ML20199L866
| ML20199L866 | |
| Person / Time | |
|---|---|
| Issue date: | 02/03/1998 |
| From: | Callan L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Ralph Beedle NUCLEAR ENERGY INSTITUTE (FORMERLY NUCLEAR MGMT & |
| References | |
| NUDOCS 9802100050 | |
| Download: ML20199L866 (7) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 306e6-0001
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I February 3, 1998 Mr. Ralph E. Beedle Senior Vice President and Chief Nuclear Officer Nuclear Generation Nuclear Energy institute 1776 i Street, NW., Suite 400
. Washington, DC 20006-3708
SUBJECT:
NUCLEAR ENERGY INSTITUTE (NEI) 97-06, " STEAM GENERATOR PROGRAM GUIDELINES"
Dear Mr. Beedle:
Thank you for your letter of December 16,1997, enclosing a copy of NEl 97-06, " Steam Generator Program Guidelines" and apprising the staff that the NEl Nuclear Strategic issues Advisory Committee had voted to adopt the NEl 97-06 guidelines. The NRC staff supports and encourages industry initiatives that enhance the efforts of both the industry and NRC to maintain steam generator (SG) tube integrity, and we recognize the effort and resources that the industry has focused on this important issue. As you know, the staff is cunently focusing its resources toward issuing a proposed generic letter on SG tube integrity that (1) informs pressurized-water
+
reactor (PWR) licensees that actions beyond current technical specifications requirements are necessary to ensure SG tube integrity and (2) requests each licensee to implement either the actions described in the generic letter, or licensee-specified actions that ensure SG tube integrity is moniteted and maintained consistent with regulatory requirements and the plant licensing -
' oases. We plan to issue this proposed generic letter and an accompanying draft regulatory guide for public comment in the near future.
Ycur letter did not request the staff to review NEl 97-06, and the staff has not performed a
- detailed review of the subject document. However, a cursory review of NEl 97-06 rev'eals some significant differences between guidance in proposed draft regulatory guide DG-1074 " Steam Generator Tube Integrity" and NEl 97-06. As an example, the structural integrity performance criterion and the accident-induced leakage performance criterion in NEl 97-06 differ significantly
'p from the c;,mria in the draft regulatory guide. Moreover, based on our cursory review, the two
[
eriteria, as stated in NEl 97-~06, may not ensure compliance with current regulations, in this regard, I recommend that PWR licensees carefully assess the NEl 97-06 guidance and ensure
/k that implementation of the guidance would be consistent with 10 CFR 50.59 to ensure that they continue to maintain and operate their facilities such as to comply with the currant regulations (10 CFR Part 50, Appendix A, " General Design Criteria" and 10 CFR Part 50 Appendix B) and 3
applicable SG technical specifications. In this respect, the staff encourages licensees to utilize ina001 9902
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Ralph E. Beedle risk information and insights wherever possible to improve their safety decisions when making changes to the current licensing basis, particularly given the safety significance of SG tube integrity.
If NEl decides to submit NEl 97-06 for staff review, possibly In response to the request for public comments on the proposed generic letter and draft regulatory guide, the staff would perform a 2
more detai'ed r wiew of the subject document, in this regard, the staff is willing to meet with NEl ani! discur,s NEl 97-06 in more detail to support such an effort.
Sincerely, seph Callan Ex r.cutive Director for Operations L
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e Ralph E. Becdle risk information and insights wherever possible to improve their safety decisions when making changes to the current licensing basis, particularly given the safety significance of SG tube integrity.
If NEl di ; ides to submit NEl 97-00 for staff review, possibly in response to the request for public comments on the proposed generic letter and draft regulatory guide, the staff would perform a more detailed reviev of the subject document. In this regard, the staff is willing to meet with NEl and discuss NEl 97-06 in more detail to support st.:h an effort.
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Sincere %1nal Signe'lby L. Joseph Cdifah Executive Director for Operations DISTRIBUTION:
Public EMCB RF Central Files EMurphy AThadani PNorry GHolahan SLong JFlack TCollins HThompson KBohrer (GT97 876)
JDonoghue JBlaha SBurns DE R/F (97-28)
Ciiairman Jackson Commissioner Dicus Tehnical Editor Rav Sanders 1/7/98 Commissioner Diaz DOCUMENT NAME: A:\\EDONEl.198 Commissioner McGaffigan
'See Previous Concurrence To receive a copy of this doc.ument, indicate in the box C = Copy v.to attachment /enciosure E = Copy with attachment / enclosure N = No copy OFFICE
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2-Ralph E. Beedle to utilize risk information and insights wherever possible to improve their safety decisions when making changes to the current licensing basis, particularly given the safety significance of SG tube integrity, if NEl decides to submit NEl 97-06 for staff review, possibly in response to the request for public comments on the proposed generic letter and draft regulatory guide, the staff would perform a more detailed review of the subject document. In this regard, the staff is willing to meet with NEl and discuss NEl 97-06 in more detail to support such an effort. Nevertheless, we must caution that implementation of the NEl 97-06 guidelines by industry may not ensure that current regulations are being met.
Sincerely, L. Joseph Callan Executive Director for Operations DISTRIBUTION:
Public EMCB RF Central Files EMurphy AThadani PNorry GHolahan SLong JFlack TCollins HThompson KBohrer (GT97-876)
JDonognue JB!a',a SBurns DE R/F (97-28) lehnical Editor Rav Sanders 1/7/98 DOCUMENT NAME: A:\\EDONEl.108
- See Previous Concurrence To receive a copy of this d.)cument, Indicate in the box C-Copy w/o attachment / enclosure E = Copy with attachment / enclosure N = No copy OFI'CE
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Ralph E. Beedle to utilize risk information and insights wherever possible to improve their safety decisions when making changes to the current licensing basis, particularly given the safety significance of SG tube integrity, if NEl decides to vise NEl 97-06 to be more consistent with discussions that the staff and industry have had ver the past few years, and submits the document for staff review, possibly in response to the r uest for public comments on tha proposed generic letter and draft 4
regulatory guide, the taff would be willing to review the subject document, in this regard, tne staff is willing to meet ith NEl end discuss NEl 97-06 in more detail to support such an effort.
The NRC staff apprecia s being apprised of your efforts concerning SG tube integrity and encourages a continuing ank and open technical dialogue with the NEl and industry on this important matter.
Sincerely, L. Joseph Callan Executive Director for Operations DISTRIBMIlRN:
Public EMCB RF Central Files Murphy AThadani PNorry GHolahan SLong JFlack -
ollins HThompson KBohrer (GT97-876)
JDonoghue JBlahu SBurns DFgg28)
. sanders 1/7/28 Tehnical Editor. Bay %\\ REED \\EDONEl.198 DOCUMENT NAME.
To receive a copy of this document, indicate in the box C = Copy o attachment / enclosure E = Copy with attachment / enclosure N = No copy N/F/9'l-/cm.
OFFICE EMCB:DE, C
tA)SC:EMCB:DE 6 (A)BC:EkB:DEC (A)DD:DE @',f Lk JStrosr'deff 6k NAME TRe sao ESullivan l/4 DATE
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EDO Principal Correspondence Control FROM:
DUE: OJ/42/7h EDO CONTROL: G970876 DOC DT: 12/16/97 FINAL' REPLY:
Ralph E. Beedle Nucicar Energy Institute (NEI)
TO:
Callan, EDO FOR SluNATURE OF :
- GRN CRC NO:
4 DESC:
ROUTING:
STEAM GENERATOR PROGRAM Callan 4
Thadani Thompson Norry Blaha Burns DATE: 12/22/97 ASPIGNED TO:
CONTACT:
_NRR Collins SPECIAL INSTRUCTIONS OR REMARKS:
For Appropriate Action.
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December 10,1997 3
Mr. L. Joseph Callan Executive Director for Operations U.S. Nuclear Regulatomy Commission Mail Code 5 EG Washington, DC 20555 0001
Dear Mr. Callan:
NEI's Nuclear Strategic Issues Advisory Committee recently voted to adopt the following formal industry position:
P Each licensee will evaluate its existing steam generator prograin and, where necessary, revise and strengthen program attributes to meet the intent of the guidance provided in NEI 97 06, Steam Generator Program Guidelines, no later than the first refueling outage starting after January 1,1999.
The objectives of the initiative include:
Providing a clear commitment from utility executives to follow the industry steam generator inspection and maintenance guidance provided through the EPRI Steam Generator Strategic Management Project; Assuring a unified industry approach to emerging steam generator issues; e
and Applying tube integrity performance criteria in vinjunction with the performance based philosophy outlined in the maintenance rule,9 50.05.
Due to concerted efforts both by the industry and the NRC, steam generator tube rupture frequency has decreased over time even with increasing levels of tubo degradation. This steam generator initiative builds upon these efforts and provides performance criteria which will be used to monitor and maintain performance. We EDO -- G970876
.1776 't 51tilf, NW
$Utti 400 WA$HINGTON, DC 20006-3708 PHONE 202.739 8088 i Ax 20'.793.18 98 i
Mr. Jouph C:ll:n December 16,1997 Page 2 l
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believe that adopting this initiative enhances industry's efforts in meeting out mutual goal of assuring steam generator tube integrity.
[
r Enclosed is a copy of NEI 97 00.
l Please call me if you have any questions regarding the industry initiatives Sincerely, h,0w.fo Ralph E. Beedle RCC/tmc Enclosure c:
Document Control Desk (w/ enclosure)
Dr. Brian Sheron (w/ enclosure)
Jack Strosnider (w/3cnclosures)
Ted Sullivan (w/3 enclosures)
Emmett Murphy (w/ enclosure)
Tim Reed (w/enclourc) i
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NEI 97 06(Original)
December 1997 AClul0WLEMMEllTS The Nuclear Energy Institute (NEI) Task Force on Steam Generator Programs developed the Steam Ger.crator Program Guldcline with oversight provided by the NEl Steam Generator Issres Working Group. We appreciate those industry contributors who reviewed and commented on this document to improve its technical content and its clarity.
4 NET also wishes to thank the Electric Power Research Institute (EPRI). EPRI, through the Steam Generator Strategic Management Project, developed the steam generator guidelines referenced in this document.
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Il0TICE
. Meither NEl, nor any ofits employees, members, supporting organizations, contractors, or consultants make any warranty, erpn ssed or implied, or assume any legal responsibility for the accuracy or completeness of, or as.ane any liability for damages resulting from any use of, any information apparatus, methods, or process disclosed in this report or that such may not infringe privately owned rights.
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NEl 97 06 [ original]
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i Nuclear Energy Institute i
i Steam Generator i
l Program Guidelines l
December 1997 1
Nuclear Energylecritute,1776 i Street N.ll'., Suite 400, If'ashington D C (202.739.8000)
L NEl 97-06 (Original)
December 1997 EXECUTIVE
SUMMARY
NEl 97-06 establishes a framework for structuring and strengthening existing steam generator programs. Provided here are the fundamental elements expected to be included in a steam generator program. These elements incorporate a balance of prevention, inspection, evaluation, repair and leakage monitoring measures.
This guideline refers licensees to EPRI guidelines for the detailed development of these programmatic attributes. EPRI will maintain these guidelines through the Steam Generator Strategic Management Project (SGMP) consensus process. Revisions to the EPRI documents will follow the pruocol as noted in Section 1.5 of this document. Other industry orgar.irations, such as NSSS Owners Groups, may also develop guidelines for implementing steam generator program elements. Development and revision of referenced documents shall follow a protocol similar to the EPRI s tocol.
ro The intent of this document is to bring consistency in application ofindustry guidelines relative to managing steam generator programs. This document and those it references recognize the need for flexibility within each plant specific program to adjust for the degree of degradation experienced and expected improvements in techniques for managing tube degradation.
Section 1, " Introduction," provides a background, discusses regulatory interface, utility responsibilities, and ; rotocol for revision of the referenced EPRI guidelines.
Section 2,"Perfomiance Criteria," dermes the performance criteria that utilities shall use to measure tube integrity. Meeting the performance criteria provides reasonable assurance that the steam generator tubing remains capable of fulfilling its intended safety function of maintaining RCPU integrity.
Section 3, " Steam Generator Program," discusses the program elements and implementing guidance for strengthening existing steam generator programs.
Section 4," Reports to NRC," lists required reports, i
i
NEl 97-06 (Original)
December 1997 i
[This page intentionally left blank.)
il
NEl 97-06 (Original)
December 1997 TABLE 0F CONTENTS
.....................................................................I Executive Summary..........................
- 1. l HT R O D U CTIO N................................
PURrOsE...........................................................................................
1.1
.................................1 1.2 HACKGROUND...............................................................................
U T I LI TY R ES PO N S I a l LI T I ES...............................................................................
1.3 R EG U L AT O RY R E Q O ' RE M ENTS.................................
1.4 1.4.1 10 CFR Part 50 Appcndix A, General Design Criteriafor Nuclear Power Plants, and Appendin U, Quality Assurance Criteriafor Nuclear Power Plants an d FueI Reprocessing Pla nts..........................................................................
1.4.2 10 C F R Q 50.65, Malnten an ce R ule..................................................................... 3 1.4.310 CFR @ 50.72, hnmediate Notification Requirementsfor Operating Nuclear Power Reactors, and 60.73, Licensee Event Report System............................
1.4.4 Plant Technical Specifications for Primary-to-Secondary Leakage...............
PREPARATION AND REVIStos PROTOCOL TOR EPRI G UIDE 1.5
ST RUCTURAL I NTEGRITY PERr0RM AN CE CRITEIUON......
2.1 ACCIDENT-1NDUCED LEAKAGE PCRFORM ANCE CRITER10N................................
6 2.7 OPERATION AL LEA KAG E P err 0RM AN CE C RITERION...
2.3
ASSESSMEN1 OF POTENTI AL DEGRADATION MECil ANISMS...................................
7 l
l 3.1 i
............................................................................8 3.2 I N S P E CT I O N........................................
iii L
c 3
NE! 97 06 (Original)
December 1997 3.3 T u n E F NT E G R I T Y A S S E5 5 51 E N T......................................................................................... 8 3.3.1 R e p a i r L i m i t s......................................................................................................... 9 3.4 5 f A I NT EN A N C E A N D R E r A I Rs......................................................................................... ! O 3.5 PRI hI A R Y-TO S ECO N D A R Y L E A KA G E hI ONI1 O H I N G...................................................... 10 3.6 S E CON D A R Y-S I D E WATE R CII Est I STRY.......................................................................... I 0 3.7 PRI 31 A RY-S I D E WAT E R Cli 051 ISTRY............................................................................... I 1 1
3.8 FORElGN h1 ATFRI AL ExcE-
.N...................................................................................12 3.8.1 S ec o n d a ry-S id c Vis u a i I n s p ec t l o n.................................................................... I 2 3.8.2 Control and blonitoring of Foreign Objects and Loose Paris........................12 3.9 51 AINTENANct Or STEA31 GENERATOR SECONDARY SIDE INTEGRITY........................12 3.1 0 S E L F A s s Es s 31 E N T............................................................................................................ I 3 4 R E P O RTS T O N R C.................................................................................
APPENDIXES -
A.RErERENCE5..........................................................................................................................A-1 B. DEFINITIONS..........................................................................................................................B-I C.AcRONYAIS...........................................................................................................................C-1 D. 12 -h ! O NTI I R E r0 RT CG N T E NT.............................................................................................. D - 1 1
i IV
NE! 97-06(Original)
December 1997 l
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- 1. INTRODUCTION
1.1 PURrost
The purpose of this doctiment is to bring consistency in applicat t
related to managing steam generator program ilities shall measures. Additionally, this document establishes performance criteria tha use under the maintenance rule.
1.2 BACKGROUND
The program elements described in this document are eviden i
es commitment to safe and reliable steam generator operation. These ele d
d the relative to the management and repair of steam generator tubing. For o industry has expended considerable resources developing guidance di generator programs to meet the challenges posed by continuing tube Chemistry control is an example of the industry's commitment to th-l i
management of steam generator degradation. By the mid 1970s, ut i
The dominant tubes at a rate that would exceed steam genera d i to all-volatile water chemistry control. This, however, resulted in conditions i
lt of corrosion of the carbon steel support plates, which led to tubing defonna denting and cracking with the same unacceptable rate of tube plu working through EPRI, met these challenges by implementing steam d
des in with aggressive improvements in control of secondary-side water che The secondary side equipment, thus essentially e ll'ater Chemistry Guide /Ines and associated supporting documents.
These chemistry guidelines have proven to be the cornerstones o maintain acceptable steam generator performance. Over time, the indu h
generator programs have matured to include i d
d ti RI Steam specific management. Building on the collecti guidelines, to incorporate technological and programmatic improvement 1.3 UTILITY REsroNstattiTits Each utility shall adopt the performance criteria contained in Section i
l criteria are (1) Structural Integrity, (2) Accident Icluced Leakage and (3' t
NEl 97 06(Original)
December 1997 Leakage. Further, each utility shall evaluate existing program elements against those described in Section 3 and revise and strengthen, where necessary, to meet the intent of this doce ent an j the referenced EPRI guidelines.
The steam generator program shall include the following elements:
assessment of potential degradation mechanisms
=
4 ins 3Xction e
integrity assessment e
maintenance and repairs a
primary-to-secondcry leakage monitoring a
secondary side water chemistry a
primary-side water chemistry e
maintenance of secondary-side integrity a
a selfassessment NRC reporting a
Section 3 further discusses these program elements.
1.4 RtcutATony REQUIREMENTS The followir$g section addresses NRC requireinents that licensees should include in the l
development and implementatior. of the plant-specific steam generator program.
1.4.1 10 CFR Part 50 Appendix A, GeneralDesign CriterlaforNuclear Power Plants, and Appendix B, Quality Assurance Criterlafor Nuclear Power Plants and Fuel Reprocessing Plants.
General Design Criteria (GDC) 1,2,4,14,30,31 and 32 of 10 CFR Part 50, Appendix A, define requirements for the reactor coolant pressure boundary (RCPB) with respect to structural and leakage integrity. Steam generator tubing and tube repairs constitute a maior fraction of the RCPB surface area. Steam generr. tor tubing and associated repair techniques and components, such as plugs and sleeves, must be capable of maintaining reactor coolant inventory and pressure.
r 10 CFR 50, Appendix B, establishes o,uality assurance requirements for the design.
^
construction and operation of safety-related components. The pertinent requirements of this appendix apply to all activities affecting the safety-related functions of these components; these include, in part, inspecting, testing, operating and maintaining. Criteria IX, XI, and XVI of Appendix B apply to the steam generator tube integrity program.
2 8-
7, NEl 97 06 (Original)
December 1997 1.4.2 10 CFR $ 50.65, Maintenance Rule Under the maintenance rule, uti!! ties classify steam generators as sa components because they are relied on to remain functional du basis events. The performance criteria in Se Rule through the performance of appropriate preventive maintenance"(Main program that provides the appropriate prev of die maintenance rule.
Steam generators are to be monitored under {(a)(2) of the main industry-established performance criteria. If the performance criteria d
cause determination of appropriate depth sha 3.3," Integrity Assessment," provides guidance for cause determina determination should identify the cause of the fadure or unacceptable and whether the failure was a maintenance preventable functional failur NUMARC 93 01 [ Reference 1J offers guidance for implementing the rule should a utility elect to incorporate additional monitoring gcals scope of this document.
1.4.3 10 CFR % 50.72,immediate Notification Requirementsfor Ope Power Reactors, and % 50.73, Licensee Event Report System Failure to meet the performance criteria should be assessed to in degradation of safety barriers. If so, the reponing requirem and {50.73(a)(2)(i) or (ii) should be reviewed to determine applicabi r
!.4.4 Plant Technical Specificaticus for Primary to-Secondary Leakage Plant technical specifications include a requirement to shut down t
exceeds an established threshold of primary to-secondary leakage.
I 1.5 PREPARATION AND REVISION PaoTocotronEPRI GUIDELINE Some of the EPRI guidelines referenced herein are directive in n d
meet the intent of the guideline. Other EPRI guidelines are non-dire I-i be used by utilities as general guidance for structuring steam genera The protocol noted here applies to the directive EPRI guidelines.
3
s NE! 97 06 (Original)
December 1997 At an interval not to exceed two years, the EPRI Nuclear Power Council (NPC) will convene a utility committee (s) to review the applicable EPRI guideline document to determine the need for revision.
Committee members include utility personnel, supplemented, as appropriate, by consultants, NSSS vendor and other supplier and/or senice vendor personnel, all with equal voting rights. The members will have expertise relevant to the particular area being addressed. These committees are responsible to, and under the charter of, a utility sponsor group that broadly represents the management of the plants to which the prepared guidance is applicable. T1.cie will be an EPRI staff member on the committee, usually the chairperson, wlio will be a non voting member. The NPC will approve the membership on the committees.
The directive guidelines referenced herein are:
PIl'R Stcam Geneo ator Examination Guidelines, [ Reference 2);
a Pil'R Primary to-Secondary Leak Guidelines, lReference 3};
Pll'R Secondary Il'ater Chemistry Guidelines, (Reference 4}; and PIl'R Primary Il'ater Chemistry Guidelines, (Reference 5}.
a The requirements in the directive EPRI guidelines represent a consensus of the committee and are experience-based in that they.. e achievable with available technology.
Requirements will be incorporated into the EPRI guideline documents when it has been successfully demonstrated that the requirement can be applied in operating plants.
Once the committee prepares a final draft, it is circulated for broad industry review. The committee then resolves all comments generated as a result of the review and prepares a final document to be approved and issued by the sponsor group.
The non directive, how to EPRI guidelines will be revised in a manner deemed appropriate by the committee originating the guideline. If at some point in time these documents become directive in nature, they will be handled via the formal committee approach described above.
The non-directive guidelines referenced herein are; Steam Generator Tube Integrity Assessment Guidelines, (Reference 6];
=
In-Situ Pressure Testing Guidelines, lReference 7);
Pll'R Steam Generator Tube Plug Assessment Document, \\ Reference 8}; and a
Pil'R Sleeving Assessment Document, (Reference 9].
a When a committee revises a directive EPRI guideline, utilities will modify their steam j
generator programs accordingly. Utilities should reflect program revisions in plant procedures for the upcoming refueling outage if that outage is greater than six months 4
I NEl97 06 (Original) l December 1997 I
ility may delay away. If the next refueling outage is less than six months away, the ut incorporating appropriate changes until the following refueling outage.
- 2. PERFORMANCE CRITERIA d
i st The steam generator perfonnance criteria described below identify th i
which performance is to be measured. Meeting the lli it speciFe safety ftmetion of maintaining RCPB integrity.
li
- ity, Performance criteria used for steam generators shall be based on l
accident induced leakare, and operational leakage as defined be ow.
2.1 STRUCTURALINTEGRITY PERI'ORMANCE CUJTERION The structural integrity performance criterion is to:
Ensure steam generator tubes will analntain adequate margin i
ycle.
under normal andpostulated accident conditionsfor the operat ng c ce The structural performance criterion is based or, ensuring that l
ostulated accident that a steam generator tube will not rupsure during norma or p il l ents to meet this conditions. Section 3.3 of this guideline establishes the essent a e em performance criterion.
ff The EPR1 Steam Generator Tube Integrity Assessment Guide id i s used to guidance for the evaluation methods, margin, an i s of istent with the safety against gross failure or rupture of the tubing. These margins a S ciety of safety factor margins implicit in the stress limit criteria of the Amencan Mechanical Engineers (ASME) Code.
b bilities The probabilistic method of Reference 6 is based on satisfyin i
i l ding known and ofinduced tube rupture due to postulated accident loading condit ons, nc li it re:
unknown degradation mechanisms. The conditional probability m s a per year that one tube ruptures during an accident,
<5 x 10 2
<2.5 x 10', per year that two to 10 tubes rupture during an accide a
a h f rmation of a primary to-secondary opening h
1 A tube rupture or burst is a gross failure of the tube such t at t e o t
with an area affiliated to that of a double-ended guillotine brea occurs.
I l-5
mm NEl 97-06 (Otiginal)
Decemb,tr 1997 0
a
<1 x 10 per year that more than 10 tubes rupture during an accident.
The conservatism in these limits provides reasonable assurance for meeting the stated performance criterion.
2.2 ACCIDENT-INDUCED LEAKAGE PERIORM ANCE CRITERION The accident induced performance criterion is to:
Ensure that the potentialprimary-to-secondary leak rate during limiting postulated events will; not exceed the total normal makeup capacity of the prirnary coolant system; a
and not exceed the offsite radiological dose consequences, per 10 CFR Part 100 a
guidelines, and the radio!ogical consequences to control room personnelper GDC-19.
The projected leakage of deraded steam generator tubes following an accident must not result in the associated radioactiQ.y releases to the environment exceeding specified values. This criterion is specified in terms of dose to the maximum exposed individual offsite and dose to control room operators. Potential accident induced leakage is defined in a plant's licensing basis. Changes to the licensing basis require appropriate regulatory reviews.
As a defense in-depth measure, a maximum leakage limit is additionally imposed in this performance criterion. The projected leakage must not exceed the normal reactor coolant l
system makeup capacity. As used in this document, normal makeup capacity refers to the ability of the makeup system to maintain reactor coolant system inventory without manual i
or automatic actuation of engineered safeguards features. Manual starting of redundant or standby pumps may be credited as normal makeup capacity if plant procedures provide for the additional pumps, in lieu of calculating offsite and control room doses for each condition monitoring and operational assessment evaluation, utilities may determine a maximum allowable accident leakage limit based on plant-specific raeteorology, dose-equivalent iodine limits, and iodine spiking considerations. This leakage limit would result in offsite and control room doses that meet the performance criterion and that are within normal makeup capacity. Once determined, the maximum allowable accident leakage would serve as a derived criterion against which the postulated accident induced leak rate would be compared.
j 6
NEl 97-06 (Origin:3)
December 1997 2.3 OPERATIONAL LEAKAGE ptRTORhlANCE CRITERION The operational leakage perfo.mance cri'erion is to:
Ensure that the operationalprimary-to-secondary leakage limitfor any steam generator does not exceed 150 GFD.
Plant technical specifications include a requirement to shut dow l k period when the plant exceeds an established threshold of prim He primary to secondary leaky,e limit referred to in the perfo I
gallons per day (GFD) at room temperature conditions from an l t addition, since some types of tube degradation may progagate rapidl h
should initiate shutdown when the leak rate increase ex d (e.g., steam period. This measurement should be confin
- 3. STEAM GENERATOR PROGRAM ld The purpo,e of a steam generator program is to ensure tube k
itoring contain a balance of prevention, inspection, evaluation and repair, and The major measures. Licensee: shall document the program through plant proced program elements a e discussed below.
3.1 ASSESSMENT
OF POTENTI AL DEGRADATION. f ECil ANISM 4
Licensees shall perfonn an assessment of both existing and poten i hin the mecht.nisms. The assessmem shall address the reactor coolant pr steam generator, e.g., plugs, sleeves, tubes a
l ider experience from other similar steam generators. Tne assessment sh engineeririg analysis of the degradation mechanisms.
The parpose o' the assessment is to identify degradation mec mechanism ic'entified:
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choose techniques to test for that degradation based on the proba a
sizing capability; establish the number of tubes to be inspected; a
establish the structural limits; and a
i establish the flaw growth rate.
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NEl 97 06 (Original)
I December 1997 The identification of these parameters allows a utility to establish the inspection or repair criterion before an outage. If a plant identifies a new degradation, or if the measured parameters change, such as growth rate, the plant may need to adjust analytical parameters i
during an inspection as the condition monitoring or operational assessment dictates.
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The assessment of potential degradation mechanisms affects both the inspection and structural components of the program. The inspection component identifies the technique's capability, including detection probability, sizing capability, and measurement uncertainty.
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It will al>o identify the sampling strategy. The structural component applies the information gathered from the inspection with flaw growth rate projections to establish the repair limit and/or cycle length.
To conduct an effective inspection, the utility should integrate the structural and inspection components. EPRI Steam Generator Tube Integrity Assessment Guidelines (Reference 6}
and EPR1 PIl'R Steam Generator Examination Guidelines (Reference 2} provide guidance for assessment of potential degradation mechanisms.
3.2 INSrtcTioN Each utility shall plan inspections according to the expected tube degradation and follow the inspection guidelines contained in the latest revision of the EPRI PWR Steam Generator Examination Guidelines (Reference 2}.
Some of the important features include:
sampling using either a prescriptive approach or a performance based approach
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obtaining the information necessary to develop degradation assessments, e.g., condition
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monitoring and operational assessments qualifying the inspection program by determining the accuracy and defming the a
elements for enhancing system performance, including technique, analysis, field analysis feedback, human performance and process controls 3.3 Time INTrcnITY ASSESS % LENT Licensees shall assess tube integrity after each steam generator inspection. The purpose of the integrity assessment is to ensure that the performance criteria have been met for the previous operating period (e.g., condition monitoring), and will continue to be met for the next period (e.g., operational assessment). The EPRI Steam Generator Tube Integrity Assessment Guideline (Reference 6] offers guidance for the evaluation methods, margins, and uncertainty considerations used to determine tube integrity.
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NEl 97-06 (Original)
December 1997 The choice of an evaluation method to verify tube integrity will depend on the uncertainty surrounding the particular degradation being assessed which can be highly dependent on the availability of data. Utilities may use activities such as m situ pressure testing or pulling tubes to supplement the tube integrity analysis. Reference 6 provides guidance as to when to conduct in situ pressure testing to address past operating period performance. The EPRI In-Situ Prc33ure Testing Guidelines [ Reference 7] provide guidance on screening criteria for candidate tube selection, as well as for test methods and testing parameters.
If a licensee determines that the structural integrity or accident leakage performance criteria have not been oatisfied during the prior operating period, an evaluation of causal factors for failing to meet the criteria shall be performed, in this event, the licensee is required to notify the NRC as discussed in Section 4.
For an unscheduled inspection due to primary to secondary leakage, the tube integrity assessment need only address the degradation mechanism that caused the leak, provided the interval between scheduled inspections is not lengthened.
Licensees shall complete a tube integrity assessment for the next operating cyc'e within 90 days after startup. Completian of this assessment may not be possible due to the complexity of the analysis within the 90 day period. In this case, a preliminary assessment is acceptable r; an interim measure. There should be reasonable assurance that the performance criteria w'll not be exceeded prior to completing and submitting the final assessment.
Tl e completed assessment shall be submitted to the NRC within 12 months following the completion of the insenive inspection. Appendix D provides the content of this assessment report.
3.3.1 REPAIR LI511TS Licensees shall establish tube repair limits for each active c. gradation mechanism.
Tube repair criteria shall be either the existing technical specification through-wall (TW), depth-based criteria (i.e.,40% TW for most plants), a voltage-based repair limit per Generic Letter 95-05, or other alternative repair criteria (ARC). Iflicensees choose to develop and implement an ARC, they should follow a steam generator degradation specific management (SGDSM) strategy. Reference 6 provides guidance for developing an ARC.
For plants experiencing a damage form or mechanism for which no depth sizing capability exists, tubes identified with such damage are " repaired / plugged-on-detection" and integrity should be assessed. Note: " Plug on-detection"is not
' considered an ARC.
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NEl 97-06 (Original)
Decernber 1997 3.4 MAINTENAscE AND REPAins Licensees shall qualify and implement repair methods in accordance with industry standards. The qualification of the repair techniques shall consider the specific steam generator conditions and mockup testing. The purpose of the repair is typically to remove degraded tubing from service, thereby redefining the reactor coolant pressure boundary.
Licensees shall clearly identify engineering prerequisites and plant conditions prior to performing the repair. Process controls shall be identified to ensure proper performance of the repair including the consideration ofpost maintenance testing. Additionally, licensees shall perform a baseline inspection of the repair consistent with the latest revision of the EPRI Pli'R Steam Generator Examination Guidelines [ Reference 2).
The EPRI Pil'R Ste: n Generator Tube Plug Assessment Document [ Reference 8} and the EPRI Pil'R Sleeving Assessment Document [ Reference 9] provide further guidance for maintenance and repair of tubing.
3.5 PnistAny-To-SEcoNnAny LEAKAGE MONITORING Licensees shall establish primary to secondary leakage monitoring procedures in accordance with the EPRI Primary to Secondary Leak Guidelines (R:lerence 3). Licensees shall initiate plant shutdown in a controlled and timely manner prior to exceeding 150 gallons per day (GPD) or when the leak rate increase exceeds 60 GPD in any one hour period. This measurement should be confirmed using a qualitative method (e.g., steam generator blowdown radiation monitors, main steam line monitors, etc.. ).
Primary-to secondary leakage monitoring is an important defense-in depth measure that assists plant staffin monitoring overall tube integrity during operation. Monitoring gives operators information needed to safely respond to situations in which tube integrity becomes impaired and significant leakage or tube failure occurs. Additionally, operational leakage is an important tool for assessing the effectiveness of a steam generator program. Plants shall assess any operational leakage to determine if the leakage is expected or unexpected.
Appropriate training shall be provided for personnel who respond to primary-to-secondary leakage events.
3.6 SEcoNoARY-SIDE WATER CIIE511STRY Each utility shall have procedures for monitoring a:.d controlling secondary side water chemistry to inhibit secondary side corrosion..iduced degradation in accordance with the EPRI Pil'R Secondary ll'ater Chemistry G ddelines [ Reference 4].
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NEl 97-06 (Original) f December 1997 This program should establish, as a minimum:
control parameters; a
a sampling schedule for the control parameters and action levels for these parameters;
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ocedurcr used to measure the values of the control parameters; a
econdary sample points, including monitoring the discharge of the condensate pumps
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for evidence of condenser in-leakage; proccdures for the recording and management of data; a
procedures for defining corrective actions for exceeding control parameter action levels;
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and procedures identifying the authority responsible for the interpretation of the data and the a
sequence and timing of administrative actions required to initiate corrective action.
3.7 PRl51ARY-SIDE WATER CIIE511STRY Each utility shall have procedures for raonitoring and controlling primary-side water chemistry to inhibit primary-side corrosion-induced degradation in accordance with the EPRI PWR Primary Water Chemistry Guidelines [ Reference 5}.
This program should establish, as a minimum:
control parameters; a
l a sampling senedule for the control parameters and action levels for these parameters; a
procedures used to measure the values of the control parameters; a
primary sample points; a
procedures for the recording and management of data; a
procedures for defining corrective actions for exceeding control parameter action levels; a
and procedures identifying the authority responsible for the interpretation of the data and the a
sequence and timing of administrative actions required to initiate corrective action.
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i NE! 97 06 (Original) l December 1997 3.8 FOREIGN M ATERI AL EXCLUSION I
Each utility shall have procedures to monitor for loose parts and control of foreign objects to inhibit fretting and wear degradation of the tubing. This program should include the attributes below.
3.8.1 Secondary-Side VisualInspection The program should define when such inspections are to be performed, the scope of inspection, and the inspection procedures and methodology to be used. Loose parts or foreign objects that are found should be nemoved from the steam generators, unless it is shown by evaluation that these objects will not cause unacceptable tube damage. This evaluation should be maintained as part of the inspection record.
Tubes found to have visible damage should be inspected non-destructively and plugged or repaired if the repair criteria are exceeded.
3.8.2 Control and Monitoring of Foreign Objects and Loose Parts The program should ine'ude procedures to preclude the introduction of foreign objects into either the primary or secondary side of the steam generator whenever it is opened (e.g., for inspections, repairs, and modifications).
Such procedures should include, as a minimum:
detailed accountability for all tools and equipment used during an operation;
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appropriate controls and accountability for foreign objects such as eyeglasses e
and film badges; cleanliness requirements; and a
accountability for components and parts removed from the intprnals of major
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components (e.g., reassembly of cut and removed components).
Utilities should have alarm response procedures for the loose part monitoring system.
3.9 M AINTENANCE OF STEAM GENERATOR SECONDARY-SIDE INTEcRITY Secondary-side steam generator components shall be monitored if their failure could prevent the steam generator from fulfilling its intended safety-related function. The monitoring shall include design reviews, an assessment of potential degradation mechanisms, industry experience for applicability, and inspections, as necessary, to insure degradation of these 12 I
NE! 97-06 (Original)
December 1997 components does not threaten tube structural and leakage integrity or the ability of the plant to achieve and maintain safe shutdmm.
3.10 SELF ASSESSMENT Licensees shall perfonn self assessments regarding the steam generator management program. This review shall be perfo:med by knowledgeable utility personnel or a contractor with independent experts selected by the utility on a periodic '> asis. An INPO assessment can be used as an adjunct to the self assessment. The self assessment should identify areas for program improvement, along with program strengths. The assessment shall include all of the essential program elements described in Section 3 above.
- 4. REPORTS TO NRC In addition to the utility-specific technical specification reporting requirements, the following reports are required:
Information to be reported Report period Failure to meet a performance criterion during in acordance with time frames tube integrity assessment specified in 10 CFR 50.72/73 The number of tubes plugged in each steam 15 days after completion of the generator insenice inspection Information contained in Appendix D 12 months after completion of the insenice inspection 13
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NEl 97-06 (Original)
December 1997 APPENDIX A References
- 1. NUMARC 93-01, Industry Guidelinesfor Afonitoring the Effectiveness ofAfaintenance at Nuclear Power Plants, (May 1993).
- 2. Pil'R Steam Generator Examination Guidelines, EPRI Report TR-107569 (Rev. 5, September 1997)
- 3. Pll'R Primary-to-Secondary Leak Guidelines, EPRI Report TR-104788 (Rev.0, June 1995)
- 4. Pli'R Secondary li'ater Chemistry Guidelines, EPRI Report TR-102134 (Rev. 4, December 1996)
- 5. Pil'R Primary li'ater Chemistry Guidelines, EPRI Repot1 TR-105714 (Rev. 3, December 1995)
- 6. Steam Generator Tsbe Integrity Assessment Guideline, EPRI Report TR-107622, (Rev 0, September 1997)
- 7. In situ Pressure Testing Guidelines, EPRI Report TR-107620 (Rev. O, October 1997)
- 8. PIl'R Steam Generator Tube Flug Assessment Document, TR-10XXXX (Rev. O, to be issued about December 1997)
- 9. EPRI Pil'R Sleering Assessment Document TR-105962, (Rev 0, December 1995)
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NEl 97-06 (Original)
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December 1997 1997 APPENDIX _8 List Of Definitions The following dermitions are provided to ensure a uniform unaerstanding of terms used in this guideline.
l3 Accident Induced Leakage ry-d Primary-to-secondary leakage that occurs in a faulted steen generator as a result of a limiting accident.
ess Condition Monitoring i
A comparison of the as-found inspection results against the performance criteria for structural integrity and accident leakage. Condition monitoring assessment is performed at the conclusion of each operating cycle.
Degrantion-Specific Repair Criterin Repair criteria devi.oped far a specific degradation mechanism and/or location, e.g., a degradation specific repair criteria for ODSCC at tube support plates or for l
PWSCC at the tube sheet expansion.
Deterministic Approach An approach that is based on the deterministic addition of parameter values to determine a limit.
Feulted The state of the steam generator in which the secondary side has been depressurized due.o a main steam lire l'reak such that protective system response such as main steam sinc isolation, reactor trip, safety injection, etc., has occurred.
Limiting Accident An accident that results in the largest differential pressure across the steam generator tubes, normally a main steam line or main feedwater line break.
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I NEl 97 06 (Original)
December 1997 -
APPENDIX B Kont'd)
Operational Assessment Forward looking prediction of the steam generator tube conditions that is used to ensure that the structural integrity and accident leakage performance criteria wili not be exceeded during the next cycle. T he operat;onal assessment needs to consider factors such as NDE uncertainty, indication growth, and degradation-specific repair limits.
Performance Criteria Criteria to provide reasonable assurance that the steam generator tubing has adequate structural and leakage integrity such that it remains capable of a t:etaining the conditions of normal operation, including anticipated operational occurrences, design basis accidents, extemal events, and natural phenomena.
Probabilistic Approach An approach that uses probabilistic simulations, e.g., Monte Carlo simulations, to determine appropriate limits.
1 Probability of Hurst (POH) i The probability of burst of a steam generator tube if a limiting accident occurs.
i Probability of Detection (POD)
The probability of detecting a flaw during a steam generator inspection.
Repair Limit An NDE parameter valua at which steam generator tube repair is required. The repe.ir limit will be determined by either subtracting margins for NDE uncertainty and growth from the structural limit or by conducting a probabilistic analysis.
Normal Makeup Capacity The ability of the makeup system to maintain reactor coolant s:, stem inventory without the manual or automatic actuation of engineered safeguards features, e.g., safety injection. Manual starting of redunaant or smndby pumps may be credited as normal makeup capacity if the additional pumps are provided for in plant procedures.
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1El 97 06(Original)
December 1997 APPENDIX B (Cont'd)
Steam Generator Degradation Specific Management (SGDSM)
The use ofinspection and/or repair criteria developed for a specific degradation mechanism, e.g., outside diameter stress corrosion cracking at tube support plates.
Steam Generator Tube Rupture (SGTR)
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A tube rupture or burst is a gross failure of the tube such that the formation of a primary-to-secondary opening with an area affiliated to that of a double-ended guillotine break occurs. For burst testing oflimited length axial cracks, approximately two inches or less in length, the phenomenon requires extension of the crack tips. In most situations, extension of the degradation is necesst.ry to achieve the level of opening needed.
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Dectmber 1997 i
APPEllDIX C 1
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List of Acronyms -
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. ARC Alternative Repair Criteria j
'American Society of Mechanical Engineers d
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AVT All Volatile Technology CFR Code of Federal Regulations EOC End of Cycle i
EPRI Electric Power Research Institute l
GDC General Design Criteria.
GPD Gallons Per Day INPO Institute of Nuclear Power Operations i
ISI Inservice Inspection i
-MSLB Main Steam Line Break MSLB SGTR Main Steam Line Break-Steam Generator Tube Rupture f-NDE Non-Destructive Examination NEI.
Nuclear Energy Institute I
NRC Nuclear f.gulatory Commission 1
NSSS Nuclear Steam Supply System ODSCC-Outer Diameter Stress Corrosion Cracking a
POB Probability of Burst POD-Probability of Detection I
' PWR -
Pressurized Water Reactor t
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NE! 97 06(Originil) e Dectmber 1997 APPENDIX C (Cont'd)
PWSCC Pressurized Water Stress Corrosion Cracking RCPB Reactor Coolant Pressure Boundary SF-Safety Factors SG Steam Generator 7
SGDSM Steam Generator Degradation Specific Management i
SGMP Steam Generator Management Project SGTR Steam Generator Tube Rupture SL Structural Limit SLB Steam Line Break 1
TR Technical Report TSP Tube Support Plate i
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NE: 97-06 (Original) e December 1997 APPENDIX D I
12 Month Report Content
- 1. Scope ofinspections performed.
- 2. Active degradation mechanisms found.
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- 3. NDE techniques utilized for nach degradation mechanism.
- 4. Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism. Repair methods utilized and tFe number of tubes repaired by each repair method.
- 5. Total number and percentage of tubes plugged and/or repaired to date and the effective plugging percentage in each steam generator.
- 6. Description of the tube integrity assessment.
- 7. Description of corrective actions implemented,if any.
- 8. Evaluation of circumstances if condition monitoring results exceeded the previous cycle operational assessment.
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