ML19341D741

Amend 31 to License DPR-54,revising Tech Specs Re Reactor High Pressure Trip Setpoint & Pressurizer Electromatic Relief Valve Setpoint to Include Auxiliary Feedwater Flow Path Verification & Sys Outage Requirements

ML19341D741
Person / Time
Site:	Rancho Seco
Issue date:	03/27/1981
From:	Stolz J Office of Nuclear Reactor Regulation
To:
Shared Package
ML19341D740	List: ML19341D739 ML19341D741 ML19341D742 ML19341D743
References
NUDOCS 8104080677
Download: ML19341D741 (47)

Text

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.t NUCLE AR REGULATORY COMM!sslON W ASHINGTON. D. C. 20556

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SACRAMENTO MUNICIPAL UTILITY DISTRICT 00gET NO. 50-312 RANCHO SECO NUCLEAR GENERATING STATION AMENDMENT TO EACILITY OPERATING __ LICENSE Amendment No. 31 License No. DPR-54 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The applications for amendment by Sacramento Municipal Utility District (the licensee) dated July 2,1979, as supplemented October 31, 1979 and April 30,1980; July 9,1980 and October 8, 1980, as supplemented Decenter 11, 1980, com;ily with the standards and raquirements of the. Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

1he forllity will operate in (:onfontit ty with the opplit at tori..

the provisfor.s of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (1.1) 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 comon defense and security or to the health and safety of the public; and E.

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

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. Accordingly, Facility Operating 1.icer se tio. DPR-54 is hereby amended by 2.

revising paragraph 2.C.(2) and adding paragraphs 2.C.(7), 2.C.(G) and 2.C.(9) as follows and by changing the Technical Specifications as indicated in the attachment to this license amen t ent:

2.C.(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 31, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

2.C.(7) Systems Integrity The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program shall include the following:

1.

Provisions establishing preventive maintenance and periodic visual inspection requirements, and 2.

Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

2.C.(8) lodine Mon _itor13 The licensee shall implement a program which will ensure the capability to accurately detennine the airborne iodine concen-tration in vital areas under accident conditions. This program shall include the following:

1.

Training of personnel, 2.

Procedures for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment.

2.C.(9). Backup Method for Determining Subcooling Margi_n The licensee shall implement a program which will ensure the tapability to accurately monitnr the Reartor Coolant System sub-cooling margin. This program shall include the following:

1.

Iraining of ' personnel, and 2.

Procedures for monitoring.

. 3.

This license amendment is effectise as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 1

Jo D. Stolz, Chief ating Reactors Branch #4 vision of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: March 27,1981

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6 ATTACHMENT TO LICENSE AMENDMENT NO. 31 FACILITY OPERATING LICENSE NO. DPR-54 I

DOCKET NO. 50-312 Revise Appernlix A as follw.:

Remove Pages Insert Pages 2-4 2-4 2-7 2-7 2-9 2-9 Figure 2.3-1 Figure 2.3-1 l

3-1 3-1

{

3-2 3-2 i

3-3 3-3 3-4 3-4 Figure 3.1.2-4 3-23 3-23 3-24: 24 3 3-25 3-26 26 l

3-26a 3-27 3-27 3-28 3-28 3-29 3-29 3-30 3-30 3.10a.

3-J4 -

l-:44 3 ~

3-35

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3-36 2-36 l 3-39:'

3-39 a

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__..___._,-.---.---_--u

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' Hemoye_ Page_s, Ins e,r_t_,Pajn 3-40 3-40 3-40a 3-41 3-41 3-42 3-42 3-42a 3-43 3-43 4-1 4-1 4-3 4-3 4-4 4-4 4-5 4-5 4-6 4-6 4-7b 4-7b 4-35 4-35 4-36 4-36 4-39 4-39 4-39a Figure 6,2-2 Figure 6.2-2 6-2 6-2 6-3 6-3 Changes on the revised pages are shown by marginal lines. Pages 3-32, 3-27, 3-36, 3-41, 4-3, 4-5 and 4-36 are unchanged and are provided for your con-venience only.

RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings 2.2 SAFETY LIMITS, REACTOR SYSTEM PRESSURE Applicability Applies to the limit on reactor coolant system pressure.

Objective To maintain the integrity of the reactor coolant system and to prevent the release of significant amounts of fission product activity.

Specification 2.2.1 The reactor coolant system pressure shall not exceed 2750 psig when there are fuel assemblies in the reactor vessel.

2.2.2 The nominal setpoint of the pressurizer code safety valves shall be less than or equal to 2500 psig.

Bases The reactor coolant system serves as a barrier to prevent radionuclides in the reactor coolant from reaching the atmosphere.

In the event of a fuel cladding failure, the reactor coolant system is a barrier against the release of fission products. Establishing a system pressure limit helps to assure the integrity of the reactor coolant system. The maximum transient pressure allowable in the reactor coolant system pressureggyssel under the ASME code, Section lit, is 110 percent of design pressure.

The maximum transient pressure allowable in the reactor coolant system piping, valves, and fittings under ANSI Section B31.7 is 110 percent of design pressure. Thus, the safety limitof275i29sig(110percentofthe2500psigdesignpressure)hasbeen andthepressurizercodesafetyvalves(2500psig)g3yssuretrip(2300psig) established The settings for the reactor high have been established to assure that the reactor coolant system pressure safety limit is not exceeded. The initial hydrostatic test was conducted at 3125 psig (125 per-cent of design pressure) to varify the Integrity of the reactor coolant system.

Additional assurance that the reactor coolant system pressure does not exceed the safety limit is provided by setting the pressurizer electromatic relief

• v.ilva at 2450 psiq.

This setpoint is above nnrmal transients limited by setting the reactor trip at ;2300 psig and sufficiently low to assure limited dependence on safety valves operation.

REFERENCES (1) FSAR, section 4 (2) FSAR, paragraph 4.3.8.1 (3) FSAR, paragraph 4.2.4 Amendment No. 31 2-4

b RANCHO SECTO UNIT 1 TECHNICAL SPr.C1F1 CATIONS

..e.tety 1.1mits and 1.imiting Sefety System Settings S.

Pump monitors The pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to (a) the loss of two reactor coolant pumps in one reactor coolant loop, and (b) loss of one or two reactor coolant pumps during two-pump operation. The pump monitors also restrict the power level to 55 percent for one reactor coolant pump operation in each loop.

C.

Reactor coolant system pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip set point is reached before the nuclear overpower trip set point. The trip setting limit shown in figure 2.3-1 for high reactor coolant system pressure (2300 peig) has l

been established to maintain the system pressure below the safety limit (2750 pais) for any design transient (1) and minimize the challenges to the EMOV and code safeties.

The low pressure (1900 psig) and variable low pressure (12.96 T

- 5834) out trip set point shown in figure 2.3-1 have been established to maintain the DNB ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction. (2,3)

Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of (12.96 T

- 5884).

out D.

Coolant outlet temperature The high reactor coolant outlet temperature trip setting limit (619 F) showa in figure 2.3-1 has been established to prevent excessive core coolant temperatures in the operating range. Due to calibration and instrumentation errors, the safety analysis used a trip set point of 620 F.

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E.

Reactor Building pressure l

The high Reactor Building pressure trip setting limit (4 peig) provides l

positive assurance that a reactor trip will occur in the unlikely l

event of a steam line failure in the Reactor Building or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.

F.

Shutdown bypass In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system. The reactor protection system segments which can be bypassed are shown in 2-7 Anmndnent No.)4731

TAa1.E 2.)-1 REACTOR PROTECTION Sf5ftM T11P $ETTI E LIMITS One Reactor Cociant Pump Four Reactor Coolant Pumpo Three taactor Coolant Pumpe operating in Each Isop Shutds.n Operating (Noalaal Operating (Nostaal (Nostaal Operating Bypass Operating Power - 2003)

Operattaa Power - 751)

Power - 491)

W 1.

leaclear power, I of rated, maa.

105.5 105.5 105.S 5.0 2.

Nuclear power based en flew 1.08 times flow staus 1.08 times flow minus 1.C8 times flow mtnus sypes.e4 emi imbalance,1 of tated, maa.

reduction doe to reJoccine des to reduction due to imbalance (s) tabelance(s) imbalance (s) 3.

Nuclear puwer based en ymy asalture, I of ratus,.42 NA na 55 Sype.s.J 4.

Bligli reactor coolaat I

eyette pressure, pelg, maa.

2300 2300 2300 IC20 'I S.

Low reactor coolmat erstem 1900 1900 1900 sypassed pressure, pels. mio.

6.

Variable low reactor castas 12.96 T - 5814 12.96 7 - 5834 12.96 7 - 5834 3FP',,.4 out out out systwo pressure, peng, et..

7.

Reactor coolant t.mp. F.,maa.

619 619 619 619 8.

liigh Reactor Bu!! ding 4

4 4

4 prenaure, pois, maa.

(1) T,g le in degrees Faasenaelt (r).

(2) Reactor m lant systes t'aw. 3.

(1) AJelnists atively custalled reductico set only Juring reactor sh.sdowe.

(4) Automatically met m atzer segnants of the RFS (as specified) are bypeas4J.

(5) Tlee pump moottors else produce a trip on (a) loss of two reactor coolant pumpa la one reactor coolant loop, anJ (b) toes of om. se twa reactor coolant pumpe during two9 tap operassoa.

2-9 Amendment No. J(, Pf, ~:

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Figure 2.3-1 Protective System Maxi' Allowable Setpoints, Pressure Vs lemperature 2600 2400- -

2 P = 2300 psig T = 619F c

3 Acceptable Operation U

2200- -

Unacceptable g

Operation 8

1. . s e*

2000

{

P - 1900 psig q

cc 1800 l

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540 560 580 600 620 640 Reactor Outlet Temperature, F I

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Arnendrnent No. JAf 31

RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3

LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR COOLANT SYSTEM Apolicability 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 Soccification 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.

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.

C.

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

3.1.1.2 Steam Generator A.

One steam generator shall be operable whenever the reactor coolant average temperature is above 280 F.

3.1.1.3 Pressurizer Safety Valves A.

The reactor shall not remain critical unless both pressurizer code safety valves are operable.

B.

When the reactor is subcritical, at least one pressurizer code safety valve shall be operable i f all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Boller and Pressure Vessel Code, Section Ill.

3.1.l.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.

Bases A reactor coolant pump or deca'y heat removal pump is required to be in opera-tion before the boron concentration is reduced by dilution with makeup water.

Either pump will provide mixing which will prevent sudden positive 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.

(I) i Amendment No.,Af 31 31

RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The decay heat removal system suction piping is designed for 300 F and 300 psig; thus, cne 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 pudo 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 overpres-sure 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 I45,000 lb/h 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 valve during tr n-!ents.

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

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 (3) F$AR paragraph 4.2 5 (4) FSAR paragraph 4.3.8.4 and 4.2.4 (5) FSAR paragraph 4.3.6 and 14.1.2.2 3 l

Anendment No.,4(31 3-2

t RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operatien 3.1.2 PRESSURIZATION, HEATUP, AND C00'.DOWN LIMITATIONS Specification 3.1.2.1 Inservice Leak and Hydrostatic Tests:

Pressure temperature limits for the first five EFP years of inservice leak and hydrostatic tests are given in Figure 3.1.2-3.

Heacup and cooldown ratee shall be restricted according to the rates specified in Figure 3.1.2-3.

3.1.2.2 Heatuo Cooldown:

For the first five EFP years of power opeations, the reactor coolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figure 3.1.2-1 and Figure 3.1.2-2 respectively. Heatup and cooldewn rates shall not exceed the rates stated on the associated figure.

3.1.2.3 The secondary side of the steam generator shall not be pressurized above 200 psig if the camperature of the steam generator shell is 0

below 130 F, 0

3.1.2.4 The pressuriser heatup and cooldown rates shall not exceed 100 F in any 1-hour period.

3.1.2.5 The spray shall not be used if the temperature difference between the pressurizer and spray fluid is greater than 4100F.

3.1.2.6 Prior to exceeding five effective full power years of operation, Figures 3.1.2-1,

-2, and -3 shall be updated for the next service period in accordance with 10 CFR 50, Appendix G, Section V.B.

The highest predicted adjusted reference temperature of all the beltline materials shall be used to determine the adjusted reference temperature at the end of the service period. The basis for this prediction shall be submitted for NRC staff review in accordance with Specification 3.1.2.7.

3.1.2.7 The updated proposed technical specifications referred to in 3.1.2.6 shall be submitted for NRC review at least 90 days prior to the end of the service period. Appropriate additional NRC review time shall be allowed for proposed technical specificationa submitted in accordance with 10 CFR 50 Appendix G.Section V.C.

3.1.2.8 Kmergency/ Faulted Operation:

In the emergency / faulted condition when there is no forced or natural circulation in the reactor coolant system and there is high pressure injection and/or makeup addition, the Reactor Coolant System temperature and pressure shall be limited in accordance with the limit line shown on Figure 3.1.2-4.

Under-the above emergency / faulted conditions, Figure 3.1.2-2 will not apply.

Amenthent No.

31 3-3

RANCIO SECO UNIT 1 TECHNICAL SPEcITICATICNS 1.imiting Conditions for Operatien The maximus allowable pressure is taken to be the icwest pressure of the three calculated pressures. The pressure limit is adjusted for the pressure dif ferential netween the point of system pressura measurement and the limiting corpocent for all reactor coolant pump concinaticus. De limit curves were prepared b ued upen limiting adjusted reference tasperature of all the beltline region sacerials the mos t at the end of the fif th ef fective full power year.

De actual shif t in ItT of the beltline region unter'.a1 will be established y

periodically during operation by removing and evalustar s, in accordance with e

Appendix B to 10 C71L 50, reactor veseal material irradiation surveillance specisans installed naar the inside wall of this or a. similar reactor vessel in the core area.

Because the neutron energy spectra at the specimen location ano at the vessel

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inner wall location are assentially the same, the seasured transition shif t for a sample can be applied with confidence to the adjacent section of the reactor vessel. De limit curves must be recalculated hen the fJCIgg; determined from the surveillance capsule is dif ferent from the calculated AEr, y; for the equiva-y lent capsule radiation exposure.

De unirradiated impact properties of the beltline region sacerials, required by Appendices C and 11 to 10 CFR 50, were determined for those materials for vnich suf ficient asiounts of material were available. The adjus ted reference temperatures are calculated by adding the radiation-induced ARIspr end the unirradiated RIND **

• The predicted t.RTgg; are calculated using the respective neutron fluence and copper and phosphorus contents in accordance with Reg. Guide 1.99.

l forgings is 63k of the closure head region is 600F and the outlet nozzle stee The assumed Er he limitations imposed on pressurizar heatup and cocidown and spray water temperature differential are gravided to assure that the pressurizer is operated within tne design criteria assumed for the fatigue analysis performed in accordanca with the ASME coe requirements.

The limitations during eriergency/f aulted operation when all reactor coolant flow and all feedwater flow is lost to the OTSG's are established to take into consideration that HPI gives false cold leg temperatures. This transi-ent is controlled by Figure 31.2-4 and the vessel beltline te< cerature is calculated using incore ther tocouples and subtracting 150 F for conserva-tism. khen the coolant flow or feedwater flow is re-established, a four hour transition period will be allowed to progress f rom Figure 3.1.2-k to Figure 3.1.2-2.

Amendment No. #, FI, R. 31 3-4

REACTOR COOLANT SYSTEM, EMERGENCY / FAULTED CONDITION-COOLDOWN LIMITATIONS, APPLICABLE FOR 5.0 EFFECTIVE FULL POWER YEARS 2800 2400 LOCI Temo *F Press. (PSIG) 220*

350 240' 446 2000 270*

652 300*

970 c-330*

1460 360*

2214

=

3 368*

2500 5:

1600 e

.2 8v b

1200 2a v

3 Restricted Permissible ju Region Operating Region 800 7_

400 Saturation Pressure f

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200 250 300 350 400 0

incore Thermocouple Temperature Amendment No.31 Figure 3.1.2-4

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation l

3.4 STEAM AND POWER CONVERSION SYSTDj Applicability Applies to the operability of the turbine cycle during normal operation and 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 deccy heat from the reactor core.

Specification 3.4.1 The reactor shall not rezain above 280F with irradiated fuel in the pressure vessel unless the following conditions are met.

3.4.1.1 Capability to supply feedwater to one steam generator at a process flow rate corresponding to a decay heat of 4-1/2 percent full reactor power from at leant one of the following means.

A.

A condensate pump and a main feed pump, or B.

A condensate pump or C.

An auxiliary feedwater pump.

The required flow rates are:

Feedwater temperature, Required flow, gpm degrees,F 40 743 60 756 90 780 3.4.1.2 Two steam system sciety valves are operable per steam generator.

3.4.1.3 The turbits bypass system to the condenser shall have one valve operable or the atmospheric dump system shall have a mir.imum of 1 of 3 valves operable per steam Renerator.

i 3.4.1.4 A minimum of 250,000 gallons of werer shall be available in the condensate storage tank.

3.4.2 In addition to the requirements of 3.4.1, the reactor shall not remain critical unless the following conditions are met:

3.4.2.1 Seventeen of the eighteen main steam system safety valves are operable.

3-23 f

a

rat!CHO SECO UhlT 1 Technical Specifications Limiting Condit ons for Operation i

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3.4.2.2 When two independent 100% capacity auxillary feedwater flow paths are not available, the capacity shall be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the plant shall be placed in a cooling node which does not rely on steam generators for coolino within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3.4.7.3 When at least one 1007. capacity auxiliary feedwiter flow path is not available, tbc reactor shall be o.ade subcritical within l'our hours und the facility piaced in a shutdoua cooling mode which does not rely on steam generators for cooling within next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

F,a s es The feedwater system and the turbine bypass system are normally used for decay heat removal and cooldowm above 280 F.

Feedwater makeup is supplied by operation of a condensate pump and main feedwater pump.

In the event of complete loss of electr! cal power, feedwater is supplied by a turbine driven auxiliary feedwater pump which takes suction from the condensate storage tank. Steam relief would be throu';h the sy'<(s's atmospheri.: relief valves.

If neither main feed pump is available, fcedwater can be supplied to the steam generators by an auxiliary feedwater pump and steam relief would be through the

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turbine bypass system to the condenser.

In order to heat the reactor coolant system above 280 F the maximum steam removal capability required is 4-l/2 percent of rated power. This is the maximum decay heat rate at 30 seconds after a reactor trip. The requirement for two steam system safety valves per steam generator provides a steam relief capability of over 10 percent per steam generator (1,341,938 lb/h).

In addition, two turbine bypass valves to the condenser or two atmospheric dump valves will provide the necessary capacity.

The 250,000 gallons of water in the condensate storage tank is the amount needed for cooiing water to the steam generators f r a period in excess of one day follow-ing a complete loss of all unit ac power.

The mi imum reflef capacity of seventeen steam system safety valves is 13,329,163

~1b/hr.' 2I This is sufficient capacity to p ect the steam system under the design overpower condition of 112 percent.

REFERENCES (I); FSAR paragraph 14.1.2.8.4 (2) FSAR paragraph.10.3.4 (3) FSAR Appendix'jA, Answer to Qucstion 3A.5 Amendment No. 31 3-25

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o RANCl10 SECO UNIY l TECHNICA-. SPECIFICATIONS Limiting Conditions for Operation 3.5 INSTP.UMENTATION SYSTEMS 3.5.1 OPERATIONAL SAFE 1Y INSTRUMENTATION Applicability _

Applies to unit instrumentation and control systems.

SLYE.ctive To delineate the conditions of the unit instrunentation and safety circui ts neces-sary to assure reactor safety.

Specifications 3 5.1.1 Startupand operation are not permitted unless the requirements of table 3 5.1-1, Columns A and B are met.

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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.

In the event the number of OPERABLE Process Instrumentation channels is less than the Total Number of Channel (s). restore the Inopes ahic channels to operahic status within 7 days, or_be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

If the number of operable channels is less than the minimum chan-nels operable, either restore the Inoperable channels to operable within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3.5.l.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.

3.5.1.5 During startup when the intermediate range instrunent cones on scale, the overlap between the intermediate range and the source range instru-mentation 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.

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 un-tripped _ 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 shutdown condition within an additional four hours.

3-25 Amendment No.31

RAtlCII0 SECO UtilT 1 TECHNICAL SPECIFICATIOriS Limiting Conditions for Operation Bases Every reasonable ef fort will be made to maintain all safety instrumentation in operation. A startup is not permitted unless three power range neutron instru-nent channels and two channels cach of the following are operable:

four reactor coolant temperature instrunient channels, four reactor coolant flow instrument channel., f our reac tor c oolant pressure instrument channels, four pressure-temperature i ns t r umen t channels, t our f lor-imhalanc e flow insteoment channel..

Ico

p..we : - run.l.c r o f pumps instiument c h.u me I s, and fout bI qh a c u. Ion bo l lill n::

picssone i ns t i on.cu t channel..

The.nf ety f eatm es actuation system must have two analuq channels f unc tioning coriectly pr ior to startup.

Operation at rated power is permitted as long as the systems have at least the redundancy requirements of Column B (tabic 3.5.1-l).

This is in agreement with redundancy and single failure criteria of IEEE 279 as described in FSAR section 7 There are four reactor protection channels. Normal trip logic is two out of four. Required trip logic for the power range instrumentation channels is two out of three. Minimum trip logic on other instrumentation channels is one out of two.

The four reactor protection channels were provided with key operated 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 hypassed.

Eath reactor protection channel key operated shutdown bypass switch is provided with alarm and lights to indicate when the shutdown bypan swi.tch 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.

The source range and intermediate range nuclear flux instrumentation scales over-

. lap by one decade. This decade overlap will be achieved,at 10-10 amps on the Intermediate range scale.

Power is normally supplied to the control rod drive mechanisms f rom two 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 time to test' the remaining trip devices and.in many cases make on-line repairs.

The OPERABILITY of the SFAS instrumentation systems and bypasses ensure that 1) the associated SFAS action will be-Initiated when the parameter' monitored by each channel or combination thereof reaches its setpoint, 2) the specified coincidence Anendment. No. 31 3-26

+

- -- -.n a

~ _ -

_ _ =.

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Bases (Continued) l logit is maintairud, 3) sufficient redundancy is raaintained to permit a ch.innel to I.c out nl sesvire for i c*. t i nr1 or si.o in t enance, and 4 ) siel I i t. i c tit systrm lunttiin.I l

'fA5 pu poses from diverse pasametess.

.ip..l. i l i t y i. ov.illable for the OPCkABILiiY of these systems is required to provide the overall reliability, i

redundancy, and diversity assurned available in the facility design for the protec-tion and mitigation of accident and transient condition,. The integrated opera-tion of each of these systems is consistent with the assurrptions used in the accident analyses.

r The OPERABILITY of the accident monitoring in;trumentation ensures that sufficier.t information' is availabic on selected plant parameters to rnonitor and assess these variables during.and following an accident. This capability is consistent with l

the reccu.endations of Regulatory Guide 1.97, " Instrumentation for Light-Water-i Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident", December 1975 and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations."

hl ! I Ill fit.!

F$AR, suba.ec t ion 7.1 I:

t i

1 e

4 Teendment No.31-3-26a

~

.M n

a

0 TABLE 3.5.1-1 m

k INSTRUMENTS OPERATING CONDITIONS (C) o (A)

(B)

Operator Action if w

Conditions of Coltanns A Minimus Operable Minimum Degree

~'

and B Cannot be Met Functional Unit Channels of Redundancy Reactor Protection System Bring to hot shutdown 1.

Manual pushbutton 1

O within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Bring to hot shutdown

-4 2.

Power range instrument channel 3(a) 1(a) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> EI M 3.

Intermediate range Bring to hot shutdown 2

instrument channels 1

0 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (b) oDI r* O Bring to hot shutdown V'

channels 1

0 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (c)

%M 4.

Source range instrument

'" n 9O Bring to hot shutdown 5.

Reactor coolant tempers'ure

[j instrument channels 2

1 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Bring to hot shutdown E MM 6.

Pressure-temperature instrument channels 2

1 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

?6*

zm 7.

Flux / imbalance / flow Bring to hot shutdown instrument channels 2

1 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> n

02 I-(* For channel testing, calibration or maintenance the minimum number of operable channels l

may be'two and a degree of redundancy of one for a maximum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

m (b)When 2 of 4 power range instrument channels are greater than 10 percent full power, o,

hot shutdown is not required.

( When 1 of 2 intermediate range instrument channels is greater than 10-10 amps, or 2 of 4 power range instrument channels are greater than 10 percent full power, hot shutdown is o

not required.

1

t

3 5

TABLE 3.5. I-I (Cont inued) z P

INSTRUMENTS OPERATING CONDITIONS y

1 (A)

(8)

(C)

Functlonal Unit Minimuni Operable Minimune Degree Operator Action if Conditions Of Channels of Redundancy Columns A and B Cannot be Met Reactor Protection Syst.eni

8. Reactor coolant pressure a.

High reactor coolant pressure Bring to hot shutdown Instrinaent channels 2

I within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> R

y IE b.

Low reactor coolant pressure Bring to hot shutdown

g Instrinnent channels 2

I within 12 leurs

'f'

9. Panser/ number of pumps em Bring to hot shutdown 2R o$

Inst rament channels 2

I wi thin 12 tours Oo

10. High Reactor Building 2E Bring to hot shutdown Q -,

pressure channels 2

I within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 2_

E ll. Luss of Main feedwater 2

i Bring to hot shutdown within 12 tours

12. Turbine / Generator l

0 Bring to hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> koT.

Safety Features

• 2.

e, n

f. liigh pressure injection, Reactor Buliding isolation, and Reactor h

building emergency cooling 0

J.

Reactor coolant pressure Bring to tot shutdan n 0,

instrimment channels 2

i within letours g

~~~TTt

~

mintamam cohdltions are not met within 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> after hot shutdown, the unit shall be placed in a colt o

6 sfustown condi t ion wi thin an addit ional 28: hours.

7'3'E 3.5 1-1 (continued)

I".

INS ~:;*I'.TS CPERATitJG CONDITIONS

,a, g

j (C)

(A)

(B)

C crator Action if i

+

f Mi-I-i

• ecrable Minimum Degree i

C:: li.: ens of Columns A Functional Unit C-anmels of Rodundancy

?

3 :: S Cannot Be f*2 t 4

i Safety FeaturesA 6

l I

6 b.

Reactor Building Pressure Bring to hot shutdown within i

instra ent channels 2

1

,12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c.

Manual pushbutton' 2

1 lEring to hot shutdown within I 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

d.

Autoratic Actuation Logic 2

1 l Bring to hot shutdown within l

9-l 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> id E I

Z -

i 1 3 2.' Low pressure injection I.

s,.

w 3

a.

Reactor coolant pressure

! Bring to hot shutdown within C

Instra:,ent channels 2

1 l12heurs

_5 b.

Reactor Building pressure Bring to hot shutdown within

>~

instru-ent channels.

2 1

' 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> l r, - -

f,eringtohotshutdownwithin

! :.; G c.

Manua! pushbutton 2

1 12 Feurs I

l'O

3. Reactor Building spray pump i

t E

6 :

t a.

~4eactor Building pressure instru,ent channel 2

1

Bring to hot shutdcwn within a

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 3

b.

Manual pushbutton 2

1

• Srir g to hot shutdown wi thin

[

i 12 Fears

!2?

1 u

e (Continued) c.

I t

t elf -:r.; e conditions are not tret. : t ~ n -3 i-c a r s a f te r ho t shu tdown, t.2 un~t e a!' be placed in a coM + :.tdun condi tion within an at::t : a! 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

_ _.~

TA3LE 3.5.1-1 (Ccntinued)

{

INSTE;*I*.~5 OPERATING CCNDITIONS o

j (C) i C: ator Action if l11 (B)

,E Mini--r-

erable Minimum Degree I

C: :i:ibns of Colu-ns Functional Unit C: a-eis of Redundancy A a-d S Cannot 3e

."e t j

w Safety Features *

1. Reactor Building spray valve 4

~

Bring to hot shutdown within a.

Reactor Building pressure instrument channel 2

1 2L. cars 1

b._ Manual pushbutton 2

1 Bring to hot shutdown within ^

f 2:s %urs (C) l 55 (A)

(B) l Operator Action if Conditions of j,

~5 Process Instrumentation Total Mtad:er of Minimian Channels l l. Pressurizer Vater Level Channels Operable

• Col'rns A and B Cannot Re Met

'i

- 5 3

1 j 5ee Se::icn 3.5.1.2

+!

g w

l a -;

O

2. Auxiliary Feedwater Flow **

1 per 100:

1 per loop

-E

$3. Reactor Coolant System Subcooling

}.

(~

j Margin Mo.iltor 2

I 1.-

+

=

' 4. E*CV Posi tion Indicator (Primary f

--.![

i Detector) power indicator ***

1/ valve 1/ valve a

l 1,5. EMOV Position Indicator (Backup l

Detector) accustic or T/C***

1/ valve 0

l

!6.EMOVBlockValvePositionIndicator 1/ valve 1/ valve I

'!~

l

• • • fety valve

!7. Sa oosition Indicator

l-1 (Pri: ary Detector) T/C 1/ valve 1/ valve I

e.

s.-

8. Safety Valve Position Indicator

. l ~.

-l2 (Backup Detector) acoustic 1/ valve

'O 4

{

f *. n u condi tions are not met w i t M -

~. -ta r s a f te r ho t shu tdown, the.-:

.:e cic cd :n c !.! s :uta :r. cc.r.d; tion wi thin an cdd:-i:-- ? 1-r our 5.

• • 0TSG 1evel may ':e ned fer fiev.

• • • ?cplies t.en E:~ is OPEPA4!.E,

TABLE 3.5.1-1 (Continued) k INSTRUMENTS OPERATING CONDITIONS R

5.

(C) r (A)

(B)

Operator Action If Total Number of Minimum Channels Conditions of Columns A w

Functional Unit Channels Operable and B Cannot Be Met Auxiliary Feedwater_

1.

Low Main Feedwater Pressure:

See Section 3.5.1.2 Start Motor Driven Peimp and i

Turbine Driven Pump 2

1 i

a 1

2.

Contact Monitor - RCS Pumps:

g Start Motor Driven and Turbine Driven Pumps 2

1

[--

?

4 Y

A 8

5

=

n

-4 C-

2. 5 E

E.

~

I i

7 2

4 2

Ti

3

o RAf;Cli0 SECO UNIT 1 TECitNICAL SPECIFICAll0NS Limiting Conditions for Operation 3.5.3 SAFETY FLATURES AtTUATION SYSTEM SETPOINTS Applicability This specification applies to the safety features actuation system actuation setroints.

0,,ldg,t i vc To provide for autonatic initiation of the safety features actuation systen 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 Reactor Building spray valv s 130 psig High Reactor Building pressure Reactor Building spray pumps 130 psig a

High pressure injection and start of Reactor Building cooling and Reactor Building isolation.

1 4 psig Low pressure injection 1 4 psig Low reactor coolant system High pressure injection and start

~

pressure **

of Reactor Building cooling and Reactor Building isointion.

1 1600 psig Low pressure injection 1 1600 psin Automatic Actuation Logic All above Not Applicable Manual All above Not Applicable Loss of all RC Pumps, Starm Auxiliary Feedwater Pumps Not Applicabic Low Feedwater Pressure Start s Auxiliary Feedwater Pumps

.1 750 psig

• May be bypassed during Reactor Building leak rate test.

• • May be bypassed below 1850 psig and is automatically-reinstated above 1850 ps1 9

• • • Five-minute time delay.

Amendnent No. 31 3-34

i e

RANCll0 SECO UNIT 1 TECilNICAI, SPECJrICATIONS 1.f mit Ing Condit f enn for Operatten Wi t h an SFAS net point less cuarervative than the values shown in the obove table,_ declare the channel inoperable and apply the applicable Operator Action requi rement (Column C) of Table 3.5.1-1.

Bases _

liigh Reactor Building Pressure The basis for the 30 psig and 4 poig setpoints for the high pressure signal is to establish a setting w'aich would be reached in adequate time in the event of k sizes and yet be far enough above nornal opera-j

.a DBA, cover a spectrum of brea tion naximum-internal pressure to prevent spurious initiation.

l 1,ow Reactor Coolant System Pressure for high and The basis for the 1600 psig low reactor coolant pressure setpoint initiation is to. establish a value which is high enough Jow pressure injection i

suels that protection is provided for; the entire spectrum of break sizes and is far enough below normal operating pressure to prevent spurious initiation. (1) t REFERENCES

-(l) FSAR, paragraph 14.2.2.5 i

b' 4

i t

f Amendnent No. 31

. 3.

e m

t.

. m

.2 m

___._____.,_s-

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.4 INCORE INSTRUMENTATION Applicability Applies to the operability of the incore instrumentation system.

Objective To specify the functional and operational requirements of the incore instru-mentation system.

Specification Above 80 percent of operating powar determined by the reactor coolant pump combination, reference table 2.3.1, at least 23 individual incore detectors shall be operable to assist in the periodic calibration of the out-of-core The detectors shall detectors ic regard to.the core imbalance trip limits.

be arranged 'as follows and may be a part of both basic arrangements.

3.5.4.1-Axial 1mbalance n ree detectors in each of 3 strings shall lie in the same A.

axial plane with 1 plane in each axial core half.

The axial planes in each core half shall be symmetrical about B.

the core mid-plane.

he detector shall not have radial symmetry.

C.

3.5.4.2 Radial Tilt Each set of Two sets of 4 detectors shall lie in each core half.

The two sets in the same core A.

4 shall lie in the same axial plane.

half may lie in the same axial plane.

Detectors in the same plane shall have quarter core radial B.

symmetry.

3-36

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limitin;; Conditions for Operation 3.6 REACTOR BOIIDING Applicability Applies to the containment when the reactor is suberitical by less than 1 percent ak/k.

Objective To assure containment integrity during startup and operation.

Specification 3.6.1 Containment integrity shall be maintained whenever all three of the following conditions exist:

A.

Reactor coolant pressure is 300 psig or greater.

B.

Reactor coolant temperature is 200 F or greater.

C.

Nuclear fuel is in the core.

3.6.2 Containment integrity shall be maintained when the reactor coolant system is open to the containment atmosphere and the requirements f or a refueling shutdow, are not met.

3.6.3 Positive reactivity insertions which would result in the reactor being suberitical by less than 1 percent ak/k shall not be made by control rod motion or boron dilution whenever the containment integrity is not intact.

3.6.4 The reactor shall not remain critical if the Resetcor Building internal pressure exceeds 1.5 Psig or vacuum exceeds -1.5 psig.

Prior to criticality following refueling shutdown, a check shall be 3.6.5 made to confirm that all manual containment isolation valves which should be closed are closed.

the safety features containment isolation valves specified in Table 3.6.6 3.6-1 shall bb OPERABLE with closure times as shown in Table 3.6-1.

If, under reactor critical operating conditions an automatic containment the other containment isolation valve is determined to be inoperable, If isolation valve in the line shall be tested to insure operability.

the inoperable vilve is not restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shal.'

to the cold shutdown condition within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> be brought or the valve will be placed in a safety features position.

hwndment No. 31 3 39

TABLE 3.6-1 l

RANCHO SECO UNIT 1 TECllNICAL SPECIFICATIONS Limiting Conditions for Operations SAFETY FEATURES CONTAIN!!CNT ISOLATION VALVES VALVE NUMBER DESCRIPTION MAXDWM CLOSURE TIME (SEC)

SFV 53612 RB Atm. & Purge Sample, AB Side.

.3

.SFV 53613 RB Atm. & Rad Sarple, AB Side.

3

. 14 SFV 60003 RC Sys. Drain Isol, AB Side.

SFV 66308 RB Normal Sump Drain, AB Side.

. 15 SFV 92520 Przr. Nitrogen Isol., AB Side.

5 5

SFV 53503 RB Purge Inlet, AB Side.

3 SFV 53604 RB Purge Outlet, AB Side.

SFV 53610 RB Press. Equali.zer, AB Side.

. 15 6

SFV 60002 RC System Vent Isol., AB Side.

. 14 SFV 60004 RC System Drain Isol., AB Side.

8 SFV 66309-RB Normal Sump Drain, AB Side.

SFV 70002 Przr. Liquid Sample Isol., AB Side.

8 SFV 72502 Przr. Gas Sample Isol., AB Side.

5 HV 20611 OTSG's Blowdown Isol., AB Side.

. 22 12 HV 20593 OTSG-A Sample Isol., AB Side 5

HV 20594 OTSG-B Sampic Isol., AB Side.

8 4

SFV 53504 RB Purge Inlet, RB Side.

9 STV 53603 RB Press. Equalizer, RB Side SFV 53605 RB Purge Outlet, RB Side.

8 12 SFV 60001 RC Sys. Vent Isol, RB Side.

18 SFV.70001 Przr. Liquid Sample Isol., RB Side.

. 21 SFV 70003 Przr. Vapor Sample Isol., RB Side.

SFV 72501 Przr. Cas Sample Isol., RB Side.

9

. 15

• SFV'46014 RB CCW Supply AB Side.

. 14

• SFV 46203 RB CCW Return, RB Side.

................ 18

• SFV 46204 RB CCW Return, AB Side.

• SFV 46906 CRD Cooling Water Supply, AB Side.

9

. 14 l

SFV 46907-CRD Cooling Water Return, RD Side.

8 SFV.46908 CRD Cooling Water Return, AB Side.

15 HV 20609 OTSG-A Blowdown Isol...RB Side.

HV. 20610-OTSG-B Blowdown 18o1., RB Side.

. 14 SFV.22021

_RC Syn. Letdown, RB Side.

. 15 6

SFV 22009 RC Sy e. I.e t down, AB Side.

SFV'24004

. RC Pump Seal Return,- RB Side

..71 SFV 24013 RC Pump Seal Return, AB Side..

.8 Manual initiation signal (no auto. initiation)-

Aaendment No.-31 3-40

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Bases The reactor coolant system conditions of cold shutdown assure that no steam will be formed and hence no pressure buildup in the containment if the reastor coolant symtem rupturen.

The selected shutdown conditions are based on the type of activities that are being carried out and will preclude criticality in any occurrence.

The Reactor Building is designed for an internal pressure of 59 psig and an external pressure 2.0 psi greater than the internal pressure. The design external pressure corresponds-to the differential pressure that could be developed if the building is sealed with an internal temperature of 120 F with a barometric pressure of 29.0 inches of Hg and the building is subse-quently cooled to an internal temperature of 80 F with a concurrent rise in barometric pressure to 31.0 inches of Hg.

When containment integrity is established, the limits of 10 CFR 100 will not be exceeded should the maximum hypothetical accident occur.

11:e OPERAHII ITY of t he cont ainment tselntlou ensuren that the containment atmosphere will be isolat ed f rom the out nlde environment in the event of a relcune of radionctive material to the cont ainment at mosphere by pren-su rization of the containment. - Containment isointion within the time limits specified ensures that the relense of radioactive material to the environment will be consistent with the assumptions used in the analyses for LOCA-KEFERENCES FSAR, section 5 i

i Amendnent No.M 3-40a I

i<.

N E

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 7.7 AUXILIARY ELECTRICAL SYSTDiS Applicability Applies to the availability of of f-site and on-site electrical power for station operation and for operation of station auxiliaries.

Objective To define those conditions of electrical power availability necessary to provide for safe reactor operation and to provide for continuing availability of engineered safety features systems in an unrestricted manner.

Specification 3.7.1 The reactor shall not be brought critical unless the following conditions are met.

All nuclear service buses, nuclear service switchgear, and A.

nuclear service load shedding systems are operable.

B.

Two 220 kV lines are in service.

One 6900 volt reactor coolant pump motors bus is energized.

C.

Emergency diesel generators are operable and at least 35,000 D.

gallons of fuel are in each storage tank.

Plant batteries are charged and in service.

E.

Two out of three battery chargers are operable for 125 volt d-e F.

buses "A" and "C", and "B" and "D".

a-c vital Three out of four inverters are operable for 120 volt G.

bus power.

Both startup transformers, No. 1 and No. 2, are in service.

H.

The reactor shall not remain critical unless all of the following 3.7.2 requirements are satisfied:

One 220 kV line shall be fully operational and capable of A.

carrying nuclear service and auxiliary power except as specified in D below.-

3-41 e

r w

y,-+-

,e--

t y

y

t l

RANCi10 SECO UNIT 1 TECilNICAL SFElliICATIONS Limiting Conditions for Operation B.

Both startup transformers shall be in service except that one will be sufficient if during the time one startup trans forrer is inoperable, a diesel generator is started and run continuously.

(

C.

Both diesel generators shall be operable except that from and after the date that one of the diesel generators is made or found to be inoperable for any reason, reactor operation is permissible for the succeeding 15 days provided that during such 15 days the operable diesel generator shall be load tested daily and both startup trans-formers are available.

If the diesel is not returned to service at the end of 15 days, the other diesel will be started and run with at least minimum load continuously for an additional 15 days.

If at the end of the second 15 days the diesel is not returned to service, the reactor shall be brought to the cold shutdown condition within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D.

If the plant is separated from the system while carrying its own auxiliaries, or if all 220 kV lines are lost, continued reactor

~

operation is permissible provided that one energency dineel i

generator is started and run continuously until a trans..sion line is restored.

1 E.

The essential nuclear service electrical buses, switchgear, load l

shedding, and automatic diesel start systems shall be operable except as provided in C above and as required for surveillance testing.

F.

Nuclear service batteries are charged and in service except that one nuclear service battery may be removed from service for not more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

G.

Both nuclear services busses are operable except that one nuclear service bus may be renoved from service for not more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided th't all equipment on the other nuclear service bus is operable 3.7.3 If both-diesel generators be, ioperable, the uni t shall be placed in the cold shutdown condition.

3.7.4 The precsurizer shall be OPERABLL with at least 126 kw of pressurizer heaters. With the pressurizer inoperable due to inoperable emergency power supplies to the -pressurizer heater either restore the inoperable emergency. power supply within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY wi thin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT-SHUTDOWN wi thin the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

i r

3 2 Amendment No. 31 i

l RANCil0 SECO UNIT I i

t TECilNICAl. SPECIFICATIONS LImitin9 Condltlon*, fc Operation 4

.i.

Bases h

i l

failure The auxillary electrical power systems are arranged so t at no s ng e can inactivate enough safety features equiprnent to jeopardize plant safety.

~,.

The normal souce of power to the redundant nuclear service loads is the two All of the startup transformers connected to the 220-kV station switchyard.

. normal. power supplied to plant auxiliary loads is provided through the two f

unit. auxiliary transformers connected to the generator bus.

Emergency power for the nucIcar service loads is obtained from two on-site diesel generators.

transformers. are sized to carry. full plant auxiliary loads.

The startup plant auxiliary power is not available from the unit auxiliary trans-I When former,it will be obtained from the startup transformers.

!+

c s

4 o i:

o i

1< :

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j.

t

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L 1;;

n 3

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['

Amendment No. 31 -

3-42a s.

r

--W-i y

-g we

- i +r v y,

s-w,-

-,rn

,n t -

.y+sy w,

te W--

,, www

- + + + +

T s's a9 r

s, *- -w e et e =

'e'e-1r'

++-c'e

-~w-T m'

t-

-w--

V

RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Limiting Conditions for Operation lines are not under the direct control of the The five 220-kV transmissionTherefore, all five cannot be assured to be available Rancho Seco station.

However, extensive reliabili ty and protective features are than one source of 220-kV power at all times.

utilized so that the probability of losing more two 220-kV lines are in service prior from faults is low.

By requiring thatimmediately available following a loss of the to startup, one circuit will bediesel power supplies and the other offsite 220-kV onsite alternating currentif there is a loss of all 220-kV remote connections, power to the

line, safety features will be supplied by the diesel generators.

The 35,000 gallons of fuel stored in each storage tank permit operation of It is cons'idered unlikely not to the two diesel generators for seven days.from an outside source during this time under the be able to secure fuel oil worst of weather condi tions.

loss of one control panelboards are arranged so that The four 120-volt d-a bus will not preclude safe shutdown or operation of safety features systems.

During periods when one plant battery is de-energized for test or maintenance, the associated 125-volt d -c bus can be supplied from its battery charger.

i Each redundant pair ("A" and "C", "B" and "D") of safety features actuation and reactor protection 125-volt d-c buses has a standby battery charger in Loss of power from one battery addition to the two bus battery chargers.

charger per pair of redundant d-c buses has no significant consequence since ir. addition, each 125-volt d-c bus a standby battery charger is available.

cco continue to receive power from its respective battery without interruption.

Sufficient redundancy is available With any three of the four 120-volt a-c vital t

such that reactor safety is assured. Every reasonable 4

power buses in service effort will be made to maintain all safety instrumentation in operation.

During periods of station operation under the condition of electrical system degradation, as described above in Specification 3.7.2, the operating action required is to start and run sufficient standby-power supplies so as not toAs se compromise the safety of the plant.is placed on operation during certain degra limit reliability of the available power supply.

y ff kw of pressurizer heaters and their associate'd controls The requi rement that 126supplied with electrical power from an emergency bus being capable of being these heaters can be energlied during a loss of provides assuranca thatof fsite power condition to maintain natural circulation at HOT S REFERENCE A

I" FSAR, section 8 r

e b.. _

343 rNNee A c: rx n Amendnent No. 31 e

k:

i

._m u.

..a

a..

.--..;~.

~

l'

RANCHO SECD UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.

SURVEILLANCE STANDARDS Applicability Applies to items directly related to safety limits and limiting conditions for operation during power operation. During cold shutdown, systems and components required to maintain safe shutdown will be tested.

Objective To specify the minimum frequency and type of surveillance to be applied to unit equipment and conditions.

4.1 OPERATIONAL SAFETY REVIEW Specification 4.1.1 The minimum frequency and type of surveillance required for reactor protection system and safety feature protection system instrumenta-tion when the reactor is critical shall be as stated in table 4.1-1.

4.1.2 Equipment and sampling test shall'be performed as detailed in tables 4.1-2 and 4.1-3, 4.1.3 A power distribution map shall be made to verify the expected power distribution at periodic Intervals on approximately every 10 effec-tive full power days using the incore Instrumentation detector

-system.

i.

l Bases Check

. Failures such as blown instrument fuses, defective Indicators, faulted ampil-fiers which result in " upscale" or "downscale" Indication can be easII'y recognized by simple observation of the functioning of an Instrument or system. Furthermore, such failures are, in many cases, revealed by alanm or annunciator action.

Comparison of output and/or state of Independent channels measuring the same variable supplements this type of built-in surveillance. Based on experience in operation of both conventional and nuclear systems, when the unit is in operation, the minimum checking frequency stated is deemed adequate for reactor system instrumentation.

t Amendment No.,36731 4-1 l

i

?

TABLE 4.1-1 INSTRUMEffr SURVEILLANCE REQUIREMENTS Channel Description Check Test Calib rat e Remarks Reactbr Protective System 1.

Source range channel S(l)

P NA (1) When in service 2.

Intermediate range channel S

P NA 3.

Power range amplifier D(1)

NA (2)

(1) Heat balance check daily.

m (2) Heat balance calibration Q3 whenever indicated neutron zy power ard core thernal 6g power ditier by more than fIO 2 percent and daily during g

non-steady-state operation.

mg f,.

-O 4.

Power range channel S

M M(1,2)

(1) Using incore instrumenta-2c tion for split detector OZ dd calib ra tion.

O*

Z (2) Imbalance, upper and lower m

chambers at equilibrium xenon after each startup p

i if not done the previous g

week.

m w

Y 5.

High reactor coolant S

M R

l pressure channel N

6.

Low reactor coolant S

M R

{

pressure channel m

a.

7.

' Reactor coolant ten-S M

R perature channel

E$

TABLE 4.1-1 (Continued) 6 INSTRUMDfT SURVEILLANCE REQUIREMENTS 5

2 rh==-1 Description Check Test Calibrate Remarks P

3 8.

Reactor coolant pressur /

S M

R tasaperature comparator 9.

Power /iabalance/ flow S

M R

cs w rator 10.

Pump / flux comparator S

M R

11.

Nigh Reactor Building D

M R

M Pressure Channels g

2>

12.

Protection channel NA M

NA gZ colecidence logic

> g-

"O 13a. CED Trip Breaker Sf4 e-I-

1) EPS Undervoltage trip NA M

NA

2) Turbine /Cenerator, qO loss of Feedvater Trip NA M

NA 6h 13b. Turbine Generator Trip Boa g

g Functtonal Teut NA ST(1)

R Zrn Safety Features System 14.

Emergency core cooling a

lajectlon, emergency 7

building eno11ng and

{

building isolation n*

analog channels N

a.

Reactor coolant S

H R

g pressure channel g

b.

Reactor Building S

M R

4 psig channel e

TABLE 4.1-1 (Continued)

INSTRUKdNT SURVEILLANCE REQUIREMENTS Channel Description Check Test.

Calib rate Remarks 15.

Reactor Building spray system analog channels a.

Spray pump NA M

R 30 psig channels m

~b.

Spray valve NA M

R Q,

z>

30 psig' channels gZ 16.

High pressere ' Injection',

-NA M

NA f O O

g

. emergency building cool-m$

o.

O ing and building isola-G8 tion digital logie 2! C channels O2 17.

Low pressure injection NA M

NA d -4 j"

digital logic channels en 18.

Reactor Bui'. ding spray NA M

NA vi pumps digital logic

}

channels m

^

19.

Reactor Building spray NA M

NA O

~

. valves digital logic

{n channels m

B

?

b I

t

D TABLE 4.1-1 (Continued) 5 l'

s INSTRITHENT SURVEILLANCE REQUIREMENTS i

-w l

Channel Description Check Test Calibrate Remarks

'N d:

20.

.High' pressure Injection, N A~.

M NA l

I'

~

j.

Reactor But Iding Isolation, i.

and Reactor'ButIding emer-gency cooling. Channel A i

manual trip.

'21 High pressure. injection NA M

NA l

Reactor Building isola-tion, and Reac tor But Id-

ing emergency; cooling M

2 Channel.B manual' trip.

5

_=

z

.22.

t.ow pressure injection NA P.

NA 22 "O

Channe1 A manual trip T

• 2-A 23 Low pressure injection NA R

NA GG Channel B manual trip'

e R 5 2 fe.. ' Reactor Building spray NA R

NA pump.Channe1 A manual 5-trip 5

Reactor BulIding spray.

NA

.R NA w

25.

pump' Channel B manual E

trip' 5,

"~

26.

Reactor Building spray NA R

NA valves Channel A manual 8*

trip l

1 5

I TABl.E 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENT 5' Ng Channel Description Check Test Calibrate Remarks 9{-

J

42. Reactor Building drain 3

accumulation tank level NA NA R

43.

Incore neutron detectors M(l)

HA NA (l)Checkfunctioning, Including

• o

(

functioning of computer read-

-\\

out and/or recorder readout.

44. Process and area radiation d

monitoring systems V

H Q

45. Emergency plant rodlation Instruments H(1)

NA R

(1) Battery check

'46.

Environmental air monitors M(1)

NA R

(1) Cneck functioning 47 Strong motion accelerometer Q(1)

NA R

(l) Eattery check 4C. Auxiliary Feedwater R

Start Circuit ig

a. Phase tabalance/Under-gg power RCP 5

H R

gg

b. Low Main Feedwater

%{

y Pressure NA NA R

k U

49. Pressurizer Vater Level H

NA R

a=

-4

50. Auxillary Feedwater Flow-Rate H

NA R

g-z 51.

Reactor Coolant System Sub-v.

cooling MarnIn Monitor M

NA R

52.

EMgVPowerPosition 7

(Nibr#3c5tector) j y

M NA R

53 EMOV PosI tion Indicator b d[

ke 59 EMOV Elock Valve Posltlon ou na Indicator M

NA R

g e,

55. Safety Valve Position in-g NA dicator(Primary DetectorT/C M

g.

g o,

56. Safety Valve Posi tion in-dicator (Backup Detector)

M NA R

A.coustic 5 = Each s'hift M = Monthly P = Prior to each startup If not done prevlous week D = Dally Q = Quarterly R = Once during the refueling Interval.

W = Weekly SY= Semiannual l

g

RANCHO SECO llNIT 1 TECllN I C AL SP f C I f I CAT l ati',

' nt vcI Ilans e 5 I anel.o il,

4.6.5 Diesel generator fuel oil supply shall be tested as follows:

A.

During the monthly diesel generator test, the diesel fuel oil transfer pumps shall be nonitored for operation.

B.

Once a month, quantity of the diesel fuel oil shall be logged and checked agains t minimum speci fications.

The tests specifi'ed will be considered satisfactory if control room indication and/or visual examination demonstrates that all components have operated properly.

4.6.6 The pressurizer shall be tested as follows:

A.

The pressurizer water level shall be determined to be within its limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

{

B.

The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by transferring power from the normal to the emergency power supply and energizing the heaters.

Bases The tests specified are designed to demonstrate that the diesel generators will provide power for operation of safety features equipment. They also assure that the emergency generator control system and the control systems for the safety features equipment will function automatically in the event of a loss of all normal a-c-station service power, or upon receipt of a safety features actuation signal. The testing frequency specified is intended to identify and permit correction of any mechanical or electrical deficiency before it can result in a system failure. The fuel oil supply, starting cir-

.i cuits and controls are continuously monitored and any faults are alarmed and

. indicated. An abnormal conditi,. in these systems would be signaled without i

having to place the diesel genersairs on test.

Precipitous failure of the pinnt battery is extremely unlikely. The surveillance spcrliled is that which hns been demonstrated over the ye,ta to provide nn Indi-cat ion of a cell beco.ning unserviceable long before It falls, t

REFERENCE (1)

IEEE 308 i

4-35 i.

~

Amendment No. 31 r

• e y

u 4,.,

R/09CHO SECO UNIT 1 l

TECHNICAL SPECIFICATIONS Surveillance Standards 4.7 EEACTOR CONTROL ROD SYSTEM TESTS 4.7.1 CONTROL ROD DRIVE SYSTEM FUNCTIONAL TESTS Applicability Applies to the surveillance of the control rod system.

Objective l

To assure operability of the control rod system.

Specification 4.7.1.1 The control rod trip insertion time shall be measured for each control rod at either full flow or no flow conditions following each refueling outage prior to return to power. The maximum control rod trip inser-tion time for an operable control rod drive mechanism, except for the axial power shaping rods (APSRs), from the fully withdrawn position to 3/4 insertion (104 inches travel) shall not exceed 1.66 seconde at hot, reactor coolant full flow conditions or 1.40 seconds for hot, no flow conditions. For the APSRs it shall be demonstrated

}

that loss of power will not cause rod movement. If the, trip inser-tion time above is not met, the rod shall be declared inoperable.

4. 7.1. 2 If a control rod is misaligned with its group average by more than an indicated nine inches, the rod shall be declared inoperable and the limits of Specification 3.5.2.2 shall apply. The rod with the greatest misalignment shall be evaluated first. The position of a rod declared inoperable due to misalignment shall not be included in computing the average position of the group for determining the operability of rods with lesser misalignments.

i 1.7.1.3 If a control rod cannot be exercised, or if it cannot be located with absolute or relative position indications or in or out limit lights, the rod shall be declared to be inoperable.

Bases The control rod trip insertion time is the total elapsed time from power interruption at the control rod drive breakers until the control rod has completed 104 inches of travel from the fully withdrawn position. The speci-fied trip tbse is based upon the safety analysis in FSAR, section 14.

Each control rod drive mechanism shall be exerciesd by a movement of approxi-mately two inches of travol every two weeks. This requirement shall apply to either a partial or fully withdrawn control rod at reactor operating conditions.

Excrcising the drive mechanisms in this manner provides assurance of reliability of the mechanisms.

4-36

o P,AtlCil0 SECO UlliT 1 if C'If41 CAL SPEC IF ICA T 10!iS Surveillmice Standards i

4.8 AUXILl/PY fi l D'.!nTf f< l'UltP PER IODI C T[ ST li10 Applicability Applies to the periodic testing of the turbine and notor driven auxiliary feederater pumps.

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

Specification 4.8.1 At least every 92 days or, a st iagered test bas!s at a time when the average reactor coolant system temperature is > 305 F, the turbine / motor driver.

and motor driven auxiliary feedwater pumps shall be operated on recircula-tion to the condenser to verify proper operation.

The 92-day test frequency requirement shall be brought current within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> af ter the aver age reactor coolant sys tem temperature is > 3050F, Acceptable performance will be indicated if the pump starts and operates for fifteen minutes at the design flow of 780 gpm.

This flow will be verified using tank level decrease ar.d pump dif ferential pressure.

1 4.8.2 At least once per 18 months during a shutdown:

1.

Verify that each automatic valve in the flow path actuates to its correct position upon receipt of each auxiliary feedwater actuction tes t signal.

2.

Verify that each auxillary feedwater pump starts as designed I

automatically upon receipt of each auxiliary fecdwater actuation test signal.

4.8.3 All valves, including those that are locked, sealed, or othemise srecured in position, are to be inspected monthly to verify they are in the proper position.

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 i

the motor-driven AFW pumps can pump water from the CST to the steam generators.

Amendment No. 31 4-39 L

D RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards

+

-Bases The quarte:1y test frequency will be sufficient to verify that the turbine / noter driven and notor di iven auxiliar y f eedwater punns are operable. Veiiritation of

cor rect operation wilI be made tioth f rom the con t rol roimi ins t ruN nt a t ion m.1

' direct visual observation of the pumps.

.The-OPERABILITY of the auxiliary feedwates system ensures that the Reactor Coolant Sys tem can be cooled down to less than 305 F f rom normal operating conoitions in the event of a total less of off-site power'.

Each electric driven auxiliary feedwater pump is capable of delivering a total feeduater flow of 780 apm at a pressure of 1050 psig to the entrance of the steam generators. Tha steam driven auxiliary feedwater pump is capable of de-livering a total feedwater flow of 780 9pm at ta pressure of 1050 psig to the

~

entrance'of the steam generators. This capacity.is sufficient to ensure that ade-quate feedwater. flow -is available to renme decay heat and reduce the Reactor Coolant' System' temperature.to less than 30f f when the Decay.. Heat Removal System may..be placed.ir.to operation.

+

r N

I i Amendment. No. 31 -

' 4-39a-b

j RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS MANACIR Cr a NUCLit.R **

OPERAT10N$

ENGINEERING AND

• PLANT

• TECMICAL 8 ADMINISTRATIVE QUALITY CONTROL SUPERINTENDENT ASSISTANT STAFF SUPERVISCR CHEMICAL & RADI ATION *

$UPERVISOR

• SUPERVl50R a SUPERVl50R NUCLEAR NUCLEAR NAINTENANCE

,0PERATIONS (St.)

CHEMISTRY & R40lAT10N SHIFT

[LECTRICAL, INSTRUNENT.

PERSONNEL SUPERvlSOR (SL)

MECHANICAL MAINTENANCE PERSONNEL 3R. POWER PLANT

• QUALITY ENGINEER CONTROL NUCLEAR / MECHANICAL PERSONNEL SENIOR CONTROL RO'1M OPERATORS (L)

CONTROL ROOM Oi'ERATORS (L)

AUXILIARY OPERATORS SL - Senior Operator License Laperator License

• Responsible for_ Fire Protection Program

• - Routine Reporting EQUIPMENT Requirer.ents on Person,el ATTENDENT5 Figure 6.2-2 Changes PLANT ORGANIZATION CHART Amendment No. JC M 31

4 RANCllO SI(.O UNIT I il CllNl CAL SI't.C i FIC A11045 TABLE 6.2-1 SHIFT CREW PcRSONNEL AND LICENSE REO.UIREMENTS REACTOR H0DE TW4CHO SECO JOB TITLE COLD OTHER TilAN SHUTDOWN COLD SilVTDOWN

=

Shifi Supervisor 1 - SL 1 - SL Sr. Control Room Operator or Control Room Operator 1-L 2 - L*

Auxiliary Operator or Equipaent Attendant 1

1 Equipment Attendant or Power Plant Helper.

I Shift Technical Advisor.

O I

Minimum Total Personnel 3

6

• 0nc licensed operator when the reactor is shut down greater than 1% Ak/k.

l_

^^ ln the event that any member of a minimum shift crew is absent of

[

ini.apacitatril due to lilnes', or injury, a quallfled replacement l-shall be designated to report onsite within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

l SL -NRC. Senior Licensed Operator

.L -NRC Licensed Operator i

l I

6-2

- Amendment No. 31 L

4 f

v p

aw 7.p-a

-,y e-*

• re y-

__~

s.

_/ D't i fil STPAT I V F CO'iTPOLS

. 6. 3 FACILITY staff NAlfFlCA710!!5 6.3.1-Eacirtvnher of. tic unit staff sha l l rnec t or erf.ced the min in:un qual i f ir a-tions of AfEl fl!8.1-l'.371 for compa r sh le pos i t ion'.. cytopt for (1) the (Chenical-Rndiat ton Supervisor.' who shal1 meet or c7cced tiu qua Ii f ica-tions of Regulat<.sry Guide 1.8, Septen.ber 1975 arid (2) the Shi f t Technical Advisor who shall have a bachelor's deurce or equivalect in a scientifir.

or engineering discipline. The STA shall receive specific training i,, p l a:.t design, and response and analysis.of the plant for transient s and accish nt s.

6.4' TRI.lli t t.G 6.4.I' A retraining and replacement training program for the operating sta!f

- shall be maintained under the direction of the Training Saperviser

' and shall meet or excced the requiremer.ts and recomendations of Sec t ion 5 5 c.f' ANSI N18.1-1971 and Appendix "A" of 10 CFP. Part 55 4

.6.4.2

'A' training program for the Fire Brigade shall be maintained urder the direction of:the Safety Technician and shall racet or excee f the -

requirczents of' Section 27 of the. NFPA Code - 1975, except refresher

- classroom training-shall be on a quarterly schedule.

6.'5 REVIE'W A!40 AUDIT' q

6.5.1 PLAfD, RE" LEU COMMITTEE (PRC)

Fil!10T I ON -

i 16.5.l.I The Plant Review Committec shalI function to advise the Manager of Nuclear Operations on all matters related to nuclear safety, f

. COMPOSIT10ft' i

6.5.1.2 The Plant Review Committee shall be composed of the:

i

- Chairman:

' Technical Assi starit.

Member:

Supervisor Nuclear Operations Member:

Engineering (, Quality. Control Supervisor

' Member:

Supervisor Nuclear Mainter.ance Member:

Chemical and Radiation Supervisor

. 0ther members :as the ~ Manager of ' Nuclear. Operations may appoin t f rom ' t irne to.t ime.

t Amendment No. )(, )4r, )(, 31

~6-3.

k T

- -