Text

Proposed Tech Specs,Incorporating Revised Heatup & Cooldown Limit Curves Applicable Up to 20 EFPYs of Svc Life
	ML17345A422
Person / Time
Site:	Turkey Point
Issue date:	09/21/1988
From:	; FLORIDA POWER & LIGHT CO.
To:	;
Shared Package
ML17345A419	List: ML17345A418; ML17345A420; ML17345A422;
References
	NUDOCS 8809300066
	Download: ML17345A422 (19)
	v • d • e

ATTACHMENT 2 PROPOSED REVISED TECHNICAL SPECIFICATION PAGES TCG4/024/7

( SS09300066 SS0921 PDR ADOCK 05000250 P PDC

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. FACILITY OPERATING LICENSE NO. DPR-31 AMENDMENT NO. FACILITY OPERATING LICENSE NO. DPR-31 DOCKET NO. 50-250 AND 50-251 Revise Appendix A as follows:

Remove Pa es Insert Pa es v v vi Vi

3. 1.2 3. 1. 2, 3.1.2a

3. 1.2a 3~1 ~ 2b Fig. 3.1-la Fig. 3 1 1 a Fig. 3.1-lb Fig. 3. 1-lb Fig. 3.1-1c Fige 3.1-1c Fig. 3.1-ld Fig. 3.1-2c Fig. 3.1-2d B3.1-2 B3.1-2 B3.1-2a B3.1-2b B3.1-2c B3.1-2d B3.1-3 B3.1-3 TCGPTL.PLA

LIST OF TABLES Table Title Operational Modes 3.5-1 Instrument Operating Conditions for Reactor Trip 3.5-2 Engineering Safety Features Actuation 3.5-3 Instrument Operating Conditions for Isolation Functions 3.5-4 Engineered Safety Feature Set Points 3.5-5 Accident Monitoring Instrumentation 3.9-1 Radioactive Liquid Waste, Sampling and Analysis Program 3.9-2 Radioactive Liquid Effluent Monitoring Instrumentation 3.9-3 Radioactive Gaseous Waste Sampling and Analysis Program 3.9-4 Radioactive Gaseous Effluent Monitoring Instrumentation 3 ~ 13 1 Deleted 3.14-1 Fire Detection System 3.14-2 Fire Hose Station 3.16-1 Primary Coolant System Pressure Isolation Valves 3 ~ 17 1 Sent Fuel Burnup Requirements for Storage in Region Spent Fuel Pit II of the 3.18-1 Auxiliary Feedwater System Operability 4.1-1 Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels 4.1-2 Minimum Frequencies for Equipment and Sampling Tests 4.1-3 Minimum Frequency for Surveillance of Radioactive Liquid Effluent Monitoring Instrumentation 4.1-4 Minimum Frequency for Surveillance of Radioactive Gaseous Effluent Monitoring Instrumentation 4.2-1 Deleted 4.2-2 Minimum Number of Steam Generators to be Inspected During Inservice Inspection 4.2-3 Steam Generator Tube Inspection 4.8-1 Diesel Generator Test Schedule 4.8-2 Battery Surveillance Requirements

4. 12-1 Radiological Environmental Monitoring Program 4.12-2 Reporting Levels for Radioactivity Concentrations in Environmental Samples 4.12-3 Detection Capabilities for Environmental Sample Analysis 4.18-1 Minimum Frequencies for Safety Related Systems Flow Path Verification 6.2-1 Minimum Shift Crew Composition B3.1-1 Reactor Vessel Toughness Data, Turkey Point Unit 3 B3.1-2 Reactor Vessel Toughness Data, Turkey Point Unit 4 Amendment Nos. and TCGPTL.PLA

LIST OF FIGURES

~Fi ure Title

2. 1-1 Reactor Core Thermal and Hydraulic Safety Limits, Three Loop Operation 2.1-la Deleted 2.1-1b Deleted 2.1-2 Reactor Core Thermal and Hydraulic Safety Limits, Two Loop Operation 3 ~ 1 1 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED POWER with the Primary Coolant Specific Activity > 1.0 Ci/gram Dose Equivalent I-131 F 1-la Reactor Coolant System Heatup Limitations (60'F/hr)

Applicable for the First 20 EFPY /

3.l-lb Reactor Coolant System Heatup Limitations (100'F/hr)

Applicable for the First 20 EFPY 3.1-1c Reactor Coolant System Cooldown Pressure Limitations Applicable for the First 20 EFPY 3 ~ 1 2 Radiation Induced Increase in Transition Temperature for A302-B Steel l 3021 Control Group Insertion Limits for Unit 4, Three Loop Operation 3.2-1a Control Group Insertion Limits for Unit 4, Two Loop Operation 3.2-1b Control Group Insertion Limits for Unit 3, Three Loop Operation 3.2-1c Control Group Insertion Limits for Unit 3, Two Loop Operation 3 ~2 2 Required Shutdown Margin 3~2 3 K (z) vs Core Height 3~2 3a Deleted 3.2-4 Maximum Allowable Local KW/FT

4. 12-1 Sampling Locations

5. 1-1 FPL Turkey Point Site Area Map

6. 2-1 Deleted

6. 2-2 Deleted B3. 1-1 Effect of Fluence and Content on Shift of RTNDT for Reactor Vessel Steels Exposed to 550'F Temperature B3.1-2 Fast Neutron Fluence (E > 1MEV) as a Function of Effective Full Power Years B3.2-1 Target Band on Indicated Flux Difference as a Function of Operating Power Level B3.2-2 Permissible Operating Band on Indicated Flux Difference as a Function of Burnup (Typical) v3.

Amendment Nos. and TCGPTL.PLA

REACTOR COOLANT SYSTEM 3.1.2 PRESSURE TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.1.2a The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.1-1a, 3.1-1b and 3.1-1c for both Unit 3 and Unit 4 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup of 100'F in any 1-hour period,

b. A maximum cooldown of 100'F in any 1-hour period, and c ~ A maximum temperature change of less than or equal to 5'F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY: 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 evaluation 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 operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS Tavg and pressure to less than 200'F and 500 psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3 ~ 1~2 Amendment Nos. and TCGPTL.PLA

PRESSURIZER LIMITING CONDITION FOR OPERATION 3.1.2.b The pressurizer temperature shall be limited to:

a~ A maximum heatup of 100'F in any 1-hour period,

b. A maximum cooldown of 200 F in any 1-hour period, and

c. A maximum spray water temperature differential of 320'F.

ACTION:

With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

,3 ~ 1 ~ 2a Amendment Nos. and TCGPTL."PLA

QFRAT MP Rh 3,1.2,1 The maderator temperature coefficient (MTC) shall bee Less poeitlve than or equal to > 0 x 10 /Ic/~ for all rods withdrawn, beginning of the cycle life (SOL), hat zero THERMAL pOWER (Hgp) conditions! and b) Less positive than oc'qual to 9A x 10-> hk/k/~F from Hgp to 7096 RATED THERMAL POWER conditio~ and Less positive than or equal to f.0 x 10-f Sc/Ic/oP fram 70% RATED THERMAL POWER decreasing linearly to less positive than or equal ta 0 Ate/Ic/OP at 100% RATED THERIACAL POWER condition~ and d) Lel negative than -3.f x IM hk/k/~F for the all rods withdrawn, end af cycle lite (EOL>> RATED THERMAL POWER condition.

B ! Specification 3.1.2.lg b, and c MODES I and 2>> only>>>>.

Specification 3.1.2.ld - MODES I, 2, and 3 only a) With the MC more positive than the limits of Specifications 3.1.2.la, b, or c above, operation in MODES 1 and 2 may proceed pravidedc

1) Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive or equal to limits described in 3 1.2.1a, b, and c above <<Ithin 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months or be in HOT STANDBY <<Ithln the next C hours. These withdrawal limits shall be in addltian to the insertion limits of specification 3.2 I,

2) The cantrol rods are maintained within the ~lthdra~al limits established above until a subsequent calculation verifies that the

%fC has been restored to within its llmlt for the ail rods withdrawn candlti~ and

3) A Special Report Is prepared and submitted to the Commission pursuant to Specification 6.9.3, <<ithln 10 da~ describing the value ot the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the hflC to within its limit for the all rods withdrawn condition.

b) 'Wth the VLCC more negative than the limit of Specificatian 3.I.'2.Id above, be In HOT SHUTDOWN <<ithln 12 ho~

>> With Keff greater than ar equal to I.

>>>> The +bove limits may be suspended during the performance of LO%'OWER PHYSICS TESTS.

3. I-2b Amendment Nas.. and

MATERIAL PROPERTY BASIS CIRCUMFERENTIAL HELP< >

f11 CONTROLLING MATERIAL:

INITIAL RTNPT'OeF RTNP T AFTER 20 EFP Y '/4T 252 5 F 3/IT, 200.4eF CURVES APPLICABLE FOR HEATUP RATES UP TO 60'F/HR FOR THE SERVICE PFRIOP UP TP 20 EFPY. NO MARGINS ARE GIVEN FOR POSSIBLE INSTRUMENT ERRORS.

2500 Lect Test Lfrtt 2250 2000 1750 Uacceytaile Oyeatlm 1500 1250 5

o. 1000 Qceetehle O Oyentlm 750 IIILIIII1 Crt ttcel 1 tF I LWt SasN laservtce 11 oe.

500 atettc Teat Tmyeretere

($ 8D'F) fer the Service 250 I

~ ertee CRT te 0 50 '100 150 200 250 $ 00 $ 50 '400 450 500 IIIIChTCD TCW'CkATUIC (DCO.F)

Reactor Coolant System Heatup Limitations (60'F/Hr)

Applicable for the First 20 EFPY Figure 3.1-1a Amendment Nos. and

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2. BASES PRESSURE TEMPERATURE LIMITS All components in the Reactor Coolant System (RCS) are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are induced by normal load transients, reactor trips and startup and shutdown operations. During RCS heatup and cooldown, the temperature and pressure changes must be limited to be consistent with design assumptions and to satisfy stress limits for brittle fracture.

During 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.

During cooldown, the thermal gradients through the reactor vessel wall produce thermal stresses which are tensile at the reactor vessel inside surface and which are compressive at the reactor vessel outside surface. Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the inside surface location. Since the neutron irradiation damage is also greatest at the inside surface location, the inside surface flaw is the limiting location. Consequently, only the inside surface flaw must be evaluated for the cooldown analysis.

The fracture toughness properties of the ferritic material in the reactor vessel were determined in accordance with the NRC Standard Review Plan, ASTM E185-73 and in accordance with additional reactor vessel requirements.

The properties are then evaluated in accordance with Appendix G of the 1983 Edition of Section XIX of the ASME Boiler and Pressure Vessel Code and the additional requirements of 10CFR50, Appendix G and the calculation methods described in Westinghouse Report GTSD-A-1.12, "Procedure for Developing Heatup and Cooldown Curves".

B3. l. 2 Amendment Nos. and TCGPTL.PLA

The heatup and cooldown limit curves, Figures 3.l-la, 3.1-1b, and 3.1-1c are composite curves prepared by determining the most conservative case with either the inside or outside wall controlling, for any heatup rate up to 100 degrees F per hour and cooldown rates of up to 100 degrees F per hour. The heatup and cooldown curves were prepared based upon the most limiting value of predicted adjusted r'eference temperature at the end of the applicable service period (20 EFPY).

The reactor vessel materials have been tested to determine their initial RT ; the results of these tests as well as other material erties are shown in Tables B3.1-1 and B3.1-2. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RTN T. Therefore, an adjusted reference temperature, based upoR the fluence and chemistry factors of the limiting Reactor Vessel material has been predicted using Regulatory Guide 1.99, Revision 2, dated May 1988 (the latest accepted NRC methodology), "Radiation Embrittlement of Reactor Vessel Materials". The heatup and cooldown limit curves of Figures 3.1-1a, 3.1-1b, and 3.1-1c include predicted adjustments for this shift in RTNDT at the end of the applicable service period The actual shift in RT D of the vessel material will be established periodicallPduring operation by removing and evaluating, in accordance with ASTM E185-73 and 10CFR Appendix H, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. The surveillance specimen withdrawal schedule is shown in Specification 4.20.1. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel.

Since the limiting beltline material (Intermediate to Lower Shell Circumferential Weld) in Unit 3 and 4 is identical, the RV surveillance program was integrated and the results from capsule testing is applied to both Units. The surveillance capsule "T" results from Unit 3 (WCAP 8631) and Unit 4 (SWRI 02-4221) and the capsule "V" results from Unit 3 (SWRI 06-8576) were used with the methodology in Regulatory Guide 1.99 Revision 2 to provide limiting material properties information for generating the heatup and cooldown curves in Figures 3.1-1a, 3.1-lb, and 3.1-1c. The integrated surveillance program along with similar identical reactor vessel design and operating characteristics allows the same heatup and cooldown limit curves to be applicable at both Unit 3 and Unit 4.

B3.1-2a Amendment Nos. and TCGPTL.PLA

The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

B3.1-2b Amendment Nos. and TCGPTL.PLA

TAKE B 3 1-2 REACIOR VESSEL 'lOUGHNESS IRTA LUEKEY EGIN UNIT 4 50 ft lb/35 mila, Miniman

- lateral BtPansion E~ol'PPer Shelf Material Trans TYPe ( F) rang Trans ( F) ( ) rang Cl. Hd. Dane A302 Gr. B 0.008 -20 30 Cl. 2 0.72 0.010 4(a) 27(a) 199 129(a)

Cl. Hd. Flange A508 Ves. Sh. Flange A508 Cl. 2 0.68 0.010 1(a) -11(a) 176 114<<>

Cl. 0.08 0.71 0.009 6O(a) Hh 60 Inlet Nozzle A508 2 Inlet Nozzle A508 Cl. 2 0.84 0.019 6O(a) 60 Inlet Nozzle A508 Cl. 2" 0 75 0.008 16(a) 13(a) 16 162 105<<)

Outlet Nozzle A508 Cl. 2 0.78 0.010 7(a) -25(a) 165 1O7(a)

Outlet Nozzle A508 Cl. 2 0.68 0.010 38(a) 16(a) 38 160 -1O4(a)

Outlet Nozzle A508 Cl. 2 0.70 0.010 60(a) 42(a) 60 143 93(a)

Ugper Shell A508 Cl. 2 0.70 0.010 40 32(a) 40 1O1(a)

Inter. Shell A508 Cl. 2 0.054 0.69 0.010 90(a) 143 93<<)

Zanier Shell A508 Cl. 2 0.056 0.74 0.010 40 38(a) 40 147 97(a)

Trans. Ring A508 Cl. 2 0 69 0.011 60(a) 30(a) 60 Bot. Hd. Dane A302 Gr. B

'.010 10 30(a) 10 NA Inter. to Inkier SAW 0.26 0.60 0.011 lo(b) 63 1O( ) NA 63 Shell Girth Weld 0 0 140

~l (a) Estimated Values (b) Value Ba.~ on NUREG-0800, Branch Technical Position - MZEB 52

TABID B3. 1-1 REACIQR VESSEL 'LOUGHNESS DLTA

'LURK'.Y EGIN UNIT 3 50 ft lb/35 mils Minim Lateral Rl<pBnsion RFg~ Upper Shel f Cu

(~) ( F) ~ Trans ( F) ( ) lang Trans Cl. Hd. Dane A302 Gr. B 0. 010 36(a) > 70 > 45.5(

Cl. 85. Flange A508 Cl. 2 0.72 0.010 44(a) 31(a) 44 >118 > 76.5(

Ves. Sh. Flange A508 Cl. 2 0.65 0 010 -23(a) -41(a) -23 >120 > 78(a)

Inlet Nozzle A508 Cl. 2 0.76 0.019 60( ) 60 NA Inlet Nozzle A508 Cl. 2 0.74 0.019 6p(a) 60 Inlet Nozzle A508 Cl. 2 0 80 0.019 60(a) 60 NA Outlet Nozzle A508 Cl. 2 0 79 0.010 27(a) 9(a) 27 >110 > 71.5(a)

Outlet Nozzle A508 Cl. 2 0.72 0.010 7(a) -22(a) > 72(a)

A Outlet Nozzle A508 Cl. 2 0.72 0.010 42(a) 23(a) 42 >140 > 91(a)

Upper Shell A508 Cl. 2 0.68 0.010 44 (a) 50 >129 > 83.5(

Inter. Shell A508 Cl. 2 0. 058 0.70 0.010 40 25(a) 40 >122 > 79(a)

?ower Shell A508 Cl. 2 0.079 0. 67, 0.010 30 2(a) 30 163 '06 (a)

Trans. Rirg 'A508 Cl. 2 0.69 0.013 60(a) (a) 60 >109 > 7p 5(a)

Bot. Hd. Dane A302 Gr. B 0.010 -10 30 Inter. to Zamr SAW 0.26 0. 60 0.011 10( ) 63 10(b) 63 Shell Girth Weld p(a)- 0 168 (a) Estimated Values Based on NURHG-0800, Branch Vertical position - HKB 52

( ) Actual Value

QOOERh'TOR TEQPBRhTt5% COP3VICENT The limitations on moderator temperature coefficient (h4TC) are provided to ensue that the value of this coefficien re~ains within the limiting condition assumed in the FSAR accident and transient analyses The 4lTC values of this speciflcati n are appHcable to a specific set of plant conditions accordingly, verification of hffC values at conditions other than those explicitly stated will iequire exttapolatlm to those conditicns in order to permit an accurate co parison The most negative VfQ value to the most positive moderator density coefficient (H!DCl, wai obtained by incrementally correcting the VIDC used in the FShR analyses to noml ml operating condltiae. These corrections involved subtracting the incremental change in the %DC associated with a core condition of iLll rods inserted (most positive hIDC) to an all rods withdrawn condlfi(xl and 4 conversion for the rate of change ot moderator density with temperture at RR,TEO THERMAL, POWER conditims R3. l -) Amendment Nos. ~d

tv" ~ I C

ML17345A422

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