• Nuclear Engineering and Technology
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• 제40권1호
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• pp.9-20
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• 2008

Since plutonium accounts for 40-50% of the power produced by uranium fuels, spent fuel contains only residual plutonium. Management of this plutonium is one of the aspects influencing the choice of a fuel cycle back-end option: reprocessing, direct disposal or wait-and-see. Different grades and qualities of plutonium exist depending from their specific generation conditions; all are valuable fissile material. Safeguard authorities watch the inventories of civil plutonium, but access to those data is restricted. Independent evaluations have led to an estimated current inventory of 220t plutonium in total (spent fuel, separated civil plutonium and military plutonium). If used as MOX fuel, it would be sufficient to feed all the PWRs and BWRs worldwide during 7 years or to deploy a FBR park corresponding to 150% of today' s installed nuclear capacity worldwide, which could then be exploited for centuries with the current stockpile of depleted and spent uranium. The energy potential of plutonium deteriorates with storage time of spent fuel and of separated plutonium, due to the decay of $^{241}Pu$, the best fissile isotope, into americium, a neutron absorber. The loss of fissile value of plutonium is more pronounced for usage in LWRs than in FBR. However, keeping the current plutonium inventory for an expected future deployment of FBRs is counterproductive. Recycling plutonium reduce the required volume for final disposal in an underground repository and the cost of final disposal. However, the benefits of utilizing an energy resource and of reducing final disposal liabilities are not the only aspects that determine the choice of a back-end policy.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 Proceedings of The 6th korea-china joint workshop on nuclear waste management
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• pp.156-164
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• 2005

The challenges China is facing in energy security are briefly discussed. Then, the development of nuclear power in China in the first half of 21 st century is envisioned, and it is expected that Generation-3 PWR nuclear power plants (NPPs) would be the leading units of nuclear power in the coming $30\~40$ years. As part of the nuclear power program, the R&D work on nuclear fuel cycle is generally proposed.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.427-438
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• 2013

In this paper we discuss spent fuel management options in the Republic of Korea (ROK) from two interrelated perspectives: Centralized dry cask storage and spent fuel pyroprocessing and burning in sodium fast reactors (SFRs). We argue that the ROK will run out of space for at-reactors spent fuel storage by about the year 2030 and will thus need to transition centralized dry cask storage. Pyroprocessing plant capacity, even if approved and successfully licensed and constructed by that time, will not suffice to handle all the spent fuel discharged annually. Hence centralized dry cask storage will be required even if the pyroprocessing option is successfully developed by 2030. Pyroprocessing is but an enabling technology on the path leading to fissile material recycling and burning in future SFRs. In this regard we discuss two SFR options under development in the U.S.: the Super Prism and the Travelling Wave Reactor (TWR). We note that the U.S. is further along in reactor development than the ROK. The ROK though has acquired more experience, recently in investigating fuel recycling options for SFRs. We thus call for two complementary joint R&D project to be conducted by U.S. and ROK scientists. One leading to the development of a demonstration centralized away-fromreactors spent fuel storage facility. The other involve further R&D on a combined SFR-fuel cycle complex based on the reactor and fuel cycle options discussed in the paper.

• Nuclear Engineering and Technology
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• 제49권8호
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• pp.1727-1732
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• 2017

Pyroprocessing is an alternative method of reprocessing spent fuel, usually involving the dissolving spent fuel in a molten salt media. The National Nuclear Laboratory designed, built, and commissioned a molten salt dynamics rig to investigate the transfer characteristics of molten lithium chloride-potassium chloride eutectic salt. The efficacy and flow characteristics of a high-temperature centrifugal pump and argon gas lift were obtained for pumping the molten salt at temperatures up to $500^{\circ}C$. The rig design proved suitable on an industrial scale and transfer methods appropriate for use in future molten salt systems. Corrosion within the rig was managed, and melting techniques were optimized to reduce stresses on the rig. The results obtained improve the understanding of molten salt transport dynamics, materials, and engineering design issues and support the industrialization of molten salts pyroprocessing.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.727-732
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• 2016

A simple and rapid spectrophotometric method has been developed to enable the determination of plutonium concentration in an irradiated fuel solution in the presence of all fission products. An excess of ceric ammonium nitrate solution was employed to oxidize all the valence states of plutonium to +6 oxidation state. Interference due to the presence of fission products such as ruthenium and zirconium, and corrosion products such as iron in the envisaged concentration range, as in the irradiated fuel solution, was studied in the determination of plutonium concentration by the direct spectrophotometric method. The stability of plutonium in +6 oxidation state was monitored under experimental conditions as a function of time. Results obtained are reproducible, and this method is applicable to radioactive samples resulting before the solvent extraction process during the reprocessing of fast reactor spent fuel. An analysis of the concentration of plutonium shows a relative standard deviation of <1.2% in standard as well as in simulated conditions. This reflects the fast reactor fuel composition with respect to uranium, plutonium, fission products such as ruthenium and zirconium, and corrosion products such as iron.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.317-326
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• 2005

The burning-out of excess plutonium from the reprocessing of spent nuclear fuel and from the dismantlement of nuclear weapons is recently emphasized due to the difficulties in securing the final repository for the spent fuel and the necessity to consume the ex-weapons plutonium. An irradiation test in the Halden reactor was launched by the OECD Halden Reactor Project (HRP) to investigate the in-pile behavior of plutonium-embedded fuel as a form of mixed oxide (MOX) and of inert matrix fuel (IMF). The first cycle of irradiation was successfully accomplished with good integrity of test fuel rods and without any undesirable fault of instrumentations. The test results revealed that the MOX fuel is more stable under irradiation environments than IMF. In addition, MOX fuel shows lower thermal resistance due to its better thermal conductivity than IMF. The on-line measured in-pile performance data of attrition milled MOX fuel are used in the analysis of the in-pile performance of the fuel with the fuel performance code, COSMOS. The COSMOS code has been developed for the analysis of MOX fuel as well as $UO_2$ fuel up to high burnup and showed good capability to analyze the in-reactor behavior of MOX fuel even with different instrumentation.

• Nuclear Engineering and Technology
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• 제38권4호
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• pp.359-374
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• 2006

Since 1991, Korea, Canada and United States have performed the direct use of spent pressurized water reactor (PWR) fuel in the Canada deuterium uranium (CANDU) reactors (DUPIC) fuel development project. Unlike the Tandem fuel cycle, which requires a wet reprocessing, the DUPIC fuel technology can directly refabricate CANDU fuels from the PWR spent fuel and, therefore, is recognized as a highly proliferation-resistant fuel cycle technology, which can be adopted even in non-proliferation treaty countries. The Korea Atomic Energy Research Institute (KAERI) has fabricated DUPIC fuel elements in a laboratory-scale remote fuel fabrication facility. KAERI has demonstrated the fuel performance in the research reactor, and has confirmed the operational feasibility and safety of a CANDU reactor loaded with the DUPIC fuel using conventional design and analysis tools, which will be the foundation of the future practical and commercial uses of DUPIC fuel.

• Journal of Radiation Protection and Research
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• 제21권1호
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• pp.59-72
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• 1996

원자력의 개발과 지속적인 이용은 방사성폐기물과 사용후핵연료의 발생을 야기시키며, 발생된 사용후핵연료의 안전하며 효율적인 관리는 1990년 초부터 중요하며 민감한 국제사회의 이슈가 되고 있다. 특히 구 소련의 해체를 포함한 최근 중부유럽의 정치적인 변화에 따른 안전한 사용후핵연료관리 문제와 현재 원자력산업이 직면하고 있는 어려움 등이 국제정치의 관점에서 그 의미를 더하고 있다. 따라서 국가별로 현재 검토 및 시행되고 있는 사용후핵연료 관리에 대한 현황을 정리하였다. 즉 국제원자력기구에서 개최하고 있는 사용후핵연료관리회의에서 발표된 나라별 관리정책에 대한 현황 및 기타 기술자료에서 발표된 최신의 사용후핵연료관리 실례에 대한 내용을 정리하였다.

• 방사성폐기물학회지
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• 제20권1호
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• pp.23-32
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• 2022

Pyroprocessing is a promising technology for managing spent nuclear fuel. The nuclear material accounting of feed material is a challenging issue in safeguarding pyroprocessing facilities. The input material in pyroprocessing is in a solid-state, unlike the solution state in an input accountability tank used in conventional wet-type reprocessing. To reduce the uncertainty of the input material accounting, a double-stage homogenization process is proposed in considering the process throughput, remote controllability, and remote maintenance of an engineering-scale pyroprocessing facility. This study tests two types of mixing equipment in the proposed double-stage homogenization process using surrogate materials. The expected heterogeneity and accounting uncertainty of Pu are calculated based on the surrogate test results. The heterogeneity of Pu was 0.584% obtained from Pressurized Water Reactor (PWR) spent fuel of 59 WGd/tU when the relative standard deviation of the mass ratio, tested from the surrogate powder, is 1%. The uncertainty of the Pu accounting can be lower than 1% when the uncertainty of the spent fuel mass charged into the first mixers is 2%, and the uncertainty of the first sampling mass is 5%.

• 방사성폐기물학회지
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• 제4권3호
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• pp.285-299
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• 2006

2005년말 현재, 전세계 32개국에서 443기의 원자력발전소가 운영되고 있다. 현재 전체발전량은 약 3,000 TWh이며 전세계 전력공급의 약 16 퍼센트를 차지하고 있다. 2004년말 사용후핵 연료는 전세계 원전의 발전용량 368 GWe에서 매년 11,000 tHM 정도 발생되고 있으며 현재 운영중인 대부분의 원전이 가동정지가 예상되는 2020년에는 445,000 tHM까지 예상되고 있다. 이러한 관점에서, 사용후핵 연료 관리는 전체 IAEA 회원국에게는 그들이 취하고 있는 후행핵 연료주기 정책과 전략에 관계없이 국제협력 등을 통해 가까운 장래에 시급히 그리고 반드시 해결해야 할 필수 사안임이 분명하다. 지난 2006년 5월 15일부터 2주간 제2차 방사성폐기물안전협약 체약국회의가 오스트리아 IAEA본부에서 개최되었다. 동 회의에서 사용후 핵연료에 대한 국가 정책 및 전략, 그리고 그들의 현황, 향후 전망, 정책에 일차적으로 고려한 인자와 이행내용 등이 심층논의되었으며, 향후 개별 국가의 노력 및 국제협력의 방향 등이 확인되었다. 본 논문에서는 상기협약에서 논의된 사용후핵 연료 관리에 대한 국가정책 및 향후 추세 둥을 자세히 기술하였다. 또한 주요국가의 최근 이행내용도 요약정리 하였다.