• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2819-2827
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• 2022

The disposal area of a deep geological repository (DGR) for the disposal of spent nuclear fuels (SNFs) is estimated considering the spacing between deposition holes and between disposal tunnels, as determined by a thermal analysis using the decay heat of a reference SNF. Given the relatively large amount of decay heat of the reference SNF, the disposal area of the DGR is found to be overestimated. Therefore, we develop a computer program using MATLAB, termed ACom (Assembly Combination), to combine SNFs when stored in canisters such that the decay heat per canister is evenly distributed. The stability of ACom was checked and the overall distribution of the decay heat per canister was analyzed. Finally, ACom was applied to disposal scenarios suggested in the conceptual design of a DGR for SNFs, and it was confirmed that the decay heat per canister could be evenly distributed and that the maximum decay heat of the canister could be much lower than that of a canister estimated using a reference SNF. ACom can be used to improve the disposal efficiency by reducing the disposal area of a DGR for SNFs by ensuringg a relatively even distribution of decay heat per canister.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.18-24
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• 2018

This study proposes a new method of analyzing the burnup credit in boiling water reactor spent fuel assemblies against various operating parameters. The operating parameters under investigation include fuel temperature, axial burnup profile, axial moderator density profile, and control blade usage. In particular, the effects of variations in one and two operating parameters on the curve of effective multiplication factor ($k_{eff}$) versus burnup (B) are, respectively, the so-called single and compound effects. All the calculations were performed using SCALE 6.1 together with the Evaluated Nuclear Data Files, part B (ENDF/B)-VII238-neutron energy group data library. Furthermore, two geometrical models were established based on the General Electric (GE)14 $10{\times}10$ boiling water reactor fuel assembly and the Generic Burnup-Credit (GBC)-68 storage cask. The results revealed that the curves of $k_{eff}$ versus B, due to single and compound effects, can be approximated using a first degree polynomial of B. However, the reactivity deviation (or changes of $k_{eff}$, ${\Delta}k$) in some compound effects was not a summation of the all ${\Delta}k$ resulting from the two associated single effects. This phenomenon is undesirable because it may to some extent affect the precise assessment of burnup credit. In this study, a general formula was thus proposed to express the curves of $k_{eff}$ versus B for both single and compound effects.

• Journal of Radiation Protection and Research
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• v.4 no.1
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• pp.1-4
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• 1979

Effective multiplication factor has been calculated for one PWR fresh fuel assembly immersed in a spent fuel storage vault on the basis of the neutron transport theory. A numerical calculation has been carried out by means of Sn approximation. The method employed in this study is that the energy domain is broken into 16 groups, the angular variable is divided into four discrete direction, i.e., $S_4$, and the spatial variable which is divided into fine meshes at the interface between different materials is discretized into 27 mesh points. The calculated $K_{eff}$ value of 0.6145 seems to be far small in comparison with the value obtained by other author for an infinite array of fuel assemblies.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.134-140
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• 2016

The Control Rod Assembly (CRA) controls the reactor power by adjusting its position in the reactor core during normal operation and should be quickly inserted into the reactor core by free drop under scram condition to shut down chain reactions. Therefore, the drop time of the CRA is one of important factors for the safety of the nuclear reactor and must be experimentally verified. This study presents the drop performance test of the CRA which has been conceptually designed for the Proto-type Generation IV Sodium-cooled Fast Reactor. During the test, the CRA was free dropped from a height of 1 m under different flow rate conditions and its drop time was measured. The results showed that the drop time of the CRA increased as the flow rate increased; the average drop times of the CRA were approximately 1.527 seconds, 1.599 seconds and 1.676 seconds at 0%, 100% and 200% of design flow rates, respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.403-409
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• 2017

For dry storage of the spent nuclear fuel (SNF) stored in the storage pool of a nuclear power plant, essentially all moisture must be removed to prevent corrosion of the assembly and canister internals and/or degradation of fuel cladding integrity after SNF canister loading operation. R&D work is now in progress on a forced gas drying system that can be used to remove residual water in canisters. In this work, preliminary design is performed to manufacture the forced gas drying system. This process includes a case study of dry methods for canister moisture removal, relative codes and standards, confirmation of adequate dryness, needs analysis at plant sites, and characteristics of SNF stored in pools. Through this preliminary design work, we obtained a conceptual flow diagram and preliminary P&ID of the forced gas drying system. The results of this study can be used to determine details of the design to manufacture the forced gas drying system.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.483-490
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• 2012

PWR spent fuels produced in the Republic of Korea are expected to be recycled by pyroprocess in the long term future. Even though pyroprocess waste amounts can be smaller than that of PWR spent fuel assembly in case of direct disposal, this process essentially will produce various and unique radioactive wastes. The goals of this article are to characterize these wastes, calculate the amount of wastes, design disposal systems for each waste and evaluate the radiation safety of each system by dose assessment. The absorbed dose results of the metal and ceramic waste for the engineering barrier system (EBS) showed $2.21{\times}10^{-2}$ Gy/h and $1.15{\times}10^{-2}$ Gy/h, which are lower than the recommended value of 1 Gy/h. These results confirmed that the newly proposed disposal systems have a safety margin for the radiation produced from each waste.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.70-75
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• 1997

A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14$\times$14 PWR fuel assembly and the results are given in this paper.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.339-346
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• 2019

Based on the $8^{th}$ Basic Plan for Electric Power Demand and Supply, an estimation has been made for inventories and characteristics of spent fuel (SF) to be generated from existing and planned nuclear power plants. The characteristics under consideration in this study are dimensions, fuel array, $^{235}U$ enrichment, discharge burnup, and cooling time for each fuel assembly. These are essentially needed for designing a disposal facility for SFs. It appears that the anticipated quantity by the end of 2082 is about 62,500 assemblies for PWR SFs. The inventories of Westinghouse-type and Korean-type SFs were revealed to be 60% and 40%, respectively as of the end of 2018. The proportion of SFs with initial $^{235}U$ enrichment below 4.5 weight percent (wt%) was shown to be approximately 90% in total as of the end of 2018. As of 2077, more than 97% of SFs generated from Westinghouse-type nuclear reactors were shown to have cooling time of over 50 years. As of 2125, more than 98% of SFs generated from Korean-type nuclear reactors were shown to have cooling time of over 45 years. Based on these results, for the efficient design of a disposal system, it is reasonable to adopt two types of reference spent fuel. SF of KSFA with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 50 years was determined as reference fuel for Westinghouse-type SFs; SF of PLUS7 with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 45 years was determined as reference fuel for Korean-type SFs.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.248-254
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• 1993

The purpose of this study is to evaluate the thermal safety for dry transport of a shipping cask. Analysis condition was based on an ambient temperature of 38$^{\circ}C$ for normal heat condition. The cask was designed to carry 4PWR spent fuel assemblies with a burnup of 38,000 MWD/MTU and 3 years of cooling time. Thermal analysis was carried out by using the COBRA-SFS code. The fuel cavity was considered to be filled with air, nitrogen or helium gas for dry transport. The results of analysis showed that the maximum temperatures of fuel rod cladding in air and helium cavity would be 277$^{\circ}C$ and 226$^{\circ}C$, respectively, for 3 years of cooling time. These values were less than the specified temperature to maintain the thermal integrity of fuel assembly for dry transport.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.661-673
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• 2020

The quarter-core simulation of BEAVRS Cycle 2 depletion benchmark has been conducted using the MCS/CTF coupling system. MCS/CTF is a cycle-wise Picard iteration based inner-coupling code system, which couples sub-channel T/H (thermal/hydraulic) code CTF as a T/H solver in Monte Carlo neutron transport code MCS. This coupling code system has been previously applied in the BEAVRS benchmark Cycle 1 full-core simulation. The Cycle 2 depletion has been performed with T/H feedback based on the spent fuel materials composition pre-generated by the Cycle 1 depletion simulation using refueling capability of MCS code. Meanwhile, the MCS internal one-dimension T/H solver (MCS/TH1D) has been also applied in the simulation as the reference. In this paper, an analysis of the detailed criticality boron concentration and the axially integrated assembly-wise detector signals will be presented and compared with measured data based on the real operating physical conditions. Moreover, the MCS/CTF simulated results for neutronics and T/H parameters will be also compared to MCS/TH1D to figure out their difference, which proves the practical application of MCS into the BEAVRS benchmark two-cycle depletion simulations.