• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.87-93
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• 2006

Inventories, and characteristics such as dimension, fuel rod array, weight, $^{235}U$ enrichment, and discharge burnup of spent nuclear fuel (SNF) generated from existing and planed nuclear power plants based on National 2nd Basic Plan for Electric Power Demand and Supply were investigated and projected to support geological disposal system design. The historical and projected inventory by the end 2057 is expected to be 20,500 and 14,800MTU for PWR and CANDU spent nuclear fuel, respectively. The quantity of SNF with initial $^{235}U$ enrichment of 4.5 wt.% and below was shown to be 96.5% in total. Average burnup of SNF revealed $\sim36$ GWD/MTU and $\sim40$ GWD/MTU for the period of 1994-1999 and 2000-2003, respectively. It is expected that the average burnup of SNF will be $\sim45$ GWD/MTU at the end of 2000's. From the comprehensive study, it was concluded that the imaginary SNF with $16\times16$ Korean Standard Fuel Assembly, cross section of $21.4cm\times21.4cm$, length of 453cm, mass of 672 kg, initial $^{235}U$ enrichment of 4.5 wt.%, discharge burnup of 55 GWD/MTU could cover almost all SNFs to be produced by 2057.

• Journal of Radiation Protection and Research
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• v.35 no.1
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• pp.21-25
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• 2010

Burnup of spent fuel should be determined accurately for the safety control of spent fuel. Especially, it is necessary to measure the burnup profile along the nuclear fuel axis. In the present work, an ionization chamber was designed and fabricated to measure the gamma ray profile inside the guide tube of spent fuel. The ionization chamber was composed of three parts; induction part, gas-inlet part, and sensor part. The sensor part had two electrodes; cathode and anode. A guide electrode was considered in the ionization chamber design to make the ionization chamber to be inserted easily into the guide tube. Pure gas (argon and xenon) was inserted into the ionization chamber, and the leakage current and saturation curve were measured to determine the operation characteristics of the ionization chamber. The gamma ray radiation was also measured in relatively high dose environment. The gamma ray profile of the spent fuel will be measured with the ionization chamber.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.34-41
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• 2024

This work presents experimental data and modelling of the release of Mo from high-burnup spent nuclear fuel (63 MWd/kgU) at two different pH values, 8.4 and 13.2 in air. The release of Mo from SF to the solution is around two orders of magnitude higher at pH = 13.2 than at pH = 8.4. The high Mo release at high pH would indicate that Mo would not be congruently released with uranium and would have an important contribution to the Instant Release Fraction, with a value of 5.3%. Parallel experiments with pure non irradiated Mo(s) and XPS determinations indicated that the faster dissolution at pH = 13.2 could be the consequence of the higher releases from metallic Mo in the fuel through a surface complexation mechanism promoted by the OH^- and the oxidation of the metal to Mo(VI) via the formation of intermediate Mo(IV) and Mo(V) species.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.272-273
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.470-475
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• 1998

The marginal nuclear criticality analysis for the high density spent fuel storage at a PWR plant was carried out by using the HELIOS and CASMO-3 codes. More than 20 % of the calculated reactivity saving effect is observed in this analysis. This mainly comes from the adoption of some important fission products and B-10 in the criticality analysis. By taking burnup and boron credits, the high capacity of the spent fuel storage rack can be more fully utilized, reducing the space of storage. Larger storage for a given inventory of spent fuel should result in remarkable cost savings and mort importantly reduce the risks to the public and occupational workers.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.37-50
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• 2020

In this paper, an overview of the scoping calculation results is provided with respect to criticality and radiation shielding of two KBS-3V type PWR SNF disposal systems and one NWMO-type CANDU SNF disposal system of the improved KAERI reference disposal system for SNFs (KRS⁺). The results confirmed that the calculated effective multiplication factors (k_eff) of each disposal system comply with the design criteria (< 0.95). Based on a sensitivity study, the bounding conditions for criticality assumed a flooded container, actinide-only fuel composition, and a decay time of tens of thousands of years. The necessity of mixed loading for some PWR SNFs with high enrichment and low discharge burnup was identified from the evaluated preliminary possible loading area. Furthermore, the absorbed dose rate in the bentonite region was confirmed to be considerably lower than the design criterion (< 1 Gy·hr^-1). Entire PWR SNFs with various enrichment and discharge burnup can be deposited in the KRS⁺ system without any shielding issues. The container thickness applied to the current KRS⁺ design was clarified as sufficient considering the minimum thickness of the container to satisfy the shielding criterion. In conclusion, the current KRS⁺ design is suitable in terms of nuclear criticality and radiation shielding.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.689-699
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• 2020

In this work, new multi-recycling options of TRU nuclides using PWR fuel assemblies comprised of MOX and FCM (Fully Ceramic Micro Encapsulated) fuels are suggested and neutronically analyzed. These options do not use a fully recycling of TRU but a partial recycling where TRUs from MOX fuels are recycled while the ones from FCM fuels are not recycled due to their high consumption rate resulted from high burnup. In particular, additional external TRU feed in MOX fuels for each cycle was considered to significantly increase the TRU consumption rate and the finally selected option is to use external TRU and enriched uranium feed as a makeup for the heavy metal consumption in MOX fuels. This hybrid external feeding of TRU and enriched uranium in MOX fuel was shown to be very effective in significantly increasing TRU consumption rate, maintaining long cycle length, and achieving negative void reactivity worth during recycling.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.312.1-312.1
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• 2003

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3723-3740
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• 2021

The harsh conditions in the reactor affect the thermal and mechanical performance of the fuel plate heavily. Some in-pile behaviors, like fission-induced swelling, can cause a large deformation of fuel plate at very high burnup, which may even disturb the flow of coolant. In this research, the emphasis is put on the thermal expansion, fission-induced swelling, interaction layer (IL) growth, creep of the fuel meat, and plasticity of the cladding for the U₃Si₂/Al dispersion fuel plate. A detailed model of the fuel meat swelling is developed. Taking these in-pile behaviors into consideration, the three-dimensional large deformation incremental constitutive relations and stress update algorithms have been developed to study its thermal-mechanical performance under normal conditions using Abaqus. Results have shown that IL can effectively decrease the thermal conductivity of fuel meat. The high Mises stress region mainly locates at the interface between fuel meat and cladding, especially around the side edge of the interface. With irradiation time increasing, the stress in the fuel plate gets larger resulting from the growth of fuel meat swelling but then decreases under the effect of creep deformation. For the cladding, plasticity deformation does not occur within the irradiation time.

• Nuclear Engineering and Technology
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• v.30 no.6
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• pp.541-554
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• 1998

A computer code COSMOS has been developed based on the CARO-D5 for the thermal analysis of LWR UO$_2$ and MOX fuel rod under steady-state and transient operating conditions. The main purpose of the COSMOS, which considers high turnup characteristics such as thermal conductivity degradation with turnup and rim formation at the outer part of fuel pellet, is to calculate temperature profile across fuel pellet and fission gas release up to high burnup. A new mechanistic fission gas release model developed based on physical processes has been incorporated into the code. In addition, the features of MOX fuel such as change in themo-mechanical properties and the effect of microscopic heterogeneity on fission gas release have been also taken into account so that it can be applied to MOX fuel. Another important feature of the COSMOS is that it can analyze fuel segment refabricated from base irradiated fuel rods in commercial reactors. This feature makes it possible to analyze database obtained from international projects such as the MALDEN and RISO, many of which were collected from refabricated fuel segments. The capacity of the COSMOS has been tested with some number of experimental results obtained from the HALDEN, RISO and FIGARO programs. Comparison with the measured data indicates that, although the COSMOS gives reasonable agreement, the current models need to be improved. This work is being performed using database available from the OECD/NEA.