Effect of Bending Load on Burst Pressure of Nuclear Power Plant Steam Generator Tubes With Uniform Wall Thinning

2019 ◽  
Author(s):  
Michael C. Liu ◽  
Robert J. Gialdini ◽  
Russell C. Cipolla ◽  
Chang-Hoon Ha ◽  
Min-Ki Cho ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Burst Pressure ◽  
Burst Strength ◽  
Wall Loss ◽  
Steam Generator Tubes ◽  
Tube Repair

Abstract Tube integrity is an important aspect for safe and reliable operation of nuclear power plant steam generators. As a U.S. industry and licensing requirement, all in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions and design basis accidents by meeting the structural integrity performance criterion (SIPC) as given in NEI 97-06. The SIPC margin shall be maintained during plant operation between tube examinations. The burst strength of tubes subjected to wall thinning will depend on the extent and mode of degradation, and the magnitude of design loads to include pressure differential across the tube wall during normal operation and postulated accident conditions. In addition, non-pressure loads that can occur during postulated accident events shall be evaluated and included in the assessment of tube integrity if determined to significantly reduce the tube burst strength. The EPRI Flaw Handbook provides burst pressure relationships for flaws which include a reduction factor that accounts for the effect of applied bending stress on circumferential degradation. However, this previous industry work was only for planar crack-like flaws and did not directly address uniform volumetric wall loss which can have both axial and circumferential extent. This paper describes a test program to determine the effect of bending loads on the burst pressure of a tube with uniform thinning over a given axial length. The uniform thinning geometry was selected since it represented a bounding case of general wall loss and is conservative for calculating a tube repair limit for volumetric degradation for a given steam generator design. Tube repair limits are required for defining an upper limit on in-service degradation for which a tube is to be removed from service. Tube repair limits are cited in the Plant Technical Specifications, which is an important part of the licensing basis.

Limit Loads for Circumferential Cracks in the Steam Generator Tube under Combined Pressure and Bending Loads

2007 ◽  
Vol 345-346 ◽  
pp. 1357-1360
Author(s):  
Hyun Su Kim ◽  
Tae Eun Jin ◽  
Hong Deok Kim ◽  
Han Sub Chung ◽  
Yoon Suk Chang ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Plastic Material ◽  
Material Behavior ◽  
Limit Loads ◽  
Integrity Assessment ◽  
Bending Loads ◽  
Steam Generator Tubes

Steam generator in a nuclear power plant is huge heat exchanger that transfers heat from reactor to make steam to drive turbine-generator. Failure of the steam generator tubes can result in the release of fission products to the secondary side. Therefore, accurate integrity assessment of the cracked steam generator tubes is of great importance for maintaining the safety of the nuclear power plant. This paper provides limit loads for circumferential through-wall cracks in steam generator tubes under combined internal pressure and bending loads. Such limit loads are developed on the basis of three dimensional finite element analyses assuming elastic-perfectly plastic material behavior. As for the crack location, both the top of the tubesheet and U-bend regions are considered. The analysis results can be directly applied to the practical integrity assessment of cracked steam generator tubes, because the comparison between experimental data and FE results shows a very good agreement.

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.

Assessment of Limit Loads for Circumferential Cracks in Steam Generator Tube Considering Constraining Effect of Tube Support

2005 ◽  
Author(s):  
Hyun Su Kim ◽  
Jong Sung Kim ◽  
Tae Eun Jin ◽  
Hong Deok Kim ◽  
Han Sub Chung
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Three Dimensional ◽  
Fission Products ◽  
Limit Load ◽  
Steam Generator Tube ◽  
Integrity Assessment ◽  
Steam Generator Tubes

The steam generator in a nuclear power plant is a large heat exchanger that uses heat from reactor to generate steam to drive the turbine generators. Rupture of a steam generator tube can result in release of fission products to environment. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining safety of a nuclear power plant. The steam generator tubes are supported at periodic intervals by support plates and rotations of the tubes are constrained. Although it was reported that the limit load for a circumferential crack was significantly affected by boundary condition of the tube, existing limit load solutions do not include the constraining effect of tube supports. This paper provides detailed limit load solutions for circumferential cracks in steam generator tubes considering the actual boundary conditions to simulate the constraining effect of the tube supports. Such solutions are developed based on three dimensional (3D) finite element analyses. The resulting limit load solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.

Pulse Electroplating of Ni-P-Nano TiO2 and ZrO2 for Steam Generator Tube Repair

2007 ◽  
Vol 26-28 ◽  
pp. 1067-1070
Author(s):  
Kee Won Urm ◽  
Seon Ho Lee ◽  
Woo Sung Kim ◽  
Chang Yeol Cho ◽  
Jong Ho Lee
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Radioactive Material ◽  
Alloy 600 ◽  
Nano Tio2 ◽  
Steam Generator Tube ◽  
Pulse Electroplating ◽  
Tube Repair

Steam generator tubes provide the pressure boundary between the primary and secondary regions of a nuclear power plant. Alloy 600 is a tube material with good corrosion resistance; however, tubes of this material have experienced damage, particularly as Stress Corrosion Cracking, under the elevated temperature and pressure environment of a nuclear power plant. These damaged tubes must be repaired to prevent leakage of radioactive material from the primary to the second regions in the nuclear steam generator. In this study, Ni-P-Nano TiO2 and ZrO2 layers were produced by pulse electroplating for steam generator tube repair. These electroplate layers were obtained from Ni sulfamate bath with an added small quantity of H3PO3 and Nano TiO2 and ZrO2 particles with an average size of 20-80nm. Results of TEM analysis in these layers show that Nano TiO2 and ZrO2 particles were uniformly distributed into the electroplated Ni matrix and the tensile strength of these layers at 800-1000MPa was higher than that of alloy 600 with a conventional pure Ni electroplate layer.

Development of an Integrity Evaluation System on the Basis of Cooperative Virtual Reality Environment for Nuclear Power Plant

2004 ◽  
Author(s):  
J. C. Kim ◽  
J. B. Choi ◽  
Y. H. Choi
Keyword(s):  
Virtual Reality ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Evaluation System ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Virtual Reality Environment ◽  
Reactor Pressure

Since early 1950’s fracture mechanics has brought significant impact on structural integrity assessment in a wide range of industries such as power, transportation, civil and petrochemical industries, especially in nuclear power plant industries. For the last two decades, significant efforts have been devoted in developing defect assessment procedures, from which various fitness-for-purpose or fitness-for-service codes have been developed. From another aspect, recent advances in IT (Information Technologies) bring rapid changes in various engineering fields. IT enables people to share information through network and thus provides concurrent working environment without limitations of working places. For this reason, a network system based on internet or intranet has been appeared in various fields of business. Evaluating the integrity of structures is one of the most critical issues in nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including regular in-service inspection, fracture mechanics analysis, etc. And thus, experts in different fields have to cooperate to resolve the integrity problem. In this paper, an integrity evaluation system on the basis of cooperative virtual reality environment for reactor pressure vessel which adapts IT into a structural integrity evaluation procedure for reactor pressure vessel is introduced. The proposed system uses Virtual Reality (VR) technique, Virtual Network Computing (VNC) and knowledge based programs. This system is able to support 3-dimensional virtual reality environment and to provide experts to cooperate by accessing related data through internet. The proposed system is expected to provide a more efficient integrity evaluation for reactor pressure vessel.

Critical Section Selection Methodology for the U.S. EPR™ Standard Nuclear Power Plant

2010 ◽  
Author(s):  
Se-Kwon Jung ◽  
Adam Goodman ◽  
Joe Harrold ◽  
Nawar Alchaar
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Structural Design ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Evaluation ◽  
Critical Section ◽  
Selection Methodology ◽  
Critical Sections ◽  
The U.S

This paper presents a three-tier, critical section selection methodology that is used to identify critical sections for the U.S. EPR™ Standard Nuclear Power Plant (NPP). The critical section selection methodology includes three complementary approaches: qualitative, quantitative, and supplementary. These three approaches are applied to Seismic Category I structures in a complementary fashion to identify the most critical portions of the building whose structural integrity needs to be maintained for postulated design basis events and conditions. Once the design of critical sections for a particular Seismic Category I structure is complete, the design for that structure is essentially complete for safety evaluation purposes. Critical sections, taken as a whole, are analytically representative of an “essentially complete” U.S. EPR™ design; their structural design adequacy provides reasonable assurance of overall U.S. EPR™ structural design adequacy.

Overview of the nuclear power plant structural integrity research in Finland

1993 ◽  
Vol 55 (1) ◽  
pp. 3-59 ◽  
Author(s):  
K. Törrönen ◽  
P. Aaltonen ◽  
H. Hänninen ◽  
K. Mäkelä ◽  
P. Karjalainen-Roikonen ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity

Leak behavior of SCC degraded steam generator tubings of nuclear power plant

2005 ◽  
Vol 235 (23) ◽  
pp. 2477-2484 ◽  
Author(s):  
Seong Sik Hwang ◽  
Hong Pyo Kim ◽  
Joung Soo Kim ◽  
Kenneth E. Kasza ◽  
Jangyul Park ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power
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