• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.135-147
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• 2020

A methodology for evaluation of mechanical and thermal effects of localized non-axisymmetric oxidation in zircaloy claddings on LWR fuel reliability is proposed. To this end, the basic capabilities of the FALCON fuel behaviour code are used. Examples of methodology application to adjustment of selected operational limits for modern BWR fuel rods, to capture effects of the excess local oxidation, are presented. Specifically, the limiting rod internal pressure for the onset of cladding lift-off is reduced, depending on initial excess oxidation spot sizes. Also, the power limits for Anticipated Operational Occurrences are adjusted, to preclude fuel melting and cladding failure due to PCMI and PCI-SCC in the affected fuel rods.

• The Magazine of the Society of Air-Conditioning and Refrigerating Engineers of Korea
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• v.17 no.4
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• pp.499-507
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• 1988

The pressure suppression pool of BWR(Boiling Water Reactor) is subjected to hydrodynamic impact in the event of a LOCA(Loss of Coolant Accident). The pressure increase in the reactor dry cell would force the existing water of a vent pipe into the suppression pool. When the water is ejected through the pipe opening into the suppression pool, an abrupt downward force is transmitted to the suppression pool floor. Consequently, many structures installed within the pool must be able to withstand these forces. In order to determine the optimum safe locations of the pool structures, numerical analysis have been carried out to investigate the hydrodynamic behavior of the water jet. In the present analysis, a two-dimensional numerical model is utilized to solve transient flow equations.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1190-1197
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• 2018

Analysis of the GRSW-A model coupled into the FALCON code is extended by simulation of central void formation in fuel pellets due to high-temperature fuel restructuring. The extended calculation is verified against published, well-known experimental data. Good agreement with the data for a central void diameter in pellets of the rod irradiated in an Experimental Breeder Reactor is shown. The new calculation methodology is employed in comparative analysis of modern BWR fuel behavior under assumed high-power operation. The initial fuel porosity is shown to have a major effect on the predicted central void diameter during the operation in question. Discernible effects of a central void on peak fuel temperature and Pellet-Cladding Mechanical Interaction (PCMI) during a simulated power ramp are shown. A mitigating effect on PCMI is largely attributed to the additional free volume in the pellets into which the fuel can creep due to internal compressive stresses during a power ramp.

• Journal of information and communication convergence engineering
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• v.16 no.4
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• pp.203-212
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• 2018

In this article, a novel design of patch antenna with wide band characteristics is presented. The proposed antenna is having electrical dimensions of $0.14{\lambda}{\times}0.11{\lambda}$ (at lower initial frequency) and footprints of $150mm^2$. Structural parameters optimization shows 3.1-23.5 GHz frequency range for a (reflection coefficient) $S_{11}{\leq}-10dB$ and simulated gain 6.8 dB is obtained. An equivalent circuit model is proposed to get an insight view of antenna. Advanced Systems Design (ADS) simulation results are obtain which confirm the validity of proposed model. Degenerated foster canonical form has been used to explain the reactance and capacitive behavior idea of simulated proposed antenna's input impedance later on an equivalent circuit model and smith chart is also suggested. HFSS and CST have been used to analyze antenna behavior. The proposed antenna can be further used for microwave image detection applications.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.401-413
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• 2022

This paper provides a simple method by which to estimate the cross-section stress profiles for nozzles designed according to ASME Code Section III. Further, this method validates the effectiveness of earlier work performed by the authors on standard nozzles. The method requires only the geometric information of the pressure vessel and the attached nozzle. A PWR direct vessel injection nozzle, a PWR outlet nozzle, a PWR inlet nozzle and a BWR recirculation outlet nozzle are selected based on their corresponding specific designs, e.g., a varying nozzle radius, a varying nozzle thickness and an outlet nozzle boss. A cross-section stress profile comparison shows that the estimates are in good agreement with the finite element analysis results. Differences in stress intensity factors calculated in accordance with ASME BPVC Section XI Appendix G are discussed. In addition, a change in the dimensions of an alternate nozzle design relative to the standard values is discussed, focusing on the stress concentration factors of the nozzle inside corner.

• 제어로봇시스템학회:학술대회논문집
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• 2003.10a
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• pp.209-214
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• 2003

We developed a new type of human-sized BWR (biped walking robot), named KUBIR1 which is driven by the closed-chain type of actuator. A new type of the closed-chain actuator for the robot is developed, which is composed of the four-bar-link mechanism driven by the ball screw which has high strength and high gear ratio. Each leg of the robot is composed of 6 D.O.F joints. For front walking, three pitch joints and one roll joint at the ankle. In addition to this, one yaw joint for direction change, and another roll joint for balancing the body are attached. Also, the robot has two D.O.F joints of each hand and three D.O.F. for eye motion. There are three actuating motors for stereo cameras for eyes. In all, a 18 degree-of-freedom robot was developed. KUBIR1 was designed to walk autonomously by adapting small 90W DC motors as the robot actuators and batteries and controllers are on-boarded. The whole weight for Kubir1 is over 90Kg, and height is 167Cm. In the paper, the performance test of KUBIR1 will be shown.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.284-289
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• 1996

The predictions of the COBRA-IV-I code with the modified turbulent mixing and void drift models have been compared with the diabatic two-phase flow data on equilibrium quality. The turbulent mixing model based on an equal mass exchange of the existing COBRA-IV-I code has been modified to that based on an equal volume exchange between adjacent subchannels, and a void drift model has been newly incorporated in the code. To evaluate the performance of the equal volume exchange turbulent mixing model and the effects of the void drift model, the diabatic steam-water two-phase flow data obtained for the 9-rod bundle test under the typical operating conditions of the boiling water reactor(BWR) conducted by the General Electric (GE) were analyzed by the modified COBRA-IV-I code. The analysis indicates that the equal volume exchange turbulent mixing model with void drift predicts the observed two-phase flow data trends better than the equal mass exchange model, and to predict the correct data trends a more physically based void drift model need to be developed.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1005-1014
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• 2009

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the $21^{st}$ century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.343-352
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• 2014

Uncertainty and sensitivity analyses with respect to nuclear data are performed with depletion calculations for BWR and PWR fuel assemblies specified in the framework of the UAM-LWR Benchmark Phase II. For this, the GRS sampling based tool XSUSA is employed together with the TRITON depletion sequences from the SCALE 6.1 code system. Uncertainties for multiplication factors and nuclide inventories are determined, as well as the main contributors to these result uncertainties by calculating importance indicators. The corresponding neutron transport calculations are performed with the deterministic discrete-ordinates code NEWT. In addition, the Monte Carlo code KENO in multi-group mode is used to demonstrate a method with which the number of neutron histories per calculation run can be substantially reduced as compared to that in a calculation for the nominal case without uncertainties, while uncertainties and sensitivities are obtained with almost the same accuracy.

• Nuclear Engineering and Technology
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• v.39 no.5
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• pp.591-602
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• 2007

The results of Reactivity-Initiated Accidents (RIA) experiments have been analysed and the main variables affecting the fuel failure propensity identified. Fuel burn-up aggravates the mechanical loading of the cladding, while corrosion, or better the hydrogen absorbed in the cladding as a consequence of corrosion, may under some conditions make the cladding brittle and more susceptible to failure. Experiments point out that corrosion impairs the fuel resistance for RIA transient occurring at cold conditions, whereas there is no evidence of important embrittlement effects at hot conditions, unless the cladding was degraded by oxide spalling. A fuel failure threshold correlation has been derived and compared with experimental data relevant for BWR and PWR fuel. The correlation can be applied to both cold and hot RIA transients, account taken for the lower ductility at cold conditions and for the different initial enthalpy. It can also be used for non-zero power transients, provided that a term accounting for the start-up power is incorporated. The proposed threshold is easy to use and reproduces the results obtained in the CABRI and NSRR tests in a rather satisfactory manner. The behaviour of advanced PWR alloys and of MOX fuel is discussed in light of the correlation predictions. Finally, a probabilistic approach has been developed in order to account for the small scatter of the failure predictions. This approach completes the RIA failure assessment in that after determining a best estimate failure threshold, a failure probability is inferred based on the spreading of data around the calculated best estimate value.