ML20196G479

Informs That NRC Considered Proposed GL Dealing with Augmented Insp of PWR Class 1 High Pressure Safety Injection Piping.Suggested Rev to List of Questions Proposed by NEI Encl

ML20196G479
Person / Time
Issue date:	12/03/1998
From:	Wessman R NRC (Affiliation Not Assigned)
To:	Cozens K NUCLEAR ENERGY INSTITUTE (FORMERLY NUCLEAR MGMT &
References
PROJECT-689 NUDOCS 9812080017
Download: ML20196G479 (4)

Text

__

Mr. Kurt Cozens December 3,1998 Nuclear Energy Institute 1776 l Street N.W., Suite 400 Washington, D.C.,20006-3708 l

SUBJECT:

CRACKING IN HIGH PRESSURE INJECTION LINES IN NUCLEAR POWER PLANTS

Dear Mr. Cozens:

Earlier this year the NRC considered a proposed generic letter dealing with augmented i

inspection of PWR Class 1 high pressure safety injection piping. This generic letter was being l

considered as a result of recent domestic and foreign reactor experience with thermal fatigue in l

reactor coolant system piping. Related to this issue was an NRC concern with ASME Section XI l

inservice inspection requirements that only mandate a surface inspection of the welds in small l

diameter ASME Section ill Class 1 piping.

l Recently, NEl agreed to undertake a voluntary initiative to develop a technical basis why l

additional measures are, or are not, necessary to assure the integrity of welds and piping in l

Class 1 systems which may be subjected to thermal fatigue. This initiative is tentatively being l

considered as part of the PWR Materials Reliability Project, coordinated by EPRI with NEl participation. In follow up to recent conversations, between NEl and the NRC staff, we have exchanged proposed lists of questions for consideration as part of this initiative, by E-mail.

l Attached is our suggested revision to the list of questions proposed by NEl.

l If you have questions, please contact Dr. M. Hartzman at (301) 415-2755, or myself at (301) 415-3288.

Sincerely yours,

/s/

Richard H. Wessman, Chief Mechanical Engineering Branch Division of Engineering Office of Nuclear Reactor Regulation

Enclosure:

Proposed Questions 1

cc w/ enclosure:

• See previous concurrences J. O'Hanlon, VEPCO l

T. Mulford, EPRIDistribution; File Center /PDR EMEB RF MMayfield EHackett EDeBec-l Mathet MMitchell LLund JStrosnider SMcGruder KWichman JWiggins RMallett JGrobe l

AHowell DOCUMENT NAME: G:\\WESSMAN\\ COZEN 2.LTR t

j To receive a copy of this document. ind;cate in the box C= Copy w/o attachmentienclosure E= Copy with l'

j attachment / enclosure N = No copy l

OFFICE EMEB:DE E

EMEB:DE E

EMCB.DE E

EMF,1) des NAME MHartzman*

KManoty*

KWichman*

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,g DATE 12/y98 12/03/98 12/03/98

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J OFFICIAL RECORD COPY

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9812080017 981203 h -j r1 - 7 [V//E PDR REVGP ERONUMRC j

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l s

Mr. Kurt Cozens Nuclear Energy Institute 1776 i Street

.W., Suite 400 Washington,

.C., 20006-3708 l

SUBJECT:

RACKING IN HIGH PRESSURE INJECTION LINES IN NUCLEAR POWER NTS l

Dear Mr. Cozens:

1 Earlier this year the RC considered a proposed generic letter dealing with augmented inspection of PWR C ss 1 high pressure safety injection piping. This generic letter was being considered as a result f recent domestic and foreign reactor experience with thermal fatigue degradation in reactor c olant system piping. Related to this issue was an NRC concern with l

ASME Section XI inservi inspection requirements that only mandates a surface inspection of I

the welds in small diamete ASME Section lil Class 1 piping welds.

Recently, NEl agreed to und ake a voluntary initiative to develop a technical basis why additional measures are, or ar not, necessary to assure the integrity of welds and piping in Class 1 systems which may be bjected to thermal fatigue. This initiative is tentatively being considered as part of the PWR M terials Reliability Project, coordinated by EPRI with NEl participation. In follow up to recent onversations, between NEl and the NRC staff, we have exchanged proposed lists of questio for consideration as part of this initiative, by E-mail.

Attached is our suggested revision to e list of questions proposed by NEl.

If you have questions, please contact Dr.. Hartzman at (301) 415-2755, or myself at (301) 415-3288.

Sincerely yours, Richard H. Wessman, Chief Mechanical Engineering Branch ivision of Engineering ice of Nuclear Reactor Regulation

Enclosure:

Proposed Questions cc w/ enclosure:

J. O'Hanlon, VEPCO T. Mulford, EPRIDistribution: File Center /PDR EMEB RF ayfield EHackett EDeBec-Mathet MMitchell LLund JStrosnider SMcGruder KWichma JWiggins BMallett JGrobe AHowell DOCUMENT NAME: G:\\WESSMAN\\ COZEN 2.LTR To receive a copy of this document. indicate in the bogC= Copy w/o attachment /e losure E= Copy with attachment / enclosure N = No copy

'l OFFICE EMEB.DE d

EMEB.DE if Ebc IE i@

EMEB DE NAME MHa KMa y RWessman DATE f) f198 4M/98 h8

/ /98 l

OFFICIAL RECORD COPY l

~

PROPOSED QUESTIONS FOR HPl LINE CRACKING INITIATIVE A)

As a result of thermal fatigue concerns, has the plant's owner conducted volumetric l

inspections of welds in small diameter (less than 4") ASME Class 1 piping connected to I

the RCS pressure boundary, beyond the requirements of IWB 25007

1) If yes:

I a.

When were these volumetric inspections conducted?

I b.

At what locations were the welds inspected?

c.

What inspection techrJques were used?

I d.

If any indications were found as a result of these inspections, was any follow up action taken, e.g., replacement of the affected weld, etc., to l

address the indications?

2)

If not, does the potential of thermal fatigue failure form the basis for the selection of welds in Class 1 piping 4" or less in diameter currently examined as part of the l

Section XI program?

a.

If not, what is the basis for the weld selection?

b.

If yes:

(1)

At what locations are the welds inspected?

(2)

What inspection techniques are used?

(3)

If any 's'ications are found as a result of these inspections, is any follow Jp action taken, e.g., replacement of the affected weld, etc.,

to adarestithe indications?

B)

Has the plant's owner ever peribrmed any base metal examinations on unisolable piping l

connected to the RCS because of thermal fatigue concerns? If yes, 1.

At what locations were these examinations performed?

l 2.

What were the criteria for the selection of these locations?

3.

Were any indications found at these locations?

C)

Does the plant have a very small number of welds (less than10), a moderate number of welds (10 to 30), or a significant number of welds (greater than 30), in unisolable piping less than 4-inch nominal attached to the RCS?

ENCLOSURE 4

l

~

2 D)

Does the plant's owner maintain a program for assessing the potential for thermal fatigue cracking within unisolable piping connected to the RCS?

j

1) '

If yes, summarize the program and indicate if the program includes:

a) temperature monitoring?

f b) pressure monitoring?

b c) valve condition (intemal leakage) assessment?

i d) any other monitoring means?

e) analytical assessment?

(1) what program is used?

2)

If not, what is the basis for not maintaining such a program?

a) analytical assessment?

l l

l (1) what program is used?

i l

(2) has the NRC reviewed this program?

(a) If yes, what was the date of acceptance?

l b) any other basis?

E)

Does the plant's owner intend to maintain the program stated in (D,1) above through the end of the facility's operating license?

i i

j 1)

If not, what is the basis for not maintaining the program?

l F)

Did the plant's owner establish a program with provisions for monitoring, as stated in l

(D,1) above, and then terminate it?

l L

1)

If yes, what was the basis for terminating the program?

I r

s i

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