ML20198S632
| 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.
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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
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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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Power Operations -
L San Diego Gas & Electric Company
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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 -
- P.O. Box 942732
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Sacramento, California 94234
Mayor -
L City of San Clemente
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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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DRP Director
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G. Goines - RIV A.97-335 & DRS Al 97-G-0015'
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RIV File
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WCFO File'
G. Sanborn
J. Lieberman, OE (07H5)
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EDISON
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October 24,1997'
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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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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
"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
l
'
. .
.
.
_ _ _
_ ._ .
.
. , _
. ... _ , . . . _ _
_ __
_ _ _ _ . _ . _
_
. _ _ _ . . _ _
. . ,
'
-
,
~
.:
.
.,
,
--
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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4
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l
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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.
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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-
y
.
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
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
.
-
r
,-
.
- - -.
.,
..
--
.- - - - .. - .- _
.
-
.:
,
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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_ _ ,,s
e
.N,
,
- . - - . - - . - . .
,-_ - - - - , ,
____ ____ _ _____ __ _ ____
.
.
.
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'
.
-
..,....w
,
.,m.
. .
,
_ _ _ _ _
_
. .
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
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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