• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.322-336
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• 2021

This paper presents a numerical prediction of the transient hydraulic loads acting on the tubes and external supports of a pressurized water reactor (PWR) steam generator (SG) during blowdown following a sudden feedwater line break (FWLB). A simplified SG model was used to easily demonstrate the prediction. The blowdown discharge flow was treated as a flashing flow to realistically simulate the transient flow fields inside the SG and the connected broken feedwater pipe. The effects of the SG initial pressure or the broken feedwater pipe length on the intensities or magnitudes of transient hydraulic loads were investigated. Then predictions of the decompression pressure wave-induced impulsive pressure differential loads on SG tubes and the transient blowdown loads on SG external supports were demonstrated and the general aspects of transient responses of such transient hydraulic loads to the FWLB were discussed.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.1029-1040
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• 2021

For the structural integrity evaluation of pressurized water reactor (PWR) steam generator (SG) tubes subjected to transient hydraulic loading, determination of the tube-to-tube gap velocity and static pressure distributions along the tubes is prerequisite. This paper addresses both computational fluid dynamics (CFD) and analytical approaches for predicting the tube-to-tube gap velocity and static pressure distributions during blowdown following a feedwater line break (FWLB) accident at a PWR SG. First of all, a comparative study on CFD calculations of the transient velocity and pressure distributions in the SG secondary sides for two different models having 30 or no tubes is performed. The result shows that the velocities of sub-cooled water flowing between any adjacent two tubes of a tubed SG model during blowdown can be roughly estimated by applying the specified SG secondary side porosity to those of the no-tubed SG model. Secondly, simplified analytical approximate solutions for the steady two-dimensional SG secondary flow velocity and pressure distributions under a given discharge flowrate are derived using a line sink model. The simplified analytical solutions are validated by comparing them to the CFD calculations.

• Journal of Energy Engineering
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• v.16 no.3
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• pp.103-112
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• 2007

Flood issue for nuclear power plants designed and built in 1970 is extremely severe for main steam header compartment and main feedwater line region of intermediate building and lower floor. A calculation for flood level at the main feedwater line isolation compartment is now performing by hand calculation. But, this methodology is quite conservative assumption. The goal of this study was to develop method to analyze flowrate using the RETRAN-3D computer code, and the developed method was applied to flood level analysis following main feedwater line break. As a result of analysis, flood level was low remarkably.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.234.1-234
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.584-585
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• 2005

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.347-360
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• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.681-682
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• 2005

• Journal of Energy Engineering
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• v.21 no.1
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• pp.75-80
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• 2012

In this paper a realistic bounding method for flooding analysis following rupture of large size of thanks and piping is defined. Mass and energy release during main feedwater line break accident is analyzed with RETRAN code. It is modeled that the main feed water control valve is closed in 5.0 seconds after reactor trip. In result of the analysis, largest mass and energy is discharged at 70% reactor power. The flood sources for main feedwater room are calculated when piping failure occurs in the high energy line and medium energy line. Based on the result of flood level (1.43m), it is investigated that all of the safety-related environmental qualification equipments are well located above the flood level.

• Journal of Energy Engineering
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• v.12 no.2
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• pp.93-100
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• 2003

An estimation methodology of the mass and energy (M/E) release due to the main steam line break (MSLB) has been developed with the RETRAN-3D code. In the case of equipment qualification (EQ), the over-estimated temperature would exceed the design limits of some cables or valves. In order to have a more flexible EQ profiles from the MSLB M/E release, the methodology with the best-estimated code was used. The major conditions affecting the MSLB M/E were found to be the initial SG level, heat transfer between primary and secondary sides, power level, operable protection system, main or auxiliary feedwater availability, and break conditions. The RETRAN-3D models were developed for the Kori unit 1 (KRN-1) which is typical two loop Westinghouse (WH) designed plant. Particularly, a detailed model of the steam generators was developed to estimate a more realistic two-phase heat transfer effect of the steam flow. After the modeling, the methodology has been developed through the sensitivity analyses. The M/E release data generated from the analyses have been used as the input to the inside containment pressure and temperature (P/T) analysis. According to the results at the point of view containment P/T, the Kori unit 1 can have more margin of 5∼15 ㎪ in pressure and 8∼15$^{\circ}C$ in temperature.

• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.597-610
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• 2012

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of PAFS, an experimental program is in progress at KAERI, which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL ($\underline{P}$AF$\underline{S}$ $\underline{C}$ondensing Heat Removal $\underline{A}$ssessment $\underline{L}$oop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. A single, nearly-horizontal U-tube, whose dimensions are the same as the prototypic U-tube of the APR+ PAFS, is simulated in the PASCAL test. The PASCAL experimental result showed that the present design of PAFS satisfied the heat removal requirement for cooling down the reactor core during the anticipated accident transients. The integral effect test is in progress to confirm the operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. As the first integral effect test, an FLB (feedwater line break) accident was simulated for the APR+. From the integral effect test result, it could be concluded that the APR+ has the capability of coping with the hypothetical FLB accident by adopting PAFS and proper set-points of its operation.