Information Notice 1995-40, Supplemental Information to Generic Letter 95-03, Circumferential Cracking of Steam Generator Tubes.

Supplemental Information to Generic Letter 95-03, Circumferential Cracking of Steam Generator Tubes.
	ML031060272
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:	09/20/1995
From:	Crutchfield D; Office of Nuclear Reactor Regulation
To:	;
References
	GL-95-003 IN-95-040, NUDOCS 9509140386
	Download: ML031060272 (9)
	v • d • e

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UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001 September 20, 1995 NRC INFORMATION NOTICE 95-40: SUPPLEMENTAL INFORMATION TO GENERIC LETTER

95-03, "CIRCUMFERENTIAL CRACKING OF STEAM

GENERATOR TUBESH

Addressees

All holders of operating licenses or construction permits for nuclear power

reactors.

Purpose

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

examination

notice to provide additional information on steam generator tube in

results from Maine Yankee Atomic Power Station as previously discussed

Generic Letter (GL) 95-03, "Circumferential Cracking of Steam Generator

for

Tubes." It is expected that recipients will review the information to

applicability to their facilities and consider actions, as appropriate, in this information

avoid similar problems. However, suggestions contained written

notice are not NRC requirements; therefore, no specific action or

response is required.

Description of Circumstances

The staff issued GL 95-03, to obtain information necessary to assess in light

compliance with requirements regarding steam generator tube integritythe staff

of the inspection findings at the Maine Yankee plant. In GL 95-03, with respect

requested that utilities (1) evaluate recent operating experiencedevelop a

to the detection and sizing of circumferential indications, (2)

safety assessment justifying continued operation until the next scheduled for the

steam generator tube inspections are performed, and (3) develop plans detection of

next inspections of steam generator tubes as they pertain to the

circumferential cracking. Since the issuance of GL 95-03, additional

analysis

information pertaining to in situ pressure testing and destructive In

for the tubes removed from the Maine Yankee plant has become available.

addition, the wrong title given to NUREG-0844 in GL 95-03 was erroneously

indicated as, "Voltage-Based Interim Plugging Criteria for Steam Generator of

Tubes." The correct title is, "NRC Integrated Program for the Resolution Tube

Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator

Integrity."

Discussion

for Maine

On July 15, 1994, Maine Yankee Atomic Power Company, the licensee

leak rate

Yankee, shut down the plant when the measured primary-to-secondarythe plant, approached 189 liters [50 gallons] per day. After shutting down

1h501.19

20027 P9R

'OqI 'Pfk

IN 95-40

September 20, 1995 94-88, the licensee tested for leaks and found four leaking tubes. INSteam Generator

Degraded

Inservice Inspection Deficiencies Result in Severely the licensee in 1994, by

Tubes," discusses in situ pressure testing performed to assess their actual

on tubes containing some of the largest indications, not be pressurized due to

burst integrity. At that time, certain tubes could the staff had not

a combination of leakage and pump capacity limitations, and to simulate an actual

reached a conclusion regarding the validity of the tests

pressure transient in the steam generators.

In 1995, the licensee performed additional steam generator inspections. Seven

three of which were from the

tubes were subjected to in situ pressure testing, four of which were

sample subjected to in situ pressure testing in 1994 and

identified at the end of the

tubes containing some of the largest indications that the tubes were

1994-to-1995 operating interval. The testing indicated

in excess of the loads for which

capable of withstanding pressure loadings with the

failure would be predicted on the basis of the size estimates which the tubes were

standard pancake coil. Furthermore, the pressures to

loads. NRC Regulatory Guide 1.121, subjected were greater than design-basis indicates that tubes

Bases for Plugging Degraded PWR Steam Generator Tubes,"

pressure' and '1.4 times main

should be able to withstand R3 times operating At Maine Yankee, 3 times

steam line break maximum pressure' without bursting.

to 34.47 MPa [5000 psi] and 1.4 operating pressure is approximately equal MPa [4057 psi]. All

times main steam line break maximum pressure equals 27.9739.30 [5700 psi]

tested tubes at Maine Yankee were subjected to at least leakageHPa and no tubes

hydrostatic pressure. Three tubes exhibited no defect

that these tests adequately bound main steam

burst. The staff has concluded

line break loads on steam generator tubes.

As stated in GL 95-03, three tubes were removed from the marginal Maine Yankee steam

plus-point

generators for destructive examination: two tubes with less than

coil responses (sized by the eddy current analysts as probably

and one with an intermediate response (sized by

40 percent through-wall depth)

40 percent through-wall

the eddy current analysts as probably greater than examined

depth). Before the tubes were removed, they were with several

as ultrasonic, fluorescent penetrant, and eddy

nondestructive methods, such The eddy current

current techniques to confirm the nature of the indications. coil, a

methods included examination with a standard rotating pancake indications were

plus-point coil, and a high-frequency pancake coil. The

sized with various techniques. The size estimates for thecalibration high-frequency

coil were obtained after of the

pancake coil and the plus-point within a

probes on electric discharge-machined (EDM) notches contained of the

standard. With the high-frequency pancake coil, the most sensitive pulled tubes

coils to the degradation at Maine Yankee, the indications on44 the percent, and

were sized with maximum through-wall depths of 36, 32, andThe average depth

average depths of 30, 21, and 27 percent, respectively.

from the

estimates obtained from the eddy current examination are calculated the maximum

maximum depth and the circumferential extent by assuming that circumferential

depth is the depth of the degradation over the entire measured

arc length and averaging this estimate over the entire tube circumference.

tubes indicated

The corresponding destructive examination results for theseaverage depths of

that the maximum depths were 45, 37, and 57 percent, with

V~ IN 95-40

September 20, 1995 24, 23, and 26 percent, respectively. The destructive examination of these

tubes indicated that numerous small cracks had initiated at various locations

about the circumference and at various elevations (axial locations) within a

1.27 mm [0.05 inch] band inthe "expansion transition region of the tubes, noncorroded ligaments existed between some of the cracks. The cracks

initiated at the inner diameter of the tubes. The licensee compared the

sizing of several of the larger indications that were inspected with both a

standard pancake coil and the high-frequency pancake coil. The high-frequency

pancake coil is,ingeneral, more sensitive than the standard pancake coil to

cracks initiating at the inner diameter. The results of this comparison

estimated by the high-frequency

indicated that the maximum and average depths maximum

pancake coil were consistently lower than the though the and average depths

length (i.e.,

estimated with the standard pancake coil even with the high-frequency coil.

circumferential extent) estimates were longer

high-frequency coil suggest that

The smaller depth estimates obtained with the structurally

many of the indications may not have been as significant as the

in IN 94-88. Furthermore, standard pancake coil suggested and as was reportedcracks were not coplanar, but

the destructive examination indicated that the over a short axial

rather of short circumferential length and staggered between the cracks. Due

region. There were, in fact, ligaments of materialbetween the cracks), the

to the nature of this cracking (i.e., the spacing by the nondestructive

ligaments of sound material could not be distinguished coil and plus-point

examination (i.e., standard and high-frequency pancake data are conservative in

coil) data; however, the nondestructive examination by the

that the tubes are most likely more structurally character estimated

sound than

of these cracks is

eddy current examination. The observed segmented examination results at

consistent with the results of fluorescent penetrant observed on

Maine Yankee and with the morphology of circumferential cracks

specimens of tubes pulled from other plants.

This information notice requires no specific action or written response. If

you have any questions about the informationor in this notice, please contact

the appropriate Office of

one of the technical contacts listed belowmanager.

Nuclear Reactor Regulation (NRR) project

Denln sM. Crutch e rector

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Kenneth J. Karwoski, NRR

(301) 415-2754 Eric J. Benner, NRR

(301) 415-1171 Attachment:

List of Recently Issued NRC Information Notices

hb d& /1 a-c

w iachment

IN 95-40

September 20, 1995 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject - Issuance Issued to

95-39 Brachytherapy Incidents 09/19/95 All U.S. Nuclear Regulatory

Involving Treatment Commission Medical

Planning Errors Licensees.

95-38 Degradation of Boraflex 09/08/95 All holders of OLs or CPs

Neutron Absorber in for nuclear power reactors.

Spent Fuel Storage Racks

95-37 Inadequate Offsite Power 09/07/95 All holders of OLs or CPs

System Voltages during for nuclear power reactors.

Design-Basis Events

95-36 Potential Problems with 08/29/95 All holders of OLs or CPs

Post-Fire Emergency for nuclear power reactors.

Lighting

95-35 Degraded Ability of 08/28/95 All holders of OLs or CPs

Steam Generators to for pressurized water

Remove Decay Heat by reactors (PWRs).

Natural Circulation

95-34 Air Actuator and Supply 08/25/95 All holders of OLs or CPs

Air Regulator Problems in for nuclear power reactors.

Copes-Vulcan Pressurizer

Power-Operated Relief Valves

93-83, Potential Loss of Spent 08/24/95 All holders of OLs or CPs

Supp. 1 Fuel Pool Cooling After a for nuclear power reactors.

Loss-of-Coolant Accident

or a Loss of Offsite Power

95-33 Switchgear Fire and 08/23/95 All holders of OLs or CPs

Partial Loss of Offsite for nuclear power reactors.

Power at Waterford

Generating Station, Unit 3

95-10, Potential for Loss of 08/11/95 All holders of OLs or CPs

Supp. 2 Automatic Engineered for nuclear power reactors.

Safety Features Actuation

OL - Operating License

CP - Construction Permit

IN 95-40

V

\_ September 20, 1995 24, 23, and 26 percent, respectively. The destructive examination of these

tubes indicated that numerous small cracks had initiated at various locations

about the circumference and at various elevations (axial locations) within a

1.27 mm [0.05 inch] band in the "expansion" transition region of the tubes, Noncorroded ligaments existed between some of the cracks. The cracks

initiated at the inner diameter of the tubes. The licensee compared the

sizing of several of the larger indications that were inspected with both a

standard pancake coil and the high-frequency pancake coil. The high-frequency

pancake coil is, in general, more sensitive than the standard pancake coil to

cracks initiating at the inner diameter. The results of this comparison

indicated that the maximum and average depths estimated by the high-frequency

pancake coil were consistently lower than the maximum and average depths

estimated with the standard pancake coil even though the length (i.e.,

circumferential extent) estimates were longer with the high-frequency coil.

The smaller depth estimates obtained with the high-frequency coil suggest that

many of the indications may not have been as structurally significant as the

standard pancake coil suggested and as was reported in IN 94-88. Furthermore, the destructive examination indicated that the cracks were not coplanar, but

rather of short circumferential length and staggered over a short axial

region. There were, in fact, ligaments of material between the cracks. Due

to the nature of this cracking (i.e., the spacing between the cracks), the

ligaments of sound material could not be distinguished by the nondestructive

examination (i.e., standard and high-frequency pancake coil and plus-point

coil) data; however, the nondestructive examination data are conservative in

that the tubes are most likely more structurally sound than estimated by the

eddy current examination. The observed segmented character of these cracks is

consistent with the results of fluorescent penetrant examination results at

Maine Yankee and with the morphology of circumferential cracks observed on

specimens of tubes pulled from other plants.

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.

orig /s/'d by DMCrutchfield

Dennis M. Crutchfield, Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Kenneth J. Karwoski, NRR

(301) 415-2754 Eric J. Benner, NRR

(301) 415-1171 Attachment:

List of Recently Issued NRC Information Notices

DOCUMENT NAME: 95-40.IN

See previous concurrence

To receive a copy of this document, Indicate In the box: 'C' = Copy without attachmentlenclosure 'E' = Copy with attachment/enclosure 'N"

No copy

OFFICE PECB:DRPM I ADM:PUB I JSC/PECB:DRPM [C/EMCB:DE I D/DE

NAME EBenner Tech Editor* EGoodwin* IJStrosnider* BSheron*

DATE 8/3/95 7/25/95 7/25/95 1 08/08/95 _ 08/16/95 OFFICE PECB:DRPM C/PECB:DRPM

NAME IRKiessel* AChaffee* DCr# field

DATE 109/11/95 09/13/95 109/ /K/ 95 _

OFFICIAL RECORD COPY

IN 95-40

.sJb September 20, 1995 24, 23, and 26 percent, respectively. The destructive examination of these

tubes indicated that numerous small cracks had initiated at various locations

about the circumference and at various elevations (axial locations) within a

1.27 mm [0.05 inch] band in the "expansion" transition region of the tubes, noncorroded ligaments existed between some of the cracks. The cracks