ML17216A734

Forwards Proposed Tech Specs,Incorporating Revised pressure-temp Overpressure Protection Limits for RCS Sys & Addressing NRC Concerns Re 860715 Submittal.Related Info Encl

ML17216A734
Person / Time
Site:	Saint Lucie
Issue date:	10/10/1986
From:	Woody C FLORIDA POWER & LIGHT CO.
To:	Thadani A Office of Nuclear Reactor Regulation
References
L-86-417, NUDOCS 8610200399
Download: ML17216A734 (40)

Text

RESUL>)Y INFORNATION DISTRIBUTI BYBTEI'1 <R IDB)

I I ACCESSION NBR 8610200399 DOC. DATE: 86/10/10 NOTARIZED: NO DOC}(ET FAC IL: 50-389 St. Lucie Plant> Unit 2> Florida Poeer Zc Light Co.

AUTH. NANE AUTHOR AFFILIATION WOODY> C. O. Florida Paver 8c Light Co.

RECIP. MANE RECIPIENT AFFILIATION 05000385'UBJECT:

THADANI,A. C. PWR Prospect Directorate 8 Forwards proposed Tech Specs> incorporating revised pressure temp ovel pressure protection 1 imits For RCS sos 5 addressing NRC concerns re 860715 submittal. Related into encl DISTRIBUTION CODE: *001D DENOTES:

TITLE: OR COFIEB RECEIVED: LTR Submittal: Qeneral Distribution ENCL / SIZE:

RECIPIENT COPIES RECIPIENT COP IES ID CODE/NANE LTTR ENCL ID CODE/NANE LTTR ENCL PWR-8 EB PWR-8 PEI CSB 2 2 PWR-8 FOB PWR-8 PD8 LA 0 PWR-8 PD8 PD 01 5 5 TOUR IQNY> E 1 PWR-8 PEICSB PWR-8 RSB 1 INTERNAL: ADN/LFNB 1 0 ELD/HDS2 1 0 SCB -. 1 NRR/ORAS 1 0 04 1 RQN2 1 EXTERNAL: EQhQ BRUSQUE> S 1 1 LPDR 03 1 1 NRC PDR 02 1 NSIC 05 1 TOTAL NUNBER OF COPIES REQUIRED: LTTR 23 ENCL 19

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FLORIDA POWER & LIGHT COMPANY OCTOBER J,P Ig86 L 41'7 Office of Nuclear Reactor Regulation Attention: Mr. Ashok C. Thadani, Director PWR Project Directorate /38 Division of PWR Licensing-B U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Dear Mr. Thadani:

Re: St. Lucie Unit 2 Docket No. 50-389 Low Tem erature Over ressure Protection (LTOP)

By letter L-86-28I, dated July l5, l986, Florida Power Bc Light Company (FPL) submitted proposed Technical Specifications to incorporate revised pressure-temperature and low temperature overpressure protection limits (LTOP) for the reactor coolant system. As a result of discussions between the NRC staff and FPL subsequent to our July l5, l 986 submittal, the proposed Technical Specifications have been revised to address NRC staff comments. The revised Technical Specifications are included in Attachment I.

Additionally, the staff requested clarification of certain design features of the shutdown cooling system relief valves, V-3666 and V-3667, used for LTOP. As discussed in Section 5.4.7.2.3 of the St. Lucie Unit 2 Updated Final Safety Analysis Report, p.5.4-24a, valves V-3666 and V-3667 each have a capacity of 2300 gallons per minute at a set pressure of 335 psig.

The staff also requested FPL to perform confirmatory supporting analyses to show that for specific LTOP alignments transient peak pressures in the reactor coolant system are within the pressure/temperature limits proposed in FPL's July l5, l986 submittal. The LTOP alignments for which confirmatory analysis will be provided are summarized in Attachment 2.

Please contact us if you have any questions about this submittal.

Very truly yours, C . Woody G oup Vice President Nuclear Energy COW/E JW/gp A t tachments cc: Dr. J. Nelson Grace, Region II, USNRC Mr. Alan Schubert, Florida Dept. of Health and Rehabilitative Services Harold F. Reis, Esquire agi0200~99 05o pp+99 PDR ADOCK PDR P PEOPLE... SERVING PEOPLE E JW I /0 I 7/ I

ATTACHMENT I Marked Up Technical Specification Pages I-4 3/4 4-3 3/4 4-5 3/4 4-29 3/4 4-30 3/4 4-3I 3/4 4-32 3/4 4-33 New Figure 3.4-2 New Figure 3.4-3 3/4 4-36 (with insert) 3/4 5-7 83/4 4-l (with insert) 4-8 '3/4 83/4 4-I I 83/4 5-2 E JW I /OI 7/2

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DEFINITIONS 1.16 'LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE ze'f GE is that operating gg5'p The LOW TEMPERATURE RCS OVERPRESSURE PROTECTIVE condition when (1) the cold leg temperature is during cooldown and

.BSBB during heatup and (2) the Reactor Coolant System has 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 opening is greater than 3.58 square inches.

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 CALCUL'ATION 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

l. 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 MODE - MODE 1.20 An OPERATIONAL MODE (i.e. MODE) shall correspond to any one inclusive combination of core reactivity 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 related instrumentation and (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

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REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 At least two of the loop(s)/train(s) listed below 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 2B and its associated steam generator and at least one associated Reactor Coolant pump,""

c. Shutdown Cooling Train 2A, d.

APPLICABILITY:

ACTION:

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Shutdown Cooling Train 2B.

aO 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 COLD 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 loop to operation.

All Reactor Coolant 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 'f5 ~

dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10'F below saturation temperature.

Reactor Coolant pump shall not be started with one or more of e Reactor Co t System cold leg temperatures less than or equal to~@& during heatup 4@Hi during cooldown unless the secondary water temperature of each steam generator is less than above each of'he Reactor Coolant System cold leg temperatures.

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REACTOR COOLANT SYSTEM COLD 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 10K indicated narrow range level.

APPLICABILITY: MODE 5 with Reactor Coolant loops filled ACTION:

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 level as soon as possible.

b. With no 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 shutdown cooling loop to operation.

SURVEILLANCE'E 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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4. 1.4. l. 2 At least one shutdown coolinf loop shall be determined to 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 10'F 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 cooling loop is OPERABLE and in operation. ZW W 0~

A ctor Coolant pump shall.not be started with one or mor of the Reactor C olan S stem cold leg temperatures less than or equal to during heatup o dS~ during coo)down unless the secondary water temperature of each steam generator is less than above each of the Reactor Coolant System cold leg temperatures.

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REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9. 1 The Re ctor Coolant System (except the pressurizer) temperature and pressure shall e limited in accordance with the limit lines shown on Figures 3.4-2+ 3.4-+am~~ during heatup, cooldown, criticality, and inservice leak and hydrostatic testing,oeBhka, APPLICABILITY: At al 1 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 operations or be in at least HOT STANDBY 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..

avg than 2004F and 500 psia, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS 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.

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REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS (Continued) 4.4.9.1.2 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 thyrse examinations shall be used to update Figures 3.4-2j~3.4-g.ee~~

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REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall

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In the event either the PORVsgor the RCS vent(s) are used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specifica-tion 6. 9. 2 within 30 days. The report shall describe the ~r g<rCJS circumstances initiating the transient, the effect of the PORVs+or vent(s) on the transient and any corrective action necessary to prevent recurrence.

'4)L - 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 the valve through one complete cycle of full travel at least once per 18 months.

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EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - T LESS THAN 325 F av LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE high-pressure safety injection pump, and

b. ,

An OPERABLE flow path capable of taking suction from the refueling water tank on a Safety Injection Actuation Signal and automatically transferring suction to the containment sump on a Sump Recirculation Actuation. Signal.

APPLICABILITY: MODES 3" and P.

ACTION:

'a 0 With n'o ECCS subsystem OPERABLE, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SHUTDOWN within the next 20 hours'n

b. the event'he ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to'date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

SURVEILLANCE RE UIREMENTS 4.5.3 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.

With pressurizer pressure less than 1750 psia.

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'/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to oper ate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.20 during all normal oper'ations and anticipated transients. In MODES 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 hour.

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

In MODE 4, and in MODE 5 with reactor 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 MODE 5 with reactor coolant loops not filled, a single shutdown, C

cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the .steam generators as a heat removing component, require 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 recognition The restrictions on start g a reactor coolant pump i MODES 4 and 5, with one or more RCS cold le les's,than or equal to during cooldown and Z~F ~3-F during heatup are provided to prevent RCS pressure transi he limits of energy additions from the secondary syste~

Appendix G to 10 CFR Part 50. The RCS wil'l be protected against overpressure transients by (1) sizing each PORV to mitigate the pressure transient of an inadvertent safety injection actuation in a water-solid RCS with pressurizer heaters energized,~d-(2) restricting startin'g of the RCPs to when the secondary water temperature of each steam generator is less than 496% above each of the RCS cold leg temperatures~

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'3/4.4.2 SAFETY VALVES 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. overpres-The relief capacity of a single. safety valve is adequate to relieve any 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 overpressur e relief capability and will prevent RCS overpressurization.

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

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REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS 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 cy'cles 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 'oper ation.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal induced compressive stresses tend to al'leviate the tensile stresses induced by the internal pressure. Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents'a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.

The heatup analysis also covers the determination of pressure-temperature limitations for the cas'e in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. Consequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

The heatup and cooldown limit curves Figures 3.4-Z~ 3.4-3 and=~~ are composite curves which were prepared by determining the most onservative case, The reactor vessel materials have been tested to determine their initial RTNDT the results of these test are shown in Table B 3/4.4-. 1. Reactor opera" tion and resultant fast neutron (E greater than 1 MeV) irradiation will cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based" upon the fluence and copper~content of the material in question, can be redi ure,B 3/4.4-1 and the recommendations of Regulatory z(~s) Guide 1.99, Revision A, "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials~" The heatup and cooldown limit curves Figures 3.4- 3.4-3y wad=~ inc de predicted adjustments for this shift in RTN at the e of the applicablC service period, as well as adjustments for possible errors in the pressure an) temperature sensing instruments.

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REACTOR COOLANT SYSTEM

'BASES PRESSURE/TEMPERATURE LIMITS Continued The actual shift in RTNDT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73 and 10 CFR Appendix H, reactor vessel material irradiation surveil-lance specimens installed 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 applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the delta RTNDT determined from the surveillance capsule is different from the calculated equivalent capsule radiation exposure. delta RTNDT for the The pressure-temperature limit lines

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~kaL 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 RTNDT for all Reactor Coolant System pressure-retaining mate" s with the exception of the reactor pressure vessel, has been determined

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of 1972) of Section III of the ASME Bo(Nr 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 20K 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 ASME Code rqquirements.

w~'- OpcZVi The OPERABILITY of two PORVs~Por an RCS vent opening of greater than I 3.58 square inches ensures that the RCS will be protected from pressure KbW transients which could exceed the limits of Appendix 8 to 10 CFR P en o or of the RCS cold legs are less than or equal to gfmpq'uring capability to protect the RCS fr overpressurization when the transient is limited to either (1) a safety i ction actuation in a water solid RCS with the pressurizer heaters energize r (2) the start of an idle RCP with the secondary water temperature of th steam generator less than or equal to&90'-4 above the RCS cold leg temperatur s with the pressurizer solid. 40'F HIp lou>'.fgssp6GWNzg e ~iog ST. LUCIE - UNIT 2 B 3/4 4"11

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EMERGENCY CORE COOLING SYSTEMS BASES ECCS.SUBSYSTEMS Continued With the RCS temperature below 3254F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

The trisodium phosphate dodecahydrate (TSP) stored in dissolving baskets located in the containment basement is provided to minimize the possibility of corrosion cracking of certain metal components during operation of the ECCS following a LOCA. The TSP provided this protection by dissolving in the sump ater and causing,its final pH to be raised to greater than or equal to 7.0.

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, although the analysis supports actuation of safety injection in a water solid RCS with pressurizer heaters energized, provide additional administrative assurance that a mass addition pressure transient can be relieved by the operation of a single PORVg oe ~/V.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensure that at a minimum, the assumptions used in the accident analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for throttle .valve position stops and flow balance testing pro-vide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration,'2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses. The requirement to dissolve a representative sample of TSP in a sample of RWT water provides assurance that the stored TSP will dissolve in borated water at the postulated post-LOCA temperatures.

3/4.5.4 REFUELING WATER TANK The OPERABILITY of the Refueling Water Tank (RWT) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWT minimum volume and boron concentration ensure that (1) sufficient water is available within containment to permit recirculation cooling flow to the core, and (2) the reactor will remain subcritical in the cold condition following mixing of the RWT and the RCS water volumes with all control rods inserted except for the

. most reactive control assembly. These assumptions are consistent with the LOCA analyses.

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LTOP SYSTEM ALIGNMENTS RCS Cold Leg ~

Required Supporting Analysis Temperature LTOP Alignment Regarding Mass Addition Transient Heat Up ~142oF SDCS in operation I) Assumptions to include; (T.S. 3.4.1.4.1) a) I SDCS RV at a lift setting Two SDCS RVs operable of 350 psia, and (T.S. 3.4.9.3.b) b) I HPSI pump plus 3 charging pumps in operation following a postulated SIS Heatup 142 F -295 F 142 F -200 F 2) Same as I) above SDCS in operation (T.S. 3.4.1.4.1)

PORVs operable (T.S. 3.4.9.3.a) 200 F -295 F 3) Same as I) above SDCS may be in (when SDCS in operation) operation (T.S. 3.4.1.3)

PORVs operable (T.S. 3.4.9.3.a) b)

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4) Assumptions to include a) I PORV at a lift setting I

470 psia HPSI pump plus 3 charging pumps in operation following a postulated SIS (when no SDCS in operation)

Cooldown SDCS in operation 5) Same as I) above (T.S. 3.4.1.4.1)

Two SDCS RV operable (T.S. 3.4.9.3.b)

Cooldown 161 F -286 F 161 F -.200 F 6) Same as I) above SDCS in operation (T.S. 3.4.1.4.1)

PORV operable (T.S. 3.4.9.3.a) 200 F -286 F 7) Same as I) above SDCS may be in (when SDCS in operation) operation 8) Same as 4) above (T.S. 3.4.1.3) (when no SDCS in operation)

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