Characteristics of Transition Boiling in Sodium-Heated Steam Generator Tubes

1979 ◽  
Vol 101 (2) ◽  
pp. 270-275 ◽  
Author(s):  
D. M. France ◽  
R. D. Carlson ◽  
T. Chiang ◽  
R. Priemer
Keyword(s):  
Thermal Stress ◽  
Steam Generator ◽  
Tube Wall ◽  
Water Pressure ◽  
Thermal Fluctuations ◽  
Straight Tube ◽  
Steam Generators ◽  
Steam Generator Tubes ◽  
Heated Length ◽  
Boiling Zone

Thermal fluctuations were measured in the tube wall in the transition boiling zone of a full-scale LMFBR sodium-heated steam generator tube. The tube had an inside diameter = 10 mm, wall thickness = 2.90 mm, heated length = 13.1 m, and material = 2 1/4 Cr-1 Mo steel. Water flowed vertically upwards inside the straight tube, and sodium flowed counter-currently in a surrounding annulus. Results of thermal, spectral, and thermal stress analyses are presented for a test within the normal operating range of LMFBR steam generators. Results of other tests are presented that show the effects and sensitivity of sodium temperature and water pressure on the severity of the thermal fluctuations.

Flow-Induced Wear in Steam Generator Tubes—Prediction Versus Operational Experience

1998 ◽  
Vol 120 (2) ◽  
pp. 138-143 ◽  
Author(s):  
M. K. Au-Yang
Keyword(s):  
Steam Generator ◽  
Tube Wall ◽  
Tube Bundle ◽  
Cross Flow ◽  
Operational Experience ◽  
Steam Generators ◽  
Support Plate ◽  
Power Operations ◽  
Full Power ◽  
Steam Generator Tubes

Many nuclear steam generators have accumulated more than 10 effective-full-power-years of operation. Eddy-current inspections revealed that a number of these steam generator tubes, notably those located in high local cross-flow regions, have indications of wear at some support plate elevations after 5 to 10 yr of effective-full-power operations. In the last 5 yr, a number of technical papers on nonlinear tube bundle dynamics has been published to address the effect of tube and support plate interactions. At the same time, test data relating wear and tube wall thickness losses for different material combinations and different support plate geometries became available. Based on the available data in the literature, as well as data obtained in the author’s affiliation, this paper assesses the cumulative tube wall wear after 5, 10, and 15 effective-full-power years of operation of a typical commercial nuclear steam generator, using different wear models. It is hoped that this study will shed some light on the probable mechanism that caused the observed wear in today’s operating nuclear steam generators.

Parametric Study of External Pressure on Steam Generator Tube Bends

2012 ◽  
Author(s):  
Mitch Hokazono ◽  
Clayton T. Smith
Keyword(s):  
Steam Generator ◽  
Parametric Study ◽  
External Pressure ◽  
Light Water Reactor ◽  
Material Strength ◽  
Steam Generators ◽  
Elliptical Cross ◽  
Light Water ◽  
Water Reactor ◽  
Steam Generator Tubes

Integral light-water reactor designs propose the use of steam generators located within the reactor vessel. Steam generator tubes in these designs must withstand external pressure loadings to prevent buckling, which is affected by material strength, fabrication techniques, chemical environment and tube geometry. Experience with fired tube boilers has shown that buckling in boiler tubes is greatly alleviated by controlling ovality in bends when the tubes are fabricated. Light water reactor steam generator pressures will not cause a buckling problem in steam generators with reasonable fabrication limits on tube ovality and wall thinning. Utilizing existing Code rules, there is a significant design margin, even for the maximum differential pressure case. With reasonable bend design and fabrication limits the helical steam generator thermodynamic advantages can be realized without a buckling concern. This paper describes a theoretical methodology for determining allowable external pressure for steam generator tubes subject to tube ovality based on ASME Section III Code Case N-759-2 rules. A parametric study of the results of this methodology applied to an elliptical cross section with varying wall thicknesses, tube diameters, and ovality values is also presented.

Experimental Investigation of Subcooled Choking Flow in a Steam Generator Tube Crack

2016 ◽  
Author(s):  
Hung Nguyen ◽  
Mark Brown ◽  
Shripad T. Revankar ◽  
Jovica Riznic
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Experimental Program ◽  
Steam Generators ◽  
Steam Generator Tube ◽  
Loss Of Coolant ◽  
Steam Generator Tubes

Steam generator tubes have a history of small cracks and even ruptures, which lead to a loss of coolant from the primary side to the secondary side. These tubes have an important role in reactor safety since they serve as one of the barriers between radioactive and non-radioactive materials of a nuclear power plant. A rupture then signifies the loss of the integrity of the tube itself. Therefore, choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant, but also to everyday operation. There is limited data on actual steam generators tube wall cracks. Here experiments were conducted on choked flow of subcooled water through two samples of axial cracks of steam generator tubes taken from US PWR steam generators. The purpose of the experimental program was to develop database on critical flow through actual steam generator tube cracks with subcooled liquid flow at the entrance. The knowledge of this maximum flow rate through a crack in the steam generator tubes of a pressurized water nuclear reactor will allow designers to calculate leak rates and design inventory levels accordingly while limiting losses during loss of coolant accidents. The test facility design is modular so that various steam generator tube cracks can be studied. Two sets of PWR steam generators tubes were studied whose wall thickness is 1.285 mm. Tests were carried out at stagnation pressure up to 6.89 MPa and range of subcoolings 16.2–59°C. Based on these new choking flow data, the applicability of analytical models to highlight the importance of non-equilibrium effects was examined.

Fitness-for-Service Assessment of Steam Generator Tubing Subject to Dealloying Degradation

2017 ◽  
Author(s):  
Rosita Mousavi ◽  
Xinjian Duan ◽  
Michael Kozluk ◽  
Min Wang ◽  
Yihai Shi
Keyword(s):  
Steam Generator ◽  
Degradation Mechanism ◽  
Life Cycle Management ◽  
Steam Generators ◽  
Nickel Copper ◽  
Steam Generator Tubing ◽  
Management Plans ◽  
Monel 400 ◽  
Technical Basis ◽  
Steam Generator Tubes

A new degradation mechanism has been observed in Monel 400 Steam Generator tubing material, a nickel-copper alloy (63Ni-28Cu-2½Fe) with the ASME material designation SB-163/N04400. The location is above the top preheater support plate of the two re-circulating steam generator in one of the units of the Pickering Nuclear Generating Station. This paper provides a brief description of the regulatory environment, OPG’s steam generator life cycle management plans, the Canadian Industry’s fitness-for-service guidelines for steam generator tubes, and the afflicted steam generators. The paper then goes on to discuss the following activities that were conducted to support the technical basis to justify that the steam generators fit to be returned to service: • Inspection scope expansion, methods, and results. • Examination of removed tubes. • Condition monitoring assessment. • Operational assessment. • Burst-pressure tests of removed tubes and of fabricated test specimens. • Degradation specific flaw model and acceptance standards. • Flaw growth rate predictions. • Plugging limit adopted.

Plastic Limit Load Solutions for Surface Crack in the Steam Generator Tubes

2005 ◽  
Vol 297-300 ◽  
pp. 1704-1712
Author(s):  
Ouk Sub Lee ◽  
Hyun Su Kim ◽  
Jong Sung Kim ◽  
Tae Eun Jin ◽  
Hong Deok Kim ◽  
...  
Keyword(s):  
Steam Generator ◽  
Safety Margin ◽  
Local Limit ◽  
Limit Load ◽  
Material Behavior ◽  
Surface Cracks ◽  
Steam Generators ◽  
Integrity Assessment ◽  
Global And Local ◽  
Steam Generator Tubes

Operating experience of steam generators has shown that cracks of various morphologies frequently occur in the steam generator tubes. These cracked tubes can stay in service if it is proved that the tubes have sufficient safety margin to preclude the risk of burst and leak. Therefore, integrity assessment using exact limit load solutions is very important for safe operation of the steam generators. This paper provides global and local limit load solutions for surface cracks in the steam generator tubes. Such solutions are developed based on three-dimensional (3-D) finite element analyses assuming elastic-perfectly plastic material behavior. For the crack location, both axial and circumferential surface cracks, and for each case, both external and internal cracks are considered. The resulting global and local limit load solutions are given in polynomial forms, and thus can be simply used in practical integrity assessment of the steam generator tubes, because the comparison between experimental data and FE solutions shows good agreement.

Effect of Support Friction in Thermal Stress of VVER-1000 RCS for NOP Condition

2014 ◽  
Author(s):  
Truong Quang Nguyen ◽  
Ihn Namgung
Keyword(s):  
Thermal Stress ◽  
Yield Stress ◽  
Steam Generator ◽  
Reactor Vessel ◽  
Inside Surface ◽  
Steam Generators ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Hexahedral Elements ◽  
Effect Of Support

The main purpose of this research is to investigate the effect of friction in the thermal stress of Reactor Coolant System (RCS) of VVER-1000. RCS is a large system connecting reactor vessel, steam generators and RC Pumps. During the heat-up of reactor, the RCS expand and during cool-down of reactor, it contracts. Because of the heavy weight of reactor and steam generator, the friction at the support of RCS affects the thermal stress of RCS. In this paper how much support friction contributes to the development of thermal stress is assessed in order to investigate the thermal stress and effect of support friction. A quarter-symmetry model of VVER-1000 RCS is developed in ANSYS and meshed with hexahedral elements to ensure better solution accuracies. The model includes reactor vessel, steam generator and reactor coolant pump. Internals of reactor vessel, steam generators and RCPs are represented by point mass to simplify the model. Temperature of inside surface of hot-leg side of reactor vessel to inlet side of steam generator is assumed same uniform hot-leg temperature, and the temperature of inside surface of outlet side of steam generator to reactor vessel is uniform cold-leg temperature. All outside surface are assumed insulated. The analysis includes neither transient thermal loading nor dynamic loadings. The analysis results show that friction at support brings little effect on the peak thermal stress. The peak thermal stress occurs at hot-leg nozzle of reactor pressure vessel and it approached near yield stress. If load combination is included the localized total stress at hot-leg nozzle could go over the yield stress. This peak stress could affect fatigue life in a long run. A recommendation is made that a detailed fatigue analysis of VVER-1000 RCS is necessary.

Critical Heat Flux Experiments and Correlation in a Long, Sodium-Heated Tube

1981 ◽  
Vol 103 (1) ◽  
pp. 74-80 ◽  
Author(s):  
D. M. France ◽  
R. D. Carlson ◽  
T. Chiang ◽  
W. J. Minkowycz
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Steam Generator ◽  
Test Section ◽  
Water Pressure ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Test Facility ◽  
Correlation Equations ◽  
Steam Generators

Critical heat flux (CHF) experiments were performed in the Steam Generator Test Facility (SGTF) at Argonne National Laboratory for application to liquid metal fast breeder reactor steam generators. The test section consisted of a single, straight, vertical, full-scale LMFBR steam generator tube with force-circulated water boiling upwards inside the tube heated by sodium flowing countercurrent in a surrounding annulus. The test section tube parameters were as follows: 10.1 mm i.d., 15.9 mm o.d., material = 2 1/4 Cr–1 Mo steel, and 13.1 m heated length. Experiments were performed in the water pressure range of 7.0 to 15.3 MPa and the water mass flux range of 720 to 3200 kg/m2˙s. The data exhibited two trends: heat flux independent and heat flux dependent. Empirical correlation equations were developed from over 400 CHF tests performed in the SGTF. The data and correlation equations were compared to the results of other CHF investigations.

Choking Flow of Subcooled Liquid in Steam Generator Tube Wall Cracks

2014 ◽  
Author(s):  
Ram Anand Vadlamani ◽  
Shripad T. Revankar ◽  
Jovica R. Riznic
Keyword(s):  
Steam Generator ◽  
Tube Wall ◽  
Channel Length ◽  
Analytical Models ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Choked Flow ◽  
Loss Of Coolant ◽  
Small Cracks ◽  
Steam Generator Tubes

Steam generator tubes have a history of small cracks and even ruptures, which lead to a loss of coolant from the primary side to the secondary side. Currently, steam generators operate under a leak-before-break approach. A rupture then signifies the loss of the integrity of the tube itself. Therefore, choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant, but also to everyday operation. Choked flow of subcooled water through small cracks such as in steam generator tube wall cracks is studied both with experiments and analytical models. The knowledge of this maximum flow rate through a crack in the steam generator tubes of a pressurized water nuclear reactor will allow designers to calculate leak rates and design inventory levels accordingly while limiting losses during loss of coolant accidents. Slits of very small channel length to hydraulics diameter ratio (L/D) were manufactured and tested upto 6.89 MPa pressure and range of subcoolings 10–40 °C. Small flow channel length was used (1.3mm) equivalent to steam generator tube thickness with differences in surface roughness. The effect of L/D on the choking flow rates was examined and was contrasted with other data in literature. Analytical models were applied highlighting the importance of non-equilibrium effects and the effects of L/D ranging from 1.3 to 400 on the chocked flow were investigated.

Correlations for the Determination of the Inception Conditions of Density-Wave Oscillations for Forced and Natural Circulation Steam Generator Tubes

1980 ◽  
Vol 102 (1) ◽  
pp. 14-19 ◽  
Author(s):  
H. C. U¨nal
Keyword(s):  
Steam Generator ◽  
Mass Velocity ◽  
Natural Circulation ◽  
Density Wave ◽  
Steam Quality ◽  
Forced Circulation ◽  
Percent Accuracy ◽  
Steam Generator Tubes ◽  
Density Wave Oscillations ◽  
Heated Length

Accurate and simple correlations are presented to determine the inception conditions of density-wave oscillations in steam generator tubes. The correlations predict the power at the start of the density-wave oscillations within about 6.5 percent accuracy for long (i.e., L ≥ 10 m) forced circulation steam generator tubes and within about 20 percent accuracy for natural circulation and short forced circulation steam generator tubes. The ranges of the operating conditions and geometries for the data used to establish the correlations are as follows: Forced circulation tubes: Geometry: circular-straight tubes and serpentines, a circular coil and a rectangular straight tube; type of heating: electrical or sodium heating; the ratio of the heated length to diameter: 153–9502; pressure: 4.1–17.3 MN/m2; outlet steam quality: 0.27–1.85; inlet subcooling: 2.8–245.9 K; mass velocity: 118–2088 kg/m2s. Natural circulation tubes: Geometry and heating conditions: electrically heated circular tubes and annuli; ratio of the heated length to diameter: 34–489; pressure: 0.2–7.1 MN/m2; outlet steam quality: 0.04–0.62; inlet subcooling: 0–244 K; mass velocity: 529–1230 kg/m2s. The number of data considered is 106 for forced circulation tubes and 110 for natural circulation tubes.

A High Reliability Straight Tube LMFBR Steam Generator Design

1980 ◽  
Vol 102 (3) ◽  
pp. 568-572 ◽  
Author(s):  
A. H. Spring ◽  
D. D. DeFur
Keyword(s):  
Thermal Expansion ◽  
Steam Generator ◽  
High Reliability ◽  
Straight Tube ◽  
Steam Generators ◽  
Geometric Size ◽  
Differential Thermal Expansion ◽  
Generator Design ◽  
Double Wall ◽  
Reliability And Availability

Straight-tube counterflow steam generators for LMFBR applications have been proposed, tested, fabricated and operated with varying degrees of success. A new embodiment of the straight-tube concept is described which incorporates a number of unique features which contribute to high reliability and availability. These features include a replaceable bellows for accommodation of differential thermal expansion between shell and tubes and a redundant, crevice-free tube-to-tubesheet joint design. The design can also easily incorporate single-wall or double-wall tubes. Single and double-wall tube versions are described whose thermal and geometric size are based on anticipated manufacturing limitations. The results of scoping tests of the tube-to-tubesheet welds are described which provide positive indications of the soundness of the weld design.

Sign in / Sign up

Export Citation Format

Share Document