• Proceedings of the Korean Radioactive Waste Society Conference
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• 2006.11a
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• pp.89-90
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• 2006

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.511.2-511.2
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.114-114
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.516.2-516.2
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.120-120
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• 2003

• 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.37 no.6
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• pp.575-586
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• 2005

In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a loss of coolant accident (LOCA) with coincident loss of emergency core cooling (LOECC), as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.901-910
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• 2021

KHNP recently has obtained the approval for the commercialization of the modified 37-element (or 37 M) fuel bundle and now is loading the 37 M fuel bundles in CANDU-6 reactors in KOREA. One of the main issues for approval was the burnup limit. Due to CANDU design and neutronic characteristics, there was no specific burnup restriction of a fuel bundle. The absence of a burnup limit does not mean that a fuel bundle can stay in the CANDU reactor without a time limit. However, the regulator requested traditional design values as well as the burnup limit reflecting the computer code uncertainty. The method for the PWR burnup limit was not applied to the CANDU fuel bundle. Since there was no approved methodology to build the burnup limit with uncertainties, KHNP introduced a Monte-Carlo method coupled with a 95/95 approach to determine the conservative burnup limit from the viewpoint of the centerline temperature, internal pressure, strain measurement deviation. Moreover, to consider the uncertainties of various computing models, a converted power uncertainty was introduced. This paper presents the methodology and puts forward the limits on burnup, evaluated for each of the existing and modified fuel bundles, in consideration of the pressure tube aging condition.

• Nuclear Engineering and Technology
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• v.43 no.3
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• pp.301-308
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• 2011

The diametral creep of pressure tubes (PTs) in CANDU (CANada Deuterium Uranium) reactors is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS). PT diametral creep leads to diametral expansion, which affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux (CHF). The CHF is a major parameter determining the critical channel power (CCP), which is used in the trip setpoint calculations of regional overpower protection (ROP) systems. Therefore, it is essential to predict PT diametral creep in CANDU reactors. PT diametral creep is caused mainly by fast neutron irradiation, temperature and applied stress. The objective of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. The linear model was optimized using a genetic algorithm and was devised based on a bundle position because it is expected that each bundle position in a PT channel has inherent characteristics. The proposed BPLM for predicting PT diametral creep was confirmed using the operating data of the Wolsung nuclear power plant in Korea. The linear model was able to predict PT diametral creep accurately.

• Nuclear Engineering and Technology
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• v.45 no.1
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• pp.29-40
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• 2013

Two different kinds of nuclear power plants produce a substantial amount of spent fuel annually in Korea. According to the current projection, it is expected that around 60,000 MtU of spent fuel will be produced from 36 PWR and APR reactors and 4 CANDU reactors by the end of 2089. In 2006, KAERI proposed a conceptual design of a geological disposal system (called KRS, Korean Reference disposal System for spent fuel) for PWR and CANDU spent fuel, as a product of a 4-year research project from 2003 to 2006. The major result of the research was that it was feasible to construct a direct disposal system for 20,000 MtU of PWR spent fuels and 16,000 MtU of CANDU spent fuel in the Korean peninsula. Recently, KAERI and MEST launched a project to develop an advanced fuel cycle based on the pyroprocessing of PWR spent fuel to reduce the amount of HLW and reuse the valuable fissile material in PWR spent fuel. Thus, KAERI has developed a geological disposal system for high-level waste from the pyroprocessing of PWR spent fuel since 2007. However, since no decision was made for the CANDU spent fuel, KAERI improved the disposal density of KRS by introducing several improved concepts for the disposal canister. In this paper, the geological disposal systems developed so far are briefly outlined. The amount and characteristics of spent fuel and HLW, 4 kinds of disposal canisters, the characteristics of a buffer with domestic Ca-bentonite, and the results of a thermal design of deposition holes and disposal tunnels are described. The different disposal systems are compared in terms of their disposal density.