R590.1 Évaluation réglementaire des fuites qui s’écoulent par des fissures dans des composants de conduites

1.0 Background

Steam generator tubes represent a major portion of the reactor primary coolant pressure boundary surface area in both CANDU reactors and Pressurized Water Reactors (PWR). These tubes have an important safety role because they constitute one of the primary barriers between the radioactive and non-radioactive sides of the plant. The ability to estimate the leak rates from through wall cracks in a steam generator tube is important to calculate source terms, assess the operational management of steam generators and to demonstrate the leak-before-break deterministic analysis methodology.

Steam generator tubes are subject to several degradation mechanisms such as; intergranular attack and outside diameter stress corrosion cracking (SCC), primary water stress corrosion cracking, tube fretting and wear, foreign material wear, pitting, high cycle fatigue, and wastage or thinning. Of these, environmentally induced degradation through intergranular SCC and intergranular attack are the most serious degradation mechanisms. These degradations commonly occur in crevice regions at tube support plates and tube sheet locations or under sludge piles. Intergranular SCC has also been observed in the free span of steam generator tubes.

Steam generator tubes must support more than 10 MPa pressure differential during normal operation. In the event of a main steam line break, where the secondary side drops to atmospheric pressure, the tube wall differential pressure can be more than double that of normal operation. The steam generator tube design requirement according to ASME code and the United States Nuclear Regulatory Commission (US NRC) is 1.4?PMSLR > 25 MPa [1]. This safety margin is based on the rupture or burst pressure of an unflawed tube. A typical unflawed Alloy 600 tube has an industry expected burst pressure of approximately 86 MPa [2].

A sufficient safety margin against tube rupture was originally the basis to confirm a suitable level of safety and reliability. Steam generator operability requires the completion of steam generator tube inspections using non-destructive techniques, usually eddy current examination, in intervals ranging from 12 to 40 months. Steam generator tube integrity assessment guidelines provide assessment procedures and criteria to evaluate for tube integrity as well as define loading definitions and design margins [3]. In support of assessed safe operation, nuclear regulators accept the leak-before-break concept where a through-wall flaw will render detection from leak monitors well before tube rupture. Much research in the area of steam generator tube integrity is therefore related to the burst characterization of flawed tubes [4].

To this present date most experimental data concerning simulated steam generator tube cracks were obtained with tube specimens ultimately leading to a burst condition. From such experiments, the break geometry characterization often failed to provide a detailed link to fracture and fluid mechanics; therefore, predicting and benchmarking leak rates through degraded steam generator tubes is challenging. Most studies of sub-cooled choking flow are related to long tubes with large L/D and nozzles [5-7]. The survey shows that this geometry has been studied over a large range of pressures and liquid sub-cooling conditions, but the data is focused on L/D geometries greater than 15. Also, all data have a channel length (wall thickness) greater than 10 mm, which is not indicative of steam generator tubing. Steam generator tubes have a wall thickness typically less than 3 mm. An earlier CNSC research project focused on leakage through mini and micro slits [9-12] to fill gaps in the available database. In view of this, an experimental program needs to be carried out where parameters and results are well characterized and conditions include choking flow of sub-cooled water tests on steam generator tube samples with actual crack-like geometries based upon tube degradation mechanisms discussed in Reference [8].

References

[1] S. Majumdar, S., 1999, “Prediction of structural integrity of steam generator tubes under severe accident conditions,” Nuclear Engineering Design, Volume 194, p. 31-55. 1999.
[2] Vaia, A. R., Stevens, J. M., Prabhu, P. J., 2005, “Laboratory examination of pulled steam generator tube with free span axial ODSCC,” 12th International Conference On Environmental Degradation of Materials in Nuclear Power Systems, Salt Lake City, Utah, Minerals, Metals and Materials Society, pp. 1113-1123, 2005.
[3] E. Fuller et al., “Steam Generator Integrity Assessment Guidelines” Revision 2 Non-Proprietary Version 1012987, Final Report. EPRI Technical Report TR-1012987, 2006.
[4] R. F. Keating, P. Hernalsteen, J. A. Begley, “Burst pressure correlation for steam generator tubes with throughwall axial cracks,” Electric Power Research Institute, Palo Alto, California, EPRI Technical Report TR-105505, 1955.
[5] R. E. Henry, “A study of one- and two-component twophase critical flows at low qualities,” Argonne National Laboratory, Report ANL-7430, 1968.
[6] R. E. Henry, H. K. Fauske, “The two-phase critical flow of one-component mixtures in nozzles, orifices, and short tubes,” Journal of Heat Transfer, 93, p. 179-187, 1971.
[7] R. E. Henry, Heat Transfer at Low Temperature, Plenum Press, New York, 1975.
[8] S. Pagan, X. Duan, M.K. Kozluk, MB. Mills, G. Goszczynski, “Characterization and structural integrity tests of ex-service steam generator tubes at Ontario Power Generation,” Nuclear Engineering Design, Volume 223, p. 477-481, 2009.
[9] S. T. Revankar, Brian Wolf and Anand Vadlamani, (2013), Assessment of Leak Rates through Steam Generator Tubes, RSP-0294, Canadian Nuclear Safety Commission, 2013, (available at http://www.nuclearsafety.gc.ca/eng/resources/research… )
[10] Revankar, S.T., Riznic, J. R., (2009), Assessment of Steam Generator Tube Flaw size and Leak Rate Models, Nuclear Technology, 167, (1), 157-168, July 2009.
[11] S. T. Revankar, B. Wolf, R.A. Vadlamani, and J.R. Riznic, Sub-Cooled Flashing Flow in a Steam Generator Tube Crack, Paper NURETH15-098, Proceedings of the 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-15, Pisa, Italy, May 12-17, 2013
[12] Revankar, S.T., Wolf, B., Riznic, J.R., (2013), Flashing Flow of Subcooled Liquid through Small Cracks, Procedia Engineering, Vol. 56, 454-461 (2013).

2.0 Objectives

To develop a consolidated steam generator tube database and leak rates models. The output from this project will propose an improved steam generator tube fitness for service assessment methodology. This methodology may form the technical basis for enhanced regulatory requirements to address steam generator tube aging and degradation.

3.0 Scope of Work

This project involves conducting a literature review, collecting open-source data, producing a test matrix, performing experimental tests, and developing quality-assured models. The work shall consider steam generator tubing used in both CANDU and Pressurized Water Reactor (PWR) technologies.

The experimental tests shall produce data for steam generator tube specimen degradations based on primary water stress-corrosion-cracks and/or fatigue cracks in tube specimens. At least one set of experiments shall be performed to replicate degradation observed on CANDU feeder piping.

The project duration is estimated at thirty-six (36) months.

4.0 Tasks to be Performed

Phase I – Literature review and data collection

4.1 Perform a literature review of the following documents concerning leakage through metallic components:

• Relevant regulatory documents developed by the US NRC (e.g. RG 1.45, NUREG reports)
• Relevant OECD Nuclear Energy Agency documents
• Work performed by Battelle, INPO, EPRI, and Oak Ridge National Laboratories
• Work in support of computer code development for SQUIRT, xLPR, and LEAPOR

4.2 Perform a literature review of the following topics concerning steam generator tube degradation:

• Thermal hydraulics heat transfer
• Fluid flow through micro and nano channels
• Applicable fractal geometry for cracks in metallic tubes
• Fracture mechanics of cracks

4.3 Consolidate CANDU and PWR steam generator data, identifying the: 1) degradation mechanism, 2) assessed source of the flaw initiation, and 3) crack growth based on fracture mechanics modelling for each case. Sources from this data shall include: EPRI, University of California – Berkley, Purdue University, Argonne National Laboratory, US NRC International Steam Generator Tube Integrity Program (ISG TIP), amongst others.

Phase II – Experiments

4.4 Develop a test matrix to simulate prototypical working parameters relevant to CANDU and PWR steam generator tubing. The test matrix shall be subject to CNSC review and approval.

4.5 Prepare a minimum of ten (10) test specimens to replicate realistic stress-corrosion-cracks; primary water stress-corrosion-cracks and/or fatigue cracks. The test specimens shall be suitable for collecting data under critical single phase and flashing leakage conditions. EDM notches or laser cut slits are not of interest for this study.

4.6 Characterize each test specimen for its three-dimensional profilometry of a leakage path through the crack. This characterization may be performed by either computer tomography or by subjecting the test specimens to a liquid nitrogen environment (or other suitable technique) to open up cracks and prepare for metallurgical, fractographic analysis.

Phase III – Modelling

4.7 Develop models to account for defect and size dependent crack growth based on experiments and current knowledge.

4.8 Develop deterministic and probabilistic assessment methodology for leak rate modelling in steam generator tubes. Combine the improved leak rates model based on experiments data with probabilistic crack (fracture) growth model into an integrated deterministic and/or probabilistic leak rate model for SGT. Implement different flow growth simultaneously and improve the probable leak rate predictions for SGT.