Text

Amend 93 to License DPR-54,revising Criteria Re Auxiliary Feedwater Sys & Adding Specific Requirements Associated W/ Emergency Feedwater Instrumentation & Control Sys
	ML20148C703
Person / Time
Site:	Rancho Seco
Issue date:	01/05/1988
From:	Knighton G; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20148C634	List: ML20148C630; ML20148C703; ML20148C732;
References
	NUDOCS 8801250241
	Download: ML20148C703 (42)
	v • d • e

4 @ C'ou f, fg UNITED STATES y g NUCLEAR REGULATORY COMMISSION g- e l WASHINGTON, D C 20535

%, ...../

SACRAMENTO MUNICIPAL UTILITY DISTRICT DOCKET N0. 50-312 RANCHO SEC0 NUC, LEAR GENERATING STATION AMENDMENT TO FACILITY OPERATINC LICENSE Amendment No. 93 License No. DPR-54

1. The Nuclear Regulatory Comission (the Comissicn) has found that:

A. The applicaticn for amendment by Sacramento Municipal Utility l District (the licensee) dated December 5, 1986, as supplemented March 26, July 31 and November 6,1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter I;

8. The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Comission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and j safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D. The issuance of this amendment will not be inimical to the common 4 defense and security or to the health and safety of the public; and )

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

8801250241 880105 PDR P ADOCK 05000312 PDR 1

-7

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-54 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 93, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. The changes in Technical Specifications are to become effective within 30 days of issuance of the amendment. In the period between issuance of the amendment and the effective date of the new Technical Specifications, the licensee shall adhere to the Technical Specifications ey.isting at the time. The period of time during change over shall be minimized.

4. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY C0?OllSSION

\

George W. Knighton, Director Project Directorate V Division of Reactor Projects - III, IV, Y and Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance: January 5, 1988

January 5,1988 ATTACHMENT TO LICENSE AMEN 0 MENT NO. 93 FACILITY OPERATING LICENSE NO. DPR-54 l DOCKET NO. 50-312 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area.of change.

Remove Insert iii *iii 1-2 1-2 1-7 1-7 3-1 3-1 3-2 3-2 3-2a 3-2a 3-15a **3-15a 3-23 3-23 3-23a I 3-24 3-24 l 3-24a 3-25 3-25 3-25a 3-26 3-26 3-26a 3-26a l 3-26b i 3-30a 3-30a 3-30b 3-30c 3-34 3-34 3-38a **3-38d 3-38b **3-38e 3-38c **3-38f 3-37d **3-38a 3-37e **3-38b 3-38f **3-38c 3-389 3-38h 4-7b 4-70 4-7c 4-7c 4-7d 1 4-7e 4-8 4-8 4-8a 4-8a 4-29 **4-29 4-39 4-39 I 4-39a 4-39a Pages issued to revise content and correct previously issued amendments.

Pages issued to correct previously issued amendments.

TECHNICAL $PECIFICATIONS TABLE OF CONTENTS (Continued)

Sec tion Page 3.1.6 Leakage 3-12 3.1.7 ;4oderator Temperature Coefficient of Radioactivity 3-15 3.1.8 Low Power Physics Testing Restrictions 3-ISb 3.1.9 Control Rod Operation 3-16 3.2 HIGH PRESSURE INJECTION, CHE4! CAL ADDITION AND LCW .

TEiPERATURE OVERPRESSURE PROTECTION (LioP) SYSTLl5 3-17 3.3 E4ERGENCY CORE COOLING, REACTOR BUILDING E4ERGENCY COOLING, AND REACTOR BUILDING SPRAY 5Y5iE45 3-19 3.4 STEAi AND POWER CONVERSION SYSTE4 3-23 I 3.5 INSTRU4ENTATION SYSTDiS 3-25 3.5.1 Operational Safety Instrumentation 3-25 1 3.5.2 Control Rod Group and Power Distribution Limits 3-31 3.5.3 Safety Features Actuation System Setooints 3-34 3.5.4 Incore Instrumentation S-36 l 3.5.5 Accident ;4cnitoring Instrumentation 3-38d 3.5.6 Emergency Feedwater Initiation and Control Setooints 3-389 3.6 '

REACTOR BUILDING 3-39 1

3.7 AUXILIARY ELECTRICAL SYST54S 3 41 3.8 FUEL LOADING AND REFUELING 3-44 3.9 SPENT FUEL POOL 3-46a l 3.10 SECONDARY SYSTEM ACTIVITY 3-47 3.11 REACTOR BUILDING POLAR CRANE AND AUXILIARY HOIST 3-49 3.12 SHOCK SUPPRESSORS (SNUBBERS) 3-51 3.13 AIR FILTER SYSTE45 3-52 3.14 FIRE SUPPRESSION 3-53 3.14.1 Instrumeid: tion 3-53 3.14.2 Water System 3-53 3.14.3 Spray and Sprinkler Systems 3-56 3.14.4 CO2 System 3-56 Amendment No. 7$,39,gg,yg,pp,9a (ii - -

i '

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions A refueling shutdown 140 F. Pressure is defined by Specification 3.1.2.

refers to a shutdown to replace or rearrange all or a portion of the fuel assemblies and/or control rods.

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.

l.2.8 Refueling Interval

18 nonths.

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.

1.2.10 Remain Critical A technical specification that requires that the reactor shall not remain critical shall mean that an uninterrupted nomal hot shutdown procedure will be completed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months unless othemise specified. l 1.2.11 Tg At operating conditions Tavg is defined as the arithmetic average of the coolant temperatures in the hot and cold leg: of the loop with the greater ,

number of reactor coolant pumps operating, if such a distinction of loops can l be made.

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

1.3 OPERABLE _

A component or system is operable when it is The capable of perfoming component its shall be or system intended function within the required range.(1) it satisfies the limiting considered to have this capability when:

conditions for operation defined in Specification 3, (2) it has been tested periodically in accordance with Specification 4, and has met its perfomance ;

requirements, (3) the system has available its nomal and energency sourcl power, and (4) its required auxiliaries are capable of perfoming their I When a system or component is determined to be inoperable intended function. l solely because its nonnal power source is inoperable or its emergency power source ,is inoperable, it may be considered OPERAB provided its redundant system or component is OPERABLE with an OPE and energency power source.

See page 1-2b -

Amendment No. JE,0,U,93 1-2

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

An 0FFSITE DOSE CALCULATION MANUAL (ODCH) shall be a manual containing the methodology and parameters to be used in the l calculation of offsite dose due to raditoactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alann/ trip setpoints and specific details of the environmental radiological monitoring program.

I 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 BOUNDARY The boundary of the SMUD owned property.

1.18 00SE EQUIVALENT I-131 The DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce tne same thyroid dose as the quantity and isotopic mixture of I-131,1-132, 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 MEMBER (S) 0F THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all individuals who by virtue of their occupational status have no fonnal 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.

1.20 VECTOR LOGIC A set of circuitry in each channel of the EFIC system which once AFM has been initiated determines whether AFW to a steam generator should be allowed or teminated and the signal output for each EFIC channel to the AFW valves associated with that channel.

Amendment No. E3,93 1-7 .

RANCHO SECO UNIT 1 TECHNTCAL $PECIF! CATIONS Limiting Conditions for Operation

3. LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR COOLANT SYSTEM Applicabili ty Applies to the operating status of the reactor ccolant 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 Specification l

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 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. l 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 A. Two steam generators shall be operable whenever the reactor 8

coolant average temperature is above 280 F, except as described in 3.1.1.2.2.

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 generator (s) inoperable due to steam generator defective tube (s), restore the inoperable generator (s) l to operable status prior to increasing reactor coolant average .

temperature above 200*F. l 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 hyr' rstatic tests in accordance with ASME Boiler ano Pressure vessel Code,Section III.

Amendment No. f ,3J ,7J ,g/ ,9 3 3-1

I RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS L"miting Conditions for Operatior, 3.1.1.4 Pressurizer Electromatic Relief Val .e A. The nominal setpoint of the pressurizer electromatic relief valve shall be 2450 psig

10 psig except when required for cold overpressure protection. !

3.1.1.5 Decay Heat Removal A. At least two of the coolant loops listed below shall be i 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, i

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

3. Decay Heat Removal Loop ( A) l

4. Decay Heat Removal Loop (B)

With less than the above required coolant loops OPERABLE, imediately initiate corrective action to return the required '

coolant loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

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 I during power operation. ,

1 C. With one of the above reactor coolant system vent paths l inoperable, STARTUP and/or POWER OPERATION may continue ;

provided the inoperable vent path is maintained closed with I power removed from the valve actuator of all the valves in the inoperable vent path; restore the f.noperable vent path to OPERABLE status within 30 days. If the status is not restorea 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 STANOBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Amendment No. $ .M ,D ,80,93 3-2

RANCHO SECO UNii 1 l

TECHNZCAL SPECIFICATIONS l

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

The decay heat removal system su'ction piping is designed for 300 F and 300 l psig; thus, the system can remove decay heat when the reactor coolant system !

is below this temperature. (2) (3)

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 hect. (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 overpressure for rod withdrawal accidents. (5) The Pressurizer code safety valve lif t 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 above the set pressure.

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

Two pump operation is limited until further ECCS analysis is performed.

When the re' actor is not critical but TAV is above 280' F, one steam generator provides sufficient heat removal capability for removing decay heat. However, single sfailure considerations require that both steam generators be operable.

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

The purpose of the high point vents is to vent noncondensible gases f rom the 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 patn has been removed.

REFERENCES (1) USAR Tables 9. 5-2, 4. 2-1, 4. 2- 2, 4. 2-4, 4. 2- 5, 4. 2- 6 (2) USAR paragraph 9.5.2.2 and 10.2.2 (3) USAR paragraph 4.2.5 (4) USAR paragraph 4.3.8.4 and 4.2.4 -

(5) USAR paragraph 4.3.6 and 14.1.2.2.3 Amendment No. 71,80,87,93 3-2a

.. :.s . s ., . g... . :

,.3 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.7 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY (co as power increases. Adj us t p<uer to the coefficient the moderator coefficient at 15 percent at any power lhvel above 15 percent.

5.

Dissolved boron concentration - This correction is for any Since the moderator coefficient is more positi I amounts of dissolved boron, the sign of the correction depends on whether boron is added or removed.

6.

Control rod insertion - This correction is for the differenc In moderator coefficients between an unrodded and ro 7.

Isothermal to distributed temperatures - The correction for found to distributed spatially moderator temperature effects has been be insignificant. Therefore, correction for l distributed ef fects is not required. l REFERENCES t

(1)

USAR, subsections 14.1 and 14.2 (2) USAR, paragraph 3.2.2.1.5.0 Amendment No. 87,93 3-15a

q RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.4 STEAM AND POWER CONVERS!CN SYSTEM Applicability Applies to the operability of the turbine cycle during normal operation anc for the removal of decay heat.

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

I Specification 3.4.1 The reactor coolant system shall not be brought or remain above 280f l with irradf ated fuel in the pressure vessel unless the following (

conditions are met:

A. Capattility te remove decay heat by use of two steam generator' as speci fied in 3.1.1.2. A.

1 B. One atmospheric dump valve per steam generator snall be l operabl e . !

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

' Two main steam system safety valves are operable per steam D.

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. l G. Four independent backup instrument air bottle supply systems for ADVs and MFW, 5FW, and AFW valves are operable.

With less than the above required components operdble, be on decay l heat cooling within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

/mendment No. 93 3-23

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.4.2 The reactor shall not be brought or remain critical unless the following conditions are met:

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

B. One atmospheric dump valve per steam generator shall be operable except that: (1) with only one atmospheric dump valve cperable, restore an inoperable valve for the other steam 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 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 no atmospheric dump valves operable, restore at least one 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 the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and on decay heat cooling within tne next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. A minimum of 250,000 gallons of water shall be available in the condensate storage tank except that with less than the minimum volume, restore the minimum volume witnin 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 cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D. Seventeen of the eighteen main steam safety valves are I operable except that with less than the minimum number of valves, restore the incperable 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 cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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 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 .

F. Both auxiliary feedwater trains (i.e. , pump and their flow path) are operable except that:

(1) With one auxiliary feedwater 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 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 .

Amendment No. 9 3 3-23a

i l

RANCHO SECO UNIT 1 !

TECHN: CAL SPECIFICATIONS l Limiting Conditions for Operatwo 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 het 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 I the reactor shall be on decay heat cooling within the j next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . l H. Two independent backup instrument air bottle supply systems i (one per steamline) for ADVs are operable except that: ;

l (1) With one system inoperable, restore the system to l operable status within 7 days or be in hot shutdown l within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and nn 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).

l I. Two independent backup instrument air bottle supply systems l (one per feed water line) for MFW, SFW, and AFW control valve- l are operable except that with either one or both system (si j inoperable, restore the inoperable system (s) within 7 ciays 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 decdy heat cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I Bases The feedwater system and the turbine bypass system are normally used for decay heat removal and cooldown above 280 F. Main feedwater is supplieo by operation of a condensate pump and main feedwater pump. If neither main feed 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 of main feedwater to match decay heat plus Reactor Coolant Pump heat input to the Re occur. gor Coolant System before solid pressurizer operation could Amendment No. 31,P/ ,9 3 3-24

~

f..

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The 250,000 gallons of water in the condensate storage tank is sufficient to remove decay heat (plus Reactor Coolant pump heat for two pumps) for 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 at a cooldown rate of 50'F/hr (1).

The minimum relie{2japacity 13,329,163 lb/hr.

of seventeen steam system safety valves is This is sufficient capacity to protec system under the design cverpower condition of 112 percent.{3Jhe steam Both trains of main feed ater isolation on each main feedwater line are !

required to be operable. Train A of main feedwater isolation is comprised of main feedwater control valves, main feedwater block valves and startup control valves. Train 8 of main feedwater isolation is comprised of the main j feedwater isolation valves, j Four independent Class 1 backup air supply systems are provioed 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 Feeawater (SFW) and AFW control valves feeding the "A" 0TSG; another system supplies power for same valves feeding the "B" OTSG. Two systems supply power for ADVs with one for the ADYs on the "A" main steam line and one for the ADVs on the "B" main steam line. Each system is sized to provide at least two hours of air supply.

REFERENCES i (1) > B and W Document 32-1141727-00, "Heat Removal Capability of SMUD CST," March 1984 (2) USAR paragraph 10.3.4 (3) USAR Appendix 3A, Answer to Question 3A.5 (4) B and W Calculation 86-1167930, "Rancho Seco: AFW Minimum Flow Analysis," (SMUD Calculaticn No. Z-FWS 101S0) e1 Amendmen t No. 7,J ,P,7,9 3 3-24a j

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5 IfiSJR_UMENTATION SYSTEMS 3.5.1 OPERATIONAL SAFETY INSTRUMENTATION A

J plinbility Applies to unit instrementation and control ' systems.

Qbjective To delineate the conditions of the unit instrumentation _ and safety circuits necessary to assure reactor safety.

Specifications 3.5.1.1 Startup and operation are not permitted unless the requirements of Table 3.5.1-1, Columns A and B are met. l 3.5.1.2 In the event the number of protection or EFIC System channels l l operable falls below the limit given under Table 3.5.1-1, Columns A l I

and B, operation shall be limited as specified in Column C.

In the event the number of operable Process Instrumentation l channels is less than the Total Number of Channel (s), restore the 1 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 operable channels is 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 i operable, the reactor shall be made subcritical within four 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 .

3.5.1.3 For on-line testing or in the event of a protection instrument or channel failure, a key operated channel bypass switch associated with each reactor protection channel will be used to lock the 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 time.

3.5.1.4 The key operated shutdown bypass switch associated with each

. reactor protection channel shall not be used during reactor power operation. l 3 3.5.1.5 During startup when the intermediate range instrument comes on scale, the overlap between the intermediate range and the source 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.

1 Amendment No. Al,E/,93 3-25

m. . ~ _ . . _ . - _ _ _ _ - _ _ - . . _ _ _ . - _ . _ _ _ . _ _ _ _ . - _

l l

l RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS l Limiting Conditions for Operation I 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 tne 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 condition within an additional four hours.

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 bypass" at any one time. j i

3.5.1.8 If a channel of the RPS is in bypass, it is pemissiele 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 pemitted unless three power range neutron instrument channels and two channels each of the following are operable:

four reactor coolant temperature instrument channels, four reactor coolant i flow instrument channels, four reactor coolant pressure instrument channels, 'l 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 3.5.1-1 must be operable. .

Operation at rated power is pemitted as long as the systems have at least the redundancy requirements of Column B (Table 3.5.1-1J. This is in j agreement with redundancy and single failure criteria of IEEE 279 as ,

described in FSAR section 7. I

. 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 alam and lights to indicate when that channel is bypas sed.

l Amendment No. 27,87,93 3-25a l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATI0h5 Limiting Conditions for Operation Bases (Continued)

Each reactor orotection channel key operated shutdown bypass switch is provided with alam and lights to indicate when 1.he shutdown bypass switch is being used. There are four shatdown 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. Nomal trip logic is two out of four. Required trip logic for the power range instrumentation channels is 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 cut of two.

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 ',oss 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 I

If one SG is below 600 psig, supply AFW to the other SG If both SGs are below 600 psig but the pressure dif ference between the two SGs exceeds 100 psig, supply AFW only to the SG with the 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 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 been automatically reset: 1) Loss of 4 RC pumps, 2) low steam generator pressure, 3) low steam generator level.

Amendment No. 37,93 3-26

1 RANCHO SECO UNIT 1 1 TECHNICAL SPECIFICATIONS l

Limiting Conditions for Operation l 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 interlocking features allow the EFIC system to take 1 a single failure in addition to having one channel in maintenance bypass.

Various RPS test features cc inhibit initiate signals to the EFIC system and degrade the EFIC system below acceptable limits if tne RPS channel is not in bypass. Therefore, no testing should be perfomed on a RPS instrument string which supplies an output to EFIC without placing that RPS channel in bypass.

The EFIC system is designed to allow testing during power cperation. The EFIC system can be tested from its input teminals to the actuated device i controllers without placing the channel in key locked "maintenance bypass.' l 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 actuation of the ccmponent. l I

Each EFIC channel key operated maintenance bypass switch is provided with 1 alann and lights to indicate when the maintenance bypass switen is being usee. l 1

The source range and intermediate range nuclear flux instrumentatico scales i overlap by one decade. This decade overlap will be achieved at 10-20 amps on the intermediate range scale.

Power is normally supplied to the control rod drive mechanisms from two i separate parallel 480 volt sources. Redundant trip devices are employed in ;

each of these soun:es. If any one of these trip devices fails in the I untripped state on-line repairs to the failed device, when practical, will be l made, and the remaining trip devices will be tested. Eight hours is ample time to test the remaining trip devices and in many cases make on-line repai rs.

l l

Amendment No. 31,93 3-26a

-l l

l

e l

RANCH 0 SEC0 UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions f or Operation The OPERABILITY of the SFAS instrumentation systems and bypasses ensure that

1) the associated SFAS action will be initiated when the parameter monitoreo 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 be out of service for testing or maintenance, and 4) sufficient system functional capability is available for SFAS purposes from diverse parameters.

The OPERABILITY of these systems .is required to provide the overall j reliability, redundancy, and diversity assumed available in tr,e facility I design for the protection and mitigation of accident and transient l conditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.

The OPERASILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and followinc an accident. This capability is consistent with the recommendations of Regulatory Guioe 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident", December 1975 and NUREG-0576, "THI-2 Lessons Learned Task Force Status Report and Short-Term Rec ommenda tions. "

l REFERENCE I

USAR, Subsection 7.1 l

1 e

l l

Amendment No. 37,RT ,9 3 3-26b l

PANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Table 3.5.1-1 (Continued) Limiting Conditions for Operation

,h , , INSTRUMENTS OPERATING CCNDIT!GNS

s m

(( (C)

(A) (B) Operator Action if j,' Functional Unit Total Nu.ber m of Minimum Channels Conditions of Columns A

Chinnels Operable and B Cannot be Met o -

gg 4 Ceactor Building Purge Isolation 2 1 Operation may continue provided the purge sw on high radiation inlet and outlet valves of the incoerable q, channel (s) are closed and their respective to breakers de-energized or comply with 4, 3.5.1.2. At cold shutdown or refueling.

ra each of the purge inlet and cutlet valves

g will be closed.

w Erergancy_fead->ter Initiation Lic qn t rol ( EEICLSy s t e, 1 A r=' I n i t i a t i e n

a. Panual 2 t",te 1) 2 (Nete 1) See Actions 3 and 4

b. Lcw level. SGA or 8 (N:te 2) 4/SG (Note 1) 3/SG See Actions 1 2 and 3. May be typassed below 750 psig OTSG pressure,

c. Lo Pressure SGA or B 4/SG (Note 1) 3/SG See Actions 1 2 and 3. May be bypasse-d oa below 750 psig OTSG pressure.

w c) d. Los s of PFW An ti c i p3-

tory deactor Trip 4 (Note 1) 3 See Actions 1 2 and 3. Loss of MFW Anticipatory Reactor Trip is ef f ectively bypassed in RPS below 20 percent pcwer.

e. Loss of 4 RC Pumps 4 (Note 1) 3 See Actions 1, 2 and 3. May be bypassed bel ow 750 psig OTSG pressure.

f. Automitic Trio Logic 2 (Note 1) 2 (Note 1) See Actions 3 and 4 2 SG-A Main Feed-ater IsolatiCn

a. Manual 2 (Ncte 1) 2 (Note 1) See Actions 3 and 4

b. Le- SGA Pressu e <' te 3) 4 (Note 1) 3 See Actions 1 2 and 3. May be typissed below 750 psig OTSG pressure.

c. Automatic Trio Lop 2 (Note 1) 2 (Note 1) See Acticns 3 and 4 Nate i for channel testing, calibration, or Paintenance the Iotal Number of Channels and/or the Minimum Channels Operable may be reduced by one for a ravirum of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months providing the remaining channels are OPEPABLE.

Note 2 Low level AFW Initittion has a ma=imum of a 10.0 second delay.

Note 3 Low pressure AFW Initiation has a maximum of a 3.0 second delay.

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 2p Limiting Conditions for Operation n3 Table 3.5.1-1 (Continued)

s C[ INSTRUMENTS OPERATING CONDITIONS en 2'

' (C)

(A) (B) Operator Action if

((

Functional Unit Total Number of Chan,els Minirum Channels Operable

onditions cf Colu n; A and B Cannot be Met e

w

3. SG-8 Main Feed =ater Isolation

a. Manual 2 (Note 1) 2 (Note 1) See Actions 3 and 4

b. Low SGB Pressure (Note 3) 4 (Note 1) 3 See Actions 1, 2 and 3. May be bypassed below 750 psig OTSG pressure.

c. Automatic Trip Logic 2 (Note 1) -2 (Note 1) See Actions 3 and 4 4 AFV Valve Commands (vector)

4. Vector Enable 2 (Note 1) 2 (Note 1) See Actions 3 and 4.

b. Vector Module (Note 4) 4 (Note 1) 3 See Actions 1 and S.

c. Control Enable 2 (Note 1) 2 (Note 1) See Actions 1 and 3.

d. Control Module 2 (Note 1) 2 (Note 1) See Actions I and 3.

w a

w o

cr Note 1 For channel testing, calibration, or maintenance the Total Number of Channels and/or the Minimum Channels Operable may be reduced by one for a m3=imum of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months providing the remaining channe's are OPERABLE.

Note 3 Low pressure AFV Initiation has a mtwimum of a 3.0 second delay.

Note 4 SG Pressure Dif f erence AFW valve Command (Vector) has a mavimum of a 10.0 second delay.

I,

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Table 3.5.1-1 (Continued)

INSTRUMENTS OPERATING CONDITIONS Action 1 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE 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 .

Action 2 - Hith the number of OPERABLE channels one less than the Minimum Channels Operable then put one of the inoperable channels in trip, and restore at least one of the inoperable channels. to 1 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 .

Action 3 - With the number of OPERABLE channels two less than the Minimum Channels Operable, be in at least hot shutdown within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and in cold shutdown within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Action 4 - Hith the number of OPERABLE channels one less than the Total Number of Channels OPERABLE, restore the inoperable channel 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 .

Action 5 - With the number of OPERABLE channels one less than the Minimum Channels Operable, restore one inoperable channel to OPERABLE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 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 .

Amendment No. 9 3 3-30c

-l l

l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS I J

Limiting Conditions for Operation j 3.5.3 SAFETY FEATURES ACTUATION SYSTEM SETPOINTS Applicability This specification applies to the safety features actuation system actuation setpoints.

Objective 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 per::lissible bypasses shall be as follows:

i Functional Unit Ac tion 5c t uoin t !

High Reactor Building Reactor Building spray valves *" 130 ps i9 p re s su re*

Reactor Building spray pumps *" 130 p sig High pressure injection, and start of Reactor Building cooling and e Reactor Building isolation. 14 psig Low pressure injection,EFIC AFW initiate 14 p s ig l Low reactor coolant system High pressure injection, and start pressure" of Reactor Building cooling and Reactor Building Isolation 11600 psig Low pressure injection,EFIC AFW initiatel160:. psio l

Automatic Actuation Logic All above Not Appi'catie

. Manual Ai1 above Nut Appi t cacie

May be bypassed during Reactor Building leak rate test. I "May be bypassed below 1850 psig and is automatically reinstatec above 1850 psig

"Five-minute time delay.

Amendment No. 3J ,93 3-34

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Figure 3.5.4-1 Incore Instrumentation Specification Axial Imbalance Indication M

- ~ , ticx R A0ut 7 - - - N -

sva:ETRT f 7 R~

<r,

(

\, /

'N "

/ / TOP ulAl CORE Half N

, /

N- -

', ~

'm s i o y E I / \ ail At PL AAE

,5

( \

x_ _-

/

f \ ~,

3 2 f- -- -- }

/ - N

/ %

[O / \ \

E *

(

x / j x -

N'. -

80TT0s ulit CORE H Alf

~~s f

/ N

/ \

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATI0MS Limiting Conditions for Operation.

Figure 3.5.4-2 Incore Instrumentation Specification Racial Flux Tilt Indication

( NOI AL /,

_ _ _ _ _ -. - - - - - L 'B 109-I i

i

% i M

R ADI AL SYutETRY g -

IN THIS PL ANE

\

x\- 4 /M/ - -

M 5

s.

/ /

$a / \

s 5e 1 /

N g / /

! N _._' ' /

E 1

~

, i

~~

$ /

RA0lAL SYuuETRY l g -- -

j IN THI$ PL ANE

~ - -

)A '

/ %

/ \

w Amendment Nn. &/ 93 3 38b

I RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Figure 3.5.4-3 Incore Instrumenta tion Sceci fication M

/"F 3

/rQW ik \

c

.M N N bd -

r/ - <

/ -

qu N j

(( I] ' Os N 7

3 M g 4N8 j

! Q --

/

5M W Y t

$7I+s M \

O gg /

N #

$N 1

% A06

- - -$ 01 m /

/

/ \,

Amendment No. 87 93 3 -38c

_ ~ _

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.5 ACCIDENT MONITORING INSTRUMENTATION Accident monitoring instrumentation channels shown in Taole 3.5 5-1 shall be OPERABLE with their alarm / trip setpoints as shown.

Acolicability As shown in Table 3.5.5-1.

Action A. With an accident monitoring instrument channel less conservative than the setpoints provided in Table 3.5.5-1, declare the channel inoperable.

B. With less than the minimum nunder of operable channels, take tne ACTION shown in Table 3.5.5-1.

Bases Table 3.5.5-1 lists the operability requirements for the various types of accident monitoring instrumentation that were installed in response to NUREG 0737, items II.F.1 and II.F.2. This new set of equipment meets or exceeds the amount of coverage outlined in Generic Letter No. 83-37, "NUREG-0737 Technical Sp ecifica tions. " Most of the instrument parameters in Table 3.5.5-1 are monitored by redundant equipment. However, a failure of any one of the radiation monitors described in items 1, 6, and 7 would place that item in an LC0 position and would require action number I. If the inoperable channel cannot be repaired within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Action Statement Number I requires that a pre-planned alternate method of monitoring be initiated. Operating procedures will be used to control the use of backup radiation equipment.

l l

Amendment No. S9,9 3 3 38d

. P.ANCHO SECO INT i TECHNICAL SPECIFICATIONS Limiting Conditions for Operation TABLE 3.5.5-1 ACCIDENT MONITORING INSTRUMENTATION OPERABILITY REQUIREMENTSII) l iotal Number M afmum Number A l a m/ i rl p of Channels of Channels Setpoint Action Instrument Ooerable

1. Containment Area 2 2 12 rad /hr I High Range Radiation Monitor

2. Wide Range Con- 2 1 N/A !!

tainment Water (Range Level 0-10 ft)

3. Containment 2 1 <4 Percent II Hydrogen Analyzer II2Conc

4. Emergency Stoo 2 ] $4 ft. (High III Leve. Ala ns or Ccmputer)

5. Containment Wide 2 1 N/A 11 Range Pressure (Range -5 to Monitor /Recordce 180 psig)

6. High Range Noble N/AI2) 1 Gas Effluent Monitors (Ragge10-7 g a) RB Exhaust StackI3) 1 1 f b) Aux Building Stack 1 1 l c) Radwaste Venti 4) 1 1

7. Main Steam Lines 2 2 <10 mr/hr 1 Radiation Monitors --

8. Subcooling Margin 2 1 No alams. !!

Monitor Procedural controls in ;

place i

9. Incore Themocouples 4/ core 2/ core (Range 200- !!!

quadrant quadrant 2300 F)

(1)

This Table applies at all times except during cold shutdown or refueling.

(2) Alarm limits are set accordiag to the Offsite Oose Calculation Hanual.

(3)

Monitoring of the RB Exhaust Stack is not required when the purge and/or equalizing valves are closed.

- I4I Monitoring of the Radwaste Vent is not required when the unit is not operating.

A,endment No. pp,9 3 3-36e !i n

RANCHO SECO Unli 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Table 3.5.5-1 (Continued)

Action I. With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, either restore the inoperable Channel (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or:

1) Initiate the pre-planned alternate method of monitoring, and

2) Prepare and submit a Special Report to the Commission pursuant to Specifica tion 6.9.5.0, within 30 days followina the event.

outlining the action taken, the cause of the Inoperability, and the corrective action and schedule for implementation.

II. a. With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels, restore the inoperable channel (s) to OPERABLE status within 30 days, or be ,

in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. Vith the number of OPERABLE accident monitoring instrueentation ,

channels less than the Minimum Wumber of Channels Operable. i restore the inoperable channel (s) to OPERABLE status within 7 days or be in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

III. With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Number of Channels Operable, restore ,

the inoperable channel (s) tv OPERABLE status within 30 days, or be l in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , l Amendment No, BCf ,9 3 3-38 f 6

RANCHO SECO UNZT 1 TECHNICAL SPECIFICATIONS Limiting Condition for Operation 3.5.6 EdERGENCY FEEDWATER INITIATION AND CONTROL SETPOINTS Applicability This specification applies to the emergency feedwater initiation and control (EFIC) setpoints.

Objective '

To provide for automatic initiation and control of auxiliary feeawater and l automatic isolation of main feedwater, j

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

Functional Unit Action Setpoint

a. Low SG Level Initiates AFW >9 inches

b. Low SG Pressure Inittetes 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 HFW

f. Vector Logic Isolates Faulted SG Yarious (3)

g. Shutdown Bypass Bypass Pemissive <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 Amendment No. 93 3-38 9 l

RAhCHO SECO UNIT 1 TECHil! CAL SPECIFICATIONS Limiting Condition for Operation 3.5.6 (continued)

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

1. all four RC pumps are tripped, or

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 stays below a minimum set point of 575 psig.

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

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

- If one SG is below 600 psig, supply AFW to the other SG If both SGs are beluw 600 psig but the pressure difference between the two SGs exceeds 100 psig, supply AFW only to the SG with the 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 initiation 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 perTnissive logic within the RPS and SFAS, respectively. ,

l l

l Amendment No. 9 3 3-38h

RANCHO SECO UNIT I

> TEC1011 CAL SPECIFICATIONS

,@ Survefilance Standards s ~ ' Table 4.1-1 (Continued)

@ INSTRUMENT SURVEILLANCE REQUIREMENTS ri 5 Channel Description Check Test Calibrate Remarks 4z. Reactor Bulldtsig drain

2

_,o accumulation tank level NA NA R 7 43. Incore neutrots detectors Mll) NA NA (1) Check functiontng, including w functioning of computer readout 7 and/or recorder readout.

m

? 44. a. Process and area radt-

@ ation monitoring system W M Q 9 b. C'ontainment Area Monttors W NA , R

45. Emergency plant radiation Instruments Mll) NA R (1) Battery check

46. Environmental air monitors Mll) NA R (1) Check functioning

47. Strong motion accelerometer Q(1) NA R (1) Battery check

48. Deleted ,

49. Pressurf rer Water Level , M NA R E 50. Auxfilary feedwater Flow Rate M NA R

51. Reactor Coolant Systen Sub-cooling Margin Monitor M NA ', R

52. Ett0V Power Position Indicator (Primary Detector) M NA R e

53. EMOV Pcattfon Indicator fBackup Detector) H !!A R T/C or Acoustic 54 Ot0V 81ock valve Position Indicator M NA R

55. Safety Valve Position in-dicator (Primary Detector) H NA R T/C l

h a RAtJCHO SECO UNIT I j' TECilNICAL $PECIFICAY10N5 s Survefilance Standards

@ TACLE 4.1 1 (Continued) c'

[NSTRUMENT SURVEILLANCE REQUINEMENTS v: -

'[ Channel Descriptf on Check Test Calibrate Remarks

@$ 56. Safety Yalve Position In-

- dicator (8ackup Detector) 5$ Acoustic M N/A R

2$ 57 Yaltage Protection 5(1) (1) Compare voltmeter readings "o a. Undervoltage M R La b. Overvoltage H R

c. Time Delay M R

58. Containment Area High 5 M(1, R (1) Test using installed source Range Monitor -

59. Wide Range Containment M N/A R Water Level

,, 60. Containment Hydrogen 5 M Q Analyzer E

61. Emergency Susp Level M N/A R

62. Containment Vide range M N/A R Pressure Monitor / Recorder

63. High Range lloble Gas 5 M R Effluent Monitors

- AB Ext.aust Stack

- Aux. Buf1 ding Stack ..

- Radwaste Vent 64 Main Steam Line Radiation 5 M(1) R (1) Test using installed source Monitors 65 Subcooling Margin Monitors M W/A R

66. Incore Thermocouples M N/A R

67. Low Temperature Over. N/A (1) R (1) Prf or to cooldown.

Pressure Protection (EH0Y)

RAIJCl:0 SECO UNIT 1 TECHNICAL SPECIFICAT10l15 A Surveillance Standards TABLE 4.1-1 (Continued) l o IllSTktitLNT SURVEILLAllCE HLQUIHEAENTS 5 -

e Channel Description Clieck Test Cal ibra te Remarks 68 AFW Initiation

a. Aanual IJ/A M N/A

b. Low Level SGA or B S M (1) R (11 (1) Include time delay module.

c. Low Pressure SGA or B S M (1) H (1) (1) Include time delay module.

d. Loss of AFW Anticipa-tory Reactor Trip S e4 N/A

e. Loss of 4 kC Pumps S si N/A

f. SFAS Actuation S R N/A

~

, , 9 Automatic Trip logic S M N/A

- o.

4 li. Bypasses S M R 69 SGA Main Feedwater Line Isolation

a. i4anual 11/A M N/A

b. Automatic Trip Logic 5 H tJ/A

70. SG8 14ain feedwater Line Isolation

a. Itanual N/A M N/A

b. Automatic Trip Logic S M IJ/A 4

i

._ _ _. _ ~ . .. = = _ _ _ _ __.-

k NAIKHO 5tCO UNIT 1

@ TECHIJICAL SPCCIFICAlf Ui45 I. Surveillance Standerds en TA8L[ 4.1-1 (Continued)

S INSTWui[NT SURYCILLANCE 1.EQUlHE.l[NTS z ..

?

, Channel Description Chect Test Calibrate Nea.a rk s sn

71. Afd val ve Comands (Vec tors

a. Vector [nable 5 M II/A

b. SGA Pressure Lou 5 N R

c. SCB Pressure Low 5 M R

d. SG Pressure Dif ference SGA Pressure > 5 M (13 R (1) (1) Ir.clude tiene delay r.iodule.

SCd Pressure SCB Pressure >

SCA Pressure 5 M (1) R (11 (1) Include time delay nodule.

72. AfW Control valve Cor. trol

a. fianual/ Auto N/A M N/A in Aanual

73. SG Level control

a. Setpois.t Selection N/A N N/A

b. Control [nable N/A 44 N/A o

C. Module Response N/A M (1) R (1) Confiru External Controller Settings 74 ADV Control valve Control

a. ilanual/ Auto N/A M N/A in 8t anual

75. SG Pressure Contrcl

a. Module Desponse N/A M (1) R (I) Confirm External Lontroller Settings 7 f, . tfackup instrurent Air Supply Systen

a. Pressure O N/A w/A Table Notations 5 . Eacts shif t M . itun tl.ly P . Prior to eaci. startup if teot d5ne previous weel D Daily 0 . Quarterly R . O..cc durf s.J the refucilng intervcl u . Ucekly SY . Sciciane.ua l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-2 HINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency

1. Control rods Rod drop times of Each refueling snutdown all full length rods

2. Control rod movement Hovement of each rod Every two weeks

3. Pressurizer Setpoint Note 3 code safety valves 4 Main Steam safety Setpoint Note 3 valves

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

6. Turbine throttle Hovement of each valve Monthly stop valves

7. Reactor Coolant Leakage Calculatea inventory weekly System Leakage check daily

8. Charcoal and high Charcoal and HEPA filter Each refueling interval and efficiency filters for iodine and particul- 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

10. Reactor Building Functional Each refueling interval ,

isolation trip I

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

l

12. Turbine Overspeed Calibration Each refueling interval Trips

13. Internals Yent Manual Actuation, III Each refueling interval Yalves Remot tion,2{isualinspec-and verify that valve not stuck open.

1 l

Amendment No. 7,16,76,80,82,93 4-8

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE 4.1-2 (Continued)

MINIMUN EQUIPMENT TEST FREQUENCY Item Test Frecuency runctional 1 *, . neactor Coolant st of Each refueling interval System High Point each valve (

Vents

15. Low Temperature Functional (5) Prior to RCS temperature Overpressure decreasing below 350"F Protection (EMOV)

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 .

Control Valves

d. Main Feedwater Ftnctional Eacn refueling interval .

Control Valves

17. Tur;bine Throttle Cycle Each refueling 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).

2. Check visually accessible surfaces to evaluate observec surface i rregul arities.

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

Amendment No. 76,EO,82,9 3 4-84

P RANCHO SECO UNIT l TECHNICAL SPECIFICATIONS Surveillance Standards 4.5.2 REACTOR BUILDING COOLING SYSTEMS Apolicability <

Applies to testing of the Reactor Building cooling systems.

Objective

To verify that the Reactor Building cooling systems are operable.

Specification 4.5.2.1 System Tests A. Reactor Building Spray System

1. During each refueling interval a system test shall be conducted to demonstrate proper operation of the system.

A manual trip signal will be applied to demonstrate actuation of the Reactcr Building spray system (except for Reactor Building motor-operated inlet valves which prevent water entering nozzles). Hater will be circulated from the borated water storage tank through the Reactor Building spray pumps and returned through the test line to the borated water storage tank.

2. The test will be considered satisfactory if visual observation and control board indication verifies that all components have responded to the actuation signal and the appropriate pump breakers shall have opened and closed, and all valves shall have completed their travel except the blocked Reactor Building inlet valve.

3. Air will be introduced into the spray headers to verify the availability of the headers and spray nozzle at least every 10 years.

B. Reactor Building Eme'rgency Cooling System

1. During each refueling interval, a system test shall be conducted to demonstrate proper operation of the system.

The test shall be performed in accordance with the procedure summarized below:

A manual trip signal will be applied to actuate the Reactor Building emergency cooling system for Reactor Building cooling operation.

Amendment No. 76,89,93 4-29 ,

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATI0liS Surveillance Stanoarcs 4.8 AUXILIARY FEEDWATER PUIP PERIODIC TESTING Applicability Applies to the periodic testing of the turbine and motor driven auxiliary feedwater pumps.

Obj ec ti ve To verify that the auxiliary feeawater pump and associated valves are operable.

Soecification 4.8.1 Monthly on a staggered test basis at a time when the average reactor coolant system temperature is >305*F, the turbine / motor driven anu motor driven auxiliary feedvater pumps shall be operated on recirculation to the condenser to verify proper operation.

Separate tests will be perfomed in order to verify the turuine driven capability and the motor driven capability of auxiliary feedwater pump P-518.

The monthly test frequency requirement shall be brcught 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 >300'F for the motor driven pumps. The turbine oriven capaoliity shaTl be brought current within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of obtaining e percent reactor power.

Acceptable performance will be indicated if the pump starts ano operates for fif teen minutes at a flow rate suf ficient to assure 475 gpm of flow to the Steam Generator at a discharge pressure surficierit to drive that flow through the most restrictive ficw path to a single steam generator which is at a pressure of 10d0 psig.

The monthly testing of the auxiliary feedwater pumps and valves shall be performed in accordance with the inservice inspection requirements of Specification 4.2.2.1.

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 fecuwater actuation test signal. j i

2. Verify that each auxiliary feedwater pump starts as designed !

automatically upon receipt of each auxiliary feeowater actuatien test signal.

4.8.3 All auxiliary feedwater system valves, including those that are locked, sealed, or othen.ise secured in position, are to ce inspected to verify tney are in the proper position following surveillances

. performed pursuant to Specifications 4.8.1, 4.8.2 and ,

4.8.4 l Amendm0M Mo. 3J ,7g , g ,9 3 4,39 ;

RANCHO SECO UNfT 1 TECHNICAL SPECIFICATIONS Surveillance Stancaros 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 genera to r.

Bases The monthly test frequency will be sufficient to verify that the turbine / motor driven and motor driven auxiliary feedwater pumps are operable. Verification of correct operation will be made both f rom 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 normal operating conditions in the event of a total loss of off-site power.

The electric driven auxiliary feedwater pumps are capable of delivering a total feedwater flow 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 600 psig to 1050 psig.

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

l Amendment No. 37,76,80,93 4-39a e

RANCHO SEC0 lJN!T 1 TECHNICAL SPECIFICATIONS l Surveillance Stancaros 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 generator.

Bases The monthly test frequency will be sufficient to verify that the turbine / motor driven and motor driven auxiliary feedwater pumps are operable. Verification of correct operation will be made both from the control rocm 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 normel operating conditions in the event of a total loss of off-site power.

The electric driven auxiliary feedwater pumps are capable of delivering a total feedwater flow 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 ficw of 475 gpm to the entrance of the steam generators over the steam generator operating range of 600 psig to 1050 psig.

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

e l l

l Amendnient No. 31,76,P0,93 4-39a

{

ML20148C703

Text

Attachment:

Navigation menu

Search