• Nuclear Engineering and Technology
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• v.25 no.4
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• pp.561-569
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• 1993

DOT4.2-QAD-CG coupling method was used to analyze the dose rates outside the side and the bottom shield system of CANDU 6 plant. The average dose rates at the main airlock and the new fuel loading area are approximately 6 $\mu$Sv/h as it is required. The calculated dose rates have a good agreement with the measurements at the operating CANDU 6 plant. The method used in this paper can be applied to the radiation shielding analysis of Wolsong 2, 3, and 4 CANDU 6 type plants which will be constructed in the near future.

• Journal of Radiation Protection and Research
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• v.20 no.1
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• pp.53-64
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• 1995

The radiation fields following the large loss of coolant accident (LOCA) have been assessed for the vital areas in the service building of Wolsong 2, 3, and 4 nuclear power plants. The ORIGEN2 code was used in calculating the fission product inventories in the fuel. The source terms were based upon the activity released following the dual failure accident scenario, i.e., a LOCA followed by impaired emergency core cooling (ECC). Configurations of the reactor building, the service building, and the ECC system were constructed for the QAD-CG calculations. The dose rates and the time-integrated doses were calculated for the time period of upto 90 days after the accident. The results showed that the radiation fields in the vital access areas were found to be sufficiently low. Some areas however showed relatively high radiation fields that may require limited access.

• Journal of Radiation Protection and Research
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• v.10 no.2
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• pp.148-154
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• 1985

Radiation shield design for a shipping cask, KSC-1, was evaluated to verify that the cask can be used in the transportation of a spent fuel assembly discharged from KNU 5 & 6. Radiation source term of the spent fuel assembly was calculated with the computer program ORIGEN-79, QAD-CG, ANISN-KA and DOT 3.5 codes Were used in the shielding calculations and the nuclear cross section data needed was extracted from the DLC-23/CASK library. It is concluded that KSC-1 shipping cask satisfies the requirements specified in the relevant regulations under normal conditions of transport and under accident conditions in transport.

• Nuclear Engineering and Technology
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• v.21 no.2
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• pp.63-72
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• 1989

An analysis for the gamma radiation fields in the research reactor MAPLE-X10 facility has been peformed under the assumption of partial loss of reactor and service pool water to assess the safety from the view point of design. Four photon source terms considered in the analysis were calculated using the ORIGEN-S code. Gamma dose rate calculations over the reactor and service pools during the water-loss accident conditions were performed using QAD-CG code. MCNP code (Monte Carlo Neuron and Photon Transport code), also, was used to assess the scattered radiation fields away from the pools, which is appropriate for calculating the scattered photon dose rates outside of the solid angle subtended by the source and pool walls.

• Journal of Radiation Protection and Research
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• v.23 no.3
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• pp.175-184
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• 1998

In order to calculate dose rates from steam generators to be replaced from Kori unit 1 in 1998, radionuclide inventories inside steam generator were evaluated from smear test results and measured dose rates from S/G tubes withdrawn for the metallographical examination of damaged tubes. Based on the inventories, contact dose rates and dose rates at 1 m from the surface of a steam generator were calculated using the QAD-CG computer code. Contact dose rates ranged from 11.5 mR/hr at the bottom of channel head to 37.7 mR/hr at the middle of shell barrel, and showed no significant difference with dose rates at 1 m from the surface of steam generator. Shielding effects of lead and carbon steel were compared to provide basic shielding data. Lead shield showed excellent shielding effects. Dose rate at 1 m from the middle of S/G shell barrel decreased from 38.6 mR/hr to 15.5 mR/hr with the lead shield of 2 mm thickness. However, carbon steel showed a poor shielding effect even with the thickness of 2.0 cm. This can be explained with the great differences in the attenuation effect and buildup factor between lead and carbon steel for low energy photons.