Information Notice 1996-38, Results of Steam Generator Tube Examinations

Results of Steam Generator Tube Examinations
	ML031060026
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	06/21/1996
From:	Grimes B; Office of Nuclear Reactor Regulation
To:	;
References
	IN-96-038, NUDOCS 9606180338
	Download: ML031060026 (8)
	v • d • e

UNITED STATES

NUCLEAR REGULATORY COMMISSION

-A OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001 June 21, 1996 NRC INFORMATION NOTICE 96-38: RESULTS OF STEAM GENERATOR TUBE EXAMINATIONS

Addressees

All holders of operating licenses or construction permits for pressurized

water reactors (PWRs).

Purposg

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information

notice to promulgate information about steam generator tube examinations. It

is expected that recipients will review the information for applicability to

their facilities and consider actions, as appropriate, to avoid similar

problems. However, suggestions contained in this information notice are not

NRC requirements; therefore, no specific action or written response is

required.

Description of Circumstances

Improved techniques and equipment are constantly developed to detect flaws in

steam generator tubes. In addition, as nuclear power plants get older, different degradation mechanisms of steam generator tubes occur. This

information notice discusses recent experiences by licensees involving these

new techniques and equipment and different degradation mechanisms.

Recent steam generator tube examinations have revealed degradation at a number

of locations, such as in dented areas, the expansion transition region, the

freespan region, and in the tubesheet crevice. The types of degradation

observed in these locations are discussed below. In addition to identifying

several degradation mechanisms, these examinations raised a number of

technical issues with respect to classifying inspection results, periodicity

of examinations, and expanding the initial inspection scope.

Axial and circumferential indications at dented tube support plates were

identified at a number of plants, including Sequoyah Nuclear Plant Unit 1, Diablo Canyon Nuclear Power Plant Unit 1, and Salem Generating Station Unit 1.

These indications are associated with minor dents (i.e., dents that can be

inspected with a standard size probe). These dented regions were examined

with Cecco probes or rotating probes with plus-point coils or pancake coils

(or both). On the basis of the examinations, the axial indications appear to

have initiated from the inside diameter of the tube, and the circumferential

indications appear to have initiated from the outside diameter of the tube.

However, at Diablo Canyon Unit 1, several circumferential indications have

initiated from the inside diameter of the tube (as evidenced by destructive

examination).

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IN 96-38 June 21, 1996 Some plants that have Combustion Engineering and Westinghouse-designed steam

generators also reported circumferential indications at the expansion

transition region. Among these are Sequoyah Nuclear Plant Unit 1, Diablo

Canyon Unit 1, Salem Unit 1, Arkansas Nuclear One Unit 2, Braidwood Unit 1, Byron Unit 1, and Callaway Unit 1. At particular plants, from tens to

thousands of indications were reported.

The circumferential indications at the expansion transition have occurred at

roll expansions, kinetic/explosive expansions, and hydraulic expansions. For

example, circumferential indications have been reported in mechanically roll- expanded tubes at Farley Unit 1, Westinghouse explosively expanded (i.e.,

WEXTEX) tubes at Sequoyah Unit 1, Salem Unit 1, and Diablo Canyon Unit 1, Combustion Engineering explosively expanded tubes (i.e., EXPLANSION tubes) at

Arkansas Nuclear One Unit 2, and in hydraulically expanded tubes at Callaway

Unit 1. The majority of these indications were seen at the hot-leg expansion

transition; however, circumferential indications were reported at the cold-leg

expansion transition at Arkansas Nuclear One Unit 2. The circumferential

cracks detected at these plants were all in Alloy 600 mill-annealed tubes.

Freespan degradation has been reported at a few plants. Freespan degradation

is degradation observed above the sludge pile region at the top of the

tubesheet and is not located at any support structure (e.g., tube support

plates including eggcrates, anti-vibration bars, and batwings). Historically, moderate amounts of freespan degradation had been observed at McGuire Units 1 and 2 and at Palo Verde Units 1, 2, and 3. During the fall outages, Arkansas

Nuclear One Unit 2, Farley Unit 1, and Point Beach Unit I reported freespan

tube degradation. In addition, Oconee Units 1, 2, and 3 reported freespan

axial indications attributed to intergranular attack.

A few plants have tubes which are only partially expanded in the tubesheet.

As a result, there is a crevice between the tube and the tubesheet for. the

portion of the tube in the tubesheet that is not expanded. Corrosion products

can accumulate in this crevice and can lead to tube degradation. Historically, tubesheet crevice region defects have been observed with the bobbin coil and

repaired, accordingly; however, many of the indications detected during

outages this fall were not found with the conventional bobbin coil probe. As

a result, extensive examinations using alternate techniques were performed

(e.g., rotating pancake coil examinations). Extensive tube repairs were

performed, such as sleeving at Zion Unit I and tube rerolling at Point Beach

Unit 1.

Discussion

Steam generators with mill-annealed Alloy 600 steam generator tubes are

susceptible to such degradation as stress corrosion cracking. Degradation has

been observed in the hot legs and cold legs of the steam generator tubes, in

the expanded portion of the tube, at the expansion transition, in the

tube-to-tubesheet crevice, in the sludge pile, in the freespan, and at tube

support structures such as the tube support plate, batwings, anti-vibration

bars, and vertical straps. The severity of the degradation and the number of

tubes affected tend to be plant specific since these depend on many factors

-IN 96-38 June 21, 1996 such as temperature, operating time, water chemistry history, and tube

mechanical properties, including microstructure. Inspections have illustrated

the importance of comprehensive steam generator tube examinations using

appropriate techniques to ensure tube integrity even if a specific type of

degradation has not been observed at a given location in the past. Previous

inspection findings do not ensure that a location/tube is not susceptible to a

particular mechanism. For example, before the inspections at Callaway Unit 1, no circumferential cracking had occurred domestically at tubes which had been

hydraulically expanded within the tubesheet. The inspections at Callaway

demonstrate that continually assessing the condition of all portions of the

steam generator tube can ensure that new forms of degradation are detected.

The recent inspections also indicate the importance of comprehensively

examining all portions of the steam generator tubes using techniques and

equipment capable of reliably detecting degradation to which the steam

generator tubes may potentially be susceptible. This experience calls into

question the effectiveness of the bobbin coil for detecting circumferential

indications or for detecting indications where significant interfering signals

exist (e.g., expansion transition locations, dented locations, and locations

with excessive deposits), as discussed in NRC Information Notice 94-88, Inservice Inspection Deficiencies Result in Severely Degraded Steam Generator

Tubes." In addition, this experience further indicates that a generically

qualified technique may need to be supplemented to account for the testing

conditions at a specific plant. Furthermore, optimizing such test variables

as probe design and frequencies for the type of degradation observed at the

plant such as inside-diameter initiated indications versus outside-diameter

initiated indications, and controlling such test variables as cable length and

capacitance within the range for which the technique was qualified can be

important in ensuring the reliable detection of degradation.

Several large indications were detected during the most recent examinations of

steam generator tubes. As a result, several licensees took additional

measures to ensure that all tubes were capable of withstanding the pressure

loadings specified in Regulatory Guide 1.121, *Bases for Plugging Degraded PWR

Steam Generator Tubes." These additional measures (insitu pressure testing

and removing tubes for destructive examination) were performed even though

many of these indications were repaired. Although methods other than removing

tubes for destructive examination exist for evaluating tube integrity, tube

removal has the advantage of assessing inspection reliability, developing

additional confidence in the ability to size indications, determining the root

cause of the degradation, and possibly identifying corrective actions.

Assessment of the inspection findings after every inspection assures that all

tubes are capable of performing their intended safety function for the planned

operating interval. In some instances, these assessments have led to

mid-cycle inspections.

When degraded tubes are left in service (i.e., for degradation mechanisms for

which qualified sizing techniques exist), assessment of the acceptable

operating interval typically involves a detailed knowledge of the growth rate

of the degradation, the scope of the examination, and the capabilities of the

inspection technique.

IN 96-38 June 21, 1996 7.

For degradation mechanisms for which there is no qualified depth sizing

technique, a tube with an indication typically has been considered defective.

In these instances, demonstrating that the largest indications detected during

an inspection were capable of withstanding specified pressure loadings

(through such techniques such in-situ pressure testing or burst and leakage

testing or both) can provide assurance that tubes currently without

indications will also be capable of withstanding specified pressure loadings

at the end of the next inspection interval, if the interval is of comparable

duration and operating parameters (e.g., water chemistry and hot leg

temperature) to the previous inspection interval.

Although only steam generators that contain tubes made from mill-annealed

Alloy 600 are discussed above, the information may have applicability to all

PWRs. This information notice requires no specific action or written

response. If you have any questions about the information in this notice, please contact one of the technical contacts listed below or the appropriate

Office of Nuclear Reactor Regulation (NRR) project manager.

Brian K. Grimes, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Kenneth J. Karwoski, NRR

(301) 415-2754 Internet:kjkl@nrc.gov

Eric J. Benner, NRR

(301) 415-1171 Internet:ejbl@nrc.gov

Attachment: List of Recently Issued NRC Information Notices

Attachment

IN 96-38 June 21, 1996 Page I of I

LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

96-37 Inaccurate Reactor Water 06/18/96 All pressurized water

Level Indication and Inad- reactor facilities holding

vertent Draindown During an operating license or a

Shutdown construction permit

96-36 Degradation of Cooling 06/12/96 All holders of OLs or CPs

Water Systems Due to Icing for nuclear power reactors

96-35 Failure of Safety Systems 06/11/96 All U.S. Nuclear Regulatory

on Self-Shielded Irradia- Commission irradiator

tors Because of Inadequate licensees and vendors

Maintenance and Training

96-34 Hydrogen Gas Ignition 05/31/96 All holders of OLs or CPs

during Closure Welding for nuclear power reactors

of a VSC-24 Multi-Assembly

Sealed Basket

96-33 Erroneous Data From 05/24/96 All material and fuel cycle

Defective Thermocouple licensees that monitor tem- Results in a Fire perature with thermocouples

96-32 Implementation of 10 CFR 06/05/96 All holders of OLs or CPs

50.55a(g) (6)(ii) (A), for nuclear power reactors

Augmented Examination

of Reactor Vessels

96-31 Cross-Tied Safety Injec- 05/22/96 All holders of OLs or CPs

tion Accumulators for pressurized water

reactors

96-30 Inaccuracy of Diagnostic 05/21/96 All holders of OLs or CPs

Equipment for Motor- for nuclear power reactors

Operated Butterfly Valves

96-29 Requirements in 10 CFR 05/20/96 All holders of OLs or CPs

Part 21 for Reporting and for nuclear power reactors

Evaluating Software Errors

OL - Operating License

CP - Construction Permit

IN 96-38 June 21, 1996 For degradation mechanisms for which there is no qualified depth sizing