ML17221A530

Proposed marked-up Tech Specs,Incorporating Revised Pressure/Temp Limits & Results of Recent Low Temp Overpressure Protection Analysis

ML17221A530
Person / Time
Site:	Saint Lucie
Issue date:	11/27/1987
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17221A528	List: L-87-482, Application for Amend to License NPF-16,incorporating Revised Pressure/Temp Limits & Results of Recent Low Temp Overpressure Protection Analysis.Nrc Approval Requested Prior to Plant Reaching 32 Efpys.Fee Paid ML17221A530
References
NUDOCS 8712030121
Download: ML17221A530 (170)

Text

ATTACNBlT 1 ST.

LIJCIE NIT 2 MARKED UP TECHNICAL SPECIFICATION PAGES 1-4 3/4 4-3 3/4 4-5 3/4 4-10 3/4 4-29 3/4 4-30 3/4 4-31}

Delete; replace with new pages:

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Replace with new page B 3/4 4-10 8 3/4 4-11 l

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1. 16 LOW TEMPERATURE RCS OVERPRESSURE P

ECTION RANGE The LOW TEMPERATURE RCS OVERPRESSURE PROTECTI RANGE is that operating condition when (1) the cold leg temperature s

MEMBER S OF THE PUBLIC 1.17 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant.

This category does not include employees of the licensee, its contractors or vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL ODCM 1.18 The OFFSITE DOSE CALCULATION MANUAL shall contain the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and shall include the Radiological Environmental Monitoring Sample point locations.

OPERABLE - OPERABILITY

1. 19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s),

and when all necessary attendant instrumentation, controls, electrical

power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MOOE -

MODE 1.20 An OPERATIONAL MODE (i.e.

MODE) shall correspond to any one inclusive combination of core reactfvity condition, power level and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and relapsed instrumentation and (1) described in Chapter 14.0 of the FSAR, (2) authorired under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

ST.

LUCIE - UNIT 2 1-4 Amendment No. 16

~

REACTOR COOLANT SYSTBl HOT SHUTQOWN LIMITING CONOITION FOR OPERATION

't 3.4.1.3 At least two of the loop(s)/train(s) listed belo~ shall be OPERABLE and at least one Reactor Coolant and/or shutdown cooling 'loops shall be in operation. "

a.

Reactor Coolant Loop 2A and its associated steam generator and at least one associated Reactor Coolant pump,""

b.

Reactor Coolant Loop 2& and its associated steam generator and at least one associated Reactor Coolant pump,~~

c.

Shutdown Cooling Train 2A, d.

Shutdown Cooling jrain 2B.

APPLICABILITY:

HOQE 4.

'\\

AC'lIOM:

a.

With less than the above required Reactor Coolant and/or shutdown cooling loops OPERABLE, immediately initiate corective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is a shutdown cooling loop, be in COLO SHUTDOWN within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With no Reactor Coolant or shutdown cooling loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant 1oop to operation.

Za+Iaa:

paso id/ 'lo~pS a~d Al Reactor oo ant pumps and shutdown cooling pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10oF below saturation temperature.

• "A Reactor. Coolant pump shall not be started wit one or more of the Reactor 5

ld1 I

h

.q unless the secondary water temperature of each steam generator is less than 4O F above each of the Reactor Coolant System cold leg temperatures.

pygmy". wt speci+

li ToLlc applaeaLlc 4ycmpihq

)cubi od ST.

LUCIE - UNIT 2 3/4 4-3 Amendment No. I6

REACTOR COOLANT SYSTEM COLO SHUTDOWN -

LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4. 1.4. 1 At least one shutdown cooling loop shall be OPERABLE and in operation",

and either:

a.

One additional shutdown cooling loop shall be OPERABLE

, or b.

The secondary side water level of at least two steam generators shall be greater than LOS indicated narrow range level.

APPLICABILITY:

MODE 5 with Reactor Coolant loops filled ACTION:

a.

, With one of the shutdown cooling loops inoperable and with less than the required steam generator level, immediately initiate corrective action to return the inoperable shutdown cooling loop to OPERABLE status or to restore the required steam generator'evel as soon as possible.

b.

With no shutdown cooling loop in operation, suspend all..operations involving a reduction in boro'n concent'ration of the Reactor Coolant System and immediately initiate, corrective action to return the required shutdown cooling loop to operation.

SURVEILLANCE RE UIREMENTS 4.4. 1.4. l. 1 The secondary side water level of at least two steam generators when required shall be determined.to be within limits at least once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> s.

4.4. 1.4. 1.2 At least one shutdown cooling loop shall be determined

.o be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The shutdown cooling pump may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided

1) no operations are permitted that would cause dilution of 'the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at least, 104F below saturation temperature.

One shutdown cooling loop may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other shutdown coolin loo is OPERABLE and in operation.

XssSKT':

%oao idle M HA Reacto~ Coolant pump shall not be star ted with+one or more of the Reactor unless the secondary water temperature of each steam generator is less than 49 F above each of the Reactor Coolant System cold leg temperatures.

ST.

LUCIE - UNIT 2 ZQssa'T:

3/4 4-5 Amendment gpeeo

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C ayplica'gg e~o Wine p+<< ~~

REACTOR COOLANT SYSTEM 3/4.4.4 PORV BLOCK VALVES LIMITING CONO'-TION FOR OPERATION 3.4.4 Each Power Operated Relief Valve (PORV) Block valve shall be OPERABLE.

No more than one block valve shall be open at any one time.

APPLICABILITY:

HOOES 1, 2, and 3.

Q~J ACTION:

a.

With one or more block valve(s) inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the block valve(s) to OPERABLE status or close the block valve(s) and remove power from the block valve(s); otherwise, be in at least HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLO SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With both block valves open, close one block valve within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, otherwise be in at least HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in.

COLO SHUTOOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c.

The provisions of specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREHENTS 4.4.4 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of.full ravel unless the block valve is closed wi h power removed in order to tpeet the requirements of Action a. or b.

above.

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ST.

LUCIE - UNIT 2 3/4 4-10

REACTOR COOLANT SYSTEM

~34. 4. 9 P RESSUREQTEHPERATURE LINITS REACTOR COOLANT SYSTEI4t LIMITING CONOITION FOR OPERATION 3.4.9. 1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures "durin

heatup, cooldown, criticality, and inservice leak and hydrostatic tes cng.

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2e Waugh

>.0 'la APPL ICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering

'valuation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than 200'F and 500 psia, respectively, within tQg following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREHENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup,

cooldown, and inservice leak and hydrostatic testing operations.

ST., LUCIE - UNIT 2 3/4 4-29 Amendment No.

I6

REACTOR COOLANT SYSTEM SURYEILLAHCE RE UIREHENTS Continued 4.4.9.l.Z The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50 Appendix H in accordance with the schedule in Table 4.4-5.

The results of these examinations shall be used to update Fig IIJ5%gt:-$.% -%ca Wve~4 LQ'-7e

(

ST.

LUCIE - UNIT 2 3/4 4-30 Amendment No. 16

FIGURE 3.4.2 ST. LUt'E 2 P/T LIMITS, 4 EFPY HEATUP ANO CORE CRITICAL QSLQT~ y Q5 Q 7-l l'Qv 6 g + ~

7Hgou 4~

5.Q -7c lllli I

I ~li I

S04 F/HR 2000 1500 N

i 1000 LOWEST SERVICE TEMP. 1684F O'z O

Soa I s04F/ HR.

35 PSIA

. CORE RITICAL t.

MAXlMUM LLOWABLEHU RATES HU oF/HR TEMP. LIMIT.4F I

MIN.BOLTUP TEMP.,

I I

I 40 50 IIII IlllfllllIIII IIII IIIII

'367 100 200 300 400 60 TC ~ INOICATEO REACTOR COOLANT TEMPERATURE, 4F ST.

CIE - UNIT 2 3/4 4-31 Anendment No. I6

4

FIGURE 3.4-2a ST. LUCIE-2 P/T LIMITS, 10 EFPY HEATUP AND CORE CRITICAL 2500

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-'::60 F/HR

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2000 1500 N

1000 4z

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LOWEST SERVICE TEMP 168oF I

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ITICAL ' -"

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BOLTUP TEMP..:"

ALLOWABLEHEATUP RATES RATE oF/HR TEMP I INIIT oF 50 AT ALLTEMPERATURES

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I 0 0 100 300 400 600 TC INDICATEDREACTOR COOLANTTEMPERATURE oF 3/4 4-31a

FIGURE 3.4-2b ST. LUCIE-2 P/T LIMITS, 10 EFPY COOLDOWN AND INSERVICE TEST 2500 INSERVICE TEST 2000 r

CO CC 1500 N

D a

1000 O

ClR O

500 LOWEST SERVICE TEMP. 168 F

THERMAL 660 PSIA 30oFlHR 50 100 100'F/H R 5. ISOTHERMAL 0

0 100 MIN BOLTUP TEMP.

200 300 400 500 TC INDICATEDREACTOR COOLANTTEMPERATURE oi 3/4 4-31b

FIG UR E 3.4-2c ST. LUCIE-2 P/T LIMITS,10 EFPY MAXIMUMALLOWABLECOOLDOWN RATES 100 80 X

60 1

IXz O

40 Cl 00 II 20 RATE oF/HR 30 50 75 100 TEMP LIMIT oF 4104 104-130 130-146

>146 80 100 120 160 180 200 Tc INDICATEDREACTOR COOLANTTEMPERATURE oF NOTE: A MAXIMUMCOOLDOWN RATE OF 100oF/HR IS ALLOWEDAT ANY TEMPERATURE ABOVE 146oF 3/4 4-31c

psg4gp S 5.Q -24

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FIOU813.4.3 IT, LIJCII2 f/T LIMITS, 4 IlLPY COOLOOWN AhlD INIIRVICITRtf I l'Il t e

2000

. LONltFIBlVICt TKQt. 1(4 i 4 IIOT

ISO I MAXNUQALLONAILICO RAT%5 Var. LIIIn;eI 400 50 15 100

~85'4 lit.

l)QM

~128 a

0 100 2N

%0; 400 Tg - NIQICATIDRIACTOltCOOLANTTIQP%llATVRI, t'T.'.UCLE UNlT 2 3)4 4-32 Anendeent No. ~~

FIGURE 3.4-3a ST. LUCIE-2 P/T LIMITS,15 EFPY HEATUP AND CORE CRITICAL 2500 50 F/HR 2000 1500 LLI N

D IX 0

1000 I

u4R O

500 50oF /HR 40'/HR LOWEST SERVICE TEMP 168oF 535 PSIA CORE CR ITICAL MIN. BOLTUP TEMP.

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ALLOWABLEHEATUP RATES RATE oF/HR TEMP LIMIT oF 40 489 50

)89 0"

100 200 300 400 500 TC INDICATEDREACTOR COOLANT TEMPERATURE oF 3/4 4-32a

FIGURE 3.4-3b ST LUCIE 2 P/T LIMITS,15 EFPY COOLDOWN AND INSERVICE TEST 2500 INSERVICE TEST 2000 4

V)i 1500 CC N

D 4

1000 I

u4P u

500 LOWEST SERVICE TEMP 168oF ISOTHERMAL 30oF/HR 50 100 100oF/H THERM Rh ISO AL:

MIN. BOLTUP TEMP.

100 300 400 500 TC INDICATEDREACTOR COOLANTTEMPERATURE' 3/4 4-32b

FIGURE 3.4-3c ST. LUCIE-2 P/T LIMITS, 15 EFPY MAXIMUMALLOWABLECOOLDOWN RATES 100 80 cx 0

60 cz 40 OO 20 RATE 'F/HR 30 50 75 100 TEMP LIMIT oF 4 115 115-141 141-157

) 157 80 100 120 160 180 200 Tc INDICATEDREACTOR COOLANTTEMPERATURE o NOTE: A MAXIMUMCOOLOOWN RATE OF 100oF/HR IS ALLOWEDAT ANY TEMPERATURE ABOVE 157oF 3/4 4-32c

FIGURE 3.4-4a ST. LUCIE.2 P/T LIMITS,20 EFPY HEAT UP AND CORE CRITICAL 2500 50oF/HR 2000 CD 1500 N

K 1000 I

O0z O

500 LOWEST SERVICE TEMP 168oF 50oF/HR 40oF/HR

..535 PSIA CORE CRITICAL ALLOWABLEHEATUP RATES MIN. BOLTUPTEMP.

RATE oF/HR 40

-50 TEMP LIMIT oF 496

) 96 100 200 300 400 500 TC INDICATEDREACTOR COOLANTTEMPERATURE 3/4 4-32d

FIGURE 3.44b ST. LUCIE-2 P/T LIMITS, 20 EFPY COOLDOWN AND INSERVICE TEST 2500 INSERVICE TEST 2000 CCD a.

1500 N

D O

1000 I

u4z

~

CJ 500 LOWEST SERVICE TEMP 168oF ISOTHERMAL 30 F/HR 50 100 100 F/HR & ISOTHERMAL 100 MIN.8OLTUP TEMP.

200 300 400 500 TC INDICATEDREACTOR COOLANT TEMPERATURE, F

3/4 4-320

FIGURE 3.4-4c ST. LUCIE.2 P/T LIMITS,20 EFPY MAXIMUMALLOWABLECOOLDOWN RATES 100 80 0

60 z

OO 40 20 RATE F/HR 30 50 75 100 TEMP LIMIT oF 4 121 121-147 147-163

) 163 80 100 120 140 160 180 200 Tc - INDICATEDREACTOR COOLANTTEMPERATURE oF NOTE: A MAXIMUMCOOLDOWN RATE OF 100oF/HR IS ALLOWEDAT ANY TEMPERATURE ABOVE 163oF 3/4 4-32f

F IGURE 3.4.5a ST. LUCIE-2 P/T LIMITS,26 EFPY HEAT UP AND CORE CRITICAL 2500 50oF/HR 2000 D

1500 N

CC Q

I-1000 4z O

lX 0

500 50oF/ HR 40'/HR'OWEST SERVICE TEMP 168oF

.535 PSIA CORE CR ITICAL ALLOWABLEHEATUP RATES RATE oF/HR TEMP LIMIT oF 40 4 102 50

) 102 100 MIN. BOLTUP TEMP.

200 300 400 500 TC INDICATEDREACTOR COOLANTTEMPERATURE oF 3/4 4-32g

FIGURE 3.4-6b ST. LUCIE.2 P/T LIMITS,25 EFPY COOLDOWN AND INSERVICE TEST 2500 INSERVICE TEST 2000 C4 LLJ 1600 ill N

Ch Lll Q

1000 uaz o

LOWEST SERVICE TEMP 168oF ISOTHERMAL 30oF/HR 50 100

-100 F/HR & ISOTHERMAL

'00 00 100 MIN. BOLTUP TEMP.

200 300 400 500 TC INDICATEDREACTOR COOLANTTEMPERATURE F

3/4 4-32h

FIGURE 3.4-5c ST. LUCIE-2 P/T LIMITS,25 EFPY MAXIMUMALLOWABLECOOLDOWN RATES 100 80 lCx LL 0

60 z

O O0 20 RATE oF/HR 20 30 50 75 100 TEMP LIMIT oF 483 83-126 126-152 152-168

> 168 80 100 120 160 180 200 Tc INDICATEDREACTOR COOLANTTEMPERATURE oF NOTE: A MAXIMUMCOOLDOWN RATE OF 100oF/HR IS ALLOWEDAT ANY TEMPERATURE ABOVE 168oF 3/4 4-32i

FIGURE 3.4%a ST. LUCIE-2 P/T LIMITS,30 EFPY HEATUP AND CORE CRITICAL 2600 50oF/HR 2000 CC 1600 N

CI 1000 I

O 0z 500 LOWEST SERVICE TEMP. 168oF 60 F/HR 0oF/HR

.-.- 536 PSIA CORE CRITICAL ALLOWABLEHEATUP RATES RATE oF/HR TFMP LIMIT oF 40 4 106 50

) 106 MIN. BOLTUP TEMP, 100 200 300 400 500 TC INDICATEDREACTOR COOLANTTEMPERATURE F

3/4 4-32j

FIGURE 3.4kb ST LUCIE-2 P/T LIMITS,30 EFP Y COOLDOWN AND INSERVICE TEST 2600 INSERVICE TEST..

2000 CC 0

1500 N

CC tL 0

Cl 1000 O0Z LOWEST SERVICE TEMP. 158oF ISOTHERMAL 30 F/HR 100oF/ HR & ISOTHERMAL:

50 600 0

0 MIN. BOLTUP TEMP.

100 200 300 400 TC INDICATEDREACTOR COOLANTTEMPERATURE oF 500 3/4 4-32k

F IGURE 3.4-6c ST. LUCIE-2 P/T LIMITS,30 EF,PY MAXIMUMALLOWABLECOOLDOWN RATES 100 80 x

60 I

Kz o

40 Cl O0 20 RATE F/HR 20 30 50 75 100 TEMP LIMIT oF 87-130 130-156 156-172

) 172 80 100 120 160 180 200 Tc - INDICATEDREACTOR COOLANTTEMPERATURE oF NOTE: A MAXIMUMCOOLDOWN RATE OF 100oF/HR IS ALLOWEDAT ANY TEMPERATURE ABOVE 172oF 3/4 4-32I

FIGURE 3.4-7a ST. LUCIE-2 P/T LIMITS,32 EFPY HEATUP AND CORE CRITICAL 2500

~

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50oF/HR 2000 Q

1500 N

Q 1000 Oz O

0 500 50'F/ HR 40oF/HR LOWEST SERVICE TEMP 168oF 535 PSIA CORE CR ITICAL ALIOWABLE HEATUP RATES RATE oF/HR TEMP LIMIT oF 40 4 108 50

) 108 MIN. BOLTUP TEMP.

00 100 200 300 400 500 TC INDICATEDREACTOR COOLANTTEMPERATURE oF 3/4 4-32m

FIGURE 3.4.7b ST. LUCIE-2 P/T LIMITS,32 EFPY

, COOLDOWN AND INSERVICE TEST 2500 INSERVICE TEST 2000 D

15OO CC N

aw 1000 O4R V

4 5oo LOWEST SERVICE TEMP. 168oF ISOTHERMAL 30oF/HR 50 100 100oF/ HR &ISOT HE AL 100 MIN. BOLTUP TEMP.

200 300 400 500 TC INDICATEDREACTOR COOLANTTEMPERATURE oF 3/4 4-32ll

FIGURE 3.4-7c ST. LUCIE-2 P/T LIMITS,32 EFPY MAXIMUMALLOWABLECOOLDOWN RATES 100 80 x

0 60 CCz O0 40 00V 20 RATE oF/HR 20 30 50 75 100 TEMP. LIMIT, F 489 89-132 132-158 158-174

) 174 80 100 120 160 180 200 Tc INDICATEDREACTOR COOLANTTEMPERATURE oF NOTE: -'A MAXIMUMCOOLDOWN RATE OF 100oF/HR IS ALLOWEDAT ANY TEMPERATURE ABOVE 174oF 3/4 4-32o

TABLE 4. 4-5 REACTOR VESSEL HATERIAL SURVEILLANCE PROGRAH - WITIIORAMAL SCHEDULE CAPSULE NUHER 2

VESSEL LOCATION 83 974 l04 2634 2224 2844 LEAD FACTOR

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MITNDRA'MAL T IHE EFPY 1.0

24. 0 STANDBY

12. 0 STANDBY-STANOBY Zvss QY:

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REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONOITION FOR OPERATION 3,4,9.3 At least one of the following overpressure protection sy tems shall be OPERABLE:

a.

Two ower-operated relief valves (PORVs),

each wi a lift setting of 1

s than or equal to 470 psia, or b.

Two sh tdown cooling relief valves (SDCRVs) w' a lift setting of less th or equal to 350 psia, or c.

The React r Coolant System depressurized ith an RCS vent of greater th or equal to 3.58 square i es.

APPLICABILITY:

a.

PORVs:

p epLAC,6

@,~ zusmT W)

) ~~g) b.

SQCRVs:

E.

In MO S 4, 5 and 6, duri g cooldown when the temperature of any CS cold leg is eater than er equal te 161'F and less tha or..equal to 6'F, and during heatup when the temperatu e of any R

cold leg is greater than or egal to 142'F a

d less an or equal to 295'F.

In HOOES 4, a

6, during cooldown when the temperature of any RCS co leg is less than or equal to 161'F, and during heatup when the temperature of any RCS cold leg is less tha o

equal to 142'F.

c.

RCS Vent:

In MO S

5 and

, dur ing cooldown when the temperature of any S cold le is less than or equal to 286'F, and durin eatup when e temperature of any RCS cold leg is 1

s than or equal to 295'F.

ACTI0[i:

a.

b.

C.

With a ORV being used for LTOP i

perable, restore the inoperable PORV OPERABLE status within 7 da s or depressurize and vent the RCS rough a greater than or equal 3.58 square inch vent(s) wi fain the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

ith both PORVs being used for LTOP ino rable, depressurize and vent the RCS through a greater than or e al to 3.58 square inch vent(s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

With a SOCRV being used for LTOP inoperable, estore the inoperable SOCRV to OPERABLE status within 7 days or depr surize and vent the RCS through a greater than or equal to 3.58 squ e inch vent(s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

I'

~

ST.

LUCIE - UNIT 2 3/4 4-35 Amendment No.

INSERT 4 1 3.4.9.3 Unless the RCS is depressurized and vented by a least 3.58 square inches, at least one of the following overpressure protection systems shall be OPERABLE:

a) Two power-operated relief valves (PORVs) with a liftsetting of less than or equal to 470 psia and with their associated block valves open.

These valves may only be used to satisfy low temperature overpressure protection (LTOP) when the RCS cold leg temperature is greater thon the temperature listed in Table 3.4-4.

b)

Two shutdown cooling relief valves (SDCRVs) with a liftsetting of less than or equal to 350 psia.

c)

One PORV with a liftsetting of less than or equal to 470 psia and with its associated block valve open in conjunction with the use of one SDCRV with a lift setting of less than or equal to 350 psia.

This combination may only be used to satisfy LTOP when the RCS cold leg temperature is greater than the temperature listed in Table 3.4-4.

APPLICABILITY:

MODES 31k, 48, 5 and 6.

ACTION:

a)

With only one of the required overpressure protection systems OPERABLE, restore at least two overpressure protection devices to OPERABLE status within 7 days or:

I) Depressurize and vent the RCS with a minimum vent area of 3.58 square inches within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; OR 2)

Be at a

temperature above the LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE of Table 3.4-3 within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b)

With none of the required overpressure protection devices

OPERABLE, within the next eight hours either:

I)

Be in a least COLD SHUTDOWN: restore at least on overpressure protection device to OPERABLE status within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or vent the RCS within the following 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; OR 2)

Be at a

temperature above the LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE of Table 3.4-3.

With cold leg temperature within the LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE of Table 3.4-3.

EJW I /033/4

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING COHOITION FOR OPERATION ACTION:

Continued:

h SDCRVs being used for vent the C

nin the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

e, depressuriz and equal to 3.58 square inch c~+

In the event either the

PORVs, SDCRVs or the RCS vent(s) are used to mitigate a

RCS pressure transient a Special Peport shall be prepared and submitted to the Coneission pursuant to Specification 6.9.2 within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the

PORVs, SOCRVs or vent(s) on the transient and any corrective action necessary to prevent recurrence.

\\

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a.

In addition to the requirements of Specification 4.0.5, operating t e through one complete cycle of full travel at least once per 18 months.

poav ST.

LUCIE - UNIT 2 3/4 4-36 Amendment No. 16

TABLE 3.4-3 LOW TEHPERATURE RCS OVERPRESSURE PROTECTION RANGE Operating

Period, EFPY During

~Heatu During Cooldown Cold Le Tem erature F

4?

10 10 -

15 15 - 20 20 - 25 25 - 30 30 - 32

< 313

< 324

< 330

< 335

< 339

< 341

< 304

< 315

< 321

< 326

< 330

< 332 TABLE 3.4-4 HINIHUH COLD LEG TEHPERATURE FOR PORV USE FOR LTOP Operating

Period, EFPY 4 -

10 10 ? 15 i5 ? 20 20 - 25 25 - 30 30 - 32 cold i 1'uring

~Heatu 156 165 172 178 182 184 Tcol d.)

During Cooldown 179 190 196 201 205 207

'0/+

4 - +7aa.

0 I

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.

1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.2O during all normal operations and anticipated transients.

In HOGES 1

and 2

with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.

In MODE 4, and in MODE 5 with. react:or coolant loops filled, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either shutdown cooling or RCS) be OPERABLE.

In %DE 5 with reactor coolant loops not filled, a single shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considergtions, and the unavailability of the steam generators as a heat removing component..requi re that at least two shutdown cooling loops be OPERABLE.

The operation of one reactor coolant pump or one shutdown cooling pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System; The reactivity change rate associated with boron reductions will, therefore, be within the 'capability of operator recogni tion and control.

The restrictionX on starting a reactor coolant pump in MODES 4 and 5, i

h RCS 1d 1

g<<

I h

provided to prevent RCS pressure tran-

sients, caused by energy additions from the secondary system from exceeding the limits of Appendix G to 10 CFR Part 50.

The RCS will be protected against overpressure transients by (1) sizing each PORV to mitigate the pressure tran-sient of an inadvertent safety injection actuation in a water-solid RCS with pressurizer heaters energized, (2) restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 40'F above each of the RCS cold leg temperatures, (3) using SDCRVs to mitigate RCP start transients and the transients caused by inadvertent SIAS actuation and charging water, and (4) rendering one HPSI pump inoperable when the RCS is at low t:emperatures.

xslsscT:

%a Spgc'>>

a e

3/4.4.2 SAFETY VALVES opple cap'(~

o pctehihq

/%joan The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia.

Each safety valve is designed to relieve 212,182 lbs per hour of saturated steam at the valve setpoint.

The relief capacity of a single safety valve is adequate to relieve any overpres-sure condition which could occur during shutdown.

In the event that no safety valves are

OPERABLE, an operating shutdown cooling loop, connected to the RCS.

provides overpressurerelief capability and will prevent RCS overpressurization.

In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

ST.

LUCIE - UNIT 2 B 3/4 4-1 lg)c, 4cl 5 Amendment, No. 16

REACTOR COOLANT SYSTEM BASES 3/4.4.4 PORV BLOCK VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.

Operation of the PORVs in conjunc-tion with a reactor trip on a Pressurizer Pressure-High signal minimizes the undesirable opening o

the spring-loaded pressurizer code safety valves.

The opening of the PORVs fulfills no safety-related function and no credit is taken for their operation in the safety analysis for MOOE 1, 2, or 3.

Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inooerable.

Since it is impractical and undesirable to actually open the PQRVs to demonstate their reclosing, it becomes necessary to verify OPERABILITY of the PORV block valves to ensure the capability to isolate a malfunctioning PORV.

As the PORVs are pilot operated and require some system pressure to operate, it is impractical to test them with the block valve closed.

The PORVs are sized to prdvide low temperature overpressure protection 3/4.4.5 STEAM GENERATORS RstaAce tuim zgsse7 y c 0~ <~Waco V'ACg The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of he RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision l.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to

design, manufacturing errors, or inser vice conditions that lead to corrosion.

ST.

LUCIE - UNIT 2 B 3/4 4-3

INSERT 42 (LTOP).

Since both PORVs must be OPERABLE when used for LTOP, both block valves will be open during operation within the LTOP range.

As the PORV capacity required to perform the LTOP function is excessive for operation in MODE 1, 2, or 3, it is necessary that the operation of more than one PORV be precluded du~ing these NOOES.

Thus, one block valve must be shut during MODES 1, 2

and 3.

The applicability of this technical specification. to only a part of MODE 3 is due to the LTOP range slightly overlapping NODE 3 in the operating periods beyond 15 EFPY.

(Refer to Table 3.4-3).

Both block valves will be open during operation in these lower temperature portions of NODE 3.

Pe REACTOR COOLANT SYSTEM BASES 3/4.4. 9 PRESSURE/TEMPERATURE LIMITS Qtli.>g

%o y~ip ALA Dmv N3 All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure)changes.

These cyclic 'loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.

The various categories of load cycles used for design purposes are provided in Section 5.2 of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

ing heatup, the thermal gradients in the reactor vessel wall prod thermal s

sses which vary from compressive at the inner wall to ten e at the outer wal These thermal induced compressive stresses tend lleviate he tensile stres induced by the internal pressure.

Theref

, a pressure-temperature curve bas n steaQ state conditions (i.e.,

thermal stresses) represents a lower bound ll similar curves for fi heatup rates when he inner wall of the vessil i treated as the gov ing location.

The heatup analysis also covers rmination of pressure-temperature limitations for the case in which the all'of the vessel becomes the ontrolling location.

The thermal adients e

lished during heatup produce ensile stresses at the outer of the vessel.

ese stresses are additive o the pressure induced te e stresses which are al present.

The hermal induced stress at the outer wall of the vessel a

tensile and are ependent on both rate of heatup and the time along the he ramp;

herefore, a

1 r bound curve similar to that described for the tup of the inner wall nnot be defined.

Consequently, for the cases in which t uter all o e vessel heromes the stress controlling location each heatu r

f crest aust be analyzed'n an individual'asis.

s hc 4 3. -l 0 omah)

The heatup and caoldown limit curves Figures

.4-2<

are compos e

curves which were prepared by determining the most conservative case, with the reactor vessel beltline or flange juncture limiting for heatup, and vessel belt-line limiting for cooldown, for the specified heatup and cooldown rates.

The reactor vessel materials have been tested to determine their initial RTNDT'he resul ts of these tests are shown in Table B 3/4.4-1.

Reactor opera-tion and resultant fast neutron (E greater than 1 NeV) irradiation will cause an increase in the RTNDT.

Therefore, an adjusted reference temperature, based upon the fluence and copper and nickel content of the material in question, can be predicted usin Figure B 3/4.4-1 and the recommendations of Regulatory Guide 1.99 Revision 2

Draft "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Naterials, The heatup and cooldown limit curves Figures 3.4-2 include predicted adjustments for this shift in RTNDT at the end of the applicable service period, as well as adjustments for possibTe errors in the pressure and temperature sensing instruments.

c,<'d~~qq~g,g~g~),q~

Eoc iwennq4ph th(eb'Act'loqj for A6jos4d @terence Teepnck(.e. 4k la)in'> dab'p4m+

~987, ST.

LUCIE - UNIT 2 8 3/4 4-8 Amendment No.16

INSERT /f3 Ouring heatup, the thermal gradients through the reactor vessel wall produce thermal stresses which are compressive at the reactor vessel inside surface and which are tensile at the reactor vessel outside surface.

Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the outside surface location.

However, since neutron irradiation damage is larger at the inside surface location when compared to the outside surface, the inside surface flaw may be more limiting.

Consequently for the heatup analysis both the inside and outside surface flaw locations must be analyzed for the specific pressure and thermal loadings to determine which is more limiting.

Ouring cooldown, the thermal gradients through the reactor vessel wall produce thermal stresses which are,tensile at the reactor vessel inside surface.and

'hich are corn'pressive at the reactor vessel outside surface.

Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the inside surface location.

Since the neutron indication damage is also greatest at the inside surface location the inside surface flaw is the limiting location.

Consequently, only the inside surface flaw must be evaluated for the cooldown analysis.

TABLE 8 3/4.4-1 REACTOR VESSEL TOUGHNESS Piece Ho.

122-102A 122-1020 122"102C 124-1028 124-102C 124-102A 142-102C 142-1028 142-102A 102-101 106-101

.128-101A 128-10lD 128-10l0 128-10IC 128-3018 128-301A 126-101 131-102A 131-102D 131-1028 131-102C 131-1010 131-IOlA 152-101 154-102 (A to F) 104-102 (A to D)

(I) Repo Code No.

H-604-1 H-604-2 H-604"3 H-605"1 H-605-2 H-605"3 H"4116-1 H-4116"2 H-4116-3 H-4110"1 H-4101-1 H-4102-1 H-4102-2 H-4lU2-3 H-4102-4 H-4103-1 H-4103-2 H-602-1 H-4104" 1 H-4104-2 H-4104" 3 H-4104-4 H-4105-1 H-4105-2 H-4112-1 H-4111-1 Haterial SA 5338 Cl 1 SA 5338 Cl 1 SR 5338 Cl 1 SA 5338 Cl 1 SA 5338 Cl 1 SA 5338 Cl 1 SA 5338 Cl 1 SA 5338 Cl 1 SA 5338 Cl 1 SA 5338 Cl 1 SA 508 Cl 2 SA 508 Cl 2 SA 508 Cl 2 SA 508 Cl 2 SA 508 Cl 2 SA 508 Cl 2 SA 508 Cl 2 SA 508 Cl 2 SA 508 Cl 1 SA 508 Cl 1 SA 508 Cl 1 SA 500 Cl 1 SA 500 Cl 1 SA 508 Cl 1 SA 5330 Cl 1 SA 5330 Cl 1 Vessel Locatio>>

Upper Shell Plate Upper Shell Plate Upper Shell Plate Intermediate Shell Plate Intermediate Shell Plate Intermediate Shell Plate Lower Shell Plate Lower Shell Plate Lower Shell Plate Closure Ilead Closure Head Flange l>>let Nozzle I>>let Nuzzle I>>leL Nozz le Inlet Nozzle Outlet Nozzle Outlet Nozzle Vessel Flange inlet Nozzle Safe End I>>let Nozzle Safe End Inlet Nozzle Safe End l>>let Nozzle Safe End Outlet Nozzle Safe End 0>>Llet Nozzle Safe End Bottom llead Dome Bottom llead Torus rted o>>ly for bel tl ine region plates.

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Direction Charpy

. Ft-lb 105 113

.-113 91 105 100

L= m x

CCD 150 IzO 100 ccl Ots 10ts NEUTRON FLUENCE. n/cm~ iE O 1 MeY)

Figure B 3/4.4-1 Nil.ductilitytransition temperature increase as a functio'n ot test iE)1 MeV) neutron fluence iS50'F irradiation)

ST.

LUCIE - U,'lkT 2'

~/'4 a-SO

150 14O 130 120 1 1 0 1 00 o

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n/cst (E

1 MeV)

ST.

LUCIE UNIT 2 Figttre B 3/4,41 Reference Temperature (RT>>T) Increases as a function of fast N

(E 1 MeV) neutron fluence (550 F irradiation) for Reactor Vessel Seltline '4laterials

REACTOR COOLANT SYSTEM BASES The actual shift in RTNOT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with AS'185-73 and 10 CFR Appendix H, reactor vessel m4ter ial irradiation surveil-lance specimens insta11ed near the inside wall of the reactor vessel in the core area.

The surveillance specimen withdrawal schedule is shown in Table 4. 4-5.

Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be app1ied with confidence to the adjacent section of the reactor vessel.

The heatup and cooldown curves must be recalculated when the delta RTNOT determined from the surveillance capsule is different from the for the equivalent capsule radiation exposure.

calculagtd delM RTNOT The pressure-temperature limit lines shown on Figures 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.

The maximum RTNOT Wor all Reactor Coolant System pressure-retaining materials, with the exception of the reactor pressure

vessel, has been'etermined

.to be 60'F.

The Lowest Service Temperature limit line shown on lgures

~

2 is based upon this RTNO7 since Article NB-2332 (Summer Addenda of 1972) of Section III of the ASME Boiler and Pressure Vessel Code requires the Lowest Service Temperature to be RTNDT + 100'F for

piping, pumps, and valves, Below this temperature, the system pressure must be limited to a maximum of 20".of the system's hydrostatic test pressure of 3125 psia.

The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pres-surizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the AS'od rem nts.

Z4Oa~:

m Cce c

The OPERABILITY of two PORVs, two OC s

o CS vent opening of greater I

than 3.58 square inches ensures th t the RCS will be protected from pressure transients which could exceed th limits of Appendix G to 10 CFR Part 50 when

<<h IICE 1d 1 gll 1

h Tke Low Temperature Overpressure Protection System has adequate relieving capability to protect the RCS from overpressur-ization when the transient is limited to either (1) a safety injection actuation in a water-solid RCS with the pressurizer heaters energized or (2) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 40'F above the RCS cold leg temperatures with the pressurize solid.

'ilh+Q7: Ltl&f

~e eppgcabl&

Mo4 muw 44lnvpA e 4urc4

~

ST.

LUCIE - UNIT 2 8 3/4 4-11 Amendment No.

16

The lead factors shown in Table 4.4-5 are the ratio of neutron flux at the urveillance capsule to that at the reactor inside surface.

ATTACHMENT2 SAFETY EVALUATION EJW I /033/3

SAFETY EVALUATION New pressure/temperature (P/T) limit curves have been generated for six operating periods starting at 4 Effective Full Power Years (EFPY) out to 32 EFPY

( in intervals of 4-10, 10-15, 15-20, 20-25, 25-30 and 30-32 EFPY).

A new Low Temperature Overpressure Protection (LTOP) analysis has also been performed to ensure that the reactor coolant pressure boundary (RCPB) integrity will be maintained in the low temperature mode of operation in accordance with the new P/T limits during each of the operating periods.

The P/T limits proposed by this Technical Specification change ensure that all components within the reactor coolant pressure boundary will be able to withstand the effects of loads due to system temperature and pressure changes without their.functions or performance being impaired.

These loads aie introduced by normal load transients, reactor trips, and startup and shutdown operations.

The proposed changes are as follows a 0 Limzting Condition for Operation (LCO) 3. 4. 9. 1 currently provides the pressure and temperature limits in Figures

3. 4-2 and 3. 4-3 for the RCS (except the pressurizer) during heatup and criticality, and during cooldown and inservice test, respectively, up to 4 EFPY.

The proposed amendment replaces Figures 3.4-2 and 3.4-3 with Figures 3.4-2a through 3.4-7c. Figures 3.4-2a, 3.4-3a, 3.4-4a, 3.4-5a, 3.4-6a and 3.4-7a depict the P/T limits for heatup and criticality for the operting periods ending at 10 EFPY, 15 EFPY, 20

EFPY, 25 EFPY, 30 EFPY and 32 EFPY, respectively.

Figures 3.4-2b, 3.4-3b, 3.4-4b, 3.4-5b, 3.4-6b and 3.4-7b depict the P/T limits for for cooldown and inservice test for the same operating periods.

Figures 3.4-2c, 3.4-3c, 3.4-4c, 3.4-5c, 3.4-6c and 3.4-7c depict the maximum allowable cooldown rates for the same operating periods.

The new figures also provide the revised maximum allowable heatup and cooldown rates versus indicated reactor coolant temperatures in a tabular form.

In addition, the proposed change revises the Lead Factor in Table 4.4-5 fr'om 1.15 to less than or equal to 1.5.

LCO 3.4.9.3 requires that at least one of the following three over-pressure protection systems be OPERABLE

(i) two gower-operated relief valves (PORVs) with a liftsetting of less than or equal to 470 psia, (ii) two shutdown cooling relief valves (SDCRVs) with a liftsetting of less than or equal to 350 psia, or (iii) the RCS depressurized with an RCS vent of greater than or equal to 3.58 square inches.

The proposed amendment revises the LCO to require that unless the RCS is depressurized with an RCS vent of at least 3.58 square

inches, at least two of following overpressure protection systems be OPERABLE

(i) one or more PORVs with a lift setting of less than or equal to 470 psia and with OPERABLE block valve(s),

when above the minimum cold leg temperature for PORV use for LTOP specified in Table 3.4-4, and (ii) one or more SDCRVs with a lift setting of less than or equal to 350 psia.

Table 3.4-4 contains the minimum cold leg temperature for PORV use for LTOP for each of the six operating periods during heatup and cooldown.

The proposed amendment also revises the APPLICABILITyby replacing the temperature limitation during heatup and cooldown for all the applicable MODES by the Table 3.4-3 for LOW TEMPERATURE OVERPRESSURE PROTECTION RANGE for MODES 3 and 4 only and,in addition, requiring that the LCO to be applicable in all of MODES 5

and, 6.

The proposed amendment 'also consolidates separate requirements for'ach'of the three overpressure protection systems into one.

Table 3. 4-3 contains the LTOP range for each of the six operating periods during heatup and cooldown.

The ACTION statements associated with this LCO have been accordingly revised to be consistent with the above revision to the LCO.

Definition 1.16 states that the LOW TEMPERATURE OVERPRESSURE PROTECTIVE RANGE is that operating condition when (1) the cold leg is less than or equal to 286 F during cooldown and less than or 0

equal to 295 F during heatup and (2) the reactor coolant system has 0

pressure boundary integrity. The reactor coolant system does not have pressure boundary integrity when the reactor coolant system is open to containment and the minimum area of the reactor coolant system is greater than 3.58 square inches.

The proposed amendment makes an administrative change to clarify the DEFINITION by removing an inappropriate reference to the pressure boundary integrity. The DEFINITION now states that the LOW TEMPERATURE OVERPRESSURE PROTECTION RANGE is that operating condition when (1) the cold leg temperature is less than or equal to that specified in. Table 3.4-3 for the applicable operating period and (2) the reactor coolant system is not vented to containment by an opening of at least 3.58 square inches.

d.

Footnotes

(*+) appended to LCO 3.4.1.3 and (NN) appended to LCO 3.4.1.4.1 state that a reactor coolant pump should not be started with one or more of the reactor coolant system cold leg temperatures less than or equal to 295 F during heatup or 286 F

0

~

0 during cooldown unless the secondary water temperature of each steam generator is less than 40 F above each of the reactor coolant system cold leg temperatures.

The proposed amendment revises the footnotes to state that a reactor coolant pump shall not be started with two idle loops and one or more reactor coolant system cold leg temperatures less than or equal to that specified in Table 3.4-3 for the applicable operating period unless the secondary water temperature of each steam generator is less than 40 F above each of the reactor coolant system cold leg temperature.

e.

The proposed amendment revises LCO 3.4.4 by appending a qualifier to the APPLICABILITYfor MODE 3. The LCO is now applicable in MODES 1 and 2, as before, and in MODE 3 when operating above the LTOP range of Table 3.4-3 for the applicable operating period.

f.

In addition, the proposed amendment revises the appropriate Bases.

g.

Finally, the proposed amendment revises the page numbers since a

significant number of pages are being added.'The.index is revised accordingly.

A detailed report of the methodology used to calculate the new ressure/temperature limit curves, and new LTOP requirements that rovide a basis for this amendment, is provided in Attachment 5.

ATTACHMENT3 NO SIGNIFICANTHAZARDS CONSIDERATION EJW I /033/5

NO SIGNIFICANT HAZARDS CONSIDERATION The standards used to arrive at a determination that a request for amendment involves no significant hazards consideration are included in the Commission s

regulation, 10 CFR 50.92, which states that no significant hazards considerations are involved if the operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences 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.

A discussion regarding each of the three standards follows.

(1)

Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

The pressure/temperature (P/T) limit curves in the Technical Specifications are conservatively generated in accordance with the fracture toughness requirements of 10 CFR 50, Appendix G

as supplemented by the ASME Code Section III, Appendix G.

The RTNDT values for the revised curves are based on Regulatory Guide 1.99, Rev.

2 (Draft) as discussed in Combustion Engineering Report

titled, "Methodology for Adjusted Reference Temperature Calculations,"

Reference (13).

The analysis of reactor vessel material irradiation surveillance specimens are used to verify the validity of the fluence predictions and the P/T limit curves.

Use of the revised curves in conjunction with the surveillance specimen program ensures that the reactor coolant pressure boundary (RCPB) will behave in a non-brittle manner and that the possibility of rapidly propagating fracture is minimized.

In conjunction with revising the P/T limit curves, a

low temperature overpressure protection (LTOP) analysis has been performed to confirm that the setpoints for the power-operated relief valves (PORVs) and shutdown cooling relief valves (SDCRVs) will provide the appropriate overpressure protection at the low temperatures.

To ensure compliance with the P/T limit curves, overpressure protection is provided to keep the reactor coolant system (RCS) pressure below the P/T limits for any given temperature following the initiation of an assumed pressure transient (energy-addition or mass-addition transient) while operating below the temperature at which the pressurizer safety valves provide overpressure protection during heatup and cooldown.

EJW/001

The revised pressure/temperature limit curves do not represent a

significant change in the configuration or operation of the plant.

Results of the LTOP analysis show that the limiting pressures for a given temperature are not exceeded for the assumed transients and that the reactor vessel and RCPB integrity is maintained.

Thus, the proposed amendment does not involve an increase in either the probability or the consequences of accidents previously evaluated.

(2)

Operation of the facility in accordance with the proposed amendment would not create the probability of a new or different kind of accident from any accident, previously evaluated.

The analysis performed has resulted in revised P/T limits based on the fracture toughness requirements of 10 CFR 50, Appendix G and the current LTOP system setpoints, which are based on the design basis energy-addition and mass-addition transients.

Since there is no significant change in the configuration or operation of the facility due to the proposed amendment, use of the revised P/T limits and/or LTOP setpoints will not create the possibility of a

new or different kind of accident from any accident previously evaluated.

(3)

Operation of the facility in accordance with the proposed amendment would not 'involve a significant reduction in a margin of safety.

The proposed change will not involve a significant reduction in a

margin of safety because the fracture toughness requirements of 10 CFR 50, Appendix G

are satisfied and conservative operating restrictions are applied for the purpose of LTOP.

In conclusion, based on the analysis performed and foregoing discussion, FPL has determined that the amendment request does not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a

new or different kind of accident from any accident previously evaluated, or (3) involve a

significant decrease in a margin of safety; and therefore does not involve a significant hazards consideration.

EJW/001

ATTACHMENT4 PRESSURE/TEMPERATURE LIMITS AND LOW TEMPERATURE OVERPRESSURE PROTECTION (4 EFPY to 32 EFPY).

REPORT EJW I /033/6

ABSTRACT Pressure/Temperature (P/T) limit and Low Temperature Overpressure Protection (LTOP) analyses were performed to ensure that Reactor Coolant Pressure Boundary integrity will be maintained in the low temperature modes of operation.

New P/T limits were developed based on fast neutron fluence predictions at 10, 15, 20, 25, 30, and 32 effective full power years (EFPY).

The design basis energy and mass addition pressure transients were analyzed based on the currently utilized means for transient mitigation that include the shutdown cooling relief valves and the PORVs.

The pressure transient analysis results were evaluated with these new P/T limits to yield a number of administrative and operational limitations to be implemented to ensure adequate LTOP.

The current LTOP system hardware and setpoints were assumed to be in effect up to the end of license which is currently at 32 EFPY.

The LTOP system was designed in accordance with the requirements set forth in the NRC Branch Technical Position RSB 5-2.

The system will prevent violation of 10 CFR 50, Appendix G limits provided the administrative controls identified in this report are implemented.

This report addresses the methodology and analytical models utilized in the

analyses, documents analysis results, and presents the administrative controls that need to be implemented in each operating period.

The report provided herein is in support of FPL's request to amend Facility Operating License No. NPF-16 for St.

Lucie, Unit No.

2.

TABLE OF CONTENTS Section Title

~Pa e No.

1.0 INTRODUCTION

2.0 LTOP SYSTEM 2-1 2.1 General 2.2 Design Criteria 2.3 Design Basis 2.4 Proposed Modifications 2-1 2-1 2-3 2-4 3.0 PRESSURE/TEMPERATURE LIMITS 3-1 3.1 General 3.2 Fast Neutron Fluence Analysis 3.3 Material Properties and Adjusted RTNDT 3.4 Pressure/Temperature Limit Analysis 3.5 Pressure and Temperature Correction Factors 3-1 3"1 3-3 3-6 3-8 4.0 PRESSURE TRANSIENT ANALYSES 4.1 General 4.2 Energy Addition Transients 4.3 Mass Addition Transients 4.4 Relief Valve Discharge Models 4.5 Results of Analyses 4-2 4-3 4-6 4-7

TABLE OF CONTENTS (Cont'd)

Section Title Pacae No.

5.0 LTOP EVALUATION 5.1 Introduction

5. 2 Controlling Pressures 5.3 Limiting Temperatures

5.4 Results

Limiting Conditions for Operation 5-1 5-1 5-3 5-11 6.0

SUMMARY

OF PROPOSED CHANGES 6-1 7.0 8.0 CONCLUS ION

'REFERENCES 7"1 8-1

Number Title LIST OF FIGURES

~Pa e No.

4-1 4-2 4-3 St.

Lucie-2 RCP Start Transient w/PORV, P

= 470 psia set St.

Lucie-2 RCP Start Transient w/SDCRV, P

= 350 psia set St.

Lucie-2 Garrett PORV Capacity Curves 4-9 4-10 4-11

Number Title LIST OF TA8 LES Pacee No.

3-1 3-2 St.

Lucie Unit 2 P/T Limit Fluences St.

Lucie Unit 2 Adjusted RTNDT Values for Plate M605-2 3-2 4-1 4-2 Mass Addition Transient Analysis Results Maximum Transient Pressures 4-8 5-1 5-2 5"3 Identification of Limiting Temperatures, 10 EFPY Identification of Limiting Temperatures, 15 EFPY Identification of Limiting Temperatures, 20 EFPY 5"5 5-6 5-7 5-4 5-5 Identification of Limiting Temperatures,

'I Identification of Limiting Temperatures, 25, EFPY 30 EFPY 5-8 5".9 5-6 5 "7 5-8 5-9 5"10 5"11 5-12 Identification of Limiting Temperatures, 32 EFPY LTOP Requirements, 10 EFPY LTOP Requirements, 15 EFPY LTOP Requirements, 20 EFPY LTOP Requirements, 25 EFPY LTOP Requirements, 30 EFPY LTOP Requirements, 32 EFPY 5-1 0 5"12 5-13 5-14 5-15 5"16 5"17

0

1. 0 I NTRODUCTION Pressure/Temperature (P/T) limits and Low Temperature Overpressure Protection (LTOP) were analyzed to ensure that Reactor Coolant Pressure Boundary integrity will be maintained in the low temperature modes of operation.

The analyses were performed for the operating periods ending at 10, 15, 20, 25, 30, and 32 effective full power years (EFPY),

(i.e., up to the end of license).

New P/T limits were developed and a number of administrative controls which need to be implemented were identified.

The current LTOP system including a combination of the shutdown cooling relief valves (SDCRVs) and power operated relief valves (PORVs), with the current setpoints, was assumed to be in effect in each operating period.

This LTOP system provides assurance that 10 CFR'50, Appendix C limits will not be violated during both normal operation and overpressurization events due to equipment malfunction or operator error.

Operation of the LTOP system will not involve a reduction in the margin of safety presently afforded by the Technical Specifications.

This report addresses the methodology and analytical models utilized in the analyses, documents analysis results, and presents the administrative controls that need to be implemented in each operating period.

The report provided herein is in support of FPL's request to amend Facility Operating License No.

NPF-16 for St.

Lucie, Unit No. 2.

1-1

2. 0 LTOP SYSTEM 2.1 GENERAL The current LTOP system makes use of two SDCRVs (i.e., V3666 and V3667) at lower RCS temperatures and two PORVs (i.e., V1474 and V1475), in the remaining part of the LTOP temperature range.

The system is described in Reference 1.

The analyses performed indicate that changes to the Technical Specifications (Reference

2) and, consequently, to operating procedures need to be implemented for the system to continue providing adequate LTQP beyond 4 EFPY.

These changes affect relief valve alignment temperatures, P/T limits, heatup and cooldown rates, and requirements for RCP operation.

The changes are addressed in the following sections of this report.

No hardware modifications are required to make the current LTOP system adequate for operation up to 32 EFPY.

2.2 DESIGN CRITERIA The LTOP system meets the design criteria set forth in Branch Technical Position RSB 5-2 (Reference

3) as follows:

1.

The system will prevent exceeding the applicable P/T limits during all anticipated overpressurization events.

The analyses performed show that by using either a SDCRV or a PORV and implementing a

number of administrative and operational controls peak RCS pressures will not exceed 10 CFR 50, Appendix G pressure limits for given temperatures, even while the RCS.is in a water-solid condition.

2.

The syst'm will be able to perform its function assuming any single active component failure in addition to the failure that initiated the pressure transient.

The most limiting single failure was assumed to 2-1

be a loss of one relief valve, be it a SDCRV or a PORV.

-Accordingly, only one relief valve was utilized for transient mitigation in pressure transient analyses.

The analyses performed also assumed the most limiting allowable operating conditions and systems configuration at the time of the postulated overpressurization event.

3.

The PORVs, isolation valves, associated interlocks and instrumentation are designed to Quality Group A, Seismic Category I requirements.

The interlocks and instrumentation associated with the PORVs satisfy the appropriate criteria of lEEE-279.

The SDCS suction line valves, interlocks and instrumentation are also designed to: Quality Group A (except for. the SDCRVs.V3666.

and V3667), Seismic Category I requirements.

The SDCRVs are designed to Quality Group B requirements.

Although the requirements of IEEE-279 are not directly applicable to the SDCS isolation valve interlocks, the latter mostly comply with Section 4 of this standard as shown in Reference 4.

A contr'ol room operator can enable the LTOP system by switching the PORV mode selector switch to the low pressure setpoint position during cooldown when appropriate pressure and/or temperature conditions occur.

An alarm will alert the operator to enable the system.

An alarm is also provided to inform the operator if the PORVs have received a signal to open, which happens when the RCS pressure reaches the PORV setpoint.

Alignment of the SDCS is accomplished via remote (from the control room or from the local control station) operated valves.

4.

The LTOP system is testable.

The testing requirements are included in the Technical Specifications (Reference 2).

2-2

S.

The PORVs were designed and manufactured in accordance with ASME Boiler and Pressure Vessel Code Section IV and are Class I

valves.

The SDCRVs were designed to 1974 ASME Code Section NC (to 1975 Summer Addenda)

Quality Group B.

6.

Each PORV is powered from a separate DC control bus.

The SDCRVs are self-actuating spring loaded liquid relief valves which do not require control circuitry and open when pressure at valve inlets exceeds their setpoint.

The LTOP system, therefore, does not depend on the availability of offsite power to perform its function.

7.

When LTOP is provided by the SDCRVs the overpressure protection function will not be defeated by interlocks because the maximum transient pressures are lower than the autoclosure setpoint.

2.3 DESIGN BASIS 1.

The LTOP system is designed to preclude violation of PIT limits which are applicable up to 32 EFPY.

The P/T limits for 10, 15, 20, 25, 30, and 32 EFPY are addressed in Section 3.0 of this report.

2.

In the design of the LTOP system, the following overpressurization events provided the design basis:

I.

Mass Addition Events (1)

Actuation of two HPSI pumps with all three charging

pumps, and (2)

Actuation of a single HPSI pump with all three charging pumps.

2-3

I I. 'ner y Addition Events (1)

Reactor coolant pump start, with a positive secondary-to-primary temperature differential, (2)

Decay heat addition, due to SDCS isolation, and (3)

Full pressurizer heater capacity.

3.

A comparison between the maximum pressures in the most limiting events and the P/T limit curves yielded a number of administrative controls.

Implementation of these limitations will provide assurance that P/T limits will not be exceeded during any overpressurization event.

2.4 PROPOSED MODIFICATIONS Hardware No modifications to the current LTOP system hardware and setpoints are needed for operation beyond 4 EFPY.

Administrative Controls Tables 5-7 through 5-12 contain the administrative controls which have to be implemented in each operating period.

These requirements are as follows:

l.

The temperatures at which the SDCRVs are required to be aligned to the RCS.

2.

The temperatures at which the PORVs are required to be aligned to the RCS.

2-4

3.

Limitations on RCS heatup rates.

4.

Limitations on RCS cooldown rates.

Technical Specifications New P/T limits and LTOP requirements for each operating period need to be incorporated into St.

Lucie Unit 2 Technical Specifications.

The changes are indicated on marked up copies of the affected technical specifications and corresponding bases.

A list of the affected technical specifications is as follows:

1.16 3.4.1.3 3.4.1.4.1 3.4.4 3/4.4.9 3.4.9.3 A complete list of the affected technical specifications and bases pages including new pages is provided in Reference 7.

2-5

3.0 PRESSURE/TEMPERATURE LIMITS 3.1 GENERAL The new P/T limits were calculated in accordance with the requirements of 10 CFR Part 50, Appendix G (Reference 8).

The limits were developed based upon the recommendations of Appendix C of the ASME Boiler and Pressure Vessel Code Section III (Reference 9).

These limits are dependent upon the initial reference nil-ductility transition temperature (RTNDT) for the limiting materials in the reactor vessel beltline and closure flange juncture regions, and upon the increase in RTNDT resulting from fast neutron irradiation damage to the beltline materials.

The P/T limits are presented in the proposed Technical Specification change'package (Reference 7).

3.2 FAST NEUTRON FLUENCE ANALYSIS The St.

Lucie Unit 2 P/T limits for 10-32 EFPY have been calculated using surface fluences as delineated in Table 3-1 at the position of the limiting material through 32 EFPY.

This data was calculated based upon a detailed octant symmetric Discrete Ordinates Transport (DOT) model of the core to vessel configuration.

The core power distribution was modeled in a detailed manner using pin wise power distribution data.

The core power distribution was chosen to bound future fuel management strategies with respect to the fluences accumulated in the reactor vessel.

Ordinary out-in and low leakage in-out fuel management strategies will be bounded.

The fluence values in Table 3-1 are quoted for the peak fluences 'at.a given vessel radius.

The 3/4t fluence values were adjusted to account for the backscatter due to the wall and cavity structures.

The values are based upon 2700 Mwt power operation and an 8.625 inch vessel wail thickness.

3-1

TABLE 3-1 St.

Lucie Unit 2 P/T Limit Fluence Peak Fluence in the Azimuthal Direction Units are 10 N/cm Irradiation Time (EFPY)

I. D.

Radial Position 1/4t 3/4t 10 15 20 25 30 32 0.91 1.8 2.7 3.6 4.5 5.4 5.8 0.49 0.97 1.5 1.9 2.4 2.9 3.1 0.11 0.22 0.32 0.43 0.54

~

0.65 0.70 Note:

The end of life fluence values used in this table are deliberately more conservative than those submitted in Florida Power 8 Light letter L-86-25 dated January 23, 1987, in order to bound future core loading patterns.

3-2

The surveillance capsule report, BAW-1880 (Reference

10) contains the analysis of the dosimetry of surveillance capsule W-83.

The calculated peak neutron fluences at 32 EFPY using the results from capsule W-83 dosimetry is 3.64 x 10 n/cm (E>1MeV).

The calculated peak neutron 19 19 2

fluences for the Pressure/Temperature limits at EOL is 5.8 x 10 n/cm (E>1MeV) based upon the pin wise power distribution data which bounds future fuel management strategies.

The fluences used for the P/T limit analyses conservatively estimate the 32 EFPY fluence at the limiting material, shell plate M605-2.

3 ~ 3 MATERIAL PROPERTIES AND ADJUSTED RTNDT The St. Lucie Unit 2 reactor vessel was manufactured to ASME Code requirements for initial material properties.

The initial RTNDT of the various reactor vessel plates, forgings, and welds, along with the initial RTNDT of materials outside of the beltline, provide the basis for the P/T limits.

The maximum RTNDT associated with the stressed region of the reactor vessel flange during boltup is 50'F and is located at the vessel flange iuncture.

This maximum RTNDT per the ASME Code Section Ill, Article G-2222 (Reference ll) when corrected for temperature instrument uncertainty of 8~F establishes the-minimum boltup temperature.

For conservatism, the minimum boltup temperature is established at 80~F.

The maximum RTNDT for the balance of the Reactor Coolant System components, excluding the reactor vessel, is 60~F.

This maximum RTN DT is associated with the Reactor Coolant System piping and controls the Lowest Service Temperature.

The Lowest Service Temperature is defined as equal to the maximum RTNDT for the balance of RCS components plus 100 F per the ASME Code Section III Article NB 2332 (Reference 12).

Therefore, the Lowest Service Temperature is 168 F when corrected for the temperature instrument uncertainty of 8 F.

3"3

The limiting material in the reactor vessel beltline is the intermediate shell plate M605-2.

The initial RT<DT for intermediate shell plate M605-2 is 10 F.

The copper and nickel content by percent weight for this limiting plate are 0.13%, and 0.62% respectively.

The margin on the adjusted RTNDT is 34 F based upon a al

= 0 and a o

= 17 F.

The 4

intermediate shell plate M605-2 is the limiting material because of its high copper and nickel content and because the results of the first St. Lucie Unit 2 Surveillance capsule removed plate M605-1 as the limiting material.

The adjusted RTNDT values for plate M605-2 as a function of time are presented in Table 3-2.

The RTNDT of the intermediate shell plate M605-2 increases over time due to fast neutron irradiation damage.

The adjusted RTNDT values are calculated using draft Regulatory Guide 1.99, Revision 02 as discussed in Combustion Engineering Report titled. "Methodology for Adjusted Reference Temperature Calculations,"

Reference (13).

3-4

TABLE 3-2 St.

Lucie Unit 2 Adjusted RT for Plate M605-2 Values EFPY ARTNDT ( F) 1/4t 3/4t Adj. RTNDT~ F 1/4t 3/4t 10 15 20 25 30 32 91 102 108 113 117 119 54 63 70 76 80 82 135 146 152 157 161 163 98 107 114 120 124 l26

3,4 PRESSURE/TEMPERATuRE LIMIT ANALYSIS The criteria from Section III of the ASME Code Article C-2000 (Reference

9) were used to calculate the P/T limits associated with a postulated flaw in the reactor vessel.

Flaws are postulated on the inside surface and on the outside surface of the pressure vessel.

The distance is measured from the inner radius of the vessel outward.

The flaws on the inside surface and outside surface of the vessel are referred to by location as 1/4 thickness (1/4t) and 3/4 thickness (3/4t), respectively.

The postulated flaw is a semi-elliptical surface flaw oriented in the axial direction with a 1:6 aspect ratio.

The flaw is assumed to have a depth of 1/4 of the reactor vessel wall thickness or 2.16 inches.

The Reference 9 procedure is based upon the principles of linear elastic fracture mechanics and involves a stress intensity factor prediction which Is a lower bound of the static, dynamic and crack arrest critical values.

This conservative methodology insures the P/T limits provide assurance that the RCPB behaves in a non-brittle manner and the probability of rapidly propagating fracture is minimized.

Several critical locations of the reactor vessel were considered.

At each location being analyzed, a maximum postulated flaw was assumed.

At the same location. the Mode I stress Intensity factor. Kl. Was calculated for the specified pressure and thermal loadings.

The sum of the Kl values was compared to a reference stress intensity value, KIR, which is the highest critical value of K based upon static, dynamic, and crack arrest fracture toughness values that can be ensured for the material and temperature involved.

The reactor vessel beltline, flange juncture, and nozzle regions have been evaluated using the postulated defects permitted by Reference 9.

P/T limits have been established for inservice hydrostatic tests, for various heatup and cooldown rates, and for core critical operation.

For inservice hydrostatic tests, the criteria from Reference 9 requires KIR to be at least 1.5 times the Kl caused by pressure.

Consequently, the 3-6

inservice hydrostatic test P/T limits are based upon the fracture mechanics expression 1.5 KIM < KIR.

The flaw on the inside surface of the intermediate shell plate is limiting for the inservice hydrostatic tests.

For non-critical operation with various heatup and cooldown rates, the criteria from Reference 9 require KIR to be at least the sum of 2.0 times the K

caused by pressure and the Kl caused by thermal gradients.

I Consequently, the P/T limits for various heatup and cooldown rates are based upon the fracture mechanics expression 2.0 KIM + KIT < KIR.

During cooldown, a flaw on the inside surface location of the intermediate shell plate is always limiting since the adjusted RT>DT is higher than that at the 3/4t location and both the thermal and pressure stresses are tensile.

During heatup, either a flaw in the intermediate shell plate inside surface or outside surface location is limiting.

Therefore, the heatup P/T limits are composite curves based upon both the 1/4t and 3/4t flaw locations.

Limits for core critical operation are provided as required by 10 CFR Part 50, Appendix G (Reference 8).

These limits are applicable when the core is critical except for when the core is critical for low power physics tests.

These core critical P/T limits are strictly based upon fracture fracture mechanics considerations.

The core critical P/T limits are defined and established as being at least 40OF above the minimum permissible temperature corresponding to the limiting heatup or cooldown P/T limit nor lower than the minimum permissible temperature for the inservice system hydrostatic test pressure.

Although the flange and nozzle regions were evaluated using a detailed fracture mechanics procedure in accordance with the ASME Code Section III, Appendix G requirements, the P/T limits for the nozzle regions were not as limiting as the reactor vessel beltline limits.

Consequently, the composite P/T limit curves are based upon beltline and flange region Appendix G considerations.

3"7

3.5 PRESSURE AND TEMPERATURE CORRECTION FACTORS The P/T limit curves have taken into account allowances for instrument, hydrostatic, and hydrodynamic errors in sensing actual pressure and temperature conditions.

The pressure correction factors used to develop these curves vary as a function of flowrate in the RCS.

The pressure correction factors utilized include instrument uncertainties and elevation head differences between the vessel beltline and pressurizer pressure instrument tap.

The pressure correction factors used to correct the analysis results to indicated pressurizer pressure are as follows:

Pressure Indicated (PSIA)

Cold Leg Temperature Correction Factor (T

oF)

(PSIA) c

> 750

> 750

< 750

< 750

> 200

< 200 200

< 200 130 116 90 75 The temperature correction factor utilized is the actual temperature instrument loop uncertainty as obtained through the statistical combination of the individual component errors.

The temperature correction factor utilized to correct the analysis results to indicated cold leg temperature is 6.0 F.

3-8

4.0 PRESSURE TRANSIENT ANALYSES 4.1 CENERAL An analysis of the most limiting energy and mass addition transients was performed to identify maximum (peak) transient pressures in the RCS that could develop in the case of an overpressurization event.

The same design basis events as those considered in the design of the original St.

Lucie 2 LTOP system (Reference

4) were re-analyzed using modified analytical models and assumptions.

These events are as follows:

.RCP start, with a positive secondary-to-primary temperature differential (energy addition),

Inadvertent actuation of two HPSI pumps with simultaneous operation of three charging 'pumps (mass addition); and Inadvertent actuation of one HPSI pump with simultaneous operation of three charging pumps (mass addition).

The energy addition transient analysis was performed for a number of PORV setpoints in order to generate a sufficient data base to help identify optimal LTOP characteristics.

For the same

reason, in addition to the design basis mass addition events, other mass addition events were analyzed as well.

In the transient analyses with a SDCRV, the current SDCRV setpoint of 350 psia (at the valve inlet) was utilized.

The maximum pressures in these transients do not exceed 1105 of the design pressure of any component in the SDCS.

This was demonstrated in Reference (6) using the RCP start transient results for comparison.

Conservative assumptions were used in the analyses to maximize transient pressures.

These assumptions are as follows:

Water-solid pressurizer, initially at saturation conditions.

4-1

1% decay heat and full pressurizer heater capacity are additional inputs to the RCS.

Pressure transients are mitigated by a single relief valve.

Letdown flow paths are isolated.

No heat absorption or metal expansion in the RCPB.

PORV discharge model takes into account liquid subcooling at the valve inlet.

The pressure transients and relief valve analytical models are addressed in the following subsections.

The analysis results, per Subsection 4.5, are those used in establishing the current St.

Lucie Unit 2 LTOP.

system.

4.2 ENERGY ADDITION TRANSIENTS The pressure transient due to a RCP start when the secondary steam generator inventory is at a higher temperature than the primary inventory is the most limiting energy addition transient.

Accordingly, only this transient was analyzed.

Assuming a secondary-to-primary temperature differential of 40~F.

The same computer code as that described in Reference (5) was used in the analysis.

The code treats the relief valves and the RCS as having the same pressure during the transients.

Since valve capacity is a function of pressure and in the code RCS pressure is used to calculate valve flow rate, pressure transient results of the code were assumed to represent pressure transients at the valve inlets.

For further application, maximum pressures at the valve inlets were translated into the pressurizer pressure.

In the case with a PORV, the correction was made by adding PORV inlet piping pressure drops during 4-2

liquid discharge to the maximum PORV inlet pressures identified from the code outputs.

In the case with a SDCRV, the maximum pressurizer pressure was found by subtracting the pressure differential due to the elevation difference between the valve and the top of the pressurizer from the maximum SDCRV inlet transient pressure.

The. results of the computer analyses are presented in Figures 4-1 and 4-2.

Note that although the analyses were performed for a number of PORV setpoints as indicated in Subsection 4.1, only the transient with P

= 470 psia is presented herein (Figure 4-1).

The reason is that set this setpoint was selected for the subject LTOP system as described in Section 2.0.

Figure 4-2 provides the pressure transient mitigated by a SDCRV with the current setpoint of 350 psia, which was also selected for the,LTOP system.

The following maximum pressurizer pressures were determined as a result of the energy addition transient analysis:

RCP start w/PORV:

P

= 535 psia max RCP start w/SDCRV:

P

= 343 psia max These pressures apply at ail RCS temperatures in the I TOP range.

4.3 MASS ADDITION TRANSIENTS A number of mass addition events based on possible combinations of HPSI and charging pumps were analyzed.

Out of those analyzed, two events provided a basis'for the LTOP system.

These include an inadvertent actuation of either one or two HPSI pump(s), with simultaneous operation of three charging pumps in each case.

4"3

The same methodology as that used in previous similar analyses for both St.

Lucie Unit 1 and Unit 2 was utilized.

More specifically, the relief valve capacity curves were superimposed on the pump delivery curves to determine the equilibrium pressures at which mass inputs into the RCS match relief valve discharges.

The maximum pressures were then determined depending on a correlation between the equilibrium pressure and the valve opening pressure (setpoint) applying the following rationale:

When the equilibrium pressure exceeds the relief valve opening pressure, the transient pressure will continue rising following valve opening until it reaches a peak equal to the equilibrium pressure.

In this case, the maximum pressurizer pressure and the equilibrium pressure are identical, for both a PORV and a SDCRV.

When the equilibrium pressure is less than the relief valve opening

pressure, the maximum pressures will equal the valve opening pressures adjusted to the top of the pressurizer elevation as described in Subsection 4.2.

The results of the mass addition transient analysis are presented in Table 4-1.

Note that the maximum pressures in the table are applicable at all RCS temperatures in the LTOP range.

4-4

TABLE 4-1 Mass Addition Transient Anal sis Results Input Equilibrium Pressure Maximum Pressure (1)

~

(1) w/PORV w/SDCRV w/PORV w/SDCRV (2)

(3) 2 HPSI S

3 Charging Pumps 535 psia 355 psia 535 psia 355 psia 1

HPSI S

3 Charging Pumps 375 psia 345 psia 492 psia 345 psia Note:

(1)

The pressures are indicated pressurizer pressures.

(2)

Based on a PORV opening pressure (setpoint) of 470 psia.

(3)

Based on a SDCRV opening pressure of 341 psia.

4-5

4,4 RELIEF VALVE DISCHARGE MODELS PORV Previous pressure transient analyses employed a "subcooled water, non-flashing" PORV discharge model that assumed a liquid flow all the way through the valve outlet.

While this model yields conservative flow rates at the higher degrees of liquid subcooling at the valve inlet, overestimated PORV capacity would result had the model been used under the lower subcooling conditions.

The latter could occur in the case of a transient when pressure rise from the initial saturation pressure up to the PORV opening pressure would not be sufficient to sustain liquid subcooling down to the valve outlet.

As a result, a flashing, or two-phase flow, at the valve outlet would occur, thus decreasing valve relieving capacity.

A new PORV discharge model was developed.

The model assumed liquid subcooling through the PORV and flashing at the outlet.

It was determined that based on the initial saturation pressure of 300 psia, the new model yields more conservative results than the existing model at the PORV inlet pressures of less than 600 psia.

Consequently, the PORV capacity curve was calculated using both models, with the existing model applied at P

> 600 psia.

The PORV capacity curve is presented in Figure 4-3 in two modifications:

Curve 1:

Flow rate as a function of PORV inlet pressure.

(1)

Initial saturation conditions in the pressurizer were conservatively assumed in the pressure transient analyses.

4-6

Curve 2:

Flow rate as a function of the pressurizer pressure.

A difference between the curves is the PORV inlet piping pressure drop.

SDCRV The SDCRV discharge model used in the pressure transient analyses is consistent with the ASME Code requirements for relief valves, i.e., the valve should open at 103't of the set pressure (setpoint) and rated lift should occur at 10% accumulation.

The SDCRV opening pressure adjusted to the pressurizer is 341 psia.

The valve was assumed to discharge subcooled water.

4.5 RESULTS OF ANALYSES The final results of the pressure transient analyses are presented in Table 4-2.

4-7

0

TABLE 4-2 Maximum Transient Pressures Transient PORV Relief Valve SDCRV RCP Start 535 psia 343 psia 2 HPSI S

3 Charging Pumps 535 psia 355 psia 1 HPSl 3 Charging Pumps 492 psia 345 psia 4-8

FIGURE 4-1 ST. LUCIE-2 RCP START TRANSIENT W/PORV, PSET = 470 PSIA b,t.p = 40~F 600 PMAX= 474.5 PSIA 450 D

V) 400 I

LQ z) 0 360 300 0 10 ELAPSED TIME, SECONDS 15 20 n g

fIGURE 4-2 ST. LUCIE-2 RCP START TRANSIENT W/SDCRV, PSf T ~ 350 PSIA 4t.p ~400F 400 PMAX 3621 CO 350 LLI z

LLl) 0 LL Lll ChOa CO 300 0

10 ELAPSED TIME, SECONDS 15 20 4-10

FIGURE 4-3 ST. LUCIE-2 GARRETT PORV CAPACITYCURVES 1200

< 800 h

D 600 NOTES:

~ Curve 1 represents flow rate as a function of PORV inlet pressure.

~ Curve 2 represents flow rate as a function of the indicated pressurizer pressure

~ hPiniet is the PORV inlet PiPing pressure drop during valve discharge

~PINLET 400 200 800 1000 1200 1400

1600 PORV FLOWRATE, GPM 1800 2000 2200

5.0 LTOP EVALUATION

5.1 INTRODUCTION

The current St.

Lucie Unit 2 LTOP system makes use of two PORVs (P

= 470 psia), two SDCRVs (P

= 350 psia),

and an RCS vent (A >

set set 3.58 sq. in.).

The LTOP system characteristics such as relief valve alignment temperatures, RCS heatup and cooldown rates, and restrictions on RCP and HPSI pump operation were established based on the P/T limits for 4 EFPY and results of analyses of the design basis'transients.

The existing LTOP system hardware and setpoints are intended to remain unchanged in the operating periods beyond 4 EFPY.

Consequently, the results of the analyses of the design bases transients, per Section 4.0, are also applicable beyond 4 EFPY.

P/T limits, however, change, i.e.,

become more restrictive with time, thus affecting LTOP system characteristics, especially valve alignment temperatures and heatup and cooldown rates.

To identify these temperatures and rates, P/T limits for 10, 15, 20, 25, 30, and 32 EFPY were evaluated together with the pressure transient analysis results.

The identified characteristics for each operating period were then included in St.

Lucie Unit 2 Technical. Specifications.

The following subsections provide details of the evaluation and results.

5. 2 CONTROLLING PRESSURES A controlling pressure generally identifies an RCS pressure limit which will not be exceeded during any overpressurization event that could occur in the corresponding temperature region. while mitigated by an applicable relief valve.

When applied to Appendix G P/T limit curves, a

controlling pressure also provides a lower bound pressure limit for these curves.

In other words, a controlling pressure is more limiting than the 5-1

P/T limit curves above it.

Therefore, no P/T limits above the I

controlling pressure will be exceeded during normal operation or an overpressurization event.

As far as the P/T limits below the controlling pressures are concerned, restrictions on heatup and cooldown rates (which are a part of LTOP requirements) prevent operation based on these limits.

In this evaluation, the controlling pressures were determined based on the data in Table 4-2.

In the PORV-mitigated transients, the controlling pressure was assumed to equal the highest maximum transient pressure of 535 psia that is applicable at all RCS temperatures at which LTOP is provided by the PORVs.

In the SDCRV-mitigated transients,

however, the second highest maximum transient pressure (345 psia) was assumed to represent the controlling pressure, for the reasons given below.

According to the Technical Specifications (Reference 2, Page 3/4 5-7),

one HPSI pump is required to be rendered inoperable at T

< 200 F.

Therefore, the mass addition transient due to 2 HPSI and 3 charging pumps could only occur at T

> 200 F, yielding a maximum pressure of c

355 psia if mitigated by a SDCRV.

However, no intersections exist between a horizontal line corresponding to this pressure and any P/T limit curve.

Thus, P = 355 psia does not control any heatup or cooldown rate.

As a result, with only one HPSI pump operable, the mass addition transient due to one HPSI and three charging. pumps yields the highest transient pressure (345 psia) between the two remaining transients mitigated by a SDCRV.

5-2

Thus, the following controlling pressures were established for use in the identification of the limiting temperatures:

P

= 305 psia, in the temperature region where LTOP is provided contr1 by the SDCRVs, P

2 = 535 psia, in the temperature region where LTOP is provided contr2 by the PORVs.

5. 3 LIMITl NG TEMPERATURES Most of the limiting temperatures were determined at the intersections between one of the controlling pressures and appropriate P/T limit curves.

Thus, the temperatures at which heatup rates can be raised or cooldown rates must be reduced were identified based on Pcontrl '

~

ntr2 determined the temperatures at which the LTOP function can be transferred from the SDCRVs to the PORVs, during heatup, or must be transferred back to the SDCRVs, during cooldown.

The maximum LTOP temperatures were determined at the intersections between a safety valve setpoint of 2500 psia and the P/T limit curves for heatup at 50 F/hr'nd cooldown at 100 F/hr.

Note that these temperatures generally identify upper bounds for the regions in which LTOP is required.

Since in St.

Lucie Unit 2 LTOP at the higher temperatures is provided by the PORVs, the maximum LTOP temperatures also identify upper temperature limits for PORV alignment.

The limiting temperatures are summarized in Tables 5-1 through 5-6, with each table covering one operating period.

The following notes apply to each table:

5-3

(1)

Minimum allowable temperature, so as not to exceed the applicable P/T curve pressure limit, for the controlling pressure indicated.

(2)

Safety valve setpoint; used to identify TLTOP for heatup and cooldown.

(3)

For 50OF/hr.

(4)

For 100~F/hr.

5-4

C O

'U OO Om U

LL 0

0 CJl L

S I

U LL LL 0

0 0

lh O

O Ml L

4 4O LL LI LL 0

0 0

O O'

Fl O

CO

~

~

V I

tQ C

I-Ui C

E Q.

Pvx LL 0

Pl U

U 0

0 tO O

lA 00

~

V I

I I

I I

I I

I I

CO IOU C

CJ'0 CO OlIIN V)

O C

OI COO IO a

N N

Q.

Q.

Q.

LA tA O

m O

lh Fl LA PV I

Cl CJ NQ OC 5-5

TABLE 5-2 identification of Limiting Tem eratures, 15 EFPY Controlling Pressure 2500 psia~

~

535 psia 345 psia 345 psia 345 psia 345 psia 324 F

165 F

89 F~3~

<80 F for 40 F/hr 315~F 190~F "

1570F 141 F for 75 F/hr 115 F for 50 F/hr

<80 F for 30 F/hr Limitin Temperature 6 P/T Limit Curve

~Heatu Cooldown For notes refer to Subsection 5.3.

TABLE 5-3 Identification of Limitin Tem eratures, 20 EFPY Limitin Tem erature 8 P/T Limit Curve (1)

Control Iin Pressure Heatup Cooldown 2500 psia

- (2) 535 psia 345 psia 345 psia 345 psia 345 psia 330'F(

)

172 F

96 F'"

<80'F for 40 F/hr 321 F

196 F

147oF fol 75oF/hr 121 F for 50 F/hr

<80 F for 30 F/hr For notes refer to Subsection 5.3.

TABLE 5-4 Identification of Limitin Tem eratures, 25 EFPY Limitin Temperature 8 P/T Limit Curve Control Iin Pressure 2500 psia

~ (2) 535 psia 345 psia 345 psia 345 psia 345 psia 345 psia

~Heatu 3350F 178 F

102~F

<80 F for 40 F/hr Cooldown 326~F 201 F

168 F

152 F for 75 F/hr 126 F for 50 F/hr 83 F for 30 F/hr

<80 F for 20 F/hr For notes refer to Subsection 5.3.

TABLE 5-5 Identification of Limitin Tem eratures, 30 EFPY Limitin Tem erature O P/T Limit Curve (1)

Control Iin Pressure

~Heatu Cooldown 2500 psia

- (2) 535 psia 345 psia 345 psia 345 psia 345 psia 345 psia 339 F

182~F 106~F

<80'F for 40 F/hr 330'F(4) 205~F(

172'F(4) 156'F for 75 F/hr 130 F for 50 F/hr 87oF for 30oF/hr

<80 F for 20"F/hr For notes refer to Subsection 5.3.

TABLE 5-6 Identification of Limitin Temperatures, 32 EFPY Limitin Tem erature I P/T Limit Curve (1)

Controllin Pressure

~Heetu Cooldown 2500 psia

- (2) 535 psia 345 psia 345 psia 345 psia 345 psia 345 psia 184~F 1080F

<80'F for 40 F/hr 332 F

207'F(4) 174~F "

158 F for 75 F/hr 132"F for 50 F/hr 89 F for 30 F/hr

<80'F for 20 F/hr For notes refer to Subsection 5.3.

5.Q RESULTS:

LIMITING CONDITIONS FOR OPERATION The temperature requirements for aligning the SDCRVs and PORVs for LTOP, and the limitations on heatup and cooldown rates are derived from the data in Tables 5-1 through 5-6.

The LTOP requirements for each operating period are summarized in Tables 5-7 through 5-12.

It should be noted that during heatup, the LTOP function can be transferred from the SDCRVs to the PORVs at any temperature above that required for SDCRV alignment (e.g.,

156~F in Table 5-7).

During cooldown, however, the SDCRVs must take over the LTOP function upon reaching the indicated temperature such as 179~F in Table 5-7.

5"11

e TABLE 5-7 LTOP Requirements, 10 EFPY Both SDCRVs are required to be aligned to the RCS as follows:

During heatup, at T

< 156 F

During cooldown, at T

< 179~F c

Both PORVs are required to be aligned to the RCS as follows:

During heatup, at 156 F

< T

< 313F During cooldown, at 179 F

< T

< 304<'F-c Heatup rates shall be limited to a maximum of:

50oF/hr, at all temperatures Cooldown rates shall be limited to a maximum of:

30~ F/hr, at T

< 104'F c

50aF/hr, at 104~F

< T

< 130~F c

75~ F/hr, at 130 F

< T

< 146 F

c 100~F/hr, at T

> 146~F c

5-12

TABLE 5-8 LTOP Requirements, 15 EFPY Both SDCRVs are required to be aligned to the RCS as follows:

During heatup, at T

< 165 F

During cooldown, at T

< 190oF c

Both PORVs are required to be aligned to the RCS as follows:

During heatup, at 165 F

< T

< 324 F

During cooldown, at 190oF

< T

< 315 F

Heatup rates shall be limited to a maximum of:

40oF/hr, at T

< 89oF c

50oF/hr, at T

> 89oF c

Cooldown rates shall be limited to a maximum of:

30oF/hr, at T

< 115oF c

5PoF/hr, at115 F< T

<141 F

c 75oF/hr at 141 F

< T

< 157 F

c 1ppoF/hr, at T

> 157oF c

5>>13

TABLE 5-9 LTOP Requirements, 20 EFPY Both SDCRVs are required to be aligned to the RCS as follows:

During heatup, at T

< 172~F During cooldown, at T

< 196~F Both PORVs are required to be aligned to the RCS as follows:

During heatup, at 172 F

< T

< 330 F During cooldown, at 196~F

< T

< 321 F

Heatup rates shall be limited to a maximum of:

40oF/hr, at T

< 96F c

50oF/hl at T

> 96~F c

Cooldown rates shall be limited to a maximum of:

30oF/ht at T

< 121'F c

50~ F/hr, at 121 F

< T

< 147'F c

75~F/hr, at 147 F

< T

< 163 F

c 100'F/hr,

) 163oF c

5-14

TABLE 5-10 LTOP Requirements, 25 EFPY Both SDCRVs are required to be aligned to the RCS as follows:

During heatup, at T

< 178~F During cooldown, at T

< 201~F c

Both PORVs are required to be aligned to the RCS as follows:

During heatup, at 178~F

< T

< 335 F c

During cooldown, at 201 F < T

< 326 F c

Heatup rates shall be limited to a maximum of:

qpoF/hr, at T

< 102~F c

50oF/hr at T

> 102~F c

Cooldown rates, shall be limited to a maximum of:

2poF/hr, at T

< 83~F c

30oF/hr at83F<

T

<126F c

5poF/hr, at 126oF

< T

< 152aF c

75~ F/hr, at 152 F < T

< 168 F c

1ppoF/ht at T

> 168~F c

5-15

TABLE 5-11 LTOP Re uirements, 30 EFPY Both SDCRVs are required to be aligned to the RCS as follows:

During heatup, at T

< 182aF During cooldown, at T

< 205'F c

Both PORVs are required to be aligned to the RCS as follows:

During heatup, at 182 F

< T

<<330 F

During cooldown, at 205 F < T

< 339 F

Heatup rates shall be limited to a maximum of:

gpoF/hr, at T

< 106aF c

50aF/hr at T

> 106aF c

Cooldown rates shall be limited to a maximum of:

20aF/hr, at T

< 87 F c

3paF/hr at 87 F

< T

< 130aF c

50oF/hr, at 130 F

< T

< 156aF c

75oF/hr at 156 F

< T

< 172 F

c 100oF/hr, at T

> 172aF c

5"16

0,

TABLE 5-12 LTOP Requirements, 32 EFPY Both SDCRVs are required to be aligned to the RCS as follows:

During heatup.

at Tc

< 1840F During cooldown, at T

< 207'F c

Both PORVs are required to be aligned to the RCS as follows:

During heatup, at 184 F < T

< 341 F

During cooldown, at 207 F

< T

<.332 F

Heatup rates shall be limited to a maximum of:

40~ F/hr, at T

< 108~F c

50oF/hr at T

) 108~F c

Cooldown rates shall be limited to a maximum of:

20oF/hr at T

< 89'F c

30oF/hr at 89 F < T

< 132 F

c 50oF/hr, at 132~F

< T

< 158~F c

75~ F/hr, at 158~F

< T

< 174~F c

100~ F/hr, at T ) 174~F c

5"17

6.0

SUMMARY

OF PROPOSED CHANCES In order to implement the subject LTOP system for operation between 4

and 32 EFPY, the following is required:

1.

Modify the Technical Specifications indicated in Subsection 2.4, based on the results of this evaluation.

2.

Modify the appropriate plant operating procedures to incorporate the changes in the Technical Specifications.

6-1

7.0 CONCLUSION

The subject LTOP system is designed in accordance with the requirements set forth in the NRC Branch Technical Position RSB 5-2 (Reference 3).

The system is adequate for preventing violations of the Appendix G P/T limits during any operating period between 4 and 32 EFPY.

Implementation of the system will not result in a reduction in the margin of safety presently afforded by the Technical Specifications while providing an adequate operating window.

7"1

8.0 REFERENCES

0 1.

NRC letter to FPL (E. G. Tourigny to C. O. Woody), Docket No.

50-389, dated October 16, 1986.

2.

St.

Lucie Unit 2 Technical Specifications.

3.

Branch Technical Position RSB 5-2, Overpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures.

St.

Lucie Unit 2 Final Safety Analysis Report.

5.

L-78-129, FPL letter to NRC (R.

E. Uhrig to V. Stello), Docket No. 50-335, dated April 13, 1978.

6.

L-86-353, FPL letter to NRC (C. O. Woody to E. C. Tourigny),

Docket No. 50-389, dated September 13, 1986.

7.

St.

Lucie Unit 2 P/T Limits and LTOP, 4 EFPY to 32 EFPY, Proposed Technical Specification Changes.

(Attached) 8.

Code of Federal Regulations Part 50 Appendix G, Fracture Toughness Requirements, May 1983.

9.

ASME Boiler and Pressure Vessel Code Section III, Appendix G, Protection Against Nonductile Failure.

10.

Babcock and Wilcox Report BAW-1880, Analysis of Capsule W-83, Florida Power arid Light Company, St.

Lucie Plant Unit No. 2, Reactor Vessel Material Surveillance

Program, September 1985.

11.

ASME Boiler and Pressure Vessel Code Section III, Article G-2222.

12.

ASME Boiler and Pressure Vessel Code Section Ill, Article NB-2332.

,'13.

Methodology for Adjusted Reference Temperature Calculations for St.

Lucie Units 1 and 2

8"1

ATTACHMENT5 METHODOLOGY FOR ADJUSTED REFERENCE TEMPERATURE CALCULATIONS EJW I /033/7

0

TABLE OF CONTENTS TXTL PAGE NO.

INTRODUCTION BACKGROUND CALCULATION OF ADJUSTED REFERENCE TEMPERATURES (ART) 6 C-E METHODOLOGY POSZTXON ON ADJUSTED REFERENCE TEMPERATURES I

ST.

LUCXE UNIT 2 ART VALUES ST.

LUCIE UNIT 1 ARTVALUES'0 REFERENCES 13

LIS OF TABLES ADJUSTED REFERENCE TEMPERATURES FOR ST.

LUCIE UNIT 2 RELEVANT SURVEILLANCE DATA FOR ST.

LUCIE UNIT 2 ADJUSTED REFERENCE TEMPERATURES FOR STi LUCIE UNXT 1 RELEVANT SURVEXLLANCE DATA FOR ST LUCIE UNIT 1

LIST OF F CURES

~PAG NO; COMPARISON OF REVISION 2 PREDICTIONS TO SURVEILLANCE CAPSULE RESULTS 19 COMPARISON OF ADJUSTED REFERENCE TEMPERATURES AT END-OF-LIFE AS A FUNCTION OF THICKNESS FOR ST.

LUCXE UNIT 2 20 RTNDT SHIFTS FROM SURVEILLANCE DATA COMPARED TO PREDICTIONS FOR ST.

LUCIE UNXT 2 2I COMPARISON OF ADJUSTED REFERENCE TEMPERATURES AT END OF-LIFE AS A FUNCTION OF THICKNESS FOR ST.

LUCZE UNIT 1 22 RTNDT SHIFTS FROM SURVEILLANCE DATA COMPARED TO PREDICTIONS FOR ST.

LUCIE UNIT 1 23

METHODOLOGY FOR ADZUSTED REFERENCE TEMPERATURE CALCULATIONS ODUC 0

The NRC review of the St. Lucie Unit 2 P/T limits for 5 EFPY, resulted in questions from the NRC concerning the values of adjusted reference temperatures (ART) used in generating these P/T limits.

The NRC questioned the conservatism of these ART values since the calculated temperatures were low'er than values for ART determined by the method detailed.in the current draft version of Regulatory Guide 1.99, Rev.

2.

The ART values cited to be of most concern are those that deviate by 54 or more on the non-conservative side.

The purpose of this report is twofold.

First, to describe the methodology that was utilized in calculating ART values for use in generating P/T limits out to 32 EFPY for both St. Lucie Unit 1 and Unit 2.

Second, to demonstrate that these ART values are in fact sufficiently conservative relative to the temperatures calculated according to the guidelines of draft Regulatory Guide 1.99, Rev.

2 when other allowable features of the methodology are exercised.

The report provided herein is in support of the subject FP&L's request to amend Facility Ope'rating License No. NPF-16 for the St. Lucie Plant, Unit No. 2, and in support of a similar request for Unit No.

1 that is intended to be filed later.

SUMPfAR Combustion Engineering (C-E) employed the guidelines of draft Regulatory Guide 1.99, Rev.

2 in calculating adjusted reference temperatures (ART's) for use in generating P/T limits for St.

Lucie Unit 1 and Unit 2.

The only exception to the Regulatory Guide procedures in CE's approach was the use of calculated fluence values at 1/4T and 3/4T locations rather than the draft Regulatory Guide 1.99, Rev.

2 correlation based on surface fluence.

The net result was adjusted reference temperature values at the 3/4T location that were non-conservative with respect to ART values determined by the Regulatory Guide procedures (i.e. more than 54 difference).

Combustion Engineering's (C-E) position on the adjusted reference temperatures used in generating P/T limits for the St; Lucie':

Units is that the values calculated for both Unit 1 and Unit 2 are sufficiently conservative with respect to the guidelines of draft Reg.

Guide 1.99, Rev.

2.

This position is based on the comparison of ART values provided for St. Lucie Units 1

& 2 with ART values determined within the guidelines of the draft Regulatory Guide and considering the current position of the NRC with respect to establishing the credibility of surveillance data.

The following provides the basis for CE's position on the adjusted reference temperatures:

St.

ucie U t

- Controlling plate material and surveillance material are very similar.

- Surveillance data shows plate behaves as predicted.

- Surveillance data from other C-E plants show behavior for C-E design and equivalent material equal to or better than draft Reg.

Guide 1.99, Rev.

2 predictions.

- Surveillance data from non-CE plants with materials similar to Unit 2 plates also show predictable irradiation response.

St.

uc e

U 't

- Surveillance data show weld behaves conservatively compared to predictions.

- Surveillance data from other plants with similar surveillance weld metal also show predictable behavior.

- Surveillance weld from Beaver Valley is same as controlling weld for St Lucia 1.

Two Beaver Valley capsules show predictable shift behavior for weld metal.

- Ex-vessel dosimetry analysis corroborates spatial fluence distribution calculations.

Therefore, by demonstrating that credible surveillance information is available which reduces uncertainty in shift predictions, reducing the amount of margin applied in calculating adjusted reference temperatures is )ustified.

When the resultant ad)usted reference temperatures are compared to the ART values used in generating P/T limits, the P/T limit values for Unit 2 are conservative by a factor of 04 to 154.

A similar comparison for Unit 1 yields P/T limit values at 1/4T which are conservative by a factor of 64 to 114, and values at 3/4T which are non-conservative by a factor of 04 to 54.

These comparisons demonstrate that the P/T limit values were reasonably

based, including the 3/4T location values which were within 54 of the surveillance data adjusted Regulatory Guide ARTs.

BACKGROUND The adjusted reference temperature is defined as the temperature shift in the Charpy curve for the irradiated material relative to that for the unirradiated material measured at the 30 foot-pound energy level.

Margin is added to the adjusted reference temperature to obtain conservatism with the use of upper bound values.

This method for determining adjusted reference temperature (ART) is given by the following expression:

ART Xnitial RTNDT + 6 RTNDT + Margin The shift, or 4 RTNDT, is a function of the irradiation fluence and a chemistry factor which is dependent on the copper and nickel content of the material in cpxestion.

Values for the margin applied to the ART were determined from the materials property data base of radiation damage used in developing the guidelines of draft Regulatory Guide 1.99, Rev.

2.

The margin is actually based on a combination of a number of variables which influence the nature and extent of radiation damage in materials.

Among the factors which contribute to the scatter in the data base, and hence the margin, are the uncertainties in the irradiation temperature and fluence, flux, neutron energy spectra and the material composition.

Surveillance capsules from operating reactors are used to provide an accurate assessment of the plant specific irradiation environment.

Results of irradiated capsule analyses yield a measured correlation of shift versus fluence as opposed to the RTNDT shifts predicted through use of the. draft Regulatory Guide 1.99, Rev.

2 guidelines.

Zn addition, the surveillance data results in a significant reduction in the RTNDT shift uncertainty, since the actual (plant specific) irradiation temperature, fluence, flux spectrum and vessel materials are involved.

An early draft of Regulatory Guide 1.99, Rev.

2 permitted the use of surveillance data for calculation of adjusted reference temperatures when two or more credible surveillance data sets became available from the reactor in question.

One of the requirements for credibility of the data was that the surveillance capsule material should be the controlling vessel material with regard to radiation damage.

If the tests for credibility of the surveillance data were satisfied the procedure for determining adjusted reference temperatures was defined as follows:

1) Calculate the chemistry factor by a least squares method to fit the shift to fluence relationship to the surveillance data.

2)

To calculate the margin, use one-half the values given for G~

3) If this procedure gives a higher value of adjusted reference temperature than the procedures of Position 1.1 of draft Regulatory Guide 1.99, Rev.

2, the surveillance data should be used.

If this procedure gives a lower value, either may be used.

Based on industry review and comments to the NRC on the first draft of Rev.

2 to Regulatory Guide 1.99 some modifications are being made concerning the applicability of surveillance data.

The current NRC position on use of surveillance data for adjusted reference temperature calculations, as related to the guidelines of draft Regulatory Guide 1.99, Rev.

2, is described as follows:

"For plants having surveillance data that are credible in all respects except the material does not represent the critical material in the vessel, the calculative procedures of Revision 2 should be used to obtain mean values of shift, 6 RTNDT, but in calculating the margin the value of ag may be reduce'd from the values given in the last paragraph of Section 1.1 by an amount to be decided on a case-by-case basis

~5~

depending on where the measured values fall relative to the mean calculated f'r the surveillance materials."

0 0

R C

ES 0DOLOG Combustion Engineering provided values for adjusted reference temperatures for use in generating pressure-temperature (P/T) operating limits for St. Lucie Units 1

& 2.

The following methodology was used to determine these adjusted reference temperatures:

1 - The controlling vessel material was identified based on. the initial RTNDT's and chemistry factors of the pl'ate and weld materials'.

2 >> Fluence values at 1/4T and 3/4T were calculated for neutron energy levels, E > 1 MeV; the spectrum adjustment factor recommended by draft Reg.

Guide 1.99, Rev.

2 was not applied.

3 - RTNDT shifts were calculated using the draft Reg.

Guide equation and tabulated chemistry factors for plate and weld.

4 - The adjusted reference temperature values were determined by the sum of the initial RTNDT, shift and margin as defined by the equation and values for margin in the draft Reg.

Guide.

Therefore, the only difference between this methodology and that of the draft Regulatory Guide is the application of the spectrum adjustment factor in the fluence calculations.

This leads to the differences between the C-E methodology and the procedure in the draft Regulatory Guide in the calculated ART values, particularly at the 3/4T location.

OS ON ON AMUS NC Combustion Engineering's position on the adjusted reference temperatures used in generating P/T limits for the St. Lucie Units is that the values calculated for both Unit 1 and Unit 2 are sufficiently conservative with respect to the guidelines of draft. Regulatory Guide 1.99, Rev.

2.

This position is based on the comparison of ART values provided for St. Lucie Units 1

2 with ART values determined within the guidelines of the Regulatory Guide and within the current position of the NRC with respect to the use of surveillance data.

The procedures defined in draft Regulatory Guide 1.99, Rev.

2 include built in conservatisms in the calculated RTNDT shift.

This observation is depicted graphically in Figure 1 as

a. plot of..-

measured shifts from surveillance data versus predicted shifts for the same materials and fluences using the procedures of draft Regulatory Guide 1.99, Rev.

2.

The review of measured shift values for surveillance materials from C-E plants shows the generally conservative nature of the shifts predicted using Reg.

Guide 1.99, Rev. 2.

Since this comparison is restricted to C-E plants, the differences in material and neutron irradiation environment (i.e. fluence, flux and spectrum) are minimal and therefore enable the use of such data to improve predictive accuracy in C-E plants.

Reduction in the margin applied to ad)usted reference temperature calculations is )ustified when relevant surveillance data is available to reduce the overall uncertainty involved in the reference temperature shift predictions.

Surveillance data from the reactor vessel in question combined with additional data from comparable plants provide a basis for accurate determination of adjusted reference temperatures.

It is also reasonable to use surveillance data from comparable plants which include the controlling material for the reactor in question.

In addition, all relevant surveillance data sets for similar materials and irradiation environments should be utilized to take advantage of measured credible data in an effort 'to reduce the overall uncertainty of the calculations while retaining adequate conservatism of the calculations to assure the continued safe operation of the plant.

Adjusted reference temperature values for the controlling material in St. Lucie Unit 2 are given in Table 1.

The table includes the ART values used. in generating the P/T limits and ART values calculated according to Procedure 1.1 of draft Regulatory Guide 1.99, Rev.

2.

Zn addition, ART.values are'hown where relevant surveillance data was incorporated into the ART calculation procedures to provide a reduction in uncertainty.

The ART values at end-of-life (EOL) as a function of depth are shown in Figure 2.

In Figure 2, the Adjusted Reference Temperatures used in generating the P/T limits for St. Lucie Unit 2 are shown to be conservative, at all locations, when credible surveillance data is utilized to reduce the uncertainty of the RTNDT shift prediction.

St. Lucia Unit 2 has one set of irradiated surveillance capsule data.

The analysis of the surveillance capsule revealed a shift of 35 F for the longitudinal Charpy curve and 21 F for the transverse Charpy curve.

The measured fluence was 1.6 x 10 18 n/cm 2.

The calculated chemistry factor of 67 for the capsule

material, based on the more conservative shift of the longitudinal base metal, compares favorably with the chemistry factor tabulated in the draft Regulatory Guide of 74 for plate with 0.114 copper and 0.614 nickel.

Surveillance data from other plants with similar plate material also compare favorably in terms of measured versus predicted shift as shown in Table 2.

Zn general, the calculated chemistry, factors for plate material, within this range of copper and nickel content, are in good agreement with the mean values predicted by the draft Regulatory Guide procedures.

The controlling material and surveillance material for St. Lucie Unit 2 compare as follows in copper and nickel content and initial RTNDT:

Controlling Plate Surveillance Plate Copper 0 ~ 134 0.114 Initial Chemistry Nickel RTNDT Factor 0'24'10 F

92 0.614

+30 F

74 Note that the method'in Regulatory Guide 1.99 for determining which vessel material is controlling has changed since the surveillance plate material was originally selected.

The properties of the controlling material were therefore used in calculating the adjusted reference temperatures.

Figure 3 shows the shift versus fluence data from Table 2 plotted against the predicted shift for the controlling vessel material, Plate M-605-2.

The adjusted reference temperature, including the margin is also plotted in Figure 3.

As shown in Figure 3, the surveillance measured shifts correlate very well with the predicted shifts.

The other curves show the effect of adding margin to the sum of initial RTNDT + shift.

Figure 3

demonstrates graphically that the margin applied to the adjusted reference temperature can be reduced by using a factor of Gn/2 as a result of greater confidence in shift prediction by comparison with surveillance data.

As shown in Figure 3, the reduced level of margin still retains a sufficient amount of conservatism in the final adjusted reference temperature values.

Based on the availability of comparable surveillance data sets shown in Table 2 along with the set of data specific to St. Lucie Unit 2, there is a demonstration of the predictability of the St.

Lucie Unit 2 beltline materials.

This reduction in uncertainty of the shift prediction justifies reducing the Margin that is added to determine the ad)usted reference temperature.

The recommended value of g> to be used for margin is one-half the value for plate given in Reg.

Guide 1.99, Rev.

2 (i.e. for plate Gz/2 ~ 8.5 F).

When this procedure is used, the ART values used in generating the P/T limits are shown to be reasonably conservative with respect to ART values determined using surveillance data to reduce. the margin.

(See Table 1 and Figure 2)

S UC Ad]usted reference temperature values for the controlling material in St. Lucie Unit 1 are given in Table 3.

As in.Table 1

the values from Procedure 1.1 of the draft Regulatory Guide and ART values calculated using surveillance data are compared with the ART values used for the St. Lucie Unit 1 pressure-temperature (P/T) operating curves.

The ART values at End-Of-Life (EOL) are shown in Figure 4 as a function of depth through the thickness.

The use of relevant surveillance data provides greater confidence in the calculated shift values, thereby reducing the amount of margin that should be applied to define the ART value.

St. Lucie Unit 1 is similar to Unit 2 in that one set of surveillance data is available and the controlling vessel material is not included in the surveillance capsule.

Analysis of the surveillance capsule revealed a shift of 74 degrees in the Charpy curve for the weld metal with a corresponding fluence of 5.5 x 10 18 n/cm 2.

The calculated chemistry factor of 89 for the surveillance weld metal is approximately 204 lower than the

tabulated chemistry factor in draft Regulatory Guide 1.99, Rev.

2 of 110 for weld metal containing 0.234 copper and 0.114 nickel.

Table 4 shows the relevant surveillance data for St, Lucie 1, including a surveillance weld from another CE designed plant with very similar weld chemistry, and two surveillance data sets for Beaver Valley Unit 1, where the capsule weld material is the same as the controlling weld material for St. Lucie Unit 1.

In

general, the measured shifts and the calculated chemistry factors are in close agreement or conservative with respect to the draft Regulatory Guide predictions.

Figure 5 shows the shift versus fluence data from the Beaver Valley surveillance capsules in Table 4 plotted against the predicted shift for the controlling vessel material, Weld 3-203 A,B,C,.

The ad)usted reference temperature, inc'luding the marg'in is also plotted in Figure 5.

Figure 5 demonstrates graphically that the margin applied to the ad)usted reference temperature can be reduced bp using a factor of 0> /2 given greater confidence in shift prediction by comparison with surveillance data.

In addition to the in-vessel irradiated capsule for St. Lucie Unit 1, an evaluation of ex-vessel dosimetry capsules has provided a greater confidence in the calculated fluence values and confirmation of the axial and azimuthal vessel fluence profiles.(8)

Based on the availability of a number of relevant surveillance data sets, the reduction in uncertainty of the shift prediction justifies reducing the margin that is added to determine the adjusted reference temperature.

The recommended value of Gg to be used for the margin is one-half the value for weld metal given in draft Regulatory Guide 1.99, Rev.

2 (i.e. for weld d~/2

~

14 F).

When this procedure is used, the 1/4T ART values used in generating the P/T limits for St. Lucie Unit 1 are shown to be

t conservative by 6 to 114 (13 to 16 F) with respect to ART values

~

~

~

~

~ determined using surveillance data to reduce the margin.

(See Table 3 and Figure 4).

At 3/4T, the p/T limit values are non-conservative but within 54 of the ART value determined using surveillance data to reduce the margin; the difference ranged from 04'o 54 (0 to 7 F).

REFERENC S

1.

A.L. Lowe, L.L. Collins, L.A. Hassler and W.A. Pavinich, "Analysis of Capsule W-83 Florida Power and Light Company St. Lucie Plant Unit No. 2," BAW-1880, September 1985.

2.

J.S. Perrin, E.O.

Fromm, D.R. -Farmelo, R.S.

Denning and R.G. Jung, "Calvert Cliffs Unit No.

1 Nuclear Plant Reactor Pressure Vessel Surveillance Program:

Capsule 263," Battelle Memorial Institute Final Report, December 15, 1980.

3.

S.E.

Yanichko et al., "Analysis of Capsule U from the Commonwealth Edison Company Zion Nuclear Plant Unit 1 Reactor Vessel Radiation Surveillance Program,"

WCAP-9890, EPRI RP1021-3 Topical Report, March 1981.

4.

J.S. Perrin et al., "Zion Nuclear Plant. Reactor Vessel Radiation Surveillance Program: Unit No.

1 Capsule T, and Unit No.

2 Capsule U," final Report to Commonwealth Edison

Company, BCL-585-4, March 1978.

5.

J.S.

Perrin et al., "Surry Unit No.

1 Pressure Vessel Irradiated Capsule Program and Analysis o'f Capsule T,"

BCL -Final Report to Virginia Electric Power Company, June 1975.

6.

J.S. Perrin et al., "Surry Unit No.

2 Pressure Vessel Irradiated Capsule Program:

Examination and Analysis of capsule X," BCL Final Report to VEPCO, September 1975.

7.

S.T.

Byrne et al., "Florida Power and Light Company St.

Lucie Unit No.

1 Evaluation of Irradiated Capsule W-97,"

Combustion Engineering, TR-F-MCM-004, December 1983.

G.P.

Cavanaugh, "Florida Power and Light Company St. Lucie Unit No.

1 Evaluation of Cycle 6 In-and Ex-Vessel Neutron Flux Dosimetry Measurements,"

Combustion Engineering, CE-NPSD-358, May 1986.

9.

S.E.

Yanichko, et al ~

g Analysis of Capsule V from the Duquesne Light Company Beaver Valley Unit No.

1 Reactor Vessel Radiation Surveillance Program," Westinghouse Electric Corporation, WCAP-9860, January 1981.

10.

S.E.

Yanich3co et al., "Analysis of Capsule U from the Duquesne Light Company Beaver Valley Unit No.

1 Reactor Vessel Radiation Surveillance Pxogram," Westinghouse Electric Corporation, WCAP-10867.

TABLE 1 ADJUSTED REFERENCE TEMPERATURES FOR ST.

LUCIE UNIT 2 (CONTROLLING MATERIAL ~ PLATE M-605-2)

QO~CT~ON 1/4 T 1/4 T 1/4 T 1/4 T 1/4 T 1/4 T 1/4 T 1.99 REV.

2 5

119 10 137 15 147 20 154 25 159 30 163 32 164 ART VALUES USED ZN 117 135 146 152 157 161 163

+12

+12

+ll

+10

+10

+11 120 130 137 142 146 147 SURV.DATA APPLIED TO

+15 102 3/4 T 5

3/4 T 10 3/4 T 15 3/4 T 20 3/4 T 25 3/4 T 30 3/4 T 32 95 111 121 128 134 138 140

-12

-12

-12

-10

-10 10 84 98 107 114 120 124 126 78 94 104 114 117 121 123 NOTE:

THE 4

COLUMNS ARE THE PERCENTAGE DIFFERENCES

~ BETWEEN THE P/T LIMIT ART VALUES AND THE REG.

GUIDE 1.99 REV.

2 VALUES DETERMINED BY EITHER PROC.

1 ~ 1 OR WITH THE USE OF SURVEILLANCE DATA TO JUSTIFY REDUC1NG P~

BY A FACTOR OF 2.

TABLE 2 RELEVANT SURVEILLANCE DATA FOR ST.

LUCIE UNIT 2 PLA~

l. 99 REV.

2 CHEMISTRY C 0 SURVE LANC DAT CHEMISTRY ST.

LUCIE-2

( 1)

~ 11 CAL.CLIFFS-1 (2), 12 ZION-1 (3)

~ 11 ZION-1 (4)

F 11 ZION-2 (4)

~ 12 SURRY-1 (5)

~ll SURRY-.2 (6)

. 11

.61

.64

~ 49

.49

.53

.50

..54 74 84 73 73 82 73 73.

.16

.62

~ 29

~ 89

~ 20

.25

~ 32 35 60 54 92 51 52 49 68 69 82 95 89 83 7'1 (Fluence Values x10 19 n/cm 2)

• CALCULATED CHEMISTRY FACTOR

• FOR SURVEILLANCE DATA ~ 81

• . FROM LEAST SQUARES BEST FZT TABLE 3 ADJUSTED REFERENCE TEMPERATURES FOR ST.

LUCZE UNIT 1 (CONTROLLING MATERIAL ~ WELD 3-203 A)B,C)

LOCAT ON 1/4 T 1/4 T 1/4 T 1/4 T 1/4 T 1/4 T 15 20 25 30 32 195 213 227 236 240 1 ~ 99 REV.2 mx 10 1.7 ART VALUES USED FOR P

LX S

161 187 205 220 229 232 SURV.

DATA APPLIED TO 1.99 REV.2

+11 145 174 192 206 214 218.

3/4 T 3/4 T 3/4 T 3/4 T 3/4 T 3/4 T 10, 115 15 139 20, 156 25 169'0 178 32 182 19

-18

-18

-16 15

~ 93 114 128 141 151 155 93 118 135 148 157 161 NOTE:

THE 0

COLUMNS ARE THE PERCENTAGE DIFFERENCES BETWEEN THE P/T LIMIT ART VALUES AND THE REG.

GUIDE 1 ~ 99 REV.

2 VALUES DETERMINED BY EITHER PROC. 1.1 OR WITH THE USE OF SURVEILLANCE DATA TO JUSTIFY REDUCING 0~

BY A FACTOR OF 2.

17-

TABLE 4 RELEVANT SURVEILLANCE DATA FOR ST.

LUCZE UNIT 1 REV.

2 KKMISTRY S

VE LANC DAT CHEMISTRY

~FLU l~C SH:gg FACTOR

('F)

ST.

LUCXE-1 (7)

. 23

. 11 110

~ 55 CAL.CLIFFS 1 (2)

~ 24

.18 119

.62

• CALCULATED CHEMISTRY FACTOR 78'*

• FROM LZAST SQUARES BEST PZT OP

• ST.

LUCXE-1 AND CALVERT. CLIFFS 1 '*

74 59 89 BEAVER VAL. (9)

. 30

. 64 BEAVER VAL. (10)

~30,. 64 200 200

~ 26

.65 150 155 237 176 (Pluence Values x10 19 n/cm 2)

• • • 4********************************

• CALCULATED CHEMISTRY FACTOR ~ 197*

• PROM LEAST SQUARES BEST FIT OF

• BEAVER VALLEY DATA SETS NOTE:

THE BEAVER VALLEY SURVEILLANCE WELD MATERIAL ZS THE SAME WELD AS THE CONTROLLING WELD ZN ST.

LUCZE UNIT 1.

(HEAT NO ~ 305424'PEC

~ B-4 MOD'LUX LINDE 1092)

<<18>>

110 100 90 0

0 0

80 70 50 Vl 0

10 20 40 60 M~~V 2 tft (g

+

Sme Metal 80 100 Figure 1

Comparison of Revision 2 Predictions to Surveillance Capsule Results

180 St. Lucie Unit 2 Plate M-605-2 Rev. 2, Proc.

1.1 160

'ART ('-F) 140 120

'Oi Rev.

2 w/surv.

data P/T Limit Values oi 100 Surface 1/4T 1/2T Thickness 3/4T Figure 2.

Comparison of Adjusted Reference Temperatures (ART) at End-of-Life (E01

)

as a Function of Thickness for St. Lucie Unit 2.

150

('F)

St. Lucie Unit 2 Plate H-605-2 g = 17 8 Surveillance Data (Table 2)

~ Initial RTNDT + Shift + 2 0<

+0+

2 2

~initial RTNDT + Shift + 2 0<

+ (0>/2) 2

~Initial RTNDT + Shift

~RTNDT Shift 100 50 po Oo Initial RTNDT = 10'F 1018 1019 Fluence 1020 Figure 3.

RTNDT Shifts from Surveillance Data Compared to Predictions for St. Lucie Unit 2.

300 St. Lucie Unit 1

Weld 3-203, A, 8, C

250 AR7F) 200

'0 Rev.

'0 Rev. 2'w/surv.~ ~

data 2, Proc.

1.1

'a 150 P/T Limit Values Surf ace 1/4T 1/2T Thickness 3/4T

~

~

Figure 4.

Comparison of Adjusted Reference Temperatures (ART) at End-of-Lif (FQL) as a Function of T.",ickness for St. Lucie Unit 22~

200 GZ =17 O~= 28 300 St. Lucie Unit 1

Weld 3-203 A, B, C

~Initial RTNDT + Shift + 2 0

+ 0 I

~RTNOT Shift Initial RTNDT + Shift + 2 0

+ (0 /2)

I

~ Initial RTNDT + Shift (F) 10

(~)

RTNDT Shift (Beaver Valley)

RTNDT Shi ft + Initial RTNDT (Beaver Val 1 ey)

Initial RTNDT = -56'F 1018 1019 10".

Fluence figure 5.

RTNDT Shifts from Beaver Valley Surveillance Data Compared to Predictions for St. Lucie Unit 1.

r, I

C