ML20107E883

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Responds to Concerns Noted from Insp Rept 50-219/74-08 on 740506-07.Corrective Actions:Util Calibr Std Located & UT Exam of Reactor Vessel Skirt Redone After Recalibr Equipment to Util Std
ML20107E883
Person / Time
Site: Oyster Creek
Issue date: 06/12/1974
From: Ross D
JERSEY CENTRAL POWER & LIGHT CO.
To: Robert Carlson
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML18039A986 List: ... further results
References
FOIA-95-258 NUDOCS 9604220126
Download: ML20107E883 (3)
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Jersey centrai Power & Light Company MADISON AVENUE AT PUNCH BOWL ROAD e MORRISTOWN, N.J. 07960

  • 201-5394111 asum av run oeneral Public Utilities Corporation am-June 12, 1974 Mr. Robert T. Carlson, Chief Facility Construction and Engineering, Support Branch United States Atomic Energy Commission Directorate of Regulatory Operations, Region 1 631 Park Avenue King of Prussia, Pennsylvania 19406

Dear Mr. Carlson:

Subject:

Oyster Creek Station Docket No. 50-219 RO Inspection Report No. 50-219/74-8 This is in reply to your letter of May 16, 1974 to Mr. I. R.

Finfrock, Jr. regarding the inspection conducted by Mr. Walton on May 6-7, 1974 at our Oyster Creek Nuclear Generating Station.

1.

AEC Concern Criterion IX, Appendix B, 10 CFR 50 states, in part,

" Measures shall be established to assure that special processes, including... nondestructive testing, are controlled and accomplished...using qualified pro-cedures..."

Contrary to the above, nondestructive test procedures were approved by the licensee for use during in-service inspection with words in the procedure which state, "This procedure is not applicable and shall not be used for in-service inspection."

JCPSL Reply The referenced procedure is a Magnaflux Testing Laboratory j

(Magnaflux) procedure and is not normally utilized for in-service inspection.

The procedure is written to satisfy the requirements of Sections III, V, and VIII of the ASME Boiler and Pressure Vessel Code which are the sections invoked for in-service inspection at Oyster Creek.

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June 32, 1974

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Mr. Robert C. Carlson R0' Inspection Report No. 50-219/74-8

-Page II Prior to its use, Magnaflux Procedure 13N, as well es all other applicable procedures, was reylewed by an independent JCP6L ASNT-TC-1A Certified Level III Representative and was found to be acceptable for in-service inspection at Oyster Creek.

During the above review, JCP6L recognized the re-

- ferenced statement, but it was determined that the procedure was adequate. The fact that the statement was not deleted from the procedure is an administrative oversight and JCPGL

-has since received Amendment No. 1 to Procedure 13N from Magnaflux which deletes Paragraph 13.2.2 Note 1.

2.

AEC Concern Amendment 68 to " Application for Reactor Construction Permit and Operating License", Paragraph 4.2.4, Criterion 32, states, in part, "... All piping, pumps, and valves defined by ASME Boiler and Pressure Vessel Code Section XI January 1, 1970 issue, to be part of the primary coolant pressure boundary and all components of the reactor vessel...are accessible for inspection d*2 ring refueling."

... Ultrasonic ASME'Section XI IS 2.3.2 states, in part, examination, using the pulse-echo method, shall be con-ducted in accordance with the provisions of Appendix IX, IX-300 and referenced in IX 340..."

Section III, Paragraph IX-343(b) of the ASME Code states, in part, "... Drilled holes shall be used as basic calibra-tion reflectors...these holes shall be located either in the production material or in a basic calibration block..."

Paragraph IX-343(c) of tie ASME Code states,

"...In lieu of the above, other calibration reflectors are permi-d,

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provided equivalent response is demonstrated..."

l Contrary to the above, a calibration block was used which was not fabricated in accordance with the above requirements.

JCPSL Reply t

i The significance of your reference to Amendment 68 in this case is unclear. JCP6L does use the pulse-echo method for in-service inspection and does have and use calibration

- blocks for equipment calibration.

The calibration block, which is used for the reactor vessel skirt weld inspection, was not available initially so Magnaflux provided a suitable block for calibrating their equipment.

The Magnaflux 11W calibration block had a known 0.060 diameter side-drilled i

hole.

The UT' unit was calibrated using a 20 db factor-while the actual weld was examined using 10 db, which we m

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Mr. Robert T. Carlson June 12, 1974 RO Inspection Report No. 50-219/74-8 Page III considered to be a conservative approach. The examination.

was conducted with the understanding that if there were any

-reflections noted, it would not, be possible to correlate the results with previous examinatien results. However, since the examination did not reveal any reflections in either the weld zone or the base metal,, there was no need-

- l for correlation.

In any event, the Magnaflux 11W calibra-tion block is in accordance with the required codes and the -

equipment calibration method was adequate and correct.

To preclude any uncertainty, the JCP6L calibration standard was located and the UT examination of the reactor vessel skirt was redone after recalibrating the equipment to the JCP6L standard. No defects were found and the results of '~~

both UT ' examinations are part of the 1974 in-service 1

inspection records.

Very truly yours, f) d d t. / f (, %. -

f Donald A. Ross Manager, Nuclear Generating Stations pk 4

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E.'D. Thereburg, chief i

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y,3 the subject.shmormal assessemos sepost.is faseesdod for settaa.

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Based en our vertow of the 16esasee's pre 14=tmary soport, it is mesammended that R0 Meadquarters transfer the Raad gesponsibility to DL for review and l

evaluation of this penetraties fa11ers involving primary ecolant leahmes i

at the reactor vessel's betten head. Additionally, DL should be regossted to evaluate potential generic aspects of this problem.

j RorI will sontinue to provide field follow up and review of any fortheeming corrective measures by the licensee in this matter.

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E..J. Brunner, Chief l

Reactor Operations Branch I

Enclosure l

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From:.

Jersey Central Pnwer 4 Light Company Oyster Cnek Nuclear Generating Statism, Docket #50-219 Porked River, New Jersey 08731 a:' '

Subject:

Abnormal 0ccurrence Report No. 50-219/74/34-n The following is a: preliminary report being submitted in compliance with the Technical Specifications

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paragraph 6.6'.2.

l Vreliminary Approval:

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Rep. ort, No. 50-219/74/ 34

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IDBtfTIPICATION

' Violation of the Technica'l Specifications, garagraph N/A,

OP OCCURRENCE:

Indications of. coolant Irakage existing in tNo area or an I

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incow flux monitor reactor vessel housing located at core coor-e dinate 28-05.

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l his event is considered to.be an abnormal occurrence as de-1 1

fined in the Technica1' Specifications, paragraph 1.15E 2

CONDITIONS PRIOR l

.TO OCCURRENCE:

. Steady State Power '

. Routine Shutdown Hot Standby Operation i

Cold Shutdown Load: Changes During i

X-Refueling Shutdown Routine Power Operation l

Reutino Startup Other, ($pecify)

Operation 1

1 De reactor was in'the REPIEL mode during'a hydrostatic test i

at'.850ipsig pressure andwith coolant' t'ogerature approximately 155'P,-

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J DESCRIPTION On Tuesday,.Nay 28, 1974, during a scheduled reactor'vossyf

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hydrostatic test to inspect the pressure boundary fo11owing refooling maintenance activities, leakage was observed in the vicinity of an incore flux monitor hube located at ths' botton of the reactor vessel. Further investigation conducted'on j

. Nednesday, May. 29,1974, showed evidence of possible leakage i

in the ama of an incom flux monitor' housing penetration l

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. located in the reactor. vessel bottom head. A second hydro-y ;" _.,r :

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?. s$,atic test was conducted'st a pressure of 850 psig at approm:1-

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observed leaking between'the monitor housing and the reactor

-vessel.. He leakage was measured eder the conditions;of 850 4

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. psig with. a temperature ofS164'P; andJealculated'to:be our the

, v order of approximately.0.02 ga11ons per'thour.

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APPARENT CAUSE Design

' Procedure OF OCCURREN2:

Manufacture

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,he cause of this event:has yet to be determined.

i ANALYSIS OP.

As stated in FDSAR Amendaant #37, a pos'tulated failure of the 1

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i flux monitor tihe would result in vessel leakage.at a rate

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$1ch would not cause. excessive claddins; temperatures and for i

l which core refloodisg is p'ssible by engineered safety features.

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I his situatica is less severs than tho. design basis accident.

To determine the cassequences of a wold f Iure at a housing

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for an in-com acnitor tube, it is mastaned that the weld between a....

the housing and the reactor vessel bottom head fails, allowing i

the'ho'asing and the in-core monitor tube to"be ejected from the vessel. %e hole provided'.in,the bottom head for the housing g

has a diameter of two inches; tinis is the assumed break size.

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He hole has a break area of.0218 ft. Assuming worst condi-tions...this results in.neak clad temperatures less than 1000*P, as gdated in PDSAR Amer.dment f67.

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acceptable. limits of the applicable ECCS criteria.

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.p, 00RNBCTIVE 1he nucIsar steam sqply wndor and the x. actor vessel manu-ACTIOt:

factuar have been contacted with agard to this conditi'on.

Dis'cussion will ensus as to the proper course of action to be j.

taken to resolve this matter. Recommendations will be forth-coming pending cosplete review of this event by'the Plant Opera-

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ticas Review Committee.

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Date:

5/30/74 Propand by:

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