• Nuclear Engineering and Technology
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• v.23 no.4
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• pp.444-455
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• 1991

An accurate and fast running NEDAR model for calculating radial power profile throughout fuel life in both solid and annular pellets for existing and advanced CANDU-PHWR-fuel was developed in this work. This model contains resultant flux depression equations and neutron depression data tables which have been developed for CANDU-PHWR fuel of pellet with the diameter 8.0 to 19.5 mm and enrichment 0.71-6.0 wt % U-235, over a bumup range of 0 to 840 MWh /kgU (35000 MWD/T). In order to obtain the neutron flux distribution in the fuel pellet, the CE-HAMMER physics code was run for a neutron flux spectrum appropriate to a CANDU-PHWR to give predictions of radial power profile for several ranges of fuel design parameters. The results, which were calculated by the CE-HAMMER physics code, were fitted to an equation which is solved in terms of Bessel and exponential functions in order to obtain the parameters, $textsc{k}$, $\beta$ and λ in the resultant equation. The present NEDAR model produce a radial profile which, when normalized to unity at the pellet surface, is slightly higher than the profile of the original ELESIM data table. The predictions of the fission gas release by KAFEPA-NEDAR are in slightly better agreement with the experiments than those of ELESIM. The NEDAR model described in this study has been shown to provide an effective, reliable, and accurate method for determining radial power profiles in CANDU-PHWR fuel rods without incurring a significant increase in computing time.

• Nuclear Engineering and Technology
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• v.17 no.3
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• pp.185-192
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• 1985

The fissile content of PWR spent fuel is higher than that of natural uranium which is normal fuel for CANDU type reactor. Investigated are the concepts of PWR spent fuel utilization in CANDU type reactor to diversify uranium resource and partially to solve storage problems of PWR spent fuel being gradually accumulated. Nuclear characteristics of uranium-plutonium mixed oxide fuel loaded in CANDU type reactor are analysed using the WIMS/D computer code. In this study, analyses are solely carried out upon the current CANDU type reactor design without changingany reactivity control devices.

• Nuclear Engineering and Technology
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• v.10 no.4
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• pp.203-213
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• 1978

A combination of computer codes such as LATREP, HWR, AXAV and CITATION is utilized in an attempt to analyze the nuclear design characteristics of the CANDU-PHWR of the Wolsung Unit 1. The major nuclear properties to be computed are tile lattice properties of CANDU fuel channel and the core channel power distribution. The computed results are compared with the PSR documentation for the Wolsung reactor. The observed discrepancies between our computation and the PSB values are discussed in terms of incomplete information on the description of the core configuration in the PSR and the different calculation methods.

• Nuclear Engineering and Technology
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• v.17 no.2
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• pp.98-104
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• 1985

A number of reactivity devices are distributed in a CANDU-PHWR core to control the power distribution and excess reactivity. The effects of these devices are represented by incremental cross sections in core analysis. The incremental cross sections are generated by the SUPERCELL code using the two-group constant set calculated by the lattice code, WIMS. The incremental cross sections are then assessed for adjusters and zone controller by core simulation. Reactivity worth and channel powers are compared to the reference values. The deviation of reactivity worth and the maximum channel power are less than 0.97% and 0.6%, respectively, for the initial and equilibrium core.

• Nuclear Engineering and Technology
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• v.12 no.1
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• pp.45-53
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• 1980

• Corrosion Science and Technology
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• v.20 no.3
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• pp.118-128
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• 2021

The oxide layer in samples taken from an irradiated Zr-2.5Nb pressure tube from a CANDU-PHWR reactor was analyzed using electron probe microanalysis (EPMA). The examined tube had been exposed to temperatures ranging from 264 to 306 ℃ and a neutron fluence of 8.9 × 10²¹ n/cm² (E > 1 MeV) for the maximum 10 effective full-power years in a nuclear power plant. Measuring oxide layer thickness generally employs optical microscopy. However, in this study, analysis of the oxide layer from the irradiated pressure tube components was undertaken through X-ray image mapping obtained using EPMA. The oxide layer characteristics were analyzed by X-ray image mapping with 256 × 256 pixels using EPMA. In addition, the slope of the oxide layer was measured for each location. A particular advantage of this study was that backscattered electrons and X-ray image mapping were obtained at a magnification of 9,000 when 20 kV volts and 30 uA of current were applied to radiation-shielded EPMA. The results of this study should usefully contribute to the study of the oxide layer properties of various types of metallic materials irradiated by high radiation in nuclear power plants.

• Nuclear Engineering and Technology
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• v.9 no.3
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• pp.151-163
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• 1977

Being based on the Segal method, calculation was carried out for the natural uranium nuclear fuel cost with Zircaloy-4 cladding having design Parameters of Wolsung Nuclear Power Plant, CANDU-PHWR (Unit 1) , currently under construction in Korea aiming at its completion in 1982. An attempt was also made for tile sensitivity analysis of each fuel component; j. e., depreciation of fuel manufacturing plant caused by its life time, its load factor, production scale expansion of plant facilities, variations of construction and operating costs of fuel manufacturing plant, fluctuation of interest rates, extent of uranium ore price increases and effect of learning factor.

• Nuclear Engineering and Technology
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• v.15 no.2
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• pp.87-97
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• 1983

The WBURN (2-D, 2-group, coarse mesh) code is developed to analyze the equilibrium core characteristics of CANDU-PHWR. The equilibrium characteristics of Wolsung reactor computed by using WBURN are compared with the values given in the Wolsung FSR. The changes of equilibrium core characteristics caused by the variation of design parameters for operating conditions are also investigated. The numerical results indicate that the average discharge irradiation in the Wolsung reactor can be increased up to about 5%.

• Nuclear Engineering and Technology
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• v.11 no.4
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• pp.263-274
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• 1979

A statistical approach is used to investigate the relative economic advantages of pressurized water reactor (PWR) and pressurized heavy water reactor (PHWR-CANDU) nuclear power plants for hypothetical 900Mwe systems with the throwaway fuel cycle to be built in the Republic of Korea. Power cost is decomposed into the cost components related to the plant capital, operation and maintenance, working capital requirements and fuel cycle operation. The calculation of construction cost is performed with the modified version of computer code ORCOST, and the modified POWERCO-50 is used to evaluate the cost components. Most of economic parameters are treated as statistical variables, each being given with a certain range. Through a random sampling procedures. the probability histograms on unit plant construction costs and power generating costs are obtained. The power cost probability histograms of the PWR and the PHWR plants overlap considerably, and the power costs of two systems appear to be almost same with the PHWR power cost being 0.4mil1/kwh lower compared with 39.4 mills/kwh for the PWR plant (July 1986 US-dollars). When a construction period of PHWR plant is longer by one year than that of PWR plant, there is no difference in the unit power cost of two plants. This comparison leads to no definite conclusion on the cost advantage of the PWR plant versus the PHWR plant. We conclude that the selection issue of nuclear power plants in Korea still remains an open question and that future effort to solve this question should be made toward economic quantification of those factors such as technology transfer and localization.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.89.1-89
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• 2001