ML20211P499

Summary of 970917 Meeting W/Vynp in Rockville,Maryland Re ASME Code Case N560.List of Meeting Attendees,Meeting Agenda & Meeting Handouts Encl

ML20211P499
Person / Time
Site:	Vermont Yankee
Issue date:	10/09/1997
From:	NRC
To:	NRC
References
TAC-M99389, NUDOCS 9710200217
Download: ML20211P499 (51)

Text

,

p uasuq fo-Z7l j~ k UNITED STATES s* NUCLEAR REGULATORY COMMIS810N WASNINGTON, D.C. 30e06 4001 October 9, 1997 k*****

LICENSEE: Normont Yankee Nuclear Power Corporation ,

FACILITY: Vermont Yankee Nuclear Power Station

SUMMARY

0F SEPTEMBER 17, 1997, MEETING AND SEPTEMBER 23, 1997. TELEPHONE CONFERENCE WITH VERMONT YANKEE POWER CORPORATION AND THEIR CONSULTANTS REGARDING THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS CODE CASE N560 - VERMONT YANKEE NUCLEAR POWER STATION (TAC N0. M99389)

Introduction On September 17, 1997, the NRC staff met with Vermont Yankee Nuclear Power Corporation (VYNPC) representatives and their consultants at One White Flint North, Rockville, Mary 1snd. The purpose of the meeting was to discuss the American Society of Mechanical Engineers (ASME) Code Case (CC)

'N560. :The case provides an alternative to the examination requirements for ASME Code,Section XI, Class 1, Category B-J pipe welds. The licensee p1kas to implement this alternative at Vermont Yankee (VY) Nuclear Power Station during the refueling outage scheduled to commence in March 1998, in addition, on September 23, 1997, the NRC staff held a telephone conference with VYNPC consultants from Yankee Atomic Electric Corporation (YAEC). Enclosure 1 lists the attendees of the September 17, 1997, meeting, Enclosure 2 lists the participants in the September 23, 1997, telephone conference, and Enclosure 3 contains the meeting agenda.

Discussion After brief introductory remarks by the NRC and VYNPC representatives regarding the objectives of the meeting, Mr. J. Duffy, YAEC, discussed the agenda of the m6eting. Mr. R. Fougerousse, Entergy, discussed the history and background of ASME CC N560. He stated that the examination zones selected for inspection are based on susceptibility to known degradation mechanisms and consequences of failure, and that the decrease in the-percentage of Category B-J welds inspected from 25% to 10% targeted for known degradation mechanisms and consequences of failure will enhance the overall safety of the plant. The viewgraphs used for his presentation are included as Enclosure 4.

Mr. P. Riccardella, Structural Integrity Associates (SIA), discussed the g methodology for the selection of welds to be inspected. He provided a degradation mechanism checklist and a detailed degradation list of results for train B of the main feedwater system. He also provided a sunniary of results for the following systems: (1) Recirculation system, (2) Main steam system including drain, reactor core isolation cooling and high-pressure coolant injection,-(3) Standby liquid control system, and (4)

Reactor water cleanup system. The viewgraphs used for his presentation are ir.cluded as Enclosure 5.

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Ms. V. Dimitrijevic', VAEC, discussed the evaluation of consequences of postulated breaks. -She stated that the VY N560 program is prioritized accord;ng to the risk associated with each location / segment. The risk is developed by qualitatively combining consequence categories and failure potential catenories (related to degradction mechanisms) in a matrix format. The v'owgraphs used for her presentation are included as Enclosure 6.

Mr. D. Girroir, VYNPC, discussed the implementation of the program at VY.

The viewgraphs used for his presentation are included as Enclosure 7.

Mr. C. Larsen YAEC, discussed the enhancements to VY nondestructive

-examination (NDE) program for CC N560. He provided viewgraphs that showed the typical volumetric examination which will be used for CC N560. The NDE will be consistent with the ASME Code,Section XI requirements. The viewgraphs used for his presentation are included as Enclosure 8.

In response ts P:RC staff's questions, the VYNPC representatives provided additional viWgraphs regarding the comparison to Draft Regulatory Guide-1061 and the impact of N560 on ASME Class 2 and 3 piping systems. These additional viewgraphs are included as Enclosure g.

Conclusion The NRC staff had several comments on the presentation regarding: (1) the scope of the analyses and the relief requests, (2) compar son to Draft Regulatory Guide-1061, (3) the 'nethodology for evaluating unknown mechanisms, (4) justification (or the number of cycles assumed in the screening criteria, (5) the inspection techniques and equipment used, (6) monitoring of the degradation, and (7) the risk impact on plant safety. In addition, the staff requested that the licensee should investigate whether or not the change in the inspection scope and examination of the welds constitutes a change to the licensing basis of the plant, a relief for the current period or a permanent change.

VYNPC committed to provide quality responses to NRC staff's comments in a timely manner. However, the staff indicated that it does not believe that the review can be completed prior to the March 1998 refueling outage. The VY submittal is the lead for boiling-water reactors and the staff's comments are extensive. The quality of VYNPC's responses to the comments

-provided during the meeting can be a significant factor in expediting the review. Mcreover, the staff's resource constraints may impact the completion of the review required to support the implementation of CC N560 at VY during the refueling outage that commences in March 1998.

Post-meetina Telephone Conference On September 23, 1997, a telephone conference was held between the NRC staff and VYNPC consultants from YAEC. The following coments were discussed:

,_ a 1) Provide a description of the groups, and the process undertaken

' by VYNPC staff, charged with reviewing and accepting key engineering assumptions, inputs, and results of the analyses that determine the safety-significance of pipe segments. For example, did VYNPC involve multi-disciplinary expert panel (s)?

2) Provide a description of the internal and external reviews of the VY probabilistic risk assessment and the findings of these reviews. Also, provide VY's responses to the to these findings.

3) Review the June 12, 1997, request for additional information (RAI) on the EPRI methodology. Identify and develop responses to the RAI applicable to the VY submittals.

4) Provide a more descriptive presentation of the consequence i

methodology and include several examples.

5) Provide a discussion of the qualitative and quantitative basis-for Table 2-2 and the use of an unavailability of 0.01 to develop " pseudo" trains in table 3-2. Discuss the relationship between the pseudo trains, the actual physical trains at the l plant, and the success criteria.

! Again, the licensee committed to provide quality responses to NRC staff's comments in a timely manner. The staff indicated that, for the reasons stated above, the review required to support the implementation of CC N560 at VY may l

not be completed prior to the refueling outage that commences in March 1998.

Enclosures:

1-2. List of Attendees

, 3. Agenda 4-9. Handouts i cc w/encls: See next page DISTRIBUTION - Encis.112 only E-mail Hard Copy- All Enclosures S. Collins /F. Miraglia (SJC1)(FJM) u Docket. File R. Zimmerman (RPZ) "PUBLIC"

B. Boger (BAB2) PDI-3 Rdg R. Eaton (RBE1) OGC S. Little ACRS T. Martin (SLM3) K. Jabbour

4. Bagchi (GXB1) C. Hehl M. Rubin (MPR)

, S. Dinsmore (SCD1)

G. Georgiev (GBG)

5. Ali (SAA3)

C. Smith (CXS1) Q0\

U T. McLellan (TKM)

K. Kennedy (KHK)

I [

, D. Jackson (DTJ) -

M. Cheok (MCC2) '. J 5 2 '8 g

~

s K. Coyne (KXC)

' P. Patnaik (DTJ)

DOCUMENT NAME: G:\JABBOUR\MTG. SUM *SEE mEVIOUS CONCURRENCE Ta receive a copy of this document,Indcate in the box; .*C" = Copy,without attachmentlenclosure "E* = Copy with attachment / enc e

• N' = No co ',

OFFICE PDI-3/PM l - PDI l/lA d Q L J U EGGB* l SPS8* (A)D:Pgi g // l NAME KJabbour t/, Vf SLitt$ 8 7"* GBagcht MRubin REntph 6'/U DATE 10/ Q /97 * ~

10/i '//97 l 10/08/97 10/0$97 M/) /97 0FflCI AL RECORD COPY '

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f 'ermont' V Yankee Nuclear Power Vemont Yankee Nuclear Power Station

, Corporation cc:

Regional Administrator, Region I Mr. Raymond N. McCandless U. S. Nuclear Regulatory Comission Vermont Division of Occusational

475 Allendale Road and Radiological Healt1 i King of Prussia, PA 19406 Administration Building Mr. David R. Lewis Shaw Pittman, Potts & Trowbridge Mr. J. J. Duffy 1

2300 N Street, N.W. Licensing Engineer

Washington, DC 20037-1128 Vermont Yankee Nuclear Power

. Corporation Mr. Richard P. Sedano, Comissioner 580 Main Street Vermont Department of Public Service Bolton, MA 01740-1398 120 State Street, 3rd Floor

Montpelier, VT 05602- Mr. Robert J. Wanczyk -

Director of Safety and Regulatory

! Public Service Board Affairs

State of Vermont Vermont Yankee Nuclear Power Corp.

120 State Street 185 Old Ferry Road l Montpelier, VT 05602 Brattleboro, VT 05301 i Chairman, Board of Selectmen Mr. Ross B. Barkhurst, President

' Town of Vernon Vermont Yankee Nuclear Power-P.O. Box 116 Corporation i Vernon, VT 05354-0116 185 Old Ferry Road

. Brattleboro, VT 05301 i Mr. Richard E. McCullough

Operating Experience Coordinator Mr. Gregory A. Maret, Plant Manager i Vermont Yankee Nuclear Power Station Vermont Yankee Nuclear Power Station

P.O. Box 157 P.O. Box 157 i Governor Hunt Road Governor Hunt Road l Vernon, VT 05354 Vernon, VT 05354 1

i G. Dana-Bisbee, Esq. Mr. Donald A. Reid L Deputy Attorney General Senior Vice President, Operations

33 Capitol Street - Vermont Yankee Nuclear Power

Concord, NH- 03301-6937- Corporation l 185 Old Ferry Road i Resident Inspector -

Brattleboro, VT 05301

Vermont Yankee Nuclear Power Station i U.S. Nuclear Regulatory Comission Ms. Deborah B. Katz P.O. Box 176 Box 83 j Vernon, VT 05354 She11burne Falls, MA 01370 4

Chief, Safety Unit Mr. Jonathan M. Block, Esq.

! Office of the Attorney General Main Street 3

One Ashburton Place, 19th Floor P.O. Box 566 Boston, MA 02108 Putney, VT 05346-0566 i

Mr. Peter LaPorte, Director ATTN: James Muckerheide Massachusatts Emergency Management Agency 400 Worcester Rd.

l P.O. Box 1496 l Framingham, MA 01701-0317 4

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NRC/VYNPC Meetina September 17, 1997 List ot' Attendees HRC VYNPC and Consultants Westinahouse K. Jabbour J. Chapman M. Osborne G. Bagchi D. Girroir N. Closky M. Rubin J. Duffy S. Dinsmore P. O'Regan G. Georgiev V. Dimitrijevic Union of Concerned Scientists S.- Ali C. Larsen C. Smith J. Moody D. Lochbaum T. McLellan S. Gosselin (EPRI)

J. Guttmann R. Fougerousse (Entergy)

D. Jackson J. Butler (NEI)

M. Check P. Riccardella (SIA)

K. Coyne P. Patnaik

. M. Anderson (INEL, by telephone)

ENCLOSURE 1 p.m+o-*-yes.e w+4w ww..w.e.o w -wee a --n - .a..

NRC/VYNPC Consultants Teleohone Conference September 23, 1997 List of Particinants EC, VYNPC Consultants from YAEC K. Jabbour J. Duffy G. Bagchi P. O'Regan S. Dinsmore V. Dimitrijevic' S. Ali C. Larsen A

ENCLOSURE 2

AGENDA Introduction m

Background on Code Case N560 r

VY Application of Code Case l e VY Degrafation Mechanism Evaluation

. VY Consequence Evaluation

. VY Results I

Implementation Inspection Schedule and Milestones Closure / Action Items ts 4

- ,, ENCt.0SURE 3

S 8 VERMONT YANKEE Code Case N560 s

Sept 17,1997 ENCLOSURE 4 O e *

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3 HISTORY OF CODE CASE N-560

. TG on ISl Optimization Formed in 1991 To Review Section XI ISI Requirements & Develop Recommended Changes Where Warranted

. ' Assigned Task To Investigate Examination Requirements For Class 1 (Category B-J Piping) in 1992

+ 20 years experience '

+ Costs & man rem

+ Correlation with current selection criteria

. TG Conducted Extensive Study Considering:

+ Survey of industry ISI results

+ Generic industry experience w/ Class 1 piping

+ Engheering evaluation of operational loadings and degradation mechanisms

+ Safety impact of failures

+ Man-rem exposure and costs of inspections

. Study is Documented in ASME White Paper 92-01-01, (July 1995)

f StructuralIntegrityAssociates,Inc.

~~ __ _ _ _ _ . _ _ _ _ _ _ _ . _

LA

CATEGORY L-J CODE CASE

. Examination Zones Selected Based on Susceptibility to Known Degradation Mechanisms and Consequences of Failure ,

. Inspection Program Based on Total Number of Examination Zones = 10% of Class 1 Piping Welds (Excluding Socket Welds)

. Examination Zones Consist of Structural Elements Such as Welds, Fittings or Pipe Segments

. Examination Volume (Length & Extent) and Examination Method Must be Consistent with Degradation Mechanism (s) of Concern (i.e., inspection for Cause)

,_ . _ Q structuralIntegrity Associates, Inc.Y

P 3 l i

CONCLUSIONS OF WHITE PAPER i

92-01-01  :

i i e Flaws have been detected in a very low percentage of Category B-J welds and a majority of these findings were due to IGSCC in BWRs.

• There is no apparent relationship between flaws detected and those

selected due to ASME Section XI"hlgh stress /high fatigue" criteria.

• Plant specific studies have determined that attemate selection i methods are available which would maintain the overall effectiveness of the inspection program while decreasing the number ofinspections.

• Over the past 20 years, only a handful of degradation mechanisms i have been shown to affect Category B-J piping, and of them, only IGSCC has proven significant.

. The most critical locations can be ranked based on numerous factors including the probability of failure, consequences of failure, system redundancy and mitigating factors (i.e., leak-before-break considerations).

• Examination zones should be adjusted as appropriate to ensure that the inspection covers all potential areas that are susceptible to the relevant degradation mechanism (s) at that particular location.

i e A decrease in the percentage of Category B-J welds inspected from 25% to 10% targeted per known degradation mechanisms and consequences of failure will enhance the overall safety of the plant.

I i

Y , f StructuralIntegrity Associates, Inc.0

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9 VERMONT YANKEE 4

N560 ENGINEERING ANALYSIS 9

ENCLOSURE 5 .

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l DAMAGE MECHANISM ASSESSMENT PROCESS

. Review System Description and Function (s)

! . Identify Class 1 Boundaries l  :. -

i l . Apply Checklists to Determine Susceptibilty

! to Degradation Mechanisms (By System or portions of System) 4

. Conduct Design Review of Preliminary Results

. Itemize Results by Weld Number for Each System

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PRS-97-044 3

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ISI ISOMETRIC -

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i Degradeders Asochenism Assessment Workshoot Condersforrt 7tfs4 Sogneont No.

Na, l Attrnates to be Cortsidered *~ l ** l

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• Remarks TASCS-f rips > f beh, and x 0 D D TASCS-2 ppe segment has a slope < 45' kom hortiontd W edbew or x D D D too hio a woricatp>e), and l TASCS &1 poternaledsts Wkw now h a p>e socion connectedto e x D D D Potentatwhot RnCUSaw wie MCC

! w.,~. .c.: aso=*,g meg erher sad conf adds, or conneeson em+ig spent storapwedbwn h horttontar segment ofFWhie B (koshoem et cess f-A> cess 2 boundery)

TASCS32 potenEnt eelsts tbrirakage scw pest e vehe f e, h4eehage, oat- O x 0 O seekage, crossJeakege) afowhg meg ethat andconf AAfs, or . .

TASCSs3 potensar e=Is's W convecean beneng h dee6endedp>e seccons u x D D 1 connected to a source ethot nukt, or 1 . TASCS-3-4 poten6at entsts W two phase (steem /werer) saw, or D x D D i

• TASCS-3-5 potonnetexists & surbunentpenetteson h branch pipe connected x 0 0 C l

to beederp>hg contakhg hot Add we tigh turbrient now, omt TASCS-4 canctieted ermeasured AT> $0*F. and x 0 D D r TASCS-5 Rschordson number > 4 0 x D D D

! In conclusion, hertzertal segments et Line B of the Main Feedwaler System are effeded t'y the TASCS Degradarian Mechaniern due to the poten5al flew at hat RWCU vta RCIC w.AL. during pierd startupPshtAdowrt.

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TT-f f operethg ferryere* . n 270*F for stainfess stes( er D D D D TT-f-2 opern6ng nemperee.,n - 220*FIbr carborn steel and x D D D potonnat WrW repid terriperature changes hcksting TT-2-1 cold waterkrec6on hto hetpipe segment. or D x D D TT-2-2 hot waterk}eceort hto coktppe segment, omf x 0 0 0 Hot Reactor werer bechnow hio Nozzles (appses6de se atLoops!-

TT-3-f ldTI > 200*F Wstahless stee( or D D D D TT-3-2 l3Tl > f50*Fhearbon steel er x 0 D D TT43 lJTl > dT afoweble (oppicable 80 both stohisss amicarbon) D D x 0 Both loops are effected by TT in close proiamity to the reactor vessel (horttordai sechens to first abow) due to het reador water beckfkm Irto cold FW piping-i, PRS-97-044 7

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!: DEGRADATION MECHANISM CHECKLIST- MAIN FW SYSTEM .

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1 CSCC 8-1 eenkoenedk accor*nce sua emissng pientnSCcpogram pw O x 0 0 notpart of ss41pogram smco carson smer

! NRC Genernc LeBer 88-01 CSCCP-t operenne senperature > 200 F. and O O O x WA enks er& se PWRs.

IGSCCP 2 suecepeute n eterior(cs@on cornent10035%L and O O O x WA sondes anny se PwRs.

CSCC-P-3 fensfe seems % reskAsaf stress)is posent, and O O O X' WA appfes er$ de PWRs.

fGSCC P-4 orygon oredizeig specks are present O O O x WA appfes enfy se PWRs.

1 on I O O O' X WA anodes enfy se PWRs.

CSCC#-S operusng serrperufure < 200"F. she sereufes a6ow apply, and

.' - IGSCC-P-8 knietig contaminents (e g,0desurate. Aroeide. chicdde) are O O O x WA appies oniyse PWRs.

.h. - ,e p ee.n, in craluelon.:he Mmen Feedesser System is nci effeckd ty the IGSCC Degradsson DAecheniern due to the tect that sw piping is conposed of corben steel.

I a rnaternet that is not sunnepehte to IGSCC.

I TGSCC-f operothy teorpereene > t$0*F. and x 0 0 0 j TGSCC-2 sensas seress dhetning reshtrat seessf is posent. and x 0 0 O f TGSCC-31 handes(e g, Aooride chloride)arepesent.or O x 0 0 krpurks are coneenedperEPPt TCSCC-3-2 cousiclNeOH)is pM and O x 0 0 waterchermey guhnahees.

TGSCC-4 erygen er enki&g species are pesent (on0 1required en be x 0 0 O pesent ks conAmcson enhafdes, not regu6ed outeesse) in conctunion, the Main Feedneter Syoiem is not affected by the TGSCC Degradstion Mechanism due to the absence et besdes and causses. per EPRI Water Chemistry GuideAnes.

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! DEGRADATION MECHANISM CHECKLIST- MAIN FW SYSTEM .

4 Cw _ "_ Afechendsm Assessment Wortsheet ConcArsions Rfs4 Segment Akt

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t ECSCC-f opernehg :..++^6 e > 150*F. and X D D D CCSCC-2 ferrsiesbesskpresenf and X D D D ECSCC-3-f an eufskie pphp sr#ece h outNb fue esmefers of a po6eede O D D X ben pem (e p. vshe stems) andis coveredsee naseistsSc i

hw met k sof h cerryssnee sieh Reg. Gesc as f.36, or t" ECSCC-3-2 en outskie pphg serface is erposed so sevahr tem cMatie D D X 0 1 6eerer ene*enments (e g, seawerer, erecAfs4 wafer. Armel in corchasion.ew Main Feedneter System is not affected by the ECSCC Degradsson MecherWom due to the fad that the system la corsposed of carten sleet

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in conesuoion. the Main Feedneter System is riot er fected by the PWSCC Degradonan Mecherusrn due to the fad that thrs system is not erposed to pre tery w er s at T* 620 F and corenins no inconet j

MFC-f operseng sornperschre < f50*F. and D X D D AWC-2 bw orhfometrent Apw. and D D X D Remote ewonf.

M9C-3 pH< 10. and D D X D

&#C-4.f pesence4nettsion of orpenic meterfsf (e p. raw wafer sysJentP. or D D X D MFC 4-2 water source is not boeted wMoodes (e g. reAmsng waterfanA) D D X D in cortremarwi, the Main Feeduler System is riot affected by the MfC Degradsman Mechenam due to the fact that tus system is eiposed to anv .:. ,

temperature yeeser then 1507.

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h e wSr l h e e A.

E t l lI

t. t t. t t

e i

=h n o n D n e

b i me lo %de 'w_r*

%i e t n a u ewtw o t c +ef N m r o f 2 3- 4 5- n F. f wFo f 2 3 ot e 2 3 w "

o a""

f- on r cn C C or C onA T.

T T-I T-f no C C- C- ce C. C- C- ct n52 A cii as P P P ic C C C Inr a E- E- E- E- E- I F InMre wH!

' 4 4

• 0-7 9-S R

O P

l

! .i!!;! l1I j  !

!! i t l,  :  :,

, \ll Illll

i t

i ~

ITEMIZED RESULTS FOR MAIN FEEDWATER LINE B t

e

System Une WeM WeM TF SCC LC .-

ID Number Number Locatfort TASCS TT IGSCC TGsCC ECSCC PWSCC h0C PET CC EC FAC 104DW-18 FW18-N4C-SW Dowristream sitt. of Reactor Vessel Nozzle N4 N Y N N WA N N N N N [N/A

[ FW FW 10fDW-18 FW1846 bpetream side d Reedor Vessel Node N4 N Y N N WA N N N N N WA FW 10fDW-18 FW1845A Downstream of Elbow. N Y N N WA N N N N W WA Upst eam of Reactor Vessef Netzte N AC FW 10-FDW-18 FW1845A-SA Downstream ski.of Elbow. N Y N N WA N N N N N WA Upstreem from Reactor Vessel Nozzte N4 FW 1MDW-18 FW1MSA-SB Upstream side of Elbow. N Y N N WA N N N N N WA Downstream of Spring Hangers FW 18 and FW-19 FW 1040W-18 FW1STS Upstream of Spnng Hangers FW-18 and FW-19 N N N N WA N N N N NIA FW 10fDW-18 FW1845 SA Downstream of Branch N N N N WA N N N .N N N/A 1

FW 1M DW-18 FW1*F4 Downstream of Branch Conrwactaan wRtt 10" FWD-20 N N N N WA N N N N ,N .Y l

! *

• FW 10-F DW-20 FW20-N4D-SW Downstream side of Reactor vessC Nozzle N-4D N Y N N WA N (N N N N .lWA FW 10fDW-20 FW20f4 t>) stream side cf Reactor Vessel Nozzle N-40 N Y N N WA N N N N N WA FW 1MDW-20 FW20f3A Downstream of Etow. N Y N N WA N N N N N WA j Upstream of Reactor Vessel Nozzie N-4D i FW 1MDW-20 FW2043A-SA Ocwnstream side of Ebow. N Y N N WA N N N N N WA Downstream of Spring Hangers FW-12 and FW-13 l' N WA N N N N N WA FW 10-FDW-20 FW20f3A-56 Upstream side of Elbow. N Y N Downstreem of Spring Hongare FW-12 and FW-13 1MDW-20 FW N3 Upetreem of Spnng HanDean FW-12 and FW-13 N N N N/A N N N N N -WA

/W .N FW 1MDW-20 FW20f3 SA Upst eam of Spring Hangers FW-12 and FW-13 .N N N N WA N N N N N WA 1MDW-20 N 2A Downstream side W Ebow. N N N N WA N N N N N WA FW Upstream of Spring Hangers FW-12 and FW-13 FW 10-FDW-20 FW24F2 Upstream side of Etow. N N N N WA N N N  : N N WA Downstreem of Spring Hanger FW-14 IMDW-20 FW2M1A Downstream of Reducer. N N N N WA N N N N N WA

. FW Upstrearr of Snubber FW-15 N N ,N N WA N N N 'N N Y FW 1MDW-20 FW2GEIB Downstream of Restrairt FW-18 l

Upstream of Spring Hanger FW-17 N N N N N/A N N N N N Y FW 10fDW-20 FW20f1 Downstream side of Reducer.

Downstream of Branch Connec5on weh 10" FWD-18 Upstream side of Reducer. N N N N N/A N N N N N Y FW 104DW-20 FW20f30

, Downstream of Branct Connection weh 17 FWD-18 PRS-97-044 11

t i

I -

I ITEMlZED RESULTS FOR MAIN FEEDWATER LINE B g

(Continued) -

System Lkse WeM WeM TF SCC LC FS O Number Nuneeber Lace 6ers TASCS TT N3 SCC TCSCC ECSCC PWSCC ANC PfT CC EC FAC j

FW 16-FDW-18 FW18-F3A Downstream of Spring Hanger FW-20 N N t. N WA M N N N N if l

upseroem of Branch FW 16-FDW-18 , FW18-F3 Dovonstream side of ettmur. N N N N WA N N N N N WA Upstream of Sprino Hanger FW-20 FW 1MDW-18 FW18-F2E Upstream skie of Ebour. N N N N WA N N N N N WA Upstreem of Spring Har'Oer FW-;V FW 16-FDW-18 FW18-F2D Downstream side of Ebour. N N N N WA N N N N WA gN Ca.a.7.of ManualValueV2-298 l FW 18-FDW-18 FW1M2C Doumstreamof Hestraint FW-21 N N N N WA N N N N 'N N/A l Upstream of Elbour FW 16-FDW-18 FW18-F2B Doumstream of Restraint FW-21 N N N N WA N N N N N WA l

N N WA N N N N N WA l [ FW 16-FDW-15 FW1842A Downstream of Manual Valve V2-298, N N t l Upstream of Restraird FW-21 N N N N WA N N N N N WA FW 16-FDW-18 FW18-F2 Doumstream side of Manual Vafwe V2-298 1 FW 16-FDW-17 FW17-F7 Upstream side of ManualVafve V2-298 N N N N WA N N N N .N N/A j

FW 16TDW-17 FW17-F6 Doumstream side of Check Vafve V2-288 Y N N N WA N N N iN N WA FW IM DW-17 FW17-FSA Upstream side of Check Vahe V2-288 Y N N N WA N N N N N WA Upstrearn side of Penetration X-98 (FDW-HD38) Y N N N WA N N N N N N/A FW 1MDW-17 FW17-MF4A FW17-MF4 Doumstream side of Chedt valve FDW-96A Y N N N N/A N N N N N N/A FW 16-FDW-17 l

t PRS-97-044 12

V  % -

i .

SUMMARY

OF RESULTS I

i l . Recirculation System

': - Thermal Fatigue Generally Not a Concem (Except 2 Welds at RHR Connections)

- Stress Corrosion Cracking Not a Concern Since Piping System  ;

! Replaced (NUREG-0313, Cat. A)  !

I

- Not Affected by FAC or Localized Corrosion Mechanisms

. Main Steam System (incl /M.S. Drain, RCIC and HPCI)

- Main Steam, RCIC, and HPIC Systems Not Affected by Thermal Fatigue

- Main Steam Drain (Small Bore Lines) Potentially Affected by TASCS -Two-Phase Flow

- Stress Corrosion Cracking Not A Concem (Carbon Steel)

- Not Affected by FAC or Localized Corrosion Mechanisms (Steam)

& ( Structurallategrity Associates, Inc.

PRS-97-044 13

V  % -

1 '

SUMMARY

OF RESULTS (continued) l

. Main Feedwater System

- Therma Fatigue a Concern at Horizontal Sections Near RPV and Containment Penetration

- FAC Mechanism Applicable at Selected Locations in System (Per Utility FAC Program)

. Residual Heat Removal System j -

System Susceptible to Thermal Fatigue

-- System Not Susceptible to SCC (NUREG-0313, Cat. A or Carbon Steel)

Local Corrosion Mechanisms Not a Concem FAC Not identified as Concem for C.S. Portions of Line

. Core Spray System Horizontal Sections Near RPV are a Therm .: Fatigue Concern

- IGSCC a Concem at Two Piping to Safe-End Welds (Balance of System NUREG-0313, Cat A)

- Not Affected by FAC (Stainless Steel)

W ( Structurallategrity Associates, Inc.

PRS-97-044 14

W i "

SUMMARY

OF RESULTS (continued)

I:

!i

. Standby Liquid Control System

- Segments Adjacent to RPV Susceptible to Thermal Fatigue

- Portion of System > 200 F Susceptible to IGSCC l

- Not Affected by Other Mechanisms b: Small Bore, Socket Welded System

! NOTE:

i l

l

. Reactor Water Cleanup System

- System Not Affected by Thermal Fatigue

- System Not Susceptible to SCC (NUREG-0313, Cat. A)

- Not Affected by.Other Mechanisms f StructuralIntegrityAssociates,Inc.

PRS-97-044 15

V  % -

SUMMARY

OF RESULTS (concluded) 1 i . Generic Observations (All Systems) .

- Not Affected by TGSCC or Pitting Because VY Follows EPRI Water Chemistry Guideiines

)

l

- PWSCC Not Active Because OP. Temps < 620 F

\  :.

- MIC Not Active Because OP. Temps. > 150 F l l

l f - ECSCC Not a Concem Because insulation on SS Systems Meets Reg.

l Guide 1.36 Guidelines

- Crevice Corrosion Generally Only Applicable at Nozzle / Safe-End Welds (Category B-F)

- Erosion-Cavitation Not Active Because OP. Temps. > 250 F

- Water Hammer Not identified as Concern in VY Class 1 Systems

( StructuralIntegrity Associates. Inc.

PRS-97-044 16

0 m Drywd1 Wa3 RCIC Movis Movis C>Q- HPCI j Movis Movi6

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I Spatial Considerations  !

(Outside Drywell) 1 LOCA outside containment requires an auto isolation or passive valve failure

' Isolation valves that mitigate breaks outside the drywell are s,

located inside the drywell (i.e. spatially separate) l Impact of break on mitigative equipment located outside -

the drywell considered

l-i System Location Impact j LPCI RBEL252 High

CS RBEL280 High SDC RBEL252 High CUW RBEL280 High l SLC RBEL318 High

! HPCI Steam Tunnel Medium RCIC Steam Tunnel Medium

! MS Steam Tunnel Medium I FW Steam Tunnel High L MSD Steam Tunnel Medium i

5

,-,-m .. -++.-e"^"' " * * " * " * ' *

• p -+ +- y, ,e--- - - --m---7---- - - - - +- - - - - , - -.--m------.,e- +r,yy pmy--y-w -y-c.-- 3 r -- -

p w e r--,-, --,-y,wwy,vr++e yw=- pis-yw e ,-t-- *

. c INTERSYSTEM LOCA 43 Locations ldentified With The Potential To Effect The Likelihood OfIntersystem LOCA e

63% ranked as High consequence 19% with identified degradation mechanism (TF)

The N560 program is designed to cover all locations / segments where potential intersystem LOCA will result in High consequences and the potential for degradation is identified.

Summary - Final Impact on Intersystem LOCA is Positive!

O 4

_7;;;; ;;;;;- -- - ~

f i

i i

SUMMARY

OF CONSEQUENCES

'i

'i r

t CONSEQUENCE

.LLOCA MLOCA PLOCA ILOCA ISLOCA LOCA-OC TFWMS SDC g

No. dWdds No. d Wdds Na, dWdds No. d Wdds No.d Wdds No. of Wdds No. dWdds No. d Wdd, CS 32 22 - 6 -

4 - -- -

69 63 - - - 4 2 -

l FW -

1 - 6 - -

i  ; HPCI 19 12 - -

MS 117 105 - - 4 - s - -

2 1 - -  !

MSD e -

5 - -

18 7 - 4 - 7 - -

RCIC -

RECIRC 69 55 14 - - -

57 22 - 11 -

17 - - 7 RHF RWCU 18 - 6 - 10 - 2 - -

' 279- 19- 21 28~. 2- ^7-2Td. [N ;407 32.7 19

~

a.

100.06s. L7.9% -4.74: 4.7%- g. 5.2% 6.9% 0.5% 1.7%

Total (%):j 68.64.. j

~

i

'n2/94 2:43o2 m

FMECA - Segment Risk Ranking Report -

Degr d en

! Number Degradation Mechanism C axe Rhk Risk

SegmentID ofWelds Lines in Segment Welds in Segment Mechanisms Category Category Catercry Rank

FW-001 1 10-FDW-18 FW18-F4 FAC LARGEEAK HIGi CATI HIGH I

FW-004 1 10-FDW-19 FW19-F4 FAC LARGEEAK HIGH CATI HIGH FW-007 3 10-FDW-20 FW20-F3B,FW20-FL FW2n-FIB FAC LARGE EAK HIGH CATI HIGH i FW-010 1 10-FDW-21 FW21-F1 FAC LARGEEAK HIGH CATI HIGH FW-021 1 16-FDW-18 FW18-F3A FAC LARGE LEAK HIGH CATI HIGH FW-023 2 16-FDW-19 FW19-F3B, FW19-F3C FAC LARGELEAK HIGH CATI HIGH FW-016 2 16-FDW-17 FW17-MF4, FW17-MF4 A TASCS SMAll,EAK HIGH CAT 2 HIG1

! FW-003 5 10-FDW-18 FW18-FSA-SB FW18-F5A-SA. TT SMALLLEAK HIGH CAT 2 HIGH FW18-F5A. FWi%-F6. FWIS-N4C-SW FW-006 5 10-FDW-19 FW19-F64-SB. FW19-F6A-SA. TF SMALL EAK HIGH CAT 2 HIGH

! FW19-F6A. FW19-F7. FW19-N4A-SW TF SMALLLEAK HIGH HIGH FW-009 5 10-FDW-20 FW20-F3A-SB, FW20-F3A-SA. CAT 2 FW20-F3A FW20-F4,FW20-N4D-2 FW-012 5 10-FDW-21 FW21-F2A-SB, FW21-F2A-SA. TF SMALL GAK HIGH CAT 2 HIGH FW21-F2A, FW21-F3 FW21-N4B-SW FW-018 2 16-FDW-17 FW17-F6, FW17-F6A TASCS SMALLLEAK IBGH CAT 2 HIGH FW-013 2 16-FDW-16 FWI6-MF7. FWI6-F8 N NONE HIGH CAT 4 MEDIUM FW-002 2 10-FDW-18 FW!8-FS-SA.FW18-F5 N NONE HIGH CAT 4 MEDIUM FW-005 5 10-FDW-19 FW19-F4A. FW19-FS.FW19-FSA, N NONE HIGH CAT 4 MEDIUM FW19-F6-SA FW19-F6 FW-008 5 10-FDW-20 FW20-F1A FW20-F2,FW20-FA' N NONE IHGH CAT 4 MEDIUM FW20-F3-SA, FW20-F3 2

E

i l

r

'{

't:2s42:43o3 m

[ FMECA -Segment Risk Ranking Report --

c :_ _

N e, De.e - C- Ra m SegawatID ofWehls Lines la Segned Welds in Seipment Mechemises Category Cseegory Cseezery Rank i

} FW-011 2 10-FDW-21 FW21-F2-SA.FW21-F2 N NONE HIGH CAT 4 MEDIUM

! FW-015 3 16-FDW-16 FW16-19, FWI6-F9A.FW16-F10 N NONE HIGH CAT 4 MEDIUM FW-019 1 16-FDW-17 FW17-F7 N NONE HIGH CAT 4 MEDIUM FW-020 7 16-FDW-18 FW18-F2, FW18-F2A.FW18-F2B, N NONE HIGH CAT 4 MEDIUM FW18-F2C.FW18-F2D.FWI8-F2E, FW18-F3 FW-022 7 16-FDW-19 FW19-F2, FW19-F2A, FW19-F2B, N NONE HIGH CAT 4 MEDIUM FW19-F2C, FW19-F2D,FW19-F3 FW19-F3A

, FW-017 1 16-FDW-17 FW17-F5A TASCS SMALLLEAK MEDIUM CATS MEDIUM

.- FW-014 1 16-FDW-16 FW16-F8A N NONE MEDIUM CAT 6 LOW l

1 3

,. ..-s. -

r V

l

SUMMARY

OFTHE PROPOSED PROGRAM VERSUS THE CURRENT PROGRAM PerSystem Category 1 Category 2 I Category 4 Category 5 Category 6 & 7

[ sg No.erWde hofWde AerWde No.of Wdh herW44 System ~ wd T*t - mm wm Tel : mm wm T*i mm em T*t -- mm %m .Tei - mm wm CS 132i S t '- - -

'20 9 8 c'6L 2 -

6 1 -

rw 169' 23 ': 9" 9 9 224J 12 7 ^N-4 -

'1- - -

'1 - -

HPCI /19$ 2 *- '

U 121 3 2 -'

- - ~7- 2 -

. . Ms E117' e t .- -

1 105' 29 .I - -

12~ 3 -

MSD -Su 10 '~

'_ ~

6 2 -

'2 - -

RCIC 118: 1  ?-- - - - - -

'18 5

RECtRC 69I 10 - - -

'2 i 2 153i 14 2 -: - -

~ 14 4 -

RHn :1570 11 '

21-I 7 6 ?lt' 4 -

~

=

18 7 -

awcu dig 3 4 ' - - - 4-.

ll 2 '. ' - 1  :-. - -

.16 5 -

Total. I'4071. 77 59- 9 9 :57 28 23 "230'.

56 9 7- 2 -

94 27 -

I m= se m

i'errormancv 9ased Irnplementation, Monitoring and Feedback -

4 Performance 13ased Implementation r

h"

• Inspect Loceons Susceptible to Degradation e Substitute Leakage Testing for Volumetric Examina'tions Based on Operatbig Performance History l

\

Monitoring and Feedback a On-Line Leakage Monitoring I . Section XI Leakage Testing

• IWB-3130(c) - results shall be compared to PSI and priot ISI

. Flaws Exceeding Acceptance Criteria (IWA-3400)

K

. Increase to include those items scheduled for this and the next ,

schedule period (IWB-2430 (a))

. Additional flaws - all items of similar design, size and function (IWB-2430 (b))

• Flaw - removed, repaired, replaced or analytical evaluation (IWB-3130/3140)

• Flaws Not Exceeding Acceptance Criteria (IWA-3400)

• Items shall be examined for the next three inspection periods (IWB-2420 (b))

ENCLOSURE 7 e-= --**=.-+,e-e-e.---..-e--.. ..m,_._...,.,,,_.,,, ,.,._ .

- -- . mmme i

. . o Monitoring & Feedback VY Tech Spec 3.6.c

. Unidentified Reactor Coolant Leakage shall not exceed 5 gpm

. Total Reactor Coolant Leakage shall not exceed 25 gpm VY. Monitors Feedwater Nozzles for bypass flow with (4) thermocouples per nozzle Drywell Monitoring:

. Radiation, e Temperature e Pressure s

e e

e m .e , w*- ma mm +w ws- **~*

IMPACT ON OTHER PROGRAMS Section XIISI Program

. Code Case N560 applies to non-socket welded B-J components only

)

• No Impact on Other SXI Requirements Generic Letter 88-01 (NRC Position on IGSCC in BWR Austenitic SS Piping c

. substantial piping replacement with IGSCC resistant material e

only three potentially susceptible locations remain (all B-F category components)

Generic Letter 89-08 (Erosion / Corrosion-Induced Wall thinning)

• No Change Leak Detection Requirements

. No Change Chemistry Control Program

. No Change O

m.wn x. . . . e. -=. =-wa- * -+ ** '- *"""

VERMONT YANKEE

,' CODE CASE N-560 "l CATEGORY B-J WELDS SAFETY INCREASE FROM e WELDS CHOSEN HAVE HIGH SAFETY CONSEQUENCE e WELDS CHOSEN HAVE MORE LIKELlHOOD OF DEGRADATION e NDE TECHNIO.UE IS TAILORED TO SPECIFIC DEGRADATION MECHANISM I

E 5

5 .

m

t

[ 1 VERMONT YANKEE -

. CODE CASE N-560 l CATEGORY B-J WE!_DS ENHANCEMENTS TO VERMONT YANKEE NDE PROGRAM FOR N-560 e REVISE ULTRASONIC TEST PROCEDURES TO ADDRESS EXPANDED VOLUME FOR PARTICULAR DEGRADATION MECHANISMS i

• PROVIDE TRAINING TO.

l  :.

e PROCEDURE REVISIONS i

! e STEER TECHNIQUE (INCLUDING EVALUATION) TOWARDS EACH '

SPECIFIC DEGRADATION MECHANISM e PROVIDE PRE-JOB BRIEFINGS PRIOR TO EACH WELD INSPECTION TO FOCUS ON THE EXPECTED PARTICULAR DEGRADATION MECHANISM FOR THE COMPONENT TO BE INSPECTED l NOTE: EXAMINATION TECHNIQUES WILL NOT PRECLUDE FINDING OTHER DEGRADATION MECHANISMS.

^ t 1 p

i VERMONT YANKEE .

l CODE CASE N-560 L CATEGORY B-J WELDS TYPICAL EXAM VOLUME FOR N-560 (EG. THERMAL FATIGUE CRACKING)

Profile of component ,

V,

/

/' Weld buttering

I ,_ f. ' (where applied) 1 t-( jiU 1

\ d

(>

\

\

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I

} t i I B. (::f _C 6

) ###s's'N6 4 h'55555?

F'55555555:$.:-:T

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4 4 1/ +3t 6'.W<4 4 4 j u i

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\D Examination t

" " Volume A-B-C-D 1/4" 1/4" Examination Volume for Thermal Fatigue Cracking in Piping Welds NPS 4 or Larger.

M i

)

i VERMONT YANKEE -

CODE CASE N-560 CATEGORY B-J WELDS i

ASME SECTION XI EXAM VOLUME FOR PIPING WELDS Profile of volve body, vessel noule, or ,

Exam. surfece [ pump concection J

)

A-8

/

4 1/2In. 4-- --> 1/2 In. ,4- /

/' Weld end buttering A 8 (where appHed) v r v a r

)

}

C D r f

O- O g "3 ' h j

f F

O dh E h 'I I 1/4 in.-> 4 -- -> 4--1/4 in.

2 C- -E F (c) NPS 4 or Larger s

6

, n

~

i -

VERMONT YANKEE CODE CASE N-560 CATEGORY B-J WELDS QUALIFICATION OF NDE INSPECTORS i

• . QUALIFIED IN ACCORDANCE WITH 1988 EDITION OF ASME SECTION XI, IWA-2300 AND SNT-TC-1 A 4

• TYPICALLY, VERMONT YANKEE UT INSPECTORS:

]

• ARE CONSIDERED AMONG THE BEST IN THE INDUSTRY

• HAVE IGSCC CERTS e HAVE PASSED PDI (FOR VY'S MOST RECENT OUTAGES)

• HOWEVER, FIRST BULLET IS ONLY COMMITMENT

FIVE PRINCIPLES'OF RISK-INFORMED REGULATION

1. Meet Current Regulations

2. Maintain Defense In Depth

3. Maintain Sufficient Safety Margin

4. Proposed Increase In Risk (including cumulative effects) Are Small, Safety Goals Not Exceeded

5. Performanced-Based Implementation & Monitoring Strategies D

ENCLOSURE 9 j

- , . - - s Meet Current Regulations 10CFR50.55a (g)(4) - Codes and Standards e throughout the service life, components which are classified as ASME Code Class 1,2,3 must meet the requirements set forth in Section XI of the ASME B&PV code (specifically Table IWB-2500)

. Code Case N560 is an acceptable alternative to the requirements ofIWB-2500 ofSection XI of the ASME B&PV code Generic Letter 88-01 (NRC Position on IGSCC in BWR Austenitic SS Piping

• substantial piping replacement with IGSCC resistant material e

only three potentially susceptible locations remain (all B-F category components)

Generic letter 89-08 (ErosionK'orrosion-Induced Wall thinning) e No Change leak Detection Requirements

. No Change

,.yw. . --~+%---+-e-,- .i-,-r -+se- N'=**~*** ' ' ~ * ' + ' " ' " * * * ' ' " ~

suis uma I e

l Consequence Categories for  :

Pipe Failures Resulting in SystemRrain Loss

,l Affected Syelome l Number of Unaffected Backup Traine

! Frequency of Exposure Time to 0 1 2 :t 3 l Challenge Chclienge i

As year . _

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(Da cet m) (1-s monthe) < ,t ve m 2- - > .

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Long AOT P

/lAS47 1.dE-06 (1-2 weeks) g ,p,q . i - 13 w .1,8 :o e .o Short AOT(<3 daye) Ek) Met] 62.7E.48'O ' 3.7E-08'" ' 39 2.7E 10D

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.,y._o m, AS40 $a- )AS40 - m

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Short AOT f<3 days)

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n

Sufficient SafetyMargins Are Maintained Safety Function = Reactor Coolant Pressure Boundary Integrity Levelop And Maintain By:

• Quality Design - No Change,

• Quality Fabrication -No change,

• Quality Construction -No Change,

• Quality Testing - No Change,

• Quality Operation - No Change,

• Quality Inspection - Fewer inspection but conducted at more appropriate locations, using better techniques and as necessary expanded volumes (see activity 3)

. ..,.m.. . . _ . .. . . . . . . - -',..4- - - - - ^ ' - - ' - - * ' '

VERMONr YANKEE PRA RESULTS:

CDF = 4.3E-06/ year Lear = 9.4E-07/ year CHANGE m CDF DUE TO CHANGE IN THE NUMBER OF INSPECTIONS CATEGORIES CDF [1hR) DIFFERENCE IN # OF ACDF(1/yR]

(UPPER VALUE) btSPECDONS (UPPER VALUE) .

.. - -z,w re:"CA,T1

.~ 4 E-09 0 0 ruv~ mca acM,wwww&1 Rv.a. CAT 2Cuua 4.E-10 -5 2.50G-09 apr k

N.o._kyhT[4d mmd.m 2.E-10

.1 -,,,,m s%,

c u . CAT,4'w, il~s 2.E-11 -47 9.40E-10 y o.- ~, -

W~

s n,~CAT~A-. 55%, 2.E-11 -2 4.00E-11 CATE 1.5-12 -21 2.10E-11 CAT 7 1.E-14 -6 6.00E-14 TOTAL -81 3.005-09 g

1

t 1.000% --

, = _= _= _= _= _= _ _= _=

._______=___=______=_ _- _ _= _= _= _= _=- _=_=_=_. _ _

, c ._

_____________- _______- a

. -- s s ._-_______________________.

_ __ _ _ _ 1 qpg- g3., C

~ _

13

. _ _ _F_ _ __

(1- ~ l)_RiffffR tiVAE0_ATlDW _ _ _-

o 100% ---_---- ._

' Jak:  !

a o -= =-= ===

== = = Ea ec=a ss e- 5 == == == == == == == ====== -

== == == == =.- mY

-- :u:::::::::::::::::::::.______,__________________.g<3 2

r. 7 2 10% <=====

5<

2

-  ::=_=_=_=_==================_

l w .____ _____________

O NON-RISK-SIGNIFICANT .__________________

z 1%  :================

c)

= __ = ___ _______

_==__=_=_=_=_==

0.1%  :

1E-06 1E-05 1E-04 1E-03 BASELINE MEAN CDF (/yr) e Quantitive Screening Criteria for Permanent Changes Impacting CDF o

k ____-___-_ _ -

0 Impact of N-560 on ASME Code  ;

Class 2 and Class 3 Piping Systems X-560 has No Impact on Class 2 or Class 3 Piping Systems

X-560 is an Alternative to Existing Code Rules for Class 1 l (Category B-J) Piping Only l All 108 Nuclear Units Operating in the U.S. today are Licensed to Existing Code Rules Existing Code Rules Provide an Adequate Level of Safety Plants Implementing X-560 will continue to use Existing Code Rules for Class 2 and Class 3 Piping Implementation of X-560 will Improve Plant Safety in Class 1 Piping while Existing Code Rules Maintain Safety for Class 2 and Class 3 Piping

~ -

W:MQ4d W3 3 7 EM'FMIE,s'dNEMf&d M $72$ iT D dMI YA,.Td Y ( M.IE.* A 2 ' t,EF METf[WN2 sw5Ugu usw T1Y@w .

• sW w-CMND Wibng&@4:;MA.quSM Comparison of Code Requirements by Class- ac  :

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• 1 Casey+ < c Requirements! ; (i) 1 (Requirementc) 7:4x Requirements d .. L ; i Req 0irements ;

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.e , . g 7- g . .g; g. . 2 I I I ASME Section XI Code Program - ) ASME Section XICode Program - g ASME Section XI Code Program - g SpecificaNy Excluded from ASME 7.5% Sampilng and also sutiect to No Sampung but sutdect to Sectkr1 XI Code Program - No 25% Samphng and also sutdect to l Leakage Testing l Leakage Testing I SampNng and No Leakage Testing Leakage Testing 1 Current s n u I Owner Defined integrity Additional Owner Defined lategrity I Additional Owner Defined integrity l Owner Defined Integrity Management Programs (e.g IGSCC l Management Programs (e.g., FAC, l Management Programs (e.g., FAC, l Management Programs (e.g FAC,

& FAC for BWR's) Feedwater Nozzles) MIC) g MIC) g g l l i ASME Section XI code Program - l l l 10% Sampling per Code Case N.668 g g g and also sut$ect to Leakage Testing

' ' ' same as current same as current same as current N560 Additional Owner Defined integrity I

! I Management Programs - Same as current l [ t f f f -

._.s

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%.' ~ h ,!-9,5

$m,w$ Impact.of.N560'on Typical. Plant Assuming. Population of 500 BJ Welds NT,  :

%[k5[ kII55!!

$$hpMQ[$ CLASS 11 $yf $h(g$[f^3fhk h CLASS III g B.O.P. @[j%L

?:iMa M$[bdibl#$p$M3d. CLASSI N[Myl k MEMnRLE# M gM;

$p r$dipd[k ph 34#ggbl MNSyfgNMN@iiMP$hhgnwmM$hhh 50M lih$

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, , s i I i Plant 1 No Change I

No Change I

No Change improved , , i Safety i I I l i I Radiation Estimated Reduction i No Change I

No Change I

No Change Exposure of 225 Man-Rem , , ,

I I I I Plant Estimated Reducti9n i I i I No Change i No Change i No Change Costs of $3 Milh.on i I I f f f

_ _ _ _ _ _ _ _ _