• Nuclear Engineering and Technology
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• v.35 no.4
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• pp.309-317
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• 2003

The mass balance of the unit processes of the Advanced spent fuel Conditioning Process was calculated to obtain basic information. Based on this mass balance, the changes in decay heat and radioactivity of the spent fuel due to the metallization in the high temperature molten salt system were estimated. The decay heat and the radioactivity were calculated by using the ORIGEN2 computer code, and the result showed that the decay heat and the radioactivity of the metallized spent fuel ingot were $24.27\%\;and\;24.24\%$, respectively, compared to those of oxide spent fuel.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.335-343
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• 2005

The Methodology of burnup calculation with EPMA test set up in this study. The spent fuel from PWR nuclear power plant was used as specimen. This $UO_2$ fuel with $3.2\%$ of enrichment had been irradiated up to 35,000 MWd/MTU(reference data). The burnup is very important factor for nuclear fuel to estimate all fuel behaviors in reactor. To measure amounts of fission products and actinides for the burnup calcualation, chemical analysis (destructive method) has been used but it mattes long experimental time and second radio-wastes. In this study, EPMA test was available to measure amount of fission products. Neodymium is able to be detected and quantified. It can be compared with the results from chemical analysis and ORIGEN-2 code calculation. Concentration of Nd from EPMA test showed good agreement with result of ORIGEN-2 code in the same burnup.

• Applied Microscopy
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• v.35 no.3
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• pp.113-119
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• 2005

The Methodology of burnup calculation with EPMA test set up in this study. The spent fuel from PWR nuclear power plant was used as specimen. This $UO_2$ fuel with 3.2% of enrichment had been irradiated up to 35,000 MWd/MTU. The burnup is very important factor for nuclear fuel to estimate all fuel behaviors in reactor. To measure amounts of fission products and actinides for the burnup calcualation, destructive method analysis has been used but it makes long experimental time and second radio-wastes. In this study, EPMA test was available to measure amount of fission products. Neodymium is able to be detected and quantified. It can be compared with the results from chemical analysis and ORIGEN-2 code calculation. Concentration of Nd from EPMA test showed good agreement with result of ORIGEN-2 code in the same burnup.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.660-665
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• 1998

The feasibility study on conceptual design methodology for accelerator-driven sodium-cooled sub-critical transmutation reactors has been conducted to optimize the design parameters from the scale laws and validates reactor performance with the integrated code system. A 1000 MWth sodium-cooled sub-critical transmutation reactor has been scale and verified through the methodology in this paper, which is referred to advanced Liquid Metal Reactor (ALMR). a Pb-Bi target material and a partitioned fuel are the liquid phases, and they are cooled by the circulation of secondary Pb-Bi coolant and by primary sodium coolant, respectively. Overall key design parameters are generated from the scale laws and they are improved and validated by the intergrated code system. Intergrated Code System (ICS) consist of LAHET, HMCNP, ORIGEN2, and COMMIX codes and some files. Through ICS the target region, the core region, and thermal-hydraulic related are analyzed once-through. Results of conceptual design are attached in this paper.

• Journal of Radiation Protection and Research
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• v.28 no.2
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• pp.109-116
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• 2003

Based on the necessity to evaluate the activation products inventory during decommissioning lot domestic nuclear power plants, a preliminary estimation of the activation products inventory for Kori unit 1, which is getting close to the end of lifetime, was carried out with ANISN and ORIGEN2 code. In order to calculate neutron nux using ANISN code, the reactor was divided into 9 zones from core to bioshield concrete for radial direction. Also :he cross-section of main nuclides were calibrated with neutron flux in the reactor pressure vessel(RPV) region. The results showed that 95 % of tile total radioactivity in RPV from reactor shutdown to 10 years came from the nuclides of $^{55}Fe,\;^{59}Ni,\;^{63}Ni\;and\;^{60}Co$. And the total radioactivity with cooling of more than 50 years after decommissioning was no more than 0.2 % of at the time of shutdown. Considering the weight of RPV is 210 tons, the total radioactivity of RPV reached to $5.25{\times}10^{6}GBq$ at shutdown time. As compared with the total radioactivity of bioshield concrete at reactor shutdown time, the radioactivity after tooling more than 10 years was below 1 %.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.847-852
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• 1995

A shielding analysis for KSC-7, the shipping cask for transporting the 7 PWR spent fuel assemblies, has been carried out. Radiation source term has been calculated on spent fuel with burnup of 50,000 MWD/MTU and 1.5 years cooling time by ORIGEN2 code. The shielding calculation for the cask has been made by using MCNP4A code with continuous cross section data library from ENDF/B-V. As a result of neutron dose rate analysis, another shielding calculational model on spent fuel shipping cask was provided which is using the Monte Carlo method.

• Advances in Energy Research
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• v.5 no.2
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• pp.91-105
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• 2017

Thermal hydraulic (TH) analysis of nuclear power reactors is utmost important. In this way, the numerical codes that preparing TH data in reactor core are essential. In this paper, a subchannel analysis of a Russian pressurized water reactor (WWER1000) core with enhanced numerical code is carried out. For this, in fluid domain, the mass, axial and lateral momentum and energy conservation equations for desired control volume are solved, numerically. In the solid domain, the cylindrical heat transfer equation for calculation of radial temperature profile in fuel, gap and clad with finite difference and finite element solvers are considered. The dependence of material properties to fuel burnup with Calza-Bini fuel-gap model is implemented. This model is coupled with Isotope Generation and Depletion Code (ORIGEN2.1). The possibility of central hole consideration in fuel pellet is another advantage of this work. In addition, subchannel to subchannel and subchannel to rod connection data in hexagonal fuel assembly geometry could be prepared, automatically. For a demonstration of code capability, the steady state TH analysis of a the WWER1000 core is compromised with Thermal-hydraulic analysis code (COBRA-EN). By thermal hydraulic parameters averaging Fuel Assembly-to-Fuel Assembly method, the one sixth (symmetry) of the Boushehr Nuclear Power Plant (BNPP) core with regular subchannels are modeled. Comparison between the results of the work and COBRA-EN demonstrates some advantages of the presented code. Using the code the thermal modeling of the fuel rods with considering the fission gas generation would be possible. In addition, this code is compatible with neutronic codes for coupling. This method is faster and more accurate for symmetrical simulation of the core with acceptable results.

• Analytical Science and Technology
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• v.16 no.2
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• pp.110-116
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• 2003

A study has been done on the separation of iodide from spent pressurized water reactor (PWR) fuels and its quantitative determination using ion chromatography. Spent PWR fuels were dissolved with mixed acid of nitric and hydrochloric acids (80 : 20 molL%) which can oxidize iodide to iodate to prevent it from be vaporized. After reducing ${IO_3}^-$ ­to $I_2$ in 2.5 M $HNO_3$ with $NH_2OH{\cdot}HCl$, Iodine was selectively separated from actinides and all other fission products with carbontetrachloride and back-extracted with 0.1 M $NaHSO_3$. Recovered iodide was determined using the ion chromatograph of which the column was installed in a glove box for the analysis of radioactive materials. In practice, spent PWR fuel with 42,000~44,000 MWd/MtU was analyzed and its quantity was compared to that calculated by burnup code, ORIGEN2. The agreement was achieved with a deviation of -8.3~-0.5% from the ORIGEN 2 data, $324.5{\sim}343.6{\mu}g/g$.

• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.27-35
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• 1993

A shielding analysis of the shipping cask for transporting the CANDU spent fuel bundles has been studied. Radiation source term has been calculated on spent fuel with burn-up of 7,800 MWD/MTU and 5 years cooling time by ORIGEN2 code. The shielding calculation for the cask capable of transporting 378 bundles of CANDU spent fuel has been made by use of 1-D ANISN and 2-D DOT 4.2 codes. As a result of analysis, the optimum shield thickness of cask was obtained. And it is proved that the safety in radiation shielding under normal transport and hypothetical accident conditions is confirmed to satisfy the allowable values specified in IAEA Safety Series No. 6 and the Korean Atomic Law.

• Journal of Radiation Protection and Research
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• v.31 no.2
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• pp.105-113
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• 2006

The oxidizing species are generated from the radiolysis of groundwater in the pore of buffer material around the canister used for the disposal of spent fuels. A mathematical model was introduced to calculate the cathodic current density induced by the oxidant around the canister, which determined the corrosion of carbon steel. An analytical solution was derived to get the cathodic current density in the cylindrical coordinate. The cathodic current densities from both the rectangular coordinate and cylindrical coordinate were compared with each other. The source terms and absorbed dose rate for the calculation of the radiolysis were calculated using the ORIGEN2 and MCNP computer code, respectively. The radius of the canister was determined with the new model in order to prevent the local corrosion. The results showed that the new solution made the cathodic current density around 25 % lower than the Marsh model.