ML20059D397

Rev 0 to Development of RCS Pressure Temp Limits for 20 EFPYs (Fluence = 2.29 X 10(19) n/cm(2)) & Recommended Surveillance Withdrawal Schedule for Waterford Unit 3, Final Rept

ML20059D397
Person / Time
Site:	Waterford
Issue date:	10/31/1993
From:	ASEA BROWN BOVERI, INC.
To:
Shared Package
ML20058M173	List: ML20058M169 ML20058M180 ML20059D397
References
C-MECH-ER-021, C-MECH-ER-021-R00, C-MECH-ER-21, C-MECH-ER-21-R, NUDOCS 9401070235
Download: ML20059D397 (62)

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{{#Wiki_filter:--. - x- .3, DEVELOPMENT OF. 1 REACTOR COOLANT SYSTEM .E PRESSURE TEMPERATURE LIMITS FOR i 'i 20 EFFECTIVE FULL POWER YEARS 2 (FLUENCE = 2.29 x 10" n/cm ) AND RECOMMENDED i SURVEILLANCE WITHDRAWAL SCHEDULE FOR-WATERFORD UNIT 3 - 3 -h Prepared For: ENTERGY OPERATIONS, INC. WATERFORD STEAM ELECTRIC STATION UNIT 3 ' KILLONA, LOUISIANA 700664751 i .By: ABB COMBUSTION ENGINEERING NUCLEAR SERVICES l . COMBUSTION ENGINEERING, INC. z REACTOR VESSEL INTEGRITY. GROUP.. 1000 PROSPECT HIII ROAD ' 2 WINDSQR, CONNECTICUT 06095 4500 g .) f OCTOBER,1993 L o FINAL REPORT J 9401070235 931214 9 q.c' C-MECH-ER-021, REV. 00 1 PDR ADDCK 050003B2: I .,] DR o

H 3 -e EXECUTIVE

SUMMARY

] This report provides reactor coolant system pressure-temperature _ limits valid through twenty (20) effective full power years' (EFPY) operation for Waterford Steam Electnc Station Unit - -3 (Waterford Unit 3). De reactor coolant system pressure-temperamre limits currently, j provided by the Waterford Unit 3 Technical Specifications are valid through 8 EFPY and have been revised to account for additional embrittlement of the reactor vessel beltline m.erials due to continued neutron irradiation. The change in the reactor vessel beltline. fracture toughness properties have been estimated in accordance with' Regulatory Guide 1.99. Revision 2 using a projected peak surface fluence of 2.29 x 10 n/cm. His fluence was-- 2 based on the tasults obtained from the W-97 surveillance capsule evaluation report. The. limiting material was idenafied as plate M-1004-2 having adjusted reference temperamres at. the 1/4t and 3/4t locations of 65.4*F and 54.0*F, respectively. The reactor coolant system - j prews-temperanut limas.. including the revised beltline analysis, have been developed in- ] accordance with 10 CFR 50 A,Wir G requiremeses. 10 CFR 50 App =tir G and 10 CFR 550.61 requirements for end-of-life (32 EFPY) fracare toughness propernes were reviewed and shown to be in compliance. Specifically, all beltline? i materials will have greater than 50 ft-Ibs upper-shelf energy projeceed at the 1/4t locations / given an end-of-life peak surface fluence of 3.69 x 10' n/cm. The values of RTm for all ^ 1 2 beltline materials are well below the screening criteria provided by 10 CFR 550.61. l i a Protection (LTOP) Enable temperamres were developed I.ow Temperamre Cwr using the g'h of Standard Review Plan 5.2.2 and ASME Boiler and Pressure Vessel-Code, Section XI, Code Case N-514 (not yet approved for use by the NRC).- l Lastly, a revised surveillance capsule withdrawal echarhde has beien developed based oni j consideration of the W-97 serveillance capsule results. - His has been developed in accordance whh the regulatory requirements of 10 CFR 50 Appendix H.' Consideration hasi g 1 been given to the fbtore ASTM E-185-93. s j . AAR Consbeslon Eng6essnag Maciasr Servicar L C. MECH-ER Otl,'Rev. 00

7 b TABLE OF CONTENTS L L SECTION IHLE Iggg Executive Summary i i List of Tables ~ . iii

j e

List of Figures iv; lj 1.0 Introduction 1-a 2.0 RCS Pressure-Temperamre Limits - 1 t 2.1 . Overview -2' 2.2 Basic Data 3-2.3 Adjusted Reference Temperature Protections '4. I 2.4 Calculation of Reactor Beltline P-T Limits 8'. J t 17-i 2.3 Core Critical Limits 2.6 Flange Limits i 17-. 2.7 Lowest Service Temperanne '18 : 1 2.8 Minimum Pressure L18 1 2.9 Minimum Boltup Temperanne 19 3.0 LTOP Enable Temperannes 20: 4.0 Pressurized Thermal Shock (PTS) Screening Criteria 22' 5.0 End-of-Life Upper Shelf Energy 23: 1y j 6.0 Surveillance Capsule Withdrawal Schedule 24: 3 7.0 Results 128 .l .i 8.0 References 42- ? Appendix A Discussion of the Development of Tachniest ' Specification =A1-Figures 3.4-2 and 3.4-3 -i -) ^ ASS Comiteenlos Engineering #1meiser Servica C MECN-ER Otl, Rev. 00 il i i-m .m.. m.-.

I TRT OF TABf At n M P. AGE i 1 T'L*7 Beltline Materials 29 2 . ART values for Beltline Materials at 20 EFPY-30 i 3 Beltline P-T Limits, Heatup 20 EFPY '31 4 Beltline P-T Limits, Cooldown 20 EFPY -32i l r 5 Beltline P.T Limits, Hydrostatic Test,20 EFPY 33 6 RTm and End of Life USE Evaluations 34 7 WSES-FSAR-Unit 3. Capsule Annembly Removal Schedule 35. 8 Pwyc.c.d Capsule Removal Mc%le Meeting 36 ASTM E185-82 Requawise i .i ~ E a i j ABB Combannon Engsneenng Nudear Services ,,,"'~ C-MECH-Ball, Jter. 00. 4 e =

.j 1 zI LIST OF FIGURES j -i R DESCRIPTION PAgg

1 Waterford Unit 3, Appendix G Beltline P-T Limits'

.' 37 1 Heatup 2 Waterford Unit 3, A'ppendix G Beltline P-T Limits , 38 L j Cooldown i 3 Waterford Unit 3, Appendix G Beltline P-T Limits 39 ] Hydrostatic

l

-t Li 4 Waterford Unit 3 Hestup Curve, Reactor Coolant System f40. Pressure-Temperature Limits,0-20 EFPY (Peak Surface 2 Fluence = 2.29 x 10 n/cm ) i 5 Waterford Unit 3 Cooldown Curve, Reactor Coolant System 41 Pressure-Tempemmre Limits, 020 EFPY (Peak Surface Fluence = 2.29 x 102' n/cm ) j 2 .q 1 l -l I c: q .i l ? \\ i ' ABS Conkaion tl ::- 4 thwiaar Senicar C'-MEG-ER 021, Jter. 00. . iv

l.0 INTRODUCTION 3 Reactor coolant system (RCS) pressure-temperature (P-T) limits provide protection ] against nonducune failure for ferritic pressure boundary components during normal operation bestup, cooldown and inservice testing. These limits are required to be ll - developed to meet the requirements of 10 CFR 50,- Apwar G (Reference 1) and must be updated periodically to account for conrimind embrittlement of the reactor 1 - vessel beltline due to neutron irradiation. The rate of embrittlement is monitored by a the withdrawal of surveillance capsules and mechanical testing of material specimens-1 '4 within the capsules'and dosimetry evaluations. 10 CFR 50, A,Wir H (Reference

2) provides the requirements for the reactor vessel material surveillance program..

The first capsule (W-97) was recently removed from the Waterford Steam Electne Station Unit 3 (Waterford Unit 3). _Upon removal and completion of maserial testing and d.

- ; f calculatinna, a report (Report No. BAW-2177, Babcock & Wilcox -

1 Nuclear Services (Ref. e 3)) annunarizing the reeubs was provided to the NRC by 1 Entergy Operations, Inc. (under Reference 4 cover). Along with the amanary was a : comminnene to update the Waterford Unit 3 RCS P-T lhnits and review the adequacy of the surveillance' withdrawal eharhile, both contained in the Waterford Unit 3 - u Technical SpeciScations (Reference 5). a This report provides revised RCS P-T limits for Waaerford Unit 3 valid through - twenty (20) effective full power years (EFPY) operation, along with a description of j the data and marhnda associated with their dia'- -- '. In addition, a revised. survedlance withdrawal eharhale has been provided along with the basis for its development. ) ,i n [ ,? q f.) .1

I 1

ABB Coahurion Engsanersag beiner Sernent 1 C-MECH-Dt-021, Jter. 00 .a

a .2.0 RCS PRESSURE-TEMPERATURE IlMITS 2.1 OVERVIEW The basis for the development of reactor coolant system pressure-temperature l limitations for the Waterford Unit 3 Steam Electric Station is provided in the subsequent subsections. ' RCS P-T limits were developed to meet the requirements associated with 10 CFR Part 50 Appendix A (Ref. 6), Design Criterion 14 and Design Criterion 31. These design criteria require that the reactor coolant pressure i boundary be designed, fabricated, erected, and tested in order to have an' extremely { low probability of abnormal leakage, rapid fadure,.and gross rupaut. The criteria. also require that the reactor coolant pressure bonadary be designed with sufficient ' margin to assure that when stressed under operating, maineenance, and testing the ~ boundary behaves in a non-brittle manner and the probability of rapidly propagating ' fracture is minimited. Specific requirements regarding the development of pressue-temperature limits are l provided by 10 CFR 50 Appaadir G (Ref.1) which fonns the general basis for these limitations. 10 CFR Part 50 Appaadir G mandates margins of safety against fracture-equivalent to those recommended in the ASME Boiler and Preseme Vessel Code, Section III, Appaadir G. Protection Against Nonductile Faihue (Ref. 7, referred to herein as ASME Code, Ap-lir G). 'Ibe general guidance provided by the ASME Code, Appaadir G procedures has been uritived to develop the revised Waterfont Unit, 3 reactor vessel beltline pressure-temperature limits with the requisite margins of safety for the beamp, cooldown and inservice hydro test candirians. Within the RCS, the reactor vessel beltline region is subjected to sufHcient neutron irradiation to aber the maserial fracane toughness properties of the beltline material. 'the afhcts of asesses irradiation has been inchidad in the development of the beltline pressuse4separaense limits and based on the irradiarian damage pr=dW methods of Reguissory Guide 1.99 Revision 02 (Ref. 8). This methodology has been used to : enicalmen the limiting mesetial Adjusted Refierence Temperasess for Waterford. Unit 3 2 and has neili,ad peak I.D. surface fluence value of 2.29 x IOP n/cm corresponding to approximately 20 Effective Full Power Years (EFPY) of operation. Ass cAmmon Eagbueng klaar Wear /CMECH Eit-021, Jter. 00 1 m =. -, ;j

l The following sections describe the methodology associated with 'the development of - the reactor vessel beltline ptessure-temperature limits. The transients analyzed were: the isothermal condition, linear heatup rates of 30, 50, and 60'F/hr, and linear cooldown rates of 10, 30, and 100*F/hr. The beltline limits for inservice hydrostatic test cori@wiing to an isothermal condition were also developed. l In addition to the beltline P-T limits, the additional requirements of 10 CFR 50 Appendix G and ASME Code have been developed to provide P-T limits appropriate j for the RCS. j 2.2 BASIC DATA Reactor Vessel Data Reference 2500 psia 9 Design Pressure = Design Temperamre = 650*F 9 1 2250 psia 9 Operating Pressure = 8.625 in 9 Beltline Thickness = Inside Radius (to wetted 86.971 in 9 surface) = 0.21875 in 9 Cladding 'Ihickness = Material-SA 533 Grada B C1 men 1 Refenmce 23.8 BTU /hr-ft *F 10 Thermal Conductivity = 28 x 108 pai '10 Youngs Modulus = 7.77 x'10%/in/*F 10 ' Coefficient of Thermal = E=== ion 0.122 BTU /lb *F. Specific Host = 490 lbm/ff 11 Density = Stainiana Samal Cladding 10.1 BTU /hr-ft *F ' 10 - Thennal Conductivity = 0.126 BTU /lb *F - Specific Heat = 493 lbm/ff ~ 11 Density = . ins rM= Engmeerug Nudent Servicar '3-. C-MECH-ER-021, Rev. 00

2 Convective film coefficient on vessel inside surface = 1000 BTU /hr-ft 'F 2.3 ADJUSTED REFERENCE TEMPERATURE PROJECTIONS-To permit the development of revised beltline P-T limits, it was &===n to calculate the adjusted reference temperamres (ARTS) for the Waterford Unit 3 beltline materials to establish the limiting matenal. The ' djusted reference temperamres'of a each reactor vessel beltline material were calculated at the pa=ad=H cracle tip inherent with the beltline P-T limit analysis (1/4t and 3/4t) locations after 20 EFPY of ' operation. The controlling material for Waterford Unit 3 was determined by - comparing ART values for each matenal. t 1 The adjusted reference W-pr. sun (ART) were calculaasd using the procedures in - i Regulatory Position 1.1 of Regulatory Guide 1.99 Revision 2 (Ref. 8). The procedure for e=Wdadae ART values for a material in the beltline is given by the ~ following expression: ~ ART. = Initial RTm + ART,er + Margin (1). l The initial RT,arr is the reference temperature for the unirradinand material. The [ ART,er is the mean value of the adjusement in the reference temperamre caused by - irradiation and is given by the following expression: ART,er = (CF) f(0.28 - 0.10 log f) (2) CF is the eb= -d factor for the bettiine maserials widch is a fbaction of the weight gms copper and nickel for the material. Regulatory Guide 1.99 Revision 2 provides chandstry thceors for welds and for base metal planes and forgings. The j tena f is the nontma fluence at any depth in the vessel... The neutron fluence at any depth is given by the following expression: r. y,(4.24x) * - (3). 2 The term f,,, is the neutron fluence (having units of 10" n/cm E >J IMeV) at the - inner wetted surface of the vessel, and x is the depth into the vessel wall from the Ass comhamien Engnennng Nncinar Services 4~ . C-MECH-B.021, Rev. 00 l

) inner wetted surface (inches). ~ In this instance, the reactor vessel stainless steel l cladding is ignored. i Margin is the quannty that is added to obtain a conservative upper bound value of j ART. 'Ihe margin term is given by the following cxpression: Margin = 2(ai + oi)v2 (4). The terms a and a4 represent the standard deviations associated'with the' initial RTm s and the snean value for the reference t +rature shift, respectively. J The following information provides the basis for the calenlarad ART values associated with Waterford Unit 3 reactor vessel beltline: 1) Unirradiated beltline material data was obtained from Tables 5.2-6 and 5.2-13. of Reference 12, including copper consent, nickel consent and initial reference t.+idure (RTm). This data is ennmarized in Table 1 for Waterford Unit l j

3. It should be noted that in the process of =rahaishing Charpy upper-shelf.-

I energy values for the Waterford Unit 3 reactor vessel beltline welds, an l inconsistency was identified for longitudinal seen 101-1428. A weld repair ' was idennfled which identified a diftlerent weld wire best and flux, along with. a different initial RTurr and nickel consent (w%). ' Ibis information was. q considered in the determinarian of the limiting beltline material. However, the - l .1 determinarian as to the extent and location of the weld repair was not within the scope of this effort. The resuhs of this inconsiseency have had no bearing j on the results of this evaluation. d 2) Pealt neutros fluence for the Waterford Unit 3 Meline region was determined n/cm (E > 1MeV) at 20 EFPY. This was calculated by 2 to be 2.29 x 108 linear interpolation using fluence values for 6 EFPY 'and 32 EFPY 6f1 1 a n/cor, respectively, ahraiand from the' d 6.47x10is f,,,a and 3.69x108 2 surveillance capsule evaluation report (Ref. 3). 3) Calculations were based on the procedures in Regulatory Position 1.1 of NRC Regulatory Guide 1.99, Rev. 2 (Ref. 8). Uncertiinty in initial RTmyr ~ was : taken as O'F for measured values of initial RTm,r (*: values for 'the beltline . ABB Combumme Engumeanng Nmcinar Servecer i 3: 1 C-MECH ER 021, Rev. 00. .)

~_ i 1 matenals were not required due to measured values of initial RTm. See ; =nk=_== discussion for the associated technical basis.) 1 4) The effect of an 8'F reduction in reactor coolant cold leg temperature on .RTm hift was considered as recommended by Reference 13. _ This reduction s in reactor coolant cold leg temperature, from 553*F to 545'F, is still within a L the irradiation temperature range for which the methods provided by in : Reference 8 are valid. Consequently, the margin tenn,"a, currently accounts a for uncertainty in the prediction of RTm hift which could result from this d s type of irradiation environment vanability and was assessed to be adequate. ~ According to Position 1.1 of Regulatory Guide 1.99, Revision 2 (Ref. 8(the? uncertainty in the value of initial RTm is to be estimated from the' precision of testL method when a " measured" value of initial RTm is available. RTm is derived in'- accordance with NB2300 of the ASME Boiler and Pressure Vessel Code, Section M. It involves both a series of drop weight (ASTM E208) and Charpy impact (ASTM j E23) tests on the material. The RTm resulting from these two test methods of .i evaluation are conservatively biased. The elements of this conservatism include: 1)' Selection of RTm is the higher of NDTT or Tcy - 60*F. ' The drop-weight test is performed to obtain NDTT and a full Charpy impact curve is developed L ~ l to obtain Tcv for a given material. 'Ihe combination of.the two test methods l gives protecnon agamst the possibility of errors in 6 either test and ; with the full Charpy curve, demonstraans that the material is typical of reactor. pressure vessel steel. Choice of the more conservative of the two (i.e., the higher of NDIT or Tcy - 60*F) assums that tests at temperemresicrave the ; j reference temperamre will yield incienseg values of toughness, and verifies. y the temperamre W of the fracare toughness implicit in the K curve 1 (ASME Code, Section E, Appendix G). t 2) Selection of the most adverse Charpy results for Tcy. In accordance withL NB2300, a teenpernaut, Tev, is established at which three Charpyl specimens j exhibit at least 35 mils lateral expansion and not less.than 50 ft-lb absorbed energy. The three specimens' will typically exhibit a range of lateral expansion j and absorbed energy consistent with the variables ~~mberent in the' test:1

i specimen temperature, testing equipment, operator, and test specimen (e.g q

l ') ABB Comhunion Espuwmg kker SeMca

C-MECH-ER 021 Rev. 00 6

~, - - -. ,.____.__.._-___________._______._._.m'

dimensional tolerance and material homogeneity). All of these variables are controlled using process and precedural controls, calibration and operator training, and they are conservatively bounded by using the lowest measurement of the three specimens. Furthermore, two related criteria are used, lateral expansion and absorbed energy, where consistency between the two measurements provides further assurance that they are realistic and the material will exhibit the intended strength, ductility and toughness implicit in the Km curve. 3) Inherent conservatism in the protocol used in performing the drop-weight test. The drop-weight test procedure was carefully designed to assure attamment of explicit values of deflection and stress concentration, eliminating a specific need to account for below nominal test conditions and thereby guaranteeing a conservative direction of these uncertainty components. In addition, the test - 4 protocol calls for decreasing temperstme until the fug failure is mmemd, followed by increasing the test k-pereture 10*F above the point where the lag failure is encountered. This in fact assures that one has biased the resulting estimate toward a low faihus probability region of the temperature versus failure rate function diagrammed below. The effect of this protocol is ~ to conservatively accommodate the W.6,3 uncertainty components. 1 l l I ~' f =.7 '- =g-yn - gggy. g =~. .y =.gs;., t,, ? K ? n%h6c - .c ' -?;;'.g mg ( [,3 . g..g e m m -v

==h4 Li./ -,%,;1 ' ^ 1W j \\, . },. / "WGlC ...'_. L _ 4,,.,

w..

...g s p-n.,,. -i., - ' *. 1 1 ?': i =.. . f= } ~ m. ?, M 3 j - + J 4 4 - c-7 . !%.m j - f 5 ;- ., 5. .. ; 2 ^ ; w;;; . "WL. :- - = ...;z ;g 4 i g yi t, C.t , A h&=?~:f_. 1 Y1%.

)

2 4 4 : + N W@ Wi: ' ) 7in .s. nwh s v. ;. r Mi ~ ^ 7N* I " - 4 e5B=$2"4E F",5 S' = .".2, G ib a. j n= g=L== = h.l \\ - 3 =x-g. cg: . a q L= i T ugw4ggE== = M=g3R-_..:.y3 g W w i c^ ABB Combanon Engunnering M SeMcas 'l C-MECH-ER-021. Rev. 00

7.3 y b-w s n Given the three elements of conservatism described above, values of initial RT,sor ,j obtained in accordance with NB2300 will result in a conservative measure;of the-l reference temperanue. The conservative bias of the NB2300 methodology and the ' j drop-weight test protocol essennally eliminam the uncettaimy which migin result from 1 the precision of an individual drop-weight or Charpy impact test.' 'Iherefore, when l measured values of RTm are available, the edimaw of uncertaimy in initial RTwar is j taken as zero. 1 i Adjusted reference temperatures for all beltline materials at the 1/4t and 3/4t locations g through 20 EFPY were calculated using Regulatory Guide 1.99 Revision 2 and the i controlling material can be established from the'results of the material evaluation l shown in Table 2. The term "comrolling" means having the highest ART for a given j time and position within th;; -essel wall. The highest, or limiting, ARTS are then j used to develop the beltline pr.msure-temperanne limits for 'the corresponding time * ' period. 1 L I In the case of Waterford Unit 3, the limiting material at bc,th the 1/4t and 3/4 4 locadons after 20 EFPY is plate M-1004-2 based on the predicasd ART values of 65.4*F and 54.0*F, respectively. 2.4 CALCULATION OF REACTOR VESSEL BELTIlNE P-T LIMITS i 2.4.1 General Method a

q The andytical procedure for developing reactor vessel pressme-temperature limim utilizan the methada of linear Elastic Fracane Mach==iem (LEFM) found l

in the ASME Boiler and Pressme. Vessel Code (ASME Code) Section III - Appendix 0 (Reference 7) in accordance with tbs requireemens: of 10 CFR Past 50 Appendix G (Reference 1). For these analyses, the Mode I (opening ' I runde) sm insensity factoo are used for the solution basis. 'Ibe general methoit utilizes Linee.' Elastic Fracane Mach==ien procedures., Linear Elastic Fractme Mechanics relates the size of a flaw with the allowable i loading which precludes crack initiation. This relation is based upon a i mathematical stress analysis of the beltline maserial fracture toughness j properties as prescribed in Reference 7. i i ) 1 l Ass Com6msnos Engusemqr &cisar Snics .. 1 C-MECMR421, Jter. 00 '8 ~ .J

l r

.c The reactor vessel beltline region is analyzed assuming'a semie ipir.' surface n flaw oriented in the axial direction with a depth of one quaner of the reactor vessel beltline thickness and an aspect ratio (depth to length) of one to six. THs poscalated flaw is analyzed at both the inside diameter location (referred j to as the 1/4t location) and the outside diameter location (refened to as the 3/4t location) to a sure the most limiting condition is achieved.lThe above flaw geometry and orientation is the maximum pomminend defect size (reference flaw) described in Paragraph G-2120 of Reference 7. -At each of the pel='M flaw tip locations, the Mode I stress insensity factor, K, produced by each of the specified loadings is calenta==i. The mammarion. j i of the K, values is then compared to a reference stress insensity, K., which is. the critical value of K, for the material and tempernaus involved. The result ; of this method is'a relation of pressure versus temperanue for each condition analyzed providing reactor ves6el operating limits which preclude brittki fracave. In accordance with the ASME Code Section III Appendix G requirements, the' general equations for hiait the allowable pressure for any assumed rate - j of tempernaut change during Service level'A and B operation are: i 2Km + Krr < K,a

(5)

J q and j 1.5Km + Kn < K. (Inservice Hydrostatic Test) (6)

where, Km = Allowable pressure suess intensity factor Ksig -

.Krr = Thermal stress imensity factor, Ksi8 'K. =. Reference stress intensity,'Ksi6 4 r 4 9 C MECH-B4D, Rev. 00

_ 7; i e !The reference stress intensity Km is defined by Paragraph G-2110 of Reference 7 as Km =~ 26.78 + 1.223 e[0.0145(T-ART + 160)]- g o - where, r reference stress intensity factor, Ksi6 Km = temperamre at the posmlated crack tip 1*F T = ' adjusted reference nil ductdity temperamre at the-ART = po*a w crack tip, 'F l At any instant during the posmlated beamp or cooldown, K is entent***d based on the metal temperamre at the tip of the flaw and the adjusend reference temperamre at that flaw location. _The temperamre gradients across the reactor vessel wall are also entent**ad for any instant during the hoseep or cooldown j i (see Section 2.4.2) and the wi=;-:--

thennel stress insensity factor, Krr.

is determined. The thermal stress insensity is subtracted fkoot the available K. l. to determine the allowable pressure stress intensity factor and meJy the j allowable pressure, j t 'Ihe pressure-temperature limits provided in tis report ace==* for the ' j temperanus difEeremial between the reactor mal base metal and the reactor. coolant bulk timid temperamre. Corrections to account fbr pressee differentials between the location of concern and the location of mammirement dos to elevation differences and RCS flow are included in the development of. the presente4amperanne limits. ~ Uncertainties for the temperneut and preseme instr==nearian loops associated with control roont indientions are also inehmend c.- 7My, the P-T limits are provided on coordinates of ? l indicated pressurimeir pressme versus indicanad RCS esaperanse. l i 1 J The reactor coolant system pressure measurement is taken from the : pressurizer. The differential pressure due to the elevation difference between j the reactor vessel beltline wall and the pressunzer was conservatively j 1 l -l ABC r"= Eagunnerung Nednar Servecar ' C-MECR Elt-021,' Rev ' 00 ' 10 )

a 4 established and equal to 36.04 psia for all temperatures. Tbc pressure differential due to the flow induced pressure drop between he reactor vessel inlet nozzle and outlet nozzle was established based on four (4) reactor coolant pumps operating and is equal to 34.71 psi. The pressure differential associated with hot leg flow induced pressure drop was estimated to be 0.1% psi. The uncertainty associated with the pressure indication instrument loop was established as t28.34 psi for the narrow range instrument and 114.89 for the wide range instrument. This information was combined to determme the following pressure correction factor utilized in the development of the P-T curves: Actual Pressure (P) Ranee Total Pressure Correction Factor (PCF) P < 200 psia -186 psi 200 psia s P < 850 psia -100 psi 850 psia s P s 3000 psia -186 psi The urx:ertainty associated with the temperamre iMWrion instrument loop was also included. This value was established to be 25.6*F (Reference 14). By explicitly acconntmg for tbc temperature differential between the flaw tip base metal temperature and the reactor coolant bulk fluid Egrature, and the pressure diffemntials between the beltline region of the reactor vessel and the pressurizer inc uding tbc uncerrsinties associated with the indication loops, the P-T limits are correctly represented on coordinate < of iMkirM pressunzer pressure and inlicated cold leg E@amre. 2.4.2 Thermal Analysis Methodology The Mode I thermal stress intensity factor is obesinM thrtxigh a detailed thermal analysis of the reactor vessel beltlirr W1 using a computer code. In this code a one dimensional thite noded isoparametric finite element is used for performing the radial conduction-convection trsnsient beat transfer analysis. The vessel wall is divided into 24 elements and an accurate distribution of temperature as a fumtion of radial location and transient time is calculated. The code utilizes a convective boundsry condition on the inside ABB Combum'on Enyneenng Nudear Savica II C-MECH ER-02L Rev. 00

_ ~ 3 wall of the vessel'and an innilatad boundary.on the outside wall of the vessel. Variation of material pi+,Ges through the vessel wall are permitted ' llowing - a for the change in material thermal properties between the' cladding and the. base snetal. In general, the temperature distribution through the reactor vessel wall is governed by a partial differential agn=rian,- BT bat + 1 87 p C, y = K y subject to the following boundary conditions at the inside and outside walli surface locanons: Atr = ri -K = h (T-T, ) f = 'O At r = r,

where,

~ 3 density, Ib/ft. = p specific heat,~ BTU /lb *F - C = P thermal cd%, BTU /br-ft *F - K = vessel wall temperamre, *F T = radius, ft r = time, hr t = 2 convective heat transfer coefficient, BTU /hr-ft,.p h = RCS coolant temperanne,.*F' T, = r,r. = inside and outside radii of vessel wall, ft' e The above is solved amnencally.using a finies element model.to deternune wall tasapernaus as a funr.: tion of radius, time, and thermal rase.- Thermal sness insensity factors are calculated using a superposition technique. and influence coefficients specifically generased for this purpose.11he ~ influence coefficients depend upon the geomeuy of the.. ..... postulated -

Ang em-Wng M M 12 C-MECH B 02!, Rev. 00

~ - d y 4 . defect, the geomeny of the reactor vessel beltline region (i.e., r /ri, alc, alt where a= crack depth, c= crack half length,' and t= vessel wall thickness), and - ~ the assened unit loading.. The alternate method employed utili=1 a third u order polynomial fit of the temperamre profile and the respective influence _q coefficient (uniform, linear, quadrauc and cubic) to calculate each profile - j contribution to Krr. The total Krr was the summarian of all the contributions. l The influence coefficients were calculated using a detailed 2 dimensional finite element model of the reactor vessel. The infhaence coefficients'were corrected j for 3-dimensional effects using ASTM Special Tachnical Publication 677. j (Reference 15). 1 l j ASME Code Section HI Appendix G recognizes the limitarians'cf.the method it provides for calculating Ker because of the assumed temneramre' profile. : An _ alternate method for emienI= ting Krr was employed as required by l Subsubparagraph G-2214.3 of Reference 7 to account for the varying temperamre profiles (and u-- - - =ly varying thermal stress intensities) resulting from a detailed heat transfer analysis. 2.4.3 Heamp Limit Analysis During a beacup transient, the thermal bending stress is compressive at the reactnr vessel inside wall and tensile at the reactor vessel outside wall. Internal pressure creates a tensile saess at both the inside wall and the outside wall locazices. Cor.; p 3y, the outside wall location has the larger total : j suuss when compared to the'inside wall. However, nonmm embrittlement, the;, q shift in maaerial RT er, and the associated reshaction in flacnue tan =haa== are ! i i greater at the inside location than.the outside. Therefore, both the inside'and j outside flaw lae=tiana must be analyzed to asses that the most limitiai-caadietan la achieved.

i As described in Section 2.4.1, the reference stress insensity factor is calculated for the metal tasaperanue at the tip of the flaw and the adjuned reference temperanne at the flaw location. During beamp, the reference stress intensity -

is calculmaad for both the 1/4t and 3/4t laratians. The temperanne profile through the wall and the metal temperanuts at the tip of the flaw are_ ~ calculatad for the transient history using the finise element method described in - j ^ t i i E ] I3 C-MECH Eit.021, Jter. 00 '

WQgn ' 4 y Section 2.4.2. This information is used to c% the thennat stress intensityj j factor at the 1/4t and 3/4t locations using the calculated wall gradient and 1 thermal influence coefficients. The allswable pressure stress intensity is then determinad by rearranging equation 5 to give q Km = (Km - Krr)/2 (8)- j The allowable pressure is then derived from the' calculated allowable pressure stress intensity factor. Influence coefficients due to unit pressure loadings j have been developed through detailed finire element analyses. These influence ; coefficients permit expedient conversion of the stress intensity factor to an allowable pressure. N It is interesting to note that a sign change occurs in the thermal stress through [ the reactor vessel beltline wall. ' Considering a reference flaw at'the 1/4t j locatiot, it can be shown that the thermal stress tends to"allevisse the' pressure stress. This would indicate that the isotherssal sesady state condition would y represent the controlling P-T limit. However, the isothennal condition may ' not always provide the limiting pressure-tesaperature limit for the 1/4t location j during a heatup transient.L This is'due to the correction of the base metal temperanut to the Reartor Coolant Symesa (RCS) fluid Ismperstme at the j inside wall by accounting for clad and flha temperanus differentials. u For a given beanap rate (non isothermal), the differential asaperanut through - ) the clad and flha increases as a function of thermal rate resulting in a higher-RCS fluid esaperseme at the inside wall than the isothermal condition for the same flaw tip asaperanne and pressue. Therefore, to ensure the accurate. l d= of the 1/4t pressure-temperanue limit during bestup, both the j ..y. isothermal and bestup rate W pressme-temperature limits are calculated l to ensus the limiting condition was' achieved. 'the limits for both heacup 'and - i cooldown accans for clad and film differential temperatures and for the - gradual buildup of wall differential temperanuts with time, d I 1 i -L Nueiner Semicar j ABB Canhastion E-C:: i 14 ' a . -C-MEG-Dt-021, Jter. 00 m

1 ~! y u At the 3/4t location the pressure suess and thermal stresses are both tensile, j resulting in the maximum stress at that location. Pressure-Wr. core limits-' ~ were calcularari for the 3/4t location accounting for clad and fihn differential j temperature and the buildup of wall temperature gradients with time using the-method described in Section 2.4.2. The allowable pressure based upon a flaw' at the 3/4t location was derived in the same way as if the flaw were at the 1/4t location. With the K mi and Kn. both calculated, equation 8 is used to solve for.- the allowable pressure stress intensity, Ks.' 'Ihe allowable pressure is then calculated using K. To develop composite pressure-temperamre limits for the beanup transient, the ~ q isothermal,- 1/4t hearup, and 3/4t beanup pressure-temperanut limits are I compared for a given thermal rate. The most restrictive pressure-temperamre U limits are then compiled for each analyzed temperarme throughout the transient ' i duration, resulting in a composite limit curve for the reactor vessel beltline forj j the heatup event. J Table 3 provides the results pressure-tempenmus limits for linear beanup raans l of 30,50, and 60*F/hr. The allowable pressus is in unis of psia, land the temperamre is in units of 'F. Figme 1~ provides a graphical pe==awarian of, - q the heatup pressure-temperstme limits found in Table 3. It is permissible to - ~ linearly interpolate between the beanup pressme-temperstme limits. l 2.4.4 Cooldown Limit Analysis j Durms cooldown, membrane and thermal bending stresses act together in" i tension at the reactor vessel inside wall. This results in the pressure stress { inensity fksor, Kee, and the thermal stress insensity factor, Krr, acting in : unison creating a high stress intensity factor. At the resceor vessel outside l wall the tensile pressure stress and the compressive thermal stress act in. opposition resaking in a lower total suess insensity factor than at the inside3 .l wall location. Also' neutron embrierlanar the shift in RTer, and the i associated rahe in fracture toughness are less severe at the 'outside wall. j when compared to the inside wall location. Consequendy. the'inside flaw - location is always more limiting and is analyzed for the cooldown event.- ] 1 Aas Canharios Enginasmg Masr Serms: . C-MECH-ER421, Rev. 00. 15

)

1 1 The reference suess miensity is once again determined using the rnaterial= l metal temperamre and adjusted reference temperanut at the 1/4t location. j From the method provided in Section 2.4.2, the through wall temperanne gradient is calculated for the nemmed cooldown' rate to determine the thermal stress intensity factor. In~ general the thermal' stress miensity factors are found using the temperature profile through the wall'as a function of transient time as.. described in Section 2.4.2. They are then subtracted from the available K. I value to find the allowable pressure stress intensity factor and' consequently the : allowable pressure. ] l Pressure-temperature curves are.e i.;.;.4 for cooldown transients the same way they are generated for heatup transients. The allowable pressure stress a intensity at the 1/4t location is calculated using equation 8, and the allowable - pressure is calculated from the allowable pressure stress insensity factor, Km. a To develop a composite pressure-temperature limit for a specific cooldown ; l event, the isothermal pressure-temperamre limit must be calcula==i. : "the j isothermal pressure-tempername limit is then compared to the limit associated with theLspecific cooling rase, and the more restrictive of the two limits is chosen to result in a composise' limit curve for the reactor vessel beltline. .i Table 4 provides the results for the isothermal condition'and the linear rates of - 9 10,30, and 100*F/hr cooldown. The allowable pressee is in units of psia,- and the temperamre is in units of 'F. : Figme 2 provides a graphical pr-aan of the cooldown presamHemperamre limits found in Table 4. It1 j is permissible to linearly imerpolate between the linear cooldown ptsssure - a temperange limits. 2.4.5 Hydrostatic Test Limit Analysis Both 10 CFR Part 50 Appendix G and the ASME Code Appendix'G ' require the development of pressure-temperanne limits which are applicable.to inservice hydrostatic tests. For hydrostatic tests performed subsequent to' j o loading fuel into the reactor vessel, the' "' n" ". test tempernaus,ist l determined by evaluating the Mode I stress intensity factors. The evaluation ~ ) of these factors is performed in the same manner as that for normal operation: J - ARE Canbasanon Enguaernng Mdear Servicar l C-MECH-EJt.021, Jter. 00 16 J =

] d 3 t heamp and cooldown conditions with two differences. Eqa=% 6 shows that : l the safety factor applied to the pressure stress intensity factor is 1.5 instead'of _ l 2.0. _ ~ Also, the' inservice hydrostatic test limit is established based upon an isothermal condition. This eliminar*< the thetnal stress intensity factor, Krr.., l from ~;aarian 6. The inservice hydrostatic test limit is provided for 20 EFPY in Table 5 and is; shown in Figure 3. "Ihe rnininnun temperstme for the inservice hydrostatic test pressure can be conservatively determined using the g*~* of ASME j Code Section XI, Subarticle 2500 (Reference 16) using the curve developed R for inservice hydrostatic test. A test temperanus (imlicatart)' equal to 219.8'F is necessary for the selected test pressure of 2475 psia (1.1 times normal - ) operaung pressure). i 2.5 CORE CRITICAL LIMITS q Pressure-temperamre limits for core critical operation are specified in 10 CFR 50, Appendix G to provide additional margin during'acmal power operation. The j pressure-temperanne limit for core critical operatica is 1 eW upon two criteria. - These criteria are that the reactor vessel must be at a teur.vais equal to or greater

• than the mininnun :emperature required for the inservice hydrostatic test'(219.8'F);

and be at least 40*F higher than the..'.'..... pressure-temperamre curve for nortnal j operation heatup or cooldown. j Note that the core critical limits established above are solely based upon fracture - mechanics considerations and do not consider core reactivity safety analyses which can control the temperstme at which the cott can be brought critical. 2.6 FLANGE IJMIT5 1 As sensed in Raisrence 1, the temperature of the closure flange regions must exceed the initial rte,r f the maderial by at least 120*F for normal.Wa 'Ibe i o ts-gr. sue must exceed the initial RTm by at least 90*F'for hydrostatic tests and ~ N leak testing when the pressure exceeds 20% of preservice hydrostatic test pressure. m Cmm-1I Ii C-MECH EH21, Rev. 00 - ~..Ja,-- l

v; e (y -l n Accouming for instrument uncertaiory, the flange linuts were calculated to be (given i an initial RT,er f 20*F) 165.6*F for normal operation and 135.6'F for hydrostatic o testing. These are the minimum temperamres in the flange region for the pressure to. l ~ exceed 20% of the preservice hydrostatic test pressure. .i A review of the original design basis was performed'which idenufied flange limits l developed using the guidance of ASME Code, Section III Appendix G. - This provided a flange limit associated with a heatup rate of 50*F/hr which was adjusted j with the correction factor from Section 2.4.1. 'Ihe flange limit is provided below in terms of indicated pressure and temperature. 50*F/hr heatup indicead T, (*F) Indicated P, (psia) 215.6 618 lo 265.6 3991 l The design basis value is more restrictive while meeting 10 CFR 50 A,Wir Gi l requirements and shall be used in the development of the Technical Specification. j d 2.7 LOWEST SERVICE TEMPERATURE The Lowest Service Temperature is the.J a..... allowable temperamre at 'which pressures can exceed the' pre-operational system hydrostatic test pressure (625 psia q uncorrected). This temperature' is defined by Paragraph NB 2332 of ASME Code Section III (Reference 17) as to the most limiting RT,er for the balance of Reactor l Coolant Sysuun (RCS) components plus 100*Fc The rnaximum RT,er for the balance of the RCS components was conservatively established to be 90*F and was associated with the reactor coolant panp.. Therefore, l the Inwest Service Temperanne is equal to 100*F + 90'F + 25.6*F = 215.6*F, I whers 25.6*F is the temperanut instrument uncertainty. a 2.8 MINIMUM PRESSURE - .l The minirnum pressure limit is defined as 20% of the pre <gerational hydrostatic test ~ pressure. Therefore, the mmimum pressure is 625 psia uncorrected. After j application of the e ywy(. ate pressure correction factor, the indicateri minimum r t]

as r"- t

m maar sama 18 -

. C-MECH-ER 021,' Rev. 00

J -e,-. 'g ptessure was calculated to be 525 psia. The minimum pressure shall not be exceeded prior to achieving a coolant temperamre equal to the lowest Service Temperature. - 2.9 MINIMUM BOLTUP TEMPERATURE 1 The minimum boltup temteramre is the rninimum allowable temperature for the flange to be stressed by the full intended bolt preload and by pressures less than or ~ y equal to 20% of the pre-operational system hydrostatic test pressure. The mammum - 1 boltup temperamre is defined by Paragraph G-2222 of ASME Code Appendix G. (Reference 7) as the initial RTer for the material'of the higher stressed region of the i reactor vessel plus.any effects for indiation.. The mariannn initial RT,er associated with the stressed region determined to be 20*F. The mininmm boltup temperature including temperstme instrument uncertainty is 20'F + 25.6*F = 45.6*F. However, for additional conservatism it is recommended that the cw.u Jf specified inriirstari temperamre of 72*F cominue to be used. i l i I I i i 1 I J i l 1 l L ABS Condeunes Engissang Acisar SmMcar R C-MECH AR421, Jtes. 00

N y ~ 3.0 LTOP ENABLE TEMPERATURES ] Standard Review Plan 5.2.2, Overpressure Protection (Reference 18) - has dermed the 5 temperanut at which the Low Temperamre Overpressure Protection (LTOP)' system - ~ -should_be operable during starmp and' shutdown conditions. This temperanue,-icnown. as the LTOP enable temperature, is dermed as the water temperanue conesponding toi 'l a metal temperamre of at least RTmrr + 90*F at the beltline location (1/4t or 3/4t) that is controlling in the Appendix G calculations. - The ASME Section XI Code Case; y N-514 (Ref.19) suggests a coolant temperamre which corresponds to a metal q temperamre of at least (ART + 50*F) at the beltline location or 200*F, whichever isl .] greater. - The results of this method (referred to as the N-514 method below)' are; [ included for informanon only at this time. ne LTOP enable temperanue for cooldown is based on the isorbermal case. The 1/4t location is limiting for both cooldown and isothermal cases. During a cooldown

j transient, the temperanus at any point in the wall is higher than the coolant j

temperanus. Therefore, it is conservative to decennine the LTOP enable temperanne for cooldown based on an isothermal case. In an isothennel condition'the water temperamre is equal to ART + 90*F. Including instnument uncertaissy, the LTOP .j enable temperanue for cooldown was established to be'181*F. 1 LTOP enable temperanuts for beamp were desennined for each heemp rate at both. the 1/4t and 3/4t locations. The values and controlling locations are provided below. ]a 1 LTOP Enable Temperannes (SRP 5.2.2) j Hemmp,20 EFPY l 1 1/4t flaw location 3/4t flaw location ] 30*F/hr 193.7'F (3/4t limiting). 192.3*F.(3/4t limiting) -i 50*F/hr 201.8'F (3/4t limiting) ' 206.8'F (3/4t limiting) - 60*F/hr 205.6*F (3/4t' limiting) 213.6*F (3/4t limiting)- i Review of these values along with consideranon of the allowable hentup rates and - -j respective temperanut ranges _ provide an LTOP enable temperanne of 206.8'F-i 007'F) including instrument uncertainty. Above this value, the primary safety valves. j provide adequate protection. l ass cm=6mnen Ear====t A dasr5- = =

C-MEG-ER 021. Jtes. 00 20

A x The LTOP enable temperatures were c*aM using the criteria (ART + 50*F) of Code Case N-514. In the case of both bestup and cooldown, the values were less ] than 200*F. Consequently, the LTOP enable Eg stures for heatup and cooldown would be 200*F based on Code Case N-514. .i 'I

1'.

} b l .j + 1 i MlluD3 ' ?l . ABB Canburrion E.l :_---4 Nancinar Servicar u.

N r 4.0 PRESSURIZED 11IERMAL SHOCKFI'S) SCREENING CRITERIA In accordance with 10 CFR 550.61 (Ref. 20), the pressurized thermal shock (PTS)- criteria were evaluated with this update of the pressure-temperature limits to ensure a ~

complete submittal. Once the value of RTm is calculated for all beltline materials, it

1 must not exceed the PTS screening criteria of 270*F.for plates, forgings, and axial-1 weld materials and 300*F for circumferential welds. As required by 10 CFR 650.61,.

lj the RTm is calculated using the following equation j .l RTm = I + M + ARTm (9)~ where I is the initial reference temperamre of the matenal, M is a margin added to 1 cover uncertainty in the initial RTm,r, and ARTm is the shiA in the initial reference. temperature caused by irradiation. _ Measured' values of I were used, so M was set to j 56'F for weld materials and 34*F for base metals. The ARTm term is enward at the vessel inside surface using the same equadon for calculating ART,er (eq. 2). j Material information from Table 1 and an estimated peak surface fluence of 3.69x10* - n/cm corWian to 32 EFPY (from Ref. 3) were used to calculate RTm for all 'l 8 the beltline matenals, and the results are prh in Table 6. -d 1 ^ 6 4 s

1 F

l l i 3, 22 1 C MECH ER-021.'Jter. 00

5.0 END-OF-LIFE UPPER SHELF ENERGY 10 CFR 50 Appendix G requires that the reactor vessel beltline material maintam a. l. Charpy upper-shelf energy (USE) of 50 ft-lbs throughout its operational lifetune. 1 Reguistory Guide 1.99 Rev. 2 (Reference 8) provides a method to predict the.' decrease in Charpy USE which is based on initial Charpy USE, accumulated fluence and copper content (w%). To ensure compliance with 10 CFR 50, Appendix G. the 4 decrease at end-of-life (32 EFPY) was calculated through strict application of Regulatory Position 1.2 of Reference 8. Values of initial Charpy USE were obtained, in most insrances, from Reference 12. To evaluate the beltline welds,1Waterford Unit 3 original fabrication record's were used to ascertain the initial Charpy USE values. The initial and end-of-life (1/4) Charpy USE values are summarized in Table 6. i I l 0 3 ABB Combasmon Eaghuertng hciner Sermns ' C-MECH-Eit421, Jtev. 00 23-

i I '6.0L _ SURVMILANCE CAPSULE WITHDRAWAL SCHEDULE g q ~ L 'Ihe Waterford 3 teactor has 6 surveillance capsules designed to monitor the changes in beltline material properties (Ref. 21). -The governing withdrawal schedule'for these capsules as required by 10 CFR 50. Appendix H is defined in Table 5.3-10 of thez F Waterford 3 Final Safety Analysis Report (FSAR) (Ref. 22). This current withdrawal. j schedule is presented in Table 7 along with the capsule identification number and original target fluence as presented in Reference 3. Capsule 2, located at the 97 degree position, (also referred to as capsule -W-97) was j removed, and the maamlated specimens were tested. A major result in the W-97 1 capsule report pertment to the capsule removal eharhile was a change in the capsule : j lead factors. The lead factor is defined as the ratio of neutron flux density at the - location of the specimens in a surveillance capsule to the neutron flux density of the = inside surface at the pealt fluence locanon (Ref. 23). For capsules-W-104 and W-284 l the lead factor was revised from 1.5 to 0.81 (Ref. 3) and for the remaining capsules j (W-83, W-97,' W-263 and W-277) the lead factor was revised from L5 to 1.26 in. j Reference 3. j A revised eharhile was developed using the lead factors provided by Reference 3 and 4 the guidance of ASTM E185-82 in accordance with current 10 CFR 50, Appendix H ~ requirements. Factors external'to the ASTM E185-82 procedure that were also - considered included: .!y 1. Coordination with the generation of P-T limits and LTOP evaluation beyond 20 EFPY. - If additional surveillance capeale information is to be used to. j support the generation of P-T limits and an LTOP evaluation beyond 20 EFPY, the next capsule withdrawal must allow for enough time to analyze the' encapsulated materials as well as develop new P-T limits and LTOP. requiremonas prior to 20 EFPY. j 2. Potential fbr use of Position 2 of Regulatory Guide 1.99, Rev. 2 - Surveillance.. capsule data may be used in conjunction with Position 2 of R.G.1.99, Rev. 2 to predict mean shift in reference temperamre (ARTm) and decrease in upper q i shelf energy (USE) once credible surveillmace data is~ obtained. - One requirement for credibility is that, "the surveillance data for the correlation-1 11 + 3' ) - C-MECH-EJt 021 itn. 00 : .a . =.. -. -.. .--...:~.-..

JP' + monitor matenal in the capsule should fall within the scaner band of the data I base for that material" (Reference 8).. j Capsule W-97 did not contain correlation material (Ref. 21), so the next! a capsule withdrawn must contain correlation matenal in order lto allow for the ; use of Position 2 of R.G.1.99, Rev. 2. The two capsules that contain correlation material are W-104 and W-263 (Ref. 21)., J 3. The reactor coolant cold leg temperature for Waterford 3 has been reduced by l g 8'F from 553*F to 545'F (Ref.13). The effect, if any, of this temperature. j reduction on the reactor vessel beltline materials must be monitored. The new operating condition was evaluated, and it was determined thai the -~ l 1 requirements of 10 CFR 50, AWM G and 10 CFR 50.61 are not affected by the temperamre reduction of the cold leg (Ref.13). However, variations in ; the adjusted reference tempernaus (ART) and upper shelf energy (USE) of the i survedlance matenal froen predicted decreases must be monbored to' verify the =; validity of the previous studies. This evaluation should be made at the time of - 1 the second surveillance capsule withdrawal, and madincariaan to the' shift in " ,j AR,T and USE predictions can be :nade if necessary. The timing of the second l capsule withdrawal should be such that signiScant variations from predictions'- can be detected early enough to ensme that the P-T limits based on the ~ ART ] predictions remam conservative. j 4. The survedlance capsule withdrawal schedule should be managed with j consideration given to plant license renewal. Enough capsules must be ' tested j to assme con 6dence in beltline material properties,- but capsules should also be.- ,{ conserved to allow for future testing beyond the current design. lifetime. l . 'Ibe guidelines provided in Section 7 "Irradiarina Requisernents" and Subsection 7.6 n j " Number of Surveillance Capsules and Withdrawal Schedule" of ASTM E185-82' ] (Ref. 23) are currently required by 10 CFR 50, Appendix H for establishing the j surveillance capsule withdrawal schedule. A' review ~of the proposed revised standards (Ref. 24) showed no A affecting the marhad for doestmining the withdrawal ~ i schedule. Therefore, future modifications to 10 CFR 50 Appendix H by reference to -j this revised ASTM E185-93 senadard are not eW to alter the capsule withdrawal ~ l 8j A38 Con 6amien Enginamng Macinar Servica . C-MEG ER421. Rn. 00 '25 -,. I

.g ; .. s requirements. According to' ASME E185-82 (Ref. 23), the peak vessel inside fluence at EOL andi the conesponding transition temperature shift must be estimated to determine the" number of capsules required for removal. Waterford 3 has a peak EOL fluence of.. 3.69 x 10 n/cm (Ref. 3) and'a 1/4t fluence of 2.20 x 102' n/cm (using equation 3 2 2 of R.G.1.99, Rev. 2 to attenuate fluence to the'1/4t location). Based on the calculations of RTm (Table 6), the' largest shift in reference temperature - (ARTm) at EOL is'59.4*F (note that the method in 10 CFR 50.61 for calculating - .iRTm and the R.G.1.99 Rev.' 2 method for e@d=*ian ARTm produce equivalent: results). Using ASTM E185-82, it was determined that 3 c=a+ must be withdrawnl in the following order. First Capsule: (Removed and tested)- Second Capsule: At 15 EFPY or at the time when the amarmlarad neutron fluence of the capsule ws@ to the approximage EOL fluence at the reactor vessel inner ' wall location, whichever comes first. Third (Final) Capsule: At EOL but not less than once or greater than twice. the peak FOL vessel fluence. This may be modified on the basis of previous tests. This capsule may be held without testing following withdrawal. The second capsule to be withdrawn could be either W-104 or.W-263 to'obtain credible surveillance data. However,. capsule W-104 has a low lead factor (0.81), whereas capsule.W-263 has a high lead factor (1.26). It is prefened to withdraw-capsule W-263 for the second capsule, as the capsule fluence would be greater'than ' the peak surface fluence teceived by the vessel. Capsule W-263 would be' expected to; receive a fluence equivalent to the EOL fluence at the reactor vessel inner wall at - 25.4 EFPY (32 EFPY/1.26). 1 Q [ C. MECH-Dt421. Jta. 00 _26

- ~ t l Given the criteria for withdrawal of the second capak (above), c=P3M W-263 j should be withdrawn at 15 EFPY. The eng!M fluence corresponding to 15 EFPY was estimated to be' 2.18 x 108 n/cm using the lead factor of 1.26 and linear j 2 2 interpolation of the EOL vessel fluence of 3.69 x 10" n/cm _ given in Reference 3. Modifying the withdrawal' schedule to meet the current edition of ASTM standards calls for the last capsule to be removed between 25.4'and 50.8 EFPY. It is' suggested ~ that capsule withdrawal occur no later than 32 EFPY tw== this time ' corresponds to : 2 the plant EOL. This'will correspond to a capak fluence of 4.65 x 10" n/cm, q Given the requirements of 10 CFR 50 Appendix H and ASTM E185-82 along 'with - i the plant-specific considerations for Waterford 3. Table 8 presents the recommended : f ~ -schedule for the Waterford 3 reactor vessel surveillance capsule removal program: This schedule meets ASTM E185-82 requirements for capsule withdrawal (Ref. 23) as j currendy required by 10 CFR 50, AWiw H. It allows for detection of any ef!Isct i on ART,er or decrease in USE which could result frona the reduction in cold leg temperamre. This schedule should make available credibia surveillance data for l analyses following Position 2 ' f Regulatory Guide 1.99, Rsvision 2 (Ref. 8), and it L o provides for capsule withdrawal and testing prior to a P-T. limit modification q following 20 EFPY. This schedule will also allow for a mafficient munber of stindby l cayM (3) to be maintained for possible license renewal or to provide for other - j future Contingencies. t I A 1 a -l j i J l ABB Con 6mmon Enginernng Nucinar Sernces 27 d . C MECH-EA-021. Rev. 00 1

~. s q li t -7.0 RESULTS 1 Revised reactor vessel beltline P-T limits associated with twenty (20) EFPY have been- -l developed for heamp, cooldown, and inservice hydrostauc test. The reactor vessel P-h T limits are provided in tabular form in Tables 3. 4 and 5 and graphically in Figures j 1, 2 and 3. The beltline limits were considered in conjunction with the other limits discussed in Sections 2.5-2.9 to produce new Technical Specification Figures 3.4-2 3 and 3.4-3 for heatup and cooldown, respectively. These Technical Specification j figures meet the requirernents of 10 CFR 50, Appendix G and gau== of ASME Section III Appendix G, and they will continue to support ths Waterford 3 Technical d Specifications through 20 EFPY. j ~ --{ The beamp and cooldown rates, along with the associated temperamre ranges remam the same as currently specified in the Technical Specifications (8 EFPY). The. J Technical Specification Figures for 20 EFPY provide a less restrictive controlling 1 pressure than the current P-T limits for 0-8 EFPY,'so the current SDCS relief valve ] setpoint and other adminitrative controls previously established for LTOP will j continue to provide ad-=== protocuon. 1 j An evaluation of the PTS screening criteria was performed in accordance with.10 I CFR 50.61. Values of RTm were calcidanad for all beltline materials at end of life-j and are shown in Table 6. The values of RTm for all beltline materials were farf y below the screemns criteria and therefore sneet the requirements of 10 CFR 50.61. l 1 The end of life USE was evaluated for all beltline maserials in'accordance with q References 1 and 8. 'Ibe decrease in USE resulting from irradiatina was calculated at - j the 1/4t location, and the results are shown in Table 6. All values for end of life j USE are well above 50 ft lbs and satisfy the requirements of 10 CFR 50, Appendix' l G. l i ~ ] A revised surveillance withdrawal schedule was developed based on the results of the-W-77 capoule evaluation report. This revised echarkde, Table 8, has been developed to meet utility goals and the requirements of 10 CFR 50, Appendix H.: j i ei .i i - ABB rmahnnm Enginnemag 1%ecient Serwear C-MECH Dt-021. Rev. 00 '. 28 .J

~ g. '} + + j i I .i I Table 1 RPV Beltline Materials' Initial - Copper Nickel Longinviinal j Product Material ID ~RT,er (*F) Consent Consent . USE(ft' Ibs). Plate M-1003-1-. -30 0.02 0.71-144-j Plate M-1003-2 -50 0.02 0.67L ,1491 Plate M-1003-3 -42 0.02 _. 0.70 138 Plate M-1004-1 -15 0.03. 0.62 163 ya Plate M-1004-2 22 0.03 0.58-1441 ] Plate M-1004 3 ' O.03. 0.62-145 j >i Weld 101-124 A- -60 0.02 L 0.96' 106!" 1 -l Weld - 101-124 B,C ' 60 . 0.02, 0.96 ~ 1315 1 l Weld' 101-142 A,B,C -80' 'O.03 <0.20 .1295. j ~ 'i Weld - 101-171 -70 ~ 0.05- ' O.16L 1665 'I Weld (repair) 101-142 B. -40'* 0.03" 0.978-110" (a) Values were eatentaead using infonnation frees weld Certified Matenal Test Repons and Weld I@_. Fonns. -l ABC 1%"= Enguanenbeg Nucinar Services ~

~ S.. Tame 2 l ~; [ ~ ART Values for Beltline Mascrials at 20 EFPY .i [ Product Matl. ID Init. ARTm ARTm ei e, a, M I/4 M 3/44 ART ART. RTm (1/44) (3/44) (1/44) (3/44) 1/4: 3/44 Plate ' M-1003 -30 21.7 -16.0 0 10.9 8.0 - 21.7 16.0 . I3.4 - . 2.0 t.l Plase M-1003-2 -50 21.7 16.0 0 -10.9 8.0 21.7 16.0 -6.6 - l 8.0 - l y- - 42 ' 21.7 16.0 0 10.9 ' 8.0 21.7 16.0 ' l.4. -10.0 Place M-1003-3 g. Place M-1004-1 -15 21.7-16.0 0 10.9 .8.0 21.7-16.0 28.4 17.0 Plate M-1004-2 22-21.7 16.0 0 10.9 8.0 21.7 16.0 65.4 54.0 Plate M-1004-3 -10 21.7 16.0 0-10.9 8.0 21.7. 16.0-33.4 22.0' Weki 101-124 '60 29.3 21.5-0 14.7 10.8 29.3 21.5 -1.4 -17.0 A,B,C Weld 101-142- -80 38.0-27.9.' O 19.0 .14.0 38.0 27.9 .-4.0 -24.2 l A,B,C Weld 101-171 -70 48.2 35.4

O'

~ 24.1

17.7

48.2

'35.4 26.4 . 0.8 1 Weld 101-142 -40 44.5- .32.7 0-22.3. 'I6.4 44.5 -32.7 -49.0 25.4 o ~ 9 (repair). B-i . Es+,ws..4 es '*,m.,. ea-ww+.n .ses. m ..m-..m- ________..__-_m_s._u-_._,__.__ u m_m<__-...--~_..__._.,m.s<a .n W

g g-Table' 3-Beltline P-T Limits Heatup, 20 EFPY (psia). RCS Temp RCS Temp RCS Temp RCS Temp - (T) Isothermal (T) 30TMr (T) 50T Mr '(Y),601/hr 72 572.0 72 572.0 72 572.0 ' 72 - 572.0 75.6 562.0 75.6 562.0 75.6. 582.0 75.6 ~ 582.0 85.6 610.8 85.6 610.8 85.6 610.8 85.6' 610.8' 95.6 644.1 96.6 618.4 96.6 598.9 96.5-594.5 106.6 682.6 105.6 632.7 106.6 594.0 106.6 563.5 115.6 727.1 115.6 659.8 115.6 601.4 115.6 563.7 ' 120.0 749.9

• 125.6 692.6 125.6

'619.9 126.6. -594.5 ' 120.1 664.0

• 136.6 746.8 136.6 648.7 136.6 ~

615.1-125.6 692.6 - 136.1 749.9

• 146.6 687.6 146.6 :

646.5: 135.6 752.1 136.2 664.0

• 158.6 738A 158.6. : 686.5 145.6 620.9 146.6 719.9 158.0 740A
• 166.6 736.1-155.6 900.4 156.6 789.7

'156.1 664.0 *- 168.0 -- 749A * - 166.6 992.4 166.6 671.9 168A-711.1 168.1 . 664.0 * - 175.6 1098.6 175.6-967.3 175.6-783.2 175.6-711.6 L 185.6 1221.5 185.6 1078.5 186.6 888.6 188.6' 785.7 - 196.6 1383.5 196.6 1206.9 .196.6 906A 196.6 873.0-199.9 1271.0

• 200.0 10 3.3
• 206.6 1527.7 206.6 1366.0 '

2 6.6 1006A. 206.6._976.1 215.6-1717.5 215.6 1528.0 ' 216.6-1222.2 215.6 1096.2 225.6 -1937.0 225.6 1727.6 226A ' ' 1300.8 225.6.1236.8 228.0 1424.6

• 236.6 2190.7 236.6 1967.6 236.6 1864.2 236.6 1399.7 245.6 2463.9 246.6 2224.5 246.6 1777.0 -

246.6 1589.5 - 246.2 2500.0

• 266.6 2523.5 258.6 2 23.4-

-256.6 :1808.8 -

> 206A 2308A 286.6 2063.8 286.6 266.6 271.4 2500.0

• 275.6 2838.5 275.6 2358.0 275.6 275.6 286A

~ 285.6 2699.7 286.6 286.6

• - reerpointedvekse ABB Combusnon Eaguseernng Nuclear Servica.

31 C-MECH-ER-021. Rev.' 00

s r j Table 4 i .q q Beltline P-T Limits. y Cooldownc 20 EFPY (psia) j (T) Isothermal -(T) RCSTemp ' . RCS Temp. l RCS Temp RCS Temp.10TMr (T) 30TMr. '(T) 1'00TMr j 65.6 526.0 65.6 465.7-65.6 '280.5 65.6' 72.0 572.0 72.0 ' 542.9 ? 72.0 '484.9

• 72.0 310.1 75.6 582.0 75.6 552.4-75.6 406.7 175.6 ~

326.7 85.6 610.8 85.6 - 583.2 85.6 530.4 85.6 380.0: ? 95.6 ~ 644.1 ' 95.6 618.5 95.6 '570.4 95.6 441.7. j 115.6 727.1 115.6 706.8 115.6 670.3 - 115.6. ~595.'3 ~

}

105.6 682.6' 106.6' 669.6 105.6. 616.9 -105.6' _513.0 < -120.0 749.9

• 123.6 749.9
• 125A 802A

'125.6 690.6 > 0 120.1 664.0

• 123.7 664.0
• 128.2 ' 740A * -

131.1. 749.9-Ii 125.6 692.6 125.6 675.7 ' 123.3 684A * - 131.2' 664.0: 1J 134.9 734.4

• 135.0 713.4
• 135.6 752.1 136.6 ;

738.8i ' 135A.. 717A 135.6 : 714.6-145.6 82.9 146.6 812.2 146.6 . 800A i 146.6' 82.9) d 155.6 900.4 155.6 896.6-156.6 ' 808.2 - 155.6' 900.4 j 165.6 992.4-166.6. 992.4 166.6 903.4 165.6-992.4 175.6 1006.6- '175.6 1000.6 175A 100SA

175.6 1096.6' 185.6 1221.5 185.6-1221.5 188A 1221.5

-185.6 : 1221.5 - 196.6 1383.5 196.6 -1363.5 195A ~13853 195.6 1363.5 200A 14 5.7

• 200.1. 1437.4
• 206.6.

1527.7- .206.6 1527.7 ESA 1827.7' 206.6: s1527.7: 215.6 1717.5 215.6 1717.5 ~ 218A 1717.5 ~ 215.6 1717.5 - 225.6-1937.0 225.6 1937.0 25A 1937.0-225.6. 1937.0 236.6 2190.7 235.6 ' 2190.7 235A - 2100.7 - 235.6 2190.7-246.6 2483.9 246.6 2483.9 248A. 3403A 246.62483.9-246.2 2500.0

• 246.2 : 2500.0
• 246.2 ^. 2500.0
• 246.2 2500.0
• - interpolmatlvalue

>^ i 3.- fj ' n .m cc m cca n. ca.

Table 5. I Beltline P-T Limits Hydrostatic Test. 20 EFPY RCS Temp Pall ( F) 72 710.2 75.6 723.3 .85.6 761.7 95.6 806.1 105.6 857.5 115.6 916.8 125.6 985.5 135.6 1064.8 145.6 1156.5 155.6 1262.6 165.6 1385.1 175.6 1526.8 185.6 1690.7 195.6-1880.0-205.6 2099.0 215.6 2352.1 219.8 2475

• 220.7 2500
• 225.6 2644.6
• - interpolated value i

ABB Combaatum Engineermt Nuclear Services C-MECH ER-021, Rev. 00 .33

-Talde 6 RTm and End of Life USE Evaluations i Product Matl. ID - Initial .iRTm RTm . Initial USE End of Life j RTm (*F) (*F) Transverse-USE (*F) . (ft Ibs) (ft Ibs)' l Plate M-1003-1 -30 26.8 30.8-93.6 72.3

e Plate M-1003-2

-50 26.8 -10.8- %.9 74.8. ~ Plate M-1003-3 -42 26.8 18.8 89.7 . 69.2 - 4 i Plate M-1004-1 -15 26.8 45.8 106.0 ^ 81.8 3 Plate M-1004-2 22. 26.8 82.8 93.6 72.3 Plate M-1004-3 -10 26.8 50.8 94.3 '.72.8 I Weld 101-124A -60 36.1 .32.1 106; 81.8 i Weld - 101-124B C -60 36.1-32.1 131-101.1 Weld 101-142A,B.C -80 46.8 22.8-129 99.6 - Weld 101-171 -70 59.4' 45.4' 166.- 128.2D I i Weld 101-142B -40 '54.9 . 70.9 - 110 84.9 (repair) -l Asa (**=aa= EC:--0.g Wideer Serncer '34 C MECH B-021. Rev. 00

a

c.y Table 7. WSES-FSAR-Unit 3, Capsule Assembly Removal Schedule Capsule Capsule Azimuthal I. cad Removal Target Fluence 2 No. I.D. Loca. tion (deg.) Factor Time (EFPY) (n/cm ) 1 1 W-83. 83 1.5

Standby 2

W-97 9T 1.5 3.5 - 4.5 0.6 x 10 ' 3 W-104 104 1.5 10 - 12 1.6 x 101' i 6 W-284 284-1.5 16 - 20 2.5 x 10 l ) 4 W-263 263 1.5 Standby 5 W-277 277 1.5 Standby i t I r i f I ~~ i t . ABB Combasmon Espneermg Madner Meer C-MECH ER421 Rev. 00 35 .. _ ~

_ msg Table 8 Fwpcd Capsule Removal Schedule Meeting ASTM E185-82 Requirements Capsule Capsule Azimuthal Lead Removal Target Fluence-2 No. I.D. Location (deg.) Factor Time (EFPY) (n/cm ) 1 W-83 83 1.26 Standby l 2* W-97 97 1.26 4.44 6.47 x 10t' 3 W-104 104 0.81' Standby 3 4 W-263 263 1.26 15 2.18 x 10*' 5 W-277 277 1.26-25.4 3.69 x 105' to 10 50.8 7.38 x 10l' ' Recommentieri Recommended s 32 54.65 x 10 ' 6 W-284 284 0.81 Standby l

• Values represent actual data on removed capel a.

1 l a ? r b ABB Combaatwn Engineering Nucinar Servicat ' C-MECH ER-021, Jter. 00 - 36.

i

I FIGURE 1 i

.WATERFORD UNIT 3 APPENDIX G BELTLINE P-T LIMITS, HEATUP 2500 I f ISO l 30 F/HR [=j' =f ~ 5b'F/HR i 6 0* F/HR = d 2000 f 2 1 i3 . Q. a Eo i m $ 1500-g E 5 N ii2 3m E 60*F/HR E 1000 3gopfgg j f 30'F/HR / 9 SO - ./ s )k 500 ~ -d ~ ~0 -l 0 50 100 150 200 250 300 350 400 j INDICATED RCS TEMPERATURE ('F) O psia < Pres < 200 pais, oP = -186 psi. 200 pois s Prcs < 850 psia, AP = -100 psi ART B50 pois s Pres s 3000 pois, AP = -186 psi 1/41 = 65.4'F l' l AT = + 25.6*F 3/4t ' = 54.O'F l j - ABB Con 6msnan Enginemng Nudner Serwcas .r A

h. 4 F l-j FIGURE 2 WATERFORD UNIT 3 . APPENDIX G BELTLINE P-T. LIMITS, COOLDOWN 2500 1 b 2000 2 UI E. m mc m 1 1500 g e Q. mg E! o m m W g 1000-c - 10* F/HR-m Q I SO - U-5 500 / p / 100* F/HR 30*F/H R d i . i O ~ 400 i O 50 100. 150 200-250-300-350 INDICATED RCS TEMPERATURE ('F) j 0 psia < Prcs < 200 poia,aP = -186 psi g ART 200 pois s Pres < 850 pois, AP = -100 psi ,/4t = 65.4

• F B50 pois s Pres s 3000 peia, AP = -186 psi

' 3/4t = 54.O'F AT = + 25,6*F ) 1 . ABB Comiwtion Enginnersag Nucinar Serness a3 .e

r mum,, FIGURE 3 WATERFORD UNIT 3 APPENDIX G BELTLINE P-T LIMITS, HYDROSTATIC l 2500 2000 E ~ i3 .E w m. o u) @ 1500 en. e N ~ iE D w [-1000 0 N 0 / ~ O z_ 500 ? ~ O .l 0 50 100 150 200 250 300 '350 400 INDICATED RCS TEMPERATURE (*F) -l O psia < Pres < 200 pois, aP = -186 psi .l ART 200 pois s Pres < 850 peia, AP = -100 psi B50 psia s Pres s 3000 pois, AP = -186 psi 1/4t = 65.4

• F j,.

3/4r - 54.o*F i AT = + 25.6*p ABB Combumon Engineersag Nuclear Services .30

4*%% ( r 2500' ) r l' r : INSERYlCE l*.... ..... HYDROSTATIC.. t._............. ... _.....T8*' r i.. LTOP ALlONMENT. TEMP 201 9F... 3 SPECIFICATION 3.4.8.3) 2000 7 4 u) Q-3 c 50?F/HR HEATUP uI E LOWEST SERVICE 3.. TEMP.. 21('F.. .T CORE CRITICAL. u) m 1500 _,f '- g a. f- -E s y m y) m m

n. 1000 m

e '~' RCS TEMP. H/U RATE o

g...

< 200*F 30*F/1 HR w 2200*F TO $345'F - 50*F/1 HR g 4 . > 34S'F 60*F/1 HR o e, .._..:.c.r 2 ..{. 4.. .~:.-~:: ~ ;;...... ~;.i::. ~ {- 4 ..a.. '.~..:.:.:. w:::.-...ul9nuuu:::" ?::' ..noLiup r -= . TEMP 3r 72*F--- *: a. .,,,.,,..,..i a...,..,...,., ...,.,..,..c..,..,..,....., 0 100 200 300 400 500 INDICATED'RCS TEMPERATURE, Tc, *F FIGURE 3.4-2 WATERFORD UNIT 3 HEATUP CURVE REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS 18 2 O - 20 EFPY (PEAK SURFACE FLUENCE = 2.29 x 10 n/cm ) 3/4 4-30 _.as'< a a-Eng W w Senwar 4

o ck e 1:6, ^ .e. 2500 i INSERVICE i;. HYDROSTATIC TEST

... ~ LTOP AUGNMENT- -

TEMP., 207 "F O' 3(SPECIFICATION 3.4.8.3) g 100*FfHM COOLDOWN 2000 i i m c. i d E i LOWEST SERYlCE L l 3 i TEMP., 216*F m g 1500 ; A m w N cc 3 e m m w -~ ^

a. 1000 E

MCSTEMP.. C/D RATE o E;. y E < 135'F 10* F/1. HR : 4 2135'F TO s200*F - 30*F/1 HR 9 i '.. > 200*F 100*F/1 HR = O s z ~ 500 1. ~~- g MINIMURA .. SOLTUP --. TEhr.,. 7. 2

• F...

O~'''' "'''~ O 100 200 -300 400 500 INDICATED RCS TEMPERATURE, Tc, 'F '1 ~ FIGURE 3.4-3 a WATERFORD UNIT 3 COOLDOWN CURVE REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS - 18 2 0 - 20 EFPY (PEAK SURFACE FLUENCE = -2.29 x 10 n/cm 3 1 3/4 4-31 i ABB Combustwn Engineering Nuclear Services

1. I.

fB4*TYdDGWrYLJ%G

A ,e 4 - 1 ')

8.0 REFERENCES

1. Code of Federal Regulations,10 CFR Pan 50,' Appendix G,' " Fracture i Toughness Requirements", dated August 31,1992. j 2. Code of Federal Regulations,10 CFR Part 50, Appendix H,'" Reactor Vessell Material Surveillance Program Requirements," dated April 30,1993; 3. Report No. BAW-2177, " Analysis of Capsule.W-97", B&W Nuclear Services : Company, dated November,1992. J 4. R. Burski to T. Murley, "Waterford 3 SES, Docket No. 50-382, License No j NPF-38,10 CFR 50 Appendix H.M.A - Reactor Vessel Material Surveillance Program Requirements'- Report of Test Results," dated November 25,1992.' 5. Waterford 3 Technical Specifications, Amendment No. 84. Section 3/4.4.8, l Reactor Coolant System Pressure Temperamre Limira. 1 6. Code of Federal Regulations,10 CFR Part 50, Ag= u A, " General Design - j Criteria for Nuclear Power Plants", dated January 1988. H J 7. ASME Boiler and Pressure Vessel Code Section M Appendix G, ? Protection-Against Nonductile Failure",1989 Edition. ' l q

i 8.

Regulatory Guide 1.99, "Radiarian Embriertement of Reactor. Vessel ~ Materials", U.S. Nuclear Regulatory _ Comminaian Revision 2, May 1988. i 9. Instruction Manual, Reactor Vessel Assembly, Waterford W. Louisiana Power and Ught, C.E. Book No. 74170, Vol. I, dated April,1977. 2 i i 10. ASME Boiler and Pressure Vessel Code,'Section W.)pir I,:" Design:1 q Stress Intensity Values, Allowable Stresses, Material Properties,' and Design. q s Fatigue Curves",1989 Edition. j 11. FnM==*ar=1= of Fant and Mm== Tr=n<fer, Incropera and DeWitt,2nd Edition. Copyright 1985. a 42h

ABB Conabannon Enginnemag Nandaar Semas MJter. 00

s } g 'I 1 6 .e" ..I' i 12.~

Revised FSAR Table 5.3-13, License Document Change Request Form,' LDCR I j

' No. 93-0001, dated August 21,1992. j ~ 13. ' ABB/CE Report No. C-MECH-ER-014 Rev. 00, "An Assessment of the ' d Waterford 3 Reduction in Operating Temperature on NSSS Structural- ] Integrity," dated September 1993. .y

j 14.

R. O'Quinn to C. Stewart, " Design Input and Assumptions, Entergy Contract, j ~ No. W-1068-0505 ESR No. C-93-003, Revision of Pressure-Temperature j Limits". l t' e 15. " Semi-Elliptical Cracks in'a Cylinder Subjected to Stress Gradients", J.- Heliot,. ] R.C. Labbens and Pellisser - Tanon ASTM Special Technical Publication 677... j August 1979. 1 1 16. ASME Boiler and Pressure Vessel Code, Section XI,." Rules for Inservice ; Inspection of Nuclear Power Plant Components," 1989 Edition. 1 1 l 17. ASME Boiler and Pressure Vessel Code, Section III. " Rules for Construction f of Nuclear Power Plant Components," 1989 Edition. d ~1 'i 18. U.S. NRC Standard Review Plan 5.2.2, " Overpressure Protection," Revision 2, dated November 1988. 1 i 19. ASME Boiler and Pressure Vessel Code Case N-514,." Low l Temperature / Overpressure Protection, Section XI, Division 1", February 12, 1992.

j

.21 20. 10 CFR 50.61, "Fracane Toughness Requiremanen for Prosecuon Against'- ( l Pressurized Thermal Shock Events" Augut 31',~.1992.L i 21. Report No. C-NLM-003, Rev.1,." Program for Irradiation Surveillance of j Waterford Unit ihree Reactor Vessel Materials," Camhneian Engineering ] Inc., dated October 30,1974. - n ' 22. "Waterford 3 SES Updated Final Safety Analysis Report," Docket No. 50-382 - Operating License NPF-38, Controlled Copy No. 222. l ABB Coahnsnon'l-l ::4 Nudear Serwca 1 ! C-MECN4R421; Rev. 00: M l

-r s 23. ASTM Designation E 185 -82, " Standard Practice for Conducting Surveillance' l Tests for Light Water Cooled Nuclear Power Reactor Vessels," Annual Book-of ASTM Standards, Vol.12.02, American Society for Testing and Materials. Philadelphia. PA. 24. ASTM Designation E 185 (Revision approved in June 1993). " Standard Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels," Annual Book of ASTM Standards. Vol.12.02,- American Society for Testing and Materials, Philadelphia, PA. ' i i 5 9 h t 1 ? i v 48 Combusn'on Engbanermg Nucinar Services C-MECH-B421, Rev. 00

APPENDIX A Discussion of the Development of Technical Specification Figures 3.4-2 and 3.4-3. - ABB Combusnon Engineering Nuclear Services. C-MECH-ER-021. Rev. 00 41'

-s .y e TECHNICAL SPECIFICATION FIGURE DEVFT OPMENT Figures Al and A2 provide a graphical' representation of all the limits which are required in - the development of the RCS P-T limits. These were developed with heamp and cooldown'- rates and associated temperature ranges which were consistent with the current Technical [ Specifications (8 EFPY). These rates and temperature ranges are summarized below; Heatup RCS Cold 12e Temnernaire HU Rate .a < 200* F s 30*F/hr 1 200*F to 345'F s50*F/hr > 345'F s60*F/hr Cooldown RCS Cold Leg Temperamre HU Rate > 200*F s 100*F/hr. 200*F tc 135'F s 30*F/hr ' < 135'F s 10*F/hr-The most limiting pressures were then determined for each temperature. range for inclusion in. the Technical Specification figures. A GWis provided below for beamp and - cooldown. Note: No consideration was given to Low Temperamre Overpressure Protection (LTOP).- However, it can be noted that the 8 EFPY Technical Specification' figures provided :

a lower controuing pressure which should have been the basis for the LTOP seapoint, heatup and cooldown raess (and associated temperature ' iervals) and LTOP adminimerative..

m procedures. C- =M, the existmg LTOP basis should be ' adequate. Hannin To develop the heatup Technical Specification figure, the most controlling pressure for a given temperature is chosen. Figure Al provides a graphical depiction of the required limits.- l I AAS Cantessnam Engineerung Nuciasr Serncer 'C MECH BR-021,~ Rev. 00; _ l2

~ ~, < e. i e a e ~ a$ Between the minimum boltup temperature (72*F) and 200'F. the allowable heatup rate has - j r 1been established as less than or equal to 30*F/hr. Figure Al shows the beltline limit- + associated with this temperature range. As depicted, the minimum pressure (525 psia) is j more restrictive than the beltline limit. Consequently, the mmimum pressure requirement is l shown on the Technical Specification figure within this temperamre range.- .l ~ tq ~ A heatup rate of 50*F/hr is permitted from 200*F to 345'F. Figure Al shows the' beltline l P-T limit associated with 50*F/hr. Over the specified temperamre range, the reactor. vessel j flange limit associated with a 50*F/hr is also depicted. Review of Figure' Al shows that; 1 below the lowest service temperature (215.6*F) the controlling pressure is associated withH the minimum pressure requirements (525 psia). At the lowest service temperature (215.6*F); y pressures are permitted to exceed 20% of preservice hydrostatic test. However, design pressure (2500 psia) cannot be obtained until 271.4*F because~of briule fracture requirement-associated with the reactor vessel flange region and the reactor vessel beltline _ region.'. The ~ j reactor vessel flange provides the controlling pressure for temperamres above 215.6'F and. l up to 227.4*F. The beltline then provides the controlling pressure up to 271.4'F. .l At temperamres greater than 345'F, a heamp rate of 60'F/hr is allowed although normal operation pressures are permitted. y i .!1 For convenience, the requirements for inservice hydrostatic test are provided.on the heatup - l figure. 10 CFR 50 Appendix G requires a temperamre of at least 135.6*F with respect to j the vessel flange region.' However, ASME Code requires a temperamre of at least 215.6*F - (the lowest service temperature) prior to exceeding the o'.'". " pressure (525 psia). In ~ addition, the beltline controls the pressure until 220.7'F where's pressure of 2500 psia-is. permitted. The purpose of this curve is solely to establish the permissible. temperature for-l inservice hydrostatic tests in accordance with' ASME Code Section XI requirements.~. This ) terw.mre has been conservatively _ assessed to be 219.8'F. Below this temperature the ncemal beamp or cooldown P-T limits apply. The preceding will provide the appropriate P-T limit for nonnal operation and inservice ; [. hydrostatic test. 'y i D Utilizing the limit developed for heatup (it is more restrictive than'cooldown), core critical limits are established per the requirements of 10 CFR 50 Appendix G. This is depicted in 48 Combustion Enguuering Nuclear Services : W - C MECH ER-021, Rev. 00

~ Figure A1., However, for graphical illustration the curve has been conservatively modified as indicated. Cooldown 'l D Figure A2 depicts the limits associated with cooldown. Selection of the controlling pressure provides the curve presented in the Technical Specification figure. 'i a As with the heatup,' the minimum pressure (525 psia) is controlling with respect to the. i beltline up to the lowest service temperature (215.6'F). At temperamres greater than the i lowest ' service temperature, the beltline P-T limit associated with 100*F/hr cooldown provides the controlling pressure. Inservice hydrostatic test limits are also provided and 'the discussion presented in the heatup. section is still valid. i i 1 .1 -1 1 a 'I $I i i 'l l .1 i .1

ABB Combumon Enginernng Nuclear Servien

^#' C MECH-ER421. Rev. 00

gg _,., 3 K 1 'y j FIGURE A1 WATERFORD UNIT 3 RCS P-T LIMITS, HEATUP l 2500 -l 't ) INSERVICE ~ HYDROSTATIC l TEST 50* M R a I 2000 LOWEST = 4 SERVICE [ TEMP., 215.6

• F l

~ r E l CORE-i CRITICAL i ] $ 1500 l 7 / cz: Q. f 12: y e a 3 FLANGE y UMIT w [ 1000 O W ll l Z 30

• f /HR f

i l [ CORE CfIITICAL e -MINIMUM P4 ESSURE

• PER 10 CFR 50 4

APPEND XG I MINIMUM '~ RECOMMENDED ?I l 90LTUP ~ FOR TIECH. SPEC. FIGUR{ 72'F O 0 50 100 150 200 250 300 350 INDICATED RCS TEMPERATURE (*F) I O psia < Pres < 200 psia, AP = -186 psi AR 200 psia s Pres < 850 psia, AP = -100 psi g4, 'g p B50 pois s Prcs s 3000 pois, AP = -186 psi 3/4t = 54.O'F - AT = + 25.6*F l ABB Combustwn EnM Nucinar Services

c. i. .~9 - 4 FIGURE A2 WATERFORD UNIT 3 RCS P-T LIMITS, COOLDOWN 2500 t INSERVICE ,f' HYDROSTATIC TEST /' 2000 f LOWEST j SERVICE 100* F/HR TEMP., 21 5.6

• F cn E

~ 0 tn l @ 1500 g cc c. CI: y E D / u) 30*T/HF; @ 1000 10*F/HR - / b / z ~ MINil AUM PRESSURE ~ '*-WINIMUM BOLTUP 7;t*F 0 O 50 100 150 200 250 300-350-INDICATED RCS TEMPERATURE (*F) O psia < Pres < 200 psia, AP = -186 psi' 200 psia K Pres < 850 psia, AP = -100 psi ART i 850 psia s Pres g 3000 psia, AP = 186 psi 1/4t = 65.4 *F j aT = + 25.6*F 3/4t = 54.0*F . ABB Combunion Engineenng'Nudaar Services

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o.: a t u. -t .i r I NPF-38-148 v ATTACHMENT D 1 Final Letter Report.' Review of Reactor Vessel Beltline Pressure-Temperature Limits..for 0 - 8 EFPY' 8 4 h ..] 'S j-l 1 y =

October 11,1993 C-MECH-93-072 Mr. R.C. O'Quinn Entergy Operations, Inc. Waterford Steam Electric Station Unit 3 P.O. Box B Killona, LA 70066-0751

Subject:

Review of Reactor Vessel Beltline Pressure-Temperature Limits for 0-8 EFPY

Dear Mr. O'Quinn:

This letter provides a summary of the results pertaining to the assessment of the reactor vessel beltline Pressure-Temperature (P-T) limits associated with the Waterford 3 Reactor Coolant System (RCS) currently documented in Reference 1. The effort was performed under Entergy Contract No. W-1068-0505, ESR C-93-003 and was prompted due to an identified non-conservatism regarding the adjusted reference temperature for the limiting beltline material as documented in Reference 2. The effort was performed in accordance with the CE Nuclear Services Quality Assurance Mmmi for Quality Class I work. The contents of this repor:. have been independently reviewed to insure the accuracy of its contents.

Background

Tne RCS P-T limits which constitute both the operating and licensing basis of the Waterford Unit 3 plant through eight (8) effective full power years (EFPY) operation are intended to provide protection against brittle fracture for the ferritic pressure boundary components. These limits are based on the materials reference nil ductility transition temperature (RTym). With respect to the beltline, the RTmyr is adjusted to account for the loss in ductility experienced by these materials over time as a result of neutron radiation. It is the adjusted reference temperature (ART) that is used in calculating the beltline P-T limits. ABB' Combustion Engineenng Nuclear Power 688 *9 0 {e

43 Correwston Eag+ee40W F

Mr. Robert O'Quinn . C-MECH-93-U72 ' October 11,1993 - Page.2 of 6 The method for calculating the adjusted reference temperature currently acceptable to the - NRC is provided by Regulatory Guide 1.99 Revision 2 (Reference 3). However, this Regulatory Guide was not the standard when the 8 EFPY RCS P-T limits.were developed. - Application of Regulatory Guide 1.99 Revision 2 provides a higher ART corresponding to the outer diameter crack tip location (3/4t) using the design basis input parameters. Consequently, the ART's were recalculated for the design basis conditions and improved fracture mechanics methods were applied to show the conservatism inherent in beltline P-T. limits. Analysis Highlights The original basis associated with the beltline P-T limits were reviewed and the limiting material was shown to be Plate M-1004-2. The following information represents the ' design ; basis input used to predict the ART and was consistent with the current information (Reference 4) associated with this plate material. Plate M-1004-2 Initial RT,er =.22*F Copper Content = 0.03 w% Nickel Content = 0.58 w% q U As documented in the design, basis calculation, a peak surface fluence of 9.2x109 'n/cm 2 was utilized to predict the ART values at the 1/4t and 3/4t locations, 97'F and 30*F_ respectively. Strict application of Regulasory Guide 1.99 Revision 2 provides 1/4t and 3/4t ART values of 55.2*F and 44.6'P. 3 l Comparison of these vaines show the ART at the 3/4t location to be non-conservative.;The

signi&a** of this is that in developing the beltline P-T limits, the 3/4 location can provide.

g the controlling pressure during a bestup transient.- Recalculation of the beltline limits'using the same method would likely provide more restrictive beltline limits. Consequently, the. beltline limits were r** Mated using improved fracture mechanics methods to show that the basis beltline curves were conservative and in compliance with the requirements of = Ap=~ii G to 10 CFR 50 (Reference 5). t _--_--_1---_____- m

d W c Mr. Robett O'Quinn C-MECH-93-072 October 11,1993 Page 3 of 6 . Appendix G to 10 CFR 50 requires that the beltline P-T limits be calculated using the guidance of ASME Boiler and Pressure Vessel Code, Section III, Appendix G (Reference 6, referred to herein as ~ASME Code Appendix G). ABB/CE has developed an improved fracture mechanics methodology.which meets the requirements of Appendix G to 10 CFR 50 ' 3 and the ASME Code Appendix G guidance. This method employs influence coefficients and the principles of superposition to determine the stress intensity factors resulting from internal pressure. and thermal loads (14 and Kn, respectively).- A transient convection / conduction. heat transfer analysis.is performed utilizing a one-dimensional isoparametnc finite element - model. The heat transfer results are utilized to detennine the allowable material stress intensity factor, Km and the applied thermal stress intensity factor, Ka, as a function of time. l The allowable pressure stress intensity factor is determined and enhaary=wly the allowable j pressure. The results of this evaluation are allowable pressue versus temperature for'the analyzed thermal rate. ] To assure the limiting condition is achieved the isothermal condition is also compared to the analyzed rate conditions.. In this instance, the rates associated with the Waterford Unit 3 Technical Specification (Reference 1) were analyzed. To properly index the limits in tenns of indicated pressurizer pressure and indicatad cold leg te+resus, it is necessary to consider the elevation difference between the reactor vessel - and the pressurizer, the flow induced pressure losses between the reactor vessel 'mlet nozzle and the surge line nozzle, and include the effects of instnnnent uncertainty. The original design basis values were uritiend. J H The beltline P-T lindes were compared to the current Waterford Unit 3 Tachnical Specification heasup figure to assess whether the current beltline curves were adequate in providing the requisies protection fhun brittle fracaire. Esadta The ART values associated with the limiting reactor vessel beltline material (Plate M-1004-2) through 8 EFPY were calculated for the 1/4t and 3/4t locations in accordance with Regulatory Guide 1.99 Revision 2 as 55.2*F and 44.6*F respectively. The beltline P-T l] o .-.m. '3

'nppg Mr. Robert O'Quinn C-MECH-93-072 October 11, 1993 Page 4 of 6 j -limits were calculated utilizing the ART values for the limiting beltline material and other pertinent design basis information. "Ihrough the use of improved fracture mechanics methods the beltiine P-T limits were re-evaluated for Waterford Unit 3. The comparison of the l design basis beltline limits, as provided by the Waterford 3 Technical Specifications, to those calculated with the improved fracture mechanics techniques are provided in Figure 1. Conclusions Review of the limiting beltline material ART values at the 1/4t and 3/4t locations showed the

• j 3/4t value to be non-conservative when predicted using the gth provided by Regulatory l

Guide 1.99 Revision 2. While the reactor vessel beltline 3/4t location can provide the ij controlling pressure during heatup, re-evaluation of the beltline P-T limits using improved fracture mechanics techaie== showed the current basis provided by the Waterford Unit 3 i Technical Specifications to be more restrictive. Since the applied method used to re-evaluate the beltline meets the requirements of 10 CFR 50 A,Wir G, the reactor vessel beltline P-T - l limits provided by the Technical Specification have provided the requisite margins of safety _ in light of the norsonnervative ART. i If you should have any questions regarding this report, please contact me at (203)-285-2294 j i or Mr. Carl Gimbrone at (203)-285-2567. q y i Sincerely, COMBUSTION ENGINEERING, INC. fM Craig Stewart _j Project F9 --i j cc: G. Bundick C. Gimbrone - 1 4 i J

a o .g.g m,n w, ~ , 1.Q:g ' i + x; ~o - Mr. Robert O'Quinn . C-MECH-93 072 October 11,1993 '

Page 5 of 6 -

.J .i References j l o (1)' Waterford Unit 3 Technical Specifications,'Section 3/4.4.8, J ' Pressure / Temperature Limits. Amendment No. 84. 1 (2) Letter No. C-MECH-92-079, " Review of Waterford 3 Appendix G RCS n Pressure-Temperanut Limits," C.D. Stewart, datal Nov.'11,1992. (3) Regulatory. Guide 1.99 Revision 2, " Radiation Embrittlement of Reactor q Vessel Materials," dated May 1988. y (4) " Updated Final Safety Analysis Report," Waterfoed 3 SES, Docket No. 50 i j a 382, Operating License NPF-38, Controlled Copy No. 222 (As amended by LDCR No. 93-0001, J. B. Perez, dated August 21,1992). j (5) 10 CFR 50 Appendix G, " Fracture Toughness Requirements'," dated August ^ ~ 31,1992. 3 (6) ASME Boiler and Pressure Vessel Code, Section III,AWir G, " Protection. i Against Nooductde Failure," 1989 Edition. ,e -) ~ .,j l a-! t- ! ^ ) 1 1 f [q y l Y -t d -d v 4 y w 97'r

~. 4.' a ','Mr. Robert O'Quinn , C-MECH-93-072 October 11,1993 Page 6 of 6 FIGURE'1 Beltline Limits for 8 EFPY, Heamp 3000 il Ii l l l l lil,J.ded i 1.1 I ll I I !IIII' (TEV 1 l l l l1ii l l j

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