• Journal of Radiation Protection and Research
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• v.39 no.4
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• pp.176-181
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• 2014

The object of this paper is to evaluate the fission product inventories and radiological doses in a non-LOCA event, based on the U.S. NRC's regulatory methodologies recommended by the TID-14844 and the RG 1.195. For choosing a non-LOCA event, one fuel assembly was assumed to be melted by a channel blockage accident. The Hanul nuclear power reactor unit 6 and the CE $16{\times}16$ fuel assembly were selected as the computational models. The burnup cross section library for depletion calculations was produced using the TRITON module in the SCALE6.1 computer code system. Based on the recently licensed values for fuel enrichment and burnup, the source term calculation was performed using the ORIGEN-ARP module. The fission product inventories released into the environment were obtained with the assumptions of the TID-14844 and the RG 1.195. With two kinds of source terms, the radiological doses of public in normal environment reflecting realistic circumstances were evaluated by applying the average condition of meteorology, inhalation rate, and shielding factor. The statistical analysis was first carried out using consecutive three year-meteorological data measured at the Hanul site. The annual-averaged atmospheric dispersion factors were evaluated at the shortest representative distance of 1,000 m, where the residents are actually able to live from the reactor core, according to the methodology recommended by the RG 1.111. The Korean characteristic-inhalation rate and shielding factor of a building were considered for a series of dose calculations.

• Nuclear Engineering and Technology
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• v.18 no.1
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• pp.17-26
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• 1986

A test is made for a method to determine reliable bias in the criticality safety analysis of spent fuel storage pool with turnup credit between the reference and rack criticality calculation methods. The spent fuel pool of Kori Unit 1 is conceptually redesigned to the most compact rack with turnup credit, and its multiplication factors are calculated depending on fuel enrichment and burnup, by the Monte Carlo code KENO-IV as a reference and by a two-dimensional collision probability code FATAC as a practical method. Then, the computed values with the help of the above two computer codes are compared to evaluate the bias and its trend in terms of multiplication factor on fuel enrichment and turnup. The result indicates that the bias can be determined with reliability basis but without any disadvantage in criticality safety margin compared with the conventional method.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.65-70
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• 1997

Computational benchmark calculations have been performed for CANDU DUPIC fuel lattice and core using a Monte Carlo code MCNP-4B with ENDF/B-V library. The eigenvalues of the DUPIC fuel lattice have been predicted by an integral transport code WIMS-AECL using ENDF/B-V library for different burnup steps and lattice conditions. The comparison has shown that the eigenvalues match those of MCNP-4B within 0.20% $\Delta$k difference between WIMS-AECL and MCNP-4B results. The calculation of a 2-dimensional CANDU core loaded with DUPIC fuel has shown that the eigenvalue predicted by a diffusion code RFSP using lattice parameters generated by WIMS-AECL matches that of MCNP-4B within 0.12%Δk and the largest bundle power prediction error is around 7.2%.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.41-47
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• 1995

A parametric study bas been performed for the various refueling schemes of CANDU 6 reactor loaded with reference DUPIC fuel. The optimum discharge burnup was determined such that the peak bundle power is minimized for the equilibrium core. Based on the results of instantaneous core calculation using patterned random age distributions, it was decided to perform the refueling simulations only for 2-bundle and 4-bundle shift refueling schemes. The 600 FPD simulation has shown that the operational margins of the channel and bundle power to the license limits are 7.9% and 17.1%, respectively, for 2-bundle shift refueling scheme. The 4-bundle shift refueling scheme also satisfies the license limits and the operational margins of the channel and bundle power are 7.1% and 9.8%, respectively. The result of refueling simulation indicate the possibility of using reference DUPIC fuel in current CANDU 6 reactor.

• Proceedings of the Korean Vacuum Society Conference
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• 2012.02a
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• pp.582-582
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• 2012

Characteristics of a fusion-driven transmutation reactor was investigated. A compact reactor concept is desirable from an economic viewpoint. For the optimal design of a reactor, a radial build of reactor components has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor components. In a transmutation reactor, design of blanket and shield play a key role in determining the size of a reactor; the blanket should produce enough tritium for tritium self-sufficiency, the transmutation rate of waste has to be maximized, and the shield should provide sufficient protection for the superconducting toroidal field (TF) coil. To determine the radial build of the blanket and the shield, not only a radiation transport analysis but also a burnup calculation were coupled with the system analysis and it allowed the self-consistent determination of the design parameters of a transmutation reactor.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.636-643
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• 1995

The accurate determination of the fuel-cladding gap conductance as functions of rod burnup and power level may be a key to the design and safety analysis of a reactor. The incorporation of a sophisticated gap conductance model into nuclear design code for computing thermal hydraulic feedback effect has not been implemented mainly because of computational inefficiency due to complicated behavior of gap conductance. To avoid the time-consuming iteration scheme, simplification of the gap conductance model is done for the current design model. The simplified model considers only the heat conductance contribution to the gap conductance. The simplification is made possible by direct consideration of the gas conductivity depending on the composition of constituent gases in the gap and the fuel-cladding gap size from computer simulation of representative power histories. The simplified gap conductance model is applied to the various fuel power histories and the predicted gap conductances are found to agree well with the results of the design model.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.21-27
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• 1996

KAERI has recently developed the nuclear design code MASTER for the application to reactor physics analyses for pressurized water reactors. Its neutronics model solves the space-time dependent neutron diffusion equations with the advanced nodal methods. The major calculation categories of MASTER consist of microscopic depletion, steady-state and transient solution, xenon dynamics, adjoint solution and pin power and burnup reconstruction. The MASTER validation analyses, which are in progress aiming to submit the Uncertainty Topical Report to KINS in the first half of 1996, include global reactivity calculations and detailed pin-by-pin power distributions as well as in-core detector reaction rate calculations. The objective of this paper is to give an overall description of the CASMO/MASTER code system whose verification results are in details presented in the separate papers.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.73-80
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• 2011

The radiation shielding analysis for a Burnup-credit (BUC) cask designed under the management of Korea Radioactive Waste Management Corporation (KRMC) was performed to examine the contribution of each radiation source affecting dose rate distribution around the cask. Various radiation sources, which contain neutron and gamma-ray sources placed in active fuel region and the activation source, and imaginary nuclear fuel were all considered in the MCNP calculation model to realistically simulate the actual situations. It was found that the maximum external and surface dose rates of the spent fuel cask were satisfied with the domestic standards both in normal and accident conditions. In normal condition, the radiation dose rate distribution around the cask was mainly influenced by activation source ($^{60}Co$ radioisotope); in another case, the neutron emitted in active fuel region contributed about 90% to external dose rate at 1m distance from side surface of the cask. Besides, the contribution level of activation source was dramatically increased to the dose rates in top and bottom regions of the cask. From this study, it was recognized that the detailed investigation on the radiation sources should be performed conservatively and accurately in the process of radiation shielding analysis for a BUC cask.

• Journal of Radiation Protection and Research
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• v.30 no.2
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• pp.61-67
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• 2005

The criticality analysis method using HELIOS-MASTER code system, which is the nuclear core analysis code system, was developed for the spent fuel storage of SMART-P reactor. We generated the macroscopic cross section of the geometric model with HELIOS and estimated the criticality of the 3-dimensional model with MASTER for SMART-P spent fuel storage. The validity of criticality analysis method for SMART-P spent fuel storage with the HELIOS-MASTER code system by 3-D MCNP calculation was also verified. The result of the criticality analysis with the HELIOS-MASTER code system is more conservative than that with the MCNP and the accuracy of this result is within the range of an allowable error. Because HELIOS-MASTER can perform the 3-D depletion calculation lot a spent fuel storage, it will be useful to perform the criticality analysis including a burnup credit in future.