Text

Tech Spec Change Requests 42 & 47 Re Mod to ECCS
	ML19241B427
Person / Time
Site:	Crystal River
Issue date:	07/13/1979
From:	; FLORIDA POWER CORP.
To:	;
Shared Package
ML19241B420	List: ML19241B419; ML19241B427;
References
	TAC-45333, NUDOCS 7907170127
	Download: ML19241B427 (25)
	v • d • e

.

Technical Specification Chance Request No. 42 (Appendix A)

Replace pages 3/4 5-3, 5 and B 3/4 5-2 with the attached revised pages 3/4 5-3, 5 and B 3/4 5-2.

Proposeo Chance Change thc minimum HPI injection flow to >350 gpm for all combinations of 3 injection legs and >500 gpm for each HPI pump.

Reason for Proposed Change The ECCS Small Break Analysis has shown that with a certain size break in the Reactor Coolant System on the discharge of the Reactor Coolant Pump.

and a simultaDeous worst Case single failure, the present HPI system would nc' be adequate to i..itigate the accident.

The modification to be able to mitigate this (and other HPI required accidents) includes the permanent opening of the HPI discharge crossover valves. Under this condition, each HPI pump will inject flow through each HPI injection leg.

The minimum HPI flow necessary for com cooling, when HPI is required, is 350 gpm. The minimum total injection flow ensures that at least 350 gpm will be injected to the .eactor (through all combinations of three (3) injection legs) assuming the fourth leg is failed and has 30% of the flow from one pump.

Safety Analysis of Proposed Chance The purpose of the surveillance requirements is to provide assurance that proper ECCS flow will be maintained in the event of a LOCA. Mainte'ance 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.

Specification 4.5.2.c ensures that flow for each HPI pump will not exceed its nJnout conditions when the system is in its minimum resistance configuration since the stop check valves are on tha gump discharge and not in the HPI injection legs. The stop check valves am set against an RCS pressure of approximately 600 psig to obtain an HPI pump flow rate of greater than 500 gpm and less than the pump runout ficw rate.

The proper flow split between injection points is provided by requiring that at least 350 gpm of the minimum require 500 gpm pump ficw be injected through all combinations of three injection legs. This results in the maximum ficw rate through any one injection leg of no more than 30% of the minimum required pump flow of 500 gpm, i.e., any one injection leg will have a flow rate i 150 gpm when an HPI pump is only pumping 500 gpm.

TSCRN42 059 7907170/2 71 40 N5[

Cinally, an acceptable level of total ECCS flow to all injection points squal to or above that assumed in the ECCS-LOCA analysis is assured by requiring the flow rate for all combinations of three injection legs to be equal to or greater than 350 gpm. This assumes the worst case single failure (loss of power to an HPI train), the subsequent opening of the injection valves on the train that lost power, and the assumption that all the flow through the injection leg with the maximum flow is lost through the small pipe break. It has been shown that this 350 gpm is sufficient to mitigata the worst case small break accident without exceeding the Final Acceptance Criteria of a Peak Clad Temperatum of 2200 F. Therefore, this proposed change does not involve an unreviewed safety question.

TSCRN42 DS9

[t b ) 2b

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T3yg_>280 F LIMITING CONDITION FOR OPERATION 3.5.2 Two diverse ECCS subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE high pressure injection 'HPI) pump,

b. One OPERABLE low pressure injection (LPI) pump,

c. One OPERABLE decay heat ;ooler, and

d. An OPERABLE flow path capable of taking suction from the borated water storage tank (BWST) on a safety injection signal and manually transferring suction to the containment sump during the recircula-tion phase of operation.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. In the event the ECCS is actuated and injects water into the Reactor Cooiant 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.

CRYSTAL RIVER - UNIT 3 3/4 5-3

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

2. Verifying the correct position of each mechanical position stop for each of the stop check valves listed in Specification 4.5.2.c.

3. Verifying that the flow switches for ti.a throttle valves listed in Specification 4.5.2.d operate properly.

4. A visual inspection of the containment emergency sump which veri-fies that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion.

5. Verifying a total leak rate 16 gallons per hour for the LPI sys-tem at:

a) Normal operating pressure or a hydrostatic *,est pressure of

> 150 psig for those parts of the system downstream of the pump suction isolation valve, and b) 2. 55 psig for the piping from the containment emergency sump isolation valve to the pump suction isolation valve.

f. At least once per 18 months, during shutdown, by

1. Verifying that each automatic valve in the flow path actuates to its correct position on a high pressure or low pressure safety injection test signal, as appropriate.

2. Verifying that each HPI and LPI pump test starts automatically upon receipt of a high pressure or low pressure safety injection test signal, as appropriate.

g. Prior to entering MODE 3 with HPI or LPI system modifications that l could have altered system flow characteristics, by performance of a flow balance test during shutdown to confinn the following injection flow rates:

HPI System - Sincle Pump LPI system - Single Pump Single pump flow rate >500 spm 1. Injection Leg A into the Reactor at 600 psig Coolant System - 2800 to 3100 gpm Flow rate for all combinations 2. Injection Leg B into the Reactor n' 3 Injection Legs into the Coolant System - 2800 to 3100 gpm Recctor Coolant System 2.350 gpm at 600 psig CRYSTAL RIVER - UNIT 3 3/4 5-5

.oo ; ?78

l EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two diverse ECCS subsystems with RCS average temperature

> 280 F ensures that sufficient emergency core cooling capability will be av_llable in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Either subsystem operating in conjunction with the core flooding tanks is capable of supplying sufficient core cooling to maintain the peak cladding temperatures within acceptable limits for all pos-tulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long term core cooling cap 1bility in the recirculation mode during the accident recovery period.

With the RCS temperature below 280 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condi-tion of the reactor and the limited core cooling requirements.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures, that, at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. The decay heat remcyal system leak rate surveillance requi rements assure that the leakage rates assumed for .he sys-tem during the recirculation phase of the low pressure injection will not be exceeded.

The purpose of these surveillance requirements is to provide assurance that prop-er 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 .unout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split betweer 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.

3/4.5.4 BORATED WATER STORAGE TANK The OPERABILIlY of the borated water storage tank (BWST) as pact 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 BWST minimum volume and boron concentration ensure that 1) sufficient water is available within containment to pe rmit recirculation cooling flow to the core, and 2) the reactor will remain subcritical in cold condition following mixing of the BWST and the RCS water vol-umes with all control rods inserted except for the most reactive control assem-bly. These assumptions are consistent with the LOCA analyses.

CRYSTAL RIVER - UNIT 3 B 3/4 5-2 9

l} b l]

TECHNICAL SPECIFICATION CHANGE. REQUEST NO. 47 (APPENDIX A)

De'c te and insert, as indicated, the following pages in Appendix A of Operating License DPR-72:

Delete Insert V V IX IX 2-6 2-6 3/4 3-2 3/4 3-2 3/4 3-2a 3/4 3-3 3/43-3 3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-7a 3/4 3-13 3/4 3-13 3/4 3-14 3/4 3-14 3/4 3-16 3/4 3-16 3/4 3-17 3/4 3-17 3/4 3-17a 3/4 3-20 3/4 3-2" 3/4 4-33 3/4 7-5 3/4 7-5 B 3/4 4-13 B 3/4 4-13 Proposed Changes The proposed changes are:

1) Addition of flow rate indicatio? for the emergency feedwater system;

2) Addition of the auto start circuitry for the emergency feedwater pumps;

3) Addition of the anticipatory reactor trips; and

4) Change in setpoint for electromatic re:ief valve actuation.

Reasons for Proposed Change In Mr. H. R. Denton's (USNRC) letter to Mr. W. P. Stewart (FPC) of July 6,1979, Mr. Denton stated tSat " Appropriate Technical Specifications for Limiting Conditions for Operation and for surveillance requirements should be ... provided to the staff within seven days ... (and) should cover:

(1) Addition of flow rate indication for the emergenc/ feedwater system; TSCRN47 D71 t, e < ?S0

Page 2 (2) Addition of the auto start circuitry for the emergency feedwater pumps; (3) Addition of the anticipatory reactor trips; and (4) Changes in setpoint for high pressure reactor trip and PORV (or electromatic relief valve) actuation."

All of the above additions and changes have been performed at Crystal River Unit 3 and this Technical Specification Change Request incorporates them into the Technical Specifications save the change in setpoint for high pressure reactor trip. This change was already incorporated into the Technical Specifications by Amendment 19 to the Crystal River Facility Operating License.

Safety Analysis of Proposed Change The addition of flow rate irdication for the emergency feedwater system will enhance the operation of this system. The operator will be able to determine if adequate feedwater is being supplied to steam generators for heat removal . This change does not diminish any of the requirements applicable to the Safety Analysis and no unreviewed safety question is involved.

The addition of the auto start circuitry for the motor-driven emergency feedwater pump enhances the response of the emergency feedwater system in that there is greater certainty that at least one emergency feedwater pump will start when an emergency feedwater actuation signal is initiated. None of the requirements applicable to the Safety Analysis are diminished by the proposed change and no unreviewed safety question is involved.

The addition of the anticipatory reactor trip insures that the reactor will trip sooner than it would have before the anticipatory trips were installed.

The events that will initiate the anticipatory trips would have resulted in a high pressure reactor trip and these anticipatory trips will therefore limit the reactor pressure to a value below the high pressure trip setpoint. None of the requirements applicable to the Safety Analysis are dimimished by the proposed change and no unreviewed safety question is involved.

The change in the setpoint of the electromatic relief valve insures that the reactor will reach its high pressure trip setpoint first. This will reduce the number of challenges to the electromatic relief valve and therefore reducer probability of failure. None of the requirements applicable to the Safety Analysis are diminished by the proposed change and no urreviewed safety question is involved.

TSCRN47 371 347 90 k u .1 LU}

INDEX LIMITING CONDITIONS FOR CPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.4 PRESSURIZER........................................ 3/4 4-5 3/4.4.5 ST E AM GE N E R ATOR S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-6 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.......................... 3/4 4-13 Operational Leakage................................ 3/4 4-15 3/4.4.7 CHEMISTRY.......................................... 3/4 4-17 3/4.4.8 SPECIFIC ACTIVITY.................................. 3/4 4-20 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System............................. 3/4 4-24 Pressurizer........................................ 3/4 4-30 3/4.4.10 STRUCTURAL INTEGRITY............................... 3/4 4-31 3/4.4.11 RELIEF VALVE - SHUTD0WN............................ 3/4 4-33 3/4.5 EMERGENCY CORE CCOLING SYSTEMS (ECCS) 3/4.5.1 CORE FLOODING TANKS................................ 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T3yg 1280 F...................... 3/4 5-3 3/4.5.3 ECCS SUBS YSTEMS -a T v9 <280 F. . . . . . . . . . . . . . . . . . . . . . 3/4 5-6 3/4.5.4 BORATED WATER STCR AGE TANK. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-7 CRYSTAL RIVER - UNIT 3 V t~ nn7 4o/ '

<ue

INDEX PASES SECTION PAGE 3/4.0 APPLICABILITY.........................'................ B 3/4 01 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTR0L................................... b 3/4 1-1 3/4.1.2 B OR AT I O N SY S TE MS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 1-3 3/4.2 POWER 3SilIBUTION LIMITS............................. B 3/4 2-1 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION.......... B 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTR" MENTATION.................................. B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION......................... B 3/4 3-2 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT L00PS.............................. B 3/4 4-1 3/4.4.2 and 3/4.4.3 SAFETY VALVES........................... B 3/4 4-1 3/4.4.4 PRESSURIZER........................................ B 3/4 4-2 3/4.4.5 STEAM GENERATORS................................... B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE,.................... B 3/4 4-4 3/4.4.7 CHEMISTRY.......................................... B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY.................................. B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS........................ B 3/4 4-6 3/4 4.10 STRUCTURAL INTEGRITY............................... B 3/4 4-13 3/4 4.11 RELIEF VALVE....................................... B 3/4 4-13 CRYSTAL RIVER - UNIT 3 IX gss no)

<o

o

$ TABLE 2.2-1 (Continued)

U y REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS S

h FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES h 8. Reactor Containment Vessal 14 psig 14 psig E Pressur< High w

" 9. fiaia Turbine Auto Stop Oil Low (2) >45 psig >45 psig

10. liain Feedwater Pump Turtpige >55 psig >55 psig A dnd C Control Oil Low (2)

11. OiSG A and B Level Low-Low (2) >l8 inches >18 inches

'?

m (1) Trip may be manually bypassed when RCS pressure 11720 psig by actuating Shutdown Bypr.ss provided that:

> a. The Nuclear Overpower Trip Setpoint is $5% of RATED THERMAL POWER

~"

b. The Shutdown Bypass RCS Pressure - High Trip Setpoint of <1720 psig is imposed, and

c. The Shutdown Bypass is removed when RCS Pressure >l800 psig.

, (2) Trip not part of React or Protection System c2 l::=

9 -

g; TABLE 3.3-1 C

r-REACTOR PROTECTION SYSTEM INSTRUMENTATION ei MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE g FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Manual Reactor Trip 1 1 1 1, 2, and

8

2. Nuclear Overpower 4 2 3 1, 2 2#

3. RCS Outlet Temperature--High 4 2 3 1, 2 3#

4. Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE 4 2(a) 3 1, 2 2#

$$ S. RCS Pressure--Low 4 2(a) 3 1, 2 3#

{l 6. RCS Pressure--High 2 3 1, 2 3#

7. Variable Low RCS Pressure 4 2(a) 3 1, 2 3#

8. Reactor Containment Pressure--High 4 2 3 1, 2 3#

9. Intermediate Range, Neutron Flux and Rate 2 0 2 1, 2 and

4

10. Source Range, Neutron Flux and Rate A. Startup 2 0 2 2## and

5 B. Shutdown 2 0 1 3, 4 and 5 6

,s 61

'J C3 LJ1

@ TABLE 3.3-1 (Continuedl REACTOR PROTECTION SYSTEh INSTRUMENTATidN a

5

" MINIMUM TOTAL NO. CliANNELS CllANNELS APPLICABLE FUNCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE f10 DES ACTION Z 11. Control Rod Drive Trip Breakers 2 per trip 1 per trip 2 per trip 1, 2 and

7#

w system system system

12. Reactor Trip Module 2 per trip 1 per trip 2 per trip 1, 2 and

7#

system system system

13. Shutdown Bypass RCS Pressure--liigh 4 2 3 2**, 3**, 6#

4**, 5**

14. Main Turbine I.'o Stop Oil--Low (1) 1 1 1 1*** 8 w 15. Main Feedwater Pump Turb s h A and B Control Oil--Low 2 2 2 l### 8

16. OTSG A and B Level--Low-Low (1) 2 2 2 1,2,3,4 8 n

CN (1) Trip not part of Reactor Protection System n)

C3 C%

TABLE 3.3-1 (Continuedl TABLE NOTATION

With the control rod drive trip breakers in the closed position and the con-trol rod drive system capable of rod withdrawal.

- When Shutdown Bypass is actuated.

- - Trip function may be bypassed in this MODE with reactor power below 20%

RATED THERMAL POWER. Bypa:s shall be automatically removed when reactor power exceeds 20% RATED THERMAL POWER.

The provisions of Specification 3.0.4 are not applicable.

1. High voltage to detector may be deenergized above 10-10 amps on both Inter-mcdiate Range channels.

1. 1. Trip function may be bypassed in this MODE with reactor power below 10%

RATED THERMAL POWER. Bypass shall be autcmatically removed when reactor power exceeds 10% RATED THERMAL POWER. l (a) Trip may be manually bypasscu when RCS pressure 11720 psig by actuating Shutdown Bypass provided that:

(1) The Nuclear Overpower Trip Setpoint is 15% of RATED THERMAL POWER, (2) The Shutdown Bypass RCS Pressure--High Trip Setpoint of 11720 psig is imposed, and (3) The Shutdown Bypass is removed when RCS pressure >1800 psig.

ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Chranels OPERABLE requirement, restore the inoper-able channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in

least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open t' -

trol rod drive trip breakers.

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels STARTUP and/or POWER OPERATION may proceed provided all of the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition wi thin one hour.

b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1,

,. ,o7

[i' "A "% LUi CRYSTAL RIVER - UNIT 3 3/4 3-3

o m

TABLE 3.3-2 Ci h REACTOR PROTECTION SYSTEM INSTRUMENTATION RESPONSE LIMES iS

.o b FUNCT10NAL UNIT RESPONSE TIMES 5

" Not Applicable

1. Manual Reactor Trip

2. Nuclear Overpower

10.3 seconds

3. RCS Outlet Temperature--liigh Not Applicable

4. :<uclear Overpowar Based on Rt,S Flow and AXIAL POWER IMBALANCE * -<l.4 seconds u

E. RCS Pressure--Low 10.5 seconds Y'

6. RCS Pressure--High 10.5 seconds

7. Variable Low RCS Pressure Not Applicable

8. React ar Contair.inent Pressure--liigh Not Applicable

9. Main Turbine Auto Stoo Oil Low ** Not Applicable

10. Main Feedwater Pump Turbines A and B Control Oil Low ** Not Applicable

^ 11. OTSG A and B Level Low-Low ** Not Applicable m

Neutron detectors are exempt from response tine testing. Response time of the neutron flux signal portion of the channel shall be ceasured from detector output or input of first electronic component in channel.

PJ ** Trip not part of Reactor Protection System l CD CO

o

$ TABLE 4.3-1 4

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS S

h CilANNEL MODES IN WillCli

, CllANNEL CllANNEL FUNCTIONAL SURVEILLANCE c FUNCTIONAL UNIT CilECK CALIBRATION TEST REQUIRED 5

1. Manual Reactor Trip N.A. N.A. S/U(1) N.A.

]

2. Nuclear Overpower S D(2) and Q(7) M 1, 2

3. RCS Outiet Temperature--liigh S R M 1, 2

4. Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE S(4) M(3) and Q(7,8) M 1, 2

5. RCS Pressure--Low S R M 1, 2

6. RCS Pressure--Iligh S R M 1, 2

{

7. Variable Low RCS Pressure S R M 1, 2

8. Reactor Containment Pressure-High S R M 1, 2

9. Intermediate Range, Neutron Flux and Rate S 9(7) S/U(1)(S) 1, 2 a nd *

10. Source Range, Neutron Flux dnd Rate S R(7) S/U(1)(S) 2, 3, 4, and 5

11. Control Rod Drive Trip Breaker N.A. N.A. M and S/U(1) 1, 2 and *

=-

2 71 P' .)

c ~)

O

o -

5 TABLE 4.3-1 (Continued)

REACTOR PROTECTI0f4 SYSTEM INSTRUMENTATI0ff SURVEILLANCE REQUIREMEf4TS N

h CHANNEL MODES IN WillCH

, CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE c FUNCTIONAL UNIT-CHECK CALIBRATION TEST REQUIRED 5

12. Reactor Trip Module N.A. N.A. M 1, 2 and *

]

13. Shutdown Bypass RCS Pressure-High S R M 2**, 3**,

4**, 5**

14. Main Turbine Auto Stop Oil--Low # S P M 1

13. Main Feedwater Pump Turbine A and B Control Oil--Low # S R M 1 R

16. OTSG A and B Level--Low-Low # S R M 1, 2, 3, 4 V'

O n

cn # Trip not part of Reactor Protection Systeni u.~ t D .)

s ^.

E! .

a d

r-TABLE 3.3-3 (Cont'd)

US ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION M

w MINIMUM E TOTAL NO. CHANNELS CHANNELS APPLICABLE Q FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES Atfl0N us

b. Steam Line Rupture Matrix

1. Low SG Pressure 2 per steam 1 per steam 2 per steam 1, 2, 3*** 10 generator generator generator

2. Automatic Actuation Logic 1 per steam 1 per steam 1 per steam 1,2,3 10 generator geriera tor generator u2 c. Emergency Feedwater

l. Main Feedwater Punp Turbines A and B Control Oil Low 2 2 2 1, 2, 3## 10

2. OTSG A and B Level low-Low 2 2 2 1,2,3,4 10 m

T'-

'u~'

P3

~2

TABLE 3.3-3 (Continued)

TABLE NOTATION Trip function may be byoassed in this MODE with RCS pressure below 900 psig. Bybass shall be automatically removed when RCS pressure exceeds 1700 psig.

- Trip function may be bypassed in this MODE with RCS pressure below 900 psig. Bypass shall be automatically removed when RCS pressure exceeds 900 psig.

- - Trip function may be bypassed in this MODE when steam generator pres-sure below 725 psig. Bypass shall be autanatically removed when steam generator pressure exceeds 765 psig.

The provisions of Specification 3.0.4 are not applicable.

1. Trip function may be bypassed in this MODE prior to stopping the operating main feedwater pump. Bypass shall be manually removed > cer starting the first main feedwater pump.

ACTION STATr"E.1TS ACTION 9 - With the number of OPERABLE Channels one less than the Total Number of Channels operation may proceed until per-formance of the next required CHANNEL FUNCTIONAL TEST pro-vided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 10 - With the number of OPERABLE channels one less than the Total Number of Channels, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; how-ever, one channel may be bypassed for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for surveillance testing per Specification 4.3.2.1.1.

ACTION 11 - With less than the Minimum Channels OPERABLE, operation may continue provided tr 7tainment purge and exhaust valves are maintained closea.

ACTION 12 - With the number of OPERABLE Channels one less than the Total Number of Channels opcration may proceed provided the inoperable channel is placed in the bypassed condition and the minimum channels OPERABLE requirement is demonstrated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; one additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for Surveillance testing per Specification 4.3.2.1.

ACTION 13 - With the number of OPERABLE Channels one less than the Total Number of Charnels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

CRYSTAL RIVER - UNIT 3 3/4 3-14 4 6 [] 292

El a

j TABLE 3.3-4 (Continued)

$ ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION TRIP SETP0lNTS M

n

' TRIP SETPOINT ALLOWABLE VALUES FUNCTIONAL UNIT E

p 3. REACTOR BLDG. SPRAY u>

a. Reactor Bldg. Pressure High-High 3 30 psig f 30 psig coincident with HPI Signal See 1.a.2, 3, 4 See 1.a ', 3, 4

b. Automatic Actuation Logic Not Applicable Not Applicable u, 4. OTHER SAFETY SYSTEMS 1

o, a. Reactor Bldg. Purge Exhaust Duct Isolation on High Radioactivity Gaseous

Not Applicable

b. Steam Line Rupture Matrix

1. Low SG Pressure > 600 psig > 600 psig

2. Automatic Actuation Logic Not Applicable Not Applicable

c. Emergency Feedwater

1. Main Feedwater Pump Turbines > 55 psi 9 > 55 psig A and B Control Oil Low

2. OTSG A and B Level Low-Low ' 18 inches > 18 inches

,3

Detennined by requirenents of Appendix "B" Tech. Specs. Section 2.4.2 - Crystal River 3 Operating

,r _

License No. DPR-72.

W

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS *

1. Manual

a. High Pressure Injection Not Applicable

b. Low Pressure Injection Nnt Applicable

c. Reactor Building Cooling Not Applicable

c. Reactor Building Isolation Not Applicable

e. Reactor Building Spray Not Applicable

f. Reactor Building Purge Isolation Not Applicable

g. Steam Line Rupture Matrix

1) Emergency Feedwater Actuation Not Applicable

2) Feedwater Isolation Not Applicable

3) Steam Line Isolation Not Applicable 2 Reactor Building Pressure-High

a. High Pressure Injection 25*

b. Low Pressure Injection 25*

c. Reactor Building Cooling 25*

d. Reactor Building Isolation 60*

3. Reactor Building Pressure High-Hige (with HPI Signal)

a. Reactor Building Spray 56*

4. RCS Pressure Low

d. High Pressure Injection 25*

5. RCS Pressure Low-Low

a. High Pressure Injection 25*

b. Low Pressure Injection 25*

CRYSTAL RIVER - UNIT 3 3/4 3-17 nn4" f F ') Li

TABLE 3.3-5 (tontinued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS *

6. Low Steam Generator Pressure

a. Feedwater Isolation 34

b. Steam Line Isolation 5

7. Main Feedwater Pumo Turbines A and B Control Oil Low

a. Emergency Feedwater Actuation Not Applicable

8. OTSG A and B Level Low-Low

a. Emergency Feedwater Actuation Not Applicable

Diesel Generator starting and sequence loading delays included. Res ponse time limit includes movement of valves and attainment of pump or blower di scha rge press ure.

/' 'a'"

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'l 5 CRYSTAL RIVER - UNIT 3 3/4 3-17a

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}{ TACLE 4.3-2 (Continued) h ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION SURVEILLANCE REQUIREMENTS B

h CHANNEL MODES IN WHICH i CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE g; FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

3. REACTOR BLDG. SPRAY u>

a. Reactor Bldg. Pressure High-High coincident with HPI Signal S R M(4) 1, 2, 3

b. Automatic Actuation Logic N/A N/A M(3) 1, 2, 3 jj 4. OTHER SAFETY SYSTEMS Y' a. Reactor Bldg. Purge Isolation E! on High Radioactivity

1. Saseous S R M All Modes

b. Steam Line Rupture Matrix

1. Low SG Pressure N/A R N/A 1,2,3

2. Automatic Actuation Logic N/A N/A M(3) 1,2,3(5)

c. Emergency Feedwater 23, 1. Main feedwater Pump furbines S R N/A 1,2,3 cn A and B Control Oil Low u

2. OTSG A and B Level Low-Low S R N/A 1, 2, 3, 4

d.

b a '

s REACTOR COOLANT SYSTEM RFLIEF VALVE - SHUTDOWN LIMITING CONDITION 'R OPERATION 3.4.11 The pressurizer electromatic relief valve shall be OPERABLE with a lift setting of <550 psig.

APPLICABILITY: MODES 4, 5 and 6*

ACTION:

With the pressurizer eiectromatic relief valve not OPERABLE, restore the valve to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or vent the Reactor Coolant System within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.11 The pressurizer electromatic relief valve shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that the pressurizer electromatic relief isolation valve is open, and

b. At least once per 18 months, during shutdown, by verifying the lift setting.

With the reactor vessel head not removed.

CRYSTAL RIVER - UNIT 3 3/4 4-33 Ac7 . o 7

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PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (CONTINUED)

2. Verifying that each valve (manual, power-operated or auto-matic) in the flow path that is not locked, sealed or other-wise secured in position, is in its correct position.

3. Verifying that the emergency feedwater ultrasonic flow rate detector is zero-checked.

b. At least once per 18 months, during shutdown, by:

1. Verifying that each automatic valve in the flow path actuates to its correct position on each emergency feedwater actuation test signal .

2. Verifying that the steam turbina driven pump and the motor driven pump start automatically:

a. Upon receipt of an emergency feedwater actuation OTSG A and B level low-low test signal, and

b. Upon receipt of an emergency feedwater actuation main feedwater pump turbines A and B control oil low test signal.

3. Verifying that the operating air accumulators for FWV-39 and FWV-40 maintain >27 psig for at least one hour when isolated from their air supply.

CRYSTAL RIVER - UNIT 3 3/4 7-5 b h!) '

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REACTOR COOLANT SYSTEM (Continued)

BASES All pressure-temperature limit curves are applicable up to the fifth effective full power yea , The protection against non-ductile failure is assured by maintaining the coolant pressure below the upper limits of Figures 3.o-2, and 3.4-3, and 3.4-4.

TF.e pressure and temperature limits shown on Figures 3.4-2 and 3.4-4 for reactor criticallity and for inservice leak and hydrostatic testing have been prr,vided to assure compliance with the minimum temperature requirements of A9pendix G to 10 CFR 50.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

The limitations imposed on pressurizer heatup and cooldown 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.

3/4.4.10 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 canponente.,

except steam generator tubes, ensure that the structural integrity sr these components will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.

The internals vent valves are provided to relieve the pressure generated by steaming in the core following a LOCA so that the core remains sufficiently c ov er ed . Inspection and manual actuation of the internals vent valves 1) ensure OPERABILITY, 2) ensure that the valves are not stuck open during normal operation, and 3) demonstrate that the valves are fully open at the forces assumed in the safety analysis.

3/4.4.11 RELIEF VALVE The OPERABILITY of the electomatic relief valve ensures chat sufficient overpressure protection is provided and that the RC System pressure stays within Appendix G limits with Tavg 1280 F.

CRYSTAL RIVER - UNIT 3 B 3/4 4-13 c> ~nq

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ML19241B427

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