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Title: Steam generator tube rupture simulation using extended finite element method

Authors:
; ;
Publication Date:
Sponsoring Org.:
USDOE
OSTI Identifier:
1359689
Resource Type:
Journal Article: Publisher's Accepted Manuscript
Journal Name:
Nuclear Engineering and Design
Additional Journal Information:
Journal Volume: 305; Journal Issue: C; Related Information: CHORUS Timestamp: 2017-10-03 21:28:59; Journal ID: ISSN 0029-5493
Publisher:
Elsevier
Country of Publication:
Netherlands
Language:
English

Citation Formats

Mohanty, Subhasish, Majumdar, Saurin, and Natesan, Ken. Steam generator tube rupture simulation using extended finite element method. Netherlands: N. p., 2016. Web. doi:10.1016/j.nucengdes.2016.06.031.
Mohanty, Subhasish, Majumdar, Saurin, & Natesan, Ken. Steam generator tube rupture simulation using extended finite element method. Netherlands. doi:10.1016/j.nucengdes.2016.06.031.
Mohanty, Subhasish, Majumdar, Saurin, and Natesan, Ken. Mon . "Steam generator tube rupture simulation using extended finite element method". Netherlands. doi:10.1016/j.nucengdes.2016.06.031.
@article{osti_1359689,
title = {Steam generator tube rupture simulation using extended finite element method},
author = {Mohanty, Subhasish and Majumdar, Saurin and Natesan, Ken},
abstractNote = {},
doi = {10.1016/j.nucengdes.2016.06.031},
journal = {Nuclear Engineering and Design},
number = C,
volume = 305,
place = {Netherlands},
year = {Mon Aug 01 00:00:00 EDT 2016},
month = {Mon Aug 01 00:00:00 EDT 2016}
}

Journal Article:
Free Publicly Available Full Text
Publisher's Version of Record at 10.1016/j.nucengdes.2016.06.031

Citation Metrics:
Cited by: 2works
Citation information provided by
Web of Science

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  • A major concern in the nuclear power industry is failure of the steam generator tubes. Failure of the tubes necessitates the plugging of the failed tubes with the result that nuclear plants are forced to operate at lower, or derated, power levels after expensive repairs. Turbulence-induced vibration is a primary cause of premature and accelerated fretting and wear of the steam generator tubes. An alternative unsteady analysis method for incompressible fluid flow problems is demonstrated. The approach employs large eddy simulation (LES) in conjunction with the finite element method (FEM). A segregated solution technique, solving for each field variable atmore » all nodes, diminishes storage requirements by eliminating the need to solve the globally assembled finite element matrix. A direct benefit is that finer nodalizations can be employed. Equal-order quadrilateral elements are used to facilitate the segregated solution algorithm. The solution scheme is accurate to higher order to mitigate the effects of numerical diffusion in the advection terms. The Smagorinsky-type closure model for the sub-grid scale turbulence is used. The model is easily implemented into this algorithm. This combination of FEM and LES is unique. The time-dependent terms are explicitly treated. The time history of a steam generator tube bundle experiment is studied. The results show the applicability of FEM/LES and determine the prospects for further development of this methodology.« less
  • The nuclear industry has initiated a program of Steam Generator Degradation Specific Management (SGDSM) to cope with the various types of corrosion that have been observed in pressurized water reactor (PWR) steam generators. In parallel, the U.S. Nuclear Regulatory Commission is promulgating revised rules on steam generator tube integrity. To support these efforts, the Electric Power Research Institute has sponsored calculations with the MAAP 4 code. The principal objective of these calculations is to estimate the peak temperatures experienced by the steam generator tubes during high-pressure severe accidents. These results are used to evaluate the potential for degraded tubes tomore » leak or rupture. Attention was focused on station blackout (SBO) accidents with loss of turbine-driven auxiliary feedwater because these generally result in the greatest threat to the tubes.« less
  • A simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermal-hydraulic code TRACE5. Test 5 performed in the Large Scale Test Facility (LSTF) reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR). The result of these simultaneous breaks is a depressurization in the secondary and primary system in loop B because both systems are connected through the SGTR. Good approximation was obtained between TRACE5 results and experimental data. TRACE5 reproduces qualitatively the phenomena that occur in this transient: primary pressure falls after the break, stagnation ofmore » the pressure after the opening of the relief valve of the intact steam generator, the pressure falls after the two openings of the PORV and the recovery of the liquid level in the pressurizer after each closure of the PORV. Furthermore, a sensitivity analysis has been performed to know the effect of varying the High Pressure Injection (HPI) flow rate in both loops on the system pressures evolution. (authors)« less
  • The iodine concentration in the steam generator secondary vapor must be determined in order to estimate the environmental consequences (i.e., iodine source term to the environment) due to a steam generator tube rupture (SGTR). Experimental evidence indicates that this concentration is sensitive to the liquid-phase pH. A thermodynamic-based calculational approach was used to model the pH during a design-bases SGTR. The EQUILIBRIUM code within the Facility for Analysis of Chemical Thermodynamics was assessed for calculation of pH by comparison with measured pH's in operating pressurized water reactors (PWRs). The pH was calculated for ten generic PWR designs (one Babcock andmore » Wilcox, three Combustion Engineering, and six Westinghouse). The calculated pH was shown to be relatively insensitive to PWR design. The pH for all designs equilibrated to a value of [approximately] 6.5.« less