• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.11a
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• pp.335-336
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• 2009

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2011.05a
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• pp.67-68
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• 2011

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.197-202
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• 1997

ORIGEN2코드의 검증계산을 통해 PWR 사용 후 핵연료 조성핵종의 핵종량에 대한 핵임계측면에서 보수성을 가지는 안전인자를 산출하였고, MCNP코드의 검증계산으로 95/95 신뢰구간에서의 계산오차를 구하였다. 이를 바탕으로 직경이 1.2567cm이고 길이가 380.5cm인 196 개 금속봉을 장전한 캐니스터 ( 금속저장체 )가 x-y 방향으로 무한히 배열된 경우에 대해 캐니스터의 두께, 간격 및 외부의 공기중 수분농도에 따른 핵임계 안전해석을 수행하였다. 그 결과, 캐니스터의 두께가 7mm일 때 공기중 수분농도가 0.30 g/㎤이고 캐니스터간의 간격이 6.0cm인 경우의 최종핵 임계도값은 0.94130로서 최대허용핵임계값 (0.942)보다 적은 값을 보였다.

• Journal of Radiation Protection and Research
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• v.14 no.2
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• pp.30-36
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• 1989

A FISSLIB, 51 group nuclear data set for 22 product nuclides, which are present in spent fuel and significantly affect the criticality of spent fuel, was generated from ENDF/B-IV using XLACS-II. The FISSLIB is ready to be used together with a data set generated from DLC-43/CSRL using AMPX system. The reliability of FISSLIB was verified by comparison with the data reported in BNL-325. Using FISSLIB, the criticality of KORI-1 spent fuel rod arranged infinitely was analyzed, and it was found that $K_{eff}$ of the spent fuel including fission products was lower by 9-14% than that calculated without fission products.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.339-346
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• 2019

Based on the $8^{th}$ Basic Plan for Electric Power Demand and Supply, an estimation has been made for inventories and characteristics of spent fuel (SF) to be generated from existing and planned nuclear power plants. The characteristics under consideration in this study are dimensions, fuel array, $^{235}U$ enrichment, discharge burnup, and cooling time for each fuel assembly. These are essentially needed for designing a disposal facility for SFs. It appears that the anticipated quantity by the end of 2082 is about 62,500 assemblies for PWR SFs. The inventories of Westinghouse-type and Korean-type SFs were revealed to be 60% and 40%, respectively as of the end of 2018. The proportion of SFs with initial $^{235}U$ enrichment below 4.5 weight percent (wt%) was shown to be approximately 90% in total as of the end of 2018. As of 2077, more than 97% of SFs generated from Westinghouse-type nuclear reactors were shown to have cooling time of over 50 years. As of 2125, more than 98% of SFs generated from Korean-type nuclear reactors were shown to have cooling time of over 45 years. Based on these results, for the efficient design of a disposal system, it is reasonable to adopt two types of reference spent fuel. SF of KSFA with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 50 years was determined as reference fuel for Westinghouse-type SFs; SF of PLUS7 with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 45 years was determined as reference fuel for Korean-type SFs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.4
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• pp.301-307
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• 2005

The morphology of the high burnup $UO_2$ spent fuel, which was oxidized and annealed in a PIA (Post Irradiation Annealing) apparatus, has been observed. The high burnup fuel irradiated in Ulchin Unit 2, average rod burnup 57,000 MWd/tU, was transported to the KAERI's PIEF. The test specimen was used with about 200 mg of the spent $UO_2$ fuel fragment of the local burnup 65,000 MWd/tU. This specimen was annealed at $1400^{\circ}C$ for 4hrs after the oxidation for 3hrs to grain boundary using the PIA apparatus in a hot-cell. In order to oxidize the grain boundary, the oxidation temperature increased up to $500^{\circ}C$ and held for 3hrs in the mixed gas (60 ml He and 100 ml STD-air) atmosphere. The amount of 85Kr during the whole test process was measured to know the fission gas release behavior using the online system of a beta counter and a gamma counter. The detailed micro-structure was observed by a SEM to confirm the change of the fuel morphology after this test. As the annealing temperature increased, the fission products were observed to move to the grain surface and grain boundary of the $UO_2$ matrix. This specimen was re-structured through the reduction process, and the grain sizes were distributed from 5 to $10\;{\mu}m$.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.262-267
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• 1997

금속형 사용후핵연료 관리모형으로 직경 3 cm, 길이 248.5cm인 금속봉을 19개 장전한 캐니스터가 x-y 방향으로 무한 격자배열된 경우에 대해 캐니스터의 두께, 간격 및 외부의 공기중 수분농도의 변화에 따른 핵임계도 분석을 수행하였다. 미임계한계치(k$_{eff}$=0.95) 근방에서 최대 핵임계도를 나타내는 각 인자값을 구하고, 미임계 상태를 유지하는 조건을 제시했다. 그 결과, 캐니스터의 두께가 7mm인 경우의 최대 핵임계도 값은 0.94401 $\pm$ 0.00050으로서, 공기중 수분농도가 0.34 g/㎤이고 간격이 4.8 cm인 경우에 나타났다. 8 mm인 경우의 최대 핵임계도 값은 0.91182 $\pm$0.00050이며, 캐니스터간의 간격이 4.4cm이고 공기중 수분농도가 0.35 g/㎤일 때 나타났다. 8 mm 캐니스터의 금속저장체 저장은 미임계 상태를 유지할 것으로 추정되었다.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.12
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• pp.815-824
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• 2016

The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermal-hydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermal-hydraulic technology and the commercialization.

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.332-340
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• 1989

Destructive examination of 14$\times$14 PWR fuel burned for 3 cycles are carried out to investigate the in-reactor fuel performance. The results obtained are as follows; 1) Grain growth is not occured at the fuel center. 2) Fuel density is decreased as the turnup increase, the density is down to 94.4% TD at burnup of 36,000 MWD/MTU. 3) Average thickness of oxide layer on cladding is less than 10 $\mu$m in the lower and middle section, while it is rapidly increased above 20 $\mu$m in the upper section. 4) The rate of hydride production in the cladding is large in the upper section than lower section and is related to the production of oxide on the cladding

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2011.10a
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• pp.161-162
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• 2011

유압동력사용 전제하에, 기계식탈피복을 고려하지 않을 때는 전단방식이 가장 유리함을 알 수 있다. 절단방식은 전단방식에 비해서 낮은 생산성이 단점이나, 높은 원형도의 연료봉 절단면이 요구되거나, 비산에 의한 칩 분리, 쿨링(cooling) 장치를 보완하면 절단방식이 유리하다. 또한 수평식 슬릿장치는 커팅 블레이드의 낮은 내구성으로 생산성이 낮은 것이 단점이나 내구성이 강한 공구를 사용하여 처리 속도를 향상한다는 전제에서 실험적 검증의 확보, 그리고 별도의 복잡한 펠릿/헐 분리장치를 보완하면 수평식 슬릿 방식이 유리하다.