• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.2
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• pp.141-154
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• 2015

In general, conventional criticality analysis for spent fuel transport/storage systems have been performed based on the assumption of fresh fuel concerning the potential uncertainties from number density calculations of actinide nuclides and fission products in spent fuel. However, these evaluation methods cause financial losses due to an excessive criticality margin. In order to overcome this disadvantage, many studies have recently been conducted to design and commercialize a transportation and storage cask applied to the Burnup Credit (BUC). This study conducted an assessment to ensure criticality safety for reactor operating parameters, axial burn-up profiles and misload accident conditions, which are the factors that are likely to affect criticality safety when the BUC is applied to the dual-purpose cask under development at the KOrea RADioactive waste agency (KORAD). As a result, it was found that criticality resulting from specific power, changed substantially and relied on conditions of low enrichment and high burn-up. Considering the end effect in the case of high burn-up produced a positive-definite result. In particular, the increment of maximum effective multiplication factors due to misloading was 0.18467, confirming that misload is a factor that must be taken into account when applying the BUC. The results of this study may therefore be utilized as references in developing technologies to apply the BUC to domestic models and operational procedures or preventing any misload accidents during the process of spent fuel loading.

• Analytical Science and Technology
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• v.17 no.5
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• pp.438-442
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• 2004

To determine the contents of $^{99}Tc$ in the spent PWR (pressurized water reactor) nuclear fuels by ICP-MS (inductively coupled plasma-mass spectrometry), a technetium separation method using an extraction chromatographic resin (TEVA Spec resin) has been established. $^{99}Tc$ was separated from a spent PWR nuclear fuel solution by this separation procedure and its concentration was determined by ICP-MS. The result agrees well with the value calculated by the program ORIGEN 2 and also the value measured by AG MP-1 resin/ICP-MS method described in our previous paper. It can be concluded that the present separation procedure is superior to the AG MP-1 resin procedure with respect to the time required for technetium separation as well as the efficiency of decontamination from other radioactive nuclides.

• Journal of Radiation Protection and Research
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• v.9 no.2
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• pp.97-102
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• 1984

Criticality analyses of the KSC-1(Korean Shipping Cask-1) spent fuel shipping cask have been performed with the help of KENO-IV Monte Carlo computer code and 19-group CSLIB 19 cross section set which was generated from AMPX modular system. The analyses followed a benchmark calculation which has been made regard to the B & W CX-10 criticality facility in order to validate the Monte Carlo code cross section set described above. The KSC-1 shipping cask seems to be safe in the criticality point of view for the transport of one PWR spent fuel assembly under the normal conditions as well as the hypothetical accident conditions.

• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.47-59
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• 1993

The criticality of the shipping cask(KSC-7) for transportion of 7PWR spent fuel assemblies has been calculated and analysised on the basis of neutron transport theory. For criticality analysis, effects of the rod pitches, the fixed neutron absorbers(borated sus+boral) were considered. The effective multiplication factor has been calculated by KENO-Va, Mote Carlo method computer code, with the HANSEN-ROACH 16 group cross section set, which was made for personal computer system. The criticality for the KSC-7 cask was calculated in terms of the fresh fuel which was conservative for the aspects of nuclear critility. From the results of criticality analysis, the calculated Keff is proved to be lower than subcritical limit during normal transportation and under hypothetical accident condition. The maximum calculated criticalities of the KSC-7 were lower the safety criticality limit 1.0 recommended by US 10CFR71 both under normal and hypothetical accident condition. Also, to verify the KSC-7 criticality calculation results by using KENO-Va, it was carried out benchmark calculation with experimental data of B & W(Bobcock and Wilcox) company. From the 3s series of calculation of the KSC-7 cask and benchmark calculation, the cask was safely designed in nuclear criticality, respectively.

• Tunnel and Underground Space
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• v.11 no.2
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• pp.132-145
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• 2001

The purpose and need of the study is to quantify the advantage or disadvantage of the environmental friendliness of the partitioning of nuclear fuel cycle. To this end, a preliminary study on the quantitative effect of the partition on the permanent disposal of spent PWR and CANDU fuel (HLW) was carried out. Before any analysis, the so-called reference radionuclide release scenario from a potential repository embedded into a crystalline rock was developed. Firstly, the feature, event and processes (FEPs) which lead to the release of nuclides from waste disposed of in a repository and the transport to and through the biosphere were identified. Based on the selected FEPs, the ‘Well Scenario’which might be the worst case scenario was set up. For the given scenario, annual individual doses to a local resident exposed to radioactive hazard were estimated and compared to that from direct disposal. Even though partitioning and transmutation could be an ideal solution to reduce the inventory which eventually decreases the release time as well as the peaks in the annual dose and also minimize the repository area through the proper handling of nuclides, it should overcome major disadvantages such as echnical issues on the partitioning and transmutation system, cost, and public acceptance, and environment friendly issues. In this regard, some relevant issues are also discussed to show the direction for further studies.

• Analytical Science and Technology
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• v.16 no.6
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• pp.431-437
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• 2003

Metalization yield of uranium oxide to uranium metal from lithium reduction process of spent pressurized water reactor (PWR) fuels was measured using thermogravimetric analyzer. A reduced metal produced in the process was divided into a solid and a powder part, and each metalization yield was measured. Metalization yield of the solid part was 90.7~95.9 wt%, and the powder being 77.8~71.5 wt% individually. Oxidation behaviour of the quartemary alloy was investigated to take data on the thermal oxidation stability necessary for the study on dry storage of the reduced metal. At $600{\sim}700^{\circ}C$, weight increments of alloy of Mo, Ru, Rh and Pd was 0.40~0.55 wt%. Phase change on the surface of the alloy was started at $750^{\circ}C$. In particular, Mo was rapidly oxidized and then the alloy lost 0.76~25.22 wt% in weight.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.331-336
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• 2003

The advanced spent fuel conditioning process(ACP) was proposed to reduce considerably the overall volume and radioactivity for effective management of the PWR spent fuel in respects on safety and economy. The ACP is under research and development, and have scheduled to perform hot test for demonstration of the ACP after several years. For hot test, hot cell facility of ${\alpha}{\gamma}$ type possess conservative safety is required essentially. A existing hot cell of ${\beta}{\gamma}$ type will be refurbished to minimize construction expenditures of hot cell facility. In this study, the design requirements are established, and the process detail work flow was analysed for the optimum arrangement to ensure effective process operation in hot cell. And also, the basic and detail design of hot cell facility and process and safety analysis was peformed to secure conservative safety of hot cell facility and process.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.195-202
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• 2018

As a rule, geological disposal is considered a safe method for final disposal of high-level radioactive waste. However, some long-lived fission products like $^{99}Tc$ and $^{129}I$ contained in spent nuclear fuel are highly mobile as less sorbing anionic species in the subsurface environment and can mainly cause exposure dose to the ecosystem by emission of beta rays in the hundreds of keV range. Therefore, if these two nuclides can be separated and converted with high efficiency into radioactively unharmful nuclides, this would have a positive effect on disposal safety. One candidate method is to transmute these two nuclides in nuclear reactors into short-lived nuclides or into stable nuclides. For this purpose, it is necessary to evaluate which reactor type is more efficient in burning these two nuclides. In this study, the simulation results of nuclear transmutation of $^{99}Tc$ and $^{129}I$ in light water reactor (PWR), heavy water reactor (CANDU) and fast neutron reactor (SFR, MET-1000) are compared and discussed.

• Journal of Radiation Protection and Research
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• v.14 no.1
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• pp.56-65
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• 1989

The nuclear criticality of the KSC-4 shipping cask which can load four assemblies of PWR spent fuel was analyzed using KENO-IV computer code and 19-group nuclear cross section set generated from 218-group neutron cross section library(DLC-43/CSRL) using AMPX system. In accordance with 10CFR71, the analysis was performed for fresh fuel assemblies, instead of the spent fuels, under both norml transportation and hypothetical accident conditions. The calculated maximum multiplication factors(Keff) of the KSC-4 cask were 0.85289 and 0.94185 for the normal transportation and hypothetical accident conditions, respectively. The highest Keff of the KSC-4 cask is within the subcritical limit prescribed in l0CFR71 and accordingly the KSC-4 cask is safely designed in terms of nulear criticality.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.343-356
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• 2016

The Korea Radioactive Waste Agency (KORAD) has developed a dual-purpose metal cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. The metal cask was designed in compliance with international and domestic technology standards, and safety was the most important consideration in developing the design. It was designed to maintain its integrity for 50 years in terms of major safety factors. The metal cask ensures the minimization of waste generated by maintenance activities during the storage period as well as the safe management of the waste. An activation evaluation of the main body, which includes internal and external components of metal casks whose design lifetime has expired, provides quantitative data on their radioactive inventory. The radioactive inventory of the main body and the components of the metal cask were calculated by applying the MCNP5 ORIGEN-2 evaluation system and by considering each component's chemical composition, neutron flux distribution, and reaction rate, as well as the duration of neutron irradiation during the storage period. The evaluation results revealed that 10 years after the end of the cask's design life, $^{60}Co$ had greater radioactivity than other nuclides among the metal materials. In the case of the neutron shield, nuclides that emit high-energy gamma rays such as $^{28}Al$ and $^{24}Na$ had greater radioactivity immediately after the design lifetime. However, their radioactivity level became negligible after six months due to their short half-life. The surface exposure dose rates of the canister and the main body of the metal cask from which the spent nuclear fuel had been removed with expiration of the design lifetime were determined to be at very low levels, and the radiation exposure doses to which radiation workers were subjected during the decommissioning process appeared to be at insignificant levels. The evaluations of this study strongly suggest that the nuclide inventory of a spent nuclear fuel metal cask can be utilized as basic data when decommissioning of a metal cask is planned, for example, for the development of a decommissioning plan, the determination of a decommissioning method, the estimation of radiation exposure to workers engaged in decommissioning operations, the management/reuse of radioactive wastes, etc.