro-Analyzer). A sound fuel rod burned under similar conditions was used to compare and analyze, the results of the failed fuel rod. In the failed fuel rod, the oxide layer represented $10{\mu}m$ of the boundary of the cladding, and $35{\mu}m$ of the region outside the cladding. By comparison, in the sound fuel rod, the oxide layer was $8{\mu}m$, observed in the cladding boundary region. The cladding inner surface corrosion and the resulting fuel-cladding bonding were investigated using an EPMA. Zirconium existed in the bonding layer of the (U, Zr)O compound beyond the pellet cladding interaction gap of $20{\mu}m$, and composition of UZr2O3 was observed in the failed fuel rod. This paper presents the results of the EPMA examination of a spent fuel specimen, and a technique to analyze fission products in the pellet-clad mechanical interaction region.

• 방사성폐기물학회지
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• 제5권1호
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• pp.53-64
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• 2007

고연소도 경수로사용후핵연료를 이용하여 voloxidation 및 소결 열처리 공정으로부터 세슘의 시간에 따른 방출 거동을 실험적으로 평가하였다. 사용후핵연료 voloxidation 공정에서는 fragment 형태의 시편을 사용하여 최대 $1,500^{\circ}C$의 산화 및 환원 분위기에 따른 세슘 방출 거동을 상호 비교하였으며, 소결 공정에서는 압분체를 이용하여 4% H2/Ar 환원분위기 에서 열처리 온도 변화에 따른 세슘방출 특성 변화를 분석하였다. 산화 분위기에서 fragment 형태의 사용후핵연료로부터 세슘 방출 온도 구간은 $800{\circ}C{\sim}1,200^{\circ}C$였으며, 환원 분위기에서 압분체로부터 방출 온도 구간은 $1,100{\circ}C{\sim}1,400^{\circ}C$로서, 산화에 의한 사용후핵 연료의 분말화가 세슘 방출 거동에 영향을 미치는 것으로 나타났다. 아울러 사용후핵 연료로부터 세슘 방출 거동에 영향을 미치는 주요 인자는 사용후핵 연료내 세슘 화합물의 화학적 형태뿐만 아니라 결정립 및 핵연료 표면으로의 확산 속도에 지배를 받음을 알 수 있었다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
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• pp.551-556
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• 1998

The correlation of isotope composition of Zr with the turnup and some heavy isotopes in PWR uranium dioxide fuel has been investigated. The total and partial ($^{235}$ U) burnup were determined by $^{148Nd}$ and by U and Pu mass spectrometric method, respectively. After separating Zr from the fuel samples, its isotope composition was measured by mass spectrometry. In addition, the quantities of the U and Pu in the spent fuel were determined by isotope di lution mass spectrometric method using $^{233}$ U and $^{242}$ Pu as spikes. The content of some heavy isotopes, $^{235}$ U, $^{239}$ Pu and $^{241}$ Pu, and the Pu Contribution to total turnup were expressed by the correlation with Zr isotope ratios, $^{91}$ Zr/$^{96}$ Zr and $^{93}$ Zr/$^{96}$ Zr The correlations by isotope compositions measured were compared wi th those calculated from ORIGEN2 code.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2004년도 학술논문집
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• pp.261-267
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• 2004

처분시스템 설계 시 기초 자료로 사용되는 국내 사용후핵연료의 발생량, 특징 및 연소이력 등의 현재 및 향후 현황을 파악하였다. 2055년까지 PWR 및 CANDU 사용후핵연료 발생량은 각각 20,500 및 14,800MTU로 나타났다.$17{\times}17$ 핵연료 집합체의 사용후핵연료 발생량비율은 2003년 기준으로 전체대비 60%를 점유하는 것으로 나타났으며, 2012년 이후부터는 .$16{\times}16$ KSFA 사용후핵연료 발생량이 .$17{\times}17$ 핵연료를 능가하기 시작하여 최종시점인 2055년에는 70% 정도를 점유할 것으로 보인다. 사용후핵연료의 평균 연소도는 90년대 후반에는 36GWD/MUT 정도, 2000년대 초반에는 40GWD/MTU를 나타냈으며, 2000년대 중ㆍ후반부터는 45GWD/MTU를 초과할 것으로 보인다. 따라서, 현재는 1997년에 선정한 제원을 기준 핵연료 제원으로 사용하되, 2010년을 기점으로 기준핵연료를 .$16{\times}16$ KSFA 4.5w/o, 55GWD/MTU로 반영하는 것이 타당해 보인다.

• Nuclear Engineering and Technology
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• 제40권4호
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• pp.319-326
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• 2008

To transport process wastes efficiently from the Advanced Spent Fuel Conditioning Pyro-process Facility (ACPF) at the Korea Atomic Energy Research Institute (KAERI), a new hot cell cask has been designed based on an existing hot cell padirac transport cask, with not only a neutron absorber for improved shielding capability, but also a docking facility for an easy docking system. In the new hot cell cask, two kinds of materials have been considered as shielding materials, polyethylene and resin. To verify the transport compatibility of the waste and spent fuel for the ACPF, neutron and photon shielding calculations were performed using the MCNPX code. The source term was evaluated by the ORIGEN-ARP code system based on spent PWR fuel. From the calculation, it was found that the maximum surface dose rates of the hot cell cask with the two candidates were estimated within the limit (2 mSv/hr).

• 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.

• 방사성폐기물학회지
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• 제19권3호
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• pp.307-322
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• 2021

This study developed an analytical methodology for the mechanical integrity of spent nuclear fuel (SNF) cladding tubes under external pinch loads during transportation, with reference to the failure mode specified in the relevant guidelines. Special consideration was given to the degraded characteristics of SNF during dry storage, including oxide and hydride contents and orientations. The developed framework reflected a composite cladding model of elastic and plastic analysis approaches and correlation equations related to the mechanical parameters. The established models were employed for modeling the finite elements by coding their physical behaviors. A mechanical integrity evaluation of 14 × 14 PWR SNF was performed using this system. To ensure that the damage criteria met the applicable legal requirements, stress-strain analysis results were separated into elastic and plastic regions with the concept of strain energy, considering both normal and hypothetical accident conditions. Probabilistic procedures using Monte Carlo simulations and reliability evaluations were included. The evaluation results showed no probability of damage under the normal conditions, whereas there were small but considerably low probabilities under accident conditions. These results indicate that the proposed approach is a reliable predictor of SNF mechanical integrity.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.363-374
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• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.

• 분석과학
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• 제11권6호
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• pp.421-428
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• 1998

사용후핵연료의 화학특성을 규명하기 위하여 시료 중에 함유되어 있는 핵분열생성물 중 Zr을 분리, 정제하는 연구를 수행하였다. 우라늄과 핵분열생성물 대신 비방사성 금속이온들로 구성된 사용후핵연료 모의 용해용액을 시료로 사용하였다. 12 M HCl 용액으로 전처리한 Dowex $1{\times}8$ 음이온교환수지관에서 Ce, Nd, Cs, Rb, Ba, Sr, Ru, Rh, Pd, Ag 및 Cd을 용리시킨 후 5 M HCl 용액으로 Zr을 95% 이상 분리, 회수할 수 있었다. 용출액에 함유되어 있는 Zr 동위원소의 동중원소인 Mo을 제거하기 위하여 5 M HCl 용액으로 전처리한 Dowex $1{\times}8$ 음이온교환수지관에서 정제하였으며, 실제 PWR 사용후핵연료에 함유되어 있는 Zr 분리, 정제에 적용하여 질량분석한 결과 Mo 및 Sr에 의한 동중원소 영향이 나타나지 않았다.