NRC Bulletin 79-13, Cracking in Feedwater System Piping

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Cracking in Feedwater System Piping

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Bulletin 79-13: Revision 2, Cracking in Feedwater System Piping

SSINS: 6830

Accession No.:

7908220135

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, D.C. 20555

October 16, 1979

IE Bulletin No. 79-13

Revision 2

CRACKING IN FEEDWATER SYSTEM PIPING

Description of Circumstances:

This revision to IE Bulletin No. 79-13 is based on the results of the radio-

graphic examinations and ongoing investigation of the subject problem to

date since the initial Bulletin was issued. The revision reduces in scope

the number and extent of the piping system welds required to be examined.

The requirements for reporting and action time frame remain unchanged.

On May 20, 1979, indiana and Michigan Power Company notified the NRC of

cracking in two feedwater lines at their D. C. Cook Unit 2 facility. The

cracking was discovered following a shutdown on May 19 to investigate

leakage inside containment. Leaking circumferential cracks were identified

in the 16-inch feedwater elbows adjacent to two steam generator nozzle elbow

welds. Subsequent radiographic examination revealed crack indications in all

eight steam generator feedwater lines at this location on both Units 1 and

2.

On May 25, 1979, a letter was sent to all PWR licensees by the Office of

Nuclear Reactor Regulation which informed licensees of the D. C. Cook

failures and requested specific information on feedwater system design,

fabrication, inspection and operating histories. To further explore the

generic nature of the cracking problem, the Office of Inspection and

Enforcement requested licensees of PWR plants in current outages to

immediately conduct volumetric examination of certain feedwater piping

welds.

As a result of these actions, several other licensees with Westinghouse

steam generators reported crack indications. Southern California Edison

reported on June 5, 1979, that radiographic examination revealed indications

of cracking in feedwater nozzle-to-pipe welds on two of three steam

generators of San Onofre Unit 1. On June 15, 1979, Carolina Power and Light

reported that radiography showed crack indications in similar locations at

their H. B. Robinson Unit 2. Duquesne Power and Light confirmed on June 18,

1979, that radiography has shown cracking in their Beaver Valley Unit 1

feedwater piping-to-vessel nozzle weld. Public Service Electric and Gas

Company reported on June 20, 1979 that Salem Unit 1 also has crack

indications. Wisconsin Public Service company decided on June 20, 1979 to

cut out a feedwater nozzle-to-pipe weld which contained questionable

indication, for metallurgical examination. As of June 22, 1979 and since May

25, 1979 seven other PWR facilities have inspected the feedwater

nozzle-to-pipe welds without finding cracking indications.

NOTE: R1 and R2 indicates lines revised or added.

The feedwater nozzle-to-pipe configurations for D. C. Cook and for San

Onofre are shown on the attached figures l and 2. A typical feedwater

nozzle-to-pipe weld joint detail showing the principal crack locations for

D. C. Cook and San Onofre are shown on the attached figure 3.

On March 17, 1977, during heat-up for hot functional testing of Diablo

Canyon Unit 1, a leak was discovered in the vessel nozzle-to-pipe butt weld

joining the 16-inch diameter feedwater piping to steam generator 1-2.

Subsequent nondestructive examination of all nozzle welds by radiography and

ultrasonics revealed an approximate 6-inch circumferential crack originating

in the weld root heat-affected zone of the leaking nozzle weld. The cause of

this crack-ing was identified as either corrosion fatigue or thermal fatigue

initiating at small cracks probably induced by the welding and postweld heat

treatment cycles. The system was repaired by replacing with a piping

component employing greater controls on the welding including maintaining

preheat temperature until postweld heat treatment.

The potential safety consequences of the cracking is an increased likelihood

of a feedwater line break in the event of a seismic event or water hammer. A

feedwater line break results in a loss of one of the mechanisms of heat

removal from the reactor core and would result in release of stored energy

from the steam generator into containment. Although a feedwater line-break

is an analyzed accident, the identified degradation of these joints in the

absence of a routine inservice inspection requirement of these feedwater

nozzle-to-pipe welds formed the basis of this-Bulletin.

To date the radiographic examinations, supplemented by ultrasonic methods,

have identified cracking in the steam generator nozzle to feedwater piping

weldments at the following W, and C. E. plants.

D. C. Cook Units 1 & 2 Salem Unit 1

Diablo Canyon Surry Unit 1

San Onofre Unit 1 R. E. Ginna

H. B. Robinson Unit 2 Millstone Unit 2

Beaver Valley Unit 1 Palisades **

Kewaunee Yankee Rowe **

Point Beach Unit 2 Maine Yankee

  • Found during hot functional testing
    • Confirmatory evaluation incomplete

An extensive metallurgical investigation has been conducted by Westinghouse

on a substantial number of cracked weldments removed from the above plants.

Results of the metallurgical analysis lead to the conclusion that a

corrosion fatigue phenomenon is the probable failure-mechanism, except for

the San Onofre, piping which has been characterized as stress assisted

corrosion.

In parallel with the above ongoing analysis, the feedwater piping at D. C.

Cook, H. B. Robinson, R. E. Ginna, Salem and other plants have been

instrumented (Thermocouples, accelerometers, strain gages, and transducers)

to collect data on the potential forcing functions contributing to cracking under steady

state and transient conditions. Preliminary unchecked results of temperature

data has identified cyclic thermal gradients may exist due to stratified

feedwater temperature conditions in the feedpipe weld region during zero and

low power operations. This gradient tends to support the fatigue aspect of

the postulated failure mechanism. No further unexpected operation loading or

forcing functions have been identified by other instrumentation.

In regard to B&W plants a total of 95 welds in the main and separate

auxiliary feedwater piping, risers and, steam generator nozzles regions have

been examined at Crystal River Unit 3 and Davis Besse. No indications of a

cracking problem was found.

In view of the findings to date, the revised inspections outlined below is

considered acceptable to meet this intent of IE Bulletin No. 79-13.

Actions to be Taken by Licensees

For all pressurized water reactor facilities with an operating license:

1. Facilities which have steam generators fabricated by Westinghouse or Combustion Engineering that have not conducted volumetric examination of feedwater nozzles since May 1979 shall complete the inspection program described below at the earliest practical time but no later than 90 days after the date of Bulletin No. 79-13.

a. Perform radiographic examination, supplemented by ultrasonic

examination as necessary to evaluate indications, of all feedwater

nozzle-to-pipe welds and of adjacent pipe and nozzle areas (a

distance equal to at least two wall thicknesses). Evaluation shall

be in accordance with ASME Section III, Subsection NC, Article

NC-5000. Radiography shall be performed to the 2T penetrameter

sensitivity level, in lieu of Table NC-5111-1, with systems void

of water.

b. In the event cracking is identified during examination of the

nozzle-to-pipe weld, all feedwater line welds up to the first

piping support or snubber outboard of the nozzle shall be

volumetrically examined in accordance with 1.a above. All

unacceptable code discontinuities shall be subject to repair

unless justification for continued operation is provided.

c. Perform a visual inspection of feedwater system piping supports

and snubbers in containment to verify operability and conformance

to design.

2. All pressurized water reactor facilities shall perform the inspection

program described below at the next outage of sufficient duration or at

the next refueling outage after the inspection required by item 1.

a. For steam generator designs with a common nozzle for both main and

auxiliary feedwater systems, perform volumetric examination of the

feedwater nozzle-to-pipe welds, the feedwater piping welds to the

first support, and the feedwater line-to-containment penetration

welds in accordance with Item 1 above. In addition, examine an

area of at least one pipe diameter of the main feedwater line

downstream at the auxiliary feedwater to main feedwater

connection.

b. For steam generator designs utilizing auxiliary feedwater systems

connected by means of welded nozzle connections, perform

volumetric examination of all auxiliary feedwater nozzle to piping

welds and the first adjacent outboard pipe-to-pipe welds (risers)

in accordance with item 1 above.

For designs utilizing auxiliary feedwater systems connected to the

steam generator by means of bolted flange connections, perform

volumetric examination of the flanged nozzle to piping and first

outboard pipe-to-pipe welds (risers) in accordance with item 1

above.

The examinations specified in 2.b above are not required provided

that during startup, hot standby or cold shutdown operations, the

feedwater level within the steam generator is maintained

essentially constant and no intermittent cold auxiliary feedwater

injection is utilized; i.e., auxiliary feedwater injection where

used, is preheated during the forementioned operating modes.

c. Perform a visual inspection of all feedwater system piping supports and snubbers in containment to verify operability and conformance to design.

3. Identification of cracking indications in feedwater nozzle or piping weld areas in one unit of a multi-unit facility shall require shutdown and inspection of other similar units which have not been inspected since May 1979, unless justification for continued operation is provided.

4. Any cracking or other unacceptable code discontinuities identified shall be reported to the Director of the appropriate NRC Regional Office within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of identification.

5. Provide a written report to the Director of the appropriate NRC Regional Office within 20 days of the date of the original Bulletin (June 25, 1979) addressing the following:

a. Your schedule for inspection if required by item 1.

b. The adequacy of your operating and emergency procedures to

recognize and respond to a feedwater line break accident.

c. The methods and sensitivity of detection of feedwater leaks in

containment.

6. A written report of the results of examination, in accordance with

requests by Regional Offices preceding this Bulletin and with Bulletin

item 1 and 2 including any corrective measures taken, shall be

submitted within 30 days of the date of the original Bulletin No. 79-13

(June 25, 1979) or within 30 days of completion of the examination,

whichever is later, to the Director of the appropriate NRC Regional

Office with a copy to the NRC Office of Inspection and Enforcement,

Division of Reactor Operations Inspection, Washington, D. C. 20555.

Actions to be Taken by Designated Applicants for Operating Licenses:

1. On completion of the hot functional testing program and prior to fuel loading, perform the inspections described in item 1 above.
2. During the first refueling outage, perform the inspections described in item 2 above.
3. Submit reports as described in Items 4, 5. and 6 above based on the date of Revision 1 to Bulletin No. 79-13 (August 30. 1979)

Approved by GAO, B180225 (R0072), clearance expires 7/31/80. Approval was

given under a blanket clearance specifically for identified generic

problems.

Attachments:

Figures 1, 2, and 3

.

DESIGNATED APPLICANTS FOR OPERATING

LICENSES

Salem 2

North Anna 2

Diablo Canyon 1 & 2

Sequoyah 1

McGuire 1

San Onofre 2

Summer

Watts Bar 1 & 2

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