ML20126K817

Amend 71 to License DPR-54,revising Pages 3-1,3-2,3-44,3-45 & 3-46 Re Limiting Conditions for Operation to Incorporate Requirements for Redundant DHR Capability During All Modes of Facility Operation

ML20126K817
Person / Time
Site:	Rancho Seco
Issue date:	07/03/1985
From:	Stolz J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20126K798	List: ML20126K795 ML20126K817 ML20126K832
References
NUDOCS 8507300491
Download: ML20126K817 (9)

Text

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[ $ UNITED STATES

p, NUCLEAR REGULATORY COMMISSION 1 5 'j WASHINGTON, D. C. 20555 j

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SACRAMENTO MUNICIPAL UTILITY DISTRICT DOCKET NO. 50-312 RANCHO SECO NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 71 License No. DPR-54

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

A. The application for amendment by Sacramento Municipal Utility District (thelicensee)datedOctober 27, 1980, as supplemented May 30, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in confonnity with the application, the the provisions of the Act, and the rules and 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 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 defense and security or 'o the health ar.d 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.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachrrent to this license amendment, i

and paragraph 2.C.(2) of Facility Operating License No. DPR-54 is hereby amended to read as follows:

8507300491 DR 850703 p ADOCK 0500031;.

PDR

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 71 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REG LATORY COMMISSION

[

Joh F. Sto z, Chief Op ating Reactors Branch #4 ision of Licensing '

Attachment:

Changes to the Technical Specifications Date of Issuance: July 3,1985

ATTACHMENT TO LICENSE AMENDMENT NO. 71 FACILITY OPERATING LICENSE NO. DPR-54 DOCKET NO. 50-312 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Remove Insert 3-1 3-1 3-2 3-2 3-2a 3-44 3-44 3-45 3-45 3-46 3-46 l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation i

3 LIMITING CONDITIONS FOR OPERATION I

}.1 REACTOR COOLANT SYSTEM l Aoolicability Applies to the operating status of the reactor coolant system.

Ob.iective To specify those limiting conditions for operation of the reactor coolant sys tem which must be me t to ensure safe reactor operations.

3.1.1 OPERATIONAL COMPSNENTS 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.l.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 if all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Soiler and Pressure vessel Code, Section Ill.

3.I.l.4 Pressurizer Electromatic Relief Valve A. The nominal setpoint of the pressurizer electromatic relief valve shall be 2450 psig i 10 psig except when reaut red for cold overpressure protection.

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

except during fuel loading and, refueling.

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Amendment No. J, pf, 71

l RANCHO SECO UNIT 1

• TECHNICAL SPECIFICATION Limiting Conditions for Operation 1 Reactor Coolant Loop (A) and its associated steam generator and at least one associated reactor coolant pump, 2 Reactor Coolant Loop (B) and its associated steam generator and at least one associated reactor coolant pump, 3 Decay Heat Removal Loop (A) 4 Decay Heat Removal Loop (8)

With less than the above required coolant loops OPERABLE, immediately initist'e corrective action to return the required coolant loops to OPERABLE status as soon as possible; be in COLD SHUTOOWN witnin 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

Bases A reactor coolant pump or decay 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 thy equivalent of the reactor coolant system volume in one half hour or less. (11 The decay heat removal system suction piping is designed for 3000F and ?00 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 crlitcal since its relieving capacity is greater than that required by the sum of the available hea curces Which are ptmp energy, pressurizer heaters, and reactor decay heat. N Soth pressurizer code safety valves are required to be in service prior to criticality to conform to the system design relief capabilities.(5)The code safety valves prevent overpres-sure for rod withdrawal accidents. The pressurizer code safet) valve lift set point shall be set at 2500 psig + 1 percent allowance for error and each valve shall be capable of relieving 345.000 lb/h of saturated steam at a pressure not greater than 3 percent above the set pressure.

The electromatic relief valve setpoint was established to preYent operation of the valve during transients.

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

When TAV is below 2800 F. a single reactor coolant loop or DHR loop provides sufficient heat removal considerations capability require that at leastfor removing decay heat; but single failure two loops be OPERABLE. Thus, if the reactor coolant loops are not CPERABLE, this specification requires two OHR loops to be OPERABLE.

3-2 Amendment Nos. J, 37, 71

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATION Limiting Conditions for Operation 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) FSAR 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 H

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Amendment No. 71 .

3-2a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATION Limiting Conditions for Operation 3.8 FUEL LOADING AND REFUELING Applicability Applies to fuel loading and refueling operations.

Objective To ensure that fuel loading and refueling operations are perfor'med in a re-sponsible manner.

Specification 3.8.1 Radiation levels in the reactor building refueling area shall be moni-tored by R15026 and R15027. Radiation levels in the spent fuel storage area shall be monitored by R15028. If any of these instruments becomes inoperable, portable survey instrumentation, having the appropriate ranges and sensitivity to fully protect individuals involved in refuel-Ing operations, shall be used until the permanent instrumentation is returned to service.

3.8.2 Core subcritical neutron flux shall be continuously monitored by at least two neutron flux monitors, each with continuous indicaticn avail-able, whenever core geometry is being changed. When core geometry is not being changed, at least one neutron flux monitor shall be in service.

3.8.3 Two decay heat removal pumps and coolers shall be operable. One decay heat removal pump and cooler shall be operable when the transfer canal water level is above 37 feet.

3.8.4 During reactor vessel head removal and while loading and unloading fuel f rom the reactor, the boron concentration shall be maintained at not less than 1850 ppm.

3.8.5 Direct communications between the control room and the refueling person-nel in the reactor building shall exist whenever changes in core geome-try are taking place.

3.8.6 During the handling of irradiated fuel in the reactor building at least one door on the personnel and emergency hatches shall be closed. The equipment hatch cover shall be in place with a minimum of four bolts securing the cover to the sealing surfaces.

3.8.7 isolation valves in lines containing automatic containment isolation valves shall be operable, or at least one shall be in a safety features position.

Amendment No. 77, 71 3'

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATION Limiting Conditions for Operation 3.8.8 When two irradiated fuel assemblies are being handled simultaneously within the fuel transfer canal, a minimum of 10 feet separation shall be maintained between the assemblies at all times. Irradiated fuel as-semblies may be handled with the auxiliary bridge crane provided no other Irradiated fuel assembly is being handled in the fuel transfer canal.

3.8.9 If any of the above specified limiting conditions for fuel loading and refueling are not met, movement of fuel into the reactor core shall cease; action shall be initiated to correct the conditions so that the specified limits are met, and no operations which may increase the re-activity of the core shall be made.

3.8.10 The reactor building purge system, including the radiation monitors, R15001A and R15001B, shall be tested and verified to be operable immed-lately prior to refueling operations.

3.8.11 Irradiated fuel shall not be removed from the reactor until the unit has been subcritical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

3.8.12 No loads will be handled over irradiated fuel stored in the spent fuel pool, except the fuel assemblies themselves. A dead weight load test at the rated load will be performed on the fuel sto' rage building handl-Ing bridge prior to each refueling.

Bases Detailed written procedures will be available for use by refueling personnel.

These procedures, the above specifications, and the design of the fuel handling equipment, as described in subsection 9.7 of the FSAR incorporating built-in Interlocks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and safety. If no change is being made in core geometry, one flux monitor is sufficient. This permits maintenance on the instrumentation. Continuous mon-Itoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The decay heat removal pump is used to maintain a uni-form boron concentration.y The refueling boron concentration indicated in Spec,ification 3.8.4 will be maintained to ensure that the more restrictive of the following reactivity conditions is met:

1. Either a k,7f of 0.95 or less.with all control rods removed from the core.

2. A boron concentration of >l800 ppm.

Specification 3.8.5 allows the control room operator to inform the reactor building personnel of any impending unsafe condition detec,ted from the main control board indicators during fuel movement.

The Specification requiring testing reactor building purge termination is to verify that these components will function as' required should a fuel handling accident occur that results in the release of significant fission products.

3-45 Amendment No. 77, 71

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATION Limiting Condi:!ons for Operation Specification 3.8.11 is required as the safety analysis for the fuel handling accident was based on the assumption that the reactor had been shut down for 72hoursandall308fuelpinsinthehottest fuel assembly fail, releasing all gap activity.

The requirement that at least one DHR loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140 F as required during the REFUELING, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the ef fect of a boron dilution incident and prevent boron strati-fication.

The requirement to have two DHR loops OPERABLE when there is less than 37 feet of water above the core ensures that a single failure of the operating DHR loop will not result in a complete loss of decay heat removal capability. With the reactor vessel head removed and 37 feet of water above the core, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating DHR loop, adequate time is provided to initiate emergency procedures to cool the core.

REFERENCES (1) FSAR, subsection 9 5 (2) FSAR, paragraph 14.2.2.3.2 .

Amendment No. 77, 71 3-46

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