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Research Papers: Fluid-Structure Interaction

Effects of a Venturi-Type Flow Restrictor on the Thermal-Hydraulic Response of the Secondary Side of a Pressurized Water Reactor Steam Generator to a Main Steam Line Break

[+] Author and Article Information
Jong Chull Jo

Fellow ASME
Korea Institute of Nuclear Safety,
Yusung-gu,
Daejeon 34142, Korea;
School of Mechanical Engineering,
Pusan National University,
Geumjeong-gu,
Busan 46241, Korea
e-mail: jcjo@kins.re.kr

Frederick J. Moody

Fellow ASME
General Electric (retired),
2125 North Olive Avenue D-33,
Turlock, CA 95382
e-mail: fmoody@goldrush.com

Contributed by the Pressure Vessel and Piping Division of ASME for publication in the JOURNAL OF PRESSURE VESSEL TECHNOLOGY. Manuscript received August 28, 2015; final manuscript received December 2, 2015; published online April 28, 2016. Assoc. Editor: Tomomichi Nakamura.

J. Pressure Vessel Technol 138(4), 041304 (Apr 28, 2016) (12 pages) Paper No: PVT-15-1204; doi: 10.1115/1.4032282 History: Received August 28, 2015; Revised December 02, 2015

A numerical analysis has been performed to simulate the transient thermal-hydraulic response to a main steam line break (MSLB) for the secondary side of a steam generator (SG) model equipped with a venturi-type SG outlet flow restrictor at a pressurized water reactor (PWR) plant. To investigate the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB, numerical calculation results for the SG model equipped with the flow restrictor are compared to those obtained for an SG model without the restrictor. Both analysis models contain internal structures. The present computational fluid dynamics (CFD) model has been examined by comparing to a simple analytical model. It is confirmed from the comparison that the CFD model simulates the transient response of the SG secondary to the MSLB physically plausibly and minutely. Based on the CFD analysis results for both cases with or without the restrictor, it is seen that the intensities of the steam velocity and dynamic pressure are considerably attenuated in the SG model equipped with the restrictor comparing to the case in the SG model without the restrictor.

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References

Shier, W. G. , and Levine, M. M. , 1980, “ PWR Steam Line Break Analysis Assuming Concurrent Steam Generator Tube Rupture,” ANS/ASME Topical Meeting on Reactor Thermal-Hydraulics, Saratoga, NY, Oct. 9, Paper No. CONF-801002-9
Gallardo, S. , Querol, A. , and Verdú, G. , 2012, “ Simulation of a Main Steam Line Break With Steam Generator Tube Rupture Using Trace,” PHYSOR 2012, American Nuclear Society, Knoxville, TN, Apr. 15–20, pp. 2131–2144.
Gallardo, S. , Querol, A. , and Verdú, G. , 2013, “ Improvements in the Simulation of a Main Steam Line Break With Steam Generator Tube Rupture,” Joint International Conference on Supercomputing in Nuclear Applications and Monte Carlo, Paris, France, Oct. 27–31, Paper No. 05104.
Kalra, S. , and Adams, G. , 1980, “ Thermal Hydraulics of Steam Line Break Transients in Thermal Reactors—Simulation Experiments,” ANS International Conference, American Nuclear Society, Washington, DC, Nov. 17–21, Vol. 35, Paper No. CONF-801107.
Wolf, L. , 1982, “ Experimental Results of Coupled Fluid–Structure Interactions During Blowdown of the HDR-Vessel and Comparisons With Pre- and Post-Test Predictions,” Nucl. Eng. Des., 70(3), pp. 269–308. [CrossRef]
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Joo, H. G. , Jeong, J.-J. , Cho, B.-O. , Lee, W. J. , and Zee, S. Q. , 2003, “ Analysis of the OECD Main Steam Line Break Benchmark Problem Using the Refined Core Thermal-Hydraulic Nodalization Feature of the MARS/MASTER Code,” Nucl. Technol., 142(2), pp. 166–179.
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Jo, J. C. , and Moody, F. J. , 2015, “ Transient Thermal-Hydraulic Responses of the Nuclear Steam Generator Secondary Side to a Main Steam Line Break,” ASME J. Pressure Vessel Technol., 137(4), p. 041301. [CrossRef]
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Jo, J. C. , Lee, S. K. , Kim, W. S. , Shin, W. K. , Kim, H. Y. , and Ha, J. T. , 1992, “ A Study on the Thermal-Hydraulic and Flow-Induced Tube Vibration Analysis of Nuclear Steam Generators,” Korea Institute of Nuclear Safety, Technical Report No. KINS/AR-198.
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Figures

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Fig. 1

Simplified analysis model of the SG with an MSL

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Fig. 2

Perforated plate modeled as the SG internal structures

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Fig. 3

Two different SG analysis models: (a) SG model 1 and (b) SG model 2

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Fig. 4

Discretized solution domains: (a) SG model 1 and (b) SG model 2

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Fig. 6

Steady velocity distributions of steam inside the two SG models during the normal reactor operation: (a) SG model 1 and (b) SG model 2

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Fig. 7

Transient velocity distributions of steam inside the SG following the MSLB accident for the SG model 1

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Fig. 8

Transient velocity distributions of steam inside the MSL following the MSLB accident at the elapsed times of (a) 0.05 s and (b) 5.0 s for the SG model 1

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Fig. 9

Transient steam velocity responses to the MSLB at the monitoring points for the SG model 1: (a) at the monitoring points 1–3 and (b) at the monitoring points 4 and 5

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Fig. 10

Transient dynamic pressure responses to the MSLB at the monitoring points for the SG model 1

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Fig. 11

Transient static pressure responses to the MSLB at the monitoring points for the SG model 1

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Fig. 12

Transient velocity distributions of steam inside the SG following the MSLB accident for the SG model 2

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Fig. 13

Transient velocity distributions of steam inside the MSL following the MSLB accident at the elapsed times of (a) 0.05 s and (b) 5.0 s for the SG model 2

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Fig. 14

Transient steam velocity responses to the MSLB at the monitoring points for the SG model 2: (a) at the monitoring points 1–3 and (b) at the monitoring points 4 and 5

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Fig. 15

Transient dynamic pressure responses to the MSLB at the monitoring points for the SG model 2

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Fig. 16

Transient static pressure responses to the MSLB at the monitoring points for the SG model 2

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Fig. 17

Simplified SG model 2 with the MSL

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Fig. 18

Enlargement of the transient static decompression disturbance at the monitoring point P3 in the SG model 2 during a time period between 0.05 s and 0.25 s

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