• Nuclear Engineering and Technology
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• v.43 no.3
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• pp.217-242
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• 2011

A variety of Generation III/III+ water-cooled reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world in efforts to solve the future energy supply shortfall. Thermal-hydraulics is recognized as a key scientific subject in the development of innovative reactor systems. Phase change by boiling and condensation in the reverse process is a highly efficient heat transport mechanism that accommodates large heat fluxes with relatively small driving temperature differences. This mode of heat transfer is encountered in a wide spectrum of nuclear systems,and thus it is necessary to determine the thermal limit of water-cooled nuclear energy conversion in terms of economic and safety. Such applications are being advanced with the introduction of new technologies such as nanotechnology. Here, we investigated newly-introduced nanotechnologies relevant to boiling and condensation in general engineering applications. We also evaluated the potential linkage between such new advancements and nuclear applications in terms of advanced nuclear thermal-hydraulics.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1804-1812
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• 2022

The aim of this paper is to assess the reliability and accuracy of the PSI standard method, used in many previous works, for the quantification of ND uncertainties in the SPERT-III RIA transient, by quantifying the discrepancy between the actual inserted reactivity and the original static reactivity worth and their associated uncertainties. The assessment has shown that the inherent S3K neutron source renormalization scheme, introduced before starting the transient, alters the original static reactivity worth of the transient CR and reduces the associated uncertainty due to the ND perturbation. In order to overcome these limitations, two additional methods have been developed based on CR adjustment. The comparative study performed between the three methods has showed clearly the high sensitivity of the obtained results to the selected approach and pointed out the importance of using the right procedure in order to simulate correctly the effect of ND uncertainties on the overall parameters in a RIA transient. This study has proven that the approach that allows matching the original static reactivity worth and starting the transient from criticality is the most reliable method since it conservatively preserves the effect of the ND uncertainties on the inserted reactivity during a RIA transient.

• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.53-62
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• 2009

A set of experiments has been conducted on the performance sensitivity of the passive residual heat removal system (PRHRS) for an advanced integral type reactor, SMART, by using a high temperature and high pressure thermal-hydraulic test facility, the VISTA facility. In this paper the effects of the opening delay of the PRHRS bypass valves and the closing delay of the secondary system isolation valves, and the initial water level and the initial pressure of the compensating tank (CT) are investigated. During the reference test a stable flow occurs in a natural circulation loop that is composed of a steam generator secondary side, a secondary system, and a PRHRS; this is ascertained by a repetition test. When the PRHRS bypass valves are operated 10 seconds later than the secondary system isolation valves, the primary system is not properly cooled. When the secondary system isolation valves are operated 10 or 30 seconds later than the PRHRS bypass valves, the primary system is effectively cooled but the inventory of the PRHRS CT is drained earlier. As the initial water level of the CT is lowered to 16% of the full water level, the water is quickly drained and then nitrogen gas is introduced into the PRHRS, resulting in the deterioration of the PRHRS performance. When the initial pressure of the PRHRS is at 0.1MPa, the natural circulation is not performed properly. When the initial pressures of the PRHRS are 2.5 or 3.5 MPa, they show better performance than did the reference test.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.728-744
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• 2024

The presence of hydrogen absorbed by zirconium-based cladding materials during reactor operation can trigger degradation mechanisms and endanger the rod integrity. Ensuring the durability of the rods in extended time-frames like dry storage requires anticipating hydrogen behavior using numerical modeling. In this context, the present paper describes a hydrogen post-processing tool for Falcon - HYPE, a PSI's in-house tool able to calculate hydrogen uptake, transport, thermochemistry, reorientation of hydrides and hydrogen-related failure criteria. The tool extracts all necessary data from a Falcon output file; therefore, it can be considered loosely coupled to Falcon. HYPE has been successfully validated against experimental data and applied to reactor operation and interim storage scenarios to present its capabilities.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.227-228
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.568-569
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.560-561
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• 2005

• Journal of computational fluids engineering
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• v.19 no.1
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• pp.64-72
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• 2014

This study aimed at assessing the analysis capability of thermal-hydraulic computer code, MARS for the behaviors of the core make-up tank (CMT). The sensitivity study on the nodalization to simulate the CMT was conducted, and the MARS calculations were compared with KAIST experimental data and RELAP5/MOD3.3 calculations. The 12-node model was fixed through a nodalization study to investigate the effect of the number of nodes in the CMT (2-, 4-, 8-, 12-, 16-node). The sensitivity studies on various parameters, such as water subcooling of the CMT, steam pressure, and natural circulation flow were done. MARS calculations were reasonable in the injection time and the effects of several parameters on the CMT behaviors even though the mesh-dependency should be properly treated for reactor applications.

• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.25-44
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• 2011

KAERI has been operating an integral effect test facility, ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation), for accident simulations of advanced PWRs. Regarding integral effect tests, a database for major design basis accidents has been accumulated and a Domestic Standard Problem (DSP) exercise using the ATLAS has been proposed and successfully performed. The ATLAS DSP aims at the effective utilization of an integral effect database obtained from the ATLAS, the establishment of a cooperative framework in the domestic nuclear industry, better understanding of thermal hydraulic phenomena, and an investigation of the potential limitations of the existing best-estimate safety analysis codes. For the first ATLAS DSP exercise (DSP-01), integral effect test data for a 100% DVI line break accident of the APR1400 was selected by considering its technical importance and by incorporating comments from participants. Twelve domestic organizations joined in this DSP-01 exercise. Finally, ten of these organizations submitted their calculation results. This ATLAS DSP-01 exercise progressed as an open calculation; the integral effect test data was delivered to the participants prior to the code calculations. The MARS-KS was favored by most participants but the RELAP5/MOD3.3 code was also used by a few participants. This paper presents all the information of the DSP-01 exercise as well as the comparison results between the calculations and the test data. Lessons learned from the first DSP-01 are presented and recommendations for code users as well as for developers are suggested.

• Nuclear Engineering and Technology
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• v.43 no.3
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• pp.257-270
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• 2011

A set of reflood tests has been performed using ATLAS, which is a thermal-hydraulic integral effect test facility for the pressurized water reactors of APR1400 and OPR1000. Several important phenomena were observed during the ATLAS LBLOCA reflood tests, including core quenching, down-comer boiling, ECC bypass, and steam binding. The present paper discusses those four topics based on the LB-CL-11 test, which is a best-estimate simulation of the LBLOCA reflood phase for APR1400 using ATLAS. Both homogeneous bottom quenching and inhomogeneous top quenching were observed for a uniform radial power profile during the LB-CL-11 test. From the observation of the down-comer boiling phenomena during the LB-CL-11 test, it was found that the measured void fraction in the lower down-comer region was relatively smaller than that estimated from the RELAP5 code, which predicted an unrealistically higher void generation and magnified the downcomer boiling effect for APR1400. The direct ECC bypass was the dominant ECC bypass mechanism throughout the test even though sweep-out occurred during the earlier period. The ECC bypass fractions were between 0.2 and 0.6 during the later test period. The steam binding phenomena was observed, and its effect on the collapsed water levels of the core and down-comer was discussed.