ML20198N609

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Rev 0 to FPC Calculation F-98-0010, FTI Calculation 32-5000218-00,Adjusted Ref Temperatures for 32 EFPY for CR-3
ML20198N609
Person / Time
Site: Crystal River Duke Energy icon.png
Issue date: 09/30/1998
From: Knoll R, Miskiewicz D
FLORIDA POWER CORP., FRAMATOME
To:
Shared Package
ML20198N568 List:
References
F-98-0010, F-98-0010-R00, F-98-10, F-98-10-R, NUDOCS 9901060167
Download: ML20198N609 (38)
v • d • e


Text

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

t FLORIDA POWER CORPORATION 4

CRYSTAL RIVER UNIT 3

. DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 1

ATTACHMENT B l

l FPC Calculation F-98-0010, Revision 0 i "FFI Calculation 32-5000218-00, Adjusted Reference Temperatures for 32 EFPY for O 3" i

i 9901060567 981231 PDR ADOCK 05000302';  ;

P ,, PDRd g .1

_._ - m._. _ . . - _ _ _ ~ _- ~ _ . _ _ _ _ . _ _ _ _ _ _ ___ __ __

Florida MI INTEROFFICE CORRESPONDENCE A-c-xuTL.PRu l Nuclear Engineering NT01 240-3019

) 0"iC8 MAC Telephone

SUBJECT:

Crystal River Unit 3

Quality Record Transmittal - Analysis / Calculation TO
Records Management - NR2A

, The following analysis / calculation package is submitted as the QA Record copy:

DOCNO (FPC DOCUMENT IDENTIFICATION NUMBERI REV. SYSTEMISI TOTAL PAGES TRANSM!TTED l F-98-0010 TITLE 0 RC p ART for 32 EFPY for CR3 KWDS (IDENTIFY KEYWORDS FOR LATER RETRIEVAU Reactor Vessel ART DXREF (REFERENCES OR FILES . LIST PR. MARY FILE FIRST)

F-97-0013 VEND (VENDOR NAME) VENDOR DOCUMENT NUMBER (0XREF) SUPERSEDED DOCUMENTS (DXREF)

FTl l 32-5000218-00 -

RCT-1 l l l l -

l I l l COMMENTS (USAGE RESTRICTIONS. PROPRIETARY. ETC.)

Provides input to reactor vessel integrity issues NOTE:

Use Tag number only for valid tag numbers (i.e., RCV-8, SWV-34, DCH-99h otherwise, use Part number field (i.e., CSC14599.

AC1459), if more space is required, write "See Attachment" and list On separate sheet.

"FOR RECORDS MANAGEMENT USE ONLY **

Quality Record Transmittal received and information entered into SEEK.

Entered by: Date (Retum copy of Quality Record Transmittal to DE Support Specialist.)

DESIGN ENGINEER DATE VERIFICATION ENGINEER DATE SUPERVISOR, DESIGN ENG DAT 0h 9 $$ Yk '

i ~ " s s-cc:

Nuclear Projects Of MhCGWR/PEERE Calculation Review forrn Part til actions required Yes No O Return to Service Related) C Yes @ No Of Yes, send copy of the Calculation to the Responsible Organizationts)

Supervisor, Config. Mgt. Info. identified in Part til on the Calculation Review form.)

c_ Mgr., Design Engineering (Original) w/ attach

$ Mgr., Radiological Emergency Planning w/ attach Yes @ No Rav E'98

/

\g

( JPowe ANALYSIS /CALCULA'lON

SUMMARY

\ %s' / senseeee..r. ACSUM F7.M COCUMENT OENTIFICATION NUMBER DISCIPUNE CONTROL NO. REVISIO' ML F 98-0010 [ 9p.Ooj o 0 TITLE CLASSIFICATION (CHECK ONE)

ART for 32 EFPY ic-CR3 Safety Related Non Safety Related MAR /SPCGWR/PEERE NUMBER VENDOR DOCUMENT NUMBER 32-5000218-00 APPROVAL PRINTED SIGNATURES NAME Design Engineer (), d, s D. N. Miskiewicz D:te 9lzglqg D V;rification Engineer #

g D:te Supervisor ' - ~

[f  % m- R. W. Knoll D:te 'f/y./? (

ITEM.s REVISED PURPOSE

SUMMARY

D termine the adjusted reference temperatures (ART) for the limiting materials and locations in the reactor vessel.

RESULTS

SUMMARY

Revised ART values are provided to support reactor vessel integrity analyses as needed.

R;v. 3/97

p-1 N JPower

/*******"" '

CALCULATION REVIEW CALCULATION NO1RE/.

F-98-0010, Rev.0 ~

ART for 32 EFPY for CR3 PARTI-DESIGN ASSUMPTION /lNPUT REVIEW: APPLICABLE Yes C No Thecalculation:

this following organizations have reviewed and concur with the design assumptions an Systems Engineering

_M ,

g-;7 7p Nuclear Plant Operations OTHER(3)

Sgnau g hk[ 9/z/</g s.gna .Da,e SgnauetDate --

PART 11 -

RESULTS REVIEW: APPLICABLE @ Yes No The following organizations have reviewed and concur with the results of this calculation and understand the actions which the organi2ations must take to implement the results.

System Engineering Comments: E%:, eud a ll cm, fs an

,MI E ,27 - 79

b. A; r kla vie n A

Nuclear Plant Operations

[ 'f/ Z/i&

Comments:

Nuclear Plant Maintenance Yes N/A "***'

Comments:

Nuclear Licensed Operator Training Yes X N/A "*"*'

Comments:

Manager, Nuclear Regulatory Compliance O Yes "'****

G N/A Comments:

Sr. Radiation Protection Engineer Yes **"*

@ N/A Comments:

Nuclear Plant EOP Group Yes . N/A "'****

l Comments:

Design Engineering Oyes N/A SenawwDate Comments:

OTHER: 0Yes GN/A Sgna .Da SgnauesDate

, non a se

f-((.y)fg. CALCULATION REVIEW

(

l cALCuLAfloN NosREY.

Page 2 of 2 F-98-0010, Rev.0 N ART for 32 EFPY for CR3 PART111 CONFIGURATION CONTROL: APPLICABLE No PCC # Yes ~

The following is a list of Plant procedures / lesson plans /other documents and N calculations which require updating based on calculation results review:

Document Date Required Responsible Organization

~

1 l

I l

Upon completion, generate a Precursor Card in accordance with CP-111 for tracking of actions identified in Part updated to reflect this Ill, if calculations are listed, a copy shall be sent to the original file and the calc impact.

PARTIV - NUCLEAR ENGINEERING DOCUMENTATION REVIEW The responsible Design Engineer must thoroughly review the below listed documents to assess if the calculation requires revision to these documents, if "Yes," the change authorizations must be listed below and issued corcurrently with the calculation.

Enhanced Design Basis Document O Yes E No A Vendor Qualification Package FSAR Yes @ NoA C Yes @ No ""'"

Improved Tech. Specfication Topical Design Basis Doc.

O Yes E No A O Yes >2 No **"'8 E&SQPM tmproved Tech. Spec. Bases Yes @ NoA O Yes P2 No t"-'"

otner oocumeats revie*ed:

Config. Mgmt. Info. System Yes b (c mqgongoq Cyes No Analysts Basis Document (CHANot OOC. REFERENCE)

Yes @ No A O Yes O No D; sign Basis Document C Yes @ No A (CMANGE OOC. REFERENCE)

O YesO No App;ndix R Fire Study scanNoe ooc arremeNcu C Yes @ No A O YesO No a:naNos occ, REFER 2NCE)

Fira Hazardous Analysis C Yes @ No A Yes C No g:anNos occ.merameNco NFPA Code Conformance Document O Yes A Q No C Yes C No ecNaNon occ.asPERENCO PART V - PLANT REVIEWS / APPROVALS FOR INSTRUMENT SETPOINT CHANGE PRC/DNPO proces3. approval is required if a setpoint is to be physically changed in the plant through the NEP-213 PRC Review Required O Yes @ 'No PRC Chairman /Date DNPO Review Required C Yes @ No DNPO /Date DE3:GN ENGINEEA/0 ATE DESIGN ENGINEER . PRIN TEo NAME

%. [

e .

RM 5/981 ,

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

20697 3(12/95)

= '

CALCULATIONAL

SUMMARY

SHEET (CSS) l- E.Rd.DNo o,y,E .

DOCUMENT IDENTIFIER 002 W O TITLE Adjusted Reference Temperatures for 32 EFPY for CR 3 PREPARED BY:

REVIEWED BY:

! NAME M.J.DeVan NAME L. B. Gross l SIGNATURE / SIGNATURE Engineer IV TITLE DATE C[1/qy TITLE Adviso Engineer

! COST CENTER 41020 REF. PAGE(S) 31-33 DATE [ 97

?URPOSE AND

SUMMARY

OF RESULTS:

TM STATEMENT: REVIEWER INDEPENDENCE

[

l Th3 32 cffective full power years (EFPY) projected adjusted reference temperature values at the %

and %-thickness (%T) locations of the Crystal River Unit 3 reactor vessel beltline region were calculated in

! ccesrdInce with Regulatory Guide 1.99, Revision 2.

The limiting %T and %T adjusted reference temperature

' vilum at the end of 32 EFPY are 213*F and 144.5'F respectively. This data will be used to calculate new pressure-timpIrtture operating limit curves for the Crystal River Unit 3 reactor vessel.

l t

i l

l l

THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT:

, CODE / VERSION / REV CODE / VERSION / REV THIS DOCUMENT CONTAINS ASSUMPTIONS THAT MUST BE

. VERIFIED PRIOR TO USE ON SAFETY-RELATED WORK NO ( X)

YES ( )

PAGE 1 OF 33 w-

FRAMATOME -

TECHNOLOGIES FTINON-PROPRIETARY 32-5000218-00 RECORD OF REVISION Revision '. ~

[.3f[CE.': bit'~,I['~9.'De's'dripiibn... f 7.g 2 .. .

00 Original Release _

PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 PAGE2

i FRAMATOME -

TECHNOLOGIES FTI NON-PROPRIETARY 32-5000218-00 1

TABLE OF CONTENTS  !

TITLE PAGE .

1 1

1. 0 i n t ro d u ctio n . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . _ . . . . . . . . . _ . . _ . . _ . _ _ . _ _ _ . . _ _ _ . . .
2. 0 S u m m a ry o f R e s u lts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3 . 0 A s s u m p ti o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ~ . . ~ . ~ . . ~!

4. 0 R e a ct o r Ve s s e l Flu e n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

4.1 Reactor Vessel Inside Surface Fluence.............................................................. 5 1 4.2 Atten uatlon Through Vessel Wall ......... ............................................................. 5 i 5.0 Adjusted Reference Temperature Calculation Where No Surveillance Data isAvailable...............................................................................................................7 5.1 I n itial RTNor~ .. . . . ~ . ---- ~ ~ . . m m m . .~ ~ ~ ~ ~ . . -- ~ ~ ..~ ~ ~ ~ . -- . ..~ . . . . . . 7 I 5.2ARTNor................................................................................................................8 l 5.2.1 C h e mis t ry Fa ct o r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 5 .2. 2 Fl u e n c e Fa c to r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

5 c2. 3 A RL ,or C alc u la t io n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . .. . . . . . . 10 5.3 Margin...............................................................................................................11 5.4 Calculation of Adjusted Reference Temperature....... ...................................... 12 6.0 Adjusted Reference Temperature Calculation Where Surveillance Data i s Av all a b l e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

6.1 Calculation of Chemistry Factor Using Surveillance Data ................................14 6.1.1 Base M etal Heat N umbe r C4344-1 .... ..... .......................... ....... ........ .. ..... . 14 6.1.2 We!d Wire Heat 71249.............................................................................15 6.1.3 Weld Wire Heat 72105.............................................................................16 6.2 Adjusted Reference Temperature Calculated Using Surveillance Data ...........19

6. 2.1 i nitial RTuor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
6. 2.2 A RTn or C a lc u latio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . 1 9

[ 6.2.3 Margin......................................................................................................20 6.2.4 Calculation of Adjusted Reference Temperature ..................................... 21 t

PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 PAGE 3

FRAMA TOME -

_ TECHNOLOGIES I FTlNON-PROPRIETARY 32-5000218 00 ~

Appendix A - Generic Copper Centent for Base Metal Forging Materiais Fabricated by the Ladish Company........................................................ 22 Appendix B - Credibility of Surveillance Data..................................................... .......... 2 5 A ppe n dix C - R e fe re n c e s . . . . . . . . . . . . . . . . . . . . . . . . . ..... . .3.0. . . . . . . . .

h i-1 i

c

' PREPARER: M.J. DeVan DATE: 06fo9197 REVIEWER: L.B. Gross DATE: 06/09/97 PAGE4

FRAMATOME -

TfE;CHNOLOGIES FTINON-PROPRIETARY 32-5000218-00 l

1.0 INTRODUCTION

The purpose of this analysis is to determine the Florida Power Corporation Crystal River Unit 3 (CR3) reactor vessel adjusted reference temperature data for 32 effective full power years (32 EFPY). This information is to be used in the preparation of pressure-temperature operating limits curves applicable through 32 EFPY.

10

SUMMARY

OF RESULTS The adjusted reference temperature values applicable through 32 EFPY for the CR3 reactor vessel beltline materials are listed in Table 1. These values were calculated in accordance with the guidelines outlined in Regulatory Guide 1.99, Revision 2.' The controlling values of the adjusted reference temperature for the CR3 reactor vessel beltline material are 213.0 F at the %-thicknras (%T) wall location and 144.5 F at the

%-thickness (%T) wall location.

3.0 ASSUMPTIONS l

l No major assumptions are containen in this report.

4.0 REACTOR VESSEL FLUENCE 4.1 Reactor Vessel inside Surface Fluence 1

The extrapolated 32 EFPY inside surface fluences for the CR3 reactor  ;

vessel beltline materials are listed in Table 2.2 ,

4.2 Attenuation Through Vessel Wall in accordance with Regulatory Guide 1.99, Revision 2, the neutron fluence at the %T and %T wall locations in the vessel, f (x 10 n/cm 2, E>1 MeV),is determined as follows:

/ = f,, (e*' ) (1) where fur (10'8 n/cm 2, E>1 MeV) is the calculated value of the neutron fluence at the inner wetted surface of the vessel, and x (in inches) is the depth into the vessel wall measured from the vessel inner surface. The CR3 reactor vessel thickness is reported in BAW-1543, Revision 4 to be 8.44 inches. The 32 EFPY fluence values for the Cr 3 reactor vessel beltline materials at the %T and %T wall locations , a presented in Table 2.

PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 PAGE 5

FRAMATOME TECHNOLOGIES' ,

FTlNON-PROPRIETARY 32-5000218-00 ,

k Table 1. ~ Data for Preparation of Pressure-Temperature Limit Curvesa for Crystal River Unit 3 - Applicable Through 32 EFPY -

Chemical r tassonalDeectansa e . .

- ART F-  ?

32 EFPY FW sulun' -

ART. F [

et 32 EfPY tensgst Assess Wesses usa . Staat ce et 32 EFPY a Samune Rogues Lacement hisee.

le biensi Chemisery insides Beissituer Tyse ee% T/4 3 Pef YM ' 3r4T . f as% RT.ee Facser Sursace TI4 3r4T - TN L L- ^- *-- 3r4 T 2

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us,m,eh.8 , e 0 13 .. -e te,ar esise Ptose c.u4.t c4x4 9 .AA-ms. o,.

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i m E.i. 4=.t. tw.= ri.,

4r8 _ ,, 1.

C4344 2 u i.. imE.te .m. . ut, La.or ense rtsee C43471 C43442 C4347-t SA-533 Gs. Bt SA-533 Gr. Bl 0 20 a t2 ES4 O SS

+20

-13 141 e - 7.00E + 88 4 7eE+1e Sm.a 8.73E+te usS-t12 S '

ne-75 9 u.

~ 34

. .i.S5 . t= 0 La-ar ches Pimme . cour 2 cour-2 . et s 8.00Eete 403Ee18 175E+le 34 ' tes S 12e 8 SA-533 er. ss at2 0 Se .45 e5 8 44 4 34 i

- e2 4 oc0E*te 4 83Este 1.75E* te 34 see se e i es e 44 4 34 u, ue ee U$ ose. Wees po 40%) sA tree rt24e 144 4 823 4 AsAunesso One est ..te set e {

t.fe te US Cbc. Wase 10D 00%) WF-tee 1 8T1554 ASA4.tauts ee 0.18 rase *te 427E+ te WA ues wA US Lange. eusse(100%) WF-4 0 83 S ISS e5 perA Se WA' 2044 NA I 8 TIPS 2 ASAAeusese - 0.20 WA t.SSEele wA St.t US Lange. Wuu(t00%) WFte 0 SS -S 35225 F 40Eele 4 4eE,le WA- as M.

eT8742 - ASA4.Inde 80 0 20 0 SS f.e2E+ te tis e 7e 2 (se4 Sl t US to LS Chc. Wate { 300%) WF 70 -S 952 25 7.40Ee to 4 4E+fe et es teI S 142 7 T2105 ASAdhidese - t e2Eele Ito 0 LS Lenyt. Weies (Ossh 100%) SA-I500 0TIFe2 AsA&ies80 0 35 0 20 O Se 0 SS M

5 210.75 F.73Eeto 4 seE+1e i SEe18 les e 79 2 ttt e -

as Se SS 10I S t42 F .

152 25 e eEeto 4 SEE+ te Se les e toe a 1.52Eele IIS.S 77.8 Se -l Ampulsessy Oises I et. Movinsen 2.Pseason 2 t eis 1Pe 0 144 4  ;

u,,e, shee ri f, h,.i kh h[jk 'hfM Jh h f f,M[th h [j %24 -i,' ' s Ag "q c4x4 9 cau .I sA.Sn or. in en au ,

.a uti imE.a 4m.a us te uS C= Wow po 40%)

i.m. = m2 83 8 u u sA.nw rues AsAo.e = Om I40 2 nie U$ te LS cheL Wete(100%) WF-70 7210$ ASApide00 esi .le me imE.a 4 m.a i nE.m nr.e =A 0.35 O Se -Se 138.e 7.73Eete 4stEete = wA posi =A 1.8eEete 307.5 F2.4 - Se Se 13F.5 102 4 i

[]- Qontrolling value' s of the adjusted reference temperatures. I t

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l PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 PAGE6 \

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_ . - _ _ . . _ . . - _ _ _ _ _ . _ - _ .__.___._-.________.._______.__m _ _ _ _ _ - _ _ . _ _ . _ - _ _ . . - _ __ - .._..m. _

FRAMATOME TECHNOLOGIES FTI NON-PROPRIETARY 32-5000218-00 Tab le 2. Crystal River Unit 3 Reactor Vessel Beltline Materials Fluence Values at 32 EFPY 32 EFPY Fluence n/ctn' Matl. Inside Ti4 Beltline Materials 3/4T ident. Surface (x = 2.11 in.) ' (x = 6.33 in.)

Lower Nozzle Belt Forging AZJ94 7.08E+18 Upper Shell Plate 4.27E+18 1.55E+18 C43441 7.90E+18 4.76E+18 1.73E+18 Upper Shell Plate C4344-2 7.90E+18 4.76E+18 1.73E+18 Lower Shell Plate C4347-1 8.00E+18 4.82E+18 1.75E+18 Lower Shell Plate C4347-2 8.00E+18 4.82E+18 1.75E+18 LNB to US Cire. Weld (ID 40%) SA 1769 7.08E+18 4.27E+18 ----

LNB to US Cire. Weld (OD 60%) WF-1691 -- --

1.55E+18 US Longit. Weld (100%) W F-8 7.40E+18 4.46E+18 1.62E+18 US Longit. Weld (100%) WF-18 7.40E+18 4.4eE+18 1.62E+18 US to LS Cire. Weld (100%) WF-70 7.73E+18 4.66E+18 1.69E+18 LS Longit. Weld (Both 100%) SA 1580 6.96E+18 4.20E+18 1.52E+18 5.0 ADJUSTED REFERENCE TEMPERATURE CALCULATION WHERE NO SURVEILLANCE DATA IS AVAILABLE The following information is required for determination of the adjusted reference temperature outlined in Regulatory Guide 1.99, Revision 2.

5.1 Initial RTNor The initial RTNor is the reference temperature for the unirradiated base metal vessel beltline material as defined in Paragr 111 of the ASME Boiler and Pressure Vessel IfCode.,aph NB-2331 of Section measured values of initial RTNor for the materialin question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.

The initial RTNor of weld wire heat 72105 is determined using an alternative method based on the material fracture toughness in the transition range. This rnethod is described in BAW-2202.5 Table 3 lists the initial RTNor values for the CR3 reactor vessel beltline materials and their applicable sources.

PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 PAGE 7

t FRAMATOME TECHNOLOGIES FTINON-PROPRIETARY

! 32-5000218-00 Table 3. Crystal River Unit 3 initial RTuor Values for Reactor Vessel Beltline Materials i l

initial l

Matt RTmr.

Beltline Materials ident. F i Reference Source  !

Lower Nozzle Belt Forging AZJ 94 +3 Upper Shell Plate Estimated value (BAW-10046P')

C4344-1 +20 7 Upper Shell Plate Measured value (BAW-1820 )

C4344 2 +20 Lower Shell Plate C43471 Measured value (BAW 1820)  :

-10 Lower Shell Plate C4347-2 Measured value (BAW 1820)

+45 Measured value (BAW-1820) l LNB to US Cire. Weld (ID 40%) SA-1769 +10 8 LNB to US Cire. Weld (OD 60%) WF 1691 -5 Measured value (EPRI NP-373 Weld "B")

US Longit. Weld (100%) Generic value (BAW-1803, Rev.1') l WF-8 -5 Generic value (BAW-1803, Rev.1)

US Longit. Weld (100%) WF-18 -5 Generic value (BAW 1803, Rev.1)

US to LS Cire Weld (100%) WF-70 -26 5 LS Longit. Welds (Both 100%) Measured value (BAW 2202 )

SA-1580 -5 Generic value (BAW 1803, Rev.1) l 5.2 ARTNor ARTuor is the mean value of the adjustment in reference temperature caused by irradiation and is calculated as follows:  !

ARTuor = (CF) *(ff) (2) where: CF = Chemistry Factor ff = fluence factor 5.2.1 Chemistry Factor The chemistry factor (CF) is a function of the material's copper and nickel content. The CF is determined from Table 1 (for weld metals) and Table 2 (for base metals)in Regulatory Guide 1.99, Revision 2. Linear interpolation is permitted. When determining the CF, the " weight percent copper" and " weight percent nickel" are best estimate values for the material, which will normally be the mean of the measured values for the material.

The copper and nickel contents for the beltline base metal materials are reported in BAW-1820,7 and the copper and nickel contents for the beltline weld metals are reported in BAW-2121P.'

The copper content for the lower nozzle belt forging is unavailable, therefore the mean copper content is determined using data from other plants with similar materials. To compensate for heat PREPARER: M.J. DeVan DATE: 06/09/97.

REVIEWER: L.B. Gross DATE: 06/09/97 PAGE 8

FRAMATOME TECHNOLOGIES FTl NON-PROPRIETARY 32-5000218 00 variability, the generic mean value is at the lower tolerance limit with 95 percent confidence that at least 95 percent of the population is greater than this tolerance limit. (See Appendix A for calculation.)

Using Tables 1 and 2 in Regulatory Guide 1.99, Revision 2, the CF for the CR3 reactor vessel beltline region materials are listed in Table 4.

Table 4. Regulatory Guide 1.99, Revision 2, Chemistry Factors for Crystal River Unit 3 Reactor Vessel Beltline Materials Matl. Cu Ni Chemstry Beltline Materials ident. wt% wt% Factor Lower Nozzle Belt Forging AZJ 94 0.13 0.72 Upper Shell Plate 94.0 04144 1 0.20 0.54 141.8 Upper Shell Plate C 4344 2 0.20 0.54 141.8 Lower Shell Plate C43471 0.12 0.58 82.6 Lower Shell Plate C4347-2 0.12 0.58 82.6 LNB to US Cire. Weld (lO 40%) SA-1769 0.26 0.61 181.6 LNB to US Cire, Weld (OD 60%) WF 1691 - 0.18 0.63 158.95 US Longit. Weld (100%) WF-8 0.20 0.55 152.25 US Longit. Weld (100%) WF-18 0.20 0.55 152.25 US to LS Cire. Weld (100%) WF 70 0.35 0.59 210.75 LS Longit. Welds (Both 100%) SA 1580 0.20 0.55 152.25 I

5.2.2 Fluence Factor in accordance with Regulatory Guide 1.99, Revision 2, the fluence factor (ff) for the %T and %T wall locations is determined as follows:

g _ po.:s-oiowin (3)

Table 5 lists the fluence factors for the %T and %T walllocations for the CR3 reactor vessel beltline materials at 32 EFPY.

1 PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 PAGE 9

FRAMATOME

\

TECHNOLOGIES FTl NON-PROPRIETARY '

32-5000218-00 Table 5. Fluence Factors for the %T and %T Wall Locations of the Crystal River Unit 3 Reactor Vessel Beltline Region 4

%T Location %T Location Fluence, Fluence, Matl.

n/cm 2 Fluence n/cm 2 '

Beltline Materials ident. Fluence (x 10") Factor (x 10") Factor i

Lower Nozzle Belt Forging AZJ 94 0.427 Upper Shell Plate 0.764 0.155 0.510 C4344-1 0.476 0.793 0.173 Upper Shell Plate 0.535 C4344-2 0.476 0.793 0.173 1.ower Shell Plate 0.535 C4347-1 0.482 0.797 0.175 Lower Shell Plate 0.538 l C4347-2 0.482 0.797 0.175 0.538 I 1

LNB to US Cire. Weld (ID 40%) SA-1769 0.427 0.764 - - - - -

LNB to US Cire. Weld (OD 60%) WF 169-1 ---- ----

0.155 0.510 US Longit. Weld (100%) WF8 0.446 0.775 0.162 0.520 US Longit. Weld (100%) WF 18 0.446 0.775 0.162 0.520 US to LS Cire. Weld (100%) WF 70 0.466 0.787 0.169 LS Longit. Welds (Both 100%) 0.530 SA 1580 0.420 0.759 0.152 0.506 1

5.2.3 ARTNor Calculation The ARTuor values for the CR3 reactor vessel beltline materials are calculated by multiplying the chemistry factors and fluence factors.

The 32 EFPY ARTuor values for the CR3 reactor vessel beltline materials are presented in Table 6. .

i PREPARER: M.J. DeVan DATE: 06/09/97 l REVIEWER: L.B. Gross DATE: 06/09/97 i- PAGE 10

1 FRAMATOME TECHNOLOGIES FTINON-PROPRIETARY 32-5000218 00 _

Table 6. ARTN or Values for the Crystal River Unit 3 Reactor Vessel Beltline Materials T/4 Location 3/4T Location Matt. l Beltline Materials ident. CF ff ARTsor CF ff ARTNor Lower Nozzle Belt Forging AZ) 94 94.0 0.764 Upper Shell Plate 71.7 94.0 0.510 47.9 C43441 141.8 0.793 112.5 141.8 0.535 Upper Shell Plate C4344 ' 75.9 141.8 0.793 112.5 141.8 0.535 Lower Shell Plate C4347- 75.9 82.6 0.797 65.8 82.6 0.538 44.4 Lower Shell Plate C4347 2 0.797 82.6 65.8 82.6 0.538 44.4 LNB .to US Cire. Weld (ID 40%) SA-1769 181.6 0.764 138.6 181.6 - - - - - - -

LNB to US Cire. Weld (OD 60%) WF-169-1 158.95 ---- ----

0.510 US Longit. Weld (100%) 158.95 81.1 WF-8 152.25 0.775 118.0 0.520 US Longit. Weld (100%) 152.25 79.2 WF-18 152.25 0.775 118.0 0.520 152.25 79.2 US to LS Cire. Weld (100%) WF 70 210.75 0.787 165.9 0.530 LS Longit. Welds (Both 100%) 210.75 111.8 SA 1580 152.25 0.759 115.5 152.25 0.506 77.1 1

5.3 Margin The " margin" is the quantity that is added to obtain conservative, upper-bound values of the adjusted reference temperature. The margin is determined by the following expression:

Margin = 2)cl + c] (4) where ci = standard deviation for the initial RTnor c3 = standard deviation for ARTN or If a measured value of initial RTuor for the materialin question is available, as is to be estimated from the precision of the test method. If generic mean values are used, ai is the standard deviation obtained from the set of data used to establish the mean.

The standard deviation for ARTuor, o3, is 28'F for welds and 17'F for base metals, except that o3 need not exceed 0.50 times the mean value of ARTuor.

l Table 7 list the margin values calculated for the CR3 reactor vessel l

beltline materials through 32 EFPY.

PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97

  • PAGE 11 an,

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TECHNOLOGIES FTl NON-PROPRIETARY 32-5000218-00 . - _

Table 7. Margin Values for the Crystal River Unit 3 Reactor Vessel Beltline Materials l

ARTer / 2 Margin

! Matl.

Beltline Materials ident. o, %T ai %T %T %T Lower Nozzle Belt Forging AZJ 94 31' 17 Upper Snell Plate 35.9 24.0 71 71 C4344-1 0 17 56.3 38.0 34 i Upper Shell Plate 34 C4344-2 0 17 56.3 38.0 34 Lower Shell Plate 34 C43471 0 17 32.9 22.2 34 Lower Shell Plate 34 C4347-2 0 17 32.9 22.2 34 34 i LNB to US Cire. Weld (ID 40%) SA 1769 0 28 69.3 -

56 -

LNB to US Cire Weld (OD 60%) WF 169-1 19.7' 28 -

40.6 -

68 US Longit. Weld (100%) WF-8 19.7' 28 59.0 39.6 68 68 US Longit. Weld (100%) WF 18 19.7' 28 59.0 Gu.6 68 68 US to LS Cire. Weld (100%) WF-70 0 28 83.0 55.9 56 LS Longit. Welds (Both 100%) 8 56 SA 1580 19.7 28 57.8 38.6 68 68 s

5.4 Calculation of Adjusted Reference Temperature i

The adjusted reference ternperature (ART) is given by the following expression:

ART = Initial RTuor + ARTuor + Margin (5)

Table 8 lists the %T and %T ART values calculated for the CR3 reactor vessel beltline materials through 32 EFPY.

l l

t I

l l-b PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 PAGE 12

l

{ FRAMATOME TECHNOLOGIES FTI NON-PROPRIETARY 32-5000218-00 i Table 8. %T and %T ART Values for the Crystal River Unit 3 Reactor Vessel Beltline Materials Adjusted Reference Temperature, F

' Matt.

Beltline Materials ident. T/4 3/4T l

Lower Nozzle Belt Forging AZJ 94 145.5 121.7 l Upper Shell Plate C43441 166.5 129.9 Upper Shell Plate C4344-2 166.5 129.9 Lower Shell Plate C43471 89.8 68.4 Lower Shell Plate C4347-2 144.8 123.4 LNB to US Circ. Weld (ID 40%) SA-1769 204.6 -

LNB to US Cire. Weld (OD 60%) WF-1691 -

144.5 US Longit. Weld (100%) WF8 181.5 142.7 US Longit. Weld (100%) WF 18 181.5 142.7 US to LS Circ. Weld (100%) WF 70 195.9 141.8 LS Lor. git. Welds (Both 100%) SA-1580 179.0 140.6 6.0 ADJUSTED REFERENCE TEMPERATURE CALCULATION WHERE SURVEILLANCE DATA IS AVAILABLE Results from plant specific surveillance programs may be integrated into the adjusted reference temperature estimate if the surveillance data have been deemed credible as judged by the following criteria: .

1. Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement according to the recommendations of Regulatory Guide 1.99, Revision 2.
2. Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-Ib temperature unambiguously.
3. When there are two or more sets of surveillance data from one reactor, the scatter of ARTuor values about a best-fit line drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1 normally should be less than 28'F for welds and 17'F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values.

b PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97' PAGE 13

hRAMATOME

  • TECHNOLOGIES FTINON-PROPRIETARY 32-5000218-00 _

4.

The irradiation temperature of the Charpy specimens in the capsule

. should match the vessel wall temperature at the cladding / base metal interface within 225'F.

5.

The suNeillance data for the correlation monitor material in the capsule should fall within the scatter band of the data b&se for that material.

j

. Appendix B provides the evaluation of the Regulatory Guide 1.99, Revision 2, credibility criteria for the availaole surveillance data for CR3.

I When two or more credible surveillance data sets are available, these data may be used to determine the adjusted reference temperature of the reactor vessel beltline materials as follows:

First, if there is clear evidence that the copper or nickel content of the l surveillance weld differs from that of the reactor vessel weld, the measured

!. values of ARTNor should be adjusted by multiplying the values by the ratio of the chemistry factor for the reactor vessel weld to that for the surveillance weld.

. Second, using the ARTN or and its corresponding fluence, the chemistry factor may be calculated by multiplying each adjusted ARTNor by the corresponding fluence factor, summing the products, and dividing by the sum of the squares of j the fluence factors:

m CF = [ ART >r *$ (6) 15' The CR3 plant specific reactor vessel surveillance program (RVSP) provides data for predicting the reference temperature shift for the base metal plate heat number C4344-1. In addition, the Master Integrated Reactor Vessel Surveillance Program'(MIRVP) described in BAW-1543, Revision 4, provides surveillance data for weld metals SA-1769 and WF-70 on predicting the reference temperature shift.

6.1 Calculation of Chemistry Factor Using Surveillance Data 6.1.1 Base Metal Plate Heat Number C4344-1 L Table 9 lists the available surveillance data for the base metal plate heat number C4344-1 and their reference sources, t

1-l l PREPARER: M.J. DeVan DATE: 06/09/97 ~

L REVIEWER: L.B. Gross DATE: 06/09/97 ~PAGE 14

l FRAMATOME TECHNOLOGIES ' FTl NONoPROPRIETARY 32-5000218 00 i

Table 9. Reactor Vessel Surveillance Material Data for Base Metal Plate Heat Number C4344-1 4

Fluence, Capsule n/cm' ARQr,F Reference CR3-B 1.17E+16 21 CR3-C BAW-2049" 6.56E+18 126

  • CR3-D 7.50E+18 97 "

CR3 F 1.08E+19 128 "

l Using the above capsule data and Equation 6, the chemistry factor for the base metal plate heat number C4344-1 is calculated to be l 118.7 F (see Table 10).

Table 10. Calculation of Chemistry Factor for Base Metal Plate Heat Number C4344-1 Using Surveillance Data Fluence Capsule ARbr, F Factor (ff) ARkr

  • ff ff*

I CR3-B 21 0.449 9.4 0.202 l 1 CR3-C 126 0.882 111.1 0.778 i CR3-D 97 0.919 89.1 0.845 CR3-F 128 1.022 130.8 1.044 SUM 340.4 2.869 I

CF = I (ARTwor

  • ff) /I (ff*) = 118.7 .

i 6.1.2 Weld Wire Heat 71249 Table 11 lists the available surveillance data for the weld wire heat number 71249 and their reference sources.

l l

1

( PREPARER: . M.J. DeVan DATE: 06/09/97

\ REVIEWER: L.B. Gross DATE: 06/09/97 - PAGE 15

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FRAMATOME -

TECHNOLOGIES FTl NON-PROPRIETARY 32-5000218-00 Table 11. Reactor Vessel Surveillance Material Data for Weld Wire Heat Number 71249 Fluence, i Capsule n/cm* ARTuor,F Reference TP3-T 7.19E+18" 155 WCAP-8631 "

TP3 V 1.23E+19 180 swr 106-8575"

! TP4 T 7.74 E+18 12 225 SwRl 02-4221" l

l l The reported copper content for the above two surveillance welds are 0.31 wt% for the Turkey Point Unit 3 data'8 and 0.30 wt% for L

the Turkey Point Unit 4 data,'7 while the best estimate copper content for this weld wire is 0.26 wt%.' Therefore, the surveillance i data for the ARTuor values are conservative with respect to the best estimate copper contents, and no adjustment to the l surveillance data is necessary.

, Using the above capsule data for weld wire heat 71249 and i

Equation 6, the chemistry factor for the weld wire heat number 71249 is calculated to be 192.6 F (see Table 12),

l i

Table 12. Calculation of Chemistry Factor for Weld Wire i

Heat Number 71249 Using Surveillance Data Fluence .

! Capsule ARTuor,F Factor (ff) ARTuor

  • ff ff*

TP3 T 155 0.907 140.6 0.823 TP3-V 180 1.058 190.4 1.119 TP4 T 225 0.928 208.8 0.861 SUM 539.8 2.803 i

2 CF = I (ARTN or

  • ff) /I (ff ) = 192.6 l

6.1.3 Weld Wire Heat 72105 Table 13 lists the available surveillance data for the weld wire heat number 72105 and their reference sources.

PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 PAGE 16

FRAMATOME . .

- TECHNOLOGIES FTl NON-PROPRIETARY 32-5000218-00_

Table 13. Reactor Vessel Surveillance Material Data for Wold Wire Heat Number 72105 Fluence, Capsule n/cm' ARTw, F Reference OC2-C 1.02E+18 " 45 BAW 2051"'

OC2-A 3.37E+18 114 BAW-2051 OC2-E ~ 1.21 E+19 179 BAW 2051 i'

OC3-A 8.10E+17 48 OC3-B BAW 2128. Rev.180 3.12E+18 64 BAW-2128 Rev.1

! OC3 D 1.45E+19 140 BAW 2128, Rev.1 TMl2-LG1 5.85E+18 123 BAW 2253P

DB1-LG1 6.63E+18 135 BAW 1920P' CR3-LG2 1.19E+19 125 BAW 2254 Pas i-Z1 T 3.10E+18 112 BAW 2082

21 U 1.02E+19 199 BAW 2082 Z1 X 1.26E+19'" 199 BAW-2002 Z1-Y 1.56E+19'8 205 BAW 2082 Z2-U 2.70E+18 128 WCAP 12396:s Z2 T 7.79E+18 175 WCAP 12396 l

Z2-Y 1.46E+19 220 WCAP 12396 I Since the cold leg operating temperatures for the Zion units are approximately 27 F lower than the coM-leg operating temperature for CR3 (i.e.,529 F versus 556 F), for conservatism, the Zion surveillance data (shown in Table 13) will not be considered in the evaluation for this weld wire heat.

The reported copper content for the Oconee Units 2 and 3 2s surveillance welds are 0.36 wt% and 0.30 wt%, respectively and the surveillance weld in the BWOG supplemental capsules (TMl2-LG1, DB1-LG1, and CR3-LG2) has a reported copper content of 0.42 wt%.2e The best estimate copper content for this weld wire

- heat is 0.35 wt%.' Therefore, the measured ARTuor values are adjusted by multiplying these values by the ratio of the chemistry factor (as determined in accordance with Regulatory Guide 1.99, Revision 2, Table 1) for the reactor vessel to that of the

  • surveillance welds. The data used to determine the ratio of the

., weld wire heat 72105 best estimate chemistry factor to that of the surveillance weld chemistry factors are presented in Table 14.

PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B; Gross' DATE: 06/09/97 ' PAGE 17 w '

FRAMATOME TECHNOLOGIES FTlNON PROPRIETARY 32-5000218-00 Table 14. Ratio of Weld Wire Heat 72105 Best Estimate Mean Chemistry Factor to That of the Surveillance Weld Chemistry Factors Chemistry Weld Wire Factor from Ratio Weld Cu Ni Heat No. RG1.99/R2 (Vessel to Identification wt% wt%

72105 Table 1 Surv. Data)

WF-209 OC2 0.36 0.58 213.5 0.987 WF-2091 - OC3 0.30 0.58 191.3 1.102 WF 70 - BWOG 0.42 0.59 229.8 0.917 i Weld Wire Best Estimate 0.3S i 0.59 210.8 --

J The adjusted measured ARTNor values are presented in Table 15, and these values are used to determine the chemistry factor in accordance with Regulatory Guide 1.99, Revision 2.

Table 15. Adjustment of Measured Surveillance Data Using Ratio Procedure of Regulatory Guide 1.99, Revision 2 -

Ratio Weld Wire Heat No. Adjusted Measured CF Ratio Measured (Weld Identification) Caosule ARTuor.F 72105 (See Table 14) ARTyor,F OC2-C 45 0.987 (WF-209 1, WF-70) 44.4 OC2-A 114 0.987 112.5 OC2-E 179 0.987 '

176.7 OC3-A 48 1.102 52.9 OC3-D 64 1.102 70.5 OC3-D 140 1.102 154.3 TMl2-LG1 123 0.917 112.8 DB1-LG1 135 0.917 123.8 CR3-LG2 125 0.917 114.6 ,

Using the above capsule data for weld wire heat 72105 and Equation 6, the chemistry factor for the weld wire heat number 72105 is calculated to be 136.6 F (see Table 16).

i PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 PAGE 18

i FRAMATOME l

TECHNOLOGIES FTI NON-PROPRIETARY 32-5000218-00 Table 16. Calculation of Chemistry Factor for Weld Wire Heat Number 72105 Using Surveillance Data Ratio Adjusted Measured Fluence .

Capsule 6RTer, F Factor (ff) ARTer

  • ff ff*

OC2 C 44.4 0.421 18.7 0.177 OC2 A 112.5 0.701 78.9 0.491 OC2 E 176.7 1.053 186.1 1.109 OC3 A 52.9 0.376 19.9 0.141 OC3B 70.5 0.680 47.9 0.462 OC3-D 154.3 1.103 170.2 1.217 TM12-LG1 112.8 0.850 95.9 0.723 DB1-LG1 123.8 0.885 109.6 0.783 CR3-LG2 114.6 1.049 120.2 1.100 SUM 847.4 6.203 CF = I (ARTN or

  • ff) /I (ff") = 136.6 6.2 Adjusted Reference Temperature Calculated Using Surveillance Data 6.2.1 Initial RTuor The initial RTNor for base metal plate heat number C4344-1 and weld metals SA-1769 and WF-70 are the same values specified in Section 5.1: '

Beltline Materials initial RTNor. F Upper Shell Plate (C4344-1) +20 NB to US Cire. Weld (SA 1769) +10 US to LS Cire. Weld (WF 70) 26 6.2.2 ARTNor Calculation The %T and %T ARTuor values for the beltline region materials where surveillance data are available are calculated in accordance with Equation 2 using the chemistry factors determined in Section 6.1 and the respective fluence factors determined in Section 5.2.2.

The 32 EFPY %T and %T ARTuor values for the CR3 reactor PREPARER: M.J. DeVan DATE: 06/09/97

^ REVIEWER: L.B. Gross DATE: 06/09/97

  • PAGE 19

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,, e-, - -,-e-,- , , e -----n--,---. .,---,-w .n..-~+-. <. ~ ~ u e-, - - -

FRAMATOME -

TECHNOLOGIES FTlNON-PROPRIETARY 3

32-5000218-00 _

i vessel beltline region materials where surveillance data are available are presented in Table 17.

l Table 17. ARTwor Values at 32 EFPY for the CR3 Reactor Vessel Beltline Materials With Surveillance Data Available Matt. %T Location MT Location Beltline Materials ident. }

CF ff CF ARTuor ff Upper Sheli Plate ARTwer )

C43441 118.7 0.793 94.2 118.7 0.535 63.6 NB to US Cire. Weld (1D 40%) SA 1769 192.6 0.764 147.0 192.6 ----- -----

US to LS Cire. Weld (100%) WF 70 136.6 0.787 107.5 136.6 0.530 72.4 l

6.2.3 Margin Applying the credibility criteria defined in Regulatory Guide 1.99, Revision 2, the scatter of the measured ARTuor values about a best-fit line drawn as described in Position 2.1 is greater than 17 F for the upper shell base metal plate and 28 F for weld wires 71249 and 72105 (see Appendix B). Regulatory Guide 1.99, Revision 2, allows reduction of aa by one half. However, to bound the scatter of the measured data about the best-fit line, os values were not so reduced (i.e.,34 F for base metals and 56 F for weld metals).

To calculate the margin values for beltline materials with surveillance data available, Equation 4 is used. Table 18 lists the margin values calculated for the CR3 vessel beltline materials where surveillance data available.

Table 18. Margin Values for the CR3 Reactor Vessel Beltline Materials With Surveillance Data Available Matt. Margin Beltline Materials ident, e, ca %T %T Upper Shell Plate C4344-1 0 17 34 34 l NB to US Cire. Weld (ID 40%) SA 1769 0

28 56 ---

1 US to LS Cire. Weld (100%) WF 70 0 28 56 56 i

PREPARER: M.J. DeVan .DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 PAGE20

, , , , . - .+ - -

i FRAMATOME TECHNOLOGIES FTlNON-PROPRIETARY 32-5000218-00 6.2.4 Calculation of Adjusted Reference Temperature The ART is calculated using Equation 5. Table 19 lists the %T and

%T ART values through 32 EFPY calculated for the CR3 reactor vessel beltline materials where surveillance data available. '

Table 19. %T and %T ART Values Through 32 EFPY for the CR3 Reactor Vessel Beltline Materials With Surveillanca Data Available Adjusted Reference Matt. Temperature, F Beltline Materials ident.

Y4T %T Upper Shell Plate C43441 148.2 117.6 NB to US Cire. Weld (ID 40%) SA 1769 213.0 ---

US to LS Cire. Weld (100%) WF 70 137.5 102.4 PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97- PAGE21

I FR' AMATOME -

TECHNOLOGIES' FTlNON-PROPRIETARY 32 500021g.co l

l

(

  • 3 APPENDIX A l

Generic Copper Content for Base Metal Forging Materials Fabricated by the Ladish Company l

4 PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 PAGE22

~

l

FRAMATOME TECHNOLOGIES FTlNON-PROPRIETARY l 32 50002gg.gg 1

For ASTM A 508, Class 2 forging materials fabricated by the Ladish Company for which copper content was not determined, a statistical determination was performed to obtain a lower tolerance generic value with 95 percent confidence that at least 95 percent of t i

~ he population is greater than this tolerance limit. The available copper content for all  !

such materialin the B&W Owners Group (BWOG) Reactor Vessel Working Group (RVWG) data base (18 entries) were complied. The mean was found to be 0.05 weight

. percent, the standard deviation is 0.03 and the one-sided factor (k) is 2.453. The generic value is calculated to be 0.13 weight percent (mean plus ka). The data is shown in Table A-1.

l l

l l

PREPARER: M.J. DeVan DATE: 06/09/97

' REVIEWER: L.B. Gross DATE: 06/09/97- PAGE23

FRAMATOME -

TECHNOLOGIES FTlNON-PROPRIETARY 32-5000218-00 Table A-1.. Copper Content Data for Ladish Company Forging Materials Used to Determine Generic Copper Content Plant Heat No. Cu, wt%

Reference Oconee Unit 1 - AHR54 0.155 BAW 1820' Oconee Unit 2 AMX 77 0.06 BAW 1820 AAW 163 0.04 BAW-1820 AWG 164 0.02 BAW-1820 TMI Unit 1 ARY 59.-

TMI Unit 2 0.08 BAW-1820 AWK 123 0.057 Material Test Report

Crystal River Unit 3 ABM 96 0.054 MaterialTest Report" ANO Unit 1 AYN 131 0.03 BAW.1820 Oconee Unit 3 AWS 192 0.01 BAW-1820 ANK 191 0.02 BAW-1820 Davis-Besse ADB 203 0.04 BAW-1820 AKJ 233 0.04 BAW-1820 BCC 241 0.02 BAW 1820 Midland Unit 1 ABZ 196 0.02 Material Test Report" ACA 197 0.02 MaterialTest Report" Zion Unit 1 AZL 116 0.06 Material Test Report

ANA 102 0.06 Material Test Report

Zion Unit 2 ADY 144 0.09 Material Test Report **

Mean = 0.05, er = 0.03, k = 2.453 Tolerance Limit = (Mean + ka) = 0.13 l-

- PREPARER: - M.J. DeVan REVIEWER:.

DATE: 06/09/97 L.B. Gross DATE: 06/09/97 PAGE24

FRAMATOME ~

TECHNOLOGIES FTlNON-PROPRIETARY 32 50002gg.co l'

I l

l i

l APPENDIX B Credibility of Surveillance Data I

?

PREPARER: M.J. DeVan DATE: 06/09/97 l REVIEWER: L.B. Gross DATE: 06/09/97 - PAGE25

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FRAMATOME -

TECHNOLOGIES FTl NON-PROPRIETARY 32-5000218-00 _

To verify that the adjusted reference temperature for each beltline material is a bounding value for the specific reactor vessel, plant specific information that could affect the level of embrittlement should be considered. This information includes, bu not limited to, the reactor vessel operating temperature and any related surveillance data that is available. The results of plant specific surveillance program data should be integrated into the adjusted reference temperature projection if the plant specific l

surveillance data has been deemed credible as judged by the criteria presented in ,

1 Section 6.0 of this calculation. This Appendix provides the review of the available i

surveillance data for CR3 against these credibility criteria.

i Criterion 1. Materials in the capsules should be those judged most likely to be ,

controlling with regard to radiation embrittlement according to the recommendations of Regulatory Guide 1.99, Revision 2.  !

I Surveillance data available include the following materials: 1 Base Metal Heat No. C4344-1 (CR3 plant-specific RVSP data) i Weld Wire Heat No. 71249 (MIRVP data)

Weld Wire Heat No. 72105 (MIRVP data)

' All these heats of material lie within the reactor vessel beltline region of the CR3 reactor vessel as defined in 10 CFR 50, Appendix G.3' Therefore, these materials could be controlling with regard to radiation embrittlement.

i Criterion 2. Scatter in the plots of Charpy energy versus temperature for the irradiated and unfrradiated conditions should be small enough to  ;

i permit the determination of the 30 ft-lb temperature unambiguously. l The available Charpy V-notch data for these surveillance data permit the determination of the 30 ft-lb temperatures and are presented in their respective RVSP reports. <

Criterion 3. When there are two or more sets of surveillance data from one reactor, the scatter of ARTuor values about a best-fit line drawn as l

described in Regulatory Guide 1.99, Revision 2, Position 2.1 normally should be less than 28 F for welds and 17 F for base metal. Even if the fluence range is large (two or more orders of magnitude), the  ;

scatter should not exceed twice those values.

The scatter of the measured ART N or values for the available surveillance data are presented in Table B-1 for the bas'e metal and Table B-2 for the weld metals. The scatter of the measured ARTuor values about the best fit line drawn as described in PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 PAGE 26

FRAMATOME TECHNOLOGIES FTl NON-PROPRIETARY 32-5000218 00 Regulatory Guide 1.99, Revision 2, Position 2.1 is greater than 17 F for the base and greater than 28 F for the weld metals.

Criterion 4. The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding / base metal interface within 25'F.

The surveillance data for the base metal C4344-1, the Oconee Units 2 and 3 (OC2 and OC3) weld metal WF-209-1, and the B&WOG weld metal WF-70 were irradiated in the CR3 with a cold leg operating temperature of 556 F. The cold leg operating temperature for the Turkey Point Unit 3 (TP3) and the Turkey Point Unit 4 (TP4) surveillance weld data (SA-1101 and SA-1094 respectively) is 546*F and is within 25 F of the cold leg operating temperature for CR3. However, the cold leg operating temperature for the Zion Unit 1 (Z1) and the Zion Unit 2 (Z2) surveillance weld data l '(WF-209-1) is 529 F and is not within 25 F of the cold leg operating temperature for CR3. Therefore, all the available MIRVP surveillance weld data for the welds SA-1769 and WF-70 with the exception of the Zion Unit 1 and Unit 2 surveillance data can be used in the chemistry factor determination.

i Criterion 5. The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the data base for that l material.-

All available surveillance data used in the chemistry factor determination have correlation monitor material data which fall within the scatter band of the available data base for that material. '

l l

l f-l PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 ' PAGE27 w se o :- + --

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FRAMATOME TECHNOLOGIES' .

FTlNON-PROPRIETARY 32-5000218-00 Table B-1. Base Metal Plate Heat Number C4344-1 Surveillance Data Credibility

~

, y _

-. E' Chemistricq e 9..,

J,MT Q c ll ' I6_

'J Factor. M fARTd.3 Base, Metal

t. ,. .gfiQ[ ,Capsu@_; QM'g j/hh h%,6"1)j Q(CFJff)$ j ;ARTeeig , Meas. ;j; Scatter;

' <117 F Crystal River Unit 3 CR3-B 1.17E+18 0.449 118.7 53.3 1

21 (TL) -32 No Plate Cd344-1. CR3-C - 6.56E+18 0.882 118.7 104.7 126 (TL) 21 No CR3-D 7.50E+18 0.919 118.7 109.1 97(TL) -12 Yes CR3-F 1.08E+19 1.022 118.7 121.3 128 (TL) 7 Yes t

r i

i t

I i

I i t t

t i  !

PREPARER: M.J DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 i PAGE28

?

I

FRAMATOME TECHNOLOGIES FTlNON-PROPRIETARY 32-5000218-00 Table B-2. Weld Metal Surveillance Data Credibility

..;,s. .. ~ johO w N .. ,h

?Mi  %.$ I.M, I$$$$.9 W~ E "$l-Wr.w en .?/,

Ratio '

Weld Wire Heat Number.

" a r

- --s 4:_

qsRM 44 w bi}N{ Chemist it g EARTM %

Adiusted (Weld identifications) . ; , e7 W[6 lC@dWif- ,%n, Fluence @ip..d

  1. ggffMN fd@yFactor rf,%ser_)_j (6 ft(CEsif)tl( i;,li$catter -

.<i28*F ' i 71249 TP3-T 7.19E+18 0.908 192.6 -174.7 155* -20 Yes (SA-1094, SA-1101, SA-1229 TP3-V . 1.23E+19 1.058' 192.6 203.7 . 180* -24 Yes and SA-1769) TP4-T 7.74E+18 0.928 192.6 178.7 225* 46 No 72105 OC2-C 1.02E+18 0.421 136.6 57.5 44.4 -13 Yes *

(WF-209-1 and WF-70) OC2-A 3.37E+18 0.701 136.6 95.7 112.5 17 Yes OC2-E 1.21E+19 1.053 136.6 143.8 176.7 33 No OC3-A 8.10E+17 0.376 136.6 51.4 52.9 2 Yes OC3-B 3.12E+18 0.600 136.6 92.9- 70.5 -22 Yes OC3-D 1.45E+19 1.103 136.6 150.7 154.2 4 Yes  !

T1 5.85E+18 0.850 136.6 116.1 112.8 -3 Yes D1 6.63E+18 0.885 136.6 120.8 123.8 3 Yes C2 1.19E+19 1.049 136.6 143.2 114.6 -29 No

- No adjustment to measured data is required (Section 6.1.2). '

i I t PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 l '

PAGE29 J

l FRAMATOME -

\ - TECHNOLOGIES FTlNON-PROPRIETARY 32 50002yg.co i

I 1

i 1

l-i l

I l

4 l

I APPEfJDIX C References l

i 0

PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 PAGE 30 l.

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W FRAMATOME TECHNOLOGIES FTINON-PROPRIETARY 32-5000218 00 I' 1. U.S. Regulatory Commission, " Radiation Damage to Reactor VesselMaterial,"

Reaulatorv Guide 1.99. Revision 2. May 1988.-

2. Framato' me Technologies Document 86-1266133-00, "CR-3 PTFluence Analysis l R6 port Cycles 7- 10," released June'1997.

J l

3. L. S. Harbison, "MasterIntegrated Reactor VesselSurveillance Program,"BAW.

1543. Revision 4. B&W Nuclear Technologies, Inc., Lynchburg, Virginia, February 1993.

p 4.'

American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section 111, Nuclear Power Plant Components, Subsection NB, Class 1 Components.

5. K. K. Yoon, " Fracture Toughness Characterization of WF-70 WeldMetal,"BAW-22Q2, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, September 1993.
6. H. S. Palme,' H. W. Benke, and W. J. Keyworth, " Methods of Compliance with

[ ^ Fracture Toughness and OperationalRequirements of to CFR 50, Appendix G,"

BAW-10046P. Babcock and Wilcox's Nuclear Power Generation Division, Lynchburg, Virginia,' March 1976.

7. IJ. D. Aadiand,' Babcock and Wilcox Owners' Group 177-Fuel Assembly Reactor

~ Vessel and Surveillance Program Materials Information,"BAW-1820. Babcock &

Wilcox's Nuclear Power Division, Lynchburg, Virginia, December 1984.

S. W. A. Van Der Sluys, .q.t3L, t "An Investigation of Mechanical Properties and

' Chemistry Within a Thick MnMoNiSubmerged Arc Weldment,"EPRI NP-373.

Electric Power Research Institute, Palo Alto, Califomia, February 1977.

L

9. A. L. Lowe, Jr., and J. W. Pegram, Correlations forPredicting the Effects of Neutron Radiation on Linde 80 Submerged-Arc Welds,"BAW-1803. Revision 1. B&W j Nuclear Technologies, Inc., Lynchburg, Virginia, May 1991.
10. L. B . Gross, " Chemical Composition of B& W Fabricated Reactor Vessel Beltline i=

Welds,"BAW-2121 P. B&W Nuclear Technologies, Inc., Lynchburg, Virginia, April 1991.

11. A. L. Lowe, Jr., .qLgL, " Analysis of Capsul 9 CR3-F Florida Power Corporation

' ' Crystal River Unit 3 Reactor Vessel Material Surveillance Program,"BAW-2049, Babcock & Wilcox's Nuclear Power Division, Lynchburg, Virginia, September 1988.

' 12. W. N. McElroy, Editor, " LWR Pressure VesselSurveillance Dosimetry Improvement

' Program: LWR Power Reactor Surveillance Physics-Dcsimetry Data Base

' Compendium,"NUREGICR-3319. Revision 1, Prepared by Hanford Engineering Development Laboratory, Richland, Washington, March 1987.

. PREPARER: M.J. DeVan . DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/S7 - PAGE 31 r . rrr. , ~.r. <e-. -- . , * -

FRAMATOME i TECHNOLOGIES FTlNON-PROPRIETARY 32-5000218-00

13. S. E. Yanichko, J. H. Phillips, and S. L Anderson, " Analysis of Capsule Florida Power and Light Company Turkey Point Unit No. 3 Reactor Vess Pennsylvania, December 1975. Surveillance Program,"WCAP 863 ,
14. P. K. Nair and E. B. Norris, " Reactor VesselMaterialSurveillance Program for Turkey Point Unit No. 3: Analysis of Capsule V,"SwRI Project No. 06-8575.

Southwest Research Institute, San Antonio, Texas, August 1986.

15. E. B. Norris, " Reactor Vessel Material Surveillance Program for Turkey Point Un No. 4 Analys/s of Capsule T,"SwRI Project No. 02-4221. Southwest Research Institute, San Antonio,-Texas, June 14,1976.
16. S. E. Yanichko, Florida Power and Light Company Turkey Point Unit No. 3 Reac! '

- Vessel Radiation Surveillance Program,"WCAP-7656, Westinghouse Eiectric Corporation, Pittsburgh, Pennsylvania,' May 1971.

17. S. E. Yanichko, " Florida Power and Light Company Turkey Point Unit No. 4 Reacto l VesselRadiation Surveillance Program,"WCAP-7660. Westinghouse Electric

. Corporation, Pittsburgh, Pennsylvania, May 1971. t p 18. Framatome Technologies Document,38-1247136-00, " Zion Units 1 and2 Fluence L Projections for Reactor Vessel Core Region Materials," released May 28,1996.

~

19. A. L. Lowe, Jr., gt.6, " Analysis of Capsule OCII-E Duke Power Company Oconee i

Nuclear Station Unit-2 Reactor VesselMeterial Surveillance Program,"BAW-2051.  ;

L )

Babcock & Wilcox's Nuclear Power Divisk,n, Lynchburg, Virginia, October 1988 l 20. A. L. Lowe, Jr., gL&, " Analysis of Capsule OClll-D Duke Power Company Oconee i

Nuclear Station Unit-3 Reactor VesselMaterial Surveillance Program,"BAW-2128.

Revision 1. B&W Nuclear Technologies, Inc., Lynchburg, Virginia, May 1992.

.i

21. M. J. DeVan, S. O. King, and K. K. Yoon,," Test Results of Caosule TMl2-LG1 B& W

. Owners Group Master Integrated Reactor VesselSurveillance Program,"BAW-

[ 2253P. B&W Nuclear Technologies, Inc., Lynchburg, Virginia, October 1995.

22.' A. L: Lowe,' Jr., s_t.6, " Analysis of Capsule DB1-LG1 Babcock & Wilcox Owners Group, Integrated Reactor Vessel Material S'irveillance Program," B AW-1920P.

{ Babcock & Wilcox's Nuclear Power Division echburg, Virginia, October 1986.

23. M. J. DeVan, S. O. King, and K. K. Yoon, " Test Results of Capsule CRS-LG2 B& W

^ Owners Group Master Integrated Reactor Vessel Surveillance Program,"BAW-

i. 2254P, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, October 1995, t

PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 PAGE 32

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FRAMATOME TEC.HNOLOGIES FTl NON-PROPRIETARY 32-5000218-00

24. A. L. Lowe, Jr.,9th, " Analysis of Capsule Y Commonwealth Edison Com Nuclear Plant Unit 1 Reactor VesselMaterialSurveillance Program,"BAW-2082.

B&W Nuclear Technologies, Inc., Lynchburg, Virginia, March 1990.

25. E. Terek, S. L. Anderson, and L. Albertin, " Analysis of Capsule Y from the Commonwealth Edison Company Zion Unit 2 Reactor VesselRadiation Surveillan Program,"WCAP-12396. Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, September 1989.
26. K. E. Moore and A. S. Heller, " Chemistry of 177-FA B& W Owners' Group Reactor Vessel Beltline Welds,"BAW-1500. Babcock & Wilcox's Nuclear Power Generation Division, Lynchburg, Virginia, September,1978.
27. The Babcock & Wilcox Company - Mt. Vemon Works, Certificate of Test, Heat No.

AWK 123 (122W240), dated December 3,1975.

28. The Babcock & Wilcox Company - Mt. Vemon Works, Certificate of Test, Heat No.

ABM 96, dated October 11,1974.

29.1969.

Ladish Company, Materials Analysis Report, Heat No. ABZ 196, September 17,

30. Ladish Company, Materials Analysis Report, Heat No. ACA 197, September 16, 1969.
31. Ladish Company, Materials Analysis Report, Heat No. AZL 116, June 14,1968. '

4

32. Ladish Company, Materials Analysis Report, Heat No. ANA 102, June 26,1968.  !
33. Ladish Company, Materials Analysis Report, Heat No. ADY 144, March 1,1969.
34. Code of Federal Regulation, Title 10, Part 50, " Fracture Toughness Requirements for Light-Water Nuclear Reactors," Appendix G, " Fracture Toughness Requirements.' '

PREPARER: M.J. DeVan DATE: 06/09/97 REVIEWER: L.B. Gross DATE: 06/09/97 ~ PAGE 33

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