• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2001년도 ICCAS
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• pp.123.3-123
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• 2001

To recycle the uranium resources in the spent nuclear fuels, all the fuel rods are extracted from the spent fuel assemblies. The remaining components of the spent fuel assembly after extracting all the rods, so called a NFBC(Non-Fuel Bearing Components), should be compacted to minimize the waste volume. To this present, KAERI (Korea Atomic Research Institute) has developed he NFBC compactor by introducing a new concept of cutting and compaction, In this paper, to achieve he maximum compaction ration of the NFBC volume while reducing compactor size, an fuzzy controller, which determines the reference force of the compactor, is proposed with using he fuzzy-inference.

• Journal of Radiation Protection and Research
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• 제9권2호
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• pp.90-96
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• 1984

장주기핵연료(長週期核燃料) 노심기법(爐心技法)에 의한 사용후핵연료(使用後核燃料)가 기존(旣存) 사용후핵연료저장시설(使用後核燃料貯藏施設)의 설계변경(設計變更)없이 동(同) 시설(施設)에 수용(受容) 가능(可能)한지를 결정(決定)하기 위하여 저장시설(貯藏施設)에서의 예상(豫想) 방사선피폭선량률(放射線被曝線量率) DLC-23/CASK (22n, 18g) 단면적(斷面積)자료(資料)와 ANISN-W 전산(電算)코드로 계산(計算)하여 설계기준치(設計基準値)와 비교(比較) 검토(檢討)하였다. 사용후핵연료내(使用後核燃料內)의 방사능량(放射能量) 및 감마선스펙트럼은 핵연료교체(核燃料交替)모델에 따라 ORIGEN 전산(電算)코드로 계산(計算)하였다. 방사선량률(放射線量率)의 계산(計算)에 있어서 저장조(貯藏槽)의 기하학적(幾何學的) 모델은 무한평판모형(無限平板模型)이며 저장(貯藏)된 사용후핵연료(使用後核燃料)의 구성물질(構成物質)과 방사선원(放射線源)은 핵연료집합체내(核燃料集合體內)에 균일(均一)하게 분포(分布)되었다고 가정(假定)하였다. 사용후핵연료저장조(使用後核燃料貯藏槽)에 저장(貯藏)된 핵연료집합체(核燃料集合體) 및 저장용수중(貯藏用水中) 방사성핵종(放射性核種)에 의한 방사선량률(放射線量率)의 계산(計算) 결과(結果)는 정상(正常) 및 사고수면시(事故水面時) 계산(計算)된 방사선량률(放射線量率)이 설계기준치(設計基準値)를 만족(滿足)시켜주는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.689-699
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• 2020

In this work, new multi-recycling options of TRU nuclides using PWR fuel assemblies comprised of MOX and FCM (Fully Ceramic Micro Encapsulated) fuels are suggested and neutronically analyzed. These options do not use a fully recycling of TRU but a partial recycling where TRUs from MOX fuels are recycled while the ones from FCM fuels are not recycled due to their high consumption rate resulted from high burnup. In particular, additional external TRU feed in MOX fuels for each cycle was considered to significantly increase the TRU consumption rate and the finally selected option is to use external TRU and enriched uranium feed as a makeup for the heavy metal consumption in MOX fuels. This hybrid external feeding of TRU and enriched uranium in MOX fuel was shown to be very effective in significantly increasing TRU consumption rate, maintaining long cycle length, and achieving negative void reactivity worth during recycling.

• Nuclear Engineering and Technology
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• 제51권3호
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• pp.884-893
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• 2019

Spent fuel racks are steel structures used in the storage of the spent fuel removed from the nuclear power reactor. Rack units are submerged in the depths of the spent fuel pool to keep the fuel cool. Their free-standing design isolates their bases from the pool floor reducing structural stresses in case of seismic event. However, these singular features complicate their seismic analysis which involves a transient dynamic response with geometrical nonlinearities and fluid-structure interactions. An accurate estimation of the response is essential to achieve a safe pool layout and a reliable structural design. An analysis methodology based on the hydrodynamic mass concept and implicit integration algorithms was developed ad-hoc, but some dispersion of results still remains. In order to validate the analysis methodology, vibration tests are carried out on a reduced scale mock-up of a 2-rack system. The two rack mockups are submerged in free-standing conditions inside a rigid pool tank loaded with fake fuel assemblies and subjected to accelerations on a unidirectional shaking table. This article compares the experimental data with the numerical outputs of a finite element model built in ANSYS Mechanical. The in-phase motion of both units is highlighted and the water coupling effect is detailed. Results show a good agreement validating the methodology.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2504-2515
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• 2023

In past years the Paul Scherrer Institute (PSI, Switzerland) and the Karlsruhe Institue of Technology (KIT, Germany)) collaborated to develop a model to account for the active role of nitrogen in the air oxidation of a Zircalloy cladding. The "PSI-KIT Nitriding Model for Zirconium based Fuel Cladding" model was implemented at PSI into PSI-MELCOR 1.8.6. In order to make a preliminary evaluation of the effect of the new model on the evolution of full-scale spent fuel pool accidents, one spent fuel pool event was analyzed using the PSI research version of PSI-MELCOR 1.8.6, which includes the nitriding model. To adapt an existing input deck for the calculations, a sensitivity study was conducted to find an optimal nodalization for the analyses. The nitriding model results were compared to those calculated with the MELCOR 1.8.6-PSI without the new nitriding model. The results demonstrate the effect of the nitriding reactions in spent fuel pool accident progression. Moreover, they confirm the impact of ZrN formation during cladding oxidation in air when the oxidation reactions lead to oxygen starvation inside the fuel assemblies. The nitriding reaction led to higher chemical heat generation during the accident and to an earlier failure of the cladding than when the effect of nitrogen reactions was not considered. It should be noted that the nitriding model, as implemented in the PSI version of MELCOR 1.8.6 has not yet been conclusively validated. Thereby the results presented in this paper should be treated as a preliminary demonstration of the capabilities of the model.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.124-129
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• 1998

A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The $^{239}$ Pu and $^{235}$ U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the fled uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%.. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel has shown that it is desirable to increase the $^{239}$ Pu and $^{235}$ U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, il is recommended to have enrichments of 0.45 and 1.00 wt% for $^{239}$ Pu and $^{235}$ U, respectively.

• Nuclear Engineering and Technology
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• 제14권2호
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• pp.86-91
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• 1982

1977년 집합체 cell 중심간의 간격을 53.34cm(162 연료 집합체 저장)에서 36cm(562 연료 집합체저장)로 줄임으로 고리 1호기 부속 기사용 연료 저장 용량을 화장하였다. 확장된 저장 시설에 대하여 Monte Carlo방법을 이용한 KENO-IV코드로 Core Performance Branch에서 제시한 비정상적인 냉각수 밀도 조건에 따라 유효증배계수를 구하였다. KENO-IV결과 밀도 0.1143g/cm에서 최대유효 증배계수 0.9958$\pm$0.0048을 얻었고, 이 값은 US NRC 기준과 CPB기준인 0.98을 초과하므로 새로운 집합체 cell중심간의 간격을 구하였다. 이 과정은 KENO-IV보다 보수적인 결과를 나타내는 확산 코드 KIDD를 이용하여 cell중심간의 간격에 따른 상관 연구로부터 새로운 cell중심간의 간격을 얻었다. 이로부터 현재의 집합체 cell중심 간의 간격 36cm를 43cm 이상으로 늘려야 비정상적인 냉각수 밀도의 감소로 인한 사고의 경우에도 안전함을 알았다.

• 한국전산유체공학회지
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• 제21권4호
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• pp.33-39
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• 2016

Generally, thermal analysis of spent fuel storage cask has been conducted using the porous media and effective thermal conductivity model to simplify the structural complexity of spent fuel assemblies. As the fuel assembly is composed of two regions; active fuel region corresponding to UO2 pellets and unactive fuel region corresponding to the top and bottom nozzle, the heat transfer performance can be influenced depending on porous media application at these regions. In this study, numerical analysis on concrete storage cask of spent fuel was performed to investigate heat transfer effects for two cases; one was porous media application only to active fuel region(case 1) and the other one was porous media to whole length of fuel assembly(case 2). Using computational fluid dynamics code, the three dimensional, 1/4 symmetry model was constructed. For two cases, maximum temperatures for each component were evaluated below the allowable limits. For the case 1, maximum temperatures for fuel cladding, neutron absorber and baskets inside the canister were slightly higher than those for the case 2. In particular, even though the helium flows with low velocity due to buoyant forces occurred at the top and bottom of unactive fuel region, treating only active fuel region as the porous media was ineffective in respect of the heat removal performance of concrete storage cask, implying a conservative result.

• Nuclear Engineering and Technology
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• 제46권1호
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• pp.73-80
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• 2014

A concrete cask is an option for spent nuclear fuel interim storage. A concrete cask usually consists of a metallic canister which confines the spent nuclear fuel assemblies and a concrete overpack. When the overpack undergoes a missile impact, which might be caused by a tornado or an aircraft crash, it should sustain an acceptable level of structural integrity so that its radiation shielding capability and the retrievability of the canister are maintained. A missile impact against a concrete overpack produces two damage modes, local damage and global damage. In conventional approaches [1], those two damage modes are decoupled and evaluated separately. The local damage of concrete is usually evaluated by empirical formulas, while the global damage is evaluated by finite element analysis. However, this decoupled approach may lead to a very conservative estimation of both damages. In this research, finite element analysis with material failure models and element erosion is applied to the evaluation of local and global damage of concrete overpacks under high speed missile impacts. Two types of concrete overpacks with different configurations are considered. The numerical simulation results are compared with test results, and it is shown that the finite element analysis predicts both local and global damage qualitatively well, but the quantitative accuracy of the results are highly dependent on the fine-tuning of material and failure parameters.

• 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.