ML20211A016

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Summary of 970828 Meeting W/B&Wog Re Mgt of Aging Effects for Reactor Vessel Internals & Activities Related to Baffle Bolting Integrity Issues.List of Attendees & Their Affiliation & Handouts Encl
ML20211A016
Person / Time
Site: Oconee, Arkansas Nuclear, Three Mile Island  Duke Energy icon.png
Issue date: 09/16/1997
From: Anand R
NRC (Affiliation Not Assigned)
To:
NRC (Affiliation Not Assigned)
References
PROJECT-683 NUDOCS 9709240080
Download: ML20211A016 (23)
v • d • e


Text

F a no j p 4 UNITED STATES g } NUCLEAR REGULATORY COMMISSION

  • * . + ,o September 16, 1997 ORGANIZATION: Babcock and Wilcox Owners Group PROJECT: Babcock & Wilcox Owners Group License Renewal Program

SUBJECT:

SUMMARY

OF MEETING BETWEEN THE U.S. NUCLEAR REGULATORY C')MMISSION AND THE BABC0CK AND WILC0X OWNERS GROUP TO DISCUSS THE MANAGEMENT OF AGING EFFECTS FOR THE REACTOR VESSEL INTERNALS AND ACTIVITIES RELATED TO BAFFLE BOLTING INTEGRITY ISSUES On August 29 397, representatives of Babcock and Wilcox Owners Group (BWOG) met witc tne Nuclear Regulatory Commission (NRC) st6ff to discuss the BWOG-report BAW-2248. " Demonstration of the Management of Aging Effects for the Reactor Vessel Internals" July 1997, and activities related to baffle bolting integrity issues. A list of meeting attendees and their affiliation is provided as Attachment 1.

The BWOG presented the scope of BWOG Reactor Vessel Internals report and activities related to baffle bolting integrity. The handouts used in the BWOG presentations are provided as Attachment 2.

The BWOG provided an overview of the re) ort (BAW-2248) which contains an evaluation of the effects of aging on tu reactor vessel internals, including the plenum assembly, the core support shield assembly, the core barrel assembly, and lower internals assemblies. The report discusses the materials of construction. 3rograms and activities that manage aging effects. and addresses applica:le time limited aging analyses associated with reactor vessel internals. BWOG stated that the reactor vessel internals evaluation applies to Arkansas Nuclear One Unit 1. Ocone? Nuclear Stations. Units 1.2.

and 3, and Three Mile Isiand Unit 1 plants.

TM BWOG discussed the B&W tlaterials Committee baffle bolt activities. Baffle bolt degradation was reported in European plants. In October 1996, the BWOG initiated a multi-)hase program to evaluate and assess the impact of potential bolt cracking in t1e B&W plants. Baffle bolt degradation has not been observed in domestic B&W plants, and Jaaanese utilities hava not recorted bo!t cracking, .The 8WOG described the core Jaffle assembly as having pressure relief holes in the baffle plates, upflow cooling and bolt cooling holes in the former plates. The pressure relief holes relieves differential pressure across baffle plate during operation and LOCA events. The holes in the former ppm l

plates provide cooling across baffle bolts to reduce the temperature.

The BWOG discussed an analytical aaproach to determine if there were concerns with regard to bolt degradation. 3WOG stat / that a very conservative analytical core baffle model was used. Maximum accident loading was ap) lied to the cure baffle model . In the structural analysis, it was assumed tlat all the baffle bolts. except the bolts in the top, bottom. and mid level former. DYD l 9709240080 970916 PDR AbOCK 05000269  !

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2 failed. The analysis indicated the baffle remained intact. It was estimated that even if 75% of the bolts located in the three formers were missing, the baffle assembl., would remain intact.

The BWOG indicated that materials and fluence may contribute to the aging process. The B&W piantr b3ffle bolts were fabricated from solution heat treated Type 304 stainless steel material, whereas the material used in the bolts in foreign plants was 316 stainless steel cold worked type.

The BWOG further stated that a Joint Baffle Bolt Task Team has been formed to evaluate baf fle bolt cracking. Westinghouse Owners Group along with other members are on the task force. The BWOG program includes: 1. Obtaining and evaluation of all available inspection data: 2. Development of an understanding and predictive capability for degradation of existing reactor vessel internals materials: 3. Development and qualifying replacement bolting material: and 4. Providing a vehicle for interaction of international and domestic working groups addressing reactor vessel internals aging effects.

The NRC staff encouraged the BWOG to continue its dialogue with the staff on the development of a ]rogram to manage reactor vessel internals aging effects and prepare for possi)le augmented baffle bolt inspection during the next inservice inspection at the Oconee Plant Unit 1. The BWOG requested the staff to provide an estimate of cost for the review of their topical report BAW-2248 and provide a draft safety evaluation in May 1998, and a final safety evaluation by December 1998. ,

G. DN.

Raj K. Anand. Project Manager License Renewal Project Directorate Division of Reactor Program Management office of Nuclear Reactor Regulation Project No. 683 Attachments: 1. Attendance List

2. Meeting Handouts cc w/ attachments: See next page 1

l l

v

- 2.- September 16, 1997 failed. "The analysis indicated the baffle remained intact. It was estimated

-that even if 75% of the bolts located in the three formers were missing, the baffle-assembly would remain intact.

The BWOG indicated that materials and fluence may contribute to the aging process. The B&W plants baffle bolts were fabricated from solution heat treated Type 304 stainless steel material, whereas the material used in the bolts-in foreign plants was 316 stainless-steel cold worked-type.

The BWOG further stated that a Joint Baffle Bolt Task Team has been formed to evaluate baffle bolt cracking. Westinghouse Owners Group along with other members are on the task force. The BWOG program includes: 1. Obtaining and evaluation of all available inspection data: 2. Development of an understanding and predictive ca) ability for degradation of existing reactor vessel-internals materials: 3 Jevelopment and qualifying replacement bolting

material: and 4. Providing a vehicle for interaction of international and domestic working groups addressing reactor vessel internals aging effects.

The NRC staff encouraged the BWOG to continue its dialogue with the staff on the development of a arogram to manage reactor vessel internals aging effects and prepare for possi)le augmented baffle bolt inspection during the next -

inservice inspection at the Oconee Plant Unit 1. The BWOG requested the staff to provide an estimate of cost for the review of their topical report BAW-2248 and provide a draft safety evaluation in May 1998. and a final safety evaluation by December 1998.

Ce .  %)

Raj K. Anand. Project Manager License Renewal Pruject Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Project No. 683 Attachments: 1. Attendance List 2.-Meeting Handouts cc w/ attachments: See next page

DOCUMENT NAME: A:B&WMEETI.MLM (it Anand/LLM Disk)

' To receks a copy of th6e document. Ind6cete in the bos: 'C' = Copy without attchment/ enclosure 'E' = Copy with attachment / enclosure

  • N' = No copy - 'SEE PREVIOUS CONCORRENCE OFFICE SPM:PDLR LA:PDLR PM:PDLR* EMEB* PT- D:PDLR C C' R .b a m l , FOR: /Td\

NAME: RAnand:ach EHyto V ' WLiu RWessman CGriT6esi DATE 09/16/97 09//b/97 09/16/97 09/16/97 09/l(/97 0FFICIAL RECORD COPY

, _ , , - - . - , , - - ~ = - . , . ~ , - -

Meetina Summary HARD COPY iDocket'Filea-PljBLIC PDLR R/F OED0 RIV Coordinator. 0-17G21 E-MAIL:

S. Collins /F, Miraglia (SJC1/FJM)

R. Zimmerman (RPZ)

M. Slossom (MMS)

S. Weiss (SHW)

P. Shemanski (PCS)

M. Pratt (MDP)

-R. Correria (RPS)

R. Wessman (RFM)

J. Strosr,1 der (JRS2)

5. Droggitis-(SCD)

S. Peterson (SRP)

G. Lainas (GCL)

8. Morris (BMM)

J. Moore (JEM)

G. Mizuno (GSM)

G. Holahan (GMH)

B. Sheron (BWS)

M. Mayfield (MEM2)

A. Murphy (AJM1)

H. Bramer (HLB)

L. Shao (LCSI)

G. Bagchi (GXB1)

R. Johnson (REJ)

D. LaBarge (DEL)

H. Berkow-(HNB)

PDLR Staff

a Project No. 683 Babcock & Wilcox Owners Group Generic License Renewal Program cc:

Mr. Robert B. Borsum Regional Administrator. Region IV Framatome Technologies U.S. Nuclear Regulatory Comission 1700 Rockville Pike 611 Ryan Plaza Drive. Suite 1000 Suite 525 Arlington. Texas 76011 Rockville. Maryland 20852 Mr. Dwicht C. Mims Michael Laggart Director. Licensing Manager. Corporate Licensing Entergy 0)erations. Inc.

GPU Nuclear Cor) oration Route 3. Box 137G One Upper Pond Road Russelville. Arkansas 72801 Parsippany. New Jersey 07054 Earnest L Blake, Jr.. Esq.

Chairman Shaw Pittman. Potts Board of County Commissioners and Trowbridge of Dauphin County 2300 N. Street. NW Daughin County Courthouse Washington D.C. 20037 Harrisburg Pennsylvania 17120 Regional Administrator Region I Mr. W. R. McCollum Jr. U.S. Nuclear Regulatory Commission Nuclear Generation Vice President 475 Allendale Road Duke Power Company King of Prussia. Pennsylvania 19406 Oconee Nuclear Station MC: ONO IVP William Dornsife. Acting Director P.O. Box 1439 Bureau of Radiation Protection Seneca. South Carolina 29679 Pennsylvania Department of Environmental Resources Mr. John R.- McGaha P.O. Bcx 2063 Vice President. Operations Support Harrisburg, Pennsylvania 17120 Entergy Operations. Inc.

P.O. Box 31995 Chairman Jacksonville. Misr.issippi 39286 Board of Supervisors of Londonderry Township Regic.ial Administrator. Region 11 R.D. #1 Geyers Church Road U.S. Nuclear Regulatory Commission Middletown. Pennsylvania 17057 101 Marietta St.. N.W. Suite 2900 Atlanta, Georgia 30323 Mr. J. E. Bachfield Compliance Mr. R. L. Gill Duke Power Company GLRP Licensing Coordinator Oconee Nuclear Site c/o Duke Power Company P.O. Box 1439 EC-12R Seneca. South Carolina 29679 '

P.O. Box 1006 Charlotte. North Carclina 28201-1006 l

ATTENDANCE LIST NRC MEETING WITH BABC0CK & WILCOX OWNERS GROUP CONCERNING AGING MANAGEMENT OF THE REACTOR VESSEL INTERNALS Auaust 28. 1997 NAMf QRDANIZATION

1. Raj Anand NRC/NRR/PDLR
2. Wan C. Liu NRC/hRR/PDLR
3. H. F. Conras NRC/NRR/DE/EMCB
4. C. 1. Grimes NRC/NRR/PDLR
5. Dick Wessman NRC/EMEB
6. David J. Firth FTI/B&WDG GLRP
7. Bob Borsun FTI
8. F. M. Gregory FTI
9. Jeff Gilreatham Duke Power
10. Richard Miller GPU Nuclear
11. Kurt Cozens NEI
12. Bill Mackay Entergy - ANO
13. Prasoon Goyal Toledo Edison
14. David Mas 1ero GPU Nuclear
15. Lee Banic NRC/NRR/DE\EMCB
16. Mildred McNeil NRC/RES/DET/EMMEB
17. Edmund Sullivan NRC/NRR/DE/EMCB '
18. Francis Grubrick NRC/NRR/DE/EMEB
19. Lambros Lois NRC/NRR/DSSA/SRXB
20. Barry Elliot NRC/NRR/DE/EMCB 21.-Steve Hoffman NRC/NRR/DRPM/PDLR
22. Ian L.W. Wilson Westinghouse NSD/WOG Tech L.
23. David E. Whitakee Duke Energy
24. H.L. Brammer NRC/NRR/DE/ECGB
25. Bill Gray FTI (B&GOG Materials Committee)
26. Greg Robinson Duke Power
27. Bob Gill .

Duke Power

28. Steve Fyfith Frontier Technologies. Inc.
29. Daniel Spond Entergy Operations Inc.

30, Kamal Manoly NRR/DE/EMEB Attachment 1

I / AGING MANAGEMENT l B&W REACTOR VESSEL INTERNALS

~ ~ ~ ~..,, ~ , - - nc .- - -.. - - ,.~n - - ,- ~~

Nuclear Regulatory Commission Staff Briefing August 28,1997 W.11. (Ilill) Mackay, Entergy Operations. Inc.

D. J. (Dase) Finh, Framatome Technologies. Inc.

F. M. (Frank) Gregory, Framatome Technologies, Inc.

S. (Stese) Fyfitch, Framatome Technologies, Inc.

d __

  • Agenda

+ Introduction (Dave / Bill)

+ B&W Owners Group (B&WOG) Generic License Renewal Program (GLRP) Reactor Vessel Internals Report (Frank)

+ RV Internals Baffle Bolting Emerging Integrity Issues (Frank)

+ B&WOG Materials Committee RV Internals Evaluation Activities (Steve / Frank) b 1 Attachment 2 .

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-a r Joint NRC/GLRP Accomplishments

~~_...-_,...-n.~.,,_..,n_--._...--

+ Reactor Coolant System Piping Report /SE -

BAW-2243A issued (6/96)

+ Pressurizer Report /SE - B AW-2244 (8/18/97)

+ Reactor Vessel Report in review cycle; all RAI responses complete - B AW-2251 (8/l 1/97)

+ Reactor Vessel Internals Repart submittal -

BAW-2248 (7/29/97)

eeuses soeur Sl hl/

~ '- Purpose of Meeting 78

Overview of GLRP RV Internals Topical Report
Recognize that of the aging issues addressed in License Renewal, baffle bolting integrity issues may be an emerging issue
Discuss B&WOG Materials Committee e o.cobaffle bolt activities d.h!.

l 2

1 i

I, 9-- Introduction

. ~ - . _ - ~ .. _ . _ . ... _ . ~ .~ ~

+ Joint B&WOG GLRP/ Materials Committee presentation

+ GLRP RV Internals Report (BAW-2248)

+ Expectations t Report review / overview

  • Solicit NRC input / questions N$d M~~ Introduction (continued) 7R
. B&WOG GLRP has identified aging effects for RV Internals components
  • Whi!e investigating these effects, and considering worldwide industr) operating experience, a potential emerging issue was recognized with bafne bolt cracking e B&WOG and industry - increased focus more detailed investigation

.: Cracking of baffle bolts is not a safety issue for the B&W-design plant based on reasonable assumptions

+ Discuss B&WOG program development and plans for

= "** future discussions N$d 3

1

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> Expectations

_,,.~__,.,-__.._.m_ . _ _ _ _ , .

+ NRC review fee cost estimate acceptable

+ Safety Evaluation (SE) development request by GLRP

+ Draft SE by May,1998 +

+ Final SE no later than December 1998

+ Target data factor in hcense renewal apphcation submittal decision / schedule

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REACTOR VESSEL INTERNALS REPORT BAW-2248 4

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Purpose a.-~~~~,-___,--,-.._n- -. _ . ~ ...-

+ Demonstrate, for the reactor vessel internals described, that the effects of aging will be adequately managed so that the intended function will be maintained consistent with the CLB for the period of extended operation

+ Time limited aging analyses, as defined by 10CFR54.21(c), have been addressed

=11

\l[ Scope of Report Addresses effects of aging for B&W-design 177-FA lowered loop reactor vessel internals for:

+ ANO- 1

< ONS-1, ONS-2, ONS-3

+TM1-1 5

._____.m_ __ _ . _ _ _ _ _ _ _ _ _ _ _ - _ . - - - - - - - - - - - _ . - -

Report Content 7.R _-m.,. _. 4-,,.m_ _ ,~. . m m_ _

+ Define component intended function (s)

+ Describe component, including materials of construction

+ Define applicable aging effects for material, environment, and stress combinations

+ Identify programs that manage aging effects

+ Address applicable TLAAs O

I

~ Reactor Vessel Internals' Functions 78 aw Provides:

+ Support and orientation of the reactor core

  • Support, orientation, guidance and protection of the control rod assemblies

+ Passageway for the distribution of the reactor coolant flow to the reactor core

+ Passageway for support, guidance, and protection of incore instrumentation o Secondary core support, limiting the downward

,,, displacement of the core support stiucture for a

~ postulated core barrel failure L:d.

6

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,e Figure 1 1 Reactor Vessel Interstals (Cross Section) ,

9 rm cm c;3 em em rra

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Plenum Cover I. -

I! Assembly Cor. 11 98 t.

ly a,

hidn N f

_.] 0 y Vent Valves (notin scope) hs V  !

Core Support M '

Plenum f4'~ - Control Rod c ( Assembly ,~.

Guide Tube Shnend -

Assembly ) Assembly e ,a i

, g

=r

. mm .. r 5 se bly

' y _ Thermal Shield g (notin scope) t  :

1 - , , . ... ... .. ... ... . . .

-', e i Core Barrel Core Barrel  ; f Assembly "'

li A

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Con 4 ,

Baffle Plates (notin scope) i'-~

4 -

~

ormer Plates l

Lower #

hk bdbd3b bbA g Lower Grid I"I'*'"

Assembly -

s. -

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{% . f.7-I!

1- 4 N n

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/ ~ Flow Distributor o e Guide Tubes u

i Design Basis

_...___...-._1.___.__e - _ . _ , . , _ . , ~

+ Pre-dates ASME design criteria for core suppoit structures

+ Qualification combination of analysis and test

+ In place testing documented in BAW-10038 and BAW-10039

+ Faulted analysis documented in B AW-10008

e TE b.l[
  • ~ Aging Fffects Considerations

+ Cracking

+ Reduction of fracture toughness

+ Loss of material

^ Loss of closure integrity

+ Mechanical distortion and/cr ratcheting *

  • Precluded due to nornulAqwet conditions well below temperatun: thresnold for creep
.e.ees esaw 7

.l ,

y B&WOG Challenges w w+- > n - ua su s.m s : u s .. ~ :v susnmen,e + n5n ..<vs su mesw.e,u.w>.x r u .xcu

+ Credit existing program to address:

  • Potential loss of material due to wear

+ Develop future programmatic actions to address:

  • Potential loss of closure integrity due to stress relaxation of bolts
  • Reduction of fracture toughness due to irradiation
embrittlement Q

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-=  :- Programs Managing Aging EfTects 7R

ASME Section XI Examination Category B-N-3

+ VT-3 visual examination of accessible surfaces

+ RVI Aging Management Program e B&WOG Program to manage aging effects, so that the intended funcuon will be maintained consistent with the CLB

+ Program will carry forward into the period of extended operation for those pursuing renewal license i

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. - Time Limited Aging Analyses

- . . .- ~,. .n._ ~.- .o ,~.-...-,n-;

Fatigue e liigh cycle fatigue (flow induced sibration) e Low cycle fatigue (design transients) e Existing analysis valid for 60 years
Fracture Toughness e Effects of irradiation on material properties for faulted loads

+ RV Internals Aging Managernent Program to address

._,_ this is.;ue K'bl.

hl/  :- Purpose of Meeting "7"8

Overview of GLRP RV Internals Topical Report
Recognize that of the aging issues addressed in License Renewal, baffle bolting integrity issues may be an emerging issue
Discuss B&WOG Materials Committee

. .,_ baffle bolt activities S!.

9

TIMELINE OF EVENTS LEADING TO IASCC AS AN EMERGING ISSUE '

Page 1 of 1 8/22/1997 1988 1989 1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003

-~

B&W Life Extension Study (BAW-2060) performed for Duke  ; L JL10/88 .

Power Company and EPRI 1/88 PWR RV internals License Renewal '

Industry Report prepared for JL JL JL EPRl/ DOE /NUMARC 9/89 12/91 7/94 .

First known publication of baffle 1

bolt failures in French nuclear A plants 9/94 -

B&WOG Steering Committee approval for Materials Committee A

, program to evaluate IASCC issues 10/95 B&WOG GLRP RV Internals Report (BAW-2248) If95 7 97 C&WOG Materials Committee Program on baffle bolt aging j( j(

issues 1/96 12/97 7

Joint Baffle Bolt (JcBB) Task Team JL j(

l 10/96 1WO1 B&WOG GLRP/ Materials l Committee preliminary safety JL Jk8/97 l

cssessment of baffle bolt cracking 1/97 l B&WOG Material Committee 7 program to address RV Internals JL aging management issues e/97 i

'4 Haffle Bolt Integrity Emerging Issue Activities .

$ 44. .+ 4 3 . t ' ise .. . 04 % ah a.r % S *4 1.2 E E L2hTSE &A AT

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+ Materials Committee has been actively evaluating RV Internals bolting issues for several years

+ International and domestic events coupled with CLRP long range renewal planning have reprioritized Materials Committee efforts

+ Materials Committee will handle future technical issues and is currently defining a new program S

hl/- Baffle Bolt Inspection Data

-e

~78

+ Indications of cracked bolts have been identified in European nuclear plants e The number of plants with indications have been restricted to the early designs e The number of bolts with indications have been relatively small

+ Cracked bolts were fabricated from cold worked Type 316 stainless steel material e Most, if not all, bolts with indications located in plants

-i - convened from downflow to upflow design b

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Baffle Bolt Inspection Data (continued)

_.~.._,m.__ _ _ , , _ _ . _ _ _ . -

+ No bolts with indications have been identified at plants with original upDow design, pressure relief holes in baffle plates, and former plate cooling holes

+ Japanese have not reported any indications of cracked bolts from their inspections

=1.!

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e:- Key B&W-Design Features 78

+ Low differential pressure across baffle plates during operation and LOCA events e Upnow design

Baffle plate pressure relief holes / slots e Fourth former plate (from the bottom) offset from baffle-allows cooling of baffle bolts at this elevation 1
eenses esow i

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~ Key B&W-Design Features (continued)

- ,.-...~. -. -;. _ _

_ . - _ . ~ . - -

+ Baffle bolts fabricated from solution heat treated Type 304 stainless steel material

+ All these design features are applicable to ONS-1, ONS-2, ONS-3, ANO-1, CR-3, D-B, and TMI-l

=i=1 hlf{* B&WOG Safety Assessment 78

+ Baffle bolt cracking is not considered to be a safety issue for all B&W-design plants

+ Assumed practical configuration based on IASCC susceptibility of baffle bolts 12

I-9(I B&WOG Safety Assessment (continued)

. _ > .. ~ _ _ , _ . _ _ ~ . . - .m.--_ m _ ,, _ ..,

+ LOCA analysis-No safety concern e Small pressure differential due to B&W design features e Pressure pulse shape of short duration e Load generally in direction of former plates

+ Seismic analysis-No safety concern

  • No basemat amplification through reactor vessel and core barrel e Displacements small and locations not coincident with fuel assembly displacement 4

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1[=- Baffle Bolt Activities 7N

.:- Joint Baffle Bolt (JoBB) Task Team has been formed to evaluate baffle bolt cracking:

Members include B&WOG. FTI, EdF, EPRI, Westinghouse, WOG Other members also being pursued

+ Major goals and objectives:

Obtain and evaluate all available inspection data Develop an understanding and predictive capability for degradation of existing reactor vessel internals materials Develop a,d qualify replacement botting material Provide a vehicle for interaction of international and domestic

. ,,,,,,,, working groups addressing RV Internals aging effects

~

13

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- Haffle Bolt Activities (continued)

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A AM$*$ 4 ,P uMfM&l n.X MA in ry m 7 Y -L T.A%s% .J/4) tpy ; $- ' rf

+ 1996 B&WOG Materials Committee Program

+ Identified bafSe bolt designs, material, and locations in B&W design plants

+ Calculated accumulated fluence of baffic bolts for 40 and 60 years of operation

+ Identified concepts for NDE of the various baffle bolt designs

=53.

= ~

d llafile llolt Activities (continued)

+ 1997 B&WOG Materials Committee Program

+ Preparation of preliminary safety assessment of baffle bolt cracking

+ Participation in JoBB activities

  • Preparation of an industry whim paper on IASCC information

+ Calculation of accumulated fluence profile for entire reactor vessel internals for 40 and 60 years of operation

+ Identification of other reactor vessel internals component items with relatively high fluence levels i

14

I Baffle Bolt Activities (continued)

W{ I

, i.---,. n . .- - .. cc n - ~ a - - - , - - - --

+ Current and ongoing B&WOG activities

  • Develop multi year program to manage reactor vessel internals degradation issues
  • Preparation for possible augmented baffle bolt inspection during next 10-year ISI interval (2003 at the earliest)
  • Maintain cognizance of ongoing industry activities

+ Maintain open dialogue with the NRC

Sd e *.v Summary and Conclusions 7R

+ Aging management review performed in B AW 2248 provides framewoit for planning for future operation

+ BAW-2248 identified baffle bolt cracking as requiring additional aging management o B&WOG recognizes this as an emerging industry issue e B&WOG initial work indicates that baffic bolt cracking is not a si:fety issue for all B&W-design plants

+ B&WOG developing a program to manage RV  ;

Internals aging effects l

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