• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1429-1435
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• 2021

The temperature distribution of an electromagnetic pump was analyzed with a flow rate of 1380 L/min and a pressure of 4 bar designed for the sodium thermo-hydraulic test in the Sodium Test Loop for Safety Simulation and Assessment-Phase 1 (STELLA-1). The electromagnetic pump was used for the circulation of the liquid sodium coolant in the Intermediate Heat Transport System (IHTS) of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) with an electric power of 150 MWe. The temperature distribution of the components of the electromagnetic pump was numerically analyzed to prevent functional degradation in the high temperature environment during pump operation. The heat transfer was numerically calculated using ANSYS Fluent for prediction of the temperature distribution in the excited coils, the electromagnet core, and the liquid sodium flow channel of the electromagnetic pump. The temperature distribution of operating electromagnetic pump was compared with cooling of natural and forced air circulation. The temperature in the coil, the core and the flow gap in the two conditions, natural circulation and forced circulation, were compared. The electromagnetic pump with cooling of forced circulation had better efficiency than natural circulation even considering consumption of the input power for the air blower. Accordingly, this study judged that forced cooling is good for both maintenance and efficiency of the electromagnetic pump.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1890-1901
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• 2022

We have assessed the applicability of the thermal-hydraulic system analysis code, SPACE, to a small modular reactor called SMART. For the assessment, the experimental data from a scale-down integral-test facility, SMART-ITL, were used. It was conformed that the SPACE code unrealistically calculates the safety injection flow rate through the CMT and SIT during a small-break loss-of-coolant experiment. This unrealistic behavior was due to the overprediction of interfacial heat transfer at the steam-water interface in a vertically stratified flow in the tanks. In this study, a special thermal-hydraulic component model has been developed to realistically calculate the interfacial heat transfer when a strong non-equilibrium two-phase flow is formed in the CMT or SIT. Additionally, we developed a special heat structure model, which analytically calculates the heat transfer from the hot steam to the cold tank wall. The combination of two models for the tank are called the special component model. We assessed it using the SMART-ITL passive safety injection system (PSIS) test data. The results showed that the special component model well predicts the transient behaviors of the CMT and SIT.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.902-907
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• 2023

To understand the eutectic reaction mechanism and the relocation behavior of the core debris is indispensable for the safety assessment of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). This paper addresses reaction products and their distribution of the eutectic melting/solidifying reaction of boron carbide (B₄C) and stainless-steel (SS). The influence of the existence of carbon on the B₄C-SS eutectic reaction was investigated by comparing the iron boride (FeB)-SS reaction by Raman spectroscopy with Multivariate Curve Resolution (MCR) analysis. The scanning electron microscopy with dispersive X-ray spectrometer was also used to investigate the elemental information of the pure metals such as Cr, Ni, and Fe. In the B₄C-SS samples, a new layer was formed between B₄C/SS interface, and the layer was confirmed that the formed layer corresponded to amorphous carbon (graphite) or FeB or Fe₂B. In contrast, a new layer was not clearly formed between FeB and SS interface in the FeB-SS samples. All samples observed the Cr-rich domain and Fe and Ni-rich domain after the reaction. These domains might be formed during the solidifying process.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2516-2533
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• 2023

Several practical methods for accelerating the depletion calculation in a GPU-based Monte Carlo (MC) code PRAGMA are presented including the multilevel spectral collapse method and the vectorized Chebyshev rational approximation method (CRAM). Since the generation of microscopic reaction rates for each nuclide needed for the construction of the depletion matrix of the Bateman equation requires either enormous memory access or tremendous physical memory, both of which are quite burdensome on GPUs, a new method called multilevel spectral collapse is proposed which combines two types of spectra to generate microscopic reaction rates: an ultrafine spectrum for an entire fuel pin and coarser spectra for each depletion region. Errors in reaction rates introduced by this method are mitigated by a hybrid usage of direct online reaction rate tallies for several important fissile nuclides. The linear system to appear in the solution process adopting the CRAM is solved by the Gauss-Seidel method which can be easily vectorized on GPUs. With the accelerated depletion methods, only about 10% of MC calculation time is consumed for depletion, so an accurate full core cycle depletion calculation for a commercial power reactor (BEAVRS) can be done in 16 h with 24 consumer-grade GPUs.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1125-1134
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• 2024

This paper presents a validation study of the subchannel analysis code SLTHEN used for the core thermal-hydraulic design of the Prototype Gen-IV sodium-cooled fast reactor (PGSFR). To assess the performance of the ENERGY model of SLTHEN, four liquid metal heating experiments conducted by ORNL, WARD, and KIT with hexagonal assemblies of wire-wrapped rod bundles were analyzed. These experiments were performed with 19-and 61-pin bundles and varying power distributions of axial and radial peaking factors up to 1.4 and 3.0, respectively. The coolant subchannel temperatures measured at different axial locations were compared with the SLTHEN predictions with the Novendstern, Chiu-Rohsenow-Todreas (CRT), and Cheng-Todreas (CT) correlations for flow split and mixing in wire-wrapped pin bundles. The results showed that the SLTHEN predicts the measured subchannel temperatures reasonably well with root-mean-square errors of ~10 % and maximum errors of ~20 %. It was also observed that the CRT and CT correlations consistently outperform the Novendstern correlation.

• 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.

• Nuclear Engineering and Technology
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• v.27 no.6
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• pp.877-887
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• 1995

YGN 3 is the first nuclear power plant to use the Core Protection Calculator (CPC) as the core protection system and the Core Operating Limit Supervisory System (COLSS) as the core monitor-ing system in Korea. The CPC is designed to provide on-line calculations of Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD) and to initiate reactor trip if the core conditions exceed the DNBR or LPD design limit. The COLSS is designed to assist the operator in implementing the Limiting Conditions for Operation (LCOs) in Technical Specifications for DNBR/Linear Heat Rate (LHR) margin, azimuthal tilt, and axial shape index and to provide alarm when the LCOs are reached. During YGN 3 initial startup testing, extensive CPC/COLSS related tests ore peformed to ver-ify the CPC/COLSS performance and to obtain optimum CPC/COLSS calibration constants at var, -ious core conditions. Most of test results met their specific acceptance criteria. In the case of missing the acceptance criteria, the test results ore analyzed, evaluated, and justified. Through the analysis and evaluation of each of the CPC/COLSS related test results, it can be concluded that the CPC/COLSS are successfully Implemented as designed at YGN 3.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.15-19
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• 2013

The auxiliary feedwater is an important to remove the heat from the reactor core when the main feedwater system is unavailable. In most initiating events in Probabilistic Safety Assessment(PSA), the operaton of this system is required to mitigate the accidents. For one of domestic nuclear power plants, a design change of a turbine-driven auxiliary feedwater pump(TD-AFWP), pipe, and valves in the auxiliary system is implemented due to the aging related deterioration by long time operation. This change includes the replacement of the TD-AFWP, the relocation of some valves for improving the system availability, a new cross-tie line, and the installation of manual valves for maintenance. The design modification affects the PSA because the system is critical to mitigate the accidents. In this paper, the safety effect of the change of the auxiliary feedwater system is assessed with regard to the PSA view point. The results demonstrate that this change can supply the auxiliary feedwater from the TD-AFWP in the accident with the motor-driven auxiliary feedwater pump(MD-AFWP) unavailable due to test or maintenance. In addition, the change of MOV's normal position from "close" to "open" can deliver the water to steam generator in the loss of offsite power(LOOP) event. Therefore, it is confirmed that the design change of the auxiliary feedwater system reduces the total core damage frequency(CDF).

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.204-214
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• 1993

Work on fracture mechanics has provided a technical basis for elimination of main coolant loop double ended guillotine breaks from the structural design basis of reactor coolant system. Without main coolant loop pipe breaks, the tributary pipe breaks must be considered as design bases until further fracture mechanics work could eliminate some of these breaks from design consideration. This paper determines the core support barrel shell responses for the 3 inch pressurizer spray line nozzle break which is expected to be the only inlet break remaining in the primary side after leak-before-break evaluation is extended to smaller size pipes in the near future. The responses are compared with those due to 14 inch safety injection nozzle break and main coolant loop pipe break. The results show that, when the leak-before-break concept is applied to the primary side piping systems with a diameter of 10 inches or over, the core support barrel shell responses due to pipe breaks in the primary side are negligible for the faulted condition design.

• Nuclear Engineering and Technology
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• v.19 no.4
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• pp.317-324
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• 1987

The purpose of this work is to present a spatial neutron kinetics computational scheme for the analysis of space-dependent transients like rod ejection accident of pressurized water reactors. In this work modified Borresen's 1.5 group coarse mesh scheme was formulated for the neutronic computation of the space-dependent transients and applied to the analysis of hypothetical rod ejection accident of KNU no. 1 PWR core at BOC, HZP. The computational accuracy of the modified Borresen's scheme is examined by comparing calculations for core power and control rod worths with startup core physics test results. Effects of such parameters as ejected rod worths and number of delayed neution group ell transient results as well as computational efficiency are also examined. OB this basis it is suggested that the modified Borresen's method is a useful scheme for the analysis of spatial neutron transients of PWR's.