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

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.715-731
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• 2021

Uncertainties are calculated for pressurized water reactor (PWR) spent nuclear fuel (SNF) characteristics. The deterministic code STREAM is currently being used as an SNF analysis tool to obtain isotopic inventory, radioactivity, decay heat, neutron and gamma source strengths. The SNF analysis capability of STREAM was recently validated. However, the uncertainty analysis is yet to be conducted. To estimate the uncertainty due to nuclear data, STREAM is used to perturb nuclear cross section (XS) and resonance integral (RI) libraries produced by NJOY99. The perturbation of XS and RI involves the stochastic sampling of ENDF/B-VII.1 covariance data. To estimate the uncertainty due to modeling parameters (fuel design and irradiation history), surrogate models are built based on polynomial chaos expansion (PCE) and variance-based sensitivity indices (i.e., Sobol' indices) are employed to perform global sensitivity analysis (GSA). The calculation results indicate that uncertainty of SNF due to modeling parameters are also very important and as a result can contribute significantly to the difference of uncertainties due to nuclear data and modeling parameters. In addition, the surrogate model offers a computationally efficient approach with significantly reduced computation time, to accurately evaluate uncertainties of SNF integral characteristics.

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

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.51-58
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• 2006

Since 2002, more than 400 PWR spent nuclear fuel assemblies have been transported and stored on-site using transport casks in order to secure the storage capacity of PWR spent nuclear fuel of Kori nuclear power plant. The complete on-site transport system, which includes KN-12 transport casks, the related equipment and transport vehicles, had been developed and provided. KN-12 transport casks were designed, fabricated and licensed in accordance with Korean and IAEA's transport regulations, and the related equipment was also provided in accordance with the related regulations. The on-site transport and storage operation using two KN-12 casks and the related equipment has been conducted, and the strict Quality Control and Radiation Safety Management through the whole process has been carried out so as to achieve the required safety and reliability of the on-site transport of spent nuclear fuel.

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

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.301.1-301.1
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• 2004

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.156-164
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• 2005

The challenges China is facing in energy security are briefly discussed. Then, the development of nuclear power in China in the first half of 21 st century is envisioned, and it is expected that Generation-3 PWR nuclear power plants (NPPs) would be the leading units of nuclear power in the coming $30\~40$ years. As part of the nuclear power program, the R&D work on nuclear fuel cycle is generally proposed.

• Nuclear Engineering and Technology
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• v.23 no.2
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• pp.165-173
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• 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
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• 2001.05a
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• pp.90.2-90
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• 2001

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2019.05a
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• pp.59-60
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• 2019