• Journal of Radiation Protection and Research
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• v.41 no.2
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• pp.117-121
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• 2016

Background: Definition and grouping of initiating events (IEs) are important basics for probabilistic safety assessment (PSA). An IE in a spent fuel reprocessing plant (SFRP) is an event that probably leads to the release of dangerous material to jeopardize workers, public and environment. The main difference between SFRPs and nuclear power plants (NPPs) is that hazard materials spread diffusely in a SFRP and radioactive material is just one kind of hazard material. Materials and Methods: Since the research on IEs for NPPs is in-depth around the world, there are several general methods to identify IEs: reference of lists in existence, review of experience feedback, qualitative analysis method, and deductive analysis method. While failure mode and effect analysis (FMEA) is an important qualitative analysis method, master logic diagram (MLD) method is the deductive analysis method. IE identification in SFRPs should be consulted with the experience of NPPs, however the differences between SFRPs and NPPs should be considered seriously. Results and Discussion: The plutonium uranium reduction extraction (Purex) process is adopted by the technics in a model reprocessing plant. The first extraction cycle (FEC) is the pivotal process in the Purex process. Whether the FEC can function safely and steadily would directly influence the production process of the whole plant-production quality. Important facilities of the FEC are installed in the equipment cells (ECs). In this work, IEs in the FEC process were identified and categorized by FMEA and MLD two methods, based on the fact that ECs are containments in the plant. Conclusion: The results show that only two ECs in the FEC do not need to be concerned particularly with safety problems, and criticality, fire and red oil explosion are IEs which should be emphatically analyzed. The results are accordant with the references.

• Nuclear Engineering and Technology
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• v.39 no.6
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• pp.731-736
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• 2007

Radioactive organic wastes containing acetone, alcohol, and particularly tributyl phosphate (TBP)/dodecane contaminated with uranium are extracted from the PUREX process and the decontamination of related equipment. An evaporation method that utilizes existing DU oxidation apparatuses and ventilation systems and a typical muffle furnace installed with an aspirating system are adopted. A separation method using phosphoric acid especially for the TBP/dodecane waste is also studied and evaluated. The results show that a simple evaporation process is utilizable for wastes containing acetone or alcohol with a lower boiling point. A modified muffle furnace is more appropriate to dispose directly of organic wastes having a higher boiling point, such as TBP/dodecane, without generating a condensed waste solution. It is recommended that, when the uranium concentration of TBP/dodecane waste is much higher than stipulated levels, separation technology should be applied to remove uranium from the mixture. Each type of solvent after separation can then be considered disposable below the regulatory limit in the modified furnace discussed in this study.

• Journal of Institute of Convergence Technology
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• v.2 no.1
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• pp.7-12
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• 2012

Spent fuels are discharged from nuclear reactors as a result of power generations. The spent fuels would be considered as a useful resources because the main constituent is uranium and some other actinides are included in them. In order to utilize the resources chemical processes should be developed to treat the spent fuels and obtain uranium and other actinides to be fueled in a fast reactor. The technologies are categorized into wet and dry processes. In this study, the current status of such technologies is summarized to give a insight and a deep understanding on nuclear fuel cycles.

• Bulletin of the Korean Chemical Society
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• v.26 no.11
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• pp.1692-1694
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• 2005

Spent nuclear fuel is reprocessed commercially by the chemical process to recover U and Pu. Recently, new salt-free reagents to separate plutonium and neptunium from uranium suitable for use in a single cycle flowsheet have been developed. Acetohydroxamic acid $(CH_3CONHOH)$ has been taken much interest in as a complexing agent capable of selective stripping of tetravalent actinides from U(VI) when actinides are present in the solvent stream of the advanced PUREX process. Additionally acetohydroxamic acid will rapidly reduce Np(VI) to inextractable Np(V) thus allowing the separation of Np from U. In this study, the rate equation for the reduction of Np(VI) to Np(V) in nitric acid aqueous solution has been determined as: $-[NpO_2^{2+}]$/dt = $k[NpO_2^{2+}]$[AHA] with k = 191.2 ${\pm}$ 11.2 $M^{-1}s^{-1}$ at 25 ${\pm}$ 0.5 ${^{\circ}C}$ and $[HNO_3]$ = 1.0 M. Comparison with other reductants available in the literature, acetohydroxamic acid is a strong one for $NpO_2^{2+}$.

• Nuclear Engineering and Technology
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• v.40 no.7
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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.

• Nuclear Engineering and Technology
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• v.40 no.3
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• pp.183-190
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• 2008

Pyroprocessing is the optimal means of treating spent metal fuels from metal fast fuel reactors and is proposed as a potential option for GNEP in order to meet the requirements of the next generation fuel cycle. Currently, efforts for research and development are being made not only in the U.S., but also in Asian countries. Electrorefining, cathode processing by distillation, injection casting for fuel fabrication, and waste treatment must be verified by the use of genuine materials, and the engineering scale model of each device must be developed for commercial deployment. Pyroprocessing can be effectively extended to treat oxide fuels by applying an electrochemical reduction, for which various kinds of oxides are examined. A typical morphology change was observed following the electrochemical reduction, while the product composition was estimated through the process flow diagram. The products include much stronger radiation emitter than pure typical LWR Pu or weapon-grade Pu. Nevertheless, institutional measures are unavoidable to ensure proliferation-proof plant operations. The safeguard concept of a pyroprocessing plant was compared with that of a PUREX plant. The pyroprocessing is better adapted for a collocation system positioned with some reactors and a single processing facility rather than for a centralized reprocessing unit with a large scale throughput.

• Nuclear Engineering and Technology
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• v.43 no.4
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• pp.317-328
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• 2011

Pyroprocessing technology was developed in the beginning for metal fuel treatment in the US in the 1960s. The conventional aqueous process, such as PUREX, is not appropriate for treating metal fuel. Pyroprocessing technology has advantages over the aqueous process: less proliferation risk, treatment of spent fuel with relatively high heat and radioactivity, compact equipment, etc. The addition of an oxide reduction process to the pyroprocessing metal fuel treatment enables handling of oxide spent fuel, which draws a potential option for the management of spent fuel from the PWR. In this context, KAERI has been developing pyroprocessing technology to handle the oxide spent fuel since the 1990s. This paper describes the current status of pyroprocessing technology development at KAERI from the head-end process to the waste treatment. A unit process with various scales has been tested to produce the design data associated with the scale up. A performance test of unit processes integration will be conducted at the PRIDE facility, which will be constructed by early 2012. The PRIDE facility incorporates the unit processes all together in a cell with an Ar environment. The purpose of PRIDE is to test the processes for unit process performance, operability by remote equipment, the integrity of the unit processes, process monitoring, Ar environment system operation, and safeguards related activities. The test of PRIDE will be promising for further pyroprocessing technology development.

• Nuclear Engineering and Technology
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• v.7 no.3
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• pp.213-222
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• 1975

Immobilization of the second-cycle radioactive liquid wastes from a Purex process was developed with the blown asphalt (manufactured by Kukdong Shell Oil Company Ltd) to eliminate the possibility that the radioactive materials will be redispersed into the environment. Attempts to incorporate these wastes directly into the asphalt martrices without any pretreatment were not successful, as it was observed that the sulphuric acid in the waste oxidised the asphalt. Hence, the waste was treated with caustic soda and made alkaline prior to bituminization, so that it was found that this pretreatment made the waste compatible to the asphalt matrices. The pure blown asphalt samples irradiated with doses of 4.0$\times$10$^{7}$ rad showed no evidence of volume increase. The suitable temperature for incorporation of the alkaline wastes into blown asphalt was 180-20$0^{\circ}C$. The Products containing 50 wt% salts represented the following good properties viz., volume reduction (about 1.4), homogeneity, teachability etc. During the period of 131 day $s^{l37}$Cs from products containing 40wt% salts was leached at rates ranging from 2.70$\times$10-4 to 8.27$\times$10-4g/cm2_day but the rate for $^{90}$ Sr was lower by one to two orders of magnitude by distilled water. The leaching rates for $^{137}$ Cs and $^{90}$ Sr by sea water were slightly lower than by distilled water. Both of the leaching rates decreased with increasing pH.H.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.4
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• pp.339-350
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• 2007

Status on the spent nuclear fuel management policy and R&D plan of the major countries is surveyed. Also the prospect of the future R&D plan is suggested. Recently so-called fuel cycle nations, which have the reprocess policy of the spent fuel, announced new spent fuel management policy based on the advanced fuel cycle technology. The policy is focused to transmute highly radioactive material and material having a very long half-life, and to recycle the Pu and U contained in the spent fuel. In this way the radio-foxily of the spent fuel as well as the amount of the high level waste to be eventually disposed can greatly be reduced. Most of countries selected the wet process as a primary option for the treatment of the spent fuel since the advanced wet process, which is based on the existing PUREX process, looks more feasible as compared with the dry process. The wet process, however, is much more sensitive in terms of proliferation-resistance compared with the dry process. The pure Pu can easily be obtained by simply modifying the process. On the other hand the pure Pu can not be extracted in the dry process based on the high temperature molten salt process such as a pyroprocess. Even though the pyroprocess technology is very premature, it has a great merit. Thus it is necessary for Korea to have a long term strategy for pursuing a spent fuel treatment technology with a proliferation resistance and a great merit for the GEN-IV fuel cycles. Pyroprocess is one of the best candidates to satisfy these purposes.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.4
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• pp.243-251
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• 2015

Nuclear power is currently the second largest power supply method in Korea and the number of nuclear power plants are planned to be increased as well. However, clear management policy for spent fuels generated from nuclear power plants has not yet been established. The back-end fuel cycle, associated with nuclear material flow after nuclear reactors is a collection of technologies designed for the spent fuel management and the spent fuel management policy is closely related with the selection of a nuclear fuel cycle. Cost is an important consideration in selection of a nuclear fuel cycle and should be determined by adding external cost to private cost. Unlike the private cost, which is a direct cost, studies on the external cost are focused on nuclear reactors and not at the nuclear fuel cycle. In this research, external cost indicators applicable to nuclear fuel cycle were derived and quantified. OT (once through), DUPIC (Direct Use of PWR SF in CANDU), PWR-MOX (PWR PUREX reprocessing), and Pyro-SFR (SFR recycling with pyroprocessing) were selected as nuclear fuel cycles which could be considered for estimating external cost in Korea. Energy supply security cost, accident risk cost, and acceptance cost were defined as external cost according to precedent and estimated after analyzing approaches which have been adopted for estimating external costs on nuclear power generation.