Stress analysis of pressurized water reactor steam generator tube denting phenomena. Interim report

Lots of efforts have been made to Research & Development of Small pressurized water reactors (SPWRs). Steam generator tube break occurs due to wear and corrosion frequently in the reactor. Among the breaks, Small Steam Generator Tube Break (SSGTB) is difficult to detect. Therefore, it is necessary to investigate the features of SSGTB. A small pressurized water reactor model has been established in this paper by Relap5. The model includes reactor core, pressurizer, steam generator, main coolant pump and auxiliary safety system. The core flow, pressure of pressurizer, core outlet temperature and secondary outlet steam temperature obtained based on steady-state calculation is compared with design data to verify the model correct. SSGTB is simulated by introducing a small break in the steam generator tube. The important parameters of reactor are recorded and analyzed. The procedure of SSGTB is analyzed and the system response features are summarized.

Characteristics of steam generator tube deposits in an operating pressurized water reactor

Journal of Nuclear Materials ◽

10.1016/j.jnucmat.2018.05.001 ◽

2018 ◽

Vol 507 ◽

pp. 371-380 ◽

Cited By ~ 8

Author(s):

Soon-Hyeok Jeon ◽

Seokmin Hong ◽

Hyuk-Chul Kwon ◽

Do Haeng Hur

Keyword(s):

Steam Generator ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Steam Generator Tube ◽

Water Reactor

Control of Alkaline Stress Corrosion Cracking in Pressurized-Water Reactor Steam Generator Tubing

CORROSION ◽

10.5006/1.3280503 ◽

1999 ◽

Vol 55 (6) ◽

pp. 616-625 ◽

Cited By ~ 5

Author(s):

I. S. Hwang ◽

I.-G. Park

Keyword(s):

Stress Corrosion ◽

Steam Generator ◽

Stress Corrosion Cracking ◽

Corrosion Cracking ◽

Alkaline Stress ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Steam Generator Tubing

Non-uniform flow distribution in the steam generator U-tubes of a pressurized water reactor plant during single- and two-phase natural circulations

Nuclear Engineering and Design ◽

10.1016/j.nucengdes.2004.02.002 ◽

2004 ◽

Vol 231 (3) ◽

pp. 303-314 ◽

Cited By ~ 30

Author(s):

Jae-Jun Jeong ◽

Moonkyu Hwang ◽

Young Jin Lee ◽

Bub Dong Chung

Keyword(s):

Steam Generator ◽

Flow Distribution ◽

Uniform Flow ◽

Pressurized Water Reactor ◽

Reactor Plant ◽

Two Phase ◽

Pressurized Water ◽

Water Reactor ◽

Uniform Flow Distribution

Corrosion Fatigue Cracking of a Steam Generator Vessel From a Pressurized Water Reactor

Handbook of Case Histories in Failure Analysis ◽

10.31399/asm.fach.v01.c9001051 ◽

1992 ◽

pp. 136-139

Keyword(s):

Steam Generator ◽

Corrosion Fatigue ◽

Fatigue Cracking ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor

Primary Water Stress Corrosion Cracking Analysis in Alloy 600 Steam Generator Nozzle of a Pressurized Water Reactor

CORROSION ◽

10.5006/0935 ◽

2013 ◽

Vol 69 (11) ◽

pp. 1051-1059 ◽

Cited By ~ 2

Author(s):

Seong Sik Hwang ◽

Yun Soo Lim ◽

Sung Woo Kim

Keyword(s):

Water Stress ◽

Stress Corrosion ◽

Steam Generator ◽

Stress Corrosion Cracking ◽

Corrosion Cracking ◽

Alloy 600 ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Primary Water

Testing Steam Generator Materials For Pressurized Water Reactor Plants

CORROSION ◽

10.5006/0010-9312-18.1.27 ◽

1962 ◽

Vol 18 (1) ◽

pp. 27t-32t ◽

Cited By ~ 5

Author(s):

J. O. H. N. W. McGREW

Keyword(s):

Steam Generator ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor

CFD Predictions of Severe Accident Steam Generator Flows in a 1/7th Scale Pressurized Water Reactor

10th International Conference on Nuclear Engineering, Volume 3 ◽

10.1115/icone10-22466 ◽

2002 ◽

Cited By ~ 1

Author(s):

Christopher Boyd ◽

Kelly Hardesty

Keyword(s):

Steam Generator ◽

Natural Circulation ◽

Severe Accident ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Steam Generator Tube ◽

Flow Phenomena ◽

Experimental Values ◽

High Degree ◽

Side Flow

Computational Fluid Dynamics (CFD) is applied to steam generator inlet plenum mixing as part of a larger plan covering steam generator tube integrity. The technique is verified by comparing predicted results with severe accident natural circulation data [1] from a 1/7th scale Westinghouse facility. This exercise demonstrates that the technique can predict the natural circulation and mixing phenomena relevant to steam generator tube integrity issues. The model includes primary side flow paths for a single hot leg and steam generator. Qualitatively, the experimentally observed flow phenomena are predicted. The paths of the natural circulation flows and the relative flow proportions are correctly predicted. Quantitatively, comparisons are made with temperatures, mass flows, and other parameters. All predictions are generally within 10% of the experimental values. Overall, there is a high degree of confidence in the CFD technique for prediction of the relevant flow phenomena associated with this type of severe accident sequence.

Development of Steam Generator Model Using RELAP5 to Simulate Two-Phase Natural Circulation for Pressurized Water Reactor

Volume 6: Nuclear Education, Public Acceptance and Related Issues; Instrumentation and Controls (I&C); Fusion Engineering; Beyond Design Basis Events ◽

10.1115/icone22-30703 ◽

2014 ◽

Author(s):

Shinya Miyata ◽

Satoru Kamohara ◽

Wataru Sakuma ◽

Hiroaki Nishi

Keyword(s):

Steam Generator ◽

Natural Circulation ◽

Injection System ◽

Pressurized Water Reactor ◽

Two Phase ◽

Pressurized Water ◽

Water Reactor ◽

Loss Of Coolant ◽

Mass Inventory ◽

Core Cooling

In typical pressurized water reactor (PWR), to cope with beyond design basis events such as station black out (SBO) or small break loss of coolant accident with safety injection system failure, injection from accumulator sustains core cooling by compensating for loss of coolant. Core cooling is sustained by single- or two-phase natural circulation or reflux condensation depending on primary coolant mass inventory. Behavior of the natural circulation in PWR has been investigated in the facilities such as Large Scale Test Facility (LSTF) which is a full-height and full-pressure and thermal-hydraulic simulator of typical four-loop PWR. Two steady-state natural circulation tests were conducted in LSTF at both high and low pressure. These two tests were conducted changing the primary mass inventory as a test parameter, while keeping the other parameters such as core power, steam generator (SG) pressure, and steam generator water level as they are. Mitsubishi Heavy Industries (MHI) plans new natural circulation tests to cover wider range of core power and pressure as test-matrix (including the previous LSTF tests) to validate applicability of the model in wider range of core power and pressure conditions including the SBO conditions. In this paper, the previous LSTF natural circulation tests are reviewed and the new test plan will be described. Additionally, MHI also started a feasibility study to improve the steam generator tube and inlet/outlet plenum model using the M-RELAP5 code [4]. Newly developed model gives reasonable agreement with the previous LSTF tests and applies to the new test conditions. The feasibility findings will also be described in this paper.