Information Notice 1993-20, Thermal Fatigue Cracking of Feedwater Piping to Steam Generators

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Thermal Fatigue Cracking of Feedwater Piping to Steam Generators
ML031080045
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  Entergy icon.png
Issue date: 03/24/1993
From: Grimes B
Office of Nuclear Reactor Regulation
To:
References
BL-79-013 IN-93-020, NUDOCS 9303180065
Download: ML031080045 (11)
v • d • e


UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555 March 24, 1993 NRC INFORMATION NOTICE 93-20: THERMAL FATIGUE CRACKING OF FEEDWATER

PIPING TO STEAM GENERATORS

Addressees

pressurized

All holders of operating licenses or construction permits for Engineering.

water reactors (PWRs) supplied by Westinghouse or Combustion

Purpose

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

notice to inform addressees of cracks found in the feedwater pipingtheto steam

generators at the Sequoyah Nuclear Power Plant, Units 1 and 2 and to review

Diablo Canyon Nuclear Power Plant, Unit 1. Recipients are expected actions, as

the information for applicability to their facilities and consider in

appropriate, to avoid similar problems. However, suggestions contained

this information notice are not NRC requirements; therefore, no specific

action or written-response is required.

Description of Circumstances

Unit 1. In

In 1992, cracks were found in the feedwater lines at Diablo Canyon feedwater

addition, a through-wall crack and other cracks were found in the

lines at Sequoyah. The cracks were attributed to thermal fatigue.

The NRC staff first learned of cracks in feedwater lines to steam generators

which resulted from thermal fatigue in 1979 when the Indiana and Michigan

Electric Company, the licensee for the Donald C. Cook Plant, reported leaks.

In dealing with this problem, the NRC staff issued the following documents:

  • a letter to PWR licensees pursuant to Paragraph 50.54(f) of Title 10 of

the Code of Federal Regulations, May 25, 1979

Feedwater System Piping,' June 25, 1979, Revision 1, August 30, 1979, Revision 2, October 17, 1979

ultrasonic examinations of feedwater lines. As a result of these

examinations, cracks were found at 18 of the 54 facilities inspected. The

staff closed the bulletin on the basis of the results of the one-time

inspection. The industry had taken the actions recommended in the bulletin

9303180065 PO I to f *tt 13MO&I 93O's

'l

IN 93-20 .-

March 24, 1993 and Instituted the augmented inservice inspection programs, which appeared to

provide for reliable detection of cracks in feedwater piping.

Other technical evaluations were documented in the following reports:

  • Investigation of Feedwater Line Cracking in PWRs" (Westinghouse, 1980)
  • NUREG-0691, "Investigation and Evaluation of Cracking Incidents in Piping

In PWRs," (PWR pipe crack study group, 1980)

These studies showed that thermal fatigue was the main cause of the cracks.

Modifications to minimize the effects of thermal stratification In feedwater

lines and augmented licensee inservice inspections were recommended in the

studies.

Recently, some licensees have again reported cracks in feedwater piping. The

licensee for the Sequoyah units reported an-actual leak despite augmented

inservice inspections. The augmented inspections using ultrasonic techniques

--showed 'indications that might earlier have revealed the cracks, but the

licensee misinterpreted these as resulting from the geometric configuration of

the pipe. -After finding the leak, the licensee performed radiography on all

feedwater nozzles of both units and found cracks in-five of the eight nozzles.

The licensee for Diablo Canyon Unit 1 reported indications with cracklike

ultrasontc-signal characteristics -infeedwater-piping to all four steam

generators. The indications varied in length up to 20 cm [7-3/4 inches] in a

circumferential direction, and many were'intermittent. Some intermittent

indications extended the full circumference with segments up to 5 cm

[2 inches] long. The licensee tried to verify the indications by radiography

but failed. Later, metallurgical analysis showed the indications to be

cracks. ' -

Only one nozzle at Diablo Canyon had been scheduled for an inspection. This

inspection was performed in accordance with Section XI of the American Society

of Mechanical Engineers Boiler and Pressure VesselTCode (ASME Code). However, information on the leak at Sequoyah led the licensee for Diablo Canyon to use

enhanced ultrasonic techniques to inspect all four lines. These techniques

were considered more appropriate for finding small cracks from thermal fatigue

than techhiques'specified by the ASME Codej which may not be adequate to

detect these types of defects. Flaw sizing by ultrasonic techniques proved to

be overly conservative at Diablo Canyon, however, presumably because

inclusions in the material led to inaccurate results: 'cracks sized at 8.9 mm

[0.35 inch] deep by ultrasonic inspection were shown by cross sectioning to be

cracks 0.94 mm (0.037 inch] deep.

Discussion

Cracks from thermal fatigue in PWR feedwater lines have proved to be a

recurring problem. The main cause of crack growth appears to be fatigue

induced by stresses from thermal stratification during cold, low-flow, feedwater injections. Other factors that contribute to crack growth are a

high oxygen content, counterbore weld preparation geometry, and thermal

IN 93-20

March 24, 1993 conditions during heatup, hot standby, and low-power operation. A favored

solution has been to replace the degraded piping. However, replacing the

degraded piping in kind without other corrective actions to eliminate or

minimize the factors which cause the cracks can leave the piping susceptible

to the same problem.

Inspection techniques specified in the ASME Code Section XI, do not appear

adequate to find cracks of this type.

This information notice requires no specific action or written response.

you have any questions about the information in this notice, please contactIf

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation (NRR) project manager.

Brian K. Grimes, Director

-Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: Lee Banic, NRR

(301) 504-2771 Robert A. Hermann, NRR

(301) 504-2768 Attachment: List of Recently Issued NRC Information Notices

I II * 0

11 I Attachment

IN 93-20

I March 24, 1993 (0 U. 0

4 Page I of I

ILU ar

(LO

LIST OF RECENTLi ISSUED c 4LL

NRC INFORMATION NOTICES

nf

0 W

Ifl U? ^ UDrto i

Notice No.

_ . . .

Subject Issuance Issued to

.

93-19 Slab Hopper Bulging 03/17/92 All nuclear fuel cycle

licensees.

93-I8 Portable Moisture-Density 03/10/93 All U.S. Nuclear Regulatory

Cauge User Responsibilittes Commission licensees that i

possess moisture-density

during Field Operations

gauges. II

93-17 Safety Systems Response 03/08/93 All holders of OLs or CPs

to Loss of Coolant and

Loss of OffsIte Power

for nuclear power reactors. Ir

93-I6 Failures of Nut-Locking 02/19/93 All holders of OLs or CPs I

Devices in Check Valves for nuclear power reactors.

93-15 Failure to Verify the 02/18/93 All holders of OLs or CPs i

Continuity of Shunt Trip for nuclear power reactors.

Attachment Contacts In

Manual Safety Injection

and Reactor Trip Switches

93-14 Clarification of 02/18/93 All licensees who possess

10 CFR 40.22, Small source material.

Quantities of Source

Material

93-13 Undetected Modification 02/16/93 All holders of OLs or CPs

of Flow Characteristics for nuclear power reactors.

in the High Pressure z

Safety Injection System 0-

Vi00

) 3-12 Off-Gassing in Auxiliary

Feedwater System Raw

02/11/93 All holders of OLs or CPs

for nuclear power reactors.

00

Water Sources

93-11 Sigle Failure Vulner- 02/04/93 All holders of OLs or CPs

ability of Engineered for nuclear power reactors.

Safety Features c~o

Actuation Systems U.I

93-10 Dose Calibrator Quality 02/02/93 All Nuclear Regulatory Com- -aU

Control mission medical licensees. 411 zU:

UJ

2 t Ui

2L . operatini L1cense

CP - Construction Permit

a-

'V < IN 93-20

March 24, 1993 conditions during heatup, hot standby, and low-power operation. A favored

solution has been to replace the degraded piping. However, replacing the

degraded piping in kind without other corrective actions to eliminate or

minimize the factors which cause the cracks can leave the piping susceptible

to the same problem.

Inspection techniques specified in the ASME Code Section XI, do not appear

adequate to find cracks of this type.

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 igned by

Brian K. Grimes

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: Lee Banic, NRR

(301) 504-2771 Robert A. Hermann, NRR

(301) 504-2768 Attachment: List of Recently Issued NRC Information Notices

KfJ A

Document Name: 93-20.IN

  • SEE PREVIOUS CONCURRENCES

D *C/OGCB:DORS:NRR*RPB:ADM

GHMarcus TechEd

03/04/93 01/14/93

  • OGCB: DORS:NRR*EMCB:DE:NRR *SC/EMCB:DE:NRR*C/EMCB:DE:NRR *D/DE:NRR

CVHodge LBanic RAHermann JRStrosnider JERichardson

01/25/93 01/26/93 01/26/93 01/27/93 02/01/93 A //C

.sI-

AK.' v>

IN 93-XX

March xx, 1993 conditions during heatup, hot standby, and low-power operation. A favored

solution has been to replace the degraded piping. However, replacing the

degraded piping in kind without other corrective actions to eliminate or

minimize the factors which cause the cracks would leave the piping susceptible

to the same problem.

Inspection techniques specified in the ASME Code Section XI, do not appear

adequate to finding cracks of this type.

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, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: Lee Banic, NRR

(301) 504-2771 Robert A. Hermann, NRR

(301) 504-2768 Attachment: List of Recently Issued NRC Information Notices

Document Name: DIANINF.IN

  • SEE PREVIOUS CONCURRENCES

(DORS:NRR *C/OGCB: DORS:NRR*RPB:ADM

BI(Grimes V,G)HMarcus TechEd

O.3/ /93 03/04/93 01/14/93

  • OGCB: DORS:NRR*EMCB:DE:NRR *Sii/EMCB:DE:NRR*C/EMCB:DE:NRR *D/DE:NRR

CVHodge LBanic RikHermann JRStrosnider JERichardson

01/25/93 01/26/93 0]1/26/93 01/27/93 02/01/93

-. I'

IN 93-XX

March xx, 1993 replace the degraded piping. However, replacing the degraded piping in kind

without other corrective actions to eliminate or minimize stratification will

leave it susceptible to cracking again.

Inspection techniques specified in the ASME Code,Section XI, appear to be

inadequate for finding the cracking.

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, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: Lee Banic, NRR

(301) 504-2771 Robert A. Hermann, NRR

(301) 504-2768 Attachment: List of Recently Issued NRC Information Notices

Document Name: DIANINF.IN

  • SEE PREVIOUS CONCURRENCES

D/DORS:NRR C/OGCB:DORS:NRR*RPB:ADM

BKGrimes GHMarcus6 a TechEd

03/ /93)'?Y 03/# /93 01/14/93

  • OGCB:DO RS:NRR*EMCB:DE:NRR *SC/EMCB:DE:NRR*C/EMCB:DE:NRR *D/DE:NRR

CVHodge LBanic RAHermann JRStrosnider JERichardson

01/25/93 01/26/93 01/26/93 01/27/93 02/01/93

r

h -' IN 93-XX

February xx, 1993 in the ASME Code,Section XI, appear to be inadequate for finding the

cracking. A favored solution has been to replace the degraded piping.

However, replacing the degraded piping in kind without other corrective

actions to eliminate or minimize stratification will leave it susceptible to

cracking again.

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, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: Lee Banic, NRR

(301) 504-2771 Robert A. Hermann, NRR

(301) 504-2768 Attachment: List of Recently Issued NRC Information Notices

Document Name: DIANINF.IN

  • SEE PREVIOUS CONCURRENCES

Do(DORS:NRR C/OGCB:DORS:NRR*RPB:ADM

BI(Grimes GHMarcus TechEd

2/ /93 02/ /93 01/14/93

  • OGCB:DORS:NRR*EMCB:DE:NRR *SI ',/EMCB:DE:NRR*C/EMCB:DE:NRR *D/DE:NRR

CVHodge LBanic PiWHermann JRStrosnider JERichardson

01/25/93 01/26/93 0:1/26/93 01/27/93 02/01/93

-r

....... IN 93-XX

January xx, 1993 in the ASME Code,Section XI, appear to be inadequate for finding the

cracking. A favored solution has been to replace the degraded piping.

However, replacing the degraded piping in kind without other corrective

actions to eliminate or minimize stratification will leave it susceptible to

cracking.

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, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: Lee Banic, NRR

(301) 504-2771 Robert A. Hermann, NRR

(301) 504-2768 Attachment: List of Recently Issued NRC Information Notices

Document Name: DIANINF.IN

  • SEE PREVIOUS CONCURRENCES

D/DORS:NRR C/OGCB:DORS:NRR RPB:ADM /

BKGrimes GHMarcus TechEd* Al

01/ /93 01/ /3k

OGCB:DORS:NRR EMCB:DE:NRR SC/EMCB:DE:NRR C/EMCB: D &R D/DE: N

CVHodge* LBanic* RAHermann* JRStrosn i JERicha son

01/25/93 01/26/93 01/26/93 01/%1/93 0)/ (/93

4/yt1,

r

<'-* IN 93-XX

January xx, 1993 Discussion

Cracking from thermal fatigue is a persistent problem. The main cause of

crack growth is fatigue induced by stresses from thermal stratification during

cold, low-flow, feedwater injections. Other factors are a high oxygen

content, counterbore weld preparation geometry, and thermal conditions during

heatup, hot standby, and low-power operation. Inspection techniques specified

in the ASME Code,Section XI, appear to be inadequate for finding the

cracking. A favored solution has been to replace the degraded piping.

However, replacing the degraded piping in kind without other corrective

actions to eliminate or minimize stratification will leave it susceptible to

cracking.

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, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: Lee Banic, NRR

(301) 504-2771 Robert A. Hermann, NRR

(301) 504-2768 Attachment: List of Recently Issued NRC Information Notices

Document Name: DIANINF.IN

  • SEE PREVIOUS CONCURRENCES

D/DORS:NRR C/OGCB:DORS:NRR*RPB:ADM

BKGrimes GHMarcus TechEd

01/ /93 01/ /93 01/14/93 OGCB:DORS; RR EMCB:DE:NRR SC/EMPB PE:NRR C/EMCB:DE:NRR D/DE:NRR

CVHodge ZtJ LBanic PV

01/h /93 RAHerin JRStrosnider

01/ /93 JERichardson

01/ /93

01 / /93 01/Ab/+

- N.. S -

- one of the technical ontacts listed below or the appro riate Office of

Nuclear Reactor Regulation (NRR) project -manager.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: Lee Banic, NRR

(301) 504-2771 Robert A. Hermann, NRR

(301) 504-2768 Attachment: List of Recently Issued NRC Information Notices

Document Name: DIANINF.IN

D/DORS:NRR C/OGCB:DORS:NRR RPB:ADM

BKGrimes GHMarcus TechEd Miah 4n

01/ /93 01/ /93 O1/I'/93 OGCB:DORS:NRR EMCB:DE:NRR SC/EMCB:DE:NRR C/EMCB:DE:NRR D/DE!: NRR

CVHodge LBanic/ RAHermann JRStrosnider JERi hardson

01/ /93 01/)L /93 01/ /93 01/ /93 01/ /93

-, . .


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