ML20247D487

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Proposed Tech Specs Re Revised pressure-temp Limit Curves
ML20247D487
Person / Time
Site: San Onofre Southern California Edison icon.png
Issue date: 07/19/1989
From:
SOUTHERN CALIFORNIA EDISON CO.
To:
Shared Package
ML13302A666 List:
References
NUDOCS 8907250213
Download: ML20247D487 (19)
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{{#Wiki_filter:- - -. - - _ - _ - - - - - - r-I l NPF-15-278 l ATTACHMENT "B" UNIT 3 PROPOSED SPECIFICATION. 8907250213 890719 ADOCK 05000362 PDR PDC p

m 4 L% LIMITING CON 0! TION FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE 'SECTION 3/4 4-3 HOT SHUTOOWN......... CO LD SHUT 00WN - LOO P S F I L L E0............................ 3/4 4-5 COLD SHUT 00WN - LOOPS NOT FILLE 0........................ 3/4 4-6 3/4.4.2 ' SAFETY. VALVES - 0PERATING............................... 3/4 4-7 1 3/4.4.3 PRESSURIZER............................................. 3/4 4-8 3/4.4.4 STEAM GENERATORS........................................ 3/4 4-s 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS............................- 3/4 4-17 OPERATIONAL LEAKAGE.................................. 3/4 4-18 3/4 4-21 3/4.4.6 CHEMISTRY............................................... 3/4.4.7 SPECIFIC ACTIVITY...................'....!.............. 3/4 4-24 3/4.4.8 PRESSURE / TEMPERATURE LIMITS 3/4 4-28 REACTOR COOLANT SYSTEM............................... 3/4 4-32 PRESSURIZER - HEATUP/C00 LOO OVERPRESSURE PROTECTION SY, M RCSTEMPERATURE<238*F.i.2Mif.;,..)................ 3/4 4-33 RCS TEMPERATURE I.3889 3CJLPF................. 3/4 4-35 3/4 4-36 3/4.4.9 STRUCTURAL INTEGRITY.................................... 3/4 4-37 3/4.4.10 REACTOR C0OLANT GAS VENT SYSTEM......................... 3/4.5 EMEME4CY CORE COOLING SYSTEMS 3/4 5-1 2/4.5.1 SAFETY INJECTION TANKS.................................. 3/4 5-3 ECCS SUBSYSTEMS - T,yg,350*F........................... 2 3/4.5.2 3/4 5-7 ECCS SUS $YSTEMS - T,yg < 350*F.......................... 3/4.5.3 3/4 5-8 .3/4.5.4 REFUELING WATER STORAGE TANK............................ _rw

o. ') 'b' REACTOR COOLANT SYSTEM NOT $NUTDOWN LIMITING CL* ! TION FOR OPERATION 3.4.1.3 At'least two of the' loop (s)/ train (s) listed below shall be OPERA 8LE and at least one Reactor Coolant and/or shutdown coolirg loops shall be in operation." Reactor Coolant Loop 1 and its associated steam generator and at

a..

least one assuciated Reactor Coolant pump ** b.- Reactor Coolant Loop 2 and its associated steam generator and at least one associated Reactor Coolant pump,"* c. Shutdown Cooling Train A, d. Shutdown Cooling Train 8. f APPLICA81LITY: MODE 4 ACTION: With'less than the above required Reactor Coolant loops and/or s. shutdown cooling trains OPERA 8LE, immediately initiate corrective action to return the required loops / trains to OPERABLE status as soon as possible; if the remaining OPERA 8LE loop is a shutdown cooling train, be in COLD SHUTDOWN within 24 hours. b. With no Reactor Coolant loop or shutdown cooling train in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and innsediately initiate corrective action to return the required coolant loop / train to' operation. All Reactor Coolant pumps and shutdown cooling pumps may be de-energized E for up to 1 hour provided (1) ne operations are permitted that would cause dilution of the Reactor Coolant Systes boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

==A Reactor Coolant pump shall not be started with one or more of the Reactor l' Coolant System cold leg temperatures less then or equal to ;-,"T CU M ;r::: ':;r.;;t:r d_- h %;; tM. OC0 ; tk f t

v. (2) the-secondary water temperature of each steam generator is kre than 100*F above l

eac of et e Coolant 5 stem cold leg temperatures. l p d b f S y ted i w Ta b S.4-3 w en gmAec j l'

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y REACTOR COOLANT SYSTEM . COLD' SWTDOW - LOOPS FILLED LIMITING COMITION FOR OPERATION ~ 3.4.1.4Il

  • a.

At least one of the following loop (s)/ trains' listed below shall be OPERA 8LE and in operation *: 1. Reactor Coolant Loop 1 and its associated steam. generator and at least one associated Reactor Coolant . Pump ** 2. Reactor Coolant Loop 2 and its associated steam generator and at least one associated Reactor Coolant Pump ** 3. Shutdown Cooling Train A 4. Shutdown Cooling Train 8 b. One additional Reactor Coola't Loop / shutdown cooling train n shall be OPERABLE, or The secondary side water level of each steam generator shall c. . be greater than 10% (wide range). t' APPLICABILITY: MODE 5, with Reactor Coolant loops filled. ACTION: With less than the above required shutdown. trains /. loops OPERA 8LE or a. with less than the required steam generator level, immediately initiate corrective action to return the required trains / loops to OPERABLE status or restore the required level as soon as possible. With no loop / train in operation, suspend all operations involving a b. reduction in baron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required loop / train to operation. "All reactor coelant pumps and shutdown cooling pumps may be de energized i for up to 1 hour provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System baron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

    • A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 2"S*" =h

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....
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e.- (21-the secondary water temperature of each steam generator is than

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,8 i ^ eac of the A ac r ool stem old leg temperatures, g WM2.4-pLm NOI[91984 l

o i REACTOR COOLANT SYSTEM b m 4 kih f 3/4.4.0 PRES 5URE/ TEMPERATURE LINITS I REACTOR COOLANT SYSTEM LIMITING CE;EITION FOR OPERATION 3.4.8.1 The Reactor Coolant System (except the pressurizer) temperature and I pressure shall be limited in accordance with the limit lines shown on Figure 3.4-2 and Figure.f.4-3 during heatup, cooldown, criticality, and inservice leak'and hydrostatic testing with: j A maximum heatup 'of 10*F -in any 1-hour period with RC cold leg. PU a. temperature less than 110*F. A maximum heatup of 30*F in any 1-hour / 3 period with RC cold leg temperature greater tnan 110*F but less than 0 330*F. A maximum heatup of 60*F in any 1-hour period with RC cold 4 leg temperature greater than 330*F. L A maximus cocidown of 10*F in any 1-hour period with RC cold leg ~I b. temperature less than 110*F. A maximum cooldown of 30*F in any i g 1-ho u period with RC cold leg temperature greater than 110*F but d less than 200*F. A maximus'cooldown of 100 F in any 1-hour period with RC cold leg temperature greater than 200*F. (lg - c. .A maximum temperature change of 10*F in any 1-hour period during inservice hydrostatic and leak., testing. operations above the heatup )s and cooldown limit curves. i ~ PLACABILITY: At all times. AP [)_fe - ACTION: ( d With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity j . -15 of the Reactor Coolant System; determine that the Reactor Coolant System W 4 remains acceptable for continued operations or be in at least HOT STAND 8Y {* and pressure to less than g t withintho'next6hoursandreducetheRCST*N11owing30 hours. y .e 200*F and 500 psia, respectively, within the ~$ J [l SURVEILLANCE REQUIREMENTS a> The Reactor Coolant System temperature and pressure shall be 10 4.4.8.1.1 b)P,j' determined to be within the limits at least once per 30 minutes during system 3 f neeemp, cooleewn, and inservice leek and nyorostatic testing operations. 4 1 ( i 4.4.S.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at I C.- 4 the intervals required by 10 CFR 50 Appendix H in accordance with the schedule in Table 4.4-5.. The results of these examinations shall be used to update c1 4 --g Figures 3.4-2 and 3.4-3. Recalculate the Adjusted Reference Temperature based f Q. 3 L ~ j on the greater of the following: >..b# The actual shift in reference temperature for plate C-6802-1 as pq a. op W determined by impact testing, or l @ The predicted shift in reference temperature for weld se s2-203A.]. b. 2+2038, or 2-203C as deteristned by Regulatory Guide 1.99, ^:;'e': 8=Matiaa " : ^ ? *,"", " O f f::^ - ** "-- * ^!9 C : M " ^^ "n M * -d n - ~ ":::^r Y;;;;? T 10-'9 " hm /bm NOV i 5 W ~ _ _

u h - ( INSERT A 3.4.8.1 With the reactor vessel head bolts tensioned the Reactor coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on figures 3.4-2, 3.4-3, 3.4-4, and 3.4-5 during heatup, cooldown, criticality,fand inservica' leak and hydrostatic testing with: A maximum heatup as specified by Figure 3.4-3 a. in any 1-hour. period with RCS cold leg. temperature less than 153

  • F.

A maximum heatup of 60* F in any 1-hour period with RCS cold leg temperature greater than 153*F. b. A maximum cooldown as specified by Figure 3.4-5 in.any 1-hour period with RCS cold leg temperature less than 126*F. A maximum cooldown of 100*F in any 1-hour period with RCS cold _ leg temperature greater than 126*F. 'A. maximum temperature change of 10*F in any c. 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown curves. d. A minimum temperature of 86*F to tension reactor vessel head bolts. With the reactor vessel head bolts detensioned, the Reactor Coolant System (except the pressurizer) temperature shall be limited to a maximum heatup or cooldown of 60*F in any 1-hour period.

I j i 1 I r2 c x j4 j = t a 5 k ,2 2 % aa f.4, s. 7 ~5 w w I J h a 2 5= SE o o o o o o ) 24 a a a a a a 2 o. W ad .a w i &2 5 lii E a5 m I d $2 E 2 5 E G R ~ ~ ~ g gga y 5-ku vue N M W o

) i l FHMIE 3.4-2 SONOS3RCS 6 uurAmousran4-ssrn ....i.... Hr m "ID (71GUE 3.4-3 ( 153*O r (,g /HR ) 153'O - e = n v i 8, y I I I 1 2500 ' l I i 1(X)O i J J J I f f F J J l I i 1 2 I / f f mg I1500 ,I ) j camcAL r y y I I / / / / / / f I / / 1000 J J r r 500 MINlWUW acLTur TINP-set -'IIIIII i g 0 50 100 150 200 250 300 350 400 o seCED RCS TEMPDIATURE( F) 1

I ek) Y ~ SONE 3 RDB PREMAK/fDRMMElASTE MAlltfUM ALLDelmE WArW MR5 (4-4 EFFT) 70 s 90 50 40 R., 30 l / 10 E 8 l a R l l. a e 3 a 80 80 100 110 120 130 140 150 180 170 180 BRICATED RDSTEMPERATURE N NOTE: A MA10WUW HEATUP RATE OF 80"F/H E ALIDWED AT AW1EMPERATURE ABOVE 1 ~~ ~ moc, a m,,

.r ---,--,-v-,-- w- _,-~_ 4 ' e F100E 3A-4 SONG 5 3 RCS PRE 55URE/fBRRATUK UMTATIONS FOR 4-5 EFPf E 5 h h E 5 E 5 N E E E 5 E C00LD0ml 7 (nouns s.4-s <in'r) o me-202r g,,,ej g,3, 3,) \\, ~ l I.500 i l 200') I ii n I 1 l1500 1 E A I f l / r i / 1000 1 mw 00LTUP l me-se/ 'IIIII 0 O 80 100 150 200 250 300 350 400 ISOCATED RCS TEMPERATURE ('F) .m ___m_ m____m.____________ __.-._.__mA...

i I k6 l N 1 P10UK 3.4-5 SONGS 3 RCS PIEBRIK/BRIMTUK LREIB { MAMMUM ALLONAAE 000LDOWN IWEB (4-8 EFFT) 110 100 90 m 1 70 [ ' :) 80 g 3 50 I.g 30 gg 10 g(. t i i i i i i 80 90 100 110 120 130 140 150 NENCMD RCS TEMPERATURE (*F) NO2 A MAXNUM 000LDOWN RATE OF 100'F/HR 5 AIMNfED AT ANY TEMPERATUK ABOVE 12S'F OM AMAM A amm m.m

New M t. N Table 3.4-3 1ow Temperature Rcs Overeratsurs Pretsetten Ranas Cold ten Temperature. *F Onoratina Period. EFPY_ During During Codidowrt gggggg 4 to 8 1 302 1 267 e e O SANONOFRE-UNIT 3 3/4 4-31b

4 REACTOR COOLANT SYSTEM' OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE I estest 3()'2 p 8 LIMITING CONDITION FOR OPERATION 3.4.8.3.1 At least one of the following overpressure protection systems shall be OPERA 8LE: The Shutdown Cooling System Relief Valve (PSV9349) with: a. 1) A lift satting of 406 2 10 psig*, and 2) Relief valve isolation valves 3HV9337, 3HV9339, 3HV9377, and-l 3HV9378 open, or, b. The Reactor Coolant System depressurized with an RCS vent of greater. than or equal to 5.6 square inches. APPLIC'4ILITY: MODE 4 when the temperature of any one RCS cold leg is less e or vessel head on. E r equal to 2WE91-E ~ ACTION: With the SDCS Relief Valve inoperable, reduce T,,, to less than l a. 200'F, depressurize and vent the RCS through a greater than or equal to 5.6 square inch vent within the next 8 hours. With onq or both SDCS Relief Yalve isolation valves in a single b. ' SOCS Relief Valve isolation valve pair (value pair 3HV9337 and 3HV9339 or valve pair 3HV9377 and 3HV9378) closed, open the closed valve (s) within 7 days or reduce T to less than 200'F, depres-surize and went the RCS through a $Nater than or equal to 5.6 inch vent within the next 8 hours. In the event either the SDCS Relief Valve or an RCS vent is used to c. mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 The report shall describe the circumstances initi-within 30 days. ating the transient, the effect of the SDCS Relief Valve or RCS vent on the transient and any corrective action necessary to prevent recurrence. The provisions of Specification 3.0.4 are not applicable. d. SURVEILLANCE REQUIREMENTS 4.4.8.3.1.1 The SOCS Relief Valve shall be demonstrated OPERABLE by: Verifying at least once per 72 hours when the SDCS Relief Valve is a. being used for overpressure protection that SDCS Relief Valve isolation valves 3NV9337, 3HV9339, 3HV9377, and 3HV9378 are open. "The lift setting pressure applicable to valve temperatures of less than or equal to 130*F. h------------_- m i 5 87-

' REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS -l RCS TEMPERATURE > M 302?F LIMITING CONDITION FOR OPERATION At least one of the following overpressure protection systems shall 3.4.8.3.2 be OPERA 8LE: The Shutdown Cooling System Relief Valve (PSV9349) with: a. 1) A lift setting of 406 2 10 psig*, and L 2) Relief valve isolation valves 3HV9337, 3HV9339, 3Hv9377, and 3HV9378 open, or, A minimum of one pressurizer code safety valve with a lift setting b. i of 2500 psia + 1%**. APPLICABILITY: MODE 4 with RCS temperature above gispA.A a Table 1.'t-3, L g. With no safety or relief valve OPERABLE, be in COLD SHUTDCVN and s. vent'the RCS through a greater than or equal to 5.6 square inch vant I within the next 8 hours, In the event the SDCS Relief Valve or an RCS vent is used to b. mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specifica-tion 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the SDCS Relief Valve code safety valve or RCS vent on the transient and any corrective action necessary to prevent recurrence. SLAVEILLANCE REQUIREMENTS 4.4.8.3.2.1 The SOCS Relief Valve shall be demonstrated OPERABLE by: Verifying at least once per 72 hours that the SDCS Relief Valve isolation valves 3HV9337, 3HV9339, 3HV9377 and 3HV9378 are open when a. the SOCS Relief Valve is being used for overpressure protection, Verifying relief volve setpoint at least once per 30 months when b. tested pursuant to Specification 4.0.5. The pressurizer code safety valve has no additional surveillance 4.4.8.3.2.2 requirements other than those required by Specification 4.0.5. The RCS vent shall be verified to be open at least once per 12 hours 4.4.8.3.2.3 when the vent is being used for overpressure protection, except when the vent pathway is provided with a vt.lve which is locked, sealed, or otherwise secured in the open position, then vurify these valves open at least once per 31 days. "The lift setting pressure applicable to valve temperatures of less than or equal to 130*F.

    • The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating toeparature and pressure.

- eAH N

g _ REACTOR Coot. ANT SYSTEM ~ 8ASif 3/4.4.1 REACTOR COOLANT LOOPS AND C00LANT' CIRCULATION The plant is designed to operate with both reactor coolant loops and j associated reactor coolant pumps (RCPs)in operation, and maintain DN8R greater l than 1.31 during all normal operations and anticipated transients. As a result, in MODES 1 and 2 with one reactor coolant loop not in operattun. this specification requires that the plant be in at least NOT STAN08Y within 1 hour since no safety analysis has been conducted for operation with iets than four reactor coolant pumps or less than two reactor coolant loops in operation. In MODE 3, o single reactor coolant loop provides sufficient heat removal-j capability for removing decay heat; however, sir.gle failure considerations j require that two loops be OPERABLE. j In MODE 4, and in MODE 5 with reactor coolant loops filled, a single ( reactor coolant loop or shutdown cooling train provides sufficient heat remova) j capability for removing decay heat; but single failure considerattora require that at least two loops / trains (either RCS or shutdown cooling) be OPERABt.E. In MODE 5 with reactor coolant loops not filled, 6. single shutdown cooling i j train provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam genera-f h tors as a heat removing component, require that at least two shutdown c9oling l 4 g trains be OPERABLE. g ) The operation of one reactor coolant pump or one shutdown cooling pump ( a J provides adequate flow to ensure mixing, prevent stratification end p ter Coolant System. The reactivity change rate associated with boron reductions l ] )) will, therefore, be within the capability of operator recognition and cont ';, p The restrictions on starting a reactor coo %t pump in MODES 4 arG t M.th % ) one or more RCS cold legs less than or equal to 266af are provided 'o temnt 8 y / which could exceed the limits of Appendix G to 10 CFR Part 50.{ RCS p The RCS will ( be protected against overpressure transients and will not exceed the limits of ( Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to cxpar.d into or (2) by (#to ) restricting starting of the RCPs to when the secondary water temperature of p ) each steam generator is less than 100*F above each of the RCS cold leg 't u) temperatures. 3/4.4.2 SAFETY YALVES The pressurizer code safety valves operate to prevent the RCS from being Each safety valve is designed pressurized above its Safety Limit of 2750 psia. to relieve 4.6 x 105 lbs per hour of saturated steam it the valve setpoint . The reitef capacity of a single safety valve is adequate plus 3% accumulation. to relieve any overpressure condition which could occur during shutdown with In the avsnt that no safety l RCS cold leg temperature greater than valves are OPERABLE and for RCS col og temperature less than or equal to 285'F, the operating shutdown co ling relief valve, connected to tra RCS, provides overpressure relief apabd lity and will prevent RCS bWPeb overpressurization. Spet.Wa1in (G*4 M 3 d m 0138(

1 1 Il REACTOR COOLANT System

    • MS SPECIFIC ACTIVITY (Continued)

Reducing T to less than 500*F prevents the release of activity should asteamgenerat$f8tuhe rupture since the saturation pressure of the primary ceslant is below the lift pressure of the atmospheric steam relief valves. The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Infomation obtained on todine spiking will be used tn assess the parameters associated with spiking phenomena. A l reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. 3/4.4.5 PRESSURE / TEMPERATURE LIMITS Ali components in the Reactor Coolant System are designed to withstand ' the effects.of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal Ioad transients, reactor trips, and startup and shutdowt1 operations. The various categories of load cycles used for design purposes are provided in Section 3.9.1.1 of the F5AR. Dueing startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistant with'the design assumptions and satisfy the stress limits for cyclic operation. .During heatup,-the thereal gre($ents in the reactor vessel well produce thermal stresses which pary rom c re tve tne inne wait ca7. ensile ac tne oute waft. se t 11y ad il e esjive s esses tend to.a eviate the ila st as*1 ed . nter pressur

  • Theref6r, a '

essur ature c va be d on a state nditions 'e., no.t rea ' stress ) resents lower und all inflar 'cuyGes fo, finite he up r ~ es who inne all o the v sel trea y the verning*I' atto. /cov de(outerwallofti

t. e det nation.

pre ure-t eratu hp an sis o vess beco s,the 1 tati for cas n whi control ng top ton. he Y gradie ' estab1 Vied d ing he up pr e tensi ' strep s at out weil of vessel. Jhese ressa re add ive to pre ed te i 4 stres which - aire pret t. Th thermally'p to 1 nduce stress the , r wall o the v elar/tensil and re depende on the e t of heat and the as al, g the stup r. ; the h tup ft)I t desc/ bed f in whic the re, a 'r d, urve s flar to , for case f th outgr in wall r.not ' defined. sequent ,wa'1 of ths vesse M echoes the.ftress c trolling" ocatiqn, each heatup rate 1 haats. af thtarast snuat ha analy,md an an indivf etum)

f u

hek 6 NOV i51981 e l k

e- ,h ---.1. nwi i ~ e are tenstie at the reactor vessel outside surface. Since reactor vessel internal pressure always produces tensite stresses at both the inside and outside surface locations, the total applied stress is greatest'at the outside surface location. However, since neutron irradiation damage is larger at the inside surface location when compared to I the outside surface, tha inside surface flaw may be more limiting. Consequently, for the heatup analysis both the inside and outside surface flaw locations must be analyzed for the specific pressure and thermal loadings to determine which is more limiting. During cooldown, the thermal gradients through the reactor vessel wall produce thermal stresses which are tensile at the reactor vessel inside surface and which are compressive at the reactor vessel outside surface. Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the inside surface location. Since the neutron irradiation damage is also greatest at the inside surface location the inside surface flaw is the limiting location. Consequently, only the in side surface flaw must be evalutted far the cooldown analysis, l l l l

4 REACTOR COOLANT SYSTEM SASH PRES $URE/ TEMPERATURE LIMITS (continued) The heetup and cooldown limit curves (Figures 3.4-2 and 3.4-3) are composite curves which were prepared by determining the most conservative case, with either the inside or outside well controlling, for any heatup rate of up to 60'F/hr or cooldown rate of up to 100*F/hr. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of ige period indicated on i ) Figures 3.4-2 and 1.4-3. Mcka The reactor vessel saterials have beenItaste to detemine their initial I 27

the results of these tests are shown in Table 8 3/4.4-1. Reactor
j. ' D ~T o,Mtion and resuitant fast neutron (E are iter than 1 Mov> irradiation wiii cause an increase in the RT Therefore,qan adjusted reference,emperature, t

( based upon the fluence and k p.er and ;h::;r;;; content of the material in l question, can be predicted using FSAR Table 5.2-5 and the recommendations of p l- ( f 7 3 gRegulatory Guide 1.99, Revision 1. !ff;;t: ;f ;;id=? Zix::t: =.. adi;t:0 [ S di ti;; E 7; te ";;;ter L ; eel ;M eriels." The heatup and cooldown limit I curves, Figures 3.4-2 and 3.4-3, include predicted adjustments for this shift 8 g9 .h -( L h-in RT at the end of the applicable service period, as well as adjustments (

  • c, ";.;*

forpbibleerrorsinthepressureandtemperaturesensinginstruments. The actual shift in RT of the vessel material will be established t )periodicallyduringoperatikbyremovingacdevaluat j I '('f;5 rt d ASTM E185-73 and 10 CFR Appendix H, reactor vessel material irradiation sur-fa vet 11ance specimens installed near the inside wall of the reactor vessel in. 9.?. The surveillance specimen withdrawal schedule is shown in ,4 ) the core area. g ( "d' f Table 4.4-5. Sir.ca the neutron spectra at the irradiation samples and vessel \\ inside radius are essentially identical, the seasured transition

r 7 f

I

f. I N

vessel taking into account the location of the sample closer to the core than r,,,,,fQ The heatup and cooldown curves the vessel wall by means nf the Lead Factor. determined from the surveillance l V)mustberer.alculatedwhenthedeltaRTcapsuleisdiffere 2,., delta RT for the epivalent capula l NOT radiation exposure. The pressure-temperature limit lines shown on Figures 3.4-2 and 3.4-3 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minista temperature requirements of Appendin 8 to 10 CFR 50. The maximus RT for all Reactor Coolant System p essure-retaining materials, with the"Neoption of the reactor pressure vessel, has been deter-eined to be 90'F. The Lowest service Temperature limit line shown on since Article M8-2332 (Summer Figures 3.4-2 and 3.4-3 is based upon this RTAddenda of 1972) of S + 100*F for piping, pumps requires the Lowest Service Temperature to be RTand valves. Below maximum of 2tX of the system's hydrostatic test pressure of 3125 psia. The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure ASME Code requirements.

p g p 9-b J -) i REACTOR C00UWIf SYSTEN' 1 i . BASES-PRESSURE / TEMPERATURE' LIMITS (Continued) The OPERABILITY of the Shutdown Cooling System relief valve or.an RCS vent opening of greater than 5.6' square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 whenfone or more of the RCS cold legs is less'than or equal / to 406ap. The. Shutdown Cobling System relief valve has adequate relieving [ ln Ecapability to protect the RCS from overpressurization when the transient is . limited to either (1) the start of an idle RCP with the secondary water ' temperature of the steam generator less than or equal to 100*F above the' RCS cold.Ieg temperatures or.(2) inadvertent safety injection actuation with two (HPSI pumps injecting-into a water _-solid RCS with full charging capacity and -letdown isolated. v v v v-v-n Spc1Eldd IdaME,3N~ N 3/4:4.9 STRUCTURAL INTEGRITY e.,v,,,,,v ss The' inservice inspection and testing programs'for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational ! readiness ofzthese components will be maintained at an acceptable level' throughout the life ofcthe plant. These programs are in accordance with .Section'XI of the'ASME Boiler and Pressure Vessel. Code and applicable Addenda as required-by 10 C8R Part 50.55a(g) excer,t where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a (g) (6) (1). Components' of tho' Reactor Coolant System were designed to provide access L to' permit inservice inspections in accordar.ca with Section XI of the' ASME ' Boiler and Pressure Vessel Code,1974. Edition and Addenda through Summer 1975. 3/4'.4.10 REACTOR COOLANT GAS VENT SYSTEM Reactor coolant system gas vents are provided to exhaust noncondensible gases from the primary system that could inhibit natural circulation core cooling following a non-design bases accident. ' The OPERA 81LITY of at least one reacter coolant system vent path from the reactor vessel head and the pressurizer steam space' ensures the capability exists to perform this function. The design redundancy of the Reactor Coolant Gas Vent System serves to 1ainimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, or control system does not prevent isolation of the vent path, The function, capabilities, and testing requirements of the Reactor Coolant r Gas Vent System are consistent with the requirements of Item II.D.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.

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on s

4 ~ TESTING-AND ANALYSIS Of UNIRRADIATED HEAT-AFFECTED . ZONE-(HAZ)-MATERIAL FROM THE SAN ONGFRE NUCLEAR i. GENERATING STATION UNIT 3L(SONGS-3) to 0 ' SOUTHERN CALIFORNIA EDI' SON COMPANY." .1 May 31, 1989 -l by i -:g M. P. Manahan and J..Garrabrandt BATTELLE 505 King Avenue Columbus, Ohio 43201-2693 and l J. Talnagi and H. Basha THE OHIO STATE UNIVERSITY 1298 Kinnear Road Columbus Ohio 43212 f z. h 4. ( ar = __.-_-:o

y-9 p t.. L 1 l l s .] .t Battelle does not engage in research for ' advertising, sales promotion, or. endorsement of Southern California Edison interests including raising investment capital . or recommending investment decisions or othar public. 'ity purposes, and this report shall ~ not be reproduced .in whole or in part for such purposes.. Because of the experimental nature of'this project, Southern California Edison assumes sole responsibility .for any use of or inability to use any information, results, process, or apparatus disclosed in this report. Battelle assumes no liabil'ity for the conse-quences:of any use of any information, results, pro-cess, or apparatus or for the accuracy, adequacy, or efficacy thereof. i i i j I ) _--_L_.

B.; t-ih_ v*- (' TABLE OF CCNTENT,5 1 i Page

1.0 INTRODUCTION

1 b 2.0 METALLOGRAPHY ANO DIMENSIONAL VERIFICATION OF i BASEL'"E SPECIMENS............,......... 3 e 12 3.0 CHARPY TE3 TING AND ANALYSIS 3.1 Specimen Preparation and Testing 12 1 NDT Analysis 19 3.2 RT 4.0

SUMMARY

25

5.0 REFERENCES

26 Appendix.A. Pre-Test' Photographs of Charpy Specimens Showing Notch Location in the Heat Affected Zone..... A-1 Appendix 8. Post-Test Photographs of Charpy Specimens Showing Fracture Surfaces.................... B-l' t iii

j LIST OF FIGURES' Pace FIGURE 2-1. LOCATION OF CHARPY V-NOTCH IMPACT SPECIMENS WITHIN THE HEAT-AFFECTED-ZONE (HAZ) METAL,......... 4 FIGURE 2-2. RESULTS OF DIMENSIONAL VERIFICATION OF CE ARCHIVE 5 UNIRRADIATED HAZ CHARPY SPECIMENS FIGURE 2-3. CE UNTESTED SPECIMEN 44E............... 7 FIGURE 2-4. CE SPECIMEN 42U TESTED AT 80 F............ 8 s FIGURE 2-5. CE SPECIMENS 471 AND 475 TESTED AT 120 F........ 9 i FIGURE 2-0. CE SPECIMEN 43K TESTED AT 160 F 10 l FIGURE 3-1. CHARPY V-NOTCH IMPACT ENERGY VERSUS TEST TEMPERATURE FOR THE UNIRRADIATED HAZ METAL SPECIMENS FOR SONGS-3. 15 FIGURE 3-2. CHARPY V-NOTCM LATERAL EXPANSION VERSUS TEST TEMPERATURE FOR THE UNIRRADIATED HAZ METAL SPECIMENS 16 FOR SONGS-3 FIGURE 3-3. CHARPY V-NOTCH PERCENT SHEAR VERSUS TEST TEMPERATURE FOR THE UNIRRADIATED HAZ METAL SPECIMENS FOR SONGS-3. 17 FIGURE 3-4. COMPARISON OF BATTELLE AND CE HAZ CHARPY DATA 21 FIGURE 3-5. COMPARISON OF CE BASE, WELD, AND HAZ DATA 22 FIGURE 3-6. COMPARISON OF BATTELLE HAZ DATA WITH CE BASE AND WELD DATA......................... 23 FIGURE 3-7. COMPARISON OF CE BASE, WELO, AND HAZ DATA FOR SONGS-2.......................24 j ) LIST OF TABLES l 1 Pane l TABLE 2-1. CE IMPACT TEST RESULTS FOR HAZ METAL PLATE C-6802-1 AND BATTELLE POST-TEST EXAMINATION RESULTS 6 TABLE 3-1. CALIBRATION OATA FOR THE HOT LABORATORY CHARPY IMPACT MACHINE USING AMMRC STANDARDIZED SPECIMENS... 13 TABLE 3-2. CHARPY V-N0iCH IMPACT RESULTS FOR UNIRRADIATED HAZ METAL SPECIMENS FOR SONGS-3............ 18 TABLE 3-3. SUMARY OF CHARPY IMPACT PROPERTIES FOR UNIRRADIATED HAZ MATERIALS FOR SONGS-2 AND SONGS-3......... 19

u a J FINAL' REPORT J i TESTING AND ANALYSIS OF UNIRRADIATED HEAT-AFFECTED-ZONE'(HAZ) MATERIAL FROM THE SAN ONOFRE NUCLEAR GENERATING STATION UNIT 3 (SONGS-3) by M. P. Manahan and J. Garrabrandt Battelle_ j and J. Talnagi and H. Basha The Ohio State University May 31, 1989

1.0 INTRODUCTION

The 50NGS-3 surveillance program materials were fabricated from SA 533-B, Class 1 steel. The welds in the beltline region'are Mil B-4 submerged arc weldments. The baseline data was measured by Combus-tion Engineering (CE) and the results were reported in Reference (1). The HAZ Charpy data reported in Reference (1) exhibited substantial scatter. As a result of this, Southern California Edison (SCE) askea Battelle to review the baseline heat-affected-zone (HAZ) data and pro-vide recommendations concerning further testing. As stated in the Ref. erence (2) letter, the CE data is believed to be invalid due to the high scatter for tests conducted above 40 F, and Batteile recommended that additional testing and analysis be done. Based on review of the CE data by SCE and the Battelle recommendation, SCE decided to authorize Battelle to examine the CE specimens. The HAZ Charpy specimens tested by CE plus three untested CE specimens were located and sent to Battelle. Metallographic analysis was performed on several specimens and a full dimensional verification ) of the three untested CE specimens was performed. These results are presented in Section 2.0 of this report. oO__________.__.___

F

c.

2 ~ ' Based on the results of Battelle's examination of the CE spec-imens.and data, SCE-decided to go forward with testing a new set of HAZ Charpy specimens. The archive weldment (plate C-6802-1) was sent to L Battelle, and HAZ Charpy specimens were prepared and tested. These data-U -are presented'in Section 3.0 of this report. The results are summarized in Section 4.0. _ _ ~ -. - - - _ _. - - - -. - _ _ - - - _ _ _ _ _ - - _ _ - - _

~ 1

. Di v

o 3

2.0 METALL0 GRAPHY AND DIMENSIONAL VERIFICATION OF BASELINE SPECIMENS During construction of the 50NGS-3 pressure vessel, surveil-lance specimens were removed from several beltline plates for inclusion in the irradiation program and for baseline data generation. HAZ metal specimens were removed from flat slabs cut parallel to both the plate inside diameter (IO) and outside diameter (00) surfaces at a deptn of one-quarter (1/4T) and three-quarter plate thickness. The axes of the specimens are parallel to the rolling direction. Figure 2-1 is an iso-metric drawing showing the Charpy specimen orientation and location within the beltline plate HAZ metal. Further details are provided in Reference (3).

A full dimensional verification of the three CE HAZ Charpy specimens was performed to test the hypothesis that the anomalous data may have been caused by improper specimen machining. As shown in Fig-ure 2-2, the specimens are within the ASTM E23 dimensional tolerances. With regard to specimen machining, there still remains the question of residual stress. Residual stress measurements were yet performed as l. part of this work. The next step was to determine whether the notches in the CE specimens were located within the HAZ region. The untested specimens, j as well as the specimens tested at 40 F (2 specimens), 80 F (2 speci-l mens), 120 F (2 specimens), 160 F (2 specimens), and 200 F (2 speci-mens), were etched in the vicinity of the notch. In all the work reported herein, Vilella's etchant was used. Since several of these specimens were tested near the upper shelf, it was necessary to mount, grind, and polish the surfaces to obtain useful micrographs. The-results of this investigation are provided in Table 2-1 and Figures 2-3 through 2-6. As shown in Figure 2-3, the notch for untested specimen 44E was machined at the weld-HAZ interface. The notches for the other two untested specimens (412, 418) were machined within the HAZ region. How-ever, the notch for speci ne, 41B was machined close to the base-HAZ interface on one side and close to the weld-HAZ interface on the other I

E. 1 4 i WenalMessi g" y10 Surtase / / /s m l n i gene Meest + l 3f4 : I / ) I a ./ f , I )!/ > l / / I / / i ' l / f i t / s JI l I I I l l l / ! / 1 4 4 / / 4/ I l l / c / k. ,/,'/ 1 -7x aa - / // / i N g ' '/ Principal Rolling Directic; = r -e t

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1 /Z V $/ FIGURE 2-1. LOCATION OF CHARPY V-NOTCH IMPACT SPECIMENS i WITHIN THE HEAT-AFFECTED-ZONE (HAZ) METAL 1 l' 1

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] J 5 ) J 4 A h c B 'R h y A u 4 8 D g -j l / >\\ 4 I t F i J L SPECID A B C D F N R W X 41B 2.1644 1.0848 45 - 0' .3946.3945 .0006 .0787(SI) .0105 .3945 All Fem r*,i .0787(S2) l o l 44E 2.1643 1.0819 44 - 30' .3944 .0011 .0794($ 1) .0105 .3943 1.ess Then .0785(S2) ~ o o I ) .0785(S2)

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412 2.1646 1.0827 44 - 55' .3946.3945 .0009 .0790(SI) .0105 . /i42 Specimen Dimension, inches FIGURE 2-2. RESULTS OF DIMENSIONAL VERIFICATION OF CE ARCHIVE UNIRRADIATED HAZ CHARPY SPECIMENS 1'

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l 11 sida. Sequential grinding after testing would be necessary to verify that the crack plane remained within the HAZ. E As indiceted in Table 2-1, most of the data tested-at or above. 80 F are in question. It is interesting to note that all of the speci-mens tested at 120 F or above come from the 3/4T position. As described later, about 60' percent of the 3/4T Charpy blanks machined at Battelle were rejected because it was impossible to notch the specimens in such a way that the crack would remain in the HAZ region during testing. This is due to the weld orientation in the sample block relative to the plate section from'which the specimens were machined. If the portion of the 1 weld sample block which CE used to prepare the 3/4T specimens'was simi-lar to the portion used at Battelle, then this could explain the anoma-lous data obtained using the 3/4T specimens. Based on the above cursory investigation, Battelle recommended that Cnarpy specimens should be prepared and tested. It is important to note that the observations reported herein are sufficient to raise doubts concerning the validity of the CE data. Extensive additional work would be needed to show conclusively that the CE data are invalid. For example, sequential grinding, polishing, and etching would be necessary to identify those specimens where the crack left the HAZ region. Residual stress measurement would be needed on the three untested samples to completely validate the CE machining operation.

.l 1 12 1 3.0 CHARPY TESTING AND ANALYSIS As discussed above, sufficient questions were raised concern-ing the CE HAZ Charpy data reported in Reference (1).to warrant additi-onal testing. The archive weld block was sent to Battelle and 1/4T and 3/4T specimens were machined in accordance with ASTM E23 and these specimens were fabricated consistent with the orientation prescribed.in Reference (1). 3.1 Specimen Preparation and Testino Fifteen specimen blanks were prepared from the 1/4T position and fifteen from the 3/4T. Prior to notching, the specimens were etched on all four surfaces to reveal the HAZ. At that stage it'was necessary to reject nine 3/4T specimens (60 percent) because it was not possible - to position the notch such that the crack plane would be contained within the HAZ. In order to ensure that an adequate number of specimens were available for testing, three additional 1/4T specimens were pre-pared. Full dimensional verification was performed and all but two specimens were within specification. These two specimens were reground and renotched and then passed the dimensional check. Photographs of the specimens showing tne notch location in the HAZ are provided in Appendix A. Charpy impact tests were conducted using a 240 ft-lb Riehle impact machine in accordance with ASTM specifications. The 240 ft-lb range was used for all tests. Velocity of the hammer at impact was 16.87 ft/sec. Calibration of the machi' was verified as specified in ASTM E23-82, and proof tested using a set of standard Charpy specimens obtained from the U.S. Army Materials and Mechanics Research Center (AMMRC) of Watertown, Massachusetts. Results of the proof tests are listed in Table-3-1. Testing of the specimens followed the ASTM E23-82 standard method for notched bar impact testing of retallic materials. ASTM procedures for specimen temperature control were utilized. The low

r i 13 TABLE 3-1. CALIBRATION DATA FOR THE HOT LABORATORY CHARPY IMPACT MACHINE USING A MRC STANDARDIZED. SPECIMENS Energy for AMRC Average of Standard) Variation Between!BCL Average Energy)ta And AMRC Standard Enercy 5 tests Group (ft-lb) (ft-lb Actual Allowed Low Energy 12.5 11.8 0.7 ft-lb 11.0 ft-lb High Energy 75.8 72.7 4.3 percent +5.0 percent (a) Established by U.S. Army Materials and Mechanics Research Center. t temperature bath-consisted of a refrigeratioit unit containing methyl alcohol. It was necessary to add liquid nitrogen to a methyl alcohol bath to reach the -120 F test temperature. The alcohol was agitated to minimize temperature variation in the bath. The liquid level of the bath was maintained so that a minimum of 1 inch of liquid over the spec-imens was maintained. Each Charpy specimen was held at temperature for at least the minimum time (11.8 F for at least 5 minutes) recommended by ASTM E23-82. Tests above room temperature were conducted in a simi- ~ 1ar manrer using a heated oil bath. Specimens were removed from the bath and impacted in less than '5 seconds. A few specimens were impacted within 6.5 seconds. A specimen with a thermocouple attached was trans-ferred from the bath to the Charpy machine support and the temperature I continuously recorded. The temperature variation at 6.5 seconds was found to be negligible in the vicinity of the V-notch. The energy required to break each specimen was recorded and plotted as a function of test temperature. l Lateral expansion was determined from measurements made with a lateral expansion gage. The amount of lateral expansion as a function j of test temperature was also plotted. Fracture appearance (percent shear) of the Charpy specimens was estimated from photographs of the fracture surface using a planimeter. o'

14 The test temperatures chosen were identical to those used by CE so that the data could be directly compared. The one exception to this is that the 250 F test temperature was not run since the upper shelf was already well defined. The remaining specimens were used to in accordance with the provide the data needed to establish the RTNDT ASME code. The C'.srpy impact data were prepared and reported in accord-ance with ASTM E186-82. The Charpy data were fit by means of a statistical analysis methodology originally developed by Battelle. The code is entitled ' Statistical Analysis Methodology for Mechanics of Fracture (SAM McFRAC)M,5) The approach is based on a two-parameter Weibull distri-butior., and has been tested against other methods for analyzing Charpy data. The method was shown to be superior to manual (french curve) or hyperbolic tangent fitting because it is not constrained to a given functional-form, it reduces human bias, and it allows the calculation of temper?ture-dependent variances and 95-percent confidence intervals. The less accurate hyperbolic tangent' fit was also plotted for comparison. The resu'ts of tests conducted for the HAZ specimens are listed in Table 3-2. The total impact energy is the amount of energy absorbed by thi sptcimen tested at the indicated temperature. Lateral expansion 's a measure of the plastic deformation produced during testing. Lateral expansion is determined by the largest change of specimen thickness in the plane of the notch which usually occurs at the back of the specimen. Fracture appearance is a visual estimate of the amount of shear (ductile type of fracture) appearing on the specimen fracture surface. Plots of the impact properties (impact energy, lateral expan-sion, and fracture appearance) versus test temperature are provided in Figures 3-1 through 3-3. These figures show the impact properties as a function of temperature. Appendix B contains the fracture surfaces of the Charpy specimens. A sumary of the test indices is given in l Table 3-3. These data show scatter consistent with that observed indus-try wide. The energy, lateral expansion, and fracture appearance data form consistent, temperature-dependent trend curves. The Battelle E __ _ _ __ _

i 4 r A i 4 e T 5 anr M aw 9 r ui u a T pa D E W 'f m D e i v A E H B w T + 0 A 03 TAD 0 M Y8 0 e P E 2 e V R r ALE 4, p+ g e d HC ( + 0 C 0 E s 1 I l ZA N I T A A A H T + RE A 0 + 3D r = Y B + Te T I t N5 0 T UE 0 S D 1 E E T SGN 0 O 0 8 2 0 0 0 0 0 O-0 6 2 8 4 2 1 1 'g >

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A Q T n m8 N m .uu D m r m eT N E wF y B + 0 A 0 S T 3 E AV I DN EC 0 M 0 Y Pm 2 es R A r A e HA d T 0 ( CA 0 E D 1 t5 Z E 1 T AC HN AR A E 0 O R r 3 A E a P TP T P INA 0 T 0 S U E 1 E R T g 8UT GC NA 0 0R 0 F 2 8 0 0 0 O 0 O-0 6 2 B 4 2 1 1

7 9-3 l W': ? q ;.y A g 18 ' developed SONG-3 HAZ data are consistent with the CE SONGS-2 HAZ data as 4 - shown;in Table 3-3. Visual observations of the fractured specimens are summarized ' in Table 3-2. For the tests conducted above 40 F, the cracks were gen-e erally planar, but not at right angles to the specimen surface. The departure from planar, vertical fracture was substantially-.less in the- ' Battelle specimens than in the CE specimens. Additional work is needed to confim that the fracture was through the HAZ.. TABLE 3-2. CHARPY V-NOTCH IMPACT RESULTS FOR UNIRRADIATED HAZ METAL SPECIMENS FOR SONGS-3 , Test Visual. 'I Temper - Impact Lateral Fracture Exami~ nation. y ' Specimen

ature, Energy, Expansion, Appearance, of Crack Identification F

ft-lb mils % shear Plane 351- -120-7.5 5.0-11.0 flat 3S31 -120 9.5 3.2 9.3-flat 1511' -80 23.5 13.0 12.4 flatJ 1512 -80 26.0 16.6 17.8 flat 1517 -40 27.5 21.8 -- 13.1 flat 354- -40 57.0 36.2 31.1 flat 355-0 39.5 31.8 49.2 fairly flat' 159 0 64.5 43.2 55.4 fairly flat-1516 20 61.5. 46.6 52.0 flat 155 20 76.0 49.8 54.7-flat L 153 20 82.0 50.6 56.3 fairly flat 1518 40' 93.0 63.8 63.1 fairly flat 1510 40 99.0 64.0 73.4 slanted ' 1513 40 105.0 64.4 72.9 slanted 158-80 113.0 72.6 100.0 slanted 156 80 115.0 65.0 100.0 slanted 154 120 122.0 83.4 100.0 slanted 1514 120 160.0 86.6 100.0 slanted 157 160-128.0 87.2 100.0 slanted 152' 160 153.0 82.2 100.0 slanted' 356 210 118.0 70.8 100.0 slanted 1515 210 132.0 75.8 100.0 slanted 1 _m-___m

f t, a j. p. r 19 TABLE 3-3. SUMARY OF CHARPY IMPACT PROPERTIES UNIRRADIATED-HAZ MATERIALS FOR SONGS-2 AND SONGS-3 30 ft-lb 50 ft-lb 35-Mil Lateral Upper Transition Transition . Expansion Shelf Temperature, Temperature,. Temperature, Energy,; Material F F F ft-lb ' HAZSONGS-2(2) -61.1 -28.1 -20.0 134.6 HAZ.50NGS-3(3)-59.7 -21.4 -19.7 133.1 (1) The SAM McFRAC code was used to fit the data. Both units have' pressure vessels fabricated from SA 5338, class 1, mil B.4 weld. (2lCEdata.- (3 Battelle' data. 3.2 RTNOT Analysis Substantially different'results were obtained by Battelle in comparison.with CE. This is illustrated in Figure 3-4. As shown in Figure 3-5, the CE HAZ data scatter above 0 F spans the base and weld data. This behavior would be expected if the notches were not machined in the HAZ region. The CE'and Battel!e data are in reasonable agreement . below about 0 F (Figure 3-4). Figure 3-6 compares Battelle HAZ data with CE base and weld data. The Battelle data is near the upper confi- ' dence band of the weld data throughout the transition region. This is consistent with the data trend observed for SONGS-2 as shown in Fig-ure 3-7. The 50NGS-2 vessel was also constructed using SA 533-B, Class 1 steel. Based on the evidente available at the present time, Battelle recommends that'the CE HAZ Charpy data be discarded. We believe that the data reported herein is accurate, consistent with known data trends, and is of the highest quality. Based on this recommendation, the ini-tial RT for the HAZ metal for SONGS-3 is -40 F. This value for the ] NOT i RTNDT was determined in strict conformance with the ASME code using the CE drop weight data (1) and the Battelle Charpy data reported here. The ______=_z_ )

4 - c. 20 NDT reported in refer 6nce (1) is -40 F and three Charpy tests, conducted at 20 F were above 50 ft-lb. l l 1 i i ^ 1 i 3

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m. 3 h.' IC-p q p q i i-25 1 I 4.0 SUPNARY j is Charpy data reported by CE for tests above 0 F exhibits ,q 'Y excessive scatter which brackets plate and. weld data in the transition l] region and on-the upper shelf. Based on an examination of tested and 1 I untested CE specimens, Battelle-generated data and data trend-analysis, we believe that there are' sufficient questions concerning the validity ' [. of the CE data to preclude its further use. These questions are largely associated with inherent geometry unique to HAZ metal. Based on CE drop weight data and Battelle Charpy data, the initial RT.WT for the SONGS-3. HAZ material is -40 F. 3

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L 26

5.0 REFERENCES

(1) R'agl, A.,'" Southern California Edisor, San Onofre Unit 3, Evaluation of Baseline Specimens", Report No. TR-S-MCM-004, Prepared by the Nuclear Power Systems, Combustion Engineering, Inc., November 30, 1979. (2) -1.etter from M. P. Manahan, Sr...Battelle, to Mr. Phil Brashear,. Southern California Edison Company, dated 2/16/89. (3)' CE TestiSpecimen Drawing-#5023-913-16-2 (E-1470-165-112), " Heat Affected-Zone Metal Test Specimens", CE Pok r Systems, March 26, 1979.. m. (4) Manahan, M. P., Quayle, S., Rosenfield, A. R., and Shetty, D. K., ,\\ " Statistical Analysis of Cleavage-Fracture Data", Presented.at the-l International Conference and Exhibition on Fatigue, Corrosion Cracking.Fractur e Mechanics, and Failure Analysis, Salt Lake City -(December 1985). (5) Menahan, M. J., et al, " Statistical Methodology _for Analysis of i E Fracture Mechanics Data", Final Report to Corporate Technical 1 Development, Battelle Memorial Institute, Columbus, Ohio (January 1985). I l l 1

( I i APPENDIX A PRE-TEST PHOTOGRAPH 3 0F CHARPY SPECIMENS SHOWING NOTCH LOCATION IN HEAT AFFECTED ZONE f i i 1

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.,2..., .st-- nn .W a. . /.L :fFhh, Nki-j[?..sh..ME([N. ,.4 -P. f'1:.f t ':.c - 91:E;En 9.F.M l. SPECIMEN IS9

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SPECIMEN IS2 .____m__,_ _____.m.__

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