• Proceedings of the Korean Nuclear Society Conference
• /
• 2001.05a
• /
• pp.371.1-371
• /
• 2001

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.20 no.2
• /
• pp.231-241
• /
• 2022

Several countries, including Korea, are considering the direct disposal of spent nuclear fuels. The radiological safety assessment results published after a geological repository closure indicate that the instant release is the main radiation source rather than the congruent release. Three Safety Case reports recently published were reviewed and the IRF values of seven long-lived radionuclides, including relevant experimental results, were compared. According to the literature review, the IRF values of both the CANDU and low burnup PWR spent fuel have been experimentally measured and used reasonably. In particular, the IRF values of volatile long-lived nuclides, such as ¹²⁹I and ¹³⁵Cs, were estimated from the FGR value. Because experimental leaching data regarding high burnup spent nuclear fuels are extremely scarce, a mathematical modelling approach proposed by Johnson and McGinnes was successfully applied to the domestic high burnup PWR spent nuclear fuel to derive the IRF values of iodine and cesium. The best estimate of the IRF was 5.5% at a discharge burnup of 55 GWd tHM^-1.

• Nuclear Engineering and Technology
• /
• v.48 no.5
• /
• pp.1109-1119
• /
• 2016

Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by $UO_2$ enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.8 no.1
• /
• pp.77-84
• /
• 2010

Reduction of oxides has been investigated for the volume reduction and recycling of the spent oxide fuel from commercial nuclear power plants. Various oxide reduction methods were proposed and KAERI (Korea Atomic Energy Research Institute) is currently developing an electrochemical reduction process using a LiCl-$Li_2O$ molten salt as a reaction medium. The electrochemical reduction process, the front end of the pyroprocessing, can connect the PWR (Pressurized Water Reactor) oxide fuel cycle to a metal fuel cycle of the sodium cooled fast reactor. This paper summarizes KAERI efforts on the development, improvement, and scale-up of the oxide reduction process.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.124-129
• /
• 1998

A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The $^{239}$ Pu and $^{235}$ U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the fled uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%.. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel has shown that it is desirable to increase the $^{239}$ Pu and $^{235}$ U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, il is recommended to have enrichments of 0.45 and 1.00 wt% for $^{239}$ Pu and $^{235}$ U, respectively.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2003.11a
• /
• pp.621-626
• /
• 2003

A DUPIC nuclear fuel is a newly developed fuel for CANDU reactors based on the concept of refabrication of spent PWR fuel by a dry process. Because a spent PWR fuel, a highly radioactive material, is used as a starting material, the experimental verification of DUPIC nuclear fuel fabrication requires an appropriate facility which should satisfy engineering requirements and guarantees safe operation. DUPIC nuclear fuel development team modified M6 hot-cell in IMEF to construct the dedicated facility(DFDF) for tile experiment. The experiment with spent PWR fuel have been conducted since January of 2000. Environmental effects of DFDF normal operation have been investigated when DUPIC nuclear fuel is fabricated with the maximum capacity of 50kg U/yr. The analysis results of the radiological safety of DFDF facility have shown that both national regulation limit and IMEF design criteria are satisfied.

• Nuclear Engineering and Technology
• /
• v.51 no.6
• /
• pp.1708-1714
• /
• 2019

The purpose of the critical evaluation of the spent fuel pool (SFP) is to verify that the maximum effective multiplication factor ($K_{eff}$) is less than the critical safety limit at 100% stored condition of the spent fuel with the maximum reactivity. At nuclear power plants, the storage standard of spent fuel, ie, the loading curve, is established to prevent criticality from being generated in SFP. Here, the loading curve refers to a graph showing the minimum discharged burnup versus the initial enrichment of spent fuel. Recently, US NRC proposed the new critical safety assessment guideline (DSS-ISG-2010-01, Revision 0) of PWR SFPs and most of utilities in US is following it. Of course, the licensed criterion of the maximum effective multiplication factor of SFP remains unchanged and it should be less than 0.95 from the 95% probability and the 95% confidence level. However, the new guideline is including the new evaluation methodologies like the application of the axial burnup profile, the validation of depletion and criticality code, and trend analysis. Among the new evaluation methodologies, the most important factor that affects $K_{eff}$ is the axial burnup profile of spent fuel. US NRC recommends to consider the axial burnup profiles presented in NUREG-6801 in criticality analysis. In this paper, criticality effect was evaluated considering three profiles, respectively: i) Axial burnup profiles presented in NUREG-6801. ii) Representative PWR axial burnup profile. iii) Uniform axial burnup profile. As the result, the case applying the axial burnup profiles presented in NUREG-6801 showed the highest $K_{eff}$ among three cases. Therefore, we need to introduce a new methodology because it can be issued if the axial burnup profiles presented in NUREG/CR-6801 are applied to the domestic nuclear power plants without any other consideration.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.7 no.4
• /
• pp.229-236
• /
• 2009

There are two types of spent fuels generated from nuclear power plants, CANDU type and PWR type. PWR spent fuels which include a lot of reusable material can be considered to be recycled. CANDU spent fuels are considered to directly disposed in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System(KRS) which is to dispose both PWR and CANDU spent fuels, the more effective CANDU spent fuel disposal systems have been developed. To do this, the disposal canister has been modified to hold the storage basket which can load 60 spent fuel bundles. From these modified disposal canisters, the disposal systems to meet the thermal requirement for which the temperature of the buffer materials should not be over $100^{\circ}C$ have been proposed. These new disposals have made it possible to introduce the concept of long tenn storage and retrievabililty and that of the two-layered disposal canister emplacement in one disposal hole. These disposal concepts have been compared and analyzed with the KRS CANDU spent fuel disposal system in terms of disposal effectiveness. New CANDU spent fuel disposal concepts obtained in this study seem to improve thermal effectiveness, U-density, disposal area, excavation volume, and closure material volume up to 30 - 40 %.

• Nuclear Engineering and Technology
• /
• v.23 no.2
• /
• pp.165-173
• /
• 1991

The recycle of PWR spent fuels in a CANDU reactor, so called the tandem fuel cycle is Investigated in this study. This scheme of utilizing Pm spent fuels will ease the shortage of spent fuel storage capacity as well as will improve the use of uranium resources. The minimum modification to the design of present CANDU reactor is seeked in the recycle. Nine different fuel types are considered in this work and are classified into two categories： refabrication and reconfiguration For refabrication, PWR spent fuels are processed and refabricated into the present 37 rod lattice structure of fuel bundle, and for reconfiguration, meanwhile, spent fuels are simply disassembled and rods are cut to fit into the present grid configuration of fuel bundle without refabrication. For each fuel option, the neutronics calculation of lattice was conducted to evaluate the allowable burnup and power distribution. The fuel cycle cost of each option was also computed to assess the economic justification. The result show that most tandem fuel cycle options considered in this study are technically feasible as well as economically viable.

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