Text

Proposed Tech Specs Revising Pressure/Temp Limits, Overpressure Protection & Reactor Vessel Matl
	ML20210U640
Person / Time
Site:	Catawba
Issue date:	09/15/1997
From:	; DUKE POWER CO.
To:	;
Shared Package
ML20210U635	List: ML20210U632; ML20210U640; ML20217A662; ML20217A670;
References
	NUDOCS 9709220005
	Download: ML20210U640 (93)
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- - _ _ _ _ _ _ _ - _ _ _ - - _ _ - - _ _ _ - _ - - - _ - _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

Attachment la Catawba Unit 1 Current Technical Specifications Marked Cory 9709220005 970915 PDR ADOCK 05000413 P PDR l nn J

LIMITING CONDITIONS FOR OPERATION AND SURVElllANCE RE0VIREMENTS SECTION EAE 3/4.4.3 PRESSURIZER . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-9 3/4.4.4 RELIEF VALVES . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-10 3/4.4.5 STEAM GENERATORS ..................... 3/4 4-12 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE IN G TED DURING INSERVICE INSPECTION ........ .... 3/4 4-17 i TABLE 4.4-lt STEAM GENERATOR TUBE INSPECTION ............ 3/4 4-18 3/4.4.6 REAJTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems . . . . . . . . . . . . . . . . . 3/4 4-19 Operational Leakage . . . . . . . . . . . . . . . . . . . . 3/4 4-20 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES . . . . 3/4 4-22 .

1 3/4.4.7 CHEMISTRY . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-24 l TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS ........ 3/4 4-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . 3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY , . . . . . . . . . . . . . . . . . . . . 3/4 4-27 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >l pCi/ gram DOSE EQUIVALENT I-131 . . . . . . . . . . . . . .'. . . 3/4 4-28 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-29 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System .................. 3/4 4-31 FIGURE 3.4-2 REACIOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP T0-& EFPY ............... 3/4 4-32 FIGURE 3.4-3 REACTOR COOLANT SYSTEM C N LIMITATIONS - -

APPLICABLE UP TO EFPY ... . . . . . . . . . . 3/4 4-33

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REACTOR COOLANT SYSTEM /f 00 gSd$ (45 len CdihMc OVERPRESSURE PROTECTION SYSTEMS

b N8I Mo LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following

' erpressure Protection Systems shall

be OPERABLE- ' .

a. Two power operated relief alves (PORVs) with a lift setting of less than or equal tot 0 pd,p, or

b. The Reactor Coolant System depressurized with a Reactor Coolant System vent of greater than or equal to 4.5 square inches. >

APPLICABILITY: MODE 4 when the temperature of any Reactor Coolant System cold leg is less than or equal to 285'F, MODE 5 and MODE 6 when the head is on the reactor vessel.

ACTION:

a. With one PORV inoperable in MODE 4, restore the inoperable PORV to OPERABLE status within 7 days or complete depressurization and venting of the Reactor Coolant System through at least a 4.5 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b. With one PORV inoperable in MODES 5 or 6. restore the inoperable PORV to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or complete depressurization and venting of the Reactor Coolant System through at least a 4.5 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c. With both PORVs inoperable, complete depressurization and venting of the Reactor Coolant System through at least a 4.5 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,

d. In the event either the PORVs or the Reactor Coolant System vent (s) are used to mitigate a Reactor Coolant System pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or Reactor Coolant System vent (s) on the transient, and any corrective action necessary to prevent recurrence.

e. The provisions of Specification 3.0.4 are not applicable.

CATAWBA - UNIT 1 3/4 4-36 t

Amendment No. % i l

l meuc -wanr ,nwwwsrw murmmun -J

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNOT, at the end of the effective full power years (EFPY) of service life as indicated on the appli-cable heatup or cooldown curves. The service life period is chosen such that the limiting RT at the 1/4T location in the core region is greater than the RT NOT of thIoIimiting unirradiated material. The selection of such a limiting RT 3.T assures that all components in the Reactor Coolant System will be operated c)onservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RTNOT; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (,E greater than 1 MeV) irradiation can cause an increase in the RT ggy. Therefore, an adjusted reference temperature, based upon the fluence, copper content, and phosphorus content of the material in question, can be predicted using figure B 3/4.4-1 and the largest value of ARTNOT-forUnit23e6djustedreferencetemperaturehasbeencomputedusing the guidance of Regulatory Guide 1.99, Revision 2. For 'Jn' t 1, th: =:ly;i; d::::ented in "CA" 11:27 :nd rcenalj cd using the guidercc of Rcguieto;y Cu;d; i.00, "cticier 2, %dicates pe heatup and cooldown limit curves in Figures 3.4-2 and 3.4-3 er applicab,e to bt' unit te predict,;the shift in RTr oT at the end of the identified service life g g.g 43h,, g usi._ s ant deteWd " thii :ne q' M 25d art" M result; from the material ;urve!'!=:0 progr=, ecl=ted =cerding te ASTM GGL, m e

nthbk. C ASTM E185-81/psules ind 10 CFRwill be50, Part removed in H.

Appendix accordance with the requirements The surveillance specimen with-of drawal schedule is shown in Table 4.4-5. The lead factor represents the rela-tionship between the fast neutron flux density at the location of the capsule and the inner wall of the pressure vessel. Therefore, the results obtained from the surveillance specimens can be used to predict future radiation damage to the pressure vessel material by using the Icad factor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ART NOT detemined from the surveillance capsule exceeds the calculated ART NOT for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Sec-tion III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in the following paragraphs.

CATAWBA - UNIT 1 B 3/4 4-9

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LlHITS (Continued) defect at the inside of the vessel wall. The thennal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the l tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the M for the 1/4T crack during heatup is lower than the Kin forthe1/4Tcrackdu$ingsteady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compres-sive thermal stresses and different Kig's for steady-state and finite heatu) rates do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant tem-perature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. -

The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, -

the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thennal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.

I Rather, each heatup rate of interest must be analyzed on an individual basis, i

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point coirparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion, finally, th; ;caposite curves for the h;; tup rate dat; and th cocida r.

retc date ere efjusted fer pos3Me errers in the pressurc or.d tcap;raturc

---sens4eg instr;= hts by the value indic+ted en the respective curve'..

CATAWBA - UNIT 1 B 3/4 4-14

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A.cfcf f*- J-MA PRESSURE / TEMPERATURE LIMITS (Continued)

Although the pressurizer operates in temperature ran

which there is reason for concern of nonductile failure, ges above those operating limits(or are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

LOW TEMTERATURE OVERPRESSURE PROTECTION The OPERABILITY of two PORVs or a Reactor Coolant System vent opening of I at least 4.5 square inches ensures that the Reactor Coolant System will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the cold legs are less than or equal to 285'F. Either PORV has adequate relieving ca) ability to protect the Reactor i Coolant System from overpressurization when tie transient is limited to either: (1) the start of an idle reactor coolant pump with the secondary water temperature of the steam generator less than or equal to 50'F above the cold leg temperatures, or (2) the start of a Safety injection pump and its injection into a water solid Reactor Coolant System.

The Maximum Allowed PORV Setpoint for the Low Temperature Overpressure Protection System (LTOPS) is derived by analysis which models the gerformance I

' of the LTOPS assuming various mass input and heat input transients. Operation with a PORY Setpoint less than or equal to the maximum (9tbMDensures that Appndix G criteria will not be violated with consideration (for a maximum pressure overshoot beyond the PORV Setpoint which can occur as a resuic vi -

time delays in signal processing and valve opening, instrument uneartainties, j and single failure. To ensure that mass and heat input transients more severe -

than those assumed cannot occur, Technical Specifications require lockout of all but one Safety Injection pump and all but one centrifugal charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of a RCP if secondary temperature is more than 50'F above primary temperature.

The Maximum Allowed PORV setpoint for the LTOPS will Le updated based on the results of examinations of reactor vessel material irradiation surveil-lance specimens performed as required by 10 CFR Part 50, Appendix li, and in accordance with the schedule in Table 4.4-5.

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Catawba Unit 2 Current Technical Specifications Marked Copy I

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I LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i

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SECTION

.P.f9.1 3/4.4.3 PRESSURIZER . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-9 3/4.4.4 RELIEF VALVES . . . . . . . . . . . . . . . . . . . ... 3/4 4-10 3/4.4.5 STEAM GENERATORS ..................... 3/4 4-12 l

TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION ...... ....... 3/4 4-18 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION ............ 3/4 4-19 ;

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems . . . . . . 3/4 4-20

.......... f Operational Leakage . . . . . . . . . . . . . . . . . . . . 3/4 4-21 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES . . . . 3/4 4-23 3/4.4.7 CHEMISTRY . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-25 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS . . . . .... 3/4 4-26 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEIL REQUIREMENTS . . . . . . . . . . . . . . . . . LANCE . .... 3/4 4-27 3/4.4.8 SPECIFIC ACTIVITY . . . . . . . . . . . . . . . . . . . . . 3/4 4 28 ,

FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ,

ACTIVITY LIMIT VERSUS PERCENT OF RATED THERHAL POWER ;

WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >l pCi/ gram

(

DOSE E00! VALENT I-131 . . . . . . . . . . . . . . . . . 3/4 4-29 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM ........................

3/4 4-30 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System . . . ............... 4-32 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP T0 FPY ,.............. 3/4 4-33 FIGURE 3.4-3 REACTOR COOLANT SYSTEM C0 WN LIMITATIONS -

APPLICABLE UP T0-t&-EFPY .

. . . . . . . . . . . . . - 3/4 4-34

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INDICATED TEMPERATURE (DE . F) 5 i

curve ArrucAs som NE ATuP wATERiat sAsis RATES UP TO P4R FOR THE t

$ERVICE Pt Rt us TO 10 EP CONTROLLING MATERIAL UNIT 2 INTERME01 ATE SHELL coNTAiNs ATE 50006-2 '

moth os 19 h AND 80 PS FOR PCSS1BLE ,

*( COPPFR CONTENT-.0.07wn N)CK EL CONTENT- 0.01 wt it I

INSTA NT k m O'R S. !

RTwortNITI AL-33*P 5

, RT,,oyAFTER to arry iteT.104* j b $. ,,.T., .

f & FIGURE 3.4-2 l

/

/ ) R CTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICASLE FOR Tile FIRST 10 EFPY i

l 1

CATAWBA - UNIT 2 3/4 e.-33 Amendment No. 2

[

2 e

MATERIAL PROPERTY BASIS LIMIITNG MATERIALS: IhTERMEDIATE SHELL, B8605 2 LIMirING ART AT 15 EFPY: 1/4-t,112.6 T 3/4 t, 96.0 7 2,500 iiiiiiiiiiii, ,

LEAK TEST LlWIT- .,_ '

I i I H I I 2,250 ' '

l l i i

r ; ;

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i I C) 2,000 ' ' '

~

1 1 (f) I I h 1.,750 UNACCEPTABLE OPERATION f f r' g

L 1.,500 I I

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j 3 / / ._.___

l g [ ,

ACCEPTABLE .----

(f) 1.,250 HEATUP FMTE

/ /

OPERATION ~_(([

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/ ,

1.,000 /

.O G -

/ i y < -

d 750 '

j O -

, . -_. /

] 500

. L

\ CRITICALITY LIMIT EMSED ON 250 INSERVICE HYDADSTATIC TEST TEWERATURE (245 *F) FOR THE SERVICE PERIOD UP TO 15 EFPY O IiII II I Iii IIIIIII O 50 100 150 200 250 300 350 400 450 500 I ndicated Temperature (Deg . F)

FIG 4RE 3. 4 -

FIGURE ii-i 4^.T *."'".'. '.",~ 2 REACTOR COOLAhT SYSTEM HEATUP LIMITATIONS (HEA~IUP RATE OF 60*F/HR) APPLICABLE FOR THE FIRST 15 EFPY (WITHOUT MARGINS FOR INS'IRUMENTATION ERRGRS)

CATAm 6A - RMir 2.

D 3l0 W 33 AmesJ sed- No, g -

n_ : -... . , n s_; s eu: . -

w- ' .: r - ._ _ _,,_,__ . _ _ , _ . . ,,, _ _,__,

.',w

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\

2250 \ l N /

2000 \ / ;

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OPE RAl lo '

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k [

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4.

1250 \ /

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5 1000 l '

5 CCOLDOWPI CEPTABLE z RATES op,rHR O RA"lO !

750 g; W

t 500 2v-0 M, '

60'f/

l 100-250 #

l I i

I. .

0 ,

0 50 100 200 250 300 0. 400 450 500 INDICATED TEMPE R ATURE (DEG, F) curve AnticAsts Po CoOtoo*N uATE Riat sAsis RATEE UP 70 tes*P POR THE SERVICE PERioO UP 10 EPPY CONTROLLING MATERIAL. UNIT NTERMEOlATE SHELL PLATE f

CONT Aless h4AR01 OF10*P COPPER CONTEN1-0A7e%

ANOto P54G PO POW 4GLE INSTRUMENT HlCKEL CONTENT *0.81 wt%

RORS.

RTNOTINITIALMP g* T NOTAPTER 10 EPPY U4T, tod*P fleu) R (4% A<))"O ee-- -a3 4 - 5

- [ FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -

APPLICABLE FOR THE FIRST 10 EFPY CATAWBA - UNIT 2 3/4 4-34 Amendment No. 2

MATERIAL PROPERTY BASIS LIMITING MATERIALS: IN'IERMEDIATE SHELL, B8605 2 LIMITING ART AT 15 EFPY: !!4 t,112.6 'F 3/4-t, %.0 'F 2.,500 i i

I i l

2.,250 '

i l n l C) 2,000

,/

([] /

O 1.,750 :.-::: UNACCEPTABLE l

OPERATION j I

L 1.,500 3 >

m i e 1.,250 j

/ ACCEPTABLE OPEAATION 1,000 --

/

O

@ . m N 3 '

(d U

750 COOLDOWN RATES 'F/ HA.

(('

w

._. ____ o -,,- >

O 20 500 4a (f[f c ----

.__ :::: so

____ 100 ___

250 0

0 50 100 150 200 250 300 350 400 450 500 I nd i cated Temperature (Deg . F)

F16 4 RE 3.4 ---f!OURE B-2 CATA","JA L?! 2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS (COOLDOWN RATES UP TO 100*F/IIR) APPLICABLE FOR THE FIRST 15 EFPY (WTTHOUT MARGINS FOR INSTRUhENTATION ERRORS)

C.ATA46A-(AAltT E " 10 3/i 4 ~' 3 Y s - + t iJ..

e tr!.:59,'j.y

.:qgegg R

l>

L TABLE 4.4-5 l'

i t

REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE' J T

f,

. CAPSULE VESSEL LEAD j

i NUMBER LOCATION FACTOR' WITHORAWN TIME (EFPY) r t

j U 58.5* 4.01 3,77 Standby 3 ,

i V 61* -3.7' 3 ,67_. ' 9 q W 121.5* 4.01 3,77 Str e; /3.S~ >

n

! f 238.5* 4.0! 3,77 Standby ,

x/ 241' 3.7" 3,Q # % 52,

, Z 301.5* 4.05- //,07 "r t R0fue! Ng O, 8 (,

r I

5 i s

?

i k

F t

L k

k fe fi 5

' CATAWBA - UNIT 2 3/4 4-35 Amendment No.'

t

- m r. -~

jd

+ -

-. w c m a g e a s s;19E,j y g j3 p ,qa g g g' i '

lf REACTOR COOLANT SYSTEM N NkbtSlen cal /6eec11 OVERPRESSURE PROTECTION SYSTEMS *b O b .I Y" b * "

6 PAl h 42.E/Si3(46hwe[h LIMITING CONDITION FOR OPERATION

/

a v

3.4.9.3 At least one of the following 0 pressure Protection Systems shall be OPERABLE:

a. Two power operated relief alves (PORVs) with a lift setting of less than or equal to 450- pig.,or

b. The Reactor Coolant System depressurized with a Reactor Coolant System vent of greater than or equal to 4.5 square inches.

APPLICABillTY: MODE 4 when the temperature of any Reactor Coolant System co'd leg is less than or equal to 285'F, MODE 5 and MODE 6 when the head is on the reactor vessel. ;

I ACTION: '

s

a. With one PORV inoperable in MODE 4, restore the inoperable PORV to ,

OPERABLE status within 7 days or complete depressurization and venting of the Reactor Coolant System through at least a 4.5 square inch vent within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . '

b. With one PORV inoperable in MODES 5 or 6, restore the inoperable PORV .

to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or complete depressurization and '

venting of the Reactor Coolant System through at least a 4.5 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . e 4

a c.

With both PORVs incperable, complete depressurization and venting of $

the Reactor Coolant System through at least a 4.5 square inch vent l

within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . '

d.

In the event either the PORVs or the Reactor Coolant System vent (s) $

are used to mitigate a Reactor Coolant System pressure transient, a :

Special Report shall be prepared and submitted to the Commission i pursuant to Specification 6.9.2 within 30 days. The report shall I describe the circumstances initiating the transient, the effect of 5 the PORVs or Reactor Coulant System vent (s) on the transient, and any corrective action necessary to prevent recurrence, {

e. The provisions of Specification 3.0.4 are not applicable.

j e

l I

l

?

s CATAWBA - UNIT 2 \

3/4 4-37 Amendment No. M

- - w c:.u _ =

p'

,.e.-_._____....._..,

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNpT, at the end ot the effective full power years (EFPY) of servi:e life as indicated on the appli-cable heatup or cooldown curves. The service life period is chosen such that the limiting RT at the 1/4T location in the core region is greater than the RTNOT of thbimiting unitradiated material. The selection of such a limiting RTuo7 assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial the results of these tests are shown in Table B 3/4.4-1. Reactor RTuo7; operat ion and resultant fast neutron (,E greater than 1 MeV) irradiation can cause an increase in the RT gor. Therefore, an adjusted reference .

temperature, based upon the fluence, copper content, and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ART NOT' TerUnit2'[headjustedreferencetemperaturehasbeencomputedusingthe i guidance of Regulatory Guide 1.99, Revision 2. - For " nit 1, th: :n lysi:

i d:: n nt:d in WC^P 11527 :nd rt = ly::d u;ing th: guider.cc of Regeleterj Guide 1.90,"evi;ien2,indi::t;;Theheatu) and cooldown limit curves in Figures 3.4-2 and 3.4-3 ;r; ;pplicabic to beti ur.its te predict the shift in RT NDT at the end of the identified service life. f ___

jgg' gg) 4.jg

^"T m..._u_ m .___dete:-i :d i- thi; =nn r =y bc :::d unti' the r:: alt:

,. _"a .u _1.,

p _;g 7= , ;;;);;; g ;__;73j;g : g7u, = , 37;

-:v:ilable. Capsules will be recoved in accordance with the requirements of ASTM E185-82 and 10 CFR Part 50, Appendix H. The surveillance specimen with-drawal schedule is shown in Table 4.4-5. The lead factor represents the rela-tionship between the fast neutron flux density at the location of the capsule and the inner wall of the pressure vessel. Therefore, the results obtained from the surveillance specimens can be used to preJict future radiation damage to the pressure vessel material by using the lead f actor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ARTg o7 determined from the surveillance capsule exceeds the calculated ART NDT Tor the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in-Sec-tion III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in the following paragraphs.

CATAWBA - UNIT 2 B 3/4 4-9

w - --- %-m..m_._.__-.....-... .. -

m.

e

$ REACTOR COOLANT SYSTEM BASES bl w

h PRESSURE / TEMPERATURE LIMITS (Continued)

NI defect at the inside of the vessel wall. The thermal gradients during heatup

+

produce compressive stresses at the inside of the wall that alleviate the l

!,: tensile stresses produced by internal pressure. The metal temperature at ti.e crack tip lags the coolant temperature; therefore, the Kp i for the 1/4T crack

, during heatup is lower than the K IR for the 1/4T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compres-sive thermal stresses and different Kig's for steady-state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, t .. both cases have to be analyzed in order to assure that at any coolant tem-

[- perature the lower value of the allowable pressure calculated for steady-state

! and finite heatup rates is obtained. .

The second portion of the heatup analysis concerns the calculation of

,p' pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface,

' the themal gradients established at the outside surface during heatup produce

stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.

Rather, each heatup rate of interest must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are .

produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. j i

The use of the composite curve is necessary to set conservative heatup ;

limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion, re. 1, ., +w.,,,m+.-,...,.. < . +w w....., ..+. a.+. ..a.w.-m a-- i vatn Ad m Em d 4((+Ir[ h[ ((cc5w1 e((Ee n + w. n((e [ [

-A + - I iE5ii5 P niU M i h di EiiEEEl iihiUd en 5:55Eh5hh'A':Er$$ ~ '

n .

i CATAWBA - UNIT 2 B 3/4 4-14 f

M cf /4Cor,fph iPS /4.5 fry ryff W<tcer for'n hiW S cts tdoll RS Corr ech'o,1S {s, gbrc f 0** (* ll9'! ?'~

o REACTOR COOLANT SYSTEM Pa4 cym lion amt t/ie sin + feeste<re.-

BASES k' 0 '#" "

h 8f/f/Q it iun cme / f4L focefr b,r ef--

D SESSURE/TEMPERATURELIMI ntinued) fg th #'

Although the pressurizer operates iiitemperature ran H

which there is reason for concern of nonductile failure, ges above those operating limits for are provided to assure compatibility of operation with the fatigue analysis

% performed in accordance with the ASME Code requirements.

W LOW TEMPERATURE OVERPRESSURE PROTECTION f The OPERAillLITY of two PORVs or a Reactor Coolant System vent opening of at least 4.5 square inches ensures that the Reactor Coolant System will be

^ protected from pressure transients which could exceed the-limits of Appendix G to 10 CFR iart 50 when one or more of the cold legs are less than or equal to 285'F. Either PORV has adequate relieving ca) ability to prctect the Reactor

? Coolant System from overpressurization when tie transient is limited to s~ either: (1) the start of an idle reactor coolant pump with the secondary

'( water temperature of the steam generator less than or equal to 50'F above the 6 cold leg temperatures, or (2) the start of a Safety Injection pucnp and its g injection into a water solid Reactor Coolant System.

+

The Maximum Allowed PORY Setpoint for the Low Temperature Overpresp ra

& of the LTOPS assuming various mass input and heat s. input tran3 Protection Sy Operation I

with a PORY Setpoint less than or equal to the maximum"" nsures that Appendix G criteria will not be violated with consideration foi a maximum t

- pressure overshoot beyond the PORY Setpoint which can occur as a~ result of time single and delays failure.

in signal processing and valve opening, instrument uncertainties, To ensure that mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require locko;.t of p all but one Safety injection pump and all but one centrifugal charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of a RCP if secondary temperature is more than 50'F above primary temperature, t

The Maximum Allowed PORV setpoint for the LTOPS will be updated based on the results of examinations of reactor vessel material irradiation surveil-lance specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule in Table 4.4-5.

,4t/ sable W he & 925ps.3 ~

as b4 )

CATAWBA - UNIT 2 B 3/4 4-15 .

_. _ m .m__-- _m._-, - y , 7,j

Attachment lc Catawba Unit 1 Improved Technical Specifications Marked Copy l

l

1 I

RCSP/TLimits

(. 3.4.3 2W --

.: \

2250 - '

l I LEAK TEST LIMi r 2000 - I i

I

/

1750 -

G g, UNACCIPTABL OPERA 1 ON l [

c $50o - l

$ I _

S N

12W _s..- '

f ,

t- i 5 1000 g

/ 37-(

m / A ,

500 i --

/ // .

CRITI LITY LIMIT SA$ l0 ON INSE VICE HYDROS"ATIC 250 TE5TTE . (24t '0FIFOR THE SERVICE P. 103 UP TO 10 EFPY O ERATION i

0 !

0 50, 150 200 250

! b ,

300 ,350 400 450 500 INDICATED TEMPERATURE (DEG, F1 CUAVt 48POAtetATUP AAT8898 0 te*F/NR POA ft4( teAtt AIAL e Asas SE Avt A400 48' TO 10 GPPY Conf AOLLING IdAttRIAL- UNet t tNf t 404Af t SH4(L C9 AlsnAA0tN OP19'F PLAtt teses.

A CO*Pt A Cowf ENT.e AFwet 3 Pt44 POA POtelett i

v. =,6 A AORT. c htCE84 CONTENT. 4.01 et n

,1p ar,,,,taire A t,,,, Ae t A At-se,Py ie e e

v.t. i .

Y ,V \

ms.ser y '

c, 9' Figure 3.4.3-1 RCS Heatup Limitations

('#)

Catawba Unit 1 3.4-10 5/2^4

i I

MATERIAL PROPERT/ BASIS .

UMrrlNG MATERLAL: LCMfER SHEU. FORGING 04 UMITING ART AT 15 EPPY: 1/4 T. 43*F 3/4 T, 26'F 3

2,500 ii44 -

-i., ii,4 o , , 4 , , , . i , , , , . . . .

if ; i ; i . , i i i , , . . , . . . .

LEAK TEST UMIT s f I ' ' i '

' ' ' ' ' i ' ' ' '

N1 .I i / . i 4 e i . .. . l . i , i . 3 i i1 i i i ii .1 i if '

2,250 I ' ' I I !

1

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I i ! I ! I I I ! I i

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J

! I i r i f , j i ; I i ,

i i i I t i i i , t i, ! i . # i . . i i

, , . ,1, . . ,! ,i ,i , ,

2,000 , , , . , i i ,

@ l i

i i

, i i

i I I i i i i e i i i 1 e < ,i I

, i i ,

c.,1,750 UNACCEPTABLE ,/ / ,

OPERATION a / , i S

h.

i iii i i i i ti i

zi i .

i i

3 1,500 e

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7, , , 1 , , . . , , . ,

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(t)

HEATUP RATE '

UP TO 80 F/HR , ,i

/ l r

/ ACCEPTABLE i

l OPERATION '

! 2 1,250 , , / ' '

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n. ll l l l l i i! z is ,i i

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ri , i i. i . i . . .

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. . . ,. . 4 . . , . . , , , . .

ii # i I ,

l i i! . ,

i. I i 6 i i i

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@ i ,Ii t i ! I i i i i i i i i i i i i i i i ! . i i e i..i i . i . .. i e i i . . . i e i i i i i D

' ! +..I I . 6 I I

,! i i ! ! I 8 ! ! i ! 4 I i , i , , i ! 1 i i. . i , !i i . . i 4 , i i . i

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~i BCRITICAUTY UMIT BASED ON ----

, . i . . . . i i INSERVICE HYDROSTATIC TEST l '

l,;; ii.i i i

.; ;j ; { TEMPERATURE (176'F) POR THE ----

4

',i,i;!

i

. i4 i .

i i . .i

.i SERVICE , ,PERIOD , .

UP TO, 15,EFPY,- , ,- ,

0

! O 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg. F)

Figu m 34.3-1 i .% M C _..% LM. i Reactor Coolant System Hostup Umitations (Heatup rates up to 60*F/hr)

Applicable for the First 15 EFPY (Wethout Margins For instrumeniation Errors) i S. 4 -/ O 3

Cocke a- Wtt 1 am d

---.-m.- _.,,_,.r -%r.-.._-,,._w, y _m-. ..-..-r-m.-..__- -

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RCS P/1 Limits 3.4.3

(

\

4:nsa 2250 -

i s

g. \

3 y r

,\

20m -

3 l .

l i k 97$g i /. I' G UNACC TTA8LE

] g OPE RAll N :

I w i600 g 1250 t

/ !

s 5

iom /

CCOLDOWPI E A TES ACCEP1Aa -

$ppgp OPERA"I 7s0 - .; s -

( ' .

$00 20-M I -

6d / \

100-250 /

s 0- - / '

o 50 100 160 200 250 3 350- 400 450 500

. INDICATED TEMPERATURE ( EG,F)

Cuavt APPLtCAS OA COOLDonN AAtt5 w to t M MA 7844 MAttAtAL BAllt St Avett Pts UPTOte(PPY Coast A eais coastAOLLesso adAtin L usett i 18:15ReesetAtt SHELL else Of te*P PLAtt sea 06-3 Asso se Pts 04 PO984GLE c0*Pt A C0estteet eatm is.st a via t Ia nons, heCatt CONTtest 0 41 et orematut.wt at 8ectII" 18 0 FPY '#8I-

Q p ( g e e, W .' N R E t d 8"I *

f.'3u re 3 4 3 - 2. bNc kb Figure 3.4.3 2 RCS Cooldown Limitations Catawba Unit 1 3.4-11 - ' at

's^ v^ g' ^n' -

MATERIAL PROPERTY BASIS ,

LIMITING MATERIAL: LOWER SHEU. FORGING 04 LIMITING ART AT 15 EFPY: 1/4 T, 43*F 3/4 T 26'F 2,500 . . , . ,,, . . . . . i . .. .

i . , , . .,,i iii, i .. ,ii, ! , i ; i . ,i . . ,

.! ' ! !i ! I I . i * * *

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!!!i ii i ! ! ! I I ! if i ii i! Ii i6 I ! i ii e

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2,250 , , , , , . . . . . , , , , . , .

i

. i e i iiii t i i i i *i i,ii iiii ! i i i i i e i ! . ii i > > i i ! ! i i . i i i ii ! I i i i i i i ii i ( [i ie ! i i i i ; i i i i i i , ! ,1l !* I i I , i '

i ii i ! .

i i i e i + 14 i i i .'

2 ,0 0 0 e i i i e

i i if i i e i i

, 1 , , , , ,

I C7) UNACCEPTABLE '! ' ' ' ' ' '

i !

u) -.- OPERATION / l ! .

! ll ll,l llll I

o.

1,750 l..,

1 , . i i

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L,,,,

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i i i i j i . ii il q i i . f. .i, . 1 i i i i i i ie i 6 , i .

] e, J ' ' ! ' , /i e + l t i i e e W . .

i , i e i , i i

i. !, I i i a'

i 4

i i

ACCEPTABLE !

i i !,'i' .

C/) i , i i ! ii !' i i < i i . OPERATION i i ia iii i i O

,_ 1,250 ',

v,

', ,' 4' ,

i i j i , i i +i/ i8iQ 1 .

. i i .

. v.

I e it

! 6 i l i , . 4 i 6 i

. . . i i i i

i i

ie i i i i i i t t i i i i

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i i 4i . i i i 1,000 . . .

-1 4

i i e i

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COOLDOWN %';! ' '

! - i 4 i 4 !i (g RATES *F/HR_.m .

i

, , i ,

i '

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l y -

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40 ,' 'i 'i

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i 500 , , ,

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. . . . ,, . i ,

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250 ' ' '

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+ '

i

. , 5 ,jii i I i i i i 9 0 ' ' ' ' ' ' '

O 50 100 150 200 250 300 350 400 450 500 Indicated Temperature 1Deg. F)

Fra a sn.3-2

";n 0 0 - - Ce.#.,. 'Je tReactor Coolant System Cookiown Umitations (Cooldown rates up to 100'F/hr) Applicable for the First 15 EFPY (Without Margins For instrumentation Errors)

3. tl - ll c.abab s mcf 1 -

m . -

L10P System 3.4.12

(

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Low Temperature Overpressure Protection (LTOP) System LCO 3.4.12 v An LTOP System shall be OPERW.E with a maximum of one charging pump or one safety injection pump capable of injecting into the RCS and the accumulators isolated and either a or b below,

a. Two power r (relief valves (PORVs) with lif t settings 450psiJg or b.

The RCS depressu ized and an RCS vent of a 4.5 square inches.

APPLICABILITY:

MODE 4 when any RCS cold leg temperature is s 285'F.

MODE 5, MODE 6 when the reactor vessel head is on.

............................N0TE----------------------------

Accumulator isolation is only required when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided in Specification 3.4.3.

00 fS$$ kS feO ca\oWeh, ex.\\owahIe- % luc. L 425*psig (a s bmOj N

Catawba Unit 1 3.4-31

--5ff0fe?---

khh n

l (f RCSP/TLimits 8 3.4.3 Y'( BASES ACTIONS C.1 and C.2 (continued)

Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perfom the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Recuired Action C.1 is insufficient because higher than analyzec stresses may have occurred and may have affected the RCPB integrity.

\

3 SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the specified limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, 1

i since temperature rate of change limits are specified in L hourly increments, 30 minutes pemits assessment and correction for minor deviations within a reasonable time.

(

Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.

This SR is modified by a Note that only requires this SR to be perfomed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LC0 3.4.2 contains a more restrictive requirement.

REFERENCES 1. 10 CFR 50, Appendix G.

2. ASME, Boiler and Pressure Vessel Code, Section Ill, Appendix G. g j

3. ASTH E 185 7 8f, -JulM98fb.

4. 10 CFR 50, Appendix H.

(continued)

Catawba Unit 1 B 3.4-16 /20[9[

i LTOP System l B 3.4.12

{

\(

BASES BACKGROUND With minimum coolant input capability, the ability to (continued) provide core coolant addition is restricted. The LCO does not require the makeu safety injection (SI)p control actuationsystem deactivated circuits blocked. Due or theto

' the lower pressures in the LTOP MODES and the expected core decay heat levels, the makeup system can provide adequate flow via the makeup control valve. If conditions require the use of more than one charging pump for makeup in the event of loss of inventory, then pumps can be made available through manual actions.

The LTOP System for pressure relief consists of two PORVs with reduced lift settings or a depressurized RCS and an RCS vent of sufficient size. Two PORYS are required for redundancy. One PORY has adequate relieving capability to keep from overpressurization for the required coolant input capability. 400 psd3 (o /c f P cs /,% W);

cLllukble. s4 lu e .fh. 425* pts's PORY Reautrements )

As designed for the LTOP System, yeh V is signaled to open if the RCS pressure reachesGO pig)when the PORYS are

( in the 'lo-press" mode of operation. ine LTOP actuation logic monitors both RCS temperature and RCS pressure. The signals used to generate the pressure setpoints originate from the wide range pressure transmitters. The signals used to generate the temperature permissives originate from the wide range RTDs. Each signal is input to the appropriate MSSS protection system cabinet where it is converted to an internal signal and then input to a comparator to generate an actuation signal. If the indicated pressure meets or exceeds the calculated value, a PORY is signaled to open, t

' This Specification presents the PORY setpoints for LTOP.

Having the set)oints of both valves within the limits ensures that t1e Reference 1 limits will not be exceeded in

, any analyzed event.

(continued)

Catawba Unit 1 B 3.4-60 1 20/97--

Attachment id Catawba Unit 2 Improved Technical Specifications Marked copy i

l l

5 - - -

RCSP/TLimits 3.4.3

(

m *

/

2250 1 i >

/

LEAK TEST LIMIT t .

I

/

1750 \

G g

'UNACCEPT 8LEl OPERAll0N l [ [

' ; l ; /

.- 1500 3 /

$ l N

{ 1250 1 \,/ 7 8

e- \

$ 1000

\

7M (/

500

/ \

C TICALITY LIMIT 8AS ED 2M AOCEPTABLE l

/

/ .

ON SERVICE 14YOROS"ATIC TE EMP. (24tiOF) FOR THE SERVI IK PERf03 UP TO 10 EFPY O ERATION / i 0 ,

0 50, 100 150 200 250 300 400 450 500 INDICATED TEMPERATURE (DEG. Fl cuave AreticAs eOnwtAfuP inat3miAt easis RATis UP TO nemPO4THE Stevice Ptn UP70108FPY CONTROLLING $4ATERIAL. UNet 3 3 Ass 401Af t SHELL CONTA48e8 4840F198P PLAft -2 aseo se 70m poggigLg COPPf A CostTterf-estwes NtCEEL CoteTENT- gat we 4 lastTRuse ? I AROA L AT teOf68etTl&L-23*P Rg htcc., 6d, % >1 (d "'aot'"'" "" '

Fi3nr e- 3,4 3-1 (Ahc Ae).) j'l *l' Figure 3.4.3-1 RCS Heatup Limitations Catawba Unit 2 3.4-10 5/20/07

MATERIAL PROPERTY B ASIS LIMITING MATERIALS: INTERMEDIATE SHEI.L. B8605 2 LIMITING ART AT 15 EPPY: let, 112.6 'F 34t, 96.0 'F i

2,500 ,,,,,,,,,,,,, , ,

i ; ; i LEAK TEST LlWIT - m '

i i .

N I I i i 2.,250 l ' '

i r ; ; 4 D I I I 6 C) 2.,000 I i

{ ' ' '

@ f f .

i j

Q l I s y 1,,750 UNACCEPTABLE r '

OPERAT10N , j t

1.,500 ' ' ' '

3 i i iii1 1 i iii 1

(f) ; ; ACCEPTABLE '----

i (f) I I OPERATION

'"~~~

1,,250 HEATUP RATE '-~~

UP TO 80 *F/HR g"r, r

/ / ,

y 1.,000 / >

^

p e d 750 '

i U -

4 ,

y e c 500

__ t

, CRITICALITY LIMIT BA!ED ON 1 250 INsEwicE mononATic ten '

TEWERATURE C245 *F) FOR THE SERVICE PER100 UP TO 15 EFPr 4

0 iiiI!Iii'IiiIIii'i'ii 0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg. F) 4 i

f:bwve. 3 43-1 FIGURE L i CATA"OA L"O 2 REAC. TOR COOLANT SYSTEM HEATUP LIMITAllONS (HEATUP RATE OF 60'F/HR) APPLICABLE FOR THE FIRST 15 EFPY OVIIIIOUT MARGINS FOR INSTRUMENTAllON ERRORS) i l C&do % .' t 2-34-/0

~

a

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

RCSP/TLicits 3.4.3

(

\\

/

2250 \ l /

f

\

/ .

i

/

1750 \ l / l i

UNACCEPT 8LE !

f" OPE RATION '

1503 a

a g aw

\ <

'\

l/ / '

a. l l 8 e-l .

l l

3 COOLDOWP.) .

w RA TES . ACCEPT LE OF/HR OPERA"40N 750 /

v .

s 8

l 1

( sw 20 M w

100*y 250

/

t 0 # j !'

0 50 100 150 200 250 300 350. 400 450 500

. INDICATED TEMPER A E13EG.F)

Cumvt AretsCAsLA 8 COO N matis ur TO M POR TME esatt meal eg StewtCE Ptas00 TotoIPPY costTROLu Ttht4L.uust? 8 estf tnestblatt SHELL CO=Taises Man esof198P PLAtt saaes.4 ase0 P..a e PO =.ett COPPE A Costiter .4A7 wet Isott a ubst est amoms, catt C0=vt=t ...i atesof tsHTlat Af Ds0tAPTSAteEP gfp lqce u> ,e ta tief, tes*p va. .e.

pgrc 3. th 3 -2. 64ka cl'<4 Figure 3.4.3-2 RCS C00ldown Limitations Catawba Unit 2 3.4 11 5/20/ W

MA*ERIAL PROPERTY B ASIS LDerING MATERIALS: DiTERMEDIATE SHELL.B8605 2

LIMITING ART AT 15 EFPY: 1/&t 112.6 7 3/&t. M.0 7 2,500 ,

1 I

I 2,250 '

I i D I O 2,000 /

([) I O 1,750 ((((: M CCEPTABLE OPERATION /

m - , _ . ). .li '

l l L 1,500 O i U) 1 g

1,250 j

[ ACCEPTABLE OPERATION

. D. l y 1,000 /

@ i W cooLDOWN A (d 750 AATES 'F/ HA. N U :p : o w

/

1 - _ -_ __. e---

y _ -_ to m >

g. 500 _--_

40 9/ .__

((: so

__ /

__-- 100 250 0

4 0 50 100 150 200 250 300 350 400 450 500 I nd i cated Temperature C Deg . F) i Wy re 3. y . 3 - 2.

Fi&dRE B.; - CATA"/0A L""i ."2 ACTOR COOLANT SYS'IBi COOLDOWW LIMITA110NS (COOLDOWN RATES UP TO 100*F/HR) APPLICABLE FOR THE FIRST 15 EFPY (WTTHOI.TT MARGINS FOR INSTRUMENTATION ERRORS)

C ~.k & % :1- e.

3. t/-ll

LTOP System 3.4.12

(

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Low Temperature Overpressure Protection (LTOP) System LCO 3.4.12 '. An LTOP System shall be OPERABLE with a maximum of one charging pump or one safety injection pump capable of injecting into the RCS and the accumulators isolated and either a or b below,

a. Two power operated relief valves (PORVs) with lift setting s 50 psig, or

b. The RCS depressuri d and an RCS vent of a 4.5 square inches.

V APPLICABILITY: MODE 4 when any RCS cold leg temperature is s 285'F.

MODE 5 MODE 6 when the reactor vessel head is on.

............................N0TE------------------------- --

Accumulator isolation is only required when accumulator pressure is greater than or equal to the maximum RCS

( pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided in Specification 3.4.3.

( 900 p.s 3 (as lefi cal.bhd,),

cd\*u)cale klue- 6. 425"pse's Gts Le));

/

Catawba Unit 2 3.4-31 5/20/07

RCSP/TLi;its B 3.4.3 BASES ACTIONS C.1 and C.2 (continued)

Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluationlimits.

allowable of the effects of the excursion outside the Restoration alone per Recuired Action C.1 is insufficient because higher than analyzec stresses may have occurred and may have affected the RCPB integrity.

SURVEILLANCE SR 3.4.3.1

! REQUIREMENTS l

Verification that operation is within the specified limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room iitdication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes pemits assessment and

( correction for minor deviations within a reasonable time.

Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definitior, given in the relevant plant procedure for ending the activity is satisfied.

This SR is modified by a Note that only requires this SR to be perfomed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LC0 3.4.2 contains a more restrictive requirement.

REFERENCES 1. 10 CFR 50, Appendix G.

2.

ASME, Boiler Appendix G. and Pressure Vessel Code,Section III.

3.

ASTM E 185-82 J uif l982.

4. 10 CFR 50, Appendix H.

(continued)

Catawba Unit 2 B 3.4-16 J0f[

i LTOP System i

B 3.4.12

( BASES BACKGROUND With minimum coolant input capability, the ability to (continued) pmvide core coolant addition is restricted. The LC0 does V

not require the makeup control system deactivated or the i safety injection (SI) actuation circuits blocked. Due to

the lower pressures in the LTOP MODES and the expected core 1 decay heat levels, the makeup system can provide adequate '

flow via the makeup control valve. If conditions require the use of more than one charging pump for makeup in the event of loss of inventory, then pumps can be made available

, through manual actions.

4

] The LT0P System for pressure relief consists of two P0 ;

i with reduced lift settings or a d6 pressurized RCS and'$Vs an RCS 1 l

Vent of sufficient size. Two PORVS are required for redundancy. One PORY has adequate relieving capability to keep from overpressurization for the required coolant input j capability. //oopsc's (as lef f cal.hfeg')j s

sflotskble Vs.l4e h, if 25'pse'p l PORY Recuirements I" S b N )

i As designed for the LTOP System, each ORY is signaled to open if the RCS pressure reachesG'A ;;@when tlie PORYS are

( in the "lo-press' mode of operation. Tne LTOP actuation

' logic monitors both RCS temperatur2 and RCS pressure. The signals used to generate the pressure setpoints originate from the wide range pressure transmitters. The signals used 4

to generate the temperature pemissives originate from the wide range RTDs. Each signal is input to the appropriate NSSS protection system cabinet where it is converted to an j

internal signal and then input to a comparator to generate an actuation signal. If the indicated pressure meets or exceeds the calculated value, a PORY is signalted to open.

i t

This Specification presents the PORV setpoints for LTOP.

Having the setpoints of both valves within the limits i

ensures that the Reference 1 limits will not be exceeded in any analyzed event.

9 i

i (continued)

Catawba Unit 2 B 3.4-60 l --5/2^/a?

4

. . _ , . , _ . , . . _ _ ,_,____4 -_ . _

Attachment 2a Catawba Unit 1 Current Technical Specifications Remove Pages Insert Pages VII VII 3/4 4-32 (Figure 3.4-2) 3/4 4-32 (Figure 3,4-2) 3/4 4-33 (Figure 3.4-3) 3/4 4-33 (Figure 3.4-3) 3/4 4-34 (Table 4.4-5) 3/4 4-34 (Table 4.4-5) 3/4 4-36 3/4 4-36 B 3/4 4-9 B 3/4 4-9 B 3/4 4-14 B 3/4 4-14 B 3/4 4-15 B 3/4 4-15 i

LIMITINGCONDITIONSFOROPERATIONANDSURVE]LLANCERE0VIREMENTS 1I0110N EASI 3/4.4.3 PRESSURIZ ER . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-9 3/4.4.4 REllEF VALVES . . . . . . . . . . . . . . . . . . . . . . . 3/4 4 10 3/4.4.5 STEAM GENERATORS ..................... 3/4 4-12 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS 10 BE INSPECTED

, DURING INSCRVICE INSPECTION .............. 3/4 4-17 i

TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION ............ 3/4 4-18 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems . . . . . . . . . . . . . . . . . 3/4 4-19 Operational Leakage . . . . . . . . . . . . . . . . . . . . 3/4 4-20 TABLE 3.4 1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES . . . . 3/4 4-22 3/4.4.7 CHEMISTRY . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-24 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS ........ 3/4 4-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . 3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY . . . . . . . . . . . . . . . . . . . . . 3/4 4-27 FIGURE 3.4-1 DOSE EQUIVALENT l-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >l uCi/ gram DOSE EQUIVALENT l-131 . . . . . . . . . . . . . . . . . 3/4 4 28 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-29 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System .................. 3/4 4 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 15 EFPY ,.............. 3/4 4-32 I FIGURE 3.4-3 REACTOR COOLANT = SYSTEM C00LDOWN LIMITATIONS -

APPLICABLE UP TO 15 EFPY ............... 3/4 4-33 l CATAWBA - UNIT 1 Vil Amendment No. d

MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITNG AFE AT 15 EFPY: 1/4 T, 43*F 3/4-T, 26*F 2,500 l i. 1--14 i.iiii -

, ii

, ,! i i r i

i .

LEAK TEST LIMIT 3j l l l lll lil l l{ 1

,i l-2,250 .

; ;,1 ,,' ll i i i ;, L 1 iiii II i iii : i I; !i! i i i i ~1 1

ji r i i! , i> f !1 . i i ! ..

ii i ! I i ! / '! I !>i! ' i 2,000 '

l .: '; '; ' : ' i . ' ; ,' l l g

i i

i i !

i

. I

/ ! i II;

! ' !I

! i

! !i

' i !

i dI i

i m ii r ir :! - !

w' ! t -

O. 1,750 UNACCEPTABLE '/ '/ ,

OPERATION ,i 1 i

i i!

,, ,!!,ii , ,1i ,

@ i i iiii iiiI i / / ii *ii ! ! i !' ' i l' I' ' '

l

$ 1,500 A' '

'l .

/ !U !!

l g HEATUP RATE ACCEPTABLE } ,

,' i!

~

t/) UP TO 60*F/HR ii i N r l OPERATION I J !Ii o

' 1,250 . ... ,. .. ,!, /. . / .i.i i,,. !! 1 l ;; : :

,ii 1.i. i iu ii / .iii iiii ii i i ti4 i i: ,

b i iii

,,i i

ii iii/

i ! i/ >

i !t'

/! ii!i iie i i!

iiii

! i !

li ! ! '

i i!

i +

+

i

+

i!

i V 1,000 O i j,

... i i',

f i

~

!iii i,ii i

. i i i i !.

1 .i 48 , ,,

i i .i .

!i . ..i i,i !,i ,

i i !!

d5 ! ' - - ! ' ' i ! ' '

i ' i, I:

o-- 750 ,

i

. ; i i,,

. ;ii V

,- ,i i , , i < i ;ii, i ii ii; ;ii C i t

i; !

i 7'! ' ' '

ii' i! ' '

500 , ,., i; i 1 .

- -- a,

. 1 ii IC

! CRITICALITY LIMIT BASED ON W

! INSERVICE HYDROSTATIC TEST 250 :

'. H. . . TEMPERATURE (176*F) FOR THE !A-2 i; , ,

i ,

7, SERVICE PERIOD UP TO 15 EFPY

[4,:

0 O 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg. FD FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS (Heatup rates up to 60'F/hr)

APPLICABLE FOR THE FIRST 15 EFPY (Without Margins For Instrumentation Errors)

CATAWBA - UNIT 1 3/4 4 32 Amendment No.

MATERIAL PROPERTY BGIS LIMITING MATERIAL: LOWER SHEU. FORGIN0104 LIMITING ART AT 15 EFPY: 1/4 T, 43'F W4 T, 26*F 2,500 . , , , . .,

4 ;

4 i i ! !

I . ! l i i i!

2,250 I

ij i

'. .'l '

i _ i i i i- ! t i I !

! I .i !, __"I i iiI 2,000 '

./

Cd 2: UNACCEFABLE / !

i! !

'5 :: OPERATION i i il l j c.1,750 ,/ ,4

',l:' ,- , , '. . .

J s

@ l I ! 5!

B 1,500 i . /

(f) j ACCEMABLE *

(/) 1 OPERATION

@ 1,250 '

,' , , / 4 i ,, i , ,i ,, .

CL !ll I'i

.M I/' t i i i!

I i!

l' '

ll

!i l

V 1 ,000 '

l.,'

hm COOLDOWN

., RATES *F/HR.

!!i! !'

ii l!!'!'!

750 -M

.O-o !' '! ! '

20

'lI D -- t 40

! ! l !

c ::1 3o ! .,, , i i . i , i 500 ,' 10o .

i..'

1 ii,i ; ii i ! 4 i i i

! 4 .! ! ! ! !i i i ! l I. I ! !! I!!' I I . ! ! !

250 :

l ': ' '

,, , l

-l!!

. ,.., i

!!:i i! i! !

i !,i 1 .

i>!!i,i i i i i i

+

! !, ! Ii!i .iI . !!! !! TT_

g . . .. , , . ,

O 50 100 150 200 250 300 350 400 450 500 Indicated Temperature LDeg. F)

FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN Lij41TATIONS (Cooldown rates up to 100*F/nr)

APPLICABLE FOR THE FIRST 15 EFPY (Without Margins For Instrumentation Errors)

CATAWBA - UNIT 1 3/4 4-33 Amendment No. l

TA8LE 4.4-5 REACTOR VESSEL MTERIAL SURVEILLANCE PROGRAM - WITHDRAM4L SCHEDULE CAPSULE VESSEL- LEAD NUPSER LOCATION FACTOR WIThDRAMI TIPE JLQ"l U 58.5* 3.91 Standby V 61* 3.66 9 W 121.5* 3.91 13 X 238.5* 3.91 Standby Y 241* 3.66 4.98 Z 301.5* 4.10 0.79 CATAWBA - UNIT I 3/4 4-34 Amendment No.

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following Overpressure Protection Systems shall be OPERABLE:

a. Two power operated relief valves (PORVs) with a lift setting of less than or equal to 400 psig (as left calibrated), allowable value less than or equal to 425 psig (as found): or

b. The Reactor Coolant System depressurized with a Reactor Coolant System vent of greater than or equal to 4.5 square inches.

APPLICABILITY: MODE 4 when the temperature of any Reactor Coolant System cold leg is less than or equal to 285'F, MODE 5 and MODE 6 when the head is on the reactor vessel.

ACTION:

a. With one PORV inoperable in MODE 4. restore the inoperable PORV to OPERABLE status within 7 days or com)1ete depressurization and venting of the Reactor Coolant System throug1 at least a 4.5 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,

b. With one p0RV inoperable in MODES 5 or 6. restore the inoperable PORV to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or complete depressurization and venting of the Reactor Coolant System through at least a 4.5 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c. With both PORVs inoperable, coglete depressurization and venting of the Reactor Coolant System through at least a 4.5 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,

d. In the event either the PORVs or the Reactor Coolant System vent (s) are used to mitigate a Reactor Coolant System pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or Reactor Coolant System vent (s) on the transient, and any corrective action necessary to prevent recurrence,

e. The provisions of Specification 3.0.4 are not applicable.

CATAWBA - UNIT 1 3/44-36 Amendment No. l

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

Heatup and cooldown limit curves are calculated using the most limiting value of the nil ductility reference temperature, Rigny, at the end of the effective full power years (EFPY) of service life as indicated on the a)pli-l cable heatup or cooldown curves. The service life period is chosen suc1 that the limiting RTnfting unirradiated material.yThe RTNDT of the lim at selection the 1/4T of location such a_inlimiting the core region is greater Rigor assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RTNDit the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RTuoi. Therefore, an adjusted reference temperature, based upon the fluence, copper content, and phosphorus content of thematerialinquestion,canbepredictedusingFigureB3/4.4-1andthe largest value of ARTwoy.

The adjusted reference temperature has been computed using the guidance of Regulatory Guide 1.99, Revision 2. The heatup and cooldown limit curves in Figures 3.4-2 and 3.4-3 include predicted adjustments for the shift f in RTsor at the end of the identified service life.

Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10 CFR Part 50. Appendix H. The surveillance specNen withdrawal schedule is shown in Table 4.4-5. The lead factor represents the relationship between the fast neutron flux density at the location of the capsult. and the -

inner wall of the pressure vessel. Therefore, the results obtained from the surveillance specimens can be used to predict future radiation damage to the pressure vessel material by using the lead fcctor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ARTun7 determined from the surveillance capsule exceeds the calculated ART NDT for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Sec-tion III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in the following paragraphs.

CATAWBA - UNIT 1 B3/44-9

REACTOR COOLANT SYSTEM BASf5 PRESSURE / TEMPERATURE LlHITS (Continued) defect at the inside of the vessel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the Kin for the 1/4T crack during heatup is lower than the Kin for the 1/4T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive

thermal stresses and different Kgs for steady-state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. -

The second portion of the heatup analysis concerns the calculation of pressure temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, .

the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are de)endent on both the rate of heatup and the time (or coolant temperature) along tie heatup ramp.

Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.

Rather, each heatup rate of interest must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are pro-duced as follows. A composite curve is constructed based on a point-by-point ccmparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside co the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

CATAWBA - UNIT 1 B 3/4 4-14

, - ~ _ ,, n..

i REACTOR COOLANT SYSTEM BASES PRESSURE /TEMPERATVRE. Linili (Continued) j Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are

~

! provided to assure compatibility of operation with the fatigue analysis perfonned in accordance with the ASME Code requirements.

LOW TEMPERATURE OVERPRESSURE PROTECTION The OPERABILITY of two PORVs or a Reactor Coolant System vent opening of at least 4.5 square inches ensures that the Reactor Coolant System will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the cold legs are less than or equal to 285'F. Either PORV has adequate relieving ca) ability to protect the Reactor Coolant System from overpressurization when (1e transient is limited to either:

(1) the start of an idle reactor coolant 3 ump witi the secondary water temperature of the steam generator less t1an or equal to 50'F above the cold i legtemperatures,or(2)thestartofaSafetyinjectionpumpanditsinjection l into a water solid Reactor Coolant System.

Thc Maximum Allowed PORV Setpoint for the Low Tempe:ature Overpressure Protection System (LTOPS) is derived by analysis which models the perfonnance of the LTOPS assuming various mass input and heat input transients and incorporates instrument uncertainties as well as corrections for Reactor Coolant Pump operation and the static pressure difference between the Reactor Vessel Beltline Region and the location of the pressure transmitters used for LTOP. Operation with a PORV Setpoint less than or equal to the maximum allowable value of 425 psig (as found) ensures that Appendix G criteria will not be violated with consideration for a maximum pressure overshoot beyond the PORV Setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. To ensure that mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lockout of all but one Safety injection pump and all but one centrifugal charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of a RCP if secondary temperature is more than 50'F above primary temperature.

The Maximum Allowed PORV setpoint for the LTOPS will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule in Table 4.4-5.

1 CATAWBA - UNIT 1 B3/44-15

. ., y ._ . - , _.-.,_y- --

y---.

Attachment 2b Catawba Unit 2 Current Technical Specifications Remove Pages: Insert Pages:

VII VII 3/4 4-33 (Figure 3.4-2) 3/4 4-33 (Figure 3.4-2) 3/4 4-34 (Figure 3.4-3) 3/4 4-34 (Figure 3.4-3) 3/4 4-35 (Table 4.4-5) 3/4 4-35 (Table 4.4-5) 3/4 4-37 3/4 4-37 B 3/4 4-9 P 3/4 4-9 B 3/4 4-14 B 3/4 4-14 B 3/4 4-15 B 3/4 4-15 1

d

LIMITING CONulTIONS FOR OPERATION AND SURVElllANCF REQUIREMENTS SECTION EAq[

3/4.4.3 PRESSURIZER . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-9 3/4.4.4 RELIEF VALVES . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-10 3/4.4.5 STEAM GENERATORS ..................... 3/4 4-12 i

IABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING !NSERVICE INSPECTION .............. 3/4 4-17 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION ............ 3/4 4-18 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems . ............... 3/4 4-19 Operational Leakage . . . . . . . . . . . . . . . . . . . . 3/4 4-20 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES . . . . 3/4 4-22 b

3/4.4.7 CHEMISTRY . . . . . . . . . . . . . . . . . . . . . . . . .

l 3/4 4-24 i

l .

TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS ........ 3/4 4-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . 3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY , . . . . . . . . . . . . . . . . . . . . 3/4 4-27 FIGURE 3.4-1 DOSE EQUIVALENT l-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >l pCi/ gram DOSE EQUIVALENT l-131 . . . . . . . . . . . . . . . . . 3/4 4-28 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYS PROGRAM . . . . . . . . . . . . . . . . . . . . . . . IS ... 3/4 4-29 3/4.4.9 PRESSURE / TEMPERATURE LIMITS l

Reactor Coolant System ........,......... 3/4 4-31 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 15 EFPY ............... 3/4 4-32 l FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -

APPLICABLE UP TO 15 EFPY ,....._......... 3/4 4-33 l

CATAWBA - UNIT 2 VII Amendment No.

l

MATERIAL PROPERTY B ASIS LINTING MATDt1ALS: DUERhE'DIATE SHELL. B8605 2 LIhDTING ART AT 15 EFPY: 1/4 t,112.6 7 3/4 t 96.0 7 2,500 iiiiiiiiiiii, ,

LEAK TEST LIWlT - m '

l l H I I 2.,250 ' '

l r i i ,

n i l l C) 2.,000 i i

([) I I h 1.,750 UNACCEPTABLE OPERATION f f g i !

L 1.,500 l

/ / ;

~

- [ [ ACCEPTABLE 22-2 g I / OPERATION

~~"'-

@ 1.,250 HEATUP RATE N q j j

~~~~

g UP TO 60 'F/HR -7 r D- l l

.O

,000 /

@ l W '

(d 750 ,

O -

[ 500 _. ,

~

RITICALITY LIMIT ERSED ON 250 INSEWlCE WOROSTATlC TER TEM)ERATURE (245 *F) FOR THE SERVICE PERICO UP TO 15 EFPr 0 iiiIIIIiiiiiIiiIIiiiIII --

O 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg. F)

FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS (Heatup rates up to 60'F/hr)

APPLICABLE FOR THE FIRST 15 EFPY (Without Margins for Instrumentation Errors)

CATAWBA - UNIT 2 3/4 4-33 Amendment No. l 4 -

MA'115t!AL PROPERTY BASIS LIMrrING MATERIALS: Dm3tMEDIATE SHEll.B8605 2 LDETING ART AT 15 EFPY: 1/4 t 1124 7 34t 96.0 7 2,500 l

2,250 .

I

n i 1

CD 2,000 U) '

O- ----'

tm ccepTAete 1,750 y ----

n,,g47,gg j

@ h l L 1,500 / -

D W !

b 1'250 / ^ C C'"^

OPERATlDN y 1,000 /

G --- ; . i .____

H COOLDOWN d?

cd 750 MATES 'F/ HR, d7 u x:: ::2. u o .mp

} 500 : :N BD N /

--~~

1CO_

250 0

0 50 100 150 200 250 300 350 400 450 500 I ndicated Temperature (Deg . F)

FIGURE 3.4-3 REACTORCOOLANTSYSTEMC00LDOWNLIMITATIONS(Cooldownratesupto100'F/hr)

APPLICABLEFORTHEFIRST15EFPY(WithoutMarginsforInstrumentationErrors)

CATAWBA - UNIT 2 3/4 4-34 Amendment No. [

l TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE CAPSULE VESSEL LEAD NUMBER LOCATION FACTOR WITHDRAWN TIME fEFPY)

U 58.5* 3.79 Standby V 61* 3.62 9 W 121.5' 3.79 13.5 Y 238.5* 3.79 Standby X 241' 3.62 4.52 Z 301.5* 4.09 0.86 6

CATAWBA - UNIT 2 3/44-35 Amendment No.

1 i

! REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION i

! 3.4.9.3 At least one of the following Overpressure Protection Systems shall be

! OPERABLE:

a. Two power operated relief valves '(PORVs) with a lift setting of it:ss thanorequalto400psig(asleftcalibrated),allowablevalueless than or equal to 425 r*') (as found); or

b. The Reactor Coolant SystJA depressurized with a Reactor Coolant System vent of greater than or equal to 4.5 square inches.

APPLICABILITY: MODE 4 when the temperature of any Reactor Coolant System cold leg is less than or equal to 285'F, MODE 5 and MODE 6 when the head is.on the reactor vessel.

ACTION:

a. With one PORV inoperable in MODE 4, restore the inoperable PORV to OPERABLE status within 7 days or com)lete depressurization and venting of the Reactor Coolant System throug1 at least a 4.5 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,

b. With one PORV inoperable in MODES 5 or 6 restore the inoperable PORV to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or complete depressurization and venting of the Reactor Coolant System through at hast a 4.5 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,

c. With both PORVs inoperable, complete depressurization and venting of the Reactor Coolant System through at least a 4.5 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ..

d. IntheeventeitherthePORVsortheReactorCoolantSystemvent(s) are used to mitigate a Reactor Coolant System pressure transient, a Special Report shall _be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or Reactor Coolant System vent (s) on the transient, and any corrective action necessary to prevent recurrence.

e. The provisions of Specification 3.0.4 are not applicable.

CATAWBA - UNIT 2 3/44-37 Amendment No.

_ ~ _ _ _..._. _ _ _ . _ . . _ _ . . _ _ _ _ _ _ _ . _ _ . . _ _ _ _ . _ . _ _ _ .

REACTOR COOLANT SYSTEM L

i BASES i

PRESSURE / TEMPERATURE LIMITS (Continued)

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTnor, at the end of the effective full power years (EFPY) of service life as indicated on the a)pli-cable heatup or cooldown curves. The service life period ir chosen suc1 that

, the limiting RTupy at the 1/4T location in the core region greater than the RTNDT of the limiting unirradiated material. The selectior 4f such a limiting l RTuoy assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

i The reactor vessel materials have been tested to determine their initial RTNOT; the results of these tests are shown-in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increise in the RTuor. Therefore, an adjusted reference .

temperature, based upon the fluence, copper content, and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ARTNOT-l The adjusted reference temperature has been computed using the guidance of Regulatory Guide 1.99, Revision 2. The heatup and cooldown limit i

curves in Figures 3.4-2 and 3.4-3 include predicted adjustments for the shift i in RT NOT at the end of the identified service life.

i

Capsules will be removed in accordance with the requirements of ASTM E185-i 82 and 10 CFR Part 50, Appendix H. The surveillance specimen withdrawal

schedule is shown in Table 4.4-5. The lead factor represents the relationship i between the fast neutron flux density at the location of the capsule and the inner wall of the pressure vessel. Therefore, the results obtained from the surveillance specimens can be used to predict future radiation damage to the pressure vessel material by using the lead factor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the

. ARTnor detennined from the surveillance capsule exceeds the calculated ARTNOT for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and

cooldown rates are calculated using methods derived from Appendix G in Sec-tion III of the ASME Boiler and Pressure Vessel Code as required by Appendix G

to 10 CFR Part 50, and these methods are discussed in detail in the following l paragraphs, i

4 CATAWBA - UNIT 2 B 3/4 4-9

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) defect at the inside of the vessel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the Km for the 1/4T crack during heatup is lower than the Km for the 1/4T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different Kg's i for steady-state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to atsure that et any coolant temperatura the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. -

l The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, .

the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are de)endent on both the rate of heatup and the time (or coolant temperature) along tie heatup ramp.

Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.

Rather, each heatup rate of interest must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are pro-duced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

The us: of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

CATAWBA - bNIT 2 B 3/4 4-14 r .

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

LOW TEMPERATURE OVERPRESSURE PROTECTION The OPERABILITY of two PORVs or a Reactor Coolant System vent opening of at least 4.5 square inches ensures that the Reactor Coolant System will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the cold legs are less than or equal to 285'F. Either PORV has adequate relieving capability to protect the Reactor Coolant System from overpressurization when the transient is limited to either:

(1) the start of an idle reactor coolant pump with the secondary water temperature of the steam generator less than or equal to 50'F abor the cold leg temperatures, or (2) the start of a Safety Injection pump and ics injection into a water solid Reactor Coolant System.

The Maximum Allowed PORV Setpoint for the Low Temperature Overpressure Protection System (LTOPS) is derived by analysis which models the perfonnance of the LTOPS assuming various mass input and heat input transients and incorporates instrument uncertainties as well as corrections for Reactor Coolant Pump operation and the static pressure difference between the Reactor Vessel Beltline Region and the location of the pressure transmitters used for LTOP. Operation with a PORV Setpoint less than or equal to the maximum allowable value of 425 psig (as found) ensures that Appendix G criteria will not be violated with consideration for a maximum pressure overshoot beyond the PORV Setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. To ensure that mass and heat input transients more severe than those assumed cannot occur Technical Specifications require lockout of all but one Safety Injection pump and all but oae centrifugal charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of a RCP if secondary temperature is more than 50 F above primary temperature.

The Maximum Allowed PORV setpoint for the LTOPS will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule in Table 4.4-5.

CATAWBA - UNIT 2 B3/44-15

Attachment 2c Catawba Unit 1 Improved Technical Specifications Remove Pages: Insert Pages:

3.4-10 (Figure 3.4.3-1) 3.4-10 (Figure 3.4.3-1) 3.4-11 (Figure 3.4.3-2) 3.4-11 (Figure 3.4.3-2) 3.4-31 3.4-31 B 3.4-16 B 3.4-16 l B 3.4-60 B 3.4-60

MATERIAL PROPERTY BASIS RCSP/TLimits- ,

3.4.3 l l

UMfTING MATERIAL: LOWER SHELL FORGING 04 l UMITING ART AT 15 EFPY: 1/4 T, 43*F l 3/4 T, 26*F l

2,500 r i i . . , , i i . , , i , , , , i ,,, . ...

I i 1 , 4,! ,.i .

LEAK TEST UMIT - f I ' ' '

s sa ,.

i r' ii , . 4

i

,, l 2,250 f' l ';

l I I I I I t4 t

I I i , '

i f I 4 46 4

It ii

_ 2,000 ; , ,

l' ;.': ,'

CD e ===

i < i i i 1 I v3 ____ r r

o. 1,750 UNACCEPTABLE / /

-'::: OPERATION i i ,

O i I y) 1,500 , , , , , ,

(t)

HEATUP RATE ' %

UP TO 60 F/HR

/ r

/ ACCEPTABLE ,

OPERATION 2 1,250 '

/ / .

D l l

> s ,

t(D 1,000 , ,. l /, , ;, , ', '

i < i ,

M ,

m ,, '

i

.9 750 l'l

D l l l l l ll, C ><<

si s,+

+ s , s.* s 500 '

i,' , l ll '

Q 'l'. !:'i !!!! '

!!! ! ! l'

,l,' 'l

l -chil1CAUTY UMIT BASED ON ::::

250

l,'l

;jl INSERVICE HYDROSTATIC TEST ----

l l l TEMPERATURE (176*F) FOR THE ----

i,ii ,iii . ----

! SERVICE PERIOD UP TO 15 EPPY g :. . ::/: : ::l :

O 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg. FD Heatup rates up to 60*F/hr applicable for the first 15 EFPY (without margins for instrumentation errors)

Figure 3.4.3-1 RCS Heatup Limitations Catawba Unit 1 3.4-10 7/2/97

RCS P/T Limits MATERIAL PROPERTY BASIS 3.4.3 UMITING MATERIAL LOWER SHELL FORGING 04 UMITING ART AT 15 EFPY: 1/4 T 43*F 3/4-T. 26*F 2,500 , ,

, . . . . . . . . . . =

, i ... i , , i i 6 , , , , , . . .

i i

a J , , i i i , . .

c

' f f , { , t t i . . , e i i F 2,250 , ,, , l. ,! ': , ' ' ' ' '

i

,' :, ,E, i ,. ,

c ., ,

. , : . l ,

, , ,, . ,c .

, , , i , . . ,

, , i , , . 6 ., ,

i i ,. ( l ,

' 6 6 ,n . , ..

' ' ' i . . i 2,000 '

/

C)

'"- , , : l ,

' ' :-'~

UNACCEPTABLE 8

5 -"-

OPERATION / i ,' l C. 1,750 , ,' ' '

, / '

,l i

i t i e i is tii i <

Q) i i i i i i ,i ,

i

] 1,500 '

,' /

iii, i

g) ' 't .

y) l

, i

/ ACCEPTABLE 7 OPERATION 2 1,250 , :' :',, l,/. ..

, , i ., i ,, ,,

ii .

, , i i, i /,t ii i i , i is iti . ,,

g 1,000 .. . .i i . .

4+

. i . .i O COOLDOWN .. .

i, i i i+

RATES F/HR. ,

,' l l '! ,l ll,l 5 '+i6 l,l i

o 750 5 ,

. o 20

..i.

.l.l .

ii

'l .

l l ,

i

, i C

40 '

, ,ii .i. , , i i

- 60 ' ' ' '

500 '

co : .'. ::'. ll i.e .

, ,.. .i ,

l' i

i,.. ..# , ) ,

.. t

, ,, i 250

. : ll:, l:,, . , , ,

, . . , ,.. ,i , . ,

, ii . . ,a e

. , i) , ii s ,

. ... i , ,

0 ' ' ' ' ' ' ' '

O 50 100 150 200 250 300 350 400 450 500 Indicated Temperature KDeg. F)

Cooldown rates up to 100'F/hr applicable for the first 15 EFPY (without margins for instrumentation errors)

Figure 3.4.3-2 RCS Cooldown Limitations Catawba Unit 1 3.4-11 7/2/97 j

j

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

, LTOP System l 3.4.12 l 3.4 REACTOR COOLANT SYSTEM (RCS)

) 3.4.12 Low Temperature Overpressure Protection (LTOP) System l

LCO 3.4.12 An LTOP System shall be OPERABLE with a maximum of one charging pump or one safety injection pump capable of l injecting into the RCS and the accumulators isolated and either a or b below.

4 --

l a. Two power operated relief valves (PORVs) with lift j

' setting s 400 psig (as left calibrated), allowable value s 425 psig (as found); or i

j b. The RCS depressurized and an RCS vent of 2 4.5 square

inches.

i APPLICABILITY: MODE 4 when any RCS cold leg temperature is s 285'F, i MODE 5,

i. MODE 6 when the reactor vessel head is on.

] ______...

_______...._____.-N0TE----------------------------

3 Accumulator isolation is only required when accumulator

! pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed i by the P/T limit curves provided in Specification 3.4.3.

i 4

Catawba Unit 1 3.4-31 8/4/97

RCSP/TLimits B 3.4.3 BASES ACTIONS C.1 and C.2 (continued)

Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS l Verification that operation is within the specified limits

' is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits assessment and correction for minor deviations within a reasonable time.

Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.

This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LC0 3.4.2 contains a more restrictive requirement.

REFERENCES 1. 10 CFR 50, Appendix G.

2. ASME, Boiler and Pressure Vessel Code,Section III.

Appendix G.

3. ASTM E 185-73, 1973.

4. 10 CFR 50, Appendix H.

(continued)

Catawba Unit 1 B 3.4-16 9/9/97 1

1

_. .._ _ ___ __ ..________________________m. .___ _ .

a LTOP System

8 3.4.12

BASES 1-

BACKGROUND With minimum coolant input capability, the ability to

! (continued) provide core coolant addition is restricted. The LC0 does

not require the makeu

! safety injection (SI)p control system actuation deactivated circuits or theto blocked. Due i the lower pressures in the LTOP MODES and the expected core decay heat levels, the makeup system can provide adequate flow via the makeup control valve. If conditions require the use of more than one charging pump for makeup in the i event of loss of inventory, then pumps can be made available i through manual actions.

1- The LTOP System for pressure relief consists of two PORVs

, with reduced lift settings or a depressurized RCS and an RCS vent of sufficient size.

Two PORVS are required for redundancy. One PORV has adequate relieving capability to keep from overpressurization for the required coolant input capability.

PORV Reauirements e

i As. designed for the LTOP System, each PORV is signaled to j open if the RCS pressure reaches 400 psig (as left

! calibrated), allowable value s 425 psig (as found), when the

PORVS are in the "lo-press" mode of operation. The LTOP actuation logic monitors both RCS temperature and RCS i pressure. The signals used to generate the pressure setpoints originate from the wide range pressure

transmitters. The signals used to generate the temperature 1

pemissives originate from the wide range RTDs Each signal j is input to the appropriate NSSS protection system cabinet where it is converted to an internal signal and then input to a comparator to generate an actuation signal. If the i indicated pressure meets or exceeds the calculated value, a PORY is signaled to open.

This Specification presents the PORV setpoints for LTOP.

s Having the setpoints of both valves within the limits ensures that the Reference 1 limits will not be exceeded-in

any analyzed event.

I 4

(continued)

Catawba Unit 1 B 3.4-60 8/4/97

Attachment 2d

, Catawba Unit 2 Improved Technical Specifications Remove Pages: Insert Pages:

3.4-10 (Figure 3.4.3-1) 3.4-10 (Figure 3.4.3-1) 3.4-11 (Figure 3.4.3-2) 3.4-11 (Figure 3.4.3-2) 3.4-31 3.4-31 B 3.4-16 B 3.4-16 B 3.4-60 B 3.4-60

RCSP/TLimits MATERIAL PROPERTY BASIS 3'4'3 4

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i Catawba Unit 2 3.4-10 7/2/97

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MATERIAL PROPERTY BASIS RCSP/TLimits 3.4.3 LIMITING MA1 TRIALS: INIERMEDIATE SHELL. B8605 2 l LIMITING ART AT 15 EFPY: 1/4-t.112.6 T 3/4-t. M.0 7

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4 l

LTOP System 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Low Temperature Overpressure Protection (LTOP) System LCO 3.4.12 An LTOP System shall be OPERABLE with a maximum of one charging pump or one safety injection pump capable of injecting into the RCS and the accumulators isolated and either a or b below,

a. Two power operated relief valves (PORVs) with lift setting s 400 psig (as left calibrated), allowable value s 425 psig (as found); or

b. The RCS depressurized and an RCS vent of a 4.5 square inches.

APPLICABILITY: MODE 4 when any RCS cold leg temperature is s 285'F, MODE 5, MODE 6 when the reactor vessel head is on.

NOTE----------------------------

Accumulator isolation is only required when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided in Specification 3.4.3.

Catawba Unit 2 3.4-31 8/4/97

'RCSP/TLimits B 3.4.3 0

BASES ACTIONS C.1 and C.2 (continued)

Condition C is modified by a F)ie requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the specified limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits assessment and correction for minor deviations within a reasonable time.

Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.

This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirenent.

REFERENCES 1. 10 CFR 50, Appendix G.

2. ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.

3. ASTM E 185-82, 1982.

4. 10 CFR 50, Appendix H.

(continued)

Catawba Unit 2 B 3.4-16 9/9/97 4 _

LTOP System B 3.4.12 BASES BACKGROUND With minimum coolant input capability, the ability to (continued) provide core coolant addition is restricted. The LCO does not require the makeup cnntrol system deactivated or the safety injection (SI) actuation circuits blocked. Due to the lower pressures in the LTOP MODES and the expected core decay heat levels, the makeup system can provide adequate flow via the makeup control valve. If conditions require i the use of more than one charging pump for makeup in the event of loss of inventory, then pumps can be made available through manual actions.

The LTOP System for pressure relief consists of two PORVs with reduced lift settings or a depressurizeu RCS and an RCS vent of sufficient size. Two PORVS are required for I redundancy. One PORV has adequate relieving capability to l

keep from overpressurization for the required coolant input capability.

PORV Reautrements.

As designed for the LTOP System, each FORV is signaled to open if the RCS pressure reaches 400 psig (as left calibrated), allowable value s 425 psig (as found), when the PORVS are in the "lo-press" mode of operation. The LTOP actuation logic monitors both RCS temperature and RCS pressure. The signals used to generate the pressure setpoints originate from the wide range pressure transmitters. The signals used to generate the temperature permissives originate from the wide range RTDs. Each signal is input to the appropriate NSSS protection system cabinet where it is converted to an internal signal and then input to a comparator to generate an actuation signal. If the indicated pressure meets or exceeds _the calculated value,- a PORV is signaled to open.

This Specification presents the PORV setpoints for LTOP.

Having the setpoints of both valves within the limits ensures that the Reference 1 limits will not be exceeded in any analyzed event.

(continued)

Catawba Unit 2 B 3.4-60 8/4/97

Attachment 3 Description of Proposed Changes and Technical Justification Introduction

, This proposed amendment will change the following Catawba Nuclear Station Units 1 and 2 Technical Specifications:

A. Figures 3.4-2 and 3.4-3 (current Technical Specifications), and Figures 3.4.3-1 and 3.4.3-2 (Improved Technical Specifications)- new Reactor Coolant 3 Heatup and Cooldown Limitations Curves for both units; B. Table 4.4-5 " Reactor Vessel Material Surveillance Program Withdrawal Schedule"- applies only to the current Technical Specifications, since the Improved Technical Specifications move this to the UFSAR; C. Technical Specification 3.4.9.3a

Overpressure Protection Systems" (current Technical Specifications) and 3.4.12a " Low Temperature Overpressure Protection (LTOP) System" (Improved Technical Specifications).

The Technical Specifications changes contained in this proposed amendment have been grouped as follows:

1. Revise the Pressure / Temperature curves to 15 EFPY, including the incorporation of the latest radiation surveillance capsule results, and removal of instrument margin and material footnotes from the Technical Specifications figures.

2. Modify the actual capsule ID listed on Table 4.4-5 " Reactor Vessel Material-Surveillance Program- Withdrawal Schedule" (for Unit 2 only) and update each unit's Lead Factors and Withdrawal Time. This proposed change applies only to the current Catawba Technical Specifications.

3. Revise the Technical Specifications requirement for the Reactor Coolant System (RCS) Overpressure Protection System during low temperature conditions.

4. Change format and enhance consistency.

Attachment 3 Description of Proposed Changes and Technical Justification Discussion The following is a discussion of the significant changes associated with each of the identified groups:

1. Revise the Pressure / Temperature curves to 15 EFPY, including the incorporation of the latest radiation surveillance capsule results, and removal of instrument margin and material footnotes from ths Technical specifications

! figures.

This change updates the Pressure / Temperature curves (heatup and cooldown curves) for bor', units. The service period for these curves has been expanced from 10 EFPY to 15 EFPY. The new pressure / temperature limits satisfy all required material embrittlement considerations including: 10CFR50, Appendix G; Regulatory Guide 1.99, Revision 2; and ASME Section III, Appendix G. The development of these curves was performed by Westinghouse Electric Corporation and was included in the appendices of the Westinghouse surveillance capsule reports WCAP-13720 and WCAP-13875. These surveillance capsule reports were submitted to the NRC for review by letters dated August 12, 1993 and March 2, 1994, This proposed amendment adopts the heatup and cooldown curves contained in WCAP-13720 and WCAP-13875 for use in the Catawba Technical Specifications. The new curves have changed when compared to the curves presently in the Technical Specifications. As shown on the curves and discussed in the WCAPs, there has been a chift in the curves and the limiting material has changed. The shift in the curves is attributed to: 1) the removal of the instrumentation uncertainty margin, 2) the use of unit specific capsule analyses (the present Unit 1 curves are based on Unit 2 capsule analysis as documented in Duke letter dated April 19, 1989 and NRC letter dated March 28, 1990), and 3) the vessel irradiation that has occurred over time. The shift of the curves for Unit 1 is in the direction of less restrictive operation, since the new curves are based upon a Unit 1 specific capsule (Capsule Y). For Unit 2, the curves have shifted in the direction of more restrictive operation.

Currently Catawba Unit 1 has experienced 9.18 EFPY and Unit 2 has 8.32 EFPY (through 9/9/97). It is pro]ected that 10 EFPY will be reached for Catawba Unit 1 during fuel cycle 11 2

h Attachment 3 Description of Preposed Changes and Tect.aical Justification in 1998, and for Catawba Unit 2 during fuel cycle 10 in 1999.

The revised Pressure / Temperature limits have been evaluated with respect to the current LTOP setpoints. This evaluation demonstrated that the 285 F enable temperature used for both units remains conservative relative to the RTnm + 90 F enable temperature criterion.

i The layout of Figures 3.4-2 and 3.4-3 (current Technical Specifications) and Figures 3.4.3-1 and 3.4.3-2 (Improved Technical Specifications) is revised by the proposed changes. Also, the material footnotes (for copper and nickel) are not shown on the new figures since the new curves from referenced WCAPs do not show this information.

The material footnotes are not considered to be necessary for use of the curves during normal day-to-day plant operations; however, this information is readily available for reference in engineering documents. These figures are the heatup and cooldown limitations curves for both units.

The key information that is found currently within each figuro continues to be provided in the new figures, however it is arranged differently. The changes discussed in this paragraph are considered administrative, in that the praposed changes do not involve any significant technical information, just the manner in which the information is plotted or shown on the page.

The WCAP-13720 and WCAP-13875 Heatup and Cooldown Data are provided directly on Pages 1, 2, and 3 of Enclonure 3a which are provided at the end of this attachment.

The existing Technical Specifications heatup and cooldown curves contain margins of 60 psig and 10 F to account for possible instrument uncertainty The assumed allowable margins of 60 psig and 10 F are consistent with the standard Westinghouse uncertainties values. Analysis of actual Catawba instrument loop uncertainties revealed that the 60 psig and 10 F is bounding. The actual Catawba nargins, to account for instrument uncertainties associated with this proposed amendment, are 51.8 psig and 7.1 F. These values are documented in a Duke engineering calculation. For additional conservatism, instrument uncertainties of 55 psig and 10oF are used in the Duke LTOP engineering calculation.

3

)

Attachment 3 Description of Proposed Changes and Technical Justification Although the margins assumed for the Catawba pressure and temperature instrument uncertainties have been changed, there were no changes or modifications made to the temperature and pressure instruments, or how they are calibrated, maintained, or used. The changes in the temperature and pressure instrument uncertainties accommodate the different raethodologies used between Duke and Westinghouse in calculating these uncertainties, while continuing to provide sufficient margin. Westinghouse used a generic value for all their plants, while Duke used the values in the industry standard ISA-RPC7.04 Part II (International Society for Measurement and Control, Nethodologies for the Determination of Setpoints for Nuclear Safety Related Instrumentation, Approval 1994) and ISA-S67.04 Part I (Setpoints for Nuclear Safety Related Instrumentation, Dated 1994).

This proposed amendment includes a request to relocate the instrument uncertainty margins from the Catawba Technical Specifications heatup and cooldown curves to other Duke documents. This change is being proposed to provide more owner control of the instrumentation. This would permit future modifications to install replacement instrumentation (if needed), or implementation of changes to the calibration criteria (if needed) without requiring a Technical Specifications amendment. Any future developments of this type could then be processed in accordance with 10CFR50.59.

Following approval of this amendment, the instrument uncertainty margins will be administratively implemented by incorporating them into the controlling procedures for unit operations and into the LTOP system setpoint selection calculations. These documents also contain various other adjustments that are based upon Duke calculations, such as-the pressure correction due to the pressure transmitters being at a different elevation than the reactor vessel beltline region, and reactor coolant pump induced pressure effects. Operational restrictions originating from Duke calculations ensure that the Pressure / Temperature limits would not be challenged.

The relocation of the instrument uncertainty to licensee controlled documents is consistent with NUREG-1431, the new Improved Standard Technical Specifications for Westinghouse plants. As prescribed within NUREG-1431, the heatup and 4

Attachment 3 Description of Proposed Changes and Technical Justification cooldown curves could be completely relocated to a licensee controlled document entitled " Pressure Temperature Limits Report (PTLR)." Changes to the heatup and cooldown curves (including instrument uncertainty) could then be performed in accordance with 10CFR50.59. However, Duke is proposing to remove only the instrumentation uncertainty from the Technical Specifications at this time, not the actual heatup and cooldown curves. This proposed change is consistent with accepted NRC practices and is even more conservative than the provisions of NUREG-1431, since the actual heatup and cooldown curves are remaining in the Catawba Technical Specifications.

Accordingly, the proposed new heatup and cooldown curves provided within this submittal do not include any margin to account for instrument uncertainties.

2. Modify the actual capsule ID listed on Table 4.4-5 " Reactor vessel Material surveillance Program - Withdrawal schedule" (for Unit 2 only) and update each units' Lead Factors and Withdrawal Time. This proposed ceange applies to the current Technical specifications only.

The present Surveillance Program - Nithdrawal Schedule required capsule Y to be removed from vessel location 241 at 5 EFPY. However, on Unit 2 the capsule in this location was determined to be capsule X not capsule Y. This situation has been documented in WCAP-1387E and in Duke PIP 2-C94-0105 (the Duke corrective action program). The important parameter in the surveillance capsule program (WCAP-10868

" Duke Power Company Catawba Unit No. 2 Reactor Vessel Radiation Surveillance Program") is vessel lo"ation, not the actual ID number affixed on the surveillance capsule.

Therefore, it is requested that the capsule ids be modified to indicate capsule X was installed in capsule Y location and capsule Y was installed in capsule X location.

WCAP-10868 " Duke Power Company Catawba Unit No.2 Reactor Vessel Radiation Surveillance Program" has been reviewed.

Each of the six material test capsules contains like specimens from the reactor vessel shell plate, representative weld metal and heat affected-zone metal. The interchange of capsule X with capsule Y at the time of construction of the plant would have no effect on the results.

5

Attachment 3 Description of Proposed Changes and Technical Justification As mentioned earlier, the important parameter in the surveillance capsule program is the reactor vessel location.

In fact, a review of WCAP-13875 Section 7.0 shows that Westinghouse recommendations for future capsule withdrawal schedules drop the capsule ID altogether and rely cnly on reactor vessel location as a means for capsule designation.

Nonetheless, the changes proposed in this amendment retain the capsule ID within the current Catawba Technical Specifications.

WCAP-9734 " Duke Power Company Catawba Unit No.1 Reactor Vessel Radiation Surveillance Program" and WCAP-10868 " Duke Power Company Catawba Unit No.2 Reactor Vessel Radiation Surveillance Program" are included as attachments within this submittal package.

Also, this change requests the Lead Factors be updated for each unit per the latest surveillance capsule results (WCAP-13720 Table 6-19 for Unit 1 and WCAP-13875 Tyble 6-17 for Unit 2). These WCAP tables are provided directly on Pages 1 and 2 of Enclosure 3b which is provided at the end of this attachment.

This proposed amendment also requests to update the withdrawal times (removal times as used in the referenced WCAPc). The withdrawal times do differ from WCAP-13720 Section 7.0 (Unit 1) and WCAP-13875 Table 7-1 (Unit 2). This is because Westinghouse projected out to 32 EFPY and 48 EFPY for the normal and extended life of the plant (40 and 60 years respectively). The proposed amendment projects out to 34 EFPY (capsule V) for the expected normal 40-year life of the plant and to 51 EFPY (capsule W) for 60-year life (extension) of the plant. These proposed times account for Catawba's higher capacity factor during the expected service life of the plant. Specifically assumed in these projected numbers are:

20 year life extension beyond current operating license

45 day outage length

95.5% unit reactor capacity factor (Excluding refueling outage) 6

Attachment 3 Description of Proposed Changes and Technical Justification The WCAPs Section 7.0/ Table 7.1 are provided directly on Pages 1 and 2 of Enclosure 3c which is provided at the end of this attachment.

These new removal schedules are proposed in order to better meet ASTM E-185-73/ ASTM E-185-82 (Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels).

3. Revise the Technical Specifications for the Reactor Coolant System (RCS) Overpressure Protection System during low temperature conditions.

This change reduces the setpoint of the PORV during low temperature conditions (RCS cold leg temperature is less than or equal to 285 F) as required in Technical Specifications 3.4.9.3a and 3.4.12a (for the current and Improved Technical Specifications, respectively). Presently, Catawba Nuclear Station conservatively calibrates the Pressurizer PORVs to 400 psig instead of the 450 psig as allowed by the current Technical Specifications.

Consequently, implementation of this proposed change will have no impact on the present operational practices at Catawba. The maximum allowable value (as found) is also being included to be consistent with other areas of the Technical Specifications (specifically Table 3.3-4).

The Technical Specifications currently state that the PORV lift setting be less than or equal to 450 psig. The proposed Technical Specifications amendment will lower the PORV lift setpoint to less than or equal to 400 psig. The proposed change specifies a more conservative setpoint, in that the PORV will open earlier during an LTOP event. This will lower the peak pressure resulting from an overpressure event.

The licensing basis for overpressure protection during low temperature, water solid modes of operation is to ensure that the RCS will be protected from pressure transients which could exceid the limits established per Appendix G of 10CFR50. This is accomplished by limiting the peak pressure as a result of a LTOP event below the pressure / temperature limits of the heatup and cooldown curves for the unit.

7

Attachmcat 3 Description of Proposed Changes and Technical Justification I. One of the proposed Technical Specifications changes provided by this submittal is the updating of the heatup and cooldown curves for both units. As such, en analysis was performed to verify the setpoint for the PORVs would be

acceptable in preventing a violation of the proposed heat up j

i and cooldown curves for both units. The analysis evaluated three possible transients. The methodology, as outlined in the Westinghouse Reports, utilized in determining the initial PORV setpoint, was also used in the present i analysis. The three possible pressure transients that were evaluated in the present analysis are:

E

1) a mass input from a safety injection pump.

4 l 2) a mass input from a centrifugal charging pump.

3) a heat input from a 50 F temperature difference between

. the steam generators and the rest of the RCS.

i For all three pressure transient cases, the following criteria and assumptions are applicable:

~

1) The resulting PORV setpoint must be low enough to mitigate the consequences of the defined mass and heat input transients without violating the Appendix G limits. The PORV setpoint for LTOP acceptance must be set such that the peak reactor vessel beltline pressure, including instrument uncertainty, is no more than the

} ASME Section III, Appendix G limits.

l j 2) The pressurizer is water solid and the pressure drop across the vessel is the same at hot or cold

. temperatures.

3) The assumed maximum instrument loop uncertainty for the RCS pressure transmitters is 55.0 psig.

4) The RCS pressure overshoot is calculated using the methodology given by the Westinghouse Report, " Pressure

, Mitigating Systems Transient Analysis Results", dated

, July 1977 and supplemented September 1977.

5) The difference between the indicated pressure (the signal actuating the PORVs) and the actual reactor vessel belt 2ine pressure is calculated. This includes 8

i i

4 i

Attachment 3 Description of Proposed Changes and Technical Justification elevation differences between the reactor vessel beltline and the reactor coolant pressure transmitters, and the differential pressure across the reactor core due to hydraulic losses (the transmitters actuating the PORVs are on the hot legs, while the RV beltline is on the cold leg side of the core). The initial analysis performed during the licensing phase did not include these correction factors.

6) The ASME Section III, Appendix G heat up and cool down limits for both units were calculated by Westinghouse.

The information in support of generating these curves is provided within Westinghouse surveillance capsule reports WCAP-13720 for Unit 1 and WCAP-13875 for r: nit 2.

These reports were provided to the NRC for review by letters dated August 12, 1993 and March 2, 1994.

The allowable value of 425 psig incorporates the setpoint of 400 psig and adjusts it for uncertainties such as reference accuracy, calculation effect, and instrument rack drift. The development of these adjustment values is consistent with development methodology for allowable values on other portions of the Technical Specifications. These adjustments have been incorporated into the maximum instrumentation loop uncertainty of 51.8 psig, which in turn has been used in the analysis to ensure that the peak reactor vessel beltline pressure is not exceeded in an LTOP event.

In conclusion, the analysis that was performed verified that a PORV setpoin; of 400 psig, with an allowable value of 425 psig, is adequate to ensure that the peak reactor vessel beltline pressure, (including instrument uncertainties, pressure corrections for the differences between the indicated pressure and the actual reactor vessel beltline pressure, and the pressure corrections for differential pressure across the reactor core due to hydraulic losses),

is less than the ASME Section III, Appendix G limits during anticipated pressure transients, provided appropriate limits on the heatup and cooldown rates are established. For both units the limiting pressure transient is the mass input transient, resulting from the inadvertent start of a safety injection pump.

Currently, Catawba's LTOP instrument calibration procedures and operating procedures utilize 400 psig as the LTOP PORV 9

i j

Attechment 3 Description of Proposed Changes and Technical Justification setpoint. It is conservative to change the setpoint in Technical Specifications 3.4.9.3a and 3.4.12a as proposed in this amendment. Additionally, this proposed change w.i.ll not have a significant impact on the operation of the plant, since Catawba's current operating practice uses the 400 psig setpoint.

4. Change format and enhance consistency. '

other changes have been made to improve consistency between Technical Specifications. These changes are considered editorial in nature and do not affect the operation of the units or the safety functions performed by the LTOP system, j

Specifically, the administrative and editorial changes provided by this amendment request are:

a) an updating of page VII of the Technical Specification Index; b) an updating of the Bases for Technical Specifications 3/4.4.9 (Pressure / Temperature Limits) and 3.4.12 (Low Temperature Overpressure Protection System) providing more information and detail. The updated Bases are included for informational purposes only.

Impact on the Updated Final Safety Analysis Report The Reactor Vessel is discussed in Section 5.3 of the Catawba UFSAR. The reactor material surveillance program is discussed in Section 5.3.1.6, and the reactor vessel pressure-temperature limits are discussed in Section 5.3.2.

No changes to either of these sections are required as a result of this amendment.

Section 5.3.4 of the Catawba UFSAR will be revised to include WCAPs 13720 and 13875 in the list of references.

Additionally, for the Improved Technical Specifications, the

" Reactor Vessel Material Surveillance Program Withdrawal Schedule" will be relocated to the UFSAR. The necessary changes to these two sections will be made in accordance with 10CFR50.71(e).

10

Enclosure 3a DATA POINTS FOR HEATUP AND COOLDOWN CURVES WITHOUT MARGINS From WCAP-13720 for Catawba Unit 1 Semedy Siens 20CD 40 CD 40CD 100 CD T F T P T 85 62120 P T P T P 35 62120 85 62120 85 90 62120 90 62120 62120 85 62140 i 90 62120 90 62120 95 42120 95 62120 95 62120 95 90 62120 100 62120 62120 95 62120 100 62120 100 62120 800 62120 105 62120 105 62120 100 62120 105 62120 105 62120 110 62120 l10 62120 Ild 62120 106 62140 115 62120 115 62120 110 62120 110 62120 115 62140 115 62120 116 62120 116 621D0 116 62120 116 105182 116 1052.41 116 1050.42 120 10P2.51 120 109034 120 109186 125 1840.18 130 18913 0 135 124 04 140 1305.15 145 13E46 150 143v.15 155 1509.15 160 1587.19 165 1670.65 110 1740.18 175 1856.22 leo 1959a9 185 2069 28 190 2t E42

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195 2311.89 200 2445.74 Page 1 of 3 l

Enclosure 3a DATA POINTS FOR llEATUP AND COOLDOWN CURVES WITHOUT MARGINS From WCAP-13720 for Catawba Unit 1 40 HU OWnitty limit Hyes IaskTen T P T P T P as 621a0 IM 080 its 2000 to 62140 IM 62120 IM 24ss 95 628A0 iM 42140 le 62tmo IM 621 m im 628Ao IM 62120 180 62880 IM 62820 113 62120 IM 62120 116 62120 IM 62120 116 91170 IM 91173 130 m l7 IM eM,11 12$ 900J1 IM teejl 130 1006.48 IM lous.48 IM 1047A9 IM 10g7At le lorL44 im im44 145 1841J3 Its 130 llN.93 leo liestidl.,ss 3

IS$ 125165 1 95 125L65 160 1313.24 200 1315.24 168 IM173 20$ 138173 110 1433JI 280 14$$Ji

!?$ 1333J2 115 IS3332 le 1617J0 220 let?J0 in 11o7J4 22s 19e?.x leo Ism.sl 230 Iml.si los loo 7m us footm 200 20l7.68 340 2017.68 20$ 2133J8 243 2135 3 8 210 1261.95 230 2261.95 213 2M6J3 233 2396J3 i

Page 2 of 3

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35 522.28 35 48300 35 4e521 35 552.E 245 0.00 224 2000 M *"

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120 621 00 120 621A0 120 62120 120 595.78 125 53234 128 571.92 q

621.00 125 62 TAD 125 621.00 125 68734 125 557 65 125 58131 245 5e331 e v ;

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I 195 8353.33 j 200 I20534 285 99634 245 99634 <: i 210 1983 43 250 tes343 C tt [

205 1261.45 3 to -

210 1321.11 - 215 1994.26 255 1994.26

215 1385.46 2B 184840 200 1848.e0 j'

220 1454.29 225 120L94 238 128927 285 IBE94 238 13927 "h

to 3 225 1523.13 235 13M56 275 13M36 ::: .

230 1607.1s 240 14e8.50 O 235 1692.15 ' 288 140530 C 240 1783.31 245 148534 285 148534 250 1568.16 290 1568.16 *"3 l 245 1880.45 255 16%46 295 1466.46 f 250 198434 28B 1751.08 4 255 2096.ts - 265 185236 308 175IAs 305 185236 f

y i

i i 260 221535 210 19E0.55 310 19e055 g !'

l 265 2342.65 275 2075.91 w 270 247330 315 20T3.91 j 200 219948 3B 219948 Z 285 233tJt 325 233IJI CD

] 290 2471.53 330 2473.53 i 1

-l j

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t

Enclosure 3b UPDATED LEAD FACTORS FOR CATAWDA UNIT 1 SURVEILLANCE CAPSULES From.

WCAP-13720 Table 6-19 for Catawba Unit 1 1

Cacaub Lead Factor l

Z 4.10*

Y 3.66*

V 3.66"

{

W 3.91* ]

X 3.91*

U 3.91M (a) Plant specNic evaluation based on end of Cycle 1 calculated fluence.

(b) Plant specNic evaluatbn based on end of Cycle 6 calmlated fluence, f

I Pnge 1 of 2

Enclosure 3b I

UPDATED LEAD FACTORS FOR CATAWBA UNIT 2 SURVEILLANCE CAPSULES From WCAP-13875 Table 6-17 for Catawba Unit 2 CAPSULE LEAD FACTOR U 3.79 Y 3.62 !

X 3.62' W 3.79 ,

Y 3.79 Z WTTHDRAWN EOC l'

BAS 15 POR THIS ANALYSIS Note here that it is assumed that the capsule X dosimetry was installed in the capsule Y location and that the capsule Y dosimetry is installed in the capsule X location.

o Page 2 of 2

_ _ -. ._ - .. .- .~,- - - - - - - - . - - - - _ - _ _ . . - - . . - . _ . . - _ . . - ._.___ - ..

Enclosure 3c SURVEILLANCE CAPSULF REMOVAL SCHEDULE Prom WCAP-13720 Section 7.0 for Catawba Unit 1 4

i The following removal schedule meets ASTM E185-82 and is recommended for future caps removed from the Catawba Unit 1 reactor vessel:

J Capsule Estimated j

Location Lead Fluence Capsule Idstl Eggg .Romoval Time W _(nbrh Z 301.5 4.10 0.79 (Removed) 3.43 x 10(Actual)

Y 241 3.66

! 4.98 (Removed) 1.35 x 10" (Actual)

V 61 3.66 8.75 W

2.52 x 10(b) 121.5 3.91 12.5 3.8 X 10" f X 238.5 3.91 Standby -

U 58.5 3.91 Standby -

1 (a) Effective Fuu Power Years (EFPY) from plant stanup.

(b)

Maximum end of boense (32 EFPY) inner vessel wsN fluence.

t l'

4 Page 1 of 2

Enclosure 3c SURVEILLANCE CAPSULE REMOVAL SCHEDULE From WCAP-13875 Table 7-1 for Catawba Unit 2 The following surveillance capsule removal schedule meets the requirements of ASTM E185 82 and is recommended for future capsules to be removed from the Catawba Unit 2 reactor vessel:

l i

l TABLE 71 Catawba Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule

' Removal Time Fluence Location Lead Factor (EFPY)"3 (n/cm', E > 1.0 MeV) 301.5' 4.09 0.86 3.44 x 10" *)

24l' 3.62 4.52 1.219 x 10" *)

238.5* 3.79 8.5 2.48 x 10" "I 61' 3.62 13.5 3,76 x 10" 121.5' 3.79 Stand-by -

58.5' 3.79 Stand By .-

(a) Effective Full Power Years (EFPY) from plant startup.

(b) Actual measured neutron Guence (c) Approximate EOL (32 EFPY) peak' vessel inner surface fluence.

Page 2 of 2

Attachment 4 No Significant Hazards Considerations Evaluation 1

, The following is a discussion of the significant changes associated with each of the four (4) identified groups:

1) Revise the Pressure / Temperature Curves to 15 EFPY, including the incorporation of the latest radiation surveillance capsule results, and removal of instrumenc j margins and material footnotes from the Technical i specifications figures.

. The Pressure / Temperature curves (heatup and cooldown 4

curves) were developed by Westinghouse Electric Corporation and are included in the appendices of WCAP-13720 and WCAP-13875 for Units 1 and 2 respectively. These surveillance capsule reports, (WCAP-13720 and WCAP-13875) were submitted to the NRC for review by letters dated August 12, 1993, and March i 2, 1994, respectively. These curves extend the service period from 10 EFPY to 15 EFPY. The new heatup and cooldown curves have satisfied all required material i

embrittlement considerations, including: 10CFR50, Appendix G; ASME Section III, Appendix G; and Reg.

1 Guide 1.99, Rev. 2. Also, the instrumentation margins 4

' have been removed from the curves. The instrument margins are implemented administrative 1y by incorporation into the controlling procedures for unit operations. Additionally, the instrument uncertainties have been incorporated into the LTOP system set point selection calculation. Removal of the material footnotes (for copper and nickel) is considered an

^

administrative change and this information is available in reference documents if needed for future use.

2) Modify the actual capsule ID listed on Table 4.4-5

" Reactor Vessel Material Surveillance Program -

Withdrawal Schedule" (for Unit 2 only) and update each units' Lead Factors and Withdrawal Time. This proposed change applies to the current Technical Specifications only.

As discussed in Attachment 3 the impetus for this _

request is to update the Lead Factors, Withdrawal Times, and to properly identify (ID) the Catawba Unit 2 i

Attachment 4 No Significant Hagards Considerations Evaluation capsules because apparently capsule X and Y were interchanged at the time of construction. ,

l These new Lead Factors and corrected capsule ID (for '

Unit 2 only) are included in the appendices of WCAP-13720 and WCAP-13875 for Unit 1 and 2, respectively. ,

These surveillance capsule reports were submitted to !

the NRC for review by letters dated August 12, 1993 and j March 2, 1994, respectively. '

3) Revise the Technical Specifications requirement for the Reactor Coolant System (RCS) Overpressure Protection system during low temperature conditions.

As discussed in Attachment 3, the impetus for this request is to update the setpoint listed in Technical Specification 3.4.9.3a/3.4.12a to less than or equal to 400 psig (as left calibrated), allowable value less than or equal to 425 psig (as found).

4) Change format and consistency.

Other changes have been made to improve consistency between Technical Specifications, incorporate the Westinghouse Improved Standard Technical Specifications format, and update applicable code references. These changes are considered editorial in nature and do not affect the way the units are operated.

Pursuant to 10CFRSO.92, this analysis concerns whether the proposed Technical Specifications amendment involves significant hazards considerations, as defined by 10CFR 50.92(c). The standards for determining that a Technical Specifications amendment request involves no significant hazards considerations requires that subsequent operation of the facility in accordance with the requested amendment will not:

1) Involve a significant increase in the probability or consequence of an accident previously evaluated; or

2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3) Involve a significant reduction in a margin of safety.

2

Attachment 4 No Significant HaEards Considerations Evaluation The following discussion is a summary of the evaluation of the changes contained in this proposed amendment against the i 10CFR50.92(c) requirements to demonstrate that all three standards are satisfied. Within each of the three standards, the four identified groups of Technical

} Specifications changes contained in this proposed amendment

{ are discussed.

i FIRST STANDARD (Amendment would not) involve a significant increase in the probability or consequences of an accident

} previously evaluated.

l 1) Revise the Pressure / Temperature Curves to 15 EFPY, i including the incorporation of the latest radiation '

{ surveillance capsule results, and removal of j- instrument margins and material footnotes from the Technical Specifications figures:

3 The proposed heatup and cooldown curves, provided by

, this amendment request, satisfy all regulatory required j material embrittlement considerations including:

10CFR50, Appendix G; ASME Section III, Appendix G; and i Regulatory Guide 1.99, Revision 2. In addition, the margins for instrument error have been removed from the I

curves. Instrument error will be administratively i handled by incorporating them into the LTOP System set a point selection calculations and into appropriate i controlling procedures for unit operations.

1 4 The proposed changes to the heatup and cooldown curves 3 are not considered to be an initiator of~LTOP events.

l The changes to the curves proposed by this amendment i request will not cause an LTOP event. The curves

define the new limits that have been defined in l accordance with regulatory requirements within which

! both units are to be operated. Accordingly, the i

proposed changes will not increase the probability or

{ consequences of any previously evaluated accident, s

2) Modify the actual capsule ID, update the Lead Factors and Withdrawal Times (current Technical Specifications only)

) The changes in this group are considered to be 1 administrative in nature. They do not affect station operability or require any modifications to the

. facility. Accordingly, the proposed changes will not i increase the probability or consequences of any previously evaluated accident.

! 3 4

l Attachment 4 No Significant Hazards Considerations Evaluation i

4

3) Revise the Technical Specifications requirement for the Reactor Coolant System (RCS) Overpressure 4

Protection System during low temperature conditions:

The analysis performed to determine the setpoint is in

, accordance with the methods used in previous evaluations that l

., were found acceptable by the NRC. The three possible j transients evaluated are: 1) a mass input from an operable j

safety injection pump; 2) a mass input from an operable centrifugal charging pump; and 3) a heat input from a 500F 3

temperature difference between the steam generators and the rest of the RCS system. The LTOP setpoint of the PORV

proposed by this Technical Specifications amendment is not considered to be an initiator of any of these three

transients. As such, the probability of an accident 4

previously evaluated would not be increased as a result of l the proposed changes. Additionally, the consequences of an LTOP event would not change as a result of the proposed amendment. This is because the proposed 400 psig setpoint is more conservative than the current Technical Specification i requirement of 450 psig and the mass addition transient would consequently be more limited.

2 l 4) Format and consistency:

The changes in this group are considered to be administrative in nature. They do not affect station operability or require any modifications to the facility. Accordingly, the proposed changes will not i increase the probability or consequences of any ,

! previously evaluated accident.

5 SECOND STANDARD (Amendment would not) create the possibility

, of a new or different kind of accident from any kind of

accident previously evaluated.

)

1

1) Revise the Pressure / Temperature Curves to 15 EFPY, d

including the incorporation of the latest radiation

surveillance capsule results, and removal of i instrument margins and material footnotes from the Technical Specifications figures

The changes in this group will provide new heatup and cooldown curves for both Units 1 and 2, which will extend the service period from 10 EFPY to 15 EFPY and will remove the-instrument error as well. The proposed

! heatup and cooldown curves were developed in accordance 4

Attachment 4 No Significant Hazards Considerations Evaluation with all regulatory required material embrittlement criteria. Instrument error has been moved to a licensee controlled engineering calculation, where it has been conservatively included in the peak pressure evaluation. Thus, operation of the units in accordance with the proposed new heatup and cooldown curves will i not create the possibility of a new or different kind l

of accident from those accidents that have been previously evaluated.

2) Modify the actual capsule ID, update the Lead Factors and Withdrawal Times (current Technical Specifications only):

The changes in this group are considered to be administrative in nature. They do not affect station l operability or require any modifications to the l facility. Accordingly, the proposed changes will not I create the possibility of a new or different kind of accident from those previously evaluated.

3) Revise the Technical Specifications requirement for the Reactor Coolant System (RCS) Overpressure Protection System during low temperature conditions:

The proposed changes to the listed LTOP setpoint are not considered to be an initiator of LTOP events. The changes to the setpoint contained in this proposed amendment will not cause an LTOP event. The proposed amendment, also, would not impact the plant operation. Although the value for the PORV pressure setting specified within the Technical Specifications would be reduced per the proposed amendment, the actual settings of the PORV are now currently being calibrated to this lower setpoint. As such, the proposed lower set point would not require any changes to the plant, nor how the plant is operated. Notwithstanding Catawba's conservative operating practices in regard to this proposed change, lowering the Technical Specifications requirement for the LTOP setpoint, is not' considered an accident initiator.

Accordingly, the proposed changes will not create the possibility of a new or different kind of accident from those previously evaluated.

4) Format and consistency:

The changes in this group are considered to be administrative in nature. They do not affect station operability or require any modifications to the facility. Accordingly, the proposed changes will not 5

Attachment 4

.No Significant Hazards Considerations Evaluation create the possibility of a new or different kind of accident from those previously evaluated.

l THIRD STANDARD 1

(Amendmont would not) involve a significant reduction in a margin of safety.

1) Revise the Pressure / Temperature Curves to 15 EPPY, including the incorporation of the latest radiation surveillance capsule results, and removal of instrument margins and material footnotes from the Technical Specifications figures:

The changes in this group provido new heatup and cooldown curves for both Units 1 and 2, which will extend the service period from 10 EFPY to 15 EFPY and will relocate the instrument error as well. The proposed heatup and cooldown curves provided by this amendment request satisfy all regulatory required material embrittlement considerations including: ASME Section XI Appendix G, 10 CFR 50 Appendix G, and L Regulatory Guide 1.99, Revision 2. The instrument error will be administrative 1y handled by incorporation into the LTOP system set point selection calculation and into the controlling procedures for unit operations.

The relocation of the instrument error to licensee controlled documents is consistent with NUREG-1431, the new Improved-Standard Technical Specifications for Westinghouse plants. As prescribed within NUREG-1431, '

the entire heatup and cooldown limitations curves could actually be relocated to a licensee controlled document entitled " Pressure Temperature Limit Report (PTLR)".

Afterwards, future changes to the heatup and cooldown curves would then be performed in accordance with 10CFR50.5S criteria. For the situation proposed by

.this amendment, updates and revisions of the instrument error associated with the heatup and cooldown curves will also be processed in accordance with 10CFR50.59.

Thus, the proposed change to relocate the instrument error to licensee controlled documents is analogous with NRC acceptable practices. Accordingly, the proposed changes will not reduce the margin of safety.

6

. . J

I i Attachment 4 No Significant 11azards Considerations Evaluation

2) Modify the actual capsule ID, update the Lead Factors and Withdrawal Times (current Technical S p e c i ,11_i c a t i o n s o n l y ) :

The changes in this group are considered to be administrative in nature. They do not affect station operability or require any modifications to the facility. Accordingly, there is no reduction in the margin of safety due to the incorporation of these editorial / administrative changes.

3) Revise the Technical Specifications requirement for the Reactor Coolant System (RCS) Overpressure Protection System daring low temperature conditions:

This proposed change will reduce the maximum PORY set point auch that, for LTOP events, the maximum press',tre in the vessel would not exceed the Pressure / Temperature limits that have been established in accordance with ASME Saction III, Appendix G. Accordingly, the proposed changes will not reduce the margin of safety.

4) Format _and consistency:

The changes in this group are considered to be administrative in nature. They do not affect station operability or require any modifications to the facility. Accordingly, there is no reduction in the margin of safety due to the incorporation of these editorial / administrative changes.

Based on the above discussion and the supporting technical justification contained in Attachment 3, Duke Energy Corporation has concluded that there are no significant hazards conniderations involved in this proposed amendment.

7

I l

Attachment 5 Environmental Assessment Pursuant to 10CFR51.22(b), an evaluation of this license amendment request has been performed to determine whether or not it meets the criteria for categorical exclusion set forth in 10CFR51.22 (c) (9) of the regulations.

This amendment to the Catawba Units 1 and 2 Technical Specifications updates the current heatup and cooldown curves, revises the Technical Specifications for the Reactor Coolant System Overpressure Protection System, and updates the Reactor Vessel Material Surveillance Program. The

resultant p. essure / temperature limitations are conservative and satisfy all required material embrittlement censiderations necessary to ensure future Reactor Coolant System integrity over an expanded period of time (from 10 EFPY to 15 EFPY). The change to the referenced overpressure protection Technical Specifications (from 450 psig to 400 psig/425 psig allowable) and the changes made to the material surveillance program are conservative. Consequently, implementation of this amendment will have no adverse impact on the Reactor Coolant System, and neither will contribute to any additional quantity or type of effluent being available

- for adverse environmental impact or personnel exposure.

It has been determined there is:

1) No significant hazards consideration (see Attachment 4);

2) No nignificant change in the types, or significant increase in the amounts, of any effluents that may be released offsite; and

3) No significant increase in individual or cumulative occupational radiation exposures involved.

1 Therefore, this amendment to the Catawba Technical Specifications, meets the criteria of 10 CFR 51.22 (c) (9) for categorical exclusion from an environmental impact statement.

l Attachment 6 Westinghouse Electric Corporation Topical Report WCAP-9734, Duke Power Company Catawba Unit.1 Reactor Vessel Radiation Surveillance Program, 9709220006 970915 PDR ADOCK 05000413-P PDR s -

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WESTINGHOUSE CLASS 3 x

In 6

I W I

DUKE POWER COMPANY

- CATAWBA UNIT NO.1 REACTOR VESSEL RADIATION d

SURVEILLANCE PROGRAM L

S. E. Yanichko July 1980

~

APPROVED:

T. R. Mager, Manager

- Metallurgicaland NDE Analysis J

Work Performed Under DCP 106 r

f WESTINGHOUSE ELECTRIC CORPORATION

{ Nuclear Energy Systems P. O. Box 355 Pittsburgh, Pennsylvania 15230 o

l~

~

d

~

PREFACE

~

This report has been technically reviewed and checked by L. R. Singer of Metallurgical and NDE Analysis.

4 1

] .

L. R. Singer Date: June 10,1980

] '

N m

N i

lii

l E

l ABSTRACT E

l A pressure vossol stoel surveillanco program por ASTM E 185 73 has been developed for the Duke Power Company Catawba Unit No.1 to obtain information on the effects of radiation on reactor pressure vessel material under operating conditions, The radiation surveillance program for the Catawba Unit No.1 is dos!gned to, and in compliance with, fodoral government regulations identified in appendix H to 10CFR, part 50, entitlsd

" Reactor Vessel Material Surveillance Program Requirements."

Following is a description of the program, a description of the material involved, the specimon and capsule design, and the preirradiat,'on test results.

E E

E '

E E

R 4

TABLE OF CONTENTS .

Section Title Page

] 1 PURPOSE AND SCOPE 11 2 CAPSULE PREPARATION 21 2 1. Pressure Vessel Material 21 2 2. Machining 21 2 3. Charpy V notch Impact Specimens 21 2 4. Tensile Specimens 23 2 5. 1/2T Compact Tension Specimens 23 2 6. Dosimeters 23 2 7. Thermal Monitors 23 2 8. Capsule Loading 29

_ 3 PREIRRADIATION TESTING 31 3 1. Charpy V notch Tests 31 3 2. Tensile Tests 31 3 3. Dropweight Tests 32

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4 POSTIRRADIATION TESTING 41 4 1. Capsule Removal 41 4 2. Charpy V notch Impact Tests 42 4 3. Tensile Tests 42 4 4. Fracture Tocchness Tests on 1/2T Compact Tension Specimens 42 4 5. Postirradiation Test Equipment 43 Appendix A CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL SURVEILLANCE MATERIAL A1 vil i- -_. _

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LIST OF ILLUSTRATIONS 4 Figure l Title Page l 21 Charpy V notch Impact Specimens 22 22 Tensile Specimen 24 23 Compact Tension Specimen 25 24 Irradiation Capsule Assembly 2 7/2 8 i 25 Dosimeter Block Assembly 2 10 l 20 Specimen Locations in the Catawba Unit No.1 Reactor Surveillance Test Capsules 213/214 31 Preirradiation Charpy V notch Impact Energy for the l Catawba Unit No.1 Reactor Pressure Vessel Intermodlate Shell Forging 05 (Tangential Orientation) 38 32 Preirradiation Charpy V notch impact Energy for the 5 Catawba Unit No.1 Reactor Pressure '!essel Intermediate Shell Forging 05 ( Axial Orientation) 39 33 Preirradiation Charpy V notch impact Energy for the Catawba Unit No.1 Reactor Pressure Vessel Core Region Weld Metal 3 10 34 Preirradiation Charpy V notch Impact Energy for the Catawba Unit No.1 Reactor Pressure Vessel Core Region Weld Heat Aflected Zone Material 3 11 35 Preirradiation Tensile Proporties for the Catawba Unit No.1 Reactor Pressure Vesselintermediate Shell Forging 05 (Tangential Orientation) 3 12 36 Preirradiation Tensile Properties for the Catawba Unit No.1 Reactor Pressure Vessellntermediate Shell Forging 05 ( Axial Orientation) 3 13 U7 Preirradiation Tensile Pronerties for the Cataba Unit No.1 Reactor Pressure Vessel Core Region Weld Metal 3 14 38 Typical Stress Strain Curve for Tensile Test 3 15 ix

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k LIST OF TABLES E Table Title Page 21 Type and Number of Specimens in the Catawba R Unit No.1 Surveillance Test Capsules 29 22 Otantity of Isotopes Contained in the Dosimeter Blocks 2 11 l 31 "reirradiation Charpy V notch Impact Data for the Cataw:m tinit No.1 Reactor Pressur3 Vossel intermediate dieell Forging 05 (Tangential Orientation) 33 l 32 Preirradiation Charpy V notch Impact Data for the Catawba Unit No.1 Reactor Pressure Vessel Intermediate Shell Forging 05 ( Axial Orientation) 34 33 Preirradiation Charpy V notch Impact Data for the Catawba Unit No.1 Reactor Pressure Vessel Core Region Wold Metal 35 34 Preirradiation Charpy V notch Impact Data for the Catawba Unit No.1 Reactor Pressure Vessel Core Region Weld Heat Aff,ected Zone Material 36 I 35 Preirradiation Tensile Properties for the Catawba Unit No.1 Reactor Pressure Vesselintermediate Shell Forging 05 and Core Region Weld Metal 37 E 41 Capsule Lead Factors 41 I

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N SECTION 1 t

PURPOSE AND SCOPE N

l The purpose of this program is to monitor radiation effects on the reactor vessel materials of the Duke Power Company Catawba Unit No.1, a four loop,3565 megawatt plant, under actual operating conditions. Evaluation of the radiation effects is based on preirradiation testing of Charpy V notch, tensile, and dropweight specimens, and postirradiation testing of Charpy V notch, tensile, and compact tension specimens.

Current reactor pressure vessel material test requirements and acceptance standards utilize the reference nil ductility temperature, RTNDT, as a basis. RT NDT is determined from the dropweight nil ductility transition temperature (NDTT) per ASTM E208 and the '

weakl11 direction 50 f t Ib Charpy V notch temperature (or the 35 millateral expansion temperature if it is greater). RT NDT si defined as the dropwe!ght NDTT or the temper-ature 60'F less than the 50 ft Ib (or 35 mil) Charpy V notch temperature, whichever is grenter.

Therefore RTNDT = NDTT,if NDTT ) T50(35) 60'F and RTNDT = T50(35) 60*F,if T50(35) 60'F > NOTT where RTNDT = Reference nil ductility temperature NDTT

= Nil ductility transition temperature per ASTM E208 T

50(35) = 50 ft Ib temperature from Charpy V notch specimens oriented in the weak direction (or the 35 mil temperature if it is greater)

1. Longitudinal aos of the specimen oriented normal to the maior working direction of the forging 11 I

An empirical relationship between RTNDT and fracture toughness for reactor vessel ,

steels has been developed in appendix G, " Protection Against Non ductile Failure," to Section 111 of the ASME Boiler and Pressure Vessel Code This re!ationship can be ;

employed to set allowable pressure temperature limitations for normal operation of reactors which are based on fracture mechanics concepts. Appendix G defines an I,

acceptable method for calculating these limitations. '

lt is known that radiation can shift the Charpy V notch impact energy curve to highcr temperatures.l1,2) Thus, the 50 ft Ib temperature and RTNDT ncrease'with I radiation exposure. The extent of the shif t in the impact energy curve, that is, radiation embrittle.

ment, is enhanced by certain chemical elements (such as copper) present in reactor vessel steels.13.41 1

The 50 f t ib temperature or RT NDT increase with service can be monitored by a surveil-lance program involving periodic checking of irradiated reactor vessel surveillanco a specimens. The surveillance program is based on ASTM E185 73 ' Standard Recom-mended Practice for Surveillance Tests for Nuclear Reactor Vessels). Compact tension ,

fracture mechanics specimens will be used in addition to Charpy V notch specimens to evaluate the effects of radiation on the fracture toughness of reactor vessel mater.

tals. [5,6,7,8.9.10,11 )

8 I

1 Porter, L. F., Radiaton Effects in Steel." in Matenals m Nuclear Apolscat,ons. ASTM STP 216 pp 141 195. Amencan Society for Testing and MaterinM. Philadelphia.1960 t 2 Steele. L E and Hawthorne, J R , "New informahon on Neutron Embrittlement and Embrittiment Relief of Reactor Pressure i VesselSteels."NRL 6160, August 1964 3 Potapovs. U and Hawthorne. J. R , "The rffect of Residual Elements on 550' F irradiation Response of Selected Pressure Vessel Steels and Weldments." NRL 6803, September 1968 J Stee6e L. E., " Structure and Compositen Effects on irradiaton Sensitivity of Pressure Vessel Steeis."interad,atton Effeels on Structural Alloys for Nucient Reactor Applecat#ons ASTM STP 484. pp 164175, Amencan Society for Testing and Matenais.

Philadelphia.1970 6 Landerman, E., Yanichko, S E. and Hazelton, W S . " An Evaluation of Radiakon Damage to Reactor Vessel Steels Using Both Transiten Temperature and Fracture Mechanics Approaches,"in The Effects of Rad,ation on Structural Metals. ASTM STP.

426, pp. 260 217. Amencan Society for Testing and Matenais, Philadelphia 1967.

6 Mantoine. M J Biaxial Bnttle Fracture Tests," Trans, Am Soc. Mech. fagra. 8 7 Senes D, 293 298(1965) 7 Porse, L.,

Reactor Vessel Design Contadonng Radiation Effects." Trans Am Soc. Mech Engrs 86, Senes D, 743 749 (1964).

8 Johnson. R E. " Fracture Mechanics A Basis for Bnttle Fracture Preventon," WAPO.TM 50$. November 1965 ,

9. Wessel. E. T. and Pryle W H , " investigation of the Applicabilit, of the Bianial Bnttle Fracture Test for Determining Fracture l Toughness." WERL-884411, August 1965 .

10 Wilson, W K., ' Analytic Determ# nation of Stress intensity Factors for the Manjoine Brittle Fracture Test Specimen," WERL-0029 3, August 1965 ;

11 Johnson, R E. and PasierD. E. J " Fracture Toughness of Irradiate <1 A302 0 Steel as influenced by Microstructure," Trans. t Amer. Nuct Soc 9, 390 392(1966) i 12

E Postirradiation testing of the Charpy impact specimens will provide a Cuide for deter-mining pressure temperature limits on the plant. Charpy impact test data will determine the shift of the reference temperature with radiation exposure at plant temperatures.

Thase data can then be reviewed to verify or revisa pressure temperature limits of the vessel during startup and cooldown (the Charpy) specimens are most nearly indicative l of the radiation exposure experienced by the vessel). This will allow a check of the predicted shift in the reference temperature. The postirradiation test results of the compact tension specimens will provide actual fracture toughness properties of the vessel material. These properties may be used to establish allowable stress intensity factors for subsequent analyses.

Six material test capsules, located in the reactor between the neutron shielding pads and

) vessel wall, are positioned opposite the cerder of the core. The test capsules are i

located in guide tubes attached to the neutron shielding pads. The capsules contain test specimens from a forging from the reactor vesselintermediate shell course adjacent to the core region, representative wold metal, and heat affected zone (HAZ) metal.

The thermal history or heat treatment given these specimens is similar to the thermal history of the raactor vessel material with the exception that the postweld heat treatment received by the specimens has been simulated (appendix A).

I I

i 13

SECTION 2 CAPSULE PREPARATION l

21. PRESSURE VESSEL MATERIAL Reactor vessel material was supplied by Rotterdam Dockyard Company, from interme-f diate shell forging 05 (Heat No. 411343). Rotterdam Dockyard Company also supplied c weldment which joined sections of material from intermediate shell forging 05 and the adjoining lower shell course forging 04. Data on the limiting core region forging, weld, and weld heat aflected zone materialare provided in Appendix A.

l 2 2. MACHINING Test material obtained from the intermodlate shell course forging 05 (after the thermal l heat treatment) wcs taken at least one forging thickness from the quench'ed ends of the forging. All test specimens were machined from the li4 thickness location of the forging af ter performing a simulated postweld stress relieving treatment on the test material and also from weld and heat affected zone metal of a stress relieved weldment joining inter-mediate shell forging 05 (Heat No. 411343) and lower shell forging 04 (Heat No.

527708). All heat affected zone specimens were obtained from the weld heat affected zone of forging 05.

2 3. Charpy V notch impact Specimens Charpy V notch impact specimens (figures 21) from intermediate shell forging 05 were machined in both the tangential orientation (longitudinal axis of specimen parallel to mujor working direction) and axial orleatation (lorigitudinal axis of specimen perpendicular to major working direction). The core region weld Charpy impact specimens were machined from the weldment such that the long dimension of the Charpy specimen was normal to the weld direction. The notch was machined such that the direction of crack propagation in the specimen was in the weld direction.

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2-4. Tensile Specimens 43 Tensile specimens (figure 2 2) from forging 05 were machined sc, J. to produce some

$ with the longitudinal axis of the specimen normal to and some perpendicular to the major working direction of the forging. Specimens from the weld were oriented normal to the weld direction.

2 5. Il2T Compact Tension Specimens Compact tension test specimens (figure 2 3) from forging 05 were machined in both the axial and tangential orientations. Compact tension test specimens from the weld metal were machined normal to the weld direction with the notch oriented in the direction of the weld. Allspecimens were fatigue precracked according to ASTM E399 2 6. DOSIMETERS Each of the six test capsules of the type shown in figure 2 4 contain dosimeters of pure copper, iron, nickel, and aluminum 0.15 weight percent cobalt wire (cadmium shielded and unshielded) and cadmium shielded NP237 and U238 which will measure the integrated flux at specific neutron energy levels.

2 7. THERMAL MONITORS The capsules contain two k,w melting point eutectic alloys to more accurately define the maximum tencerature attained by test specimens during irradiation. The thermal monitors are sealed in Pyrex tubes and then inserted in spacers located as shown in figure 2 4. The tw ader ic alloys and their melting points are the following:

. 2.5 percent Ag,97.5 percent Pb Melting point: 579* F 1.75 percent Ag,0.75 percent Sn,97.5 percent Pb Melting point: 590* F 23

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TABLE 21 TYPE AND NUMBER OF SPECIMENS IN THE CATAWBA UNIT NO.1 SURVEILLANCE TEST CAPSULES I

Capsules U, V, W, X, Y, and Z Material Charpy Tensile CT Forging 05 (Tangential) 15 3 4 (Axial) 15 3 4 Weld Metal 15 3 4 HAZ 15 - -

Dosimeters of pure copper, iron, nickel, aluminum 0.15 weight percent cobalt, and cad-mium shielded aluminum-cobalt wires io secured in holes drilled in spacers located at capsule positions shown in figure 2 4. Each capsule also contains a dosimeter block (figure 2 5) located at the center of the capsule. Two cadmium oxide shielded capsules, each containing isotopes of either of U2aa or Np237, are loca:ad in the dosimeter block.

The double containment afforded by the dosimeter assembly prevents loss and contamination by the U238 and Np237 and their activation products. Each dosimeter block contains approximately 12 milligrams of U238 and 17 milligrams of Np237 (table 2-

2) held in a 3/8 inch long by 1/4 inch outside-diameter sealed stainless steel tube, respectively. Each tube was placed in a 1/2-inch diameter hole in the dosimeter block 29 L

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(one U238 and one Np237 tube per block), and the space around the tube was filled with cadmium oxide. After placement of this material, each hole was blocked with two 1/16-inch-thick aluminum spacer discs and an outer 1/8 inch thick steel cover disc welded in place.

The numbering system for the capsule specimens and their locations is shown in figure 2 6. The specimens are seal-welded into a square capsule of austenitic stainless steel to prevent corrosion of specimen surfaces during irradiation. The capsules were hydro-statically tested in demineralized water to collapse the capsule on the specimens so that optimum thermal conductivity between the specimens and the reactor coolant is obtained. The capsules were helium leak tested as a fini11 inspection procedure. Fabri-cation details and testing procedures are listed in figure 2 4.

l TABLE 2-2 QUANTITY OF ISOTOPES CONTAINED IN THE DOSIMETER BLOCKS lsotope Weight (mg) Compound Weight (mg)

Np237 17 i 1 NpO2 20 1 U238 12,0 U08 3 14.25 2-11

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L SECTION 3 N \

PREIRRADIATION TESTING

31. CHARPY V NOTCH TESTS Charpy V notch impact tests were performed per ASTM E 23 with specimens from the vesselintermediate shell forging 05. Specimens of both axial and tangential orientations were tested at various test temperatures in the range -100 to 210* F yielding a full Charpy V notch transition curve in both orientations (tables 31 and 3 2 and figures 31 and 3 2). Tests were also performed on weld metal and HAZ metal at various temper-atures from -200 to 210* F. The results are reported in tables 3 3 and 3 4 and figures 3 3 and 3 4.

The specimens were tested on a Sontag SI 1 impact machine which is inspected and calibrated every 12 months. Charpy V notch impact specimens of known energy values, supplied by the Watertown Arsenal, are used for the calibration.

3 2. TENSILE TESTS Table 3-5 and figures 3 5,3 6, and 3 7 give results of tensile tests (per ASTM E 8 and E 21 test criteria) from vessel intermediate shell forging 05 and from the weld metal.

Specimens from the shell forging were tested at room temperature,300* F, and 550 F in both the axial and tangential directions.

An Instron TT C tensile testing machine was used with the standard Instron gripping devices. A Baldwin Lima Hamilton Class B 1 extensometer and chart recorder provided a full stress strain curve for each specimen. The chart recorder was calibrated to the Class B 1 extensometer. The measurement and control of speeds in the tests conformed to ASTM A370-68 (Mechanical Testing of Steel Products). The instron TT C and the Baldwin Lima-Hamilton extensometer are calibrated by test equipment which has been certified by the National Bureau of Standards. A typical stress strain curve is shown in figure 3 8.

3-1

. mm. o . < . m m ._ . _ _ . . . .- _ . . _ - _ . _ _ ___ _ _ - .

I1l

- 3 3. DROPWEIGHT TESTS The nil ductility transition temperaturo (TNDT) was determined for forging 05 and the core region weld metal and heat affected zone by dropweight tests (ASTM E 208) performed at Rotterdam Dockyard Co. The following results were obtained:

Material TNOT (' F)

Forging 05 - 40 Weld Metal - 76 HAZ - 67 6

I f

N l

e I '

32 ?

.._-.---._----------..L*1T"

1 4 TABLE 31 f

PRElRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FORGING 05(TANGENTIAL ORIENTATION) l Test Temperature impact Energy Shear Lateral Expansion E

(

F) (ft Ib) (%) (milu)

- 100 12 0 1

- 40 15 0 10.5

- 40 11 0 3 l

0 15 23 12 0 50 27 37 0 17 33 15 10 17 42 14 10 24 33 19,5 20 85 45 56 20 85 48 62 20 75 52 51 75 126 73 82 75 116 66 80 120 139 81 85 120 130 81 86 210 158 100 94 210 177,5 100 83 210 168 100 88 33

[ . _ . . . . . _ . . .

TABLE 3 2 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FORGING 05(AXlAL ORIENTATION)

Test Temperature impact Energy Shear Lateral Expansion

(

F) (ft ib) (%) (mils)

- 100 2 O O

- 40 8 0 5.5

- 40 16 0 11

- 15 48 25 34 0 27 20 19 i O 35 30 25 0 51 30 38 20 60 34 44 20 56 40 44 20 69 45 52 75 81 52 60 75 89 61 67 120 111 82 77 120 119 81 79.5 150 135 100 90 .

210 137 100 88 210 132 100 91 9

210 133 100 87 .

34

1 TABLE 3 3 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION

- WELD METAL m

w% '

Test Temperature impact Energy Shear Lateral Expansion

(

F) (ft Ib) (%) (mils)

N - 100 12 13 6 I - 60 13 28 9.5

- 60 15 33 11.5

' - 40 37 33 26

- 40 26 28 18

- 16 40 42 31 16 60 37 45.5

] -

16 5

54 44 50 54 39 38 m 25 91 72 66

~ 25 98.5 87 74 75 119 87 86 75 110 95 77 120 132 100 90.5 120 119 100 87

, 210 133 100 90 210 130 100 90 210 124 100 89 I

35 I

u E' '7,u. .'Ab,1 -,3 .7 w.aA T v 3 ,

.w.*-m.r5. _ _ . . . - ,

II I

TABLE 3 4 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD HEAT AFFECTED ZONE MATERIAL l

Test Temperature Impact Energy Shear Lateral Expansion

('F) (ft 16) (%) (mils)

- 200 13 0 4

- 170 35 33 22

- 150 96 80 54

- 100 82 38 43

- 40 74 54 48

- 40 108 50 58 l -7 135 81 74

-7 110 77 60

-7 127 81 73 25 159 100 85 25 147.5 89 81 75 150 100 83 75 138.5 93 80.5 120 154 100 84 120 162 100 83 210 180 100 81.5 210 159 100 85 210 166 100 83 es 36 m

] {a ~. _ 2CC u L ..cm LA. 2 LL.~.J

p ;

f I TABLE 3-5 h PREIRRADIATION TENSILE PROPERTIES FOR THE CATAWBA UNIT NO.1 l f REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FORGING 05 h AND CORE REGION WELD METAL U

i i O.2 % Ultimate l

Test Yield Tensile Fracture Fracture Uniform Total Reductior

{ Temp. Strength Strength Load Stress Elongation Etongation in Area Material *F (ksi) (ksi) (ib) (ksi) % % %

Forging 05 (Ht. 411343) 70 69.0 89.1 2580 191.4 16.2 29.6 72.6 r (Tangential Orientation) 70 68.7 89.0 2600 198.9 15.6 28.6 73.4' 300 61.9 81.2 2500 177.3 12.4 24.4 71.4 300 60.9 80.3 2700 1ES.O 13.0 24.2 70.3 550 61.2 85.6 2650 167.4 13.5 25.3 67.9 550 60.3 85.1 2625 154.7 13.4 24.8 65.6 N Forging 05 (Ht. 411343) 70 66.5 87.1 2600 198.9 17.0 29.0 73.4 s (Axial Orientation) 70 66.5 86.9 2700 191.5 16.4 28.2 71.4 300 62.0 80.8 2425 157.6 8.3 17.2 68.7 f[

300 550 60.4 80.2 84.6 2700 150.9 11.2 23.0 63.6 58.8 2825 162.0 14.7 25.5 64.7 550 59.7 85.0 2725 160.6 15.0 25.3 65.6 Weld Metal 70 75.3 88.1 2575 194.0 15.0 27.6 73.2 70 75.0 88.0 2600 195.9 14.7 27.4 73.2 300 68.6 79.3 2375 171.0 11.3 24.3 71.8 300 67.0 79.2 2450 171.2 11.1 24.0 71.0 550 64.8 80.2 2550 156.6 11.7 22.4 67.0 550 65.2 80.3 2500 164.8 11.6 22.8 69.3 l

200 O i 160 -

O E

J O O -

120 -

.I 1 0 ?

80 -

-O w I t

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O l EP O I l l I

-100 0 100 200 300 TEMPERATURE (

F) l FIGURE 31 PREIRRADIATION CHARPY V NOTCH IMPACT ENERGY FOR THE CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FORGING 05 (TANGENTIAL ORIENTATION) 38 I

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A l 80 -

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F)

FIGURE 3-2 PREIRRADIATION CHARPY V-NOTCH IMPACT ENERGY FOR THE CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FORGING 05 (AXIAL ORIENTATION) 39

160 120 -

o 2 E ,

b (

h 80 -

E W

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0 I I I I 0

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FIGURE 3 3 PREIRRADIATION CHARPY V-NOTCH IMPACT ENERGY 'FOR THE CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD METAL 3-10 l

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29 -200 -100 0 100 200 300 J TEMPERATURE (* F) l FIGURE 3 4 PREIRRADIATION CHARPY V NOTCH IMPACT l ENERGY FOR THE CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD HEAT-AFFECTED-ZONE MATERIAL 3 11

100 l l l l l 80 -

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F)

FIGURE 3-5 PREIRRADIATION TENSILE PROPERTIES FOR THE CATAWBA UNIT N,0.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FORGING 05 (TANGENTIAL ORIENTATION) 3 12

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FIGURE 3-6 PREIRRADIATION TENSILE PROPERTIES FOR THE CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FORGING 05 (AXIAL ORIENTATION) l 3 13

)

'UU l l l l l

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g O-20 -

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UNIFORM ELONGATION O O O I l 0 100 200 300 400 500 600 TEMPERATURE (* F)

FIGURE 3-7 PREIRRADIATION TENSILE PROPERTIES FOR THE CATAWBA UNIT. NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD METAL 3 14

. , ....-- .--y,w- _--war --m-m- =wer w:m

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SECTION 4 j POSTIRRADIATION TESTING n

41. CAPSULE REMOVAL Srecimen capsules should be removed from the reactor only during normal refueling perods. Table 4 1 lists the capsule identification and their respective leac factors.

Each opecimen capsule, removed after radiation exposure, will be transferred to a post-irradiation test facility for disassembly and testing of cll the specimens. Although the lead factors are slightly greater than the limits specified in Appendix H to 10CFR Part 50, the l capsule results will not be affected. The first capsule (Capsule U) should be removed at

! the end of the first core cycle. Subsequent capsules should be removed per the new schedules identified in ASTM E185 and Appendix H to 10CFR Part 50 which are

, currently being revised. ,

I I

TABLE 41 CAPSULE LEAD FACTORS i

Multiplying Factor By Which the Capsule Leads Capsule identification Vessel Maximum Exposure U 4.05 X 4.05 V 3.37 l Y 3.37

\N 4.05

, Z 4.05 g ..,

i 4 2. CHARPY V NOTCH IMPACT TESTS The testing of the Charpy impact specimens from the intermediate shell forging, weld metal, and HAZ metalin each capsule can t'a done singly at approximately ten different temperatures. The extra specimens should be used to run duplicate tests at tempera-tures of interest to develop the complete Charpy impact energy transition curve.

The in%1 Charpy specimen from the first capsule removed should be tested at room tempeioture. The test value for this temperature should be compared with preirradiation test data. The test temperature for the remaining specimens should then be adjusted higher or lower so as to develop a complete transition curve. For succeeding tests af ter longer irradiation periods, the test temperature in each case should be chosen in the light of results from the previous capsule.

4-3. TENSILE TESTS The tensile tests of specimens for each of the irradiated materials should be performed at room temperature, 300* F, and 550' F, and in accordance with ASTM E 8 and E 21 testing criteria.

4-4.

FRACTURE TOUGHNESS TESTS ON 1/2T COMPACT TENSION SPECIMENS j

in light of current requirements of 10 CRF, Part 50, appendix G,1/2T compact tension j

(CT) specimens should be tested dyr amically to adequately characterize the fracture toughness proporties of the reactor vessel up to the initiation of the fracture toughness upper shelf. The CT specimens for each of the irradiated materials should be tested in accordance with ASTM E399 74 with appropriate modifications necessary for dynamic tests. Testing dynamically in the fracture toughness ductile to brittle transition region and at upper shelf initiation temperatures results in not only lower bound data but also provides an opportunity for obtaining validlil fracture toughness data up to the onset of upper shelf. This results from nonlinear cleavage behavior which occurs only in dynamic testing at these temperatures. The load-displacement curve exhibits a clear drop in load at the onset of crack initiation, thereby eliminating any possible doubt as to the start of crack initiation, as is the case in static loading conditions at these temperatures. Recom-mended test temperatures are equal to or lower than those characteristic of the upper tracture toughness shelf initiation temperature.

1. Rccarcetta, P. C. and Swediow, J. L., "A Combined AnalytcabExpenrnental Fracture Study of the Two leading Theones o Elaste Plaste Fracture (J4ntegraland Equivalent Energy)," HSST TR 33, WCAP 8224. october 1973. l 4-2 1

% n

-p l

1 -

An lysis should bo p rform:d using the J. Integral or Equivalont Energy Concept.[t21 Testing at temperatures characteristic of the fracture toughness bpper shelf is not suggested due to the uncertainty of the point of crack initiation even when dynamic l[ testing is performed. At these temperatures, static Jic testing appears to be most indica-tive of conservative upper shelf fracture toughness properties. Research in this area is currently being conducted by Westinghouse Research and Development Laboratory, ASTM E24, NRC, and others. Use of this technique will be further evaluated as it applies to surveillance specimen testing.

4 5. POSTIRRADIATION TEST EQUIPMENT Required minimum equipment for the postirradiation testing operations is as follows:

E Milling machine or special cutoff wheel for opening capsules, dosimeter blocks and spacers B Hot cell tenslie testing machine w'th:

J_ 1) pin type adapter for testing tensile specimens n

Q E Hot cell dynamic CT testing machine with clevis and appropriate measuring

.*f equipment associated with dynamic testing M Hot cell Charpy impact testing machine E Sodium iodide scintillation detector and pulse height analyzer for gamma counting of the specific activities of the dosimeters 1

Riccardella. P. C. and Swediow, J. L. "A Combined Analytical-Expenmental Fracture Study of the Two Leading Theones of Elastic Plastic Fracture (J. integral ana Eouivalent Energy)," HSST TR-33, WCAP 8224. October 1973

2. Buchalet. C. and Mager, T. R , "Fxpenmental Venfication of Lower Bound K Values Utilaing the Equivalent Energy Concept,"in Progress sn Flaw Grovyth and Fracture Toughness Testing, ASTM STP 536. pp. 281296. Amencan Society for Testing and Matenals, Philadelphia.1973 43

-' JA

h R

l l APPENDIX A CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL SURVEILLANCE MATERIAL P For the reactor vessel radiation surveillance program, Rotterdam Dockyard Company supplied Westinghouse with sections of SA508 Class 2 forging used in the core region of the Catawba Unit No.1 reactor r ressure vessel, specifically, from the 8% inch intermediate shell course forging 05 of the pressure vessel. Also supplied was a weld-ment made from sections of intermediate shell forging 05 and lower shell course forging 04, using weld wire representative of that used in the original fabrication for the closing girth weld seam between the intermediate and lower shell courses. The surveillance weldment is identical to the closing girth seam weldment between forgings 04 and 05.

The closing searn used weld wire heat no. 895075 with Type Grau L.O. (LW320) flux,

. IG ~ 46, except for the 1 inch root pass at the ID of the vessel. This root pass used weld wire of heat no. 899680 with type Grau L.O. (LW320) flux, lot P23, with an as-deposited copper and phosphorous content of 0.03 and 0.009, respectively. The surveillance specimens were not removed from this root area. The forgings were produced by Klockner Werge AG The heat treatment history and the chemical analysis of the pressure vessel surveillance materials are shown in tables A 1 and A 2, respectively.

TABLE A 1 HEAT TREATMENT HISTORY Temperature Time Material (*F) (hr) Cooling Intermediate shell 1679 1697 3% Water quenched forging 05 1220 1247 6 Air-cooled 1140i 25 22 Furnace cooled Weldment 1140 t 25 15 Furnace cooled A1 i ~~~~-, _

I l

TABLE A 2 CHEMICAL ANALYSIS OF MATERIALS l

l Content in Indicated Material (weight %)

Element Intermediate shell W old Forging ostal Metal (a)

C .20 .049

.73 1,73 Mn P .014 .015 S .004 .006 SI .35 .27 Ni 34 .71 Mo .53 .56 Cr .38 .036 Cu .10 .066 Al .046 .025 Co .024 .013 Pb .001 .003 W .010 .008 TI < .001 .002 Zr .002 .002 V .003 .002 Sn .005 .001 As .023 .011 Cb < 002 <.002 N2 .009 .008 Zn .010 .010 Mg .002 .002 Ag .001 <.0005 B <.0005 < 0005 _

a. Analysis conducted by Westinghouse A2 1

J

l Attachment 7 Westinghouse Electric Corporation Topical P.eport WCAP-10868, Duke Power Company Catawba Unit 2 Reactor Vestel Radiation Surveillanco Program.

9709220008 970915 PDR ADOCK 05000413 P PDR

l WCAP.10868 WESTINGHOUSE CLASS 3 DU 't' '

hIl. MARo h%

O "s,. I U cd,/'?'"S .

M J'~ Qig3',

.#gy Wf DUKE POWER COMPANY CATAWBA UNIT NO. 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM Lob ccky I DOCUMENT,

. CONTROL: DATE L. R. Singer FEB 2 41986.'

DUKE POWER COMPA83 DE5lGN ENGINEtluND APPROVED: 3 4.3%W T. A. Meyer, Mana'ger Structural Materials and Reliability Technology

~~~~

Work Performed Under DDPJ 106 I"gy~ LEU 4 tgR 1 81080

$ $ [$.'"

WESTINGHOUSE ELECTRIC CORPORATION '

Nuclear Energy Systems ~ ~ ~ ~ - -

P. O. Box 355 Pittsburgh, Pennsylvania 15230 CNN 1201 ro 0- 0110 '" 00t*O

I PREFACE This report has been technicaly reviewed and checked by S. E. Yanichko of Structural Materials and Rollability Technology.

M S.

a

. Yanichko Date: November 5,1985 ,

5 9

ill a

AB8 TRACT 1

A pressure vessel steel surveilley-e program per ASTM E 18542 has been developed for l the Duke Power Company, Carmtw Valt No.2 to obtain information on the effects of radia-tion on reactor pressure vessel medy m der operating conditions. The radiation surveillance program for the Catawba Unit No. 2 !s oesigned to, and in compliance with, federal govern-('

ment regulations identified in appendix H to 10CFR, part 50 entitled " Reactor Vessel Material Surveillance Program Requirements."

Following is a description of the program, a description of the material involved, the specimen and capsule design and fabrication, and the preirradiation test results. ,/

4 v

i TABLE OF CONTENTS Section Title Page 1 PURPOSE AND SCOPE 11 2 CAPSULE PREPARATION 21 l 21. Pressure Vessel Material 21 l 2 2. Machining 21 l 2 3. Charpy V notch impact Specimens 23

24. Tenslie Specimens ,

23 2 5. 1/2T Compact Specimens 23

26. Dosimeters 23 2.7. Thermal Monitors 23 2.8. Capsule Loading 29 l

3 PREIRRADIATION TESTING 31

31. Charpy V notch Tests 31 3 2. Tensile Tests 31 3-3. Dropweight Tests 32 4 POSTlRRADIATION TESTING 4-1 4-1. Capsule Removal 41 4-2. Charpy V-notch Impact Tests 42 4-3. Tensile Tests 42 4-4. Fracture Toughness Tests on 1/2T Compact Specimens 42

. 4-5. Postirradiation Test Equipment 4-3 Appendix A DESCRIPTION AND CHARACTERIZATION,OF THE CATAWBA UNIT NO. 2 REACTOR VESSEL BELTLINE AND SURVEILLANCE MATERIALS A1 4

vil

LIST OF ILLUSTRATIONS l

Pigure Title Page 11 Location of the Irradiation Test Capsules in the Catawba Unit No. 2 Reactor Vessel 1-4 21 Charpy V notch Impact Specimen 22

^

22 Tensile Specimen 2-4 2-3 Compact Specimen 25 24 Irradiation Capsule Assembly 2 7/2-8 25 Dosimeter Block Assembly 2 10 26 Specimen Locations in the Catawba Unit No. 2 Reactor Surveillance Test Capsules- 213/214 31 Preirradiation Charpy V-notch impact Energy for the Catawba Unit No. 2 Reactor Pressure Vessel Intermediate Shell Plate B86051 (Longitudinal Orientation) S9 S2 Proirradiation Charpy V-notch Impact Fnergy for the Catawba Unit No. 2 Reactor Pressure v'essel Intermediate Shell Plate B86051 (Transverse Orientation) S9 33 Proirradiation Charpy V notch impact Energy for the Catawba Unit No. 2 Reactor Pressure Vessel Core Region Wold Metal 3-10 3-4 Preirradiation Charpy V-notch impact Energy for the Catawba Unit No. 2 Reactor Pressure Vessel Core Region Wold Heat-Affected Zone Material 3 10 35 Proirradiation Tensile Properties for the Catawba Unit No. 2 Reactor Pressure Vessel Intermediate Shell Plate B88051 (Longitudinal Orientation) . 3 11 36 Preirradiation Tensile Proporties for the Catawba Unit No. 2 Reactor Pressure Vessel Intermediate Shell Plate B86051 '

(Transverse Orientation) 3 12 S7 Preirradiation Tensile Properties for the Catawba Unit No. 2 Reactor Pressure Vessel Core Region Wold Metal 3 13 3-8 Typical Stress-Strain Curve for Tensile Test 3 14 lx

LIST OF TABLES

. Table Title Page 21 Type and Number of Specimens in the Catawba Unit No.

2 Surveillance Test Capsules 29 22 Quantity of isotopes Contained in the Dosimeter Blocles 2 11 31 Preirradiation Charpy V notch Impact Data for the i Catawba Unit No. 2 Reactor Pressure Vessel Intermediate Shell Plate B86051 (Longitudinal Orientation) 33 1 32 Preirradiation Charpy V notch Impact Data for the 4

Catawba Unit No. 2 Reactor Pressure Vessel

! Intermediate Shell Plate B86051 (Transvorse Orientation) 3-4 33 Preirradiation Charpy V notch Impact Data for the Catawba Unit No. 2 Reactor Pressure Vessel Core Region Weld Metal . 3-5 34 Preirradiation Charpy V notch Impact Data for the Catawba Unit No. 2 Reactor Pressure Vessel Core Region Weld Heat Affected Zone Material 3-6

. 35 Summary of the Catawba Unit No. 2 Reactor Pressure

! Vesse! Impact Test Results for Intermediate Shell Plate B8605-1 and Core Region Weld and Heat Affected Zone Material 37 1 3-6 Preirradiation Tenslie Properties for the Catawba Unit No.

2 Reactor Pressure Vessel Intermediate Shell Plate B8605-1 and Core Region Weld Metal 3-8 4-1 Surveillance Capeule Removal Schedule 41 A1 Chemical Analysis of the Intermediate Shell Plates used in the Core Region of the Catawba Unit No. 2 Reactor Pressure Vessel A-2 A2 Chemical Analysis of the Lower Shell Plates used in the Core Region of the Catawba Unit No. 2

Reactor Pressure Vessel A3

, A-3 Chemical Analysis of the Weld Metal used in the Core Region Weld Seams of the Catawba l Unit No. 2 Reactor Pressure Vessel A-4 :

A-4 TNDT,RTN DT and Upper Shelf Energy for the !

Catawba Unit No. 2 Reactor Pressure Vessel Core Region Shell Plates and Weld Meta! A-5 i

A5 Heat Treatment History of the Catawba Unit No. 2 Reae.

tor Pressure Vessel Core Region Shell Plates and Weld Seams A6 xi i

SECTION 1 i PURPOSE AND SCOPE I l

The purpose of this program is to monitor radiation effects under actual operating con-ditions of the core region reactor vessel materials in the Duke Power Company, Catawba Unit No. 2, a fourloop, nuclear power plant with a thermal output rating of 3427 megawatte. Evaluation of the radiation effects is based on preirradiation testing of Charpy V notch, tensile, and dropweight specimens, and postirradiation testing of Charpy V notch, tensile, and compact specimens.

Current reactor pressure vessel material test requirements and acceptance standards utilize the reference nil-ductility temperature, RTNDT, as a basis. RTNDT is determined from the dropweight nil-ductility transition temperature (TNDT) p6r ASTM E208 and the weakI 'l direction 50 ft Ib Charpy V notch temperature (or the 35-mil lateral expan-sion temperature if it is greater). RT NDT is defined as the dropweight TNDT or the temperature 60*F less than the 50 ft Ib (or 35 mil) Charpy V notch temperature, whichever is greater.

Therefere RTNDT = TNDT, if TNDT 4T50(35) - 60*F and RTNDT = T50(35) - 60*F, if T50(35) - 60aF > TNDT RNDT = Reference nilductility temperature TNDT = Nil-ductility transition temperature per ASTM E208 T = 50 ft Ib temperature from Charpy V notch specimens oriented 50(35) in the weak direction (or the 35-mil temperature if it is greater) 1, Longetudinal axis of the specimen oriented normal to the major wortdng directkm of the plate.

11

b J(

An emperical relationship betwoon RTNOT and fracture toughness for reactor vessel A' steels has been developed in Appendix G, " Protection Against Non ductile Failure," to 1 Section lli of the ASME Poller and Pressure Vesse! Code. This relationship can be empicyd to set allowable pressure temperature limitations for normal operation of '

reactors which are based on fracture mechanics concepts. Appendix G defines an acceptable method for calculating these limitations, it is known that radiation can shift the Charpy V notch impact energy curve to higher temperatures, ll and thus cause the RTNDT to increase with radiation exposure. The extent of the shift in the impact energy curve, that is, radiation embrittlement, is enhanced 1 by certain chemical olomonts (such as copper) present in reactor vessel steels.I8'*3 l The adjustment in RTNDT with service can be monitored by a surveillance program ;

involving periodic checking of Irradiated reactor vessel surveillance specimens. The sur-veillance program is based on ASTM E185-82 (Standard Practice for Conducting Sur.

veillance Tests for Light Water Cooled Nuclear Power Reactor Vesseis), Compact fracture l mechanics specimens will be used in addition to Charpy V-notch specimens to evaluate the effects of radiation on the fracture toughness of reactor vessel rnatorials. ,

Postirradiation testing of the Charpy V-notch impact specimens will provide a guide for determining prosaure temperature limits on the plant. Charpy impact test data wi!I deter- >

l mine the shift of the reference temperature *l with radiation exposure at plant tempora;ures,

a. The reference temperature as defined by 10CFR Part 50, Appendix G, Section I! E is as follows:

" Adjusted reference temperature" means the reference temperature as adjusted for irradiation effects by adding to RTNOT the temperature shift, measured at the 30 ft Ib (41 J) level.

. 1 Porter. L F., "Radiablen Effects in Steel," in Mosenede h Nuodear p ASTM-STP 276, pp.147196, Amencen

, Soceoly for Testing and Mater 6elo, PNiedelphes,1900.

2. Steele, L E. and Hawtheme, J. R.,"New iniormellon on Neutron Emtunellement and Emtwttelement Mellet of Remotor Pressure Veessi Steels," NRL 4100. August 1984.

3. Potapovo, U. and Hawthome, J. R.,"The Ellect of Reeldual Elemente on 500*F treedletion Meeponse of Seisoted Prosaure Vessel Stooie and Weidments," NRL 4003, Septemtier 1988.

4. Sleses, L E., " Structure and compoelton Eflects on irredletion SenenNety of Pressure Vessel Stesis," in irradioman f aects on Stucturef ANoys for Nuoiser Reestor t _- . ASTM-STP 444, pp.164175, Ame doen Society for Testing and Meterteis.

PNiedelpNo.19&

12 4

b

These data can then be reviewed to verify or revise pressure temperature limits of the vessel during heatup and cooldown and will allow a check of the prodded shift in the reference temperature. The postirradiation test results of the compact specimens will l

provide actual fracture toughness properties of the vessel material. These properties may ,

be used to establish allowable stress intensity factors for subsequent analyses, I Six material test capsules are fabricated containing specimens from the reactor vessel shell plate identified as being most likely to limit the operation of the reactor vessel.

The specimens contained in the Catawba Unit No. 2 test capsules are from the in-termediate shell plate of the reactor vessel and representative weld metal and heat-affected zone (HAZ) metal.

The thermal history or heet treatment given these specimens is similar to the thermal history of the reactor vessel material with the exception that the postweld heat treatment received by the specimens has been simulated (Appendix A).

The six mateIlal test capsules are then installed in the reactor in guide tubes attached to the neutron shield pads which are located in the reactor between the core barrel and the reactor vessel wall opposite the conter of the core as shown in Figure 11.

1

\

13 4

O' i 1 REACTOR VESSEL i '

CORE BARREL NEUTRON PAD 4

(301.5 ') Z CAPSULE U (58.5')

M"58,5 ' V (61 ')

58.5' % L )

61' .

f d1 1

I 270' - 90' f

(241') Y ] [

f (238.5') X

REACTOR VESSEL 180' PLAN VIEW l

9 WALL Q[ VESSEL

', ( CAPSULE

CORE $

s OIII!IlllI

[

CORE Q MIDPLANE t

{' a %

FIGURE 1 1, N i

LOCATION OF THE IRRADIATION TEST CAPSULES IN THE

( '

' R s

NEUTRON PA CATAWBA UNIT NO. 2 ) CORE BARRE REACTOR VESSEL f ELEVATION VIEW 14 v

)

SECTION 2 CAPSULE PREPARATION 2 1. PRESSURE VESSEL MATERIAL Reactor vessel material was supplied by Combustion Engineering, Inc. from interme-dlate shell plate B86051, Heat No. C05431. Combustion Engineering, Inc., also supplied a weldment which joined sections of material of the Intermediate shell plate B8605 2 (See Note) and the adjacent lower shell plate B88061. Heat No. C22881.

Data on the limiting core region plate (B86051), weld, and weld heat affected zone material are provided in Appendix A.

Note: The limiting material for the Catawba Unit No. 2 reactor vessel beltline region is intermediate shell plate B86051. This is based on the highest ARTNDT shift (94'F) as calculated using the latest ASTM revisions.

The original material selected in 1978 was intermediate shell plate B8605 2. This selection was based at the time on the highest initial RTNDT. Therefore weld test plate "D" furnished to Westinghouse at that time was made up of plates B8605 2 and B88061.

2 2. MACHINING Test material obtained fn~n the intermodlate shell plate (after the thermal heat treat-ment and forming of the plate) was taken at least one plate thickness from the quenched ends of the plate. All test specimens were machined from the % and % thickness location of the plate after performing a simulated postweid, stress-relieving treatment o

' n the test material and also fiom wold and heat-affected zone metal of a stress relieved weldment joining intermedia's shell plate B8605-2 and adjacent lower shell plate B88061. All heat affected zone specimens were obtained from the weld heat affected zone of intermediate shell plate B8605 2.

21 f

A

S i

i 46' 44' l

l f 1 I

0.011 0.009" i

I 0.395 90' 10' O.363 898 to'

, y l v

n y==j . 0. m ----

0.303

U f d 0.314 -

0.314 : 1.043 1.063 2.125 =

2.105 ,

ALL Ovtt UNLits OTN(RWISC $P(ClFIED Figure 21, Charpy V notch Impact Specimen 22

2.3 Charpy V notch impact Spoolmene Charpy V notch impact specimens corresponding to ASTM A370 Type A (Figure 21) were machined from intermediate shell plate 886051 in both the longitudinal orientation (longitudinal axis of specimen parallel to major rolling direction) and transverse orientation (longitudinal axis of specimen normal to major rolling direction). The core region weld Charpy impact specimens were machined from the weldment such that the long dlmon-sion of the Charpy specimen was normal to the weld direction. The notch was machined l such that the direction of crack propagation in the specimen was in the welding direction.

2-4. Tenelle Specimens Tensile specimens (Figure 2 2) from shell plate B86051 were machined in both the longitudinal and transverse orientation. Tensile specimens from the weld were oriented normal to the welding direction.

2-5. 1/2T Compact Spoolmens Compact test specimens (Figure 2-3) from shell plate B86051 were machined in both the longitudinal and transverse orientations. Compact test specimens from the weld metal were machined with the notch oriented in the direction of welding. All specimens were fatigue procracked according to ASTM E399,

24. DOSIMETERS Each of the six test capsules of the type shown in Figure 2 4 contain dosimeters of copper, iron, nickel and aluminum 0.15 weight percent cobalt wire (cadmium shielded and unshielded) and cadmium-shielded Np* and U" which will measure the integrated flux at specific neutron energy levels.

2 7. THERMAL MONITORS The capsules contain two low-melting-point eutectic alloys to more accurately define the maximum temperature attained by test specimens during irradiation. The thermal monitors are sealed in Pyrex tubes and then inserted in spacers located as shown in Figure 2 4. The two autoctic alloys and their molting points are the,following:

2.5 percent Ag,97.5 percent Pb- Melting point: 304*C (579'F) 1.5 percent Ag,1.0 percent Sn,97.5 percent Pb Molting bint: 310*C (590*F) 23

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Figure 2 4. Irradiation Capsule Assembly From Westinghouse Dwo 1453E10 m_J '

0 -

- . 2 7/2 8 l

(, l

_a mi

. 28. CAPSULE LOADING The six test capsules coded U, V, W, X, Y, and Z are positioned in the reactor between the neutron shleiding pads and vessel wall at the locations shown in Figure 2-4. Each capsule contains 60 Charpy V notch specimens,9 tensile specimens and 12 compact specimens. The relationship of the test material to the type and number of specimens in each caps Jle is shown in Table 21.

TABLE 21 TYPE AND NUMBER OF SPECIMENS IN THE CATAWBA l UNIT NO. 2 SURVEILLANCE TEST CAPSULES l

Capsules U, V, W, X, Y, and Z Material Charpy Tensile Compact Plate B86051 15 Longitudinal ise.c=.n. e.m c.p.*> 3 4 Transverse 15 3 4 Weld Metal 15 3 4 HAZ 15 - -

Dosimeters of copper, iron, nickel, aluminum 0.15 weight percent cobalt, and cadmium-shielded aluminum cobalt wires are secured in holes drilled in spacers located at capsule positions shown in Figure 2-4. Each capsule also contains a dosimeter block (Figure 2 5) located at the center of the capsule. Two cadmium-oxide shielded tubes, one containing an isotope of U 28 and the other an isotope of Np237 , are located in the dosimeter block.The double containment afforded by the dosimeter assembly prevents loss and contamination by the U23e and Np 237 and their activation products. Each dosimeter block contains approximately 12 milligrams of U 2aa and 17 milligrams of Np237(Table 2 2) held in a 3/s-inch long by % inch outside diameter sealed stainless steel tube, respectively. Each tube was placed in a % inch-diameter hole in the dosimeter block (one U23e and one Np237 tube per block), and the space around the tube was 29

98 . ~.

.-l* <

5 lI ll

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a

. , a g .~e ,.. 3 3

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filled with cadmium oxide. After placement of this material, each hole was blocked with two 1/irinch thick aluminum spacer discs and an outer 1/rinch thick steel cover disc welded in place.

The numbering system for the capsule specimens and their locations is shown in Figure 2 6. The specimens are seal welded into a square capsule of austenitic st.alnless steel to prevent corrosion of specimen surfaces during Irradiation. The capsules are hydro-I statically compressed in domineralized water to collapse the capsule on the specimens so that optimum thermal conductivity between the specimens and the reactor coolant is obtained. The capsules are then leak tested with helium after pressurization and then i dye penetrant tested as a final inspection procedure. Fabrication details and testing i procedures are listed in Figure 2-4.

TABLE 2 2 QUANTITY OF ISOTOPES CONTAINED IN THE DOSIMETER BLOCKS l

1

lootope Weight (mg) Compound Weight (mg) 237 Np 17 i i NpO, 20 t 1 Ue 23 12.0 U3 0, 14.25 t

2 11

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- - - - - _ .m

= w. De we m me. av DDP - - - - - - --

Z - - = = ~ ~ ~ ~ ~ - - = = = = ~

z =,.

m.

=

==

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NN DDP - - - - - - --

Y y

- ~ ~ ~ ~"

- = = = =a -

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x =

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=.

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== 0,

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e, = D,<n =,o we m w, ..

LEGEND:DL . INTERMEDIATE SHELL PLATE B8605-1 (LONGITUDINAL) .

DT INTERMEDIATE SHELL PLATE B86051 (TRANSVERSE)

DW . WELD METAL

, DH . HEAT AFFECTED-ZONE MATERIAL

/

. - . . .s ,

l t

n

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=.i i au mn an a e Mc an m. == m., a., no I

met mes m.e att ot. Eso m7 ass atu au otst asi ota mes ofta m, ? u. au m. a. m. a nu an of. an a m., mi. ma m .. ma en

_._ __. _ ___ _ _=._ _ _ _._ _ __

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_= i ._

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l,

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r lo, =i m, ai ma au m. u on as m. a. mi ai on l

Figure 2-6. Specimen Location in the i

ANSTEC Catawba Unit No. 2 APERTURE Reactor survemance Test capsuies CARD .

213/214 Also Availablo ort Aperture Card 97o9z #

' ~

,.- p

1 i

SECTION 3 PREIRRADIATION TESTING

31. CHARPY V NOTCH TESTS Charpy V notch impact tests were performed according to ASTM E23 with specimens from the vescsl intermediate shell plate B86051. Specimens of both longitudinal and transverse orientations were tested at various test temperatures in the range from - 62*C to 160*C (- 80*F to 320'F), yielding a full Charpy V notch transition temperature curve in both orientations (Tables 31 and 3 2 and Figures 31 and 3 2). Tests were also performed on the weld metal and HAZ metal at various temperatures from - 118'C to 160*C (- 180'F to 320*F) and are shown in Tables 3-3 and 3-4 and Figures 3-3 and 3-4.

A summary of the Charpy V-notch impact tests results including upper shelf energy (USE),

t 41. joule (30 ft Ib), 68-joule (501t Ib), and 35 mils (0.89mm) lateral expansion index temperatures are presented in Table 3 5.

l The specimens were tested on a Sontag Universal Model Number SI 1 Impact machine with a hammer energy capacity of 240 foot pounds and a striking velocity of 17 feet per second. The machine is calibrated every 6 months using Charpy V notch impact specimens of known energy values supplied by Watertown Arsenal. Specimen condi-tioning for high temperature testing is maintained ; sing a Fisher ISO Temperature Oven, Model 350. For low temperature specimen conditioning either liquid nitrogen or dry ice in isopropanol is used. The specimen temperatures are monitored by use of a "J" type thermocouple or a thermometer.

3 2. TENSILE TESTS Table 3-6 and Figures 3-5,3-6, and 3-7 show the results of tensile tests (per ASTM E8

'and E 21 test criteria) from vessel intermediate shell plate B8605-1 and from the weld metal. Specimens from plate B8605-1 and the weldment were tested at 24*C (75'F),

149'C (300'F) and 288'C (550'F) in both the longitudinal and transverse directions.

3-1

An Instron Universal tensile testing machine Model TTD (20K 50K) was used with an instron load cell (Serial number 059SN and 044SN) which is calibrated daily and verified annually to the National Bureau of Standards. The gripping mechanism utilizes thread- (

od adapters to pull rods attached to the cross head / load cell and frame. The recording device utilizes an Instron Model 3124 strip chart in console calibrated to the Instron Class B-1 extensometer, Model 4929. The extensometer is calibrated by test equipment which has been certified by the National Bureau of Standards. The measurement and control of speeds in the tests conform to ASTM A370 77 (Mechanical Testing of Steel Products).

A typical stress strain curve is shown in Figure 3 8.

33. DROPWilGHT TESTS The nil ductility transition temperature (TNDT) was determined for plate B86051 and the core region weld metal and heat affected zone by dropwoight tests (ASTM E 208) l performed at Combustion Engineering, Inc. From this test data the RTNDT was calculated using the methods as described in Section 1. The TNOT and RT NOT for intermediate l shell plate B8605-1, well metal and heat affected zone (HAZ) are as follows:

l Note: TNOT and RTNOT for .N the bestNne shou pielee le gN.n in Appendis A.

i Material TNOT (*F) RTNDT ('F)

Plate B8605-1 - 10'I l

+ 15 1

WelM Metal one.rm.smo and to.or snes - 80163 - 80 wr m s mo and a6.was oinn s m>

HAZ - 80'I l

- 80

a. Combustion Engineering Materials Certification Report,

b. Combustion Engineering Welding Material Qualification Test.

.c. Combustion Engineering Surveillance Wold Test Plate "C" Materials Test Report f'

32

e TABLE 31 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE B86051 (LONGITUDINAL ORIENTATION) l l

Temperature impact Energy Lateral Expansion Shear

('C) (* F) (J) (ft Ib) (mm) (mils) (%)

- 62 - 80 5.5 4.0 0.03 1.0 3

- 62 - 80 8.0 6.0 0.08 3.0 3

- 40 - 40 18.0 13.0 0.20 8.0 9

- 40 - 40 26.0 19.0 0.36 14.0 9

- 40 - 40 37.0 27.0 0.51 20.0 14

- 18 0 43.0 32.0 0.46 18.0 25

- 18 0 51.5 38.0 0.79 31.0 30

- 18 0 57.0 42,0 0.69 27.0 25

- 7 20 69.0 51.0 1.12 44.0 34

- 7 20 69.0 51.0 0.97 38.0 29

- 7 20 95.0 70.0 1.22 48.0 23 4 40 91.0 67.0 1.09 43.0 40 4 40 111.0 32.0 1.45 57.0 48 4 40 111.0 82.0 1.52 60.0 45 27 80 104.0 77.0 1.35 53.0 56 27 80 117.0 86.0 1.45 57.0 50 27 80 123.0 91.0 1.52 60.0 54 49 120 115.0 85.0 2.01 79.0 81 49 120 118.0 87.0 1,80 71.0 75 49 120 142.0 112.0 1.83 72.0 77 66 150 159.0 117.0 2.01 79.0 94

, 66 150 182.0 . 134.0 2.13 84.0 100 82 180 180.0 133.0 2.13 84.0 100 82 180 183.0 135.0 2.24 88.0 100 82 180 192.5 142.0 2.24 88.0 100 1

116 240 183.0 135.0 2.24 88.0 100 116 240 188.5 139.0 2.18 86.0 100 160 320 188.5 139.0 -2.16 85.0 100 160 320 197.0 145.0 2.18 86.0 100 3-3

L TABLE 3 2 PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE CATAWBA UNIT NO 2 REACTOR

[- PRESSURE VESSEL INTERMEDIATE SHELL PLATE 88605-1 (TRANSVERSE ORIENTATI L

Temperature Imaad Energy Lateral Expansion

('C) Shear

('F) (J) (ft Ib)

- 62 (mm) (mils) (%)

1

- 80 7.0 j s

- 62 5.0 0.05 l - 80 8.0 2.0 3

- 40 0.0 0.13 i

- 40 12.0 5.0 3

- 40 9.0 0.15

- 40 22.0 6.0 13

- 40 16.0 0.25

- 40 24.5 10.0 18

- 18 18.0 0.33 0 46.0 13.0 9

- 18 34.0 0.53 i 0 47.5 21.0 29 35.0

- 18 0 O.69 27.0

( 4 40 56.0 47.5 41.0 0.89 35.0 25 29 4 35.0 0.69 40 70.5 27.0 38 4 52.0 1.04 40 73.0 41.0 43 '

27 54.0 1.02 80 80.0 59,0 40.0 34 27 80 1.19 47.0 87.0 64.0 41 1.22 27 80 48.0 96.0 71.0 44 38 100 1.27 50.0 72.0 53.0 45 38 100 1.09 43.0 89.5 66.0 47 38 100 1.27 50.0 104.5 77.0 55 49 120 1.37 54.0 -

89.5 6f).O 59 49 120 1.19 47.0 114.0 84.0 62 49 120 1.55 61.0 127.5 94.0 68 82 180 1.78 70.0 125.0 92.0 100 82 180 1.60 64.0 130.0 96.0 100 82 180 1.83 72.0 134.0 99.0 100 116 240 1.80 71.0 136.0 1GO.0 100 116 240 138,0 1.80 71.0 102.0 100

.160 320 - 1.98 78.0 114.0 84.0 100 160 320 1.65 65.0 136.0 100.0 100 1.90 75.0 100 M

l

?

l l

TABLE 3-3 PREIRRADIATION CHARPY V NOTCH IMPACi DATA FOR THE CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL CORE REGION WELD METAL 1

l Temperature impact Energy Lateral Expansion Shear

('C) (* F) (J) (ft Ib) (mm) (mils) (%)

- 96 -?40 5.5 4.0 0.03 1.0 9

- 96 - 140 7.0 5.0 0.05 2.0 13

- 62 - 80 8.0 6.0 0.03 1.0 13 '

- 62 - 80 11.0 8.0 0.08 3.0 18

- 62 - 80 35.0 26.0 0.28 11.0 18

- 51 - 60 15.0 11.0 0.15 6.0 28

) - 51 - 60 20.0 15.0 0.25 10.0 33

- 51 - 60 20.0 15.0 0.15 6.0 28

. - 40 - 40 62.0 46.0 0.79 31.0 47

- 40 - 40 79.0 58.0 1.04 41.0 40

- 40 - 40 99.0 73.0 1.30 51.0 52

- 18 0 79.0 58.0 1.14 45.0 65

- 18 0 130.0 96.0 1.52 60.0 73

- 18 0 137.0 101.0 1.83 72.0 71 4 40 164.0 121.0 2.01 79.0 96 4 40 169.5 125.0 1.98 78.0 93 4 40 183.0 135.0 2.03 80.0 84 27 80 178.0 131.0 2.01 79.0 93 27 80 187.0 138.0 2.24 88.0 96 27 80 199.0 147.0 2.18 86.0 94 49 120 192.5 142,0 2.24 88.0 100 49 120 198.0 146.0 2.24 88.0 100 49 120 205.0 1 E1.0 2.21

87.0 100 104 220 188.5 139.0 2.18 86.0 100 104 220 201.0 148.0 2.31 91.0 100 160 320 206.0 152.0 2.29 '90.0 100 6~ 160 120 222.0 164.0 2.18 86.0 100 3-5

TABLE 3-4 ,

PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE CATAWBA UNIT NO. 2 REACTOR PRESSURE

, VESSEL CORE REGION WELD HEAT AFFECTED-ZONE MATERIAL Temperature impact Energy Lateral Expansion Shear

('C) (*F) (J) (ft Ib) (mm) (mils) (%)

- 118 - 180 9.5 7.0 0.08 3.0 3

- 118 - 180 12.0 9.0 0.05 2.0 3 l - 118 - 180 15.0 11.0 0.05 2.0 9

- 84 - 120 20.0 15.0 0.05 2.0 10

- 84 - 120 31.0 23.0 0.18 7.0 29

- 84 - 120 38,0 28.0 0.25 10.0 25

- 62 - 80 12.0 9.0 0.13 5.0 29

- 62 - 80 S8.0 28.0 0.36 14.0 29

- 62 - 80 84.0 62.0 0.74 29.0 43

- 51 - 60 37.0 27.0 0.43 17.0 32 1

- 51 - 60 79.0 58.0 0.86 34.0 50

- 51 - 60 90.0 66.0 0.86 34.0 47

- 40 - 40 72.0 53.0 0.79 31,0 56

- 40 - 40 108.5 80.0 1.19 47.0 59 ,

I

- 40 - 40 127.5 94.0 1.37 54.0 68

- 18 0 138.0 102.0 1.55 61.0 73

- 18 0 142.0 105.0 1,42 56.0 90

- 18 0 183.0 135.0 1.78 7.0.0 100 4 40 167.0 123.0 2,06 81.0 100 4 40 186.0 138.0 1.98 78.0 100 4 40 201.0 148.0 2.03 80.0 100 27 80 163.0 120.0 1.78 70.0 100 27 80 174.0 128.0 1.80 71.0 100 27 80 197.0 145.0 1.90 75.0 100 60 140 184.5 136.0 2.08 82.0 *100 60 140 207.5 153.0 1.98 78.0 100 93 200 176.0 130.0 2.03 80.0 100 93 200 199.0 147.0 2.01 79.0 100 3-6

TABLE 3-5

SUMMARY

OF CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL IMPACT TEST RESUL1S FOR INTERMEDIATE SHELL PLATE B8605-1 AND CORE REGION WELD AND HEAT-AFFECTED-ZONE MATERIAL Upper Shelf 41-J 68-J 0.89 mm Energy (30-ft Ib) (50-ft Ib) (35 mils)

Material (USE) Index Temp Index Temp index Temp (J) (ft Ib) (*C) (*F) ('C) ('F) ('C) (*F)

Plate B86051 (Longitudinal 187 138 - 26 -15 -7 20 -9 15 l

Orientation)

Plate B8605-1 (Transverse 130 96 - 21 -5 4 40 4 40 Orientation) ,

Weld 202 149 - 46 - 50 - 40 - 40 - 34 - 30 Heat Affected 184.5 136 - 71 - 95 - 54 - 65 - 43 - 45

- Zong 3-7

TABLE 34 PRENWtADIATIOtt TEfe88LE PROPERTIES FOR THE '

CATAWBA UNIT 900. 2 REACTOR PRESSURE VESSEL INTEntdED8 ATE SHELL PLATE 38805-1 AND CORE REGION WELD 00ETAL Rossuctsen o.2% tailmene in Fracture Fracture Fracture UnIIenn Teest Test Ylead Tenene Strees Serength Enlongselon Eniengselon Aree T- n;; Strength Strength Lead tietortel l tape) (%) (%) (%)

C *F pes) (edPs) pel) (ISPs) (Idpl (98) pel) (ISPs) pel) 180A 1303.0 56.0 386.0 16.0 31.0 71A 24 75 68 0 489.0 90.0 620.5 2.8 12.454 Piste B8605-1 13314 57A 383.0 16.0 30.0 71 4 71.0 489.5 80.0 814 0 2.8 12.454 193.0 (Longstudinal 24 75 13.0 24.0 71 A 83.0 572 0 2.7 12.010 184.0 1200.0 54.0 372.0 Orientation) 149 300 62.0 427.5 69.0 180.0 1241.0 55.0 3790 13.0 24.0 149 300 62.0 427.5 83.0 572.0 2.7 12.010 176 0 1213.0 61A 421A 13.0 23.0 86 0 288 550 61.0 421.0 88 0 807.0 3.0 13.344 g 288 550 52.0 358.5 88.0 607.0 30 13.344 176.0 1213 0 50 0 407.0 14.0 24A 67.0 0o 167.0 1151.0 80.0 414.0 16.0 27.0 64.0 Piste 88605-1 24 75 67.0 462.0 90.0 620.5 3.0 13.344 170.0 1172.0 64.0 4414 16.0 26.0 62A 24 75 67.0 462.0 88.0 807.0 3.0 13.344 (Transverse 153.0 1054.0 58.0 4004 13.0 23.0 62.0 149 300 61.0 421.0 82.0 565 0 2.9 12.899 Orientation) 150.0 1096.0 50.0 407.0 14.0 24.0 83.0 {

149 300 61.0 421.0 82.0 565.0 2.9 12.899 167.0 1151.0 87.0 462.0 15.0 22A 80.0 288 550 80.0 414.0 86.0 593.0 33 14.678 56.0 150A 1034.0 67A 462A 14A 21.0 288 550 60.0 414.0 87.0 600.0 3.3 14,.678 198.0 1365 0 53.0 365.0 15.0 28.0 74.0 24 75 75.0 517.0 87.0 000.0 2.5 11.120 194.0 1330 0 51A 352.0 15.0 30.0 74.0 24 75 73 0 503.0 88.0 607.0 2.5 11.120 177.0 1220 0 51.0 352.0 13.0 26.0 71 4 149 300 68 0 469.0 82.0 565 0 2.5 11.120 Weld Metal 1255.0 53.0 365_0 12.0 24.0 71.0 69.0 476.0 82.0 565.0 2.6 11.565 182.0 149 300 400.0 13.0 24 4 64 4 87.0 800.0 2.9 12.899 163.0 1124.0 56.0 288 550 66 0 455.0 12.0 23 4 68.0 2.8 12.454 179.0 1234.0 57.0 303 0 288 550 85.0 448.0 87A 000 0 6,}^ f f 4

TEMP 2RATURE (*C) 100 50 0 50 100 150 200 1"

l I I I I I I 3eo - - 220

- 200 I" ~

8

)

Oh"b - 1%

_ 120 -

160 O O E 2(Specimens) g 100 -

/

/ - 140 s 80 -

g - 120 2

g 2 O

O - 100 ym W

? - M 40 - _ 80

_ 40 20 -

_ 20 0 'O I I I I !

O 200 100 0 100 200 300 400 TEMPERATURE (*F)

FIGURE 31. PREIRRADIATION CHARPY V-NOTCH IMPACT ENERGY FOR THE CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE 886051 (LONGITUDINAL ORIENTATION)

TEMPERATURE (*C) 100 -50 0 50 100 150 200 120 ; ; ; ; ; ; ;

160 100 - '

A O - 1#

O F a f O

_ 120

- 80 -

- 100

%e e g

/g go a

s

& O

_ m 20 -

_ 20 0 I I I I I O

-200 -100 0 100 200 300 400 TEMPERATURE (*F)

FIGURE 3 2. PREIRRADIATION CHARPY V-NOTCH IMPACT ENERGY FOR

' THE CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE B86051 (TRANSVERSE ORIENTATION) 3-9

TEMPERATURE (*C) 100 -50 0 50 100 150 200

I I I I I I I _ m

180 -

_ 240 1M - O - m t

O O _ 200 (

1" ~

O O/g _ 180 y 120 pb

- 160 E, 1#

3on _

g -

9 g _ 12 80 --

joo i M -

po - #

40 -

6 _ 60 O 2

20 -

_ 20 i

o 4- 1 I I I I O 200 100 0 100 200 300 400 TEMPERATURE (*F)

FIGUR7 4. PREIRRADIATION CHARPY V-NOTCH IMPACT ENERGY FOR THE CATAW8A UNIT NO. 2 REACTOR PRESSURE VESSEL 4

CORE REGION WELD METAL I (

TEMPERATURE (*C) 150 100 50 0 50 100 150

'80 2*

I l I I I L l 1 --

i 160 - -

220 O

l 1" -

Oo O

O -

m l O O -

180 O

,- 120 -

Ob - 160

,oo _ 8 - ia l [ 9 -

120 $ ,

i O 80 -

E - 100

I @ - O _ M w O l m -

~ * -

p Oo - #

20 -

g _ 20 I I I I I 0 0 l 300 -200 100 0 100 200 300 TEMPERATURE (*F)

FIGURE 3 4. PREIRRADIATION CHARPY V-NOTCH IMPACT ENERGY FOR !

THE CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL {

WELD HEAT-AFFECTED-ZONE MATERIAL i l

, 3-10

TEMPERATURE (SC) 0 50 100 150 200 250 300 Ho l I I I I I l 100 -

700 2

1 2 (SPECIMENS)

- 90 -

Q\ -

600 l

]#

80 - l

$ ULTIMATE TENSILE STRENGTH E l ul g 500 3 70 -

E m 60 -

O N- 400 50 -

0.2% YlELD STRENGTH O 40 l l l l l - 300 0 100 200 300 400 500 600 1 TEMPERATURE (*F) l 2

TEMPERATURE (*C) !

i 0 50 100 150 200 - 250 300 80 l l l l l l l l ;

l' 70 -

\ @

REDUCTION IN AREA g l

] 2

, - so -

50 -

[

F !

2 40 -

)

F 8

o 30 -

9 No 2 -9 20 -

2 TOTAL ELONGATION 9

0 -

UNIFORM ELONGATION l I I I I I l 0

0 100 200 300 400 500 s 600 TEMPERATURE (*F)

FIGURE 3-5. PREIRRADIATION TENSILE PROPERTIES FOR THE CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE B8605-1 (LONGITUDINAL ORIENTATION) 3 11 l l

TEMPERATURE (oC) 0 50 100 150 2u0 250 300 l l l l l l l 700 100 -

f" -b N _ 600 m 80 - -

ft) 2 ULTIMATE TENSlLE STRENGTH #

[x 600

! E 70 -

2 $

]

60 - N-2 0.2% YlELD STRENGTH 400 50 -

40 l I I I I -

300 0 100 200 300 400 500 600 TEMPERATURE (*F) 4 TEMPERATURE (*C) 0 50 100 150 200 250 300

. 80 ; ; ; ; ,

i i i r i 70 -

8-

- 60 - 9 NO

$ REDUCTION IN AREA O

[H 50 -

1 2 40 -

$ 30

= -

g-o 9 l

20 - 2 TOTAL ELONGATION

_g 0 0 10 -

UNIFORM ELONGATION I I I I I 0

O 100 200 300 400 500 600 i

TEMPERATURE (*F)

FIGURE 3-6. PREIRRADIATION TENSILE PROPERTIES FOR THE <

3 CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE B8605-1 (TRANSVERSE ORIENTATION) l 3-12 i

TEMPERATURE ('C) 9 0 50 100 150 200 250 300 i 11 l l l l l l l 00 -

700

!)

p90 so g

2( k_ 300 M

i e

y 70 -

9N ULTIMATE TENSILE STRENGTH

-g 500 3

e 4-4 H 9 l # 60 -

0.2% YlELD STRENGTH 400 50 -

4o l I I I I - 300

O 100 200 300 400 500 600 l TEMPERATURE (*F) !

i TEMPERATURE (*C)

! O 50 100 150 200 250 300

r 80 ; ; ; ; ; ; ;

70 -

2 NO !

2 o

_ REDUCTION IN AREA l

' b 50 -

l

> l a

40 -

l f h 30 -

= -8 g 20 -

1 TOTAL ELONGATION 10'

- b UNIFORM ELONGATION o l l l l - l 0 100 200 300 400 500 600 i TEMPERATURE (*6 1

FIGURE 3-7. PREIRRADIATION TENSILE PROPERTIES FOR THE is CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL

CORE REGION WELD METAL 3-13 l l

l a

a hh.,d. A.A_ -.am- - . __ _ A -m-__aya wa-%. s+,hm.,AAL-+- .-m.ma_. .--_.m__ .J*sh4_.$a---__,.A.A__uw-_a_.- e._W.,Aem.A&_ a s Aa 4 4..maa , .

t s-4 1 m o

N 4

I i

J

~

"m G

' H

Im C

l c 8

d v

Z O E

E k1: 1 i

H 9m en m .

C w

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a b ,

e e

~

i p

d i

I 4 SS381S

. t i

i a

4 l

3 14 i

a SECTION 4 POSTIRRADIATION TESTING 4-1. CAPSULE REMOVAL The first capsule (Capsule U) should be removed at the end of the first core cycle (1st refueling) as shown in Table 4-1. Subsequent capsules should be removed at 6,9, and 15 EFPY (Effective Full Power Years) as indicated. Each specimen capsule, removed after exposure, will be transferree to a postirradiation test facility for disassembly and testing of all the specimens, 1

I '

TABLE 4-1

, SURVEILLANCE CAPSULE REMOVAL SCHEDULE j Orientation I

Capsule of Lead Removal Expected Capsule identification CapsulesI 'l Factor *l Time 2 Fluence (n/cm )

U 58.5* 4.00 1st Refueling 3.28 x 10 18 Y 241

3.69 6 EFPY 1.45 x 10'M l V 61

3.69 9 EFPY 2.18 x 10'Mdl X 238.5* 4.00 15 EFPY 3.94 x 10

W 121.5* 4.00 Stand-By Z 301.5* 4.00 Stand-By

a. Reference Irradiation Capsule Assembly Drawing, Figure 2-4.

b. The factor by which the capsule fluence leads the vessels maximum inner wall fluence.

y c. Approximate Fluence at %-wall thickness at End-of-Life,

d. Approximate Fluence at vessel inner wall at Endef-Life.

4-1

i i

i 4-2. CHARPY V-NOTCH IMPACT TESTS c

i The testing of the Charpy impact specimens from the intermediate shell plate B86051 g weld metal, and HAZ metal in each capsule can be done singly at approximately ten l r

! different temperatures. The extra specimens should be used to run duplicate tests at i temperatures of interest to develop the complete Charpy impact energy transition curve.

l- The initial Charpy specimen from the first capsule removed should ha tested at room i temperature. The test value of this temperature should be compared with preirradiation

! test data. The test temperature fur the remaining specimens should then be adjusted l higher or lower so as to develop a complete transition curve, For succeeding tests after

! longer irradiation periods, the test temperature in each case should be chosen in the light of results from the previous capsule.

) 4.3 TENSILE TESTS A tensile test specimen from each of the selected irradiated materials shall be tested l

at a temperature represonative of the upper end of the Charpy energy transition region.

l The remaining tensile specimens from each material shall be tested at the service j ternperature (550*F) and the midtransition temperature.

1 l

( -

4.4 FRACTURE TOUGHNESS TESTS ON 1/2 COMPACT SPECIMENS l

in light of current requirements of 10CFR, Part 50, Appendix G and applications of ASME f Socition lil, Appendix G and Section XI, Appendix A, the %4nch thick compact specimens i

should be tested in such a manner as to determine both static, crack initiation, and l

i propagation parameters throughout the temperature range of interest with emphasis on the sharp fracture toughness transition and upper shelf regions consistent with l

i specimen availability. The specimens should thus be statically tested in accordance

with ASTM E399-81 procedures modified to account for the size of the specimens i available.UI Specific test procedures should include unloading compilance and data j interpretation should utilize the Equivalent Energy and J-Integral concepts.12.3.41 i

1. Witt. F. J., " Fracture Toughnese Parameters Otiteined from Single SmeN Specimen Teste". WCAP 9307, october 1978.

! 2. suchenet, c. and unger. T. R. "superimonimi veriscamon of Lo.or sound x,vesues uisizing the wu : Energy concept."

In Progress in Flew Growth and Fracture Toughness Teseng. ASTM.STP-638, pp. 281296. Amortcen Society for Testing er:d l Metertels. Philadelph6e.1973. '

! 3. Landes. J. D. and Begiey, J. A., "Recent Developmente in J. Teodng". In Dewedopmenes in Fracture Mecheruce Test Methode j Stenderdlaedon. ASTM-STP432, pp. 57 81. Amortcen Soc 6ety for Teeung and Maternels. Philadelphia,1977.

4. McCabe. D s.. "Determenemon of R4urves for Structural Meter 6 ele Using Noneneer Mecheruce Methods," in Flow Growth and t

Fracture. ASTM-STP431, pp. 245 226, Amorteen Society for Teeung and Matertels. Ph6tedelphia,1977.

42 I $

1 h

Fracture toughness data so obtained will be Kw, Je and dJ/da or engineering estimates thereof. Advantages should be taken of the Charpy impact and tensile data in the selec-tion of initial test temperatures. Test procedures actually performed on the specimens will reflect state of the-art at the time of testing.

4.5 POSTIRRADIATION TEST EQUIPMENT Required minimum equipment for the postirradiation testing operations is as follows:

E Milling machine or special cutoff wheel for opening capsules, dosimeter

~

blocks and spacers.

E Hot cell tensile testing machine with pin-type adapter for testing tensile specimens.

5 Hot cell static CT testing machine with clevis and appropriate measuring equipment modified to account for the size of the specimens. I E Hot cell Charpy impact testing machine.

5 Sodium iodide scintillation detector and pulse height analyzer for gamma counting of the specific activities of the dosimeters.

e e

9 e

4 4-3 O

APPENDIX A i DESCRIPTION AND CHARACTERIZATION OF THE CATAWBA UNIT NO. 2 REACTOR VESSEL BELTLINE AND SURVEILLANCE MATERIALS Based on the initial RTNDT, chemical compostion (copper and phosphorus) and the ,

end-of life neutron fluence, the reactor vessel intermediate shell plate B8605-1 is i expected to have the highest-end-of-life A RTNDT using the prediction methods of Regulatory Guide 1.99 Revision 1 and latest ASTM revisions. This material is therefore considered to be the limiting vessel beltline region material and has been used in the reactor vessel surveillance program.

For the surveillance program Combustion Engineering, Inc., suppIled Westinghouse with sections of the A533 Grade B Class 1 Steel plate produced by Lukens Steel Company. This steel was used in the fabrication of the Catawba Unit No. 2 reactor pressure vessel, specifically, from the 9%-inch intermediate shell plate B8605-1. Also l supplied was a submerged arc weldment made from sections of intermediate shell i plate B8605 2kl and adjacent lower shell plate B8806-1. This test weldment was fabricated using inch Mil B-4 weld filler wire, heat number 83648 and Linde 0091 l*

flux, lot number 3536 and 5 !dentical to that used by Combustion Engineering, Inc.

in the Catawba Unit No. 2 reactor vessel fabrication process specifically the closing girth seam between the intermediate and lower shell plates, and all longitudinal weld seams of both the intermediate and lower shell plates.

The chemical analyses,TNDT, RTNDT, upper shelf energy and heat treatment history of all the core region pressure vessel shell plates used in the fabrication of the Catawba Unit No. 2 reactor pressure vessel are summarized in Tables A-1 thru A-5 respectively.

This data is as reported in the vessel fabricators (Combustion Engineering, Inc.)

certification reports or from subsequent Westinghouse analyses of similar materials used for the Catawba Unit No. 2 surveillance program. Weld material identical to that used in the fabrication of the core region beltline weldsIbl have been correlated with the Westinghouse surveillance program test weldment and available Combustion Engineering, bc. weld certification reports and their surveillance program test weldment.

This data is also reported in Tables A-3 thru A-5 of this Appendix.

I a. The limiting plate material selectec in 1978 was intermediate shell plate B8S05 2. This selection was based at the time on the highest initial RTc. Therefore weld test plate "D" 'umished to Westinghouse at that time was made up of plates 88605-2 and B8806-1.

b. The beltline welds are considered to include the intermediate and lower shell plate longitudinal seams and the closing intermediate to lower shell girth seam.

A-1

['

i TABLE A 1 CHEMICAL ANALYSIS OF THE INTERMEDIATE SHELL PLATES \ ;

l USED IN THE CORE REGION OF THE CATAWBA ,,

UNIT NO. 2 REACTOR PRESSURE VESSEL 1

Chemical Compositon (weight %)

Plate E

~

Plate @l Plate @l

. B8605-1@l B8605-1I 'l B8605-2 88616-1 0 .25 .22 .24 .24 Mn 1.40 1.37 1.35 1.39 P .011 .012 .009 .010 S .013 .013 .013 .021 SI .28 .29 .26 .27 i Ni .63 .59 .61 -

.59 Mo .57 .57 .57 .54 Cr .085 .05 % i l

Cu .09 .071 .07 .05 Al .042 .043 Co .007 .006 Pb .001 < .001 W < .01 < .01 Ti .004 < .01 Zr < .002 .001 V not o.i.ci.o .002 .003 not o.i.ct.o Sn .007 .003 As .008 .005 Cb <.002 < .01 N2 .008 < .01

,B < .001 < .001

a. Surveillance program test plate,

b. Chemical Analysis b'; Combustion Engineering, Inc,

c. Chemical Analysis by Westinghouse.

A-2

_ __s

I i

TABLE A 2

/

CHEMICAL ANALYSIS OF THE LOWER SHELL PLATES USED IN THE CORE REGION OF THE CATABWA UNIT NO. 2 REACTOR PRESSURE VESSEL

] Chemical CompositonI 'l Element

. Plate Plate Plate I B8806-1 B8806 2 88806-3

) C .23 .19 .20

! Mn 1.40 1.33 1.35 l P .009 .007 .006 i S .016 .013 .013 Si

.22 .23 .23 NI .56 .59 .59 Mo .57 .55 .55

,j Cr .12 .03 .03 4 Cu .05 .05 .05 l Al .021 .027 .027 i Co .006 .006 .005 Pb < .001 < .001 <.001 W < .01 < .01 <.01 Ti. < .01 < .01 <.01-

Zr .002 .001 .001 V .004 .003 .002 Sn .001 .002 .001 As .003 .004 .002 Cb '< .01 < .01 < .01 N2 .007 .006 .006 l B '< .001 <.001 <.001

a. Chemical Analysis by Combustion Engineering, Inc.

A-3

I TABLE A-3 {

l CHEMICAL ANALYSIS OF THE WELD METAL USED lN THE CORE REGION WELD SEAMS OF THE %

CATAWBA UNIT NO 2 REACTOR PRESSURE VHSSEL woTe n. car. r.gion m w. n. can.id-.o m inam m. ininm.sm. .nd io.w v s pieu iongitumn i m. .no m. pining ini-m.m.m a io w.a oirm m.

M cor region W w.w. ww. fe*W.d u.ing W.id Wlr. H.m No. 83H8. Und.

0001 Run. L.ot No. M38.

Chem p ton Element Wire Flux Actual Production Westinghouse Test Wold Wold (Lower Surveillance SampleI 'l Shell Longitudinal Program Test Seem 101-142A)l*3 Weidment D#l C .13 .14 .15 Mn 1.23 .88 1.20 P .005 .008 .010 S .009 .012 ,010 Si .13 .12 .15 Ni .14 .14 (

Mo .59 .44 .60 t

Cr .03 .052 Cu .04 .04 .036 Al .001 .002 Co .017 .009 Not detected Pb -

.001 W .01 < .01 Ti <.01 .005 Zr .002 < .002 V .006 .004 .004 Sn .008 .005 As .013 .002 Cb .017 < .002 N2 .007 .004 B < .001 < .001

a. Chemical Analysis of Wire-Flux Weld Sample, Test Number D32255 and Chemistry Data Sheet 101-142A by Combustion Engineering, Inc.

b. Chemical Analysis by Westinghouse of Test Sample Supplied by Combustion

~

Engineering, Inc. Representative of the Closing Girth Seam Weld, Weld Wire Heat No.83648,0091 Flux, Lot No. 3536.

A-4

l

\

1 TABLE A-4 )

TNDT,RTNDT AND UPPER SHELF ENERGY FOR THE CATAWBA UNIT NO,2 REACTOR PRESSURE VESSEL CORE REGION SHELL PLATES i

AND WELD METAL l

I*I I Average TNDT RTNDT upper shetfall's Material Energy i

1

('C) ('F) ('C) (* F) (J) (ft Ib)

_ Intermediate Shell Plates:

B8605-1 10 -9 15 121 89 B8605 2 20 1 33 111 82 B8616-1 - 18 0 - 11 12 125 92 Lower Shell Plates:

88806-1 - 51 - 60 - 14 6 113 83

, B8806 2 40 - 23 -10 138 102 B8806-3 40 -13 8 142 105

a. Data obtained from Combustion Engineering, Inc. Reactor Vessel Material Certification Reports.

b. Drop weight data obtained from the transverse material properties (normal to the major working direction),

c. From impact data obtained from the transverse material properties (normal to the major working direction).

NDT(q Upper ShelfM RTNDT Energy Material

('C) (*F) (*C) ('F) (J) (ft Ib) i Intermediate and Lower Shell Longitudinal Weld Seams and Closing Girth 80 - 62 - 80 176 130 Weld Seam (Weld Wire Heat No. 83648, Linde 0091 Flux, Lot No. 3536)

[ d. Data obtained from Combustion Engineering, Inc. Wire / Flux Weld Deposit Material Certification Test No.1332.

A5 l

TABLE A 5 HEAT TREATMENT HISTORY OF THE CATAWBA i UNIT NO. 2 REACTOR PRESSUR2 VESSEL r CORE REGION SHELL PLATES AND WELD SEAMS Temperature Timel*l Material ('F) (hr) Cooling Austenitizing: 4 Water quenched 1600 t 25 Intermediate (871 *C)

Shell Plates Tempered: 4 Air-cooled i B86051 1225 t 25 B8605-2 (663*C)

B86161 Stress Relief: 20lbl Furnace cooled 1150 t 50 (621'C)

Austenitizing: 4 Water-quenched 1600 i 25 Lower (871 *C)

Shell Plates Tempered: 4 Air cooled B8806-1 1225 i 25 B8806-2 (663*C) 88806-3 Stress Relief: 17(bl Furnaco-cooled 1150 t 50 3 (621 *C) l Intermediate Shell Longitudinal Stress Relief: 20tbl Furnace-cooled Seam Welds 1150 t 50 (621'C)

Lower Shell Longitudinal 17Ibl Furnace-cooled Seain Welds Local Intermediate to Stress Relief: 11 Furnace-cooled Lower Shell Girth 1150 i 50 Seam Weld (621'C) _

Surveillance Program Test Material Surveillance Program Weldment Test Post Weld (a%

P d, Stress Relief: 1111 Furnace-cooled cwng oirm s m) 1150 t 50 (

(621 *C)

i*.O*1'=' M TO"'O CW,.est .ee, - T-

6. The Strees Renet Heat freennent recorved try use Surweenance Test Wetament has tieen sunusased A-6

ML20210U640

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ML20210U640

Text

=a =

. e =. = =

m ar ,. m. m. m. an = of , == otn

SUMMARY

Navigation menu

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