• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 추계학술발표회요약집
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• pp.251-251
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• 1998

• Nuclear Engineering and Technology
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• 제18권2호
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• pp.117-121
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• 1986

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.35-40
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• 1995

Considering the higher discharge burnup, lower channel refuelling rate, lower linear element rating(LER), lower coolant void reactivity and axial power shape, CANFLEX-RU fuel bundle is optimized for CANDU-6 by grading the fissile composition in the ring-wise of the bundle and by applying fuel management scheme appropriately. The fissile composition of the fuel bundle is graded as the recovered uranium (0.9 w/o U-235) in the outer and intermediate elements, depleted Uranium (0.2 w/o U-235) in the center element, natural uranium (0.71 w/o U-235) in the inner elements. Enrichment is not required for these fuel. The fissile composition is optimized by lattice calculation and by time-averaged reactor simulation. CANFLEX-RU optimized for CANDU-6 resulted to be the 15% lower channel refuelling rate, acceptable axial power profile and power envelope, 70% higher discharge burnup, 15% lower LER and not increase coolant void reactivity compared with the 37-element natural uranium bundle for CANDU-6.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.907-915
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• 2024

Various core designs with multi-batch fuel management (FM) are proposed and optimized for an innovative small modular reactor (iSMR), focusing on enhancing the inherent safety and neutronic performance. To achieve soluble-boron-free (SBF) operation, cylindrical centrally-shielded burnable absorbers (CSBAs) are utilized, reducing the burnup reactivity swing in both two- and three-batch FMs. All 69 fuel assemblies (FAs) are loaded with 2-cylindrical CSBA. Furthermore, the neutron economy is improved by deploying a truly-optimized PWR (TOP) lattice with a smaller fuel radius, optimized for neutron moderation under the SBF condition. The fuel shuffling and CSBA loading patterns are proposed for both 2- and 3-batch FM with the aim to lower the core leakage and achieve favorable power profiles. Numerical results show that both FM configurations achieve a small reactivity swing of about 1000 pcm and the power distributions are within the design criteria. The average discharge burnup in the two-batch core is comparable to three-batch commercial PWR like APR-1400. The proposed checker-board CR pattern with extended fingers effectively assures cold shutdown in the two-batch FM scenario, while in the three-batch FM, three N-1 scenarios are failed. The whole evaluation process is conducted using Monte Carlo Serpent 2 code in conjunction with ENDF/B-VII.1 nuclear library.

• 방사선산업학회지
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• 제18권3호
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• pp.249-253
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• 2024

For the safe management and disposal of spent fuel, it is essential to accurately determine the quantities of nuclides present within the spent fuel. In this study, a Gamma/Neutron detection system was developed as a part of basic research to measure the burnup profile of spent fuel, and a performance was evaluated using major nuclides. The prototype of the Gamma/Neutron detection system consisted of a CZT sensor and a ³He chamber. For quantitative evaluation, studies were conducted using calibrated ¹³⁷Cs, ¹³⁴Cs, ¹⁵⁴Eu and ²⁵²Cf sources. In the performance evaluation, a field applicability was verified by analyzing the detection characteristics according to the nuclide.

• 방사성폐기물학회지
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• 제13권2호
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• pp.141-154
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• 2015

경수로 사용후핵연료 수송/저장용기의 핵임계 해석은 사용후핵연료내의 악티나이드핵종 및 핵분열생성물 함유량에 대한 불확실성을 이유로 신연료로 가정된 가상의 연료를 선정하여 평가해오고 있다. 그러나 이러한 평가방법은 용기 설계 시 과도한 임계여유도를 유도하여 경제적 손실을 유발할 수 있는 단점이있다. 이와 같은 단점을 극복하기 위하여 최근 연소도이득효과(burnup credit, BUC)를 반영한 수송저장용기의 설계 및 상용화를 위한 연구가 추진되었다. 이에 본 연구에서는 한국원자력환경공단에서 개발중인 금속겸용용기를 대상으로 연소도 이득효과적용 시 핵임계 안전성(criticality safety)에 영향을 미칠 것으로 예상되는 '노심 운전인자', '축방향 연소도 분포', '오장전 사고상황'에 대하여 핵임계 평가를 수행하였다. 그 결과 노심운전인자 중 저농축, 고연소도일 때 비출력에 따른 핵임계 변화가 크게 평가되었으며, 고연소도 사용후핵연료에서 End effect가 양의 값을 나타내었다. 특히 오장전에 의한 유효증배계수는 최대 0.18467증가하였으므로, 연소도이득효과를 적용 할 경우 필수고려사항임을 확인하였다. 본 연구결과는 국내모델(금속겸용용기)의 연소도 이득효과 적용기술 개발 및 사용 후핵연료 장전 시 일어날 수 있는 오장전 사고를 방지하기 위한 운영절차 개발에 참고자료로 활용될 수 있다.

• Nuclear Engineering and Technology
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• 제24권1호
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• pp.1-13
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• 1992

ORIGEN2 전산코드를 위한 연구로용 1군 단면적 데이타베이스를 개발하였다. 여기에는 EN-DF/B-lV와 -V가 기 본 핵 자료로 사용되었고 이들은 NJOY 코드시스템을 사용하여 69군으로 처리되었다. 1군 축약을 위한 가중함수는 핵연료의 연소에 따른 KMRR의 중성자 스펙트럼을 WIMS-KAERI코드로 계산하여 사용하였다. 새로 개발된 데이타베이스는 KMRR핵연료의 연소에 따른 악티나이드 생성량 평가를 통해 상세 다군 수송계산 결과와 잘 일치함이 입증되었다. 그리고 새로운 데이타베이스를 이용하여 KMRR의 사용후 핵연료 특성을 분석하였다.

• Nuclear Engineering and Technology
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• 제25권2호
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• pp.259-264
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• 1993

조사후핵연료의 연소도측정에 1-octanesulfonate 를 양이온 교환체로 사용하고 $\alpha$-hydroxyisobutyric acid를 용리액으로 사용하는 동적계의 이온크로마토그래피를 적용하였다 Pu, U 및 Nd의 최적 분리조건을 찾기위해 분리조건들을 변화하였다. 이들 원소들을 $\alpha$-hydroxyisobutyric acid 용리액을 0.05 M과 0.40 M을 혼합시키는 기울기용리법으로 개별 분리한후 분취하여 동위원소희석 질량분석법으로 각각 정량하였다. 본 방법에 의래 구한 연소도 값을 기존의 음이온교환수지법에 의한 값과 비교한 결과 3.5 %차이 이내에서 두 값이 서로 일치하였다.

• Nuclear Engineering and Technology
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• 제41권1호
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• pp.99-106
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• 2009

The contents of transuranic elements ($^{237}Np$, $^{238}Pu$, $^{239}Pu$, $^{240}Pu$, $^{241}Am$, $^{244}Cm$, and $^{242}Cm$) in high-burnup spent fuel samples ($35.6{\sim}53.9\;GWd/MtU$) were determined by alpha spectrometry. Anion exchange chromatography and diethylhexyl phosphoric acid extraction chromatography were applied for the separation of these elements from the uranium matrix. The measured values of the nuclides were compared with ORIGEN-2 calculations. For plutonium, the measurements were higher than the calculations by about $2.6{\sim}32.7%$ on average according to each isotope, and those for americium and curium were also higher by about $35.9{\sim}63.1%$. However, for $^{237}Np$, the measurements were lower by about 52% on average for the samples.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3133-3150
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• 2021

This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.