ML20198S632

Discusses Which Denied Violation of 10CFR50.65, Issued on 970910,in Conjunction W/Insp Repts 50-361/97-15 & 50-362/97-15.NRC Decided Not to Pursue Enforcement Action Relative to Apparent Criterion Xvi Violation

ML20198S632
Person / Time
Site:	San Onofre
Issue date:	01/15/1998
From:	Merschoff E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	Ray H SOUTHERN CALIFORNIA EDISON CO.
References
50-361-97-15, 50-362-97-15, EA-97-414, NUDOCS 9801260243
Download: ML20198S632 (6)

See also: IR 05000361/1997015

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ARUNGioN TEXAS 7t,0114064 -

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January 15, 1998

EA 97-414

Harold D. Ray, Executive Vice President

Southem Califomia Edison Co.

San Onofre Nuclear Generating Station

P.O. Box 128

- San Clemente, California 92674-0128

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SUBJECT:

RESPONSE TO NRC INSPECTION REPORT 50-301/97-15; 50-362/97-15 AND

DENIAL OF NOTICE OF VIOLATION

Dear Mr. Ray:

This is in reference to your letter dated October 24,1997, in which you denied a violation of

10 CFR 50.65, " Requirements for Monitoring the Effectiveness of Maintenance at Nuclear

Power Plants." Your letter was in response to a Notice of Violation issued September 10,

1997, in conjunction with NRC Inspection Report 50-361/97-15; 50-362/9715. This also is

in reference to the predecisional enforcement conference conducted in the NRC's Arlington,

Texas, office on September 30,1997, with you and other representatives of Southem

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Califomia Edison Co. (SCE). The conference was conducted to discuss an apparent violation of

10 CFR Part 50, Appendix B, Criterion XVI (Corrective Action), which was described in the same

inspection report. Both issues relate to recurring cracking of Inconel 600 reactor coolant system

nonles at San Onofre Nuclear Generating Station (SONGS). A copy of the conference

transcript and related conference material were sent to SCE and to the NRC's Public Document

Room by separate correspondence on December 9,1997. The NRC also considered the

additionalinformation SCE provided following the conference by letters dated October 3 and 31,

1997.

' The NRC has decided not to pursue enforcemed action relative to the apparent Criterion XVI

violation. Our decision is based on our conclusion that the maintenance rule violation, which

you have denied, and the apparent corrective action violation are closely related, and that many

of the same arguments you made in denying the maintenance rule violation also are applicable

to the apparent corrective action violation. The corrective action issue is, in our view, subsumed

by your arguments related to the maintenance rule violation. Thus, in the interest of conserving

b nh NRC and SCE resources, we have determined that we will pursue enforcement only for the

maintenance rule violation. Accordingly, we have addressed your response to the maintenance

rule violation below.

.

We acknowledge that you have implemented a comprehensive inspection program to detect

primary weter stress corrosion cracking of reactor coolant system instrument nozzles during

outages and have pe. formed appropriate repairs upon detection. However, as noted in your

response, instrument nonle performance problems continue to occur as evidenced by the

several nonles that wsre found to be cracked during noule inspections performed before and

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after recent refueling outages As you also acknowledged in your response, nonle

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replacements for penetration leakage that had been performed under construction work orders

were not considered by SCE when it developed its maintenance rule program. Thus, SCE's

decision to place the reactor coolant system under 10 CFR 50.65(a)(2) was based on

incomplete data.

As a result of our review of the response to the violation, we found that you have provided no

additional relevant information beyond what was already obtained during the inspection process.

in particular, you have continued to indicate that you do not consider identification of evidence of

penetration leakage to represent functional failures of the reactor coolant system under

10 CFR 50.65, notwithstanding the fact that the Technical Specifications do not permit continued

operation until such leakage is repaired. As you know, part of a defense-in-depth apprcach to

reactor safety is maintaining the integrity of the reactor coolant system to provide a t:arrier for

fission products. Your response to the violation concluded that the reactor coolant system

remained capable of performing its intended function (i.e., to maintain structural integrity and

meet the allowable leakage limits of the technical specifications ). We disagree with this

conclusion, in part, because the staff view is that cracking of instrument nonles does represent

a functional failure of the pressure boundary to prevent leakage as defined by the facility

Technical Specifications, Section 1.1. Therefore, we conclude that a violation of 10 CFR 50.65

occurred at the San Onofre Nuclear Generating Station as stated in the Notice of Violation dated

September 10,1997. This conclusion has been coordinated with the Office of Nuclear Reactor

ReCulation and the Office of Enforcement.

W2 na agree with SCE's point that there was evidence of through-wallleakage in three

inw.nces and not four, as discussed in the Notica of Violation. The NRC will issue a correction

to the Notice of Violation in conjunction with our issuance of a letter noting corrections to the

September 10,1997, inspection report.

in your October 24,1997, letter, you identified that, because of several recent problems, the

reactor ;oolant systems of both SONGS units have been placed in a 10 CFR 50.65 (a)(1)

category. You also stated in this letter that S'

'limplement strategies to replace

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penetrations over time which are considered more susceptible to leakage than others, with

ALARA considerations requiring this to be done selectively. In addition, we note from review of

. Document 90022, Tusceptibility of Reactor Coolant System Alloy 600 Nonles to Primary Water

._ Stress Corrosion Cracking and Replacement Program Plan," Revision 2, which was submitted

by letter dated October 31,1997, that you have established specific replacement plans for

reactor coolant system inconel 600 nonle penetrations. Section 6.2 of Document 90022,

Revision 2, indicates all reactor coolant system piping instrument nonles are currently

- scheduled for half nonle repairs during both the Cycle 9 mid-cycle outage (upper 45' hot leg

and cold leg nonles) and the Cycle 10 refueling outage (90' and 135' orientation nonles).

Section 6.1 of Document 90022, Revision 2, indicates that you currently plan to install

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Southern California Edison Co.

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. ABB Combustion Engineering-designed mechanical nozzle seal assemblies (MNSAs) on

remaining inconel 600 pressurizer water space nozzles during the Cycle 9 mid-cycle outages.

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Section 6 2 also states that any nozzle locations identified with evidence of leakage during the

mid-cycle outages ihat are not scheduled for repair will have a MNSA installed. Use of MNSAs -

is contingent on NRC approval of a replacement request for relief from ASME Code, Section lil,

requirements, which was submitted by letter dated December 12,1997. This submittal

responded to an NRC request for additional information dated November 20,1997, and also

clarified and revised information in the onginal request for relief which was submitted by letter

dated July 11,1997.

We consider the placing of the reactor coolant systems for both SONGS units in a

10 CFR 50.65(a)(1) category, together with the olanned scope and time frame of repairs,

provides an appropriate framework of corrective actions for the Notice of Violation which

was issued September 10,1997, in conjunction with NRC Inspection Report 50-361;-362/97-15.

~ However, in accordance with 10 CFR 2.201, the NRC requests that SCE submit a formal

response to the original Notice of Violation issued September 10,1997. Please follow the

instructions specified _in the Notice when preparing your response. The NRC will use your

response, in part, to determine whether further enforcement action is necessary to ensure

compliance with regulatory requirements.

in accordance with 10 9 2.790 of the NRC's " Rules of Practice," a copy of this letter and your

response wi'l be placed in the NRC Public Document Room (PDR).

Sincerely,

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Ellis W. Mersch

Regional Admirastrator

Docket Nos.: 50-361;50-362

License Nos.: NPF-10; NPF-15

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cc:

Chairman, Board of Supervisors

County of San Diego

1600 Pacific Highway, Room 335

San Diego, Califomia 92101

Alan R. Watts, Esq.

Woodruff, Spradlin & Smart

-701 Sc Parker St. Suite 7000

Orange, California 92868-4720

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Sherwin Harris, Resource Project Manager;

Public Utilities Department

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City of Riverside

3900 Main Street -

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Riverside,Califomia 92522:

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' R. W.- Krieger, Vice President J

Southem Califomia Edison Company

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San Onofre Nuclear Generating Station .

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P.O. Box 128

San Clemente, Califomia 92674-0128

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(Stephen A. Woods, Senior Health Physicist -

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Division of Drinking Water and .

Environmental Management -

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Nuclear Emergency Response Program

Califomia Department of Health Services

- P,0. Box 942732, M/S 396

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Sacramento, Califomia 94334 7320

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Terry Wnter, Manager

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P.O. Box 1831

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San Diego, Califomia 92112

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Mr. Steve Hsu-

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Radiological Health Branch

State Department of Health Services -

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Sacramento, California 94234

Mayor -

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- 100 Avenida Presidio

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- San Clemente, Califomia - 926721

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Mr. Truman Bums \\Mr. Robert Kinosian .

Califomia Public Utilities Commission

2 505. Van Ness, Rm. 4102 -

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U. S. Nuclear Regulatory Commission -

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Gentlemen:

Subject:

Docket Nos. 50-361 and 50-362

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Reply to a Notice of Violation

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San Onofre Nuclear Generating Station, Units 2 and 3

- References: 1)

Letter, Mr. A. T. Howell lli (USNRC) to Mr. Harold B. Ray (SCE),

dated September 10,1997

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2):

Letter, Mr. G. T. Gibson (SCE) to Mr. A. T. Howell lli (USNRC),

dated July 22,1997

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3).

Meeting, NRC/SCE meeting on Maintenance Rule aspects of-

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RCS nozzle PWSCC (Handouts), dated August 21,1997

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Meeting, NRC Predecisional Enforcement Conference (Handouts),

dated September 30,1997

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Letter, Mr. Dwight E. Nunn (SCE) to Mr. E. W. Merschoff (USNRC),

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dated October 3,1997

- Reference 1 transmitted the results of NRC Inspection Report No. 50-361 and 50-

362/97-15, which concems an inspection conducted at the Southen. California Edison -

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(SCE) San Onofre Nuclear Generating Station, Units 2 and 3. The enclosure to the

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• Refere%e 1 letter contained a Notice of Violation (362/9715-02) which states that,

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contrary to the requirements of 10_CFR 50.65, SCE failed to demonstrate that the

condition of the reactor coolant system was being effectively controlled through the

performance of appropriate preventive maintenance, such that the system remained

capable of performing its intended function. : SCE does not agree or admit that this

violation occurred, as discussed below and in the enclosure to this letter.

- The basis for our conclusion that no violation occurred is that the reactor coolant

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system demonstrably _did remain capable of performing its intended function,

notwithstanding the circumstances cited in the inspection report. The preventive

maintenance performed to effectively control the candition of the reactor coolant system

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October 24,1997

was in accordance with the guidance available from the NRC itself, and did assure that

the system remained capable of performing its intended function. -Indood, we continue

to follow that guidance currently and believe it fully satisfies NRC requirements.

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In order for us to conclude that we failed to demonstrate that the reactor coolant system

remained capable of performing its intended function, we believe we would need to be '

able to identify corrective action, which, if taken, would achieve such demonstration.

We specifically do not believe that the establishment of goals in accordance with -

- 10 CFR 50.65(a)(1), applicable to primary water stress corrosion cracking (PWSCC) of

reactor coolant system Alloy 600 penetrations, would achieve this result, even if the

establishment of such goals were practical. Thus, in the absence of the ability to set

goals applicable to PWSCC of Alloy 600 penetrations which would achieve

demonstration of intended system function, and given that we conclude that there was

-no loss of intended function resulting from the PWSCC of Alloy 600 penetrations, _

based upon the extensive inspection and preventive maintenance conducted which we

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- believe meets and exceeds applicable NRC gu! dance, SCE concludes that no violation

occurred.

An important consideration in reaching this conclusion is our understanding of the

purpose of the requirement in Technical Specification 3.4.13 prohibiting leakage from

- the reactor coolant system pressure boundary. As noted in the bases for this Technical

Specification, leakage from joints and interfaces is anticipcted during plant life, through

either operational wear or mechanical deterioration. When even minute leakage is

detected from the reactor coolant system pressure boundary, prompt shutdown and

- repair is required. ; Haever, total unidentified leakage of as much as 1 gpm is

considered to not compromise safety. As amply describe't in the NRC guidance

- concerning PWSCC of Alloy 600 penetrations (which is described in Appendix A of the

enclosure to this letter), inspection for evidence of leakage during plant shutdowns, and

replacement or repair of penetrations found leakingi provides adequate assurance that -

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tnere will be no loss of intended function of the reactor coolant system. References (2)E

through (5) provide additional relevant information in this regard.

Finally, while we conclude that we continue to manage the consequences of PWSCC of

Alloy 600 penetrations in full compliance with the Technical Specifications, and both

NRC and industry guldunce, we also independently conclude that the preventive

maintenance we perform in this regard conservatively assures that the retWor coolant

system remains capable of performing its intended function. Because of the

- operational inconvenience resulting from penetration leakage, we continue to

aggressively examine strategies for replacement of penetrations prior to any leakage

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occurring, consistent with the need to maintain radiation exposure resulting from

penetration replacement ALARA.

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ENCLOSURE

>EPLY TO A NOTICE OF VIOLATION

The enc!osure_to Mr. A. T. lowell's letter dated September 10,1997 states, in part:

"During an NRC inspection conducted on June 30 through September 2,1997, one

violation of NRC requirements was identified. in accordcnce with the ' General

Statement of Policy and Procedure for NRC Enforcement Actions,' NUREG-1600,

the violation is listed below.

"10 CFR 50.65(a)(1) states, in part, that each holder of a license to operate a

nuclear plant shall monitor the performance of structures, syctems, or components,

against licensee-established goals, in a manner sufficient to provide reasonable

assurance that such structures, systems, and components, as de ined in

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paragraph (b), are capable of fulfilling their intended functions.- Such goals shall be

established commensurate with safety ar,ti, where practical, take into account

industry wide operating experience.

"10 CFR 50.65(a)(2) states, in part, that monitoring as specified in paragraph (a)(1)

is not required where it has been demonstrated that the performance or condition of

a structure, system or component is being effectively controlled through the

performance of appropriate preventive maintenance, such that the structure, system

or component remains capable of performing its intended function.

• 10 CFR 50.65(c) states that the requirements of this section shall be implemented

by each licensee no later than July 10,1996.

" Contrary to the above, as of July 10,1996, the time when the licensee elected to

not monitor the performance or condition of the reactor coolant system against

licensee-established goals pursuant to the requirements of Section (a)(1), the

licensee failed to demonstrate that the condition of this system was being effectively

controlled through the pedormance of appropriate preventive maintenance, such

that the system remained capabla of performing its intended function. Specifically,

the licensee inadequately evaluated the appropriateness of the performance of

preventive maintenance prior to placing the Unit 3 reactor coolant system under a

10 CFR 50.6S(a)(2) category (i.e., the licensee did not consider in its evaluation the

identification in 1995 of through-wall cracking in four reactor coolant system nozzle

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- penetrations, which represented multiple failures of the barrier function of the

reactor coolant system).

"This is a Severity Level IV violation (Supplement 1) (50-362/9715-02)."

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Enclosure -

L1. Reply to the Violation

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For convenience in review of the response, the violation is addressed in three parts, as

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follows: (Note: The response considers primary water stress corrosion cracking

(PWSCC) of Alloy 600 reactor coolant system nozzles only, and this is not repeated

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throughout the response.)

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Part A of the Violation: SCE did not consider in its evaluation the identification in 1995

of throughesil cracking in "four' [ sic] reactor coolant system nozzle penetrations which -

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represented multiple failures of the barrier function of the reactor coolant system.

'(Note: SCE review of data indicates that there was evidence of through-wall leakage in:

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three instances; not four.)

Response to Part A' The " barrier function' of the reactor coolant system is n7t a

_ defined term. it can only be understood by reference to the Technical

- Specifications and to General Design Criterion (GDC) 14. The bases of Technical

Specification 3.4.13 includes the following.

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' Component joints are made by welding (emphasis added), bolting... During

plant life the joint and valve 'nterfaces can produce varying amounts of

reactor coolant LEAKAGE, through either normal operational wear or

mechanical deterioration. The purpos'; cf the RCS Operational LEAKAGE

LCO (1 gpm unidentified leakage] is to limit system operation in the

. presence of LEAKAGE from these sources to amounts that do not

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compromise safety."

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Section 1.1 of the Technical Specifications defines Pressure Boundary LEAKAGE

as, " LEAKAGE (except SG LEAKAGE) through a nonisolable fault in an RCS

component body, pipe wall or vessel wall." Technical Specification 3.4.13 itself

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also provides that there shall be "No Pressure Boundary LEAKAGE".

GDC 14 states that, 'The reactor coolant pressure boundary shall be designed,

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faoricated, erected, and tested so as to have an extremely . low probability of '

abnormal leakage [ emphasis added), of rapidly propagating failure,.and of gross

rupture".

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Taken together, SCE understands the foregoing to require that any pressure -

boundary leakage which is identified must be promptly corrected. SCE has not

only complied with this requirement at all times,' but has implemented extensive -

and aggressive ~ inspection measures ,often involving significant maintenance

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Enclosure

activities - to assure that any indication of Pressure Boundary LEAKAGE at

penetrations is identified during plant shutdowns.

Extensive NRC guidance has been issued concerning PWSCC of Alloy 600

ponstrations. At no time has this guidance indicated that measures other than

inspection and repair during outages is needed, although this possibility has been

addressed. (Please refer to Appendix A to this Enclosure.) SCE concludes that

the condition resulting in this instance is not considered by the NRC to be a " failure

of the barrier function' of the reactor coolan; system, as stated in Part A of the

Violation. Moreover, SCE believes this to be reasonable, in that - as discussed in

the NRC guiaance at length - the penetration will remain intact and will continue to

provide a substantial barrier against leakage from the reactor coolant system, such

that it continues to perform its intended function as described in the bases for the

Technical Specifications and GDC 14.

Accordingly, consistent with the referenced NRC guidance, SCE does not consider

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identification of evidence of penetration leakage due to PWSCC of Alloy 600,

which may be identified during inspections conducted for the purpose of early

detection of such leakace, to represent functional failures of the reactor coolant

system under 10 CFR a0,65, notwithstanding the fact that the Technical

Specifications do not p ermit continued operation until such leakage is repaired.

(We are required to evaluate stra turcs, systems, or components (SSCs) against

the established performance criteria t 'ing historical plant data, and industry data

where applicable, to determine if the SSCs met the performance criteria.

Performance criteria for the reactor coolant system consist of functional failures

and system availability.)

Thus, the fact that SCE did not consider the 1995 instances of penetration leakage

prior to placing the Unit 3 reactor coolant system under a 10 CFR 50.65(a)(2)

category as of July 10,1996, is not a violation of requirements because these

instances were not considered functional failures of the reactor coolant system.

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Part B of the Violation: SCE inadequately evaluated the appropriateness of the

performance of preventive maintenance prior to placing the Unit 3 reactor coolant

system under a 10 CFR 50.65(a)(2) category.

. Response to Part B: In accordance with the Regulatory Guide 1.160 section titled,

"Use of Existing Licensee Programs," and NUMARC 93 01, Section 7.0, " Utilization

of Existing Programs", SCE used existing program results to support the

determination that reactor coolant system performance was being effectively

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Enclosure

controlled through preventive maintenance. That is, SCE's existing program, which

is documented in Program Plan 90022, " Susceptibility of Reactor Coolant System

Alloy 600 Nozzles to Primary Water Stress Corrosion Cracking and Replacement

Program," provided an adequate evaluation of the appropriateness of preventive

maintenance in this instance, and there was no need for another evaluation to be

performed. Thus, the lack of another evaluation is not a violation of NRC

requirements.

Part C of the ViolatioD: The licensee failed to demonstrate that the condition of the

reactor coolant system was being effectively controlled through the performance of

appropriate preventive maintenance, such that the system remained capable of

performing its intended function.

Response to Part C: SCE's preventive maintenance program for reactor coolant

system penetrations is implemented in its Alloy 600 program referenced above.

The program includes requirements for identifying inspection frequencies and

corrective actions to be taken when indications of leakage are detected. Most

importantly, the program is based on extensive industry and NRC evaluations of

the significance of PWSCC of Alloy 600 penetrations on the capability of the

reactor coolant system to perform its intended function. As documented in

Appendix A to this Enclosure, these evaluations support the use of inspection and

repair of leakage as providing assurance that the intended function will be

maintained.

SCE has considered what alternative action it might have taken. The NRC inspection

report and Notice of Violation indicate that the reactor coolant system should have

been placed in category 10 CFR 50.65 (a)(1) on July 10,1996, such that the

performance of the penetrations would have been monitored against goals which we

would have established commensurate with safety and industry-wide operating

experience. We believe that use of the existing program, which is fully permitted by

Regulatory Guide 1.160, entirely satisfied this purpose of category (a)(1). Thus, SCE

did demonstrate that the system was being effectively controlled through the

performance of appropriate maintenance, as permitted by category (a)(2) no further

goal-setting was needed, and the fact that the leakage of the penetrations did not result

in placement of the system in category (a)(1)is not a violation of NRC requirements.

Finally, SCE has considered what additional licensee-established goals it might

establish under 10 CFR 50.65 (a)(1) for the penetrations, consistent with safety (as

evaluated by the NRC), industry experience, and ALARA. We cannot identify such

goals, but we would welcome NRC guidance in this regard, applied on an industry-wide

basis.

4

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. .

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_ _ _

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Enclosure

'

l2.0 ctions Taken'

A

As noted in _the inspection report, the penetration replacements were accomplished

_

l

under construction work orders. The failure history screening performed as part of the '

.

- Maintenance Rule implementation prior to July 10,1996, did not include review of .

.

_

~

- construction work ordsrs. SCE has completed review of construction work orders

implemented since July 1993, and determined that this was an isolated occurrence.

'

SCE policy is to proactively replace any penetrations that can reasonably be predicted

'to leak, prior to such a leak developing. In addition, in order to minimize future impacts

,

_

1

to ' operational reliability, SCE will implement strategies to replace penetrations over

time which are considered more susceptible to leakage than others. - However,' ALARA

considerctions require that this be done selectively.

!

As a result of several recent problems, including steam generator tube leakage, steam

E

" generator manway gasket leakage, and a shutdown cooling valve plug leak, the reactor

coolant systems of both units have been placed in category (a)(1). The Alloy 600

penetrations will continue to be managed under updates of the program referenced

.

above, and this program will be referenced in other appropriate documents.

"

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APPENDIX A '

ALLOY 600 RCS PENETRATION NOZZLES:

NRC AND INDUSTRY PERSPECTNES-

As noted in the chronology below, Alloy 600 RCS penetration nozzle PWSCC first :

occurred at San Onofre in 1986. ' Alloy 600 RCS penetration nozzle.PWSCC has

occurred at numerous facilities and has been subsequently observed in RCS head

penutrations, RCS Pressurizer penetrations, and finally in RCS piping (both hot leg and

cold leg) penetrations.

!>

From a review of the regulatory record, the limiting case, from a safety significence

perspective, is considered to be the RCS head penetrations. This is due to their larger

physical size (worst case in the event of catastrophic failure) and the difficulties in

,

performing visual inspections due to interferorces. Therefore, the RCS head

penetrations have been and continue to be considered the primary focus, and the

-

bounding case for accident and safety significance analyses.

1

~

1

.

_ As provided below, the NRC and industry's guidance on the safety significance of

!

Alloy 600 RCS penetration PWSCC are consistent:

PWSCC is a known phenomena.' Alloy 600 RCS penetration nozzles are

o

,

susceptible to PWSCC. There is no reliable predictive methodology for

' explicitly predicting individual nozzle susceptibility to PWSCC. PWSCC is

a function of time, temperature, residual stress in the nozzle, weld,

i

'

microstructure, and water chemistry.

.

PWSCC in Alloy 600 RCS penetrations which are not roll expanded (EdF

[

e

nozzles are roll expanded) results in only axial cracking due to the stresses

involved;

e

PWSCC axial cracks are short in length, and crack growth beyond the

initial weld region is very slow since operating stresses in the region are

low,

,

_ Augmented visual inspections for cracks and indications (boric acid

o

residue) are relied upon to identify PWSCC, and upon discovery repair

{

and/or replacement is effected to the identified nozzle.

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Appendix A -

The following six examples are best illustrative of the NRC's independent review and

guidance regarding Alloy 600 RCS penetration PWSCC:

Eramole 1: Januarv 1995 NRC Petition Denial D.D. 95 2

On January 26,1995, the Director, Office of Nuclear Reactor Regulation, denied a

petition filed on behalf of Greenpeace intemational, to shutdown plants based on

PWSCC. The NRC's denial states, in part:

"In 1990, the NRC Staff identified to the Commission primary water stress

corrosion cracking (PWSCC) of Alloy 600 in components other than steam

generator tubing as an emerging technical issue after cracking was noted in

pressurizer heater sleeve penetrations at a domestic PWR facility. At that

'

time, the Staff reviewed the safety significance of the cracking as well as the

repair and replacement activities at the affected facility,

j

'The Staff determined that the safety significance of the cracking was low

because the cracks were axial, had a low growth rate, were in a material

l

with an extremely high flaw tolerance (high fracture toughness) and,

accordingly, were unlikely to propagate very far. These factors also

demonstrate that any cracking would result in a detectable leak before a

penetration broke."

" Based on the owners groups safety assessments, a leak in a VHP [ vessel

head penetrationj would be detected before significant damage could occur

to the VHP or the reactor vessel. This would result in the deposition of boric

acid crystals on the vessel head and surrounding area that would be

i

detected during surveillance walkdowns. Consequently, the concerns raised

>

by ihe Petitioner do not raise any immediate safety concerns., immediate

inspections are not required since there is no immediate safety concern....

"CEOG submitted the detailed findings of it's Alloy 600 component PWSCC

program initiated in 1990 to the Staff in a proprietary repoit on Februany 26,

1992. The conclusions of the report, which focused primarily on pressurizer

heater sleeves and instrument nozzles (em,chasis added), in part, follow:

"1) Circumferential cracking of the heater sleeves and the

instruiuentation nozzles [ emphasis added) is not a credible failure

mode...

-2-

.

-

---

-

.- .

-

. _ .

- = - -

-

-__ _-

.

-

..

Appendix A

"3) Visual inspection is the best method for detecting a leaking sleeve or

nozzle... [ emphasis added)

,

'The Staff has reviewed the report, and finds that it's results and

recommended inspections, coupled with field experience, provide a

sufficient basis to conclude that loss of structural integrity and ejection of

components with respect to pressurizers are highly unlikely."

EEam91e 2: SECY 97-063. March 1997

Proposed NRC Generic Letter: " Degradation of Control Rod Drive Mechanism

4

and Other Vessel Closure Head Penetrations"

"Beginning in 1986, leaks have been reported in several Alloy 600

pressurizer instrument nozzles [ emphasis added) at both domestic e d

foreign reactors... Tine NRC staff identified primary water stress corrosion

cracking (PWSCC) as an emerging technical issue to the Commission in

1989, after cracking was noted in Alloy 600 pressurizer heater sleeve

penetrations at a domestic PWR facility. The NRC staff reviewed the safety

significance of the cracking that occurred, as well as the repair and

replacement activities at the affected facilities. The NRC staff determined

that the cracking was not of immediate safety significance because the

cracks were axial, had a low growth rate, were in a material with an

extremely high flaw tolerance (high fracture toughness) and, accordingly,

were unlikely to propagate very far. These factors also de'nonstrated

that any cracking would result in detectable leakage and the

- opportunity to take corrective action before a penetration would fall."

Eramole 3: Generic Letter 97-01. Aoril 1997

Generic Letter 97-01 addresses the issue of the potential for cracking in Alloy 600

CRDM nozzles and other vessel head closure penetrations (VHP).

~

"The NRC staff determined that the cracking was not of immediate safety

significance because the cracks were axial, had a low growth rate, were in a

material with an extremely high flaw tolerance (h'ah fracture toughness), and

accordingly,_were unlikely to propagate very far. These factors also

demonstrated that any cracking would result in detectable leakage and the

opportunity to take corrective action before a penetration would fail."

-3-

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.

3 -

4

=m

.+

,

,

.

-

.

.

.

Appendix A

The Generic Letter also states:

"After considering this information, the NRC staff has concluded that VHP

cracking does not pose an immediate or near term safety concern."

Example 4: November 1993 NRC Letter

in a November 19,1993, letter from William T. Russell to William Raisin, the NRC

responded to NUMARC's June 16,1993, letter regarding Alloy 600 CRD

CEDM

W

head penetrations. The NRC's conclusion is:

" Based on the overseas inspection findings and the review of your

analyses, the staff has concluded that there is no immediate safety concern

for cracking of the CRDM/CEDM penetrations."

" Based upon infom1ation received from overseas regulatory authorities, your

analyses, and staff reviews, the staff believes that catastrophic failure of a

penetration is extremely unlikuly. Rather, a flaw would leak before it

reached the critica! flaw size...."

Example 5: NRC Information Notice IN 9010

"The cracking to date in the thermal sleeves and the instrument nozzles

(emphasis added] of the domestic PWRs has been reported as being only axially

oriented. The safety implication of an axial crack is not considered a significant

threat to the structural integrity of the components and most likely will result in a

small leak...Circumferential cracking poses a more serious safety concern

because if it were to go undetected it could lead to a structural failure of a

component rather than to a limited leak."

" ..it may be prudent for licensees of all PWRs to review their Alloy 600

.

applications in the primary coolant pressure boundary, and when necessary,

to implement an augmented Inspection program."[ emphasis added)

Examola 6: NRC Status Reoort to the Commission

On May 12,1993, the NRC staff provided a status report to the Commission

regarding PWSCC of Alloy 600 components. The NRC concluded the following at

that time:

4

.

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.

- - -.

.,

..

--

.- - - - .. - .- _

.

-

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,

Appendix A

l

"Having reviewed the information to date, including the inspection results

'

and findings, the s'aff maintains its view that this issue is of low safety

J

significance sir'.a all cracks reported to date, with perhaps one exception

- (a.k.a. Edr i-rench reactor), are short in length and axially oriented in an

'

4

eminely flaw-tolerant material."

'

Finally, the following is a compendium of the Alloy 600 RCS penetration PWSCC

history, involving the NRC and the industry (including SCE):

1.

1984 NRC SER for Nine Mile Point, Unit 1 (8/29/84)

2.

November 1988, LER 86-003 and 86-003 Rev 1

3.

March 1987 Nine Mile Point Code Relief (3/25/87)

4,

1989 NRC Calvert Cliffs Confirmatory Action Letter Closure

5.

November 1989, CEN 393-P, NP (11/3/89) " Pressurizer Heater Sleeve

,

Susceptibility to PWSCC' [ Issued to NRC on 11/17/89)

6.

February 1990 NRC IN 90-10 (2/23/90), " Primary Water Stress Corrosion

Cracking (PWSCC) of Alloy 600"

7.

March 1990, CE NPSD 555 (3/2/90), " Pres'surizer Heater Performance'

8.

March 1990. EPRl/CEOG PWSCC Meeting (3/14/90) Rockville. MD

9.

August 1990, CE NPSD 832 (8/15/90), " Pressurizer Heater Sleeve Examinations'

10. September 1990, PWSCC Coordinating Group Meeting (9/12/90) Parsippany, NJ

.

11. November 1990, CE NPSD-618 (11/5/90), "Intraspect/ET20 Eddy Current

Imaging Development for Pressurizer Heater Sleeve Inspection for FPL, St. Lucie

Unit 2"

--12. -January 1991, PWSCC Coordinating Group Meeting (1/8/91) Palo Alto, CA

13. February 1991, CE NPSD-817-P (2/25/91)," Destructive Examination of

- Pressurizer Instrument Nozzles from Calvert Cliffs Unit 2 and Evaluation of Similar

Nozzles"

-5-

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,

- . - - . - - . - . .

,-_ - - - - , ,

____ ____ _ _____ __ _ ____

.

.

.

Appendix A

14. March 1991, CE NPSD449 P (3/18/91), "Information Package on Alloy 600

Primary Pressure Boundary Penetrations"-Listing all Alloy 600 penetrations for

RCS and Pressurizer (less Rx Vessel) for all CEOG members.

15. March 1991, CE NPSD432 Part 2 (3/28/91), " Residual Stress Measurements on

Calvert Cliffs 2 Pressurizer Heater Sleeves *

16. April 1991, CE NPSD448 P (4/25/91), ' Corrosion and Corrosion / Erosion Testing

of Pressurizer Shell Material Exposed to Borated Water"

17. o se 1991, CE NPSD446 (6/5/91), CEOG Pressurizer Heater Sleeve Thermal

Analysis'

18. June 1991, CEN-406-P (6/6/91),' Status Report on CEOG Activities Concerning

Primary Water Stress Corrosion Cracking of inconel-600 Penetrations' [Sent to

NRC on 5/31/91 via CEOG-91-300)

19. September 1991, CE NPSD459 P (9/25/91), ' Additional Pressurizer Heater

Sleeve Examinations'

20. November 1991,1" EPRI PWSCC Workshop (10/9-11/91), Charlotte, NC

21. November 1991, CEOG Letter to EPRI (11/12/91), "CEOG Task 692 Near Term

Activities"

22. January 1992, CE NPSD-690 P (1/20/92)' Evaluation of Pressurizer Penetrations

and Evaluation of Corrosion After Unidentified Leakage Develops Pressurizer

Inspection Recommendations' [Provided to NRC on 2/26/92 via CEOG-92-052)

23. February 1992, PVNGS LER 192-001 (2/3/92), APS reports a Unit 1 pressurizer

steam space instrument nozzle leak

24. March 1992, San Onofre LER 2-92-004-00 (3/19192); LER 2 92-004-01 (5/18192)

25. March 1992, NRC/CEOG Meeting (03/25/92)

26. April 1992, NRC Inspection Report 92-06 (SONGS)- Noted that SCE identified

- three Unit 3 nozzle leaks in the pressurizer, resulting from PWSCC. Noted SCE's

effort to resolve the problems were professional and effective. Also discussed

was the meeting held in Walnut Creek where SCE presented information on the

nozzle replacement to the NRC.

-6-

.

_ _ _ _ _ _ .

_ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ . _ _ _ _ _ _ _ _

..

. . - .

.

..._-..

-

.. .-

.

..

-

--

,-

.

.

Appendix A

27. May 1992, NRC inspection Repcrt 9212 (SONGS) - The inspectors reviewed

work with associated Unit 2 nozzle repair, and questioned SCE's effort to

aggressively complete the operability assessment and failure m tanism.

IFl 9212 03 was opened.

)

28. July 1992, NRC inspection Report 92-18 The inspectors reviewed and closed

out LER 2-92-004 revisions 0 and 1 on Pressurizer Nozzle Cracks.

29. August 1992, PWSCC Coordinating Group Meeting (8/11/92), Juno Beach, FL

,

30. October 1992, NUMARC PWSCC Meeting (10/2/92), Washington, DC

31. October 1992, Nozzle Integrity Assessment Meeting (10/21/92), Washington, DC

.

32. November 1992, NUMARC PWSCC Meeting (11/11/92), Washington, DC

33. November 1992 NRC/NUMARC Alloy 600 Nozzle Meeting (11/20/92),

Rockville, MD

d

34. December 1992,2 EPRI PWSCC Workshop (12/1 3/92), Orlando, FL

35. December 1992, Nozzle Integrity Assessment Meeting (12/2/92), Orlando, FL

36. December 1992, PWSCC Integrity Assessment /EdF Meeting (12/4/92),

g

Orlando, FL

'

37. December 1992 - NRC Inspection Report 92-29 (SONGS)- This report closed

IFl 9212-03 related to Unit 2 pressurizer nozzle repair. The inspector (s) identified

concems with timeliness in completing the assessment of the impact of the

leakage. The CEOG evaluation was discussed with SCE and the inspector closed

the IFl.-

38. _ January 1993, PWSCC Integrity Assessment Meeting (1/13/93),

,

Juno Beach, FL

39. February 1993 - NRC Inspection Report 92-28 - SONGS SALP - Stated in

.

general maintenance and surveillance activities conducted more effectively, citing

SCE's effort to repair Unit 3 pressurizer nozzles.

-

40. February 1993, NUMARC PWSCC Meeting (2/19/93), Washington, DC

L

7-

-

l'

.

-

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,

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. .

,

_ _ _ _ _

_

. .

1

.

.

Appendix A

41. March 1993, CE NPSD-903 P (3/22/93), "CEDM Phase 1, Nozzle Evaluation" -

This report provided data on nozzle material heats and configurations for each

member plant.

42. March 1993, NRC/NUMARC Alloy 600 Nozzle Meeting (3/3/93), Rockville, MD

43.' March 1993, CE NPSD-904-P (3/22/93), "CEDM Phase 1, World Follow" -

Documented information from cracking at Bugey and status of other EdF and

World Wide inspections through the beginning of 1993.

44. April 1993 (4/13/93) NRC Inspection Report 93 08 (St. Lucie)

45. April 1993, Nozzle Integrity Assessment Meeting (4/15/93), Charlotte, NC

46. April 1993, EPRl/EdF PWSCC Meeting (5/6/93), Herndon, VA

47. May 1993, NRC Status Report to the Commission (5/12/93)

48. May 1993, Nozzle Integrity Assessment Meeting (5/13/93), Charlotte, NC

49. May 1993, CEN 607 (5/28/93)" Safety Evaluation For ID Axial Cracking" - This

report was developed and issued to the NRC (via NUMARC) in May,1993. It

concluded that ID axial cracking of CEDM/ICI penetrations was not an immediate

safety concem. Results documented in this report were largely based on

conclusions from the Dominion Engineering Report (del-357) also funded unJer

Task 744.

50. June 1993, del-357 (6/4/93), " Dominion Engineering Report on Stress Analysis" -

This report documented the results of finite element analyses on CEOG CEDM

penetrations.

51

June 1993, NUMARC Letter to NRC (6/16/93)-Three PWR Owners Group's

safety assessments provided addressing Alloy 600 CRDM/CEDM VHP cracking

issue. NUMARC's conclusion was, "The reports confirm that the potential for

cracking does not pose an immediate safety concern."

52. July 1993, NRC/NUMARC Alloy 600 Nozzle Meeting (7/15/93), Rockville, MD

- 53. October 1993, Nozzle integrity Assessment Meeting (10/01/93), Charlotte, NC

-8-

.

- - - - - - -

- - -

- -

I

...

4

.

Appendix A

54. November 1993 NRC Letter to NUMARC (11/19/93)

55. December 1993, CEN 614 (12/30/93)" Safety Evaluation For OD Circumferential

Cracking" - This report, like CEN 607, was issued via NUMARC to the NRC. It

documented analyses showing that propagation of an OD crack in a CEDM/ICI

j

penetration would require from 80 to 100 years to grow to a point where structural

integrity of the penetration would be in jeopardy.

56. January 1994, NUMARC Letter to NRC (1/31/94) - The conclusion of this letter

was that "neither the potential for circumferential cracking nor the existence of

circumferential cracks pose an unreviewed or immediate safety issue." This letter

included revised safety assessments from each of the 3 PWR Owners Groups in

support of this conclusion.

- 57. February 1994, CE NPSD 905 P, Revision 1 (2/15/94),'CEDM Phase I,

Susceptibility Ranking" - Compared the properties, fabrication processes and

environmental conditions of CEOG CEDM/ICI nonles with nonles from foreign

plants which had experienced cracking.

58. March 1994, CE NPSD 927-P_ (3/30/94), " Stress Analysis Sensitivity Study" -

Compared the results of analyses with both nugget cooling and heat transfer

modeln of welding to address differences between WOG and CEOG safety

analyses. Concluded that CEOG method was appropriate and that the results

reached in CEN-607 were valid.

59. April 1994, CE NPSD 918 P (4/11/94), ' Phase 2, inspection Timing Model" - This

report supersedes CE NPSD-905-P relative to individual nonle timing for -

susceptibility to cracking and crack propagation.

60. April 1994, CE NPSD 919P (4/11/94), " Phase 2, inspection Strategy and Repair

Report" - Report identified an inspection strategy for CEOG member vessel head

penetrations, and repair requirements for shallow and deep cracks initiated from

-nonle ID locations.

- 61. April 1994 (4/28/94), NRC Inspection Report 94-10 (St. Lucie)

62. _ June _1994, CE NPSE-938-P, Revision 1 (6/10/94), ' Alloy 600 Bar Stock

Procurement - Material Specifications, Certified Test Reports & Inspection

Certificates'

'

9

!

-

.

-

-

-

.

..

.

Appendix A

63. June 1994, CE NPSE 948 (6/23/94), ' Leak Detection Methods Evaluation" -

Documented ABB review of available literature on leak detection methods,

including a report made available by the B&WOG on the same subject. Reported

that Nitrogen-13 detection systems showed the most promise.

64. July 1994, CE NPSD 947-P (7/13/94), 'PWSCC Miti ation Methods" - Report

0

evaluated several mitigation methods including weld overlay, shot peening, and

nickel plating as mitigation methods for CEDM/ICI cracking.

65. July 1994, EPRI TR 103696, "PWSCC of Alloy 600 Materials in PWR Primary

System Penetrations" - EPRI states, "It is important to note that none of the

Alloy 600 penetration PWSCC incidents which have occurred to date have posed

a significant safety problem at the plants involved. This is because most of the

cracks have been short and axial, and the laakage rates from the cracks have

been ' wy low...in summary,... cracking of Alloy 600 primary loop penetrations

does not pose a significant safety problem...The NRC has concurred with the

industry position that there is no immediate safety concern for cracking of

CRDM nozzles provided that visual inspections for boric acid leakage are

performed per Generic Letter 88.05." (emphasis added]

66. October 1994 NUREG/CR-6245

67. November 1994,3d EPRI PWSCC Workshop, Tampa, FL

-

68. November 1994, CE NPSD 949 P (11/28/94),' Evaluation of Boric Acid Corrosion

'

of RV Heads Resulting from Leaking CEDM Nozzles"- Concluded that undetected

leakage from cracks in adjacent CEDM nozzles could exist for almost nine years

before ASME code requirements for reinforcemont would be violated. A more

realistic case showed more than 15 years of leakage could exist. Report justified

l

that undetected leakage did not present an immediate safety concern.

69. January 1995 Petition Denial D.D.-95 2 (1/26/95)

70. August 1995, SONGS LER 3-95-001

71. October 1995, CE NPSD-1028 (10/3/95), " Fabrication of Ten Pressurizer Nozzle

Assemblies'

72. October 1995, CE NPSD-1017 (10/06/95)," Assessment of Grain Boundary

Carbide Distribution in Aitoy 600 CEDM and ICE Nozzles"

-10-

.

l

_ _ _ _ _ - _ _ _ _ _ _ - _ _ -

. _ _

_ _ . ,

,c.

~ ' :_*

Appendix A _

73.- November 1995, CEN 406-NP (11/2/95), 'A Status Report On CEOG Activities

Concerning Primary Water Stress Corrosion Cracking of inconel 600

Penetrations" (report sent to NRC from Palisades)

74. December 1995, CE NPSD-1019 (12/27/95), " Summary Report of Stress

Evaluation for a Deep Crack Repair of Alloy 600 CEDM Penetrations'

75.' April 1996 - NRC inspection Report 96-02 (SONGS) - Review and closure of

LER 3-95 001-00 on RCS nozzle leakage. Additionally, the inspector (s) evaluated

the acceptability of welding materials used on repairs of RCS nozzles and

identified inconsistencies with UFSAR tables. NCV on LER.

176. July 1996, CE NPSD-1032 (7/15/96), .*CEDM Repair Procedure"

-77. July 1996, CE NPSD 1013-P (Tl19/96), " Development of a Deep Crack Repair

Capability for Alloy 600 CEDM Penotrations'

78.' October 1996, SONGS LER 3 96 004 (10/23/96) - Reports leakage indications on

-

three pressurizer instrument nozzles found during a nozzle inspection at the

beginning of the Cycle 8 refueling outage. The cause was identified as PWSCC.

All Unit 3 pressurizer nozzles were inspected and four nozzles were replaced.

The outer portion of the Alloy 600 nozzle had been previously replaced with

Alloy 690 material,- but the weld filler maMrial was equivalent to Alloy 600. When

replaced, new filler material equivalent to Alloy 690 was used.

79. December 1996, WCAP 13929, Rev. 2 (12/9/96), ' Crack Growth and

Microstructural Characterization of Alloy 600 Head Penetration Materiais"

80. February 1997,4* EPRI PWSCC Workshop (2/25 27/97), Datona Beach, FL -In

St. Lucle's presentation, " EXPERIENCE WITH DETECTION AND REPAIR OF

PWSCC FLAWS IN PWR PRESSURIZER AND RCS LOOP ALLOY 600

PENETRATIONS AT ST. LUCIE UNIT 2," St. Lucie concluded: 1) the observed

cracks were determined stable by fracture mechanics; 2) stress analysis shows

cracking will be axial; and 3) ejection, confirmed by field observation, is unlikely.

They also concluded the only safety concern was the boric acid corrosion from

long term unidentified leaks which are being managed by inspection. -Therefore,

PWSCC nozzle cracking is not a safety issue; however, there are economic

concerns of unplanned repairs.

81. - March 1997, SECY 97-063 (3/18/97)

_

-11-

.

_ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ . . _ _ _

_ _ _ _ _ _ _ _ -

.

.

.i

Appendix A

82. April 1997 Generic Letter 97-01 (4/1/97)

83. April 1997, NRC Inspection Report 97-05 (SONGS) - The inspectors observed

work related to Alloy nozzle replacement, and found the work thoroughly

"ermed. The report discussed the details of the repair activities.

84. April 1997, SONGS LER 2 97-004 (4/2/97) - Reports leakage from the Unit 2

pressurizer, This leakage was found during a mode 4 walk down as part of the

unit's return to power following the Cycle 9 refueling outage. The outer portion of

the Alloy 600 nozzle was replaced with Alloy 690 material. PWSCC was identified

as the cause.

85. May 1997, SONGS LER 3 97-001 (5/9/97) - Reports leakage from five Unit 3

nozzles found as part of the initial walk down at the beginning of the Cycle 9

refueling outage. The outer portion of the Alloy 600 nozzle was replaced with

Alloy 690 material. The LER acknowledges PWSCC as the likely cause.

86. June 1997, NRC Inspection Report 97-09 (SONGS)- Reports the results of

resident inspector activities, including observations of nozzle replacement. The

inspectors noted the licensee identified the potential leakage in accordance with

established plans.

87. July 1997, NRC inspection Report 97-08 (SONGS)-ISI AND BORIC ACID

INSPECTION - The inspectors noted the Boric Acid control program was being

implemented in accordance with the established program. IFl 9501-01 related to

containment inspections on Boric Acid was also closed out.

88. July 1997, CE NPSD-1085 (7/20/97)'CEOG Response to NRC GENERIC

LETTER 97-01, ' Degradation of CEDM Nozzle And Other Vessel Closure Head

Penetrations - Provided the CEOG response to GL 97-01.

89. July 1997, SONGS LER 3 97-002 (7/30/97) - Reports leakage from four Unit 3

nozzles during the planned inspections as part of the units return to power at the

end of Cycle 9 refueling. The outer portion of the Alloy 600 nozzle was replaced

with Alloy 690 material. The LER acknowledges PWSCC as the cause and

credits SCE's inspect and replace program for finding these nozzles that weren't

found at the beginning of the outage.

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Appendix A

90. September 1997, NRC Inspection Report 97-15 (SONGS) - The report also

notes that though the Cycle 9 RFO, Unit 2 has experienced 4 nozzle cracks and

Unit 3,14 cracks, it was also noted that 2 heats experienced 4 cracks each. The

report also states there is no current nozzle replacement plan due to development

of in-house capabilities, and that these actions to develop the capabilities were

not started until the 3rd quarter 1996. Also, an apparent violation of 10 CFR 50

Appendix B, Criterion XVI for failure to implement actions to preclude recurrsnce,

was stated.

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