• Nuclear Engineering and Technology
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• 제56권1호
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• pp.167-179
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• 2024

This study explores the feasibility of employing (U, Th)-based accident tolerant fuels (ATFs), specifically (_0.8UO₂, _0.2ThO₂), (_0.8UN, _0.2ThN), and (_0.8UC, _0.2ThC). The investigation assesses the overall performance of these proposed fuel materials in comparison to the conventional UO₂, focusing on deep neutronic and thermal-hydraulic (Th) analyses. Neutronic analysis utilized the MCNPX code, while COMSOL Multiphysics was employed for thermal-hydraulic analysis. The primary objective of this research is to overcome the limitations associated with traditional UO₂ fuel by exploring alternative fuel materials that offer advantages in terms of abundance and potential improvements in performance and safety. Given the limited abundance of UO₂, long-term sustainable nuclear energy production faces challenges. From a neutronic standpoint, the U-Th based fuels demonstrated remarkable fuel cycle lengths, except (_0.8UN, _0.2ThN), which exhibited the minimum fuel cycle length and, consequently, the lowest fuel burn-up. Regarding thermal-hydraulic performance, (_0.8UN, _0.2ThN) exhibited outstanding performance with significant margins against fuel melting compared to the other materials. Overall, when considering the integrated performance, the most favourable results were obtained with the use of the (_0.8UC, _0.2ThC) fuel configurations. This study contributes valuable insights into the potential benefits of (U, Th)-based ATFs as a promising avenue for enhanced nuclear fuel performance.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.309-316
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• 2005

The contents of a new book currently in preparation are described. The dearth of books in the field of nuclear materials has left both students in nuclear materials classes and professionals in the same field without a resource for the broad fundamentals of this important sub-discipline of nuclear engineering. The new book is devoted entirely to materials problems in the core of light-water reactors, from the pressure vessel into the fuel. Key topics deal with the $UO_2$ fuel, Zircaloy cladding, stainless steel, and of course, water. The restriction to LWR materials does not mean a short monograph; the enormous quantity of experimental and theoretical work over the past 50 years on these materials presents a challenge of culling the most important features and explaining them in the simplest quantitative fashion. Moreover, LWRs will probably be the sole instrument of the return of nuclear energy in electric power production for the next decade or so. By that time, a new book will be needed.

• Nuclear Engineering and Technology
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• 제40권7호
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• pp.581-592
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• 2008

In this study, a conceptual pyroprocess flowsheet has been devised by combining several dry-type unit processes; its applicability as an alternative fuel cycle technology was analyzed. A key point in the evaluation of its applicability to the fuel cycle was the recovery yield of fissile materials from spent fuels as well as the proliferation resistance of the process. The recovery yields of uranium and transuranic elements (TRU) were obtained from a material balance for every unit process composing the whole pyroprocess. The material balances for several elemental groups of interest such as uranium, TRU, rare earth, gaseous fission products, and heat generating elements were calculated on the basis of the knowledge base that is available from domestic and foreign experimental results or technical information presented in open literature. The calculated result of the material balance revealed that uranium and TRU could be recovered at 98.0% and 97.0%, respectively, from a typical PWR spent fuel. Furthermore, the anticipated TRU product was found to emit a non-negligible level of $\gamma$-ray and a significantly higher level of neutrons compared to that of a typical plutonium product obtained from the PUREX process. The results indicate that the product from this conceptual pyroprocessing should be handled in a shielded cell and that this will contribute favorably to retaining proliferation resistance.

• 방사성폐기물학회지
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• 제17권4호
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• pp.405-418
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• 2019

국내 원자력발전소에서 발생하는 사용후핵연료의 제원 및 방출시점 등 특성과 현재의 고준위 방사성폐기물 기본계획에 근거한 처분시나리오를 도출하여 기존 심층 처분시스템을 바탕으로 처분효율과 경제성을 향상시킨 개선된 처분시스템을 제안하였다. 이를 위하여 국내 원자력발전소에서 발생하는 사용후핵연료의 길이에 따라 2종류의 처분용기 개념을 도출하고, 사용후핵연료 발생 년도와 현재의 기본계획에 근거한 처분 시나리오 설정에 따른 처분시점에서의 냉각기간을 고려하여 처분용기내 수용 가능한 붕괴열 량을 결정하였다. 그리고 2종류의 처분용기에 대한 처분시스템과 결정된 붕괴열을 바탕으로 열적 안정성 분석을 통하여 제안된 처분시스템의 설계요건에 대한 적합성 여부를 확인하고, 처분효율을 평가하였다. 개선된 처분시스템은 기존 처분시스템에 비하여 처분면적은 약 20% 감소되고 처분밀도는 약 20% 향상됨을 확인하였고, 처분용기와 완충재 재료도 상당량 절감됨을 확인하였다. 본 연구의 결과는 향후 사용후핵연료 관리정책 수립 및 실제 사업을 위한 처분시스템 설계를 위한 자료로 활용될 수 있다.

• 대한화학회지
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• 제45권4호
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• pp.304-311
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• 2001

경수로 사용후 핵 연료에 미량 함유되어 있는 핵분열생성물을 유도 결합 플라스마 원자방출분광법(ICP-AES)으로 분석하기 위하여 우라늄으로부터 학분열생성물을 추출 크로마토그래피로 분리, 회수하는 방법을 검토하였다. 우라늄 분리 분야에서 잘 알려져 있는 tri-n-butyl phosphate(TBP)를 추출제로 사용하여 몇 가지 Amberlite XAD 다공성 수지들에 대한 침윤능을 비교한 후 TPB침윤양이 가장 큰 Amberlite XAD-16을 지지체로 선택하였다. 사용후핵연료 용해용액과 화학조성이 유사한 모의 사용후핵연료 용해용액을 사용하여 TBP 침윤수지에 대한 핵분열생성물 원소들의 흡착거동을 조사하고, 분리에 미치는 여러 변수들을 최적화 하였다. Pd 및 Ru을 제외한 대부분의 핵분열생성물 원소들을 정밀도 3.1% 이하의 범위에서 정량적으로 회수할 수 있었다.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.411-415
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• 2018

Nuclear fusion may be among the strongest sustainable ways to replace fossil fuels because it does not contribute to acid rain or global warming. In this context, activated cobalt materials in corrosion products for fusion energy are significant in determination of dose levels during maintenance after a coolant leak in a nuclear fusion reactor. Therefore, cross-section studies on cobalt material are very important for fusion reactor design. In this article, the excitation functions of some nuclear reaction channels induced by proton particles on $^{59}Co$ structural material were predicted using different models. The nuclear level densities were calculated using different choices of available level density models in ALICE/ASH code. Finally, the newly calculated cross sections for the investigated nuclear reactions are compared with the experimental values and TENDL data based on TALYS nuclear code.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4393-4411
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• 2022

In recent years, beryllium oxide has been widely utilized in multiple compact nuclear reactors as the neutron moderator, the neutron reflector or the matrix material with dispersed nuclear fuels due to its prominent properties. In the past 70 years, beryllium oxide has been studied extensively, but rarely been systematically organized. This article provides a systematic review of the application history, thermal properties, mechanical properties, corrosion behavior and fabrication methods of beryllium oxide. Data from previous literature are extracted and sorted out, and all of these original data are attached as the supplementary material, so that subsequent researchers can utilize this paper as a database for beryllium oxide research in reactor design or simulation analysis, etc. In addition, this review article also attempts to point out the insufficiency of research on beryllium oxide, and the possible key research areas about beryllium oxide in the future.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.597-609
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• 2020

Accident Tolerant Fuels have been widely studied since the Fukushima-Daiichi accident in 2011 as one of the options on how to further enhance the safety of nuclear power plants. Deposition of protective coatings on nuclear fuel claddings has been considered as a near-term concept that will reduce the high-temperature oxidation rate and enhance accidental tolerance of the cladding while providing additional benefits during normal operation and transients. This study focuses on experimental testing of Zr-based alloys coated with Cr-based coatings using Physical Vapour Deposition. The results of long-term corrosion tests, as well as tests simulating postulated accidents, are presented. Zr-1%Nb alloy used as nuclear fuel cladding serves as a substrate and Cr, CrN, Cr_xN_y layers are deposited by unbalanced magnetron sputtering and reactive magnetron sputtering. The deposition procedures are optimized in order to improve coating properties. Coated as well as reference uncoated samples were experimentally tested. The presented results include standard long-term corrosion tests at 360℃ in WWER water chemistry, burst (creep) tests and mainly single and double-sided high-temperature steam oxidation tests between 1000 and 1400℃ related to postulated Loss-of-coolant accident and Design extension conditions. Coated and reference samples were characterized pre- and post-testing using mechanical testing (microhardness, ring compression test), Thermal Evolved Gas Analysis analysis (hydrogen, oxygen concentration), optical microscopy, scanning electron microscopy (EDS, WDS, EBSD) and X-ray diffraction.

• Nuclear Engineering and Technology
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• 제34권3호
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• pp.187-201
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• 2002

The KN-12 cask is designed to transport 12 PWR spent nuclear fuels and to comply with the requirements of Korea Atomic Energy Act, IAEA Safety Standards Series No.57-1 and US 10 CFR Part 71 for a Type B(U)F package. It provides containment, radiation shielding, structural integrity, criticality control and heat removal for normal transport and hypothetical accident conditions. W.H 14$\times$14, 16$\times$16 and 17$\times$17 fuel assemblies with maximum allowable initial enrichment of 5.0 wt.%, maximum average burn-up of 50,000 MWD/MTU and minimum cooling time of 7 years being used in Korea will be loaded and subsequently transported under dry and wet conditions. A forged cylindrical cask body which constitutes the containment vessel is closed by a cask lid. Polyethylene rods for neutron shielding are arranged in two rows of longitudinal bore holes in the cask body wall. A fuel basket to accommodate up to 12 PWR fuel assemblies provides support of the fuels, control of criticality and a path to dissipate heat. Impact limiters to absorb the impact energy under the hypothetical accident conditions are attacked at the top and at the bottom side of the cask during transport. Handling weight loaded with water is 74.8 tons and transport weight loaded with water with the impact limiters is 84.3 tons. The cask will be licensed in accordance with Korea Atomic Energy Act 3nd fabricated in Korea in accordance with ASME B&PV Code Section 111, Division 3.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.21-26
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• 1997

Recently, thorium is receiving increasing attention as an important fertile material for the expanding nuclear power programs around the world. The superior nuclear and physical properties of thorium-based fuels could lead to very low fuel cycle cost and make thorium reactors economically attractive. In addition, the use of thorium in reactors would permit more efficient utilization of low cost uranium reserves and reduction nuclear wastes. In this work, the nuclear characteristics of a new type thorium fueled reactor (Radkowsky Thorium Reactor) consisting seed-blanket assemblies are addressed and compared with those typical assemblies of a PWR (CE type). Also, an assessment on several advantages of thorium fueled reactors is provided. All these results are based on the HELIOS code calculation.