• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.248.2-248
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.187-187
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.199-199
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.130-130
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.131-131
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• 2003

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.63-77
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• 1994

The objective of this work is to investigate fluid mixing phenomena related to pressurized thermal shock(PTS) in a pressurized water reactor(PWR) vessel downcomer during transient cooldown with direct vessel injection(DVI) using test models. The test model designs were based on ABB Combustion Engineering(C-E) System 80＋ reactor geometry. A cold leg small break loss-of-coolant accident(LOCA) md a main steam line teak were selected as the potential PTS events for the C-E System 80＋. This work consist of two parts. The first part provides the visualization tests of the fluid mixing between DVI fluid and existing coolant in the downcomer region, and the second part is to compare the results of thermal mixing tests with DVI in the other test model. Row visualization tests with DVI have clarified the physical interaction between DVI fluid and primary coolant during transient cooldown. A significant temperature drop was observed in the downcomer during the tests of a small break LOCA Measured transient temperature profiles agree well with the predictions by the REMIX code for a small break LOCA and with the calculations by the COMMIX-1B code for a steam line break event.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.980-992
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• 2024

Steam line break accident (SLB) in the nuclear reactor is one of the representative Non-LOCA accidents in which thermal-hydraulics and neutron kinetics are strongly coupled each other. Thus, the multi-scale and multi-physics approach is applied in this study in order to examine a realistic safety margin. An entire reactor coolant system is modelled by system scale node, whereas sub-channel scale resolution is applied for the region of interest such as the reactor core. Fuel performance code is extended to consider full core pin-wise fuel behaviour. The MARU platform is developed for easy integration of the codes to be coupled. An initial stage of the steam line break accident is simulated on the MARU platform. As cold coolant is injected from the cold leg into the reactor pressure vessel, the power increases due to the moderator feedback. Three-dimensional coolant and fuel behaviour are qualitatively visualized for easy comprehension. Moreover, quantitative investigation is added by focusing on the enhancement of safety margin by means of comparing the minimum departure from nucleate boiling ratio (MDNBR). Three factors contributing to the increase of the MDNBR are proposed: Various geometric parameters, realistic power distribution by neutron kinetics code, Radial coolant mixing including sub-channel physics model.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3256-3263
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• 2021

The main steam line break accident is an essential initiating event of the pressurized water reactor. In present work, the fuzzy set theory and the signal-based fault detection method has been used to detect the occurrence and diagnosis of the location and break area for the small scale MSLB. The models are validated by the AP1000 accident simulator based on MAAP5. From the test results it can be seen that the proposed approach has a rapid and proper response on accident detection and location diagnosis. The method proposed to evaluate the break area shows good performances for small scale MSLB with the relative deviation within ±3%.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.402-409
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• 2019

When a loss of coolant accident (LOCA) occurs in a nuclear power plant, accident scenarios which can prevent core damage are defined based on break size. Current probabilistic safety assessment evaluates that core damage can be prevented under small-break LOCA (SBLOCA) and steam generator tube rupture (SGTR) with rapid cool down (RCD) strategy when all safety injection systems are unavailable. However, previous research has pointed out a limitation of RCD in terms of initiation time. Therefore, RCD success criteria estimation based on allowable coping time under a SBLOCA or SGTR when all safety injection systems are unavailable was performed based on time-line and thermal-hydraulic analyses. The time line analysis assumed a single emergency operating procedure flow, and the thermal hydraulic analysis utilized MARS-KS code with variables of break size, cooling rate, and operator allowable time. Results show while RCD is possible under SGTR, it is impossible under SBLOCA at the APR1400's current cooling rate limitation of 55 K/hr. A success criteria map for RCD under SBLOCA is suggested without cooling rate limitation.

• Nuclear Engineering and Technology
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• v.14 no.4
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• pp.186-195
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• 1982

The steam line break accident for Kori Unit 1 is analyzed by a code SYSRAN which calculates nuclear power and heat flux using the point kinetics equation and the lumped-parameter model and calculates system transient using the mass and energy balance equation with the assumption of uniform reactor coolant system pressure. The 1.4 f $t^2$ steam line break accident is analyzed at EOL (End of Life), hot shutdown condition in which case the accident would be most severe. The steam discharge rate is assumed to follow the Moody critical flow model. The results reveal the peak heat flux of 38% of nominal full power value at 60 second after the accident initiates, which is higher than the FSAR result of 26%. Trends for the transient are in good agreement with FSAR results. A sensitivity study shows that this accident is most sensitive to the moderator density coefficient and the lower plenum mixing factor. The DNBR calculation under the assumption of $F_{{\Delta}H}$=3.66, which is used in the FSAR with all the control and the shutdown assemblies inserted except one B bank assembly and of Fz=1.55 shows that minimum DNBR reaches 1.62 at 60 second, indicating that the fuel failure is not anticipated to occur. The point kinetics equation, the lumped-parameter model and the system transient model which uses the mass and energy balance equation are verified to be effective to follow the system transient phenomena of the nuclear power plants.lear power plants.