Information Notice 1997-49, B&W Once-Through Steam Generator Tube Inspection Findings

B&W Once-Through Steam Generator Tube Inspection Findings
	ML031050389
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, 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:	07/10/1997
From:	Slosson M; Office of Nuclear Reactor Regulation
To:	;
References
	IN-97-049, NUDOCS 9707030183
	Download: ML031050389 (12)
	v • d • e

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001 July 10, 1997 NRC INFORMATION NOTICE 97-49: B&W ONCE-THROUGH STEAM GENERATOR TUBE

INSPECTION FINDINGS

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 notice to present

the findings from the examination of tubes in Babcock and Wilcox (B&W\) once-through steam

generators (OTSGs). 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.

DescriDtion of Circumstances

Licensees using B&W OTSGs have historically observed very little service-induced

degradation in steam generator tubes. During the last few years, however, more degradation

has been observed and this degradation has been seen at a variety of locations, such as

dented (dinged) areas, the expansion transition region, the freespan region, the sludge pile

region, and the sleeve joints. Pertinent inspection findings from steam generator tubes at

several plants with OTSGs are discussed.

Degradation at Dented Locations

Indications of degradation associated with dented (dinged) areas have been found at several

plants-Arkansas Nuclear One, Unit 1 (ANO-1), Oconee Unit 1, and Crystal River Unit 3.

These indications have been axial, circumferential, or volumetric in nature. At ANO-1 (in

1993), two volumetric indications with circumferentially oriented cracklike indications were

found at dents on the secondary face of the upper tubesheet (UTS). These indications were

initially found with a bobbin coil probe and were confirmed to be present with a rotating

pancake coil eddy current inspection probe. At ANO-1 (in 1996), two axially oriented eddy

current indications associated with dented areas in the tube's freespan region were observed.

Similar to the 1993 indications, these indications were also initially found with a bobbin coil

probe. Rotating pancake coil probe inspection of one of these indications confirmed that the

indication initiated from the outside diameter of the tube and that the indication was offset

relative to the tube axis by approximately 35 degrees. At Oconee Unit 1 (in 1995), a

volumetric and a circumferential indication were detected at dents located at the 15th tube

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"> ~IN 97-49 July 10, 1997 support plate. At Crystal River Unit 3 (in 1996), a volumetric eddy current indication was

found at a dented area. This indication was detected with the bobbin coil and confirmed with

a pancake and plus-point coil.

Degradation at the Expansion Transition Region

The expansion transition region of the tubes in B&W OTSGs were heat treated to reduce

residual stresses from tube fabrication and installation, and to increase resistance to primary

water stress corrosion cracking (PWSCC). This heat treatment resulted from a full furnace

stress relief of the entire tube bundle. During the manufacturing process, however, several

tubes were re-rolled into the tubesheet following the full furnace stress relief to temporarily

seal the tube during the shop hydrostatic tests. As a result, a limited population of tubes was

not stress relieved at the expansion transition region (i.e., fewer than 200 tubes are known to

have not been stress relieved). Axial indications associated with the expansion transitions of

both stress-relieved and nonstress-relieved transitions were recently noted in several B&W

units (Davis-Besse, Crystal River 3, ANO-1, and Oconee 3). The inspection findings at these

plants are discussed below.

At Davis-Besse (in the spring of 1996), an axially oriented indication was detected during the

examination of what was believed to be a nonstress-relieved roll transition. This indication

was in the roll transition in the UTS (i.e., the hot leg). Subsequent review of shop records

showed that the expansion transition had not been re-rolled and was, therefore, stress

relieved. The licensee removed the roll transition portion of this tube for destructive

examination. The destructive examination showed that the indication was caused by

PWSCC. To ascertain whether the tube had been stress relieved, the licensee performed

additional analyses and testing. As a result of this testing, the licensee concluded that the

roll transition was not stress relieved (i.e., it had been re-rolled following the full bundle stress

relief process).

At Crystal River 3 (in the spring of 1996), a single axial indication was detected in the roll

transition in a tube that had been re-rolled following the full bundle stress relief (i.e., a

nonstress-relieved transition), and a multiple axial indication was detected in the tube end, above the shop re-roll in the same tube. These indications were located in the roll transition

in the UTS (i.e., the hot leg). The eddy current data clearly indicated that the tube had been

rolled multiple times. The licensee attributed the indication to PWSCC.

At ANO-1 (in the fall of 1996), 24 axially oriented and volumetric indications were detected in

stress-relieved roll transitions in the UTS (i.e., the hot leg). The licensee attributed the axial

indications to inside diameter-initiated stress corrosion cracking (i.e., PWSCC). The

volumetric indications had been initiated on the outside diameter, pointing perhaps to

IN 97-49 July 10, 1997 intergranular attack (IGA) or to closely spaced cracks. To further characterize the nature and

cause for the upper roll transition indications (and other indications), the licensee for ANO-1 removed several tube sections for destructive examination. The destructive examination

findings from the one roll transition indication that was removed confirmed that the indication

was attributable to PWSCC. This transition had been stress relieved.

At Oconee Unit 3 (in the fall of 1996), 19 tubes were identified by eddy current testing as

having PWSCC at the roll transition region in the UTS. Of the 19 indications, 15 were axial

indications in the roll transition region, 3 were axial indications in the rolled area , and 1 was

a volumetric indication at the roll transition region. One tube was removed for laboratory

analysis of the indication at the upper roll transition region. The laboratory destructive

examination findings were not available at the time this notice was prepared.

Degradation at Freespan Locations

Axially oriented degradation in the freespan region has been observed in several B&W

OTSGs (Oconee 1, Oconee 2, Oconee 3, and ANO-1). Freespan degradation is degradation

observed above the sludge pile region and not located at any support structure (e.g., tube

support plates). A freespan axial indication was first identified at Oconee I in May 1994.

This indication was identified with a bobbin coil and confirmed to be present with a rotating

pancake coil probe. This tube along with six others were removed for destructive

examination. The destructive examination confirmed the presence of freespan axially

oriented IGA in all seven tubes. The tube with the indication detected with a bobbin coil was

the most significant with a through-wall depth of 47 percent and a burst pressure of

7400 pounds per square inch (psi), well above the structural criteria specified in Regulatory

Guide 1.121. The IGA in the remaining tubes ranged from 5 percent to 28 percent through- wall.

Subsequent bobbin coil inspections at Oconee Units 1, 2, and 3 identified additional tubes

with freespan axial indications. For example, 9 tubes with indications were detected at

Oconee 2 in October 1994, 22 tubes with indications were detected at Oconee 3 in June

1995, 40 tubes with indications were detected at Oconee 1 in November 1995, and 173 tubes

with indications were detected at Oconee 2 in April 1996. During the Oconee 2 inspection

outage in April 1996, four tubes were removed for destructive examination. The selection

criteria for these, tubes included small and large indications, the number of indications per

tube, and a sampling across the tube bundle. The burst pressures for these tubes ranged

from 5700 psi to 11000 psi. In November 1996, the most recent steam generator inspection

outage at an Oconee unit, 67 tubes with confirmed bobbin coil indications were identified at

Oconee 3. An assessment performed by the licensee, based on previous tube pull analysis, indicated that these tubes had adequate structural integrity. All tubes with axially oriented

freespan IGA, which were confirmed to be present with a rotating pancake coil probe, were

removed from service upon detection. During the Oconee 3 outage in November 1996, three

tubes with IGA were removed for destructive examination. The laboratory destructive

examination findings were not available at the time this notice was prepared.

IN 97-49 July 10, 1997 The root cause analysis from the Oconee I tube pull analysis did not identify any unique

feature to this degradation mechanism that would indicate that the problem was limited to the

Oconee Units. That is, the base material properties met specific values, no high residual

stresses were measured, and no detrimental environmental or chemical species were

identified. These results indicate that all B&W OTSGs are potentially susceptible to this

mechanism. In September/October 1996, the licensee for ANO-1 detected freespan axial

indications similar to those observed at the three Oconee units. Approximately 13 tubes with

freespan axial indications were identified and plugged at ANO-1 during this outage. These

indications were initially detected with a bobbin coil probe. In-situ pressure testing of two of

the more severe indications (as identified by nondestructive examination) indicated burst

pressures for these freespan axial indications in excess of 4550 and 5750 psi. No leakage

was observed from either of these two indications during the in-situ test. Ea:ch Vf the Oconee

units and ANO-1 inspected 100 percent of the inservice tubes with a bobbin coil during their

last inspection outage.

Degradation in the Sludge Pile Region

At ANO-1 (in the fall of 1996), nine axially oriented indications were observed above the

lower tubesheet. These indications were in the sludge pile region, approximately 0.25-inch

above the lower secondary face of the tubesheet The indications were found with a bobbin

coil and were confirmed with a motorized rotating pancake coil inspection. Two tubes were

removed for laboratory examination. The laboratory destructive examination pointed to these

indications being initiated from the outside diameter of the tube and were a result of axially

oriented IGSCC. The licensee also observed areas of shallow intergranular corrosion

initiating from the outside diameter of the tube near the fracture faces. This corrosion was

three-dimensional in nature, similar to IGA regions; however, many of the affected grains

were no longer present, resulting in the removal of tube material and appearance of shallow

wastage. These "IGA wastage" regions were relatively shallow (less than 24-percent

through-wall) and in the form of meandering grooves or gullies. The metallurgical results

suggested to the licensee that the axial cracks originated at the bottom of these IGA wastage

zones. On the basis of nondestructive examination, these meandering grooves appeared to

be located at or near sharp edges of surface deposits.

Degradation at Sleeved Locations

B&W mechanical sleeves have been installed in all operating B&W OTSG plants in order to

mitigate tube leaks caused by high-cycle fatigue and to repair tubes with other indications of

degradation. The number of sleeves in service at these plants varies from a few hundred to

approximately one thousand. These sleeves, fabricated from either alloy 600 or alloy 690,

have three roller-expanded joints to seal them into the parent tube (one at the top of the

sleeve and two at the bottom of the sleeve). These joints have not undergone any type of

process to relieve stress.

IN 97-49 July 10, 1997 Axial, circumferential, and volumetric indications were detected in the joints of B&W

mechanical sleeves at ANO-1 (in 1996) although no tubes were removed to learn the nature

of the degradation. The indications were found at the joints of both alloy 600 and alloy 690

sleeves with a plus-point coil. The licensee believes that 9 of the 10 indications detected are

associated with the parent tube rather than with the sleeve itself. The degradation has been

observed at both the upper joint (located within the UTS) and the lower joints (in the tube

freespan region); 8 of the 10 indications were observed at the upper joint. One

circumferential indication was detected at an alloy 600 sleeve joint at Oconee Unit 3 (in 1996)

with a plus-point coil. This indication was associated with the upper of the two lower joints

(i.e., the upper lower joint). The licensee for Oconee Unit 3 believes that the indication could

be the result of a scratch made during the rolling process; however, this cannot be confirmed, since current technology does not permit the sleeve to be removed from the steam generator

for destructive examination because of its location.

Discussion

The inspection findings from B&W OTSGs indicate that a number of locations are susceptible

to degradation. In addition, studies of removed tubes have confirmed in several instances

that the eddy current indications are attributable to such degradation mechanisms as IGA and

stress corrosion cracking. Frequently these indications can only be reliably detected with

specialized probes such as rotating probes (e.g., roll transition indications, indications in

sleeve joints). In addition, the depth of many of these types of indications cannot be reliably

determined.

To effectively manage the degradation mechanisms being observed, a variety of actions have

been taken by licensees. These actions include inspecting locations potentially susceptible to

degradation with techniques capable of reliably detecting these forms of degradation (or

using the best available technique) and ensuring that the frequency and scope of inspection

are sufficient at identifying and removing degradation from service to prevent the degradation

from progressing to the point at which tube integrity is impaired. For example, the sleeve

joints at ANO-1 and Oconee 3 were examined with a plus-point coil, and 100 percent of the

tubes were examined with a bobbin coil at ANO-1 and Oconee Units 1, 2, and 3 during their

last outage. Other actions taken by licensees include removing tubes from service based

upon detection when the degradation cannot be reliably depth-sized (unless an alternative

tube repair criterion has been approved by the NRC), and assessing significant indications in

steam generator tubes to determine whether adequate structural and leakage integrity was

maintained during the previous cycle. Specific actions taken by licensees to assess the

structural and leakage integrity of tubes include removing tubes for destructive examination

as was done at ANO-1, Davis-Besse, and Oconee 1, 2, and 3, and performing in situ

pressure testing.

IN 97-49 July 10, 1997 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.

Marylee M. Slosson, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Kenneth J. Karwoski, NRR

301-415-2754 E-mail: kjkl@nrc.gov

Eric J. Benner, NRR

301-415-1171 E-mail: ejbl nrc.gov

Attachment: List of Recently Issued NRC Information Notices

Attachment

IN 97-49 July 10, 1997 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

97-48 Inadequate or Inappro- 07/09/97 All holders of OLs or CPs

priate Interim Fire for nuclear power reactors

Protection Compensatory

Measures

97-47 Inadequate Puncture 06/27/97 All "users and fabricators'

Tests for Type B of type B transportation

Packages Under 10 CFR packages [as defined in

71.73(c)(3) 10 CFR 171.16(10)(B)]

97-46 Unisolable Crack in 07/09/97 All holders of OLs or CPs

High-Pressure for nuclear power reactors

Injection Piping

96-44, Failure of Reactor 07/02/97 All holders of OL permits

Supp. 1 Trip Breaker from for nuclear power reactors

Cracking of Phenolic

Material in Secondary

Contact Assembly

97-45 Environmental 07/02/97 All holders of OLs or CPs

Qualification for nuclear power reactors

Deficiency for

Cables and Contain- ment Penetration

Pigtails

97-44 Failures of Gamma 07/01/97 All holders of OLs or CPs

Metrics Wide-Range for test and research

Linear Neutron Flux reactors

Channels

97-43 License Condition 07/01/97 All holders of OLs or CPs

Compliance for nuclear power reactors

OL = Operating License

CP = Construction Permit

IN 97-49 July 10, 1997 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.

original signed by S.H. Weiss for

Marylee M. Slosson, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Kenneth J. Karwoski, NRR

301-415-2754 E-mail: klki@nrc.gov

Eric J. Benner, NRR

301-415-1171 E-mail: ejbl nrc.gov

Attachment: List of Recently Issued NRC Information Notices

A *"MJV'kPF la agkea

DOCUMENT NAME: 97-49.IN

Tech Editor reviewed and concurred on 5/19/97 *See previous

To receive a copy of this document, Indicate In the box: "C" = Copy without

attachment/enclosure "E" = Copy with attachmentlenclosure "N" = No copy

gOFFICE Contacts C/PECB:DRPM

CDDE DIDRPM

NAME KKarwoski* BSheron* AChaffee* MSIlsson

EBenner* _ _ _ __l

DATE 05/27/97 06/19/97 06/23/97 07/2J97

05/22/97 OFFICIAL RECORD COPY

_

K>

IN 97- July , 1997 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.

Marylee M. Slosson, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Kenneth J. Karwoski, NRR

301-415-2754 E-mail: kjkl@nrc.gov

Eric J. Benner, NRR

301-415-1171 E-mail: ejbl nrc.gov

Attachment: Ust of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\EJB1\B&WSG.IN

Tech Editor reviewed and concurred on 5/19/97 *See previous

To receive a copy of this document, Indicate In the box: "C" = Copy without

attachment/enclosure "E" = Copy with attachmentlenclosure "N" = No copy

OFFICE I D/DE C/PECB:DRPM D/DRPM

NAME KKarwoski* BSheron* AChaffee* MSlosson

EBenner*

DATE 05/27/97 06/19/97 06/23/97 07/ /97

05/22/97 UNFICIAL RECORD cqJY

1///1 7

, IN 97-xx

June xx, 1997 specialized probes such as rotating probes (e.g., roll transition indications, indications in

sleeve joints). In addition, the depth of many of these types of indications cannot be reliably

determined.

To effectively manage the degradation mechanisms being observed, a variety of actions have

been taken by licensees. These actions include inspecting locations potentially susceptible to

degradation with techniques capable of reliably detecting these forms of degradation (or

using the best available technique) and ensuring that the frequency and scope of inspection

are sufficient at identifying and removing degradation from service to prevent the degradation

from progressing to the point at which tube integrity is impaired. For example, the sleeve

joints at ANO-1 and Oconee 3 were examined with a plus-point coil, and 100% of the tubes

were examined with a bobbin coil at ANO-1 and Oconee Units 1, 2, and 3 during their last

outage. Other actions taken by licensees include removing tubes from service based upon

detection when the degradation cannot be reliably depth-sized (unless an alternative tube

repair criterion has been approved by the NRC), and assessing significant indications in

steam generator tubes to determine whether adequate structural and leakage integrity was

maintained during the previous cycle. Specific actions taken by licensees to assess the

structural and leakage integrity of tubes include removing tubes for destructive examination

as was done at ANO-1, Davis-Besse, and Oconee 1, 2, and 3, and performing in situ

pressure testing.

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.

Marylee M. Slosson, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Kenneth J. Karwoski, NRR Eric J. Benner, NRR

(301) 415-2754 (301) 415-1171 E-mail: kjklnrc.gov E-mail: ejblnrc.gov

Attachment:

List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\EJB1\B&WSG.IN

Tech Editor reviewed and concurred on 5/19/97 *See previous

To receive a copy of this document, indicate in the box: "C" = Copy without

attachment/enclosure "E" =,lCr with attachment/enclosure "N" = No copy

OFFICE Contacs E C/PECB:DRPM

NAME KKarwoski* BSV o AChaffee IMSlosson

EBenner* __ __ _ __ __C) _ _ _ _ _

DATE 05/27/97 (0 l/497 /97 I /97

05/22/97 __lv ___. l

OFFICIAL R6CRD~Y

v,- KJ

IN 97-xx