• Nuclear Engineering and Technology
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• v.13 no.1
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• pp.1-11
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• 1981

The KINS (KAERI-Improved Nodal Simulation) program, a three-dimensional nodal simulation code for pressurized water reactors, has been developed and benchmarked against the first cycle of the Kori-1 reactor. The KINS program is based on the computational model used in FLARE code and has been modified to represent the PWR characteristics more explicitly. The critical boron concentration and three-dimensional power distribution at the beginning of life hot zero power have been calculated and compared with the operating data. A three-dimensional depletion calculation at the intervals of 1000 MWD/MTU turnup steps has been performed. As the result of comparison, our calculation is shown to be in excellent agreement with the operating data. It is displayed that, incorporated with the computing time, the KINS program is an effective and powerful tool for PWR core management.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.21 no.1
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• pp.180-187
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• 1997

A formula for estimating the elastic stiffness of TW-HDS with a uniformly tapered width from w$_{0}$ to w$_{1}$ over the length, has been analytically derived based on Euler beam theory and Castigliano's theorem. Elastic stiffnesses of the TW-HDSs designed in the same dimensional design spaces as the KOFA HDSs have been estimated from the derived formula, in addition, a sensitivity study on the elastic stiffness of the TW-HDSs has been carried out. Analysis results show that elastic stiffnesses of the TW-HDSs have been by far higher than those of the KOFA HDSs, and that, as the effects of axial and shear force on the elastic stiffness have been 0.15-0.21%, most of the elastic stiffness is attributed to the bending moment. As a result of sensitivity analysis, the elastic stiffness sensitivity at each design variable is quantified and design variables having remarkable sensitivity are identified. Among the design variables, leaf thickness is identified as that of having the most remarkable sensitivity of the elastic stiffness.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.21 no.8
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• pp.1276-1290
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• 1997

The previous elastic stiffness formulas of leaf type holddown spring assemblies(HDSs) have been corrected and extended to be able to consider the point of taper runout for the TT-HDS and all the strain energies for both the TT-HDS and the TW-HDS based on Euler beam theory and Castigliano'stheorem. The elastic stiffness sensitivity of the leaf type holddown spring assemblies was analyzed using the derived elastic stiffness formulas and their gradient vectors obtained from the mid-point formula. As a result of the sensitivity analysis, the elastic stiffness sensitivity at each design variable is quantified and design variables having remarkable sensitivity are identified. Among the design variables, leaf thickness is identified as that of having the most remarkable sensitivity of the elastic stiffness. In addition, it was found that the sensitivity of the leaf type HDS's elastic stiffness is exponentially correlated to the leaf thickness.

• Journal of Energy Engineering
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• v.13 no.4
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• pp.242-258
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• 2004

The duplex burnable poison absorbers concept was suggested by Korea Atomic Energy Research Institute. This BP rod is composed of inner region of natural U-Gd$_2$O$_3$ and outer shell of enriched UO$_2$-Er$_2$O$_3$. It is expected that this burnable absorber has same reactivity control capability with gadolinia burnable absorber used in extened fuel cycle. In order to evaluate the nuclear feasibility of duplex BPs, the nuclear design characteristics were compared with that of four types of burnable absorbers; gadolinia, erbia, IFBA, dysprosia duplex BP on 24 months fuel cycle for Korean Standard Nuclear Power plants. According to the evaluation results of nuclear characteristics, the duplex BPs were better than other BPs on k-infinitives, reactivity holddown worth (RHW), pin power peaking and moderator temperature coefficient (MTC). The possibility of nuclear core design was also confirmed based on the optimized fuel assemblies which were searched for a sensitivity analysis. Characteristics of core design with duplex BPs was compared with that of reference core with gadolinia BPs for cycle length, power peaking and MTC. The duplex BP core had a little longer cycle length by 4 to 7 days because of increased amount of fissile in enriched uranium at the outer shell of duplex BP In case of power peaking F$\_$Q/ of duplex BP core was reduced from 1.5773 to 1.5335. MTC was also less -0.48 pcm/C than that of reference core. Finally, evaluation of fuel cycle economy was performed for the manufacturing feasibility test and fuel cost evaluation with duplex BPs. Fuel cycle economy of duplex BP core almost was equivalent with that of gadolinia BP core.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.347-353
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• 2010

Source terms of metal waste comprising a spent fuel assembly are relatively important when the spent fuel is pyroprocessed, because cesium, strontium, and transuranics are not a concern any more in the aspect of source term of permanent disposal. In this study, characteristics of radiation source terms for each structural component in spent fuel assembly was analyzed by using ORIGEN-S with a assumption that 10 metric tons of uranium is pyroprocessed. At first, mass and volume for each structural component of the fuel assembly were calculated in detail. Activation cross section library was generated by using KENO-VI/ORIGEN-S module for top-end piece and bottom-end piece, because those are located at outer core with different neutron spectrum compared to that of inner core. As a result, values of radioactivity, decay heat, and hazard index were reveled to be $1.40{\times}10^{15}$ Bequerels, 236 Watts, $4.34{\times}10^9m^3$-water, respectively, at 10 years after discharge. Those values correspond to 0.7 %, 1.1 %, 0.1 %, respectively, compared to that of spent fuel. Inconel 718 grid plate was shown to be the most important component in the all aspects of radioactivity, decay heat, and hazard index although the mass occupies only 1 % of the total. It was also shown that if the Inconel 718 grid plate is managed separately, the radioactivity and hazard index of metal waste could be decreased to 20~45 % and 30~45 %, respectively. As a whole, decay heat of metal waste was shown to be negligible in the aspect of disposal system design, while the radioactivity and hazard index are important.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.35 no.1
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• pp.23-31
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• 2011

At supercritical pressure, the physical properties of fluid change substantially and the heat transfer at a temperature similar to the critical or pseudo-critical temperature improves considerably; however, the heat transfer may deteriorate due to a sudden increase in the wall temperature at a certain condition of a mass and heat flux. In this study, the heat transfer rates in $CO_2$ flowing vertically upward and downward in a circular tube with a diameter of 4.57 mm under various conditions were calculated by measuring the temperature of the outer wall of the tube. The published heat transfer correlations were analyzed by comparing their prediction values with 7,250 experimental data. By introducing a buoyancy parameter, a heat transfer correlation, which could be applied only to a normal heat transfer regime, was extended such that it can be applied to regime of heat transfer deterioration. The published criteria for heat transfer deterioration were evaluated against the conditions obtained from the experiment in this study.

• Resources Recycling
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• v.21 no.5
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• pp.18-30
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• 2012

Zirconium is one of the most important material used as cladding of fuel rods in nuclear reactors because of its high dimensional stability, good corrosion resistance and especially low neutron-absorbing cross section. However, Hf free nuclear grade Zr sponge is commercially produced by only three countries including USA, France and Russia. So, Zr has been thoroughly managed as a national strategic material in Korea. Most of the zirconium is used for Korean nuclear industry as nuclear fuel cladding materials manufactured from Hf free Zr alloy raw material. Also, there are some other applications such as alloying element and detonator. In this review, zirconium production and recycling technologies have been reviewed and current industrial status was also analyzed. And recent achievements in innovative reduction technologies such as electrolytic reduction process and molten oxide electrolysis were also introduced.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.562-570
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• 1995

The recent ABB/CE(Asea Brown Boveri Combustion Engineering) type pressurized oater reactor-s have the on-line monitoring system, i.e., the COLSS(core operating limit supervisory system), to prevent the specified acceptable fuel design limits from being violated during normal operation and anticipated operational occurrences. One of the main functions of COLSS is the on-line monitoring of the DNB(departure from nucleate boiling) overpower margin by calculating the MDNBR(mini-mum DNB ratio) for the measured operating condition at every second. The CETOP-D model, used in the MDNBR calculation of COLSS, is benchmarked conservatively against the TORC mod-el using an inlet flow factor of hot assembly in CETOP-D as an adjustment factor for TORC. In this study, a technique to optimize the CETOP-D inlet flow factor has been developed by elim-inating the excessive conservatism in the ABB/CE's. A correlation is introduced to account for the actual variation of the CETOP-D inlet flow factor within the core operating limits. This technique was applied to the core operating range of the YongGwang Units 3&4 Cycle 1, which results in the increase of 2％ in the DNB overpower margin at the normal operating condition, compared with that from the ABB/CE method.

• Computational Structural Engineering
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• v.7 no.3
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• pp.95-103
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• 1994

The safety related components in the nuclear power plant should be designed to withstand the seismic load. Among these components the integrity of reactor internals under earthquake load is important in stand points of safety and economics, because these are classified to Seismic Class I components. So far the modelling methods of reactor internals have been investigated by many authors. In this paper, the dynamic behaviour of reactor internals of Yong Gwang 1&2 nuclear power plants under SSE(Safe Shutdown Earthquake) load is analyzed by using of the simpled Global Beam Model. For this, as a first step, the characteristic analysis of reactor internal components are performed by using of the finite element code ANSYS. And the Global Beam Model for reactor internals which includes beam elements, nonlinear impact springs which have gaps in upper and lower positions, and hydrodynamical couplings which simulate the fluid-filled cylinders of reactor vessel and core barrel structures is established. And for the exciting external force the response spectrum which is applied to reactor support is converted to the time history input. With this excitation and the model the dynamic behaviour of reactor internals is obtained. As the results, the structural integrity of reactor internal components under seismic excitation is verified and the input for the detailed duel assembly series model could be obtained. And the simplicity and effectiveness of Global Beam Model and the economics of the explicit Runge-Kutta-Gills algorithm in impact problem of high frequency interface components are confirmed.