Information Notice 1996-09, Damage in Foreign Steam Generator Internals

Damage in Foreign Steam Generator Internals

ML031210490
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:	02/12/1996
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:
References
IN-96-009, NUDOCS 9602060170
Download: ML031210490 (8)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555 February 12, 1996 NRC INFORMATION NOTICE 96-09: DAMAGE IN FOREIGN STEAM GENERATOR INTERNALS

Addressees

All holders of operating licenses or construction permits for pressurized

water reactors (PWRs).

Purpose

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

notice to alert addressees to recent findings of damage to steam generator

internals, namely support plates and wrapper, at foreign PWR facilities. 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

In April 1995 during a routine eddy current inspection of the steam generator

tubing at a foreign facility, anomalous support plate signals were observed at

the uppermost support plate. The steam generators are similar but not

identical to Westinghouse model 51 steam generators. The support plates are

of the drilled hole type and fabricated from carbon steel. Video camera

inspections were conducted to investigate the anomalous signals and revealed

that a significant portion of the support plate had wasted away. Pieces of

the affected region of the support plate were found resting on the next lower

support plate.

Subsequent investigation has identified chemical cleaning performed in 1992 as

the cause of the support plate damage. Review of previous eddy current data

shows that the anomalous support plate signals were present in inspections

dating back to 1993 when the first inservice inspection following chemical

cleaning was performed. Support plate signals obtained immediately prior to

the chemical cleaning were normal. The foreign regulatory authority believes

that pipes used to direct the chemical solution into the steam generators were

installed incorrectly, too close to the upper support plate. This caused an

excessively high impingement velocity of the cleaning solution against the

support plate which is believed to have been sufficient to render ineffective

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IN 96-09 February 12, 1996 the corrosion inhibitor in the cleaning solution. U.S. industry representa- tives stated during recent meetings with the NRC staff that chemical cleanings

which have been performed in the U.S. involve different cleaning agents and

inhibitors than that used at the foreign facility and involve less risk for

producing similar damage.

The support plate damage at the foreign facility effectively eliminated

lateral support to tubes within the affected region. Lateral support provides

vibrational stability and the ability to sustain earthquake and loss-of- coolant-accident loadings. Accordingly, all tubes found not to be supported

at the uppermost support plate were plugged.

Based on this experience, the foreign utility carefully examined the support

plate eddy current signals at other PWR facilities. At one of these units, with steam generators similar but not identical in design to Westinghouse

model 51 steam generators, eddy current signals indicative of support plate

ligament cracks were found at the uppermost support plate. The support plates

are of the drilled hole type and are fabricated from carbon steel. Subsequent

visual inspection confirmed the presence of ligament cracks near the periphery

of the support plate. Part of the support plate periphery was observed to be

entirely broken away in the vicinity of a radial seismic support. The steam

generators at this facility have not been chemically cleaned. Review of past

eddy current results indicates that the indications of ligament cracks date

back at least 9 years. It is not clear whether the ligament cracks were

present prior to initial service or whether the cracks may have developed

shortly thereafter. The cause of these cracks is under investigation by the

foreign utility and steam generator manufacturer. Tubes whose lateral support

was potentially affected by these cracks have been plugged. Press reports

indicate that similar indications of support plate ligament cracks have

recently been found at other facilities in the same country with similar steam

generators.

Visual inspections conducted in June 1994 at a foreign PWR facility revealed

the bottom of the wrapper had dropped down by 20 millimeters in one steam

generator and by 5 millimeters in another steam generator. The steam

generators are similar but not identical to Westinghouse model 51 steam

generators. The visual inspections were performed through handholes located

above the tubesheet. Further investigation revealed that wrapper welds at

each of six vertical supports in the first steam generator and at three of six

vertical supports in the second steam generator had failed, allowing the

downward displacement of the wrapper. The cause of this occurrence is under

investigation by the foreign utility and the steam generator manufacturer.

Their preliminary assessment is that unanticipated axial restraint against

differential thermal expansion between the wrapper and steam generator

pressure vessel shell led to significant loading of the wrapper vertical

IN 96-09 February 12, 1996 supports. This unanticipated restraint between the wrapper and shell may have

been due to differential thermal expansion between support plate number 7 and

the shell, preventing relative axial motion between the wrapper and shell at

this elevation, during transients involving the auxiliary feedwater. Poor

quality of the wrapper welds at the vertical support may also have been a

contributing factor.

Implications of a complete fall of the wrapper have been assessed by the

foreign utility to include the potential for loss of feedwater, damage to the

largest radius tube u-bends, loose parts, and tube rupture. Accordingly, the

foreign utility has implemented temporary repairs to stabilize and monitor the

wrappers pending further investigation regarding long-term resolution of this

matter.

Discussion

As illustrated by the foreign experience, support plate signal anomalies

during eddy current testing of the steam generator tubes may be indicative of

support plate damage or ligament cracking. The signal anomalies at the

foreign units were present for several years before they were first identified

by the data analysts. The Electric Power Research Institute (EPRI) has

initiated an effort, in response to the foreign experience, to develop a

qualified procedure for detecting support plate ligament cracks.

The steam generator tube support plates function to support the tubes against

lateral displacement and vibration and to minimize bending moments in the

tubes during accidents. Damage and/or cracking of the support plates can

impair the ability of the support plates to perform this function and, thus, may potentially impair tube integrity. In addition, the staff has recently

approved a 3 volt plugging criteria for two U.S. plants based, in part, upon

evidence from inspections using EPRI preliminary procedures that the tube

support plates are capable of locally constraining the tubes against tube

rupture.

Known instances of support plate cracking/damage in the U.S. have generally

involved support plates with significant denting. The potential for support

plate cracks has tended not to be of significant concern in recent years since

the steam generators most affected by denting have been replaced and, in

addition, the industry has been successful in controlling denting progression

in operating steam generators. The foreign experience serves to highlight

that there are other mechanisms which may lead to support plate damage and/or

cracking.

Based on the information available to the NRC staff, it is not yet known

whether steam generators in the U.S. are vulnerable to the type of wrapper

damage observed at the foreign unit.

IN 96-09 February 12, 1996 The staff will continue to monitor information on support plate and wrapper

damage as it becomes available from foreign authorities.

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.

Dennis M. Crutc e rector

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Emmett L. Murphy, NRR

(301) 415-2710

Internet:elm@nrc.gov

Eric J. Benner, NRR

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

Attachment: List of Recently Issued NRC Information Notices

pev I's G--b r I&S. Attachment

IN 96-09 q5 -03 ale,0 I Mof; r.so .,d February 12, 1996 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

96-08 Thermally Induced Pres- 02/05/96 All holders of OLs or CPs

sure Locking of a High for nuclear power reactors

Pressure Coolant Injec- tion Gate Valve

96-07 Slow Five Percent Scram 01/26/96 All holders of OLs or CPs

Insertion Times Caused for boiling water reactors

By Viton Diaphragms in

Scram Solenoid Pilot

Valves

96-06 Design and Testing 01/25/96 All holders of OLs or CPs

Deficiencies of Tornado for nuclear power reactors

Dampers at Nuclear Power

Plants

96-05 Partial Bypass of Shutdown 01/18/96 All holders of OLs or CPs

Cooling Flow from the for boiling water reactors

Reactor Vessel

96-04 Incident Reporting Require- 01/10/96 All radiography licensees

ments for Radiography and manufacturers of radio- Licensees graphy equipment

96-03 Main Steam Safety Valve 01/05/96 All holders of OLs or CPs

Setpoint Variation as a for nuclear power reactors

Result of Thermal Effects

OL - Operating License

CP - Construction Permit

IN 96-09 February 12, 1996 The staff will continue to monitor information on support plate and wrapper

damage as it becomes available from foreign authorities.

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

Dennis M. Crutchfield, Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Emmett L. Murphy, NRR

(301) 415-2710

Internet:elm@nrc.gov

Eric J. Benner, NRR

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

Attachment: List of Recently Issued NRC Information Notices

ft 6 I C Hke

-tffeV6",t

Reviewed by Tech Editor prior to receipt by PECB

Coordination of foreign information was handled by MCullingford, NRR

IN reviewed and concurred on by foreign regulatory agency prior to

receipt by PECB

DOCUMENT NAME: 96-09.IN *SEE PREVIOUS CONCURRENCES

To. -1-. -.

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ACE - Cnnv without enclosures WE Coov with enclosures ENS - No copy

OFFICE

CONTACT

S I C:EMCB I C:PECB I D:DRRMM

NAME EJBenner* JStrosnider* AEChaffee* OM hfie

ELMurphy* ____

DATE

_01/11/96

01/04/96 01/19/96 01/24/96 OFFICIAL RECORD COPY

02/7 /96

IN 96-xx

January xx, 1996 Discussion As illustrated by the foreign experience, support plate signal anomalies

during eddy current testing of the steam generator tubes may be indicative of

support plate damage or ligament cracking. The signal anomalies at the

foreign units were present for several years before they were first identified

by the data analysts. In addition, EPRI has initiated an effort, in response

to the foreign experience, to develop a qualified procedure for detecting

support plate ligament cracks.

The steam generator tube support plates function to support the tubes against

lateral displacement and vibration and to minimize bending moments in the

tubes during accidents. Damage and/or cracking of the support plates can

impair the ability of the support plates to perform this function and, thus, may potentially impair tube integrity. In addition, the staff has recently

approved a 3 volt plugging criteria for two U.S. plants based, in part, upon

evidence from inspections using EPRI's preliminary procedures that the tube

support plates are capable of locally constraining the tubes against tube

rupture.

Known instances of support plate cracking/damage in the U.S. have generally

involved support plates with significant denting. The potential for support

plate cracks has tended not to be of significant concern in recent years since

the steam generators most affected by denting have been replaced and, in

addition, the industry has been successful in controlling denting progression

in operating steam generators. The foreign experience serves to highlight

that there are other mechanisms which may lead to support plate damage and/or

cracking.

Based on the information available to the NRC staff, it is not yet known

whether steam generators in the U.S. are vulnerable to the type of wrapper

damage observed at the foreign unit.

The staff will continue to monitor information on support plate and wrapper

damage as it becomes available from foreign authorities.

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.

Dennis M. Crutchfield, Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Emmett L. Murphy, NRR

(301) 415-2710

Internet: ELM@NRC.GOV ForthL;

Eric J. Benner, NRR FrS>-

(301) 415-1171 Internet: E11RNRC.GOV

Attachment: List of Receri4,yIssued NRC Inf.Xormion Notices

Reviewed by Tech Editox-j1oior to receipt4b PECB

Coordination of Freplrinformation washandled by MCullingford, NRR

IN reviewed and concurred on by Frefth regulatory agency prior to

receipt by PECB

DOCUMENT NAME: G:\EJB1\FOREIGN.SG *See previous concurrence

To receive acopy of this docurnent, Indicate In the box: 'C'" Copy without enclosures Ed - Copy with enclosures N - No copy

lOFFICE Contacts l C:EMCB I PECB C:PECB l jD:DRPMj I

NAME EJBenner* JStrosnider* EFGoodwin* AEChaffee < DMCrutchfield

l.__ ELMurphy* _

DATE 1/04/96 1/11/96 01/19/96 01/19/96 01/Zq/96 Sb 01/ /96 OFFICIAL RECORD COPY /

, -1

'k

IN 96-xx

January xx, 1996 Discussion As illustrated by the foreign experience, support plate signal anomalies

during eddy current testing of the steam generator tubes may be indicative of

support plate damage or ligament cracking. The signal anomalies at the

foreign units were present for several years before they were first identified

by the data analysts. In addition, EPRI has initiated an effort, in response

to the foreign experience, to develop a qualified procedure for detecting

support plate ligament cracks.

The steam generator tube support plates function to support the tubes against

lateral displacement and vibration and to minimize bending moments in the

tubes during accidents. Damage and/or cracking of the support plates can

impair the ability of the support plates to perform this function and, thus, may potentially impair tube integrity. In addition, the staff has recently

approved a 3 volt plugging criteria for two U. S. plants based, in part, upon

evidence from inspections using EPRI's preliminary procedures that the tube

support plates are capable of locally constraining the tubes against tube

rupture.

Known instances of support plate cracking/damage in the U.S. have generally

involved support plates with significant denting. The potential for support

plate cracks has tended not to be of significant concern in recent years since

the steam generators most affected by denting have been replaced and, in

addition, the industry has been successful in controlling denting progression

in operating steam generators. The foreign experience serves to highlight

that there are other mechanisms which may lead to support plate damage and/or

cracking.

Based on the information available to the NRC staff, it is not yet known

whether steam generators in the U.S. are vulnerable to the type of wrapper

damage observed at the foreign unit.

The staff will continue to monitor information on support plate and wrapper

damage as it becomes available from foreign authorities.

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.

Dennis M. Crutchfield, Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Emmett L. Murphy, NRR

(301) 415-2710

E-mail: ELM@NRC.GOV

Eric J. Benner, NRR

(301) 415-1171 E-mail: EJBINRC.GOV

Attachment: List of Recently Issued NRC Information Notices

Reviewed by Tech Editor prior to receipt by PECB

Coordination of French information was handled by MCullingford, NRR

IN reviewed and concurred on by French regulatory agency prior to

receipt by PECB

DOCUMENT NAME: G:\EJB1\FOREIGN.SG

To receive a copy of this document Indicate In the box: 'C' - Copy without enclosures 'E' = Copy with enclosures 'No = No copy

OFFICE

NAME

ntacts

EJBenner

C: EMC

r [s 7 I PECB

EFGoodwin

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AEChaffee

Il D:DRPM Il

DMCrutchfield

_DATE_/_ /96 1/jt/96 0 9i/

6 __1/1lY96 01/ /96 01/ /96 OFFICIAL RECORD COPY