• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2534-2546
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• 2023

Gas-cooled space reactor, which adopts He-Xe gas mixture as working fluid, is a better choice for megawatt power generation. In this paper, thermal-hydraulic characteristics of He-Xe gas mixture in 2×2 rod bundle wrapped with helical wires is numerically investigated. The velocity, pressure and temperature distribution of the coolant are obtained and analyzed. The results show that the existence of helical wires forms the vortexes and changes the velocity and temperature distribution. Hot spots are found at the contact corners between helical wires and fuel rods. The highest temperature of the hot spots reach 1600K, while the mainstream temperature is less than 400K. The helical wire structure increases the friction pressure drop by 20%-50%. The effect extent varies with the pitch and the number of helical wires. The helical wire structure leads to the reduction of Nusselt number. Comparing thermal-hydraulic performance ratios (THPR) of different structures, the THPR values are all less than 1. It means that gas-cooled space reactor adopting helical wires could not strengthen the core heat removal performance. This work provides the thermal-hydraulic design basis for He-Xe gas cooled space nuclear reactor.

• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.455-467
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• 2002

RELAP5/MOD3.2 was modified to perform the thermal-hydraulic safety analysis for HANARO transients. Several aspects of RELAP5/MOD3.2 were modified or replaced by new features to properly simulate the unique HANARO characteristics such as the finned fuel element, the cooling mechanisms by both plate type heat exchanger and the natural circulation. Especially, the heat transfer packages were modified to be more appropriate for the safety analysis and the heat transfer models were developed for the plate type heat exchanger as well as natural circulation through the pool water. This modified version of RELAP5/MOD3.2 is renamed as RELAP5/HANARO. The thermal-hydraulic simulations of the single fuel pin test and plate type heat exchanger were peformed to assess the realistic predicting capabilities of RELAP5/HANARO and compared with experimental results and manufacturer＇s data in this paper. In addition, the natural circulation experiment using the scaled bundle was simulated to validate the capability of RELAP5/HANARO. The simulation results show almost similar trend with experimental data. Therefore, it is proved that RELAP5/HANARO has a confidence to use for the safety analyses of HANARO.

• Nuclear Engineering and Technology
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• v.33 no.6
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• pp.638-651
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• 2001

The present study is to assess the applicability of the best-estimate thermal-hydraulic code, MARS 1.4, for the analysis of thermal-hydraulic behavior in PWRs during natural circulation conditions. The code simulates a natural circulation test, BETHSY test 4. la, which was conducted on the integral test facility of BETHSY. The test represented the cooling states of the primary cooling system under single-phase natural circulation, two-phase natural circulation and the reflux condensation mode with conditions corresponding to the residual power, 2% of the rated core power value and 6.8 MPa at the secondary system. Based on MARS 1.4 calculations, the major thermal-hydraulic behaviors during natural circulation are evaluated and the differences between the experimental data and calculated results are identified. The calculated results show generally good behavior with regard to the experimental results; the region of single-phase natural circulation is 100-92% of the initial mass inventory, two-phase natural circulation is 84-63 %, and the reflux condensation mode occurred below 58 %. U-tubes empty and the core uncovery are obtained at 39 % and 34 % of the initial mass inventory, respectively.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.469-474
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• 1997

To provide tile information ell severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore was performed to investigate the effect of thermal hydraulic design parameters ell severe accident progression of pressurized water reactors (PWRs), Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among mile parameters. For training. different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3&4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout(SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to tile other six parameters.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.287-298
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• 2007

As part of the $AP1000^{TM}$ pressurized water reactor design certification program, a series of integral systems tests of the nuclear steam supply system was performed at the APEX-1000 test facility at Oregon State University. These tests provided data necessary to validate Westinghouse safety analysis computer codes for AP1000 applications. In addition, the tests provided the opportunity to investigate the thermal-hydraulic phenomena expected to be important in AP1000 small-break loss of coolant accidents (SBLOCAs). The APEX-1000 facility is a 1/4-scale pressure and 1/4-scale height simulation of the AP1000 nuclear steam supply system and passive safety features. A series of eleven tests was performed in the APEX-1000 facility as part of a U.S. Department of Energy contract. In all, four SBLOCA tests representing a spectrum of break sizes and locations were simulated along with tests to study specific phenomena of interest. The focus of this paper is the SBLOCA tests. The key thermal-hydraulic phenomena simulated in the APEX-1000 tests, and the performance and interactions of the passive safety-related systems that can be investigated through the APEX-1000 facility, are emphasized. The APEX-1000 tests demonstrate that the AP1000 passive safety-related systems successfully combine to provide a continuous removal of core decay heat and the reactor core remains covered with considerable margin for all small-break LOCA events.

• Journal of the Korean Society of Safety
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• v.36 no.2
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• pp.111-119
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• 2021

Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3140-3153
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• 2022

A numerical feasibility study was conducted to investigate the thermal-hydraulic characteristics of a steam generator with corrugated plates for a small modular reactor. Accordingly, a one-dimensional thermal-hydraulic analysis code was developed based on the existing state-of-the-art thermal-hydraulic models and correlations for corrugated plate heat exchangers. Subsequently, the pressure loss, heat transfer, and instability characteristics of the steam generator with corrugated plates were investigated according to the chevron angle and mass flux. Additionally, the characteristics of rectangular and disk-type corrugated plate steam generators with equivalent heat transfer areas were analyzed. The steam generator with disk-type corrugated plates exhibited better performance in terms of pressure loss and heat transfer rate than the rectangular type. In addition, when the mass flux decreased from the onset of boiling points, reverse gradients of the total pressure change were observed in both types. Thus, it was confirmed that Ledinegg instability could occur in the steam generator with corrugated plates. However, it was dependent on the chevron angle, and the optimal chevron angle to minimize instability was 45° under the conditions of the present analysis.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1860-1873
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• 2022

Nuclear thermal-hydraulic system analysis codes have been developed to comprehensively model nuclear reactor systems to evaluate the safety of a nuclear reactor system. For analyzing complex systems with finite computational resources, system codes usually solve simplified fluid equations for coarsely discretized control volumes with one-dimensional assumptions and replace source terms in the governing equations with constitutive relations. Wall boiling heat transfer models are regarded as essential models in nuclear safety evaluation among many constitutive relations. The wall boiling heat transfer models of two widely used nuclear system codes, RELAP5 and TRACE, are analyzed in this study. It is first described how wall heat transfer models are composed in the two codes. By utilizing the same method described in Part 1 paper, heat fluxes from the two codes are compared under the same thermal-hydraulic conditions. The significant factors for the differences are identified as well as at which conditions the non-negligible difference occurs. Steady-state simulations with both codes are also conducted to confirm how the difference in wall heat transfer models impacts the simulation results.

• Journal of the Korean Society for Geothermal and Hydrothermal Energy
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• v.1 no.1
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• pp.1-8
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• 2005

Aquifer Thermal Energy Storage (ATES) can be a cost-effective and renewable geothermal energy source, depending on site-specific and thermohydraulic conditions. To design an effective ATES system having the effect of groundwater movement, understanding of thermohydraulic processes is necessary. The heat transfer phenomena for an aquifer heat storage are simulated by using FEFLOW with the scenario of heat pump operation with pumping and waste water reinjection in a two layered confined aquifer model. Temperature distribution of the aquifer model is generated, and hydraulic heads and temperature variations are monitored at the both wells during 365 days. The average groundwater velocities are determined with two hydraulic gradient sets according to boundary conditions, and the effect of groundwater flow are shown at the generated thermal distributions of three different depth slices. The generated temperature contour lines at the hydraulic gradient of 0.001 are shaped circular, and the center is moved less than 5 m to the direction of groundwater flow in 365 days simulation period. However at the hydraulic gradient of 0.01, the contour center of the temperature are moved to the end of east boundary at each slice and the largest movement is at bottom slice. By the analysis of thermal interference data between two wells the efficiency of the heat pump system model is validated, and the variation of heads is monitored at injection, pumping and no operation mode.

• Journal of the Korean Geotechnical Society
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• v.19 no.5
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• pp.59-67
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• 2003

In this study, the change in hydraulic conductivity according to the changes in the contact area, aperture, confining pressure and temperature was observed to improve the reliability of the analysis of underground water flow. Also, the mechanical and thermal properties of domestic crystalline rocks in a great depth were obtained. It was found that the averaged intial aperture ranged from 544.33${\mu}{\textrm}{m}$ to 898.62${\mu}{\textrm}{m}$ and it followed a log-normal distribution. The hydraulic conductivity decreased with the increase of normal stress on the fracture surface and the hydraulic conductivity decreased as temperature increased. The change in hydraulic conductivity was strongly correlated with the change in contact area. It was verified by experiments that hydraulic conductivity was inversely proportional to the contact area. The measured mechanical and thermal properties were very close to the existing typical properties of domestic granites.