Text

Proposed Tech Specs,Clarifying Proposed Limiting Conditions of Operation & Addl Surveillance Requirements for Emergency Feedwater Initiation & Control Sys
	ML20236M353
Person / Time
Site:	Rancho Seco
Issue date:	07/31/1987
From:	; SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	;
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ML20236M344	List: ML20236M339; ML20236M353; ML20236M366;
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	TAC-64359, NUDOCS 8708110031
	Download: ML20236M353 (46)
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ENCLOSURE 2 Proposed Technical Specifications Amendment No. 152 I

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. . i LIST OF EFFECTIVE PAGES PROPOSED AMENDMENT NO. 152 EAGE IMGE i

iii 3-28 1-2 3-30a 1-7 3-30b 3-1 3-34 3-2 3-38d 3-2a 3-38e l I

3-23 4-7b 3-23a 4-7c 3-24 4-7d 3-24a 4-7e 3-25 4-8 3-25a 4-8a 3-26 4-39 3-26a 4-39a l 3-26b i

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i RANCHO SECO UNIT 1

- - TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)

Page Sec tion 3.1.6 Leakage 3-12 3.1.7 l4oderator Temperature Coef ficient of Radioactivity 3-15 3.1.8 Low Power Physics Testing Restrictions 3-ISb 3.1.9 Control Rod Operation 3-16 f 3.2 HIGH PRESSURE INJECTION, CHEiICAL ADDITION AND LOW TE4PERATURE OVERPRESSURE PROTECTION (LTOP) SYSTE45 3-17 3.3 E4ERGENCY CORE COOLING, REACTOR BUILDING E4ERGENCY COOLING, AND REACTOR BUILDING SPRAY SYSTE45 3-19 3.4 STEA4 AND POWER CONVERSION SYSTE4 3-23 3.5 INSTRU4ENTATION SYSTE4S 3-25 3.5.1 Operational Safety Instrumentation 3-25 l 3.5.2 Control Rod Group and Power Distribution Limits 3-31 3.5.3 Safety Features Actuation System Setpoints 3-34 3.5.4 Incorc Instrumentation 3-36 3.5.5 Accident ;4cnitoring Instrumentation 3-38a 152>< 3.5.6 Emergency Feedwater Initiation and Control Setpoints 3.6 REACTOR BUILDING 3-59 3.7 AUXILIARY ELECTRICAL SYSTE4S 3-41 i

3.8 FUEL LOADING AND REFUELING 3-44 3.9 Deleted 3.10 SECONDARY SYSTE4 ACTIVITY 3-47 3.11 REACTOR BUILDING POLAR CRANE AND AUXILIARY HOIST 3-49 3.12 SH0CK SUPPRESSORS (SNUBBERS) 3-51 3.13 AIR FILTER SYSTEiS 3-52 3.14 gN 3-53 FIRE SUPPRESST_'

3.14.1 Instrumentation 3-53 3.14.2 Water System 3-53 3.14.3 Spray and Sprinkler Systems 3-56 3.14.4 C0p System 3-56 Proposed Amendment No. 152 iii

RANCHO SECO UNIT 1 I

TECHNICAL SPEC 1FICATIONS Definitions 140 F. Pressure is defined by Specification 3.1.2. A refueling shutdown refers to a shutdown to replace or rearrange all or a portion of the fuel assemblies and/or control rods.

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1.2.7 Refueling Operation An operation involving a change in core geometry by manipulation of fuel or control rods when the reactor vessel head is removed. j 1.2.8 Refueling Interval

Time between normal refuelings of the reactor, not to exceed 24 months for the first refueling and 18 months thereafter without prior approval of the NRC. l 1.2.9 Startup The reactor shall be considered in the startup mode when the shutdown margin is reduced with the intent of going critical.

I 1.2.10 Remain Critical A technical specification that requires that the reactor shall not remain critical shall mean that an uninterrupted normal hot shutdown procedure will i 152x be completed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months unless otherwise specified. l

' l 1.2.11 Tavg {

At operating conditions TAVG is defined as the arithmetic average of the l coolant temperatures in the hot and cold legs of the loop with the greater I number of reactor coolant pumps operating, if such a distinction of loops can be made.

1.2.12 Heatup - Cooldown Mode The heatup-cooldown mode is the range of reactor coolant temperature greater j than 200 F and less than 525 F. l l

1.3 OPERABLE A component or system is operable when it is capable of performing its l intended function within the required range. The component or system snail be considered to have this capability when: (1) it satisfies the limiting conditions for operation defined in Specification 3, (2) it has been tested periodically in accordance with Specification 4, and has met its performance requirements, (3) the system has available its normal and emergency sources of power, and (4) its required auxiliaries are capable of performing their intended function. When a system or component is determined to be inoperable solely because its normal power source isinoperable or its emergency power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation provided its redundant system or component is OPERABLE with an OPERABLE normal and emergency power source.

See page 1-2b Proposed Amendment No. 152 1-2

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions 1.15 0FFSITE 00SE CALCULATION MANUAL (ODCM)

An 0FFSITE DOSE CALCULATION MANUAL (CDCM) shall be a manual containing the methodology and param9ters to be used in the I calculation of offsite dose due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints and specific details of the environmental radiological monitoring program.

1.16 RESTRICTED AREA That portion of the site property, the access to which is controlled by security fencing, equipment and personnel.

1.17 SITE B0UNDARY The boundary of the SMUD owned property. l 1.18 DOSE EQUIVALENT I-131 l l

The DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites".

1.19 FEMBER(S) 0F THE PUBLIC l

MEMBER (S) 0F THE PUBLIC shall include all individuals who by virtue of their occupational status have no formal association with the plant. This category shall include non-employees of the licensee who are permitted to use portions of the site for recreational, occupational, or other purposes not associated with plant functions. This category shall not include non-employees such as vending machine servicemen or postmen who, as part of their formal job function, occasionally enter an area that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

152> 1.20 VECTOR LOGIC A set of circuitry in each channel of the EFIC system which once AFW has been initiated determines whether AFW to a steam generator should be allowed or terminated and the signal output for each

< EFIC channe'l 'do the AFW valves associated with that channel .

Proposed Amendment No. 152 1-7

RANCHO SECO UNIT 1 . .

TECHN1 CAL SPECIFICATIONS Limiting Conditions for Operation

3. LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR COOLANT SYSTEM Applicability Applies to the operating status of the reactor coolant system.

Objective To specify those limiting conditions for operation of the reactor coolant system which must be met to ensure safe reactor operations.

3.1.1 OPERATIONAL COMPONENTS l Specification i 3.1.1.1 Reactor Coolant Pumps A. Pump combinations permissible for given power levels shall be as shown in specification table 2.3-1. l l

B. The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant.

I C. Operation at power with two pumps shall be limited to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 30 day period.

3.1.1.2 Steam Generators 152> A. Two steam generators shall be operable whenever the reactor coolant average temperature is above 280 F, except as described in 5.1.1.2.B.

B. With one or more steam generator (s) inoperable due to excessive leakage per 3.1.6.9, bring the ' reactor to cold shutdown conditions within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

C.

With one or more steam g(3nerator(s) inoperable due to steamg to operable status prior to increasing reactor coolant average temperature above 200*F.

3.1.1.3 Pressurizer Safety Yalves A. The reactor shall not remain critical unless both Pressurizer Coolant System code safety valves are operable.

B. When the reactor is subcritical, at least one Pressurizer code safety valve shall be operable if all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Boiler and Pressure Vessel Code,Section III.

Proposed Amendment No.152 3-1

l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation (

3.1.1.4 Pressurizer Electromatic Relief Valve A. The nominal setpoint of the pressurizer electromatic relief valve shall be 2450 psig

10 psig except when required for cold overpressure protection. l l

3.1.1.5 Decay Heat Renoval l A. At least two of the coolant loops listed below shall be operable when the coolant average temperature is below 280"F.

except during fuel loading and refueling.

1. Reactor Coolant Loop ( A) and its associated steam generator and at least one associated reactor coolant pump, 1

2. Reactor Coolant Loop (B) and its associated steam generator and at least one associated reactor coolant pump,

3. Decay Heat Removal Loop (A)

4. Decay Heat Removal Loop (B)

With less than the above required coolant loops OPidABLE, ,

immediately initiate corrective action to return the required (

coolant loops to OPERABLE status as soon as possible; be in COLD I SHUTDOWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months . l 3.1.1.6 Reactor Coolant System High Point Vents A. The vent path on Loop A and vent path on Loop B shall be operable and closed during power operation.

B. The vent path on the pressurizer shall be operable and closed during power operation.

C. With one of the above reactor coolant system vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days. If the status is not restored to operable in 30 days, be in HOT STANDBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-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 .

D. With two or more of the above reactor coolant system vent paths inoperable; maintain the inoperable vent paths closed with power removed from the valve actuators of all the valves in the inoperable vent paths, and restore at least (two) of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If the status is not restored to operable in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , be in HOT STANDBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-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 .

Proposed Amendment No. 152 3-2

RANCHO SECO UNIT 1 TECHNICAL SPEC 1FICATIONS Limiting Conditions for Operation

( Bases A reactor coolant pump or decay heat removal pump is required to be in operation before the boron concentration is reduced by dilution with makeup l water. Either pump will provide mixing which will prevent sudden positive j reactivity changes caused by dilute coolant reaching .the reactor. One decay heat removal pump will circulate the equivalent of the reactor. coolant system volume in one half hour or less. (1) ;

The decay heat removal system suction piping is designed for 300 F and 300 thus, the system can remove decay heat when the reactor coolant system f psig; is be low this temperature. (2) (3) ]

4' One Pressurizer code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat. (4) Both Pressurizer code' ,

safety valves are required to be in service prior to criticality to conform to !

the system design relief capabilities. The code safety valves prevent J overpressure for rod withdrawal accidents. (5) The Pressurizer code safety valve lift set point shall be set at 2500 psig

1 percent allowance for error and each valve shall be capable of relieving 345,000 lb/hr of saturated steam at a pressure not greater than 3 percent alsove tne set pressure.

The electromatic relief valve setpoint was established to prevent operation of s the Safety Valves during transients.

Two pump operation is limited until further ECCS analysis h perfermed.

152> When the reactor is not critical but TAV is above 280* F, one steam generator provides sufficient heat removal capability for removing decay heat. However, 4 single failure considerations require that both steam generators be operable.

When TAV is below 280*F, a single reactor coolant loop or DHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two DHR loops to ;

be OPERABLE. l i

The purpose of the high point vents is to vent noncondensible gases from the i RCS which may inhibit core cooling during natural circulation. In compliance with 10CFR50 Appendix R the power to all the valve actuators in the vent path has been removed.

REFERENCES (1) FSAR Tables 9.5-2, 4.2-1, 4.2-2, 4.2-4, 4.2-5, 4.2-6

, (2) FSAR paragraph 9.5.2.2 and 10.2.2 l

k (3) FSAR paragraph 4.2.5 I (4) FSAR paragraph 4.3.8.4 and 4.2.4 (5) FSAR paragraph 4.3.6 and 14.1.2.2.3 Proposed Amendment No.152 3-2a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.4- STEAM AND POWER CONVERSION SYSTEM Applicability Applies to the operability of the turbine cycle during normal operation ana l for the removal of decay heat.

Objective To specify minimum conditions of the turbine cycle equipment necessary to assure the required steam relief capacity during normal operation and the capability to remove decay heat from the reactor core.

Specification ,

152> 3.4.1 The reactor coolant system shall not be brought or remain above 280F with irradiated fuel in the pressure vessel unless the following conditions are met:

A. Capability to remove decay heat by use of two stean generators as specified in 3.1.1.2. A.

B. One atmospheric dump valve per steam generator snall be operable.

C. A minimum of 250,000 gallons of water shall be available in the condensate storage tank.

D. Twn main steam system safety valves are operable per steam generator.

E. Both auxiliary feedwater trains (i.e., pumps and their flow paths) are operable.

F. Both trains of main feedwater isolation-on each main feedwater line are operable.

G. Four independent backup instrument air bottle supply systems for ADVs and MFW, SFW, and AFW valves are operable.

With less than the above required components operable, be on decay

< heat cooling within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Proposed Amendment No. 152 3-23 f

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1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS l l

Limiting Conditions for Operation {

152> 3.4.2 The reactor shall not be brought or remain critical unless the following conditions are met: i 1

A. Capability to remove decay heat by use of two steam generators l as specified in 3.1.1.2. 1 i

B. One atmospheric dump valve per steam generator shall be operable except that: (1) with only one atmospheric dump valve l

operable, restore an inoperable valve for the other steam j generator within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in hot shutdown within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (

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and on decay heat cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;- (2) wi th j no atmospheric dump valves- operable, restore at least one l inoperable valve within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in hot shutdown within i the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and on decay heat cooling within the next 12 l hours l

C. A minimum of 250,000 gallons of water shall be available in i the condensate storage tank except that with less than the j minimum volume, restore the minimum volume witnin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or l be in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and on decay heat I cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . l D. Seventeen of the eighteen main steam safety valves are operable except that with le::s than the minimum number of valves, restore the inoperable valve (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be

! in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and on decay heat J cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . j i

E. Four turbine throttle stop valves are operable except that with less than the minimum number of valves, restore the inoperable valve (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in hot shutdown l within the next 6 houre and on decay heat cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

F. Both auxiliary feedwater trains (i.e., pump and their flow path) are operable except that: j (1) With one auxiliary feedwater train inoperable, restore I the train to operable status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; or be in hot l shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and on decay heat cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . :

(2) With both auxiliary feedwater trains inoperable, the reactor shall be made subcritical within four. hours and the reactor shall be on decay heat cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , i

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Proposed Amendment No. 152 3.-23a

4 4 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 152> 3.4.2 G. Both trains of main feedwater isolation on each main feedwater line are operable except that:

(1) With one main feedwater isolation train inoperable, restore the train 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 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and on decay heat cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(2) With both main feedwater isolation trains inoperable, the reactor shall be made subcritical within four hours and the reactor shall be on decay heat cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

H. Two independent backup instrument air bottle supply systems (one per steamline) for ADVs are operable except that:

(1) With one system inoperable, restore the system to operable status within 7 days or be in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and on decay heat cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(2) With two systems inoperable, restore at least one system within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . With one system restored to operable

< status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , follow 3.4.2.H.(1).

I. Two independent backup instrument air bottle supply systems )

i (one per feed water line) for MFW, SFW, and AFW control valves are operable except that with either one or both system (s) inoperable, restore the inoperable system (s) within 7 days or !

be in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and on decay heat j cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . 1 Bases The feedwater system and the turbine bypass system are rormally used for !

152> decay heat removal and cooldown above 280 F. Main feedwater is supplied by l If neither main feed operation of a condensate pump and main feedwater pump.

pump is available, feedwater can be supplied to the steam generators by an auxiliary feedwater pump. Steam relief capability is provided the system's atmospheric dump valves.

The auxiliary feedwater system is designed to provide sufficient flow on loss I of main feedwater to match decay heat plus Reactor Coolant Pump heat input to i i

the Re occur.ggpor Coolant System before solid pressurizer operation could l

l Proposed Amendment No.152 l 3-24 l I

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. . 1 RANCHO SECO UNIT 1 l

TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 152> The 250,000 gallons of water in the condensate storage tank is sufficient to remove decay heat (plus Reactor Ceolant pump heat for two pumps) for l

approximately 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months . This volume provides sufficient water to remove the

! decay heat for approximately 5.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and to subsequently cool

< the plant to the DHR system pressure ct a cooldown rate of 50*F/hr (1) .

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The minimum relie{21 apacity This is of seventeen sufficient steam to capacity system safety protect tile valves steamis I 13,329,163 lb/hr. l ystem under the design overpower condition of 112 percent.t3; 152> Both trains of main feeduater isolation on each main feedwater line are l required to be operable. Train A of main feedwater isolation is comprised of l main feedwater control valves, main feedwater block valves and startup l ccatrol valves. Train B of main feedwater isolation is comprised of the main feedwater isolation valves.

l Four independent Class 1 backup air supply systems are provided to assure power available to certain air operated valves in the event of the loss of

normal air supply. One system supplies power for the MFW, Startup Feedwater l

' (SFW) and AFW control valves feeding the "A" 0TSG; another system supplies !

l power for same valves feeding the "B" OTSG. Two systems supply power for !

ADVs with one for the ADVs on the "A" main steam line and one for the ADVs on l the "B" main steam line. Each system is sized to provide at least two hours

< of air supply.

REFERENCES 152> (1) B and W Document 32-1141727-00, " Heat Removal Capability of SMUD

< CST," March 1984 (2) FSAR paragraph 10.3.4 (3) FSAR Appendix 3A, Answer to Question 3A.5 152> (4) B and W Calculation 86-1167930, " Rancho Seco: AFW Minimum Flcw

< Analysis," (SMUD t alculation No. Z-FWS 10150)

Proposed Amendment No. 152 3-24a

RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS Limiting Conditions for Operation l 3.5 INSTRUMENTATION SYSTEMS {

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3.5.1 OPERATIONAL SAFETY INSTRUMENTATION l

Appl,cability j I

Applies to unit instrumentation and control systems.

Objective a To delineate the conditions of the unit instrumentation and safety circuits !

necessary to assure reactor safety.

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Specifications )

3.5.1.1 Startup and operation are not permitted unless the requirements of f Table 3.5.1-1, Columns A and B are met, j i

I 3.5.1.2 In the event the number of protection channels operable falls below the limit given under Table 3.5.1-1, Columns A and B, operation shall be limited as specified in Column C.

ifs In the event the number of operable Process Instrumentation or EFIC

< system channels is less than the Total Number of Channel (s), restore the inoperable channels to operable status within 7 days, or be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If the number of 152> operable channels is one less than the minimum channels operable, either restore the inoperable channels to operable within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If the number of operable channels is two less than the minimum channels operable, the reactor shall be made subcritical within foue hours ]

< and on decay heat cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ;

l 3.5.1.3 For on-line testing or in the event of a protection instrument or j channel failure, a key operated channel bypass switch associated with each reactor protection channel will be used to lock the j channel trip relay in the untripped state as indicated by a light. :

Only one channel shall be locked in this untripped state at any one l time. ;

i 3.5.1.4 The key operated shutdown bypass switch associated with each reactor j protection channel shall not be used during reactor power operation. ;

I 3.5.1.5 During startup when the intermediate range instrument comes on scale, the overlap between the intermediate range and the source j range instrumentation shall not be less than one decade. If the {

overlap is less than one decade, the flux level shall be maintained in the source range until the one decade overlap is achieved. j l

Proposed Amendment No. 152 3-25 i

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.1.6 In the event that one of the trip devices in either of the sources supplying power to the control rod drive mechanisms fails in the untripped state, the power supplied to the rod drive mechanisms through the failed trip device shall be manually removed within 30 minutes. The condition will be corrected and the remaining trip devices shall be tested within eight hours. If the condition is not corrected and the remaining trip devices are not tested within the eight-hour period, the reactor shall be placed in the hot shutcown i condition within an additional four hours.

152> 3.5.1.7 For calibration or maintenance of an Emergency Feedwater Initiation and Control (EFIC) channel, a key operated " maintenance bypass" switch associated with each channel will be used which will prevent the initiate signal from being transmitted to the Channel A and B ,

trip logic. Only one channel shall be locked into " maintenance i bypass" at any one time.

3.5.1.8 If a channel of the RPS is in bypass, it is permissible to bypass only the corresponding channel of EFIC.

Bases Every reasonable effort will be made to maintain all safety instrumentation in operation. A startup is not permitted unless three power range neutron i instrument channels and two channels each of the following are operable:

four reactor coolant temperature instrument channels, four reactor coolant l

flow instrument channels, four reactor coolant pressure instrument channels, four pressure-temperature instrument channels, four flux-imbalance flow instrument channels, four power-number of pumps instrument channels, and four high reactor building pressure instrument channels. The safety features actuation system must have two analog channels functioning correctly prior to startup. EFIC system instrumentation as required by Table 3.6.1-1 must be operable.

Operation at rated power is permitted as long as the systems have at least .

the redundancy requirements of Column B (Table 3.5.1-1). This is in agreement with redundancy and single failure criteria of IEEE 279 as described in FSAR section 7.

! 152.= The four reactor protection channels were provided with key operated maintenance bypass switches interlocked to allow on-line testing or maintenance on only one channel at a time during power operation. Each channel is provided alarm and lights to indicate when that channel is

< bypassed. ,

i Proposea Amendment No. 152 3-25a i

i i ,

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Bases (Continued) i Each reactor protection channel key operated shutdown bypass switch is provided with alarm and lights to indicate when the shutdown bypass switch is being used. There are four shutdown bypass keys in the control room under the administrative control of the shif t supervisor. The keys will not be used during reactor power operation. ;

There are four reactor protection channels. Normal trip logic is two out of four. Required trip logic for the power range instrumentation channels is 152> two out of three. The EFIC trip logic is two times one-out-of-two taken twice. Minimum trip logic on other instrumentation channels is one out of i two. 1 l

The EFIC system is designed to automatically initiate AFW when:

1. all four RC pumps are tripped,

2. RPS has tripped the reactor on anticipatory trip indicating loss of main feedwater,

3. the level of either steam generator is low,

! 4. either steam generator pressure is low, or

5. SFAS ECCS actuation (high RB pressure or low RCS pressure).

The EFIC system will isolate main feedwater to any steam generator when the pressure goes below 600 psig.

The EFIC system is also designed to isolate or feed AFW according to the following logic:

- If both SGs are above 600 psig, supply AFW to both SGs

- If one SG is below 600 psig, supply AeW to the other SG

- If both SGs are be'ow 600 psig but the pressure difference between the two SGs exceeJs 100 psig, supply AFW only to the SG with the i higher pressure l 1

- If both SGs are below 600 psig and the pressure difference is less j than 100 psig, supply AFW to both SGs At cold shutdown conditions all EFIC initiate and isolate functions are manually or automatically bypassed. When pressure in both steam generators ;

is greater than 750 psig, the following bypassed initiation signals will have l been automatically reset: 1) Loss of 4 RC pumps, 2) low steam generator }

< pressure, 3) low steam generator level. !

Proposed Amendment No. 152 3-26 j i

I i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 152> Since the EFIC receives signals from the RPS it is important that only corresponding channels be placed in " maintenance bypass." If a channel of RPS is in maintenance bypass, only the corresponding channel of EFIC can be bypa s sed. An interlock feature also prevents bypassing more than one EFIC channel at a time. These ir.terlocking features allow the EFIC system to take a single failure in addition to having one channel in maintenance bypass.

Various RPS test features can inhibit initiate signals to the EFIC system and l

i degrade the EFIC system below acceptable limits if the RPS channel is not in l

bypass. Therefore, no testing should be performed on a RPS instrument string which supplies an output to EFIC without placing that RPS channel in bypass.

l The EFIC system is designed to allow testing during power operation. The EFIC system can be tested from its input terminals to the actuated device l controllers without placing the channel in key locked " maintenance bypass."

A test of the EFIC trip logic will actuate one of two relays in the !

I controllers. The two relays are tested individually to prevent automatic l

actuation of the component.

Each EFIC channel key operated maintenance bypass switch is provided with alarm and lights to indicate when the maintenance bypass switch is being used.

4 1 sca s l The source overlap range by one and intermediate decade. This decaderange nuclear overlap will flux instrumentation be achieved at 10- 0 amps ;

l on the intermediate range scale, Power is normally supplied to the control rod drive mecnanisms from two I separate parallel 480 volt sources. Redundant trip devices are employed in each of these sources. If any one of these trip devices fails in the ;

untripped state on-line repairs to the failed device, when practical, will be ;

made, and the remaining trip devices will be tested. Eight hours is ample l time to test the remaining trip devices and in many cases make on-line repai rs.

Proposed Amendment No. 152 3-26a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The OPERABILITY of the SFAS instrumentation systems and bypasses ensure that

1) the associated SFAS action will be initiated when the parameter monitorea by each channel or combination thereof reaches its setpoint, 2) the specified coincidence logic is maintained, 3) sufficient redundancy is maintained to permit a channel to ha out of service for testing or maintenance, and 4) sufficient system funct.'nal capability is available for SFAS purposes from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent ,

with the assumptions used f r the accident analyses. l

)

The OPERABILITY of the accident monitoring instrumentation ensures that i sufficient information is available on selected plant parameters to monitor l and assess these variables during and following an accident. This capability is consistent with the recommendations of Regulatory Guiae 1.97,

" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident", December 1975 and NUREG-0578, i "TMI-2 Lessons Learned Task Force Status Report and Short-Term '

Recommendations."

l REFERENCE l FSAR, Subsection 7.1 l

l l

l Proposed Amendment No. 152 3-26b

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i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS l Limiting Conditions for Operation 3.5.3 SAFETY FEATURES ACTUATION SYSTEM SETPOINTS Applicability This specification applies to the safety features actuation system actuation i setpoints. l Objective l

To provide for automatic initiation of the safety features actuation system )

in the event of a breach of reactor coolant system integrity.

Specification The safety features actuation setpoints and permissible bypasses shall be as follows:

Functional Unit Action Setpoint High Reactor Building Reactor Building spray valves *** 530 psig l pressure

Reactor Building spray pumps *** 130 psig High pressure injection, and start of Reactor Building cooling and Reactor Building isolation. 14 psig 152>< Low pressure injection,EFIC AFW initiate 14 psig Low reactor coolant system High pressure injection, and start pressure ** of Reactor Building cooling and Reactor Building Isolation 11600psigl 152>< Low pressure injection,EFIC AFW initiatel1600 psig)

Automatic Actuation Logic All above Not ApplicaDie:

Manual All above Not Applicaole 152><

May be bypassed during Reactor Building leak rate test.

- May be bypassed below 1850 psig and is automatically reinstated above 1650 psig

- - Five-minute time delay.

Proposed Amendment No. 152 3-34 u

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS f Limiting Condition for Operation i

152> 3.5.6 E4ERGENCY FEEDWATER INITIATION AND CONTROL SETPOINTS l Applicability This specification applies to the emergency feedwater initiation and control (EFIC) setpctnts.

Objective To provide for automatic initiation and control of auxiliary feedwater and automatic isolation of main feedwater.

Specification The emergency feedwater initiation and control setpoints and bypasses shall be as follows: i i

I I Functional Unit Action Setpoint l l

I

a. Low SG Level Initiates AFW >9 inches i

b. Low SG Pressure Initiates AFW >575 psig and Isolates MFW

c. Loss of All RCP Initiates AFW N/A

d. SFAS Actuation Initiates AFW N/A (1)

e. RPS Actuation on Initiates AFW N/A (2)

Loss Of MFW j

f. Vector Logic Isolates Faulted SG Various (3)

g. Shutdown Bypass Bypass Pennissive <750 psig (1) Refer to Specification 3.5.3 for SFAS setpoint (2) Refer to Table 2.3-1 for RPS setpoint (3) Refer to Bases below for description of vector setpoints Proposed Amendment 152 3-38d

1 RANCHO SECO UNIT 1 '

TECHNICAL SPECIFICATIONS Limiting Condition for Operation i

102n 3.5.6 (continued)

Bases The EFIC system is designed to automatically initiate AFW when:

I

1. all four RC pumps are tripped, or i

2. RPS has tripped the reactor on anticipatory trip indicating loss of main feedwater, or

3. the level of either steam generator is low, or

4. either steam generator pressure is low, or

5. SFAS ECCS actuation (high RB pressure or low RCS pressure).

The EFIC system will initiate main feedwater isolation to any steam generator as the pressure goes and steys below a minimum set point of 575 psig.

The EFIC system is also designed to isolate or feed AFW according to the following vector logic. Setpoints are nominal and subject to instrument i inaccuracies:

- If both SGs are above 600 psig, supply AFW to both SGs 1

- If one SG is below 600 psig, supply AFW to the other SG

- If both SGs are below 600 psig but the pressure difference between the two SGs exceeds 100 psig, supply AFW only to the SG with the l higher pressure

- If both SGs are below 600 psig and the pressure difference is less than 100 psig, supply AFW to both SGs At cold shutdown conditions all EFIC automatic initiate and isolate functions are manually or automatically bypassed. Prior to a pressure of greater than 750 psig in both steam generators, the following bypassed .nitiation signals automatically reset: 1) Loss of 4 RC pumps, 2) low steam generator pressure. 3) low steam generator level.

Bypassing of automatic AFW initiation on Loss of MFW Anticipatory Trip or SFAS actuation is controlled by bypass permissive logic within the RPS and SFAS, respectively.

Proposed Amendment No. 152 3-38e

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1 l

RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS Surveillance Standards l TABLE 4.1-2 l MINIMUM EQUIPMENT TEST FREQUENCY Item lest Frequency l

l 1. Control rods Rod urop times of Each refueling shutdown l

all f ull length rods

2. Control rod movement Movement of each rod Every two weeks l 3. Pressurizer Setpcint Note 3 code safety valves

4. Main Steam safety Setpoint Note 3 valves }

5. Refueling system Functional Each refueling intervai interlocks prior to handling fuel I

152x 6. Turbine throttle Movement of each valve Monthly i stop valves j

7. Reactor Coolant Leakage Calculated inventory weekly System Leakage check daily ,l

8. Charcoal and high Charcoal and HEPA filter Each refueling interval and efficiency filters for iodine and parti::ul- at any time work on filters ate removal efficiencies. could alter their integrity DOP test on HEPA filters.

Freon test on charcoal filter units.

9. Fire pumps and power Functional Monthly i supplies l

10. Reactor Building Functional Each refueling interval isolation trip

11. Spent fuel cooling Functional Each refueling interval system prior to fuel handling

12. Turbine Overspeed Calibration Each refueling interval Trips

13. Internals Vent Manual Actuation, III Each refueling interval Valves Remoty2{isualinspec-tion, and verify that valve not stuck open.

Proposed Amendment No. 152 4-8 i

l

.1 RANCHO SECO UNIT 1 {

TECHNICAL SPECIFICATIONS Surveillance Standards i TABLE 4.1-2 (Continued)

MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency f

14. Reactor Coolant Functional tpst of Each refueling interval I System High Point each valve (43 Vents

15. Low Temperature Functional (5) Prior to RCS temperature ,

Overpressure decreasing below 350"F Protection (EMOV) 152> 16. Main Feedwater Isolation Valves

a. Main Feedwater Functional Each refueling interval, Isolation Valves ,

b. Main Feedwater Functional Each refueling interval.

Block Valves

c. Startup Feedwater Functional Each refueling interval, j Control Valves {

d. Main Feedwater Functional Each refueling interval, j

Control Valves

)

17. Turbine Throttle Cycle Each refteling interval.

Stop Valves

18. Backup Instrument Functional Each refueling interval

< Air Supply System

1. Verifying through manual actuation that the valve is fully open with a force of < 400 lbs. (applied vertically upward). I i 2. Check visually accessible surfaces to evaluate observed surface ;

irregularities, j 1

! 3. Tested in accordance with Section XI of the ASME Boiler and Pressure '

l Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by ~ '

the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).

4. Cycle each valve in the vent path through at least one complete cycle of full travel from the control room and verify the flow of gas through the system vent path. Verify all manual isolation valves in each vent path are locked in the open position.

5. EMOV block valve closed during test.

Proposed Amendment No. 152 4-8a

i RANCHO SECO UllIT 1 TECHNICAL SPECIFICATIONS Surveillance Stanaarcs l

4.8 AUXILIARY FEEDWATER Pd4P PERIODIC TESTING l

_A_pp l i c aDil i ty Applies to the periodic testing of the turbine and motor driven auxiliory f eedwater pumps.

Obj ective To verify that the auxiliary feedwater pump and associated valves are operable.

Specification 4.8.1 Monthly on a staggered test basis at a tilne when the average reactor {

j coolant system temperature is >305"F, the turbine / motor driven and motor driven auxiliary feedwatir pumps shal'i be operated on !

recirculation to the condenser to verify proper operation, j Separate tests will be performed in order to verify the turuine driven capability and the motor driven capability of auxiliary feedwater pump P-318.

The monthly test frequency requirement shall be brought current within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after the average reactor coolant system temperature is >305"F for the motor driven purnps. The turbine driven capability shaTl be brought current within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of obtaining d percent reactor power, 152> Acceptable performance will be indicated if the pump starts ana operates for fifteen minutes at a flow rate 2ufficient to assure ~75 gpm of flow to the Steam Generator at a discharge pressure sufficient to drive that flow through the most restrictive flow path to a single steam g;.1erator which is at a pressure of 1060 psig. l The monthly testing of the auxiliary feedwater pumps and valves shall l be performed in accordance with the inservice inspection requirements of Specification 4.2.2.1. i

< 4.8.2 At least once per 18 months:

1. Verify that each automatic valve in the flow path actuates to its correct position upon receipt of eacn auxiliary fecawater a:tuation test signal.

2. Verify that each auxiliary feedwater pump starts as designed automatically upon receipt of each auxiliary feeawater actuaticn test signal.

152> 4.8.3 All auxiliary feeduater system valves, including those that are locked, sealed, or otherwise secured in position, are to ce inspected to verify tney are in the proper position following surveillance performed pursuant to Specifications 4.8.1, 4.8.2 and

4.8.4 Proposed Amendment flo.152 4-39

I RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4,8.4 Prior to startup following a refueling shutdown or any cold shutdown of longer than 30 days duration, conduct a test to demonstrate that the motor-driven AFW pumps can pump water from the CST to the steam generater.

152><

l Bases The monthly test frequency will be sufficient to verify that the turbine / motor driver and motor driven auxiliary feedwater pumps are operable. Verification of correct operation will be made both from the control room instrumentation and direct visual observation of the pumps.

The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 305'F from normai operating conditions in the event of a total loss of off-site power.

The electric driven auxiliary feedwater pumps are capable of delivering a 152> total feedwater flew of 475 gpm at a pressure of 1050 psig to the entrance of the steam generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 475 gpm to the entrance of the steam generators over the steam generator operating range of 800 psig to 1050 psig.

This capacity is utilized as analytical input to the Loss of Main Feedwater 4 Analysis which is the design basis event for AFW flow requirements.

l l

l Proposed Amendment No. 152 ,

-39a i

l

1 ENCLOSURE 3 l

Design Basis Report - ECN A5415 ).;

I l

i j

I l

r i

t l

i l i

1 i

.c ga w

v t e Letter Sequence Other
TAC:64359, (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance... Supplement Administration Meeting, Meeting Questions 4 December 1987 Answers Results Approval Other: ML20154G705, ML20154J794, ML20205H758, ML20210K842, ML20212D134, ML20212D154, ML20212D166, ML20236M353, ML20236M366, ML20236P351, NL-86-672, Forwards Rev 1 to A-5415, Design Basis Rept, as Requested. Review of Current Emergency Feedwater Initiation & Control Sys Design Should Address Inconsistencies w/1983 SER Not Technical Adequacy
MONTHYEAR1986-09-26 [Table View]

ML20236M353

Text

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