• Nuclear Engineering and Technology
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• v.49 no.8
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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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• v.40 no.1
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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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.110-116
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• 2020

As the management plan for domestic spent nuclear fuel is delayed, the storage of the operating nuclear power plant is approaching saturation, and the Kori 1 Unit that has reached its end of operation life is preparing for the dismantling plan. The first stage of dismantling is the transfer of spent nuclear fuel stored in storage at plants. The spent fuel management process leads to temporary storage, interim storage, reprocessing and permanent disposal. In this paper, the technical issues to be considered when transporting spent fuel in this process are summarized. The spent fuels are treated as high-level radioactive waste and strictly managed according to international regulations. A series of integrity tests are performed to demonstrate that spent fuel can be safely stored for decades in a dry environment before being transferred to an intermediate storage facility. The safety of spent fuel transport container must be demonstrated under normal transport conditions and virtual accident conditions. IAEA international standards are commonly applied to the design of transport containers, licensing regulations and transport regulations worldwide. In addition, each country operates a physical protection system to reduce and respond to the threat of radioactive terrorism.

• Nuclear Engineering and Technology
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• v.45 no.4
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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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• v.37 no.4
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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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.1
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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%.

• Journal of Radiation Protection and Research
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• v.21 no.1
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• pp.59-72
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• 1996

The continuous expansions and development of nuclear power have led to generation of the significant volume of spent fuels and radioactive wastes. And so, safe and effective management of the spent fuel has been becoming internationally sensitive and significant issue since the early 1990s. Especially, more importance would be added in the view point of international politics, because of recent political changes in the countries of Eastern Europe including dissociation of the former Soviet Union and the difficulties faced by the nuclear industries worldwide. Accordingly, this paper is proposed to show an overview of national strategies and Policies on the spent fuel management, that are being assessed and carried out worldwide at this time. The overview is based on recent developments of the national strategies, their implementations and some related experiences presented in IAEA International meetings and some technical papers.

• Nuclear Engineering and Technology
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• v.48 no.3
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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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• v.38 no.4
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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 Korean Society of Environmental Engineers
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• v.37 no.5
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• pp.312-323
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• 2015

This paper presents future directions for spent nuclear fuel and high-level radioactive waste management. The successes and failures of siting nuclear waste repository experienced by the United States and other countries are reviewed with the current policy stance. Further, the needs for establishing management policy, considering the high-level radioactive waste produced by the dismantlement, nuclear security concerns, and cost-effectiveness analysis for the total nuclear fuel cycle, are emphasised. Technical discussions are organised into three main topics: interim storage, permanent disposal, and reprocessing. Licensing regimes are also investigated to suggest strategic plans for research and development programmes in the Republic of Korea.