ML20153B285

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Amend 97 to License DPR-54,modifying Various Sections of Tech Specs to More Closely Resemble B&W STS
ML20153B285
Person / Time
Site: Rancho Seco
Issue date: 03/15/1988
From: Knighton G
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20153B271 List:
References
NUDOCS 8803220027
Download: ML20153B285 (73)
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{{#Wiki_filter:"toq t .b A, - UNITED STATES - .,f-g NUCLEAR REGULATORY COMMISSION O i8 WASHING TON, D. C. 20655 1, SACRAMENTO MUNICIPAL UTILITY DISTRICT DOCKET NO. 50-312 RANCHO SEC0 NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPFRATING LICENSE Amendment No. 97 License No. DPR-54 1.

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

A. The application for amendment by Sacramento Municipal Utility District (the licensee) dated October 1,1987 as supplemented November'25 and. December 3.- 1987 complies with the standards andrequirementsoftheAtomicEnergyActof1954,asamended(the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in confo/mity'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 safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in a::cordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 8803220027 080315 DR p ADOCK 05000312 PDR

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: (2)TechnicalSpecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 97, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This Amendment to the Technical Specification shall become effective within 30 days of the issuance date or prior to reactor criticality following the 1986/87 outage, whichever is first. The implementation delay is provided to allow time for modification of affected procedures and promulgation of these changes to personnel. FOR THE NUCLEAR REGUL TORY COMMISSION 1 / gr )eY/ bV kl ' George W. Knighton[ irector Project Directorate V Division of Reactor Projects - III, IV, Y and Special Projects Attachcent: Changes to the Technical Specifications Date of Issuance: March 15,1988

March 15,1988 m-ATTACHMENT TO LICENSE AMENOMENT NO. 97 FACILITY OPERATING LICENSE NO.~DPR-54 DOCKET N0. 50-312 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Arrendment number and contain vertical lines indicating the area of change. Remove Insert i i iia ifa lii iii v v vi vi .vii vii ix ix x x 1-1 1-1 1-2 1-2 1-2a 1-2a 1-2c 3-0 3-0a 3-1 3-1 3-la 3-2 3-2 3-3 3-3 3-6 3-6 3-7 3-7 3-8 3-8 3-12 3-12 3-13 3-13 3-15 3-15 3-16 3-16 3-17 3-17 3-19 3-19 3-19a 3-20 3-20 3-21 3-21 3-22 3-22 3-22a 3-22a 3-27 3-27 3-30a 3-30a ,w ,,v4

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,f F ~;\\.: ATTACHMENT TO LICENSE AMENDMENT NO. 97 (CONTINUED) FACILITY OPERATING LICENSE NO. OPR-54 1 DOCKET-NO. 50-312 Remove I"'"# 3-31 3-3' 3-36 3-36 3-381 3-38j 3-39 3-39 3-46a 3-46a 3-46b 3-47 3-47 4-0 4-0a 4-1 4-1 4-7b 4-7b 4 7f 4-7g 4-7g 4-8 4-8 4-8a. 4-8a 4-13 4-13 4-15 4-15 4-16 4-16 4-16a 4-17 4-17 4-18 4-18 4-18a 4-18a 4-34h 4-34h 4-341 4-341 4-34j 4-34j 4_34k 4-34k 4-341 4-341 4-34m 4-41b 4 41b 6-12 6-12 6-12f 6-12f Pages 4-35 and 4-41c are reissued to add a vertical line omitted in Amendment 94 and tn consolidate changes in previous amendments.

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS I Section Eagt 1 DEFINITIONS 1-1 1.1 RATED P0HER 1,1 1.2 REACTOR OPERATING CONDITIONS 1-1 1.2.1 Cold Shutdown 1-1 1.2.2 Hot Shutdown 1-1 1.2.3 Reactor Critical 1-1 1.2.4 Hot Standby 1-1 1.2.5 Power Ooeration 1-1 1.2.6 Refuelina Shutdown 1-1 1.2.7 Refuelino Ooeration 1-2 1.2.8 Refuelina Interval 1-2 1.2.9 Startuo 1-2 1.2.10 Remain Critical 1-2 1.2.11 T 1-2 g 1.2.12 Heatuo - Cec 1down Mode 1-2 1.2.13 Action 1-2 1.3 DEEF6HLE 1-2 1.4 PROTECTION INSTRUMENTATION LOGIC 1-2 1.4.1 Instrument Channel 1-2 1.4.2 Reactor Protection System 1-2a 1.4.3 2rotection Channel 1-3 1.4.4 Reactor Protection System logic 1 -3, 1.4.5 Safety Features System logic 1-3 1.4.6 Degree of Redundancy 1-3 Amendment No. 26,97 i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued) Section Erge 1.16 RESTRICTED AREA 1-7 1.17 SITE 80Vf0ARY 1-7 1.18 DOSE E0VIVALENT I-131 1-7 1.19 MENER(S) 0F THE PUBLIC 1-7 2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2-1 2.1 SAFETY LIMITS. REACTOR CORE 2-1 2.2 SAFETY LIMITS. REACTOR SYSTEM PRESSURE 2-4 2.3 LIMITING SAFETY SYSTEN SETTINGS. PROTECTIVE INSTRlNENTATION 2-5 3 LIMITING CONDITIONS FOR OPERATION 3-0 3.0 GENERAL LIMITING CONDITIONS FOR OPERATION 3-0 3.1 REACTOR COOLANT SYSTEM 3-1 3.1.1 Doerational Comoonents 3-1 3.1.2 Pressurization. Heatuo. and Cooldown Limitations 3-3 3.1.3 Minirean Conditions for Criticality 3-6 3.1.4 Reactor Coolant System Activity 3-8 3.1.5 Chenistry 3-10 3.1.6 Leakace 3-12 3.1.7 Moderator Temoerature Coefficient of Radioactivity 3-15 3.1.8 Low Power Physics Testino Restrictions 3-15b l 3.1.9 Control Rod Operation 3-16 3.2 HIGH PRESSURE INJECTION. CHENICAL ADDITION AND LOW l I TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEMS 3-17 3.3 EMERGENCY CORE COOLING. REACTOR BUILDIfG EMERGENCY COOLING. AND REACTOR BUILDING SPRAY SYSTEMS 3-19 3.4 STEAM AND POWER CONVERSION SYSTEM 3-23 Amendment No. 28,53,97 iia

7-RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued) Section Eage 3.5 INSTRlNENTATION SYSTEMS 3-25 3.5.1 Ooerational Safety Instranentation 3-25 3.5.2 Control Rod Gmuo and Power Distribution Limits 3-31 3.5.3 Safety Features Actuation System Setooints 3-34 3.5.4 Incore Instrunentation 3-36 3.5.5 Accident Monitorina Instrunentation 3-38d 3.5.6 Emercency Feedwater Initiation and Control Setooints 3-38g 3.5.7 Emeraency Shutdown Instrunentation 3-38i 3.6 REACTOR BUILDING 3-39 3.7 AUXILIARY ELECTRICAL SYSTEMS 3-41 3.8 f]!EL LOADING AND REFUELING 3-44 3.9 SPENT FUEL POOL 3-46A 3.10 SECONDARY SYSTEM ACTIVITY 3-47 3.11 REACTOR BUILDING POLAR CRANE AND AUXILIARY HOIST 3-49 3.12 SH0CK SUPPRESSORS (SNUBBERS) 3-51 3.13 AIR FILTER SYSTESM 3-52 3.14 FIRE SUPPRESSION 3-53 3.14.1 Instrunentation 3-53 3.14.2 Water Systen 3-53 3.14.3 Sorav and Sorinkler Systens 3-56 3.14.4 C09 System 3-56 Amendment No. 28,J9,80,82,84,93,97 iii a-

~ RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued) Section EASA 4 SURVEILLANCE STANDARDS 4-0 4.0-GENERAL SURVEILLANCE REOUIREMENTS 4-0 4.1 OPERATIONAL SAFETY REVIEW 4-1 4.2 SURVEILLANCE OF ASME CODE CLASS 1. 2. AND 3 SYSTEMS 4-10 4.2.1 Reactor Vessel Surveillance Soecimens 4-10 4.2.2 Inservice Ity)ection 4-11 4.2.3 Leakage Surveillance 4-13 4.3 TESTING FOLLOHING OPENING OF SYSTEM 4-14 4.4 REACTOR BUILDING 4-15 4.4.1 Containment Leakage Tests 4-15 4.4.2 Structural Intearity 4-21 4.4.3 This Specification has been deleted 4-25 4.5 EFERGENCY CORE COOLING AND REACTOR BUILDING 4-26 COOLING SYSTEM PERIODIC TESTING 4.5.1 Emergency Core Cooling System 4-26 4.5.2 Reactor Buildina Cooling Systems 4-29 4.5.3 Decay Heat Removal System and Reactor Buildino Sorav System Leakaae 4-32 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTING 4-34 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4-36 4.7.1 Control Rod Drive System Functional Tests 4-36 4.7.2 Control Rod Program Verification (Group vs. Core Positions) 4-37 4.8 AUXILIARY FEEDWATER PUHP PERIODIC TESTING 4-39 4.9 REACTIVITY ANOMALIES 4-40 4.10 EHERGENCY CONTROL ROOH FILTERING SYSTEM 4-41 Amendment No. 28,55,5/,/6,W5,97 y

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued) Section EAga 4.11 REACTOR BUILDING PURGE EXHAUST FILTERING SYSTEM 4-42 4.12 AUXILIARY AND SPENT FUEL BUILDING FILTER SYSTEMS 4-43 4.13 AUGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH 4-44 ENERGY LINES OUTSIDE OF CONTAINMENT 4.14 SHOCK SUPPRESSORS (SNUBBERS) 4-47 4.15 RADIOACTIVE MATERIALS SOURCES 4-48 4.16 Reserved 4-49 4.17 STEAM GENERATORS 4-51 4.17.1 Steam Generator Samole Selection and Insoection 4-51 4.17.2 Steam Generator Tube Samole Selection and Insoection 4-51 4.17.3 Insoection Freauencies 4-52 4.17.4 Accentance criteria 4-53 4.17.5 Reoorts 4-54 4.17.6 OTSG Auxiliary Feedwater Header Surveillance 4-54 4.17.7 Inspection Acceptance Criteria and Corrective Actions 4-55 4.17.8 Reports 4-55 4.18 FIRE SUPPRESSION SYSTEM SURVEILLANCE 4-58 4.19 RADIOACTIVE LIOUID EFFLUENT INSTURMENTATION 4-63 4.20 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 4-65 4.21 LIOUID EFFLUENTS 4-69 4.21.1 Concentration 4-69 4.21.2 Dose Calculations 4-72 4.21.3 Liauid Holduo Tanks 4-73 4.22 GASEOUS EFFLUENTS 4-74 4.22.1 Dose Rate 4-74 Amendment No. 75,5),$$,97 Vi l

RANCHO SECO LNIT 1 TECM ICAL SPECIFICATIONS TABLE OF CONTENTS (Continued) Section Ese 4.22.2 Noble Gases 4-77 4.22.3 Iodine-131. Trititm and Radionuclides in Particulate Fonn 4-78 4.23 GASEOUS RADWASTE TREATMENT 4-79 4.24 GAS STORAGE TANKS 4-80 4.25 SOLID RADIGACTIVE WASTES 4-81 4.26 RADIOLOGICAL ENVIR0tNENTAL HJNITORING 4-83 4.27 LAto USE CENSUS 4-86 4.28 EXPLOSIVE GAS MIXTURE 4-87 4.29 FUEL CYCLE DOSE 4-89 4.30 INTERLABORATORY C0iPARISON PROGRAM SURVEILLANCE RE0VIRNENT 4-90 4.31 NUCLEAR SERVICE ELECTRICAL BUILDING BERGENCY HEATING VENTILATION Ato AIR CO@ITIONING 4-91 4.32 TDI DIESEL GENERATOR CONTROL ROCH ESSENTIAL VENTILATION SYSTEM 4-92 4.34 METER 0 LOGICAL K)NITORING It6TRtNENTATION 4-93 4-35 HYDROGEN RECCE INERS 4-95 5 DESIGN FEATURES 5-1 5.1 SIIE 5-1 5.2 CClfTAIPNENT 5-2 5.2.1 Reactor Buildino 5-2 5.2.2 Reactor Buildina Isolation Systen 5-3 5.3 REACTOR 5-4 5.3.1 Reactor Core 5-4 5.3.2 Reactor Coolant Systen 5-4 5.4 NEW Ato SPENT FUEL STORAGE FACILITIES 5-6 5.4.1 New Fuel Insoection and Temorary Storace Rack 5-6 5.4.2 New and Spent Fuel Storage Racks and Failed 5-6 Fuei Storage Container Rack 5.4.3 New and Spent Fuel Temporary Storage 5-6 5.4.4 Spent Fuel Pool and Storage Rack Design 5-6 AmendmentNo.)6,ps,g,J4,94,9) vii 97 -P

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS ~ LIST OF TABLES Section Eage 1.2-1 OPERATIONAL MODES 1-2c l 2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS 2-9 3.0-1 APPLICABILITY OF SPECIFICATIONS 3.0.3 AND 3.0.4 3-0a 3.3-1 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES 3-22a 3.5.1-1 INSTRlNENTS OPERATING CONDITIONS 3-27 3.5.5-1 ACCIDENTMONITORINGINSTRLNENTATIONOPERABILITYREQUIREMENTS 3-3Ee l 3.5.7-1 EMERGENCY SHUTDOWN INSTRLMENTATION 3-38f l 3.6-1 SAFETY FEATURES CONTAIfNENT ISOLATION VALVES 3-40 3.7-1 VOLTAGE PROTECTION SYSTEM RELAY TRIP VALUES M la 3.7-2 VOLTAGE PROTECTION SYSTEM LIMITING CONDITIONS 4-42k 3.12-1 SAFETY RELATED HYRAULIC SNUBBERS 3-51a-e 3.14-1 FIRE DETECTION INSTRlNENTS FOR SAFETY SYSTEMS 3-35 3.14-2 WATER SUPPRESSION ZONES 3-SGba 3.14-3 CARBON DIOXIDE SUPPRESSION ZONES 3-SM,e l 3-14-4 FIRE HOSE STATIONS 3-57a,b 3.15-1 RADI0ACTIVELIQUIDEFFLUENTMONITORINGINSTRlNENTATION 3-61 3.16-1 RADI0 ACTIVE GASES EFFLUENT M]NITORING INSTRlNENTATION 3-64 l 3.22-1 RADIOLOGICAL ENVIR0fNENTAL MONITORIfE PROGRAM 3-83 l 3.22-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS 3-86 IN ENVIR0fNENTAL SAMPLES 3.29-1 METER 0 LOGICAL MONITORIfE INSTRLNENTATION 3-95 4.0-1 APPLICABILITY OF SPECIFICATIONS 4.0.2 AfD 4.0.3 4-0a 4.1-1 INSTRlNENT SURVEILLANCE REQUIREMENTS 4-3 4.1-2 MINIMLN EQUIPMENT TEST FREQUENCY 4-8 ~ 4.1-3 MINIMLN SAMPLING FREQUENCY 4-9 4.2-1 CAPSULE ASSEMBLY WITHDRAWAL SCHEDULE AT DAVIS-BESSE 1 4-12b Amendment No. 2 6,53,66,68,76,77,f0 ' iX 85,81,88,95,97

r RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS LIST OF TABLES Section Eage 4.6-1 DIESEL GENERATOR TEST SCHEDULE 4-34j 4.6-2 ADDITIONAL RELIABILITY ACTIONS 4-34k 4.14-1 SNUBBERS ACCESSIBLE DURING POWER OPERATIONS 4-47c 4.17-1 MINIRN NLfEER OF STEAM GENERATORS TO BE 4-56 INSPECTED DURING INSERVICE INSPECTION 4.17-2A STE41 GENERATOR TUBE INSPECTION 4-57 4.17-2B STEAM GENERATOR TUBE INSPECTION (SPECIFIC LIMITED AREA) 4-57a 4.17-3 OTSG AUXILIARY FEEDWATER HEADER SURVEILLANCE 4-57b,c 4.19-1 RADI0 ACTIVE LIUQID EFFLUENT M0hlTORING INSTRtNENTATION 4-64 SURVEILLANCE REQUIREMENTS 4.20-1 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRlNENTATION 4-66 SURVEILLANCEREQUIREMFETS 4-21-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AfD ANALYSIS PROGRAM 4-70 4.22-1 RADI0 ACTIVE GASE0US WASTE SAMPLING AND ANALYSIS PROGRAM 4-75 4.26-1 MAXIHN VALUES FOR THE LOWER LIMITS OF DETECTION (LLD) 4-84 4.28-1 EXPLOSIVE GAS MIXTURE INSTRtNENTATION SURVEILLANCE 4-88 REQUIREMENTS 4.34-1 METER 0 LOGICAL MONITORING INSTRlNENTATION 4-94 6.2-1 MINIMLN SHIFT CREW COMPOSITION 6-2 d Amen sen: n. n, m is,is, / x m ,_m._ ,. ~.. _ _ _. - _ _ _ _ _ _ _ _.

. RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS l 1. DEFINITIONS The following terms are defined for uniform interpretation of these specifications. 1.1 RATED POWER Rated power is a sterdy reactor core output of 2772 MHt. 1.2 REACTOR OPERATING CONDITIOdS (OPERATIONAL MODE OR MODE) An operational mode or mode shall correspond to any one inclusive co:nbination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.2-1. 1.2.1 Cold Shutdown The reactor is in the cold shutdown condition when k fr is 10.99, and Ta g is no more than 200*F. Pressure is defined by Specification 3.1.2. See Table 1.2-1 1.2.2 Hot Shutdown The reactor is in the hot shutdown condition when keff is 1 0.99, and Tavg is at or greater than 525'F. See Table 1.2-1. 1.2.3 Reactor Critical The reactor is critical when t'he neutron chain reaction is self-sustaining ~ ' and keff > 0.99. l 1.2.4 Hot Standby The reactor is in the hot standby condition when all of the following conditions exist: (See Table 1.2-1) l A. Tavg is greater than 525'F. B. keff is >0.99. l C. Indicated neutron power on the power range channels is less than 2 percent of rated power. 1.2.f Power Ooeration The reactor is in a power operation condition when the indicated neutron power is above 2 percent of rated power as indicated on the power range channels. See Table 1.2-1. 1.2.6 Refueling Shutdown The reactor is in the refueling shutdown condition when the reactor core is 1 0.95 and the coolant temperature at the decay heat reactivity, keff, ion is no more than 140'F. Pressure is defined by removal pump suct Specification 3.1.2. A refueling shutdown refers to a shutdown. to replace or rearrange all or a portion of the fuel assemblies and/or control rods. l See Table 1.2-1. Amendment No. 97 1-1

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions 1.2.7 Refuelino Ooeration An operation involving a change in core geometry by manipulation of fuel or control rods when the reactor vessel head is removed. 1.2.8 Refuelina Interval

  • 18 months.

1.2.9 Startuo 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 normal hot shutdown procedure will be completed within 12 hours.unless otherwise specified. 1.2.11 Tg At operating conditions Tavg is defined as the arithmetic average of the coolant temperatures in the hot and cold legs of the loop with the greater i number of reactor coolant pumps operating, if such a distinction of loops can be made. 1.2.12 Heatuo - Cooldown The reactor is in heatup-cooldown when the range of reactor coolant temperature is greater than 200*F and less than 525'F. 1.2.13 Action Action including time requirements shall be that part of a specification which prescribes remedial measures required under designated conditions. 1.2.14 Leikagg A. IDENTIFIED LEAKAGE shall be: a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or 'See Page 1-2b Amendment No. 38,6I,8/,WA,97 l-2

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions c. Reactor coolant system leakage through a steam generator to the secondary system. B. UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE. C. PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component' body, pipe wall or vessel wall. D. CONTROLLED LEAKAGE shall be that leakage from the reactor coolant pump seals. 1.3 OPERABLE A component or system is operable when it is capable of performing its intended function within the required range. The component or system shall be considered to have this capability when: (1) it satisfies the limiting conditions for operation defined in Specification 3, (2) it has been tested periodically in accordance with Specification 4, and has met its performance requirements, (3) the system has available its normal and emergency sources of power, and (4) its required auxiliaries are capable of performing their intended function. When a system or component is determined to be inoperable solely because its normal power. source 1.s inoperable or its emergency power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation provided its redundant system or component is OPERABLE with an OPERABLE normal and emergency power source. 1.4 PROTECTION INSTRUMENTATION LOGIC l 1.4.1 Instrument Channel l l An instrument channel is the combination of sensor, wires, amplifiers and ouput devices which are connected for the purpose of measuring the value of a process variable for the purpose of observation, control and/or protection. An. instrument channel may be either analog or digital. 1.4.2 Reactor Protection System l The reactor protection system is shown in Figures 7.1-1 and 7.2-2 of the FSAR. It-is that combination of protective channels and associated l circuitry which forms the automatic system that protects the reactor by control rod trip. It includes the four protection channels, their associated instrument channel inputs, manual trip switch, all rod driv 9 control protective trip breakers and activating relays or coils. l l l Amendment No. SI,87,97 ,1-2a . ~

TABLE 1.2-1 OPERATIONAL MODES Operational Reactivitly Coolant Indicated Remarks Mode Condition Temperature Neutron keff Tavg*F Power (1 of Rated Power) Power >0.99 >525 12 Operation Hot >0.99 >525 <2 Standby Startup >0.99 >525 <2 <1.00 Hot 10.99 1525 0 Shutdown Heatup - 10.99 >200 0 Cooldown <525 Cold 10.99 1200 0 RCS Pressure as Shutdown defined in 3.1.2. Refueling

  • 10.95 1140 at 0

RCS Pressure Shutdown DHR pump, as defined suction in 3.1.2. Refueling Reactor Vessel head removed and Refueling Shutdown conditions operation

  • Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

4. Amendment No. 97 l-2c

~ RANCHO SECO.UNTI 1 TECHNICAL SPECIFICATIONS Limiting Conditiors for Operation 3. LIMITING CONDITIONS FOR OPERATION 3.0 General Limiting Conditions For Operation 3.0.1 Compliance with the Limiting Conditions for Operatio'n contained in the succeeding Specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Opration, the associated Action including time requirements shall be met. 3.0.2 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated Action including time requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the Action including time requirements is not required. 3.0.3 Hhen a Limiting Condition for Operation is not met, except as provided in the associated Action including time requirements, within I hour action shall be initiated to place the unit in a H00E in which the Specification does not apply to placing it, as applicable, in: 1. At least HOT STANDBY'within the next 6 hours, i 2. At least HOT SHUTDOHN within the following 6 hours, and l 3. At least COLD SHUTDOHN within the subsequent 24 hours. Where corrective measures are completed that permit operation I under the Action including time requirements, the Action including time requirement may be taken in accordance with the specified time limits as measured from the time of fai'ure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications and Table 3.0-1. 3.0.4 Entry into an OPERATIONAL N00E or other specified condition j shall not be made when the conditions for the Limiting Conditions for Operation are not met and the associated Action including time requires a shutdown if they are not met within a specified time interval. Entry.into an OPERATIONAL N00E or specified conditten may be made in accordance with Action including time requirements when conformance to them permits continued operation of the facility for an unlimited period of time. This provisio'n shall not prevent passage through or to' OPERATIONAL H0 DES as required to comply with Action requirements. Exceptions to these requirements are stated in the individual specifications and Table 3.01. Amendment No. 97 3-0 1 w -. =. - -

TABLE 3.0-1 Applicability of Specifications 3.0.3 and 3.0.4. (The "NA" indicates that the provisions of Specification (s) 3.0.3 and/or 3.0.4 are not applicable to the sections identi fi ed). Secti,on Specification 3.0.3 Specification 3.0.4 3.1 3.2.1 3.2.2 N/A 3.3 3.4 3.5.1 3.5.2 3.3.3 3.5.4 N/A N/A 3.5.5 N/A 3.5.6 N/A N/A 3.5.7 N/A 3.6 N/A 3.7 3.8 N/A 3.9 N/A ' N'/A 3.10 3.11 N/A N/A 3.12 3.13 3.14 N/A N/A l 3.15 W/A N/A l 3.16 N/A N/A 3.17 N/A N/A 3.18 N/A N/A-3.19 N/A N/A 3.20 N/A N/A l 3.21 N/A N/A 3.22 N/A N/A l 3.23 N/A N/A l 3.24 N/A N/A l 3.25 N/A N/A 3.26 N/A N/A

  1. The provisions of Specification 3.0.3 are not cpplicable when the reactor is in Refueling Shutdown or Refueling Operation.

l Amendment No. 97 3-Oa

k RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1 REACTOR COOLANT SYSTEM L Aeolicability' -All operational modes. -Ob.iective To specify those limiting conditions for operation of the reactor coolant system which must'be met to ensure safe reactor operations. 3.1.1 OPERATI'ONAL COMPONENTS Soecification 3.1.1.1 ' Reactor Coolant Pumps L A. Pump combinations permissible for given power levels shall be as shown in 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 d: cay heat removal pump is circulating reactor coolant. C. Operation at power with two pumps shall be limited to 24 hours-in any 30 day period. D. At Isast one RCP shall be in operation when reactor coolant temperature is equal to or greater than 280*F. 3.1.1.2 Steam Generator A. Two steam generators shall be operable whenever the reactor coolant average temperature is above 280 F, except as described e in 3.1.1.2.2. 8. With one or more steam generator (s) inoperable due to excessive i leakage per 3.1.6.g. bring the reactcr to cold shutdown conditions within 48 hours. C. With one or more steam generator (s) inoperable due to steam . generator defective tube (s), restore the inoperable generator (s) to operable status prior to increasing reactor coolant average temperature above 200*F. 3.1.1.3 Prersurizer Swfety Yalves A. The reactor shall not remain critical unless both Pressurizer Coolant System code safety valves are operable. 8. When the reactor is subcritical, at least one Pressuriier cede safety valve shall be operable if all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Boiler and Pressure Vessel Code, Section III. 3-1 Amendment No.1,31,77,E7,13,97

l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Action In COLD SHUTOOWN with no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place.an OPERABLE decay heat removal loop into operation in the shutdown cooling mode. When above the hot shutdown condition, with one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minuti,s or be in a least HOT SHUTDOWN within 6 hours. 3.1.1.4 Pressurizer Electromatic Relief Valve A. The nominal setpoint of the pressurizer electromatic relief valve shall be 2450 psig z 10 psig except when reouired for cold overpressure protection. b. If the EHOV and its associated block valve are not OPERABLE whenever the reactor is in HOT STANDBY or critical, the' following actions shall be taken: 1. Hith the EHOV itoperable, witisin 60 minutes either restore I the EHOV to OPERABLE status or close the associated block valve and remove power from.the block valve; otherwise, be in a least HOT STANDBY within the next 6 hours and in. COLD SHUTDOWN within the following 30 hours. The requirement for an operable low pressure setpoint for the EHOV for LTOP in Specification 3.2.2 is not applicable if the EHOV is inoperable. 3.1.1.5 Decay Heat Removal A. At least two of the coolant loops listed below shall be operable except during fuel loading and refueling. One loop shall be in operation when the coolant average temperature is below 280*F. The one operating coolant loop required need not be in operation for a maximum of one hour provided (1) no operatiens are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature. 1. Rear. tor Coolant Loop (A) and !ts associated stehm generator and at least one associated reactor coolant

pump, 2.

Reactor Coolant Loop (B) and its associsted steam generator and at least one associated reactor coolant ptimp, Amendmenc No. 4,U,77,59,93,97 3-la

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3. Decay Heat Removal Loop (A) 4. Decay Heat Removal Loop (B) B. With less than the above required coolant loops OPERABLE, l immediately initiate corrective action to return the required coolant loors tr0PERABLE status as soon as possible; he in COLD Shul 00HN within 20 hours. 3.1.1.6 Reactor Coolant System High Point Vents A. The vent path on Loop A and vent path on Loop B shr.11 be operable and closed during po.ser operation. B. The vent path on the pressurizer shall be operable and closed during power operation. C. With one of the above reactor coolant system vent paths inoperable, STA2 TUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with l l power removed from the valve actuator of all the valves in the inoperable vent path; restore the inopersble vent path to OPERABLE status within 30 days. If the status is not restored to operable in 30 days, be in HOT STANDBY within 12 hours Nd l in COLD SHUTDOHN within the following 30 hours. D. Hith two or more of the above reactor coolant system vent paths inop?rable; maintain the inoperable vent paths closed with power removed from the valve actuators of all the valves in the inoperable vent paths, and restere at ler.st (two) of the verst paths to OPERABLE status within 72 hours. If the status is not restored to operable in 72 hours, be in HOT STANDSY within 12 hours and in COLD SHUTDOWN within the l following 30 hours. 1 1 I l l l l Amendment No. J.JJ,7%,$9,9),97 3-2

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.2 PRESSURIZATION, HEATUP, AND C00LDOWN LIMITATIONS Foecification 3.1,.2.1 Inservice Leak and Hydrostatic Testsi Pressure temperature limits for the first eight Effective Full Power Years (EFPY) of inservice leak and hydrostatic tests are given in Figure 3.1.2-3. Heatup and cooldown rates shall be restricted according to the rates specified in Figure 3.1.2-3. 3.1.2.2 Heatuo Cooldown: For the first eight EFP years of power operations, 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. The Reactor Coolant System temperature and pr. essure chall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. Heatup and cooldown rates shall not exceed the rates stated on the associated figure. 'dM ~ If heatup and cooldown rates are exceeded, stabilize the temperature and restr.re the temperature and/or pressure to within the limits within 30 minutes, and perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System. Through this evaluation, determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT SHUTDOHN within the next 6 hours and reduce RCS Tavg and pressure to less than 200'F and 500 psig, respectively, within the following 30 hours. This action applies to Specifications 3.1.2.4 and 3.1.2.5, below. 3.1.2.3 The secondary side of the steam generator shall not be pressurized above 200 psig if the temperature of the steam generator shell is below 130*F. 3.1.2.4 The pressurizer heatup and cooldnwn 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 410*F. F Amendment No. 33,3),5),97,97 3-3

RANCl!O SECO UNIT 1 TECHNICAL SPECIFICATIONS limiting Conditions for Operation 3.1.3 HINIMUM CONDITIONS FOR CRITICALITY Sp1cifjeations 3.1.3.1 The reactor coolant temperature shall be above 525'F except for portions of low power physics testing when the requirements of Specification 3.1.8 shall apply. 3.1.3.2 Reactor coolant temperature shall be above Dtctility Transition Temperature (DTT) + 10 F. 3.1.3.3 When the reactor coolant temperature is below the minimum temperature specified in 3.1.3.1 above, except for portions of low power physics testing when the requirements of Specificatien 3.1.8 shall apply, the reactor shall be subcritical by an amount equal to or greater than the calculated reactivity insertion due to depressurization. Action Hith the reactor suberitical by less than the required amount, immediately initiate and continue boration until the required SHUTDOWN MARGIN is restored. 3.1.3.4 The reactor shall be maintained subcritical by at least 1 percent Ak/k until a steam bubble is formed and an indicated water level between 10 and 316 inches is established in the pressurizer. Action Hith the reactor subcritical by less than the required amount. immediately initiate and continue boration until the required SHUTDOWN HARGIN is restored. 3.1.3.5 Except for physics tests and as limited by 3.5.2.1 and 3.5.2.5, safety rod groups shall be fully withdrawn prior to any other reduction in shutdown margin by deboration or regulating rod withdrawal during the approach to criticality. Following safety rod withdrawal, the regulating rods shall be positioned within their position limits as defined by specification 3.5.2.5 prior to deboration. BAlti At the beginning of life of the initial fuel cycle, the moderator temperature coefficient is expected to be slightly positive at operating temperatures with the operating configuration of control rods. (1) Calculations show that above 525'F the positive moderator coefficient is acceptable. Since the moderator temperature coefficient at lower temperaturee will be less negative or more positive than at operating temperature, (2) startup and operation of the reactor when reactor coolant. temperature is less than 525'F is prohibited except where necessary for low power physics tests. Aur.d:sent No. 3,87,97 3-6

R.',NCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Bases (Continued) The potential reactivity insertion due to the moderator pressure coefficient .(2) that could result from depressurizing the coolant from 2185 psia to saturation pressure of 885 psia is approximately 0.1 percent Ak/k. During physics tests, special operating precautions will be taken. In addition, the strong negative Doppler coefficient (1) and the small integrated Ak/k would limit the magnituda of a power excursion resulting-from a reduction of moderator density. The requirement that the reactor is not to be made critical below DTT + 10 F provides increased assurances that the proper relationship between primary coolant pressure and temperatures will be maintained relative to the NOTT of the primary coolant system. Heatup to this to perature will be accomplished by operating the reactor coolant pumps. The DTT at Beginning,of Life (BOL) for the most ilmiting component in the reactor cuolant system is less than j +100 F. If the shutdown margin required by Specification 3.5.2 is maintained, thare is no possih111ty of an accidental criticality as a result of a decrease of coolant pressure. The requirement for pressurizer bubble fonnation and speciffod water level when the reactor is Tess than 1 percent suxritical will assure that the reactor coolant system cannot become solid in the event of a rod withdrewal accident or a start-up accident and that the water level is above the minimum detectable level. The requirement that the safety rod groups be fully withdrawn before criticality ensures shutdown capability during startup. This does not prohibit rod latch confinnation, f.e., withdrawal by group to a maximum of 3 ~ inches withdrawn of all seven groups prior to safety rod withdrawal. REFERENCES (1) USAR, section 3 (2) USAR, paragraph 3.2.1.4 l l Amendment No, d/,97 3-7 x. n

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Cor.ditions for Operation 3.1.4 REACTOR COOLANT SYSTEM ACTIVITY Snecification 3.1.4.1 The specific activity of the reactor coolant due to nuclides with l half lives longer than 30 minutes shall not exceed 43/E microcuries per gm whenever the reactor is critical. E is the average (mean) beta and gamma energies per disintegration, in MeV, weighted in proportion to the measured activity of the radionuclides in reactor coolant samples. Action Hith the specific activity of the reactor coolant greater than 43/E microcuries/ gram, be in at least HOT SHUTDOHN within 6 hours. BAlti The above specification is based on limiting the consequences of a postulated accident involving the double-ended rupture of a steam generator tube. The rupture of a steam generator tube enables reactor coolant and its associated activity to enter the secondary system where volatile isotopes could be discharged to the atmosphere through condenser air-ejectors and through steam safety valves (which may lift momentarily). Since the major portion of the activity entering the secondary system is due to noble gases, the bulk of the activity would be discharged to the atmosphere. The activity release continues until the operator stops the leakage by reducing the reactor coolant system pressure below the set point of the steam safety valves and isclates the faulty steam generator. The operator can identify a faulty steam generator by using the off-gas monitors on the condenser air ejector lines; thus he can isolate the faulty steam generator within 34 minutes after thg tube break occurred. During that 34 minute period, a maximum of 2740 ft3 of hot reactor coolant will have leaked into the secondary system; s this is equivalent to a cold volume of 1980 ft3 The controlling dose for the steam generator tube rupture accident is the whole-body dose resulting from immersion in the cloud of released activity. To insure that the public is adequately protected, the specific activity of the reactor coolant will be limited to a value which will insure that the whole-body annual dose at the site boundary will not exceed 0.5 rem, the limit in 10 CFR Part 20 for whole body dose in an unrestricted area. Although only volatile isotopes will be released from the secondary system, the following whole-body dose calculation conservatively assumes that all of the radioactivity which enters the secondary system with the reactor coolant is released to the atmosphere. Both the beta and gamma radiation from these isotopes contribute to the whole-body dose. The gamma dose is dependent ori the finite size and configuration of the cloud. However, the analysis employs the simple model of a semi-infinite cloud, which gives an upper limit to the potential gamma dose. The semi-infinite cloud model is applicable to the beta dose because of the short range of beta radiation in air. It is further assumed that meteorological conditions during the course of the accident correspond to Pasquill Type F and 0.6 meter per second wind speed, resulting in a X/Q value of 8.51 x 10-4 sec/m3 3-8 Amendment No. $1,97

i J, RANCHO SECO UNIT 1-TECHNICAL SPECIFICATIONS Limiting Conditions for Operation, [erformance discharge test may be performed in lieu of the sattery service test required by Emergency Power System Periodic Testing Specification 4.6.4.d provided that the performance discharge test is performed in the "as-found" condition. f. Each vital 125 volt DC and vital 120 volt AC bus listed in Specifications 3.7.1H and I shall be determined OPERABLE and ener'ized at least once per 7 days by verifying correct g breaker alignment and indicated power availability with an overall voltage of greater than or equal to 125 volts DC and 120 volts AC, respectively. 4 Amendment No. 97 4-341 y

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS ~ p, Surveillance Standards ELECTRIC POWER' SYSTEMS Table 4.6-1 DIESEL GENERATOR TEST SCHEDULE Number of Failure in Last-20 Valid Tests

  • Test Frecuency 1

At least MONTHLY 2 At least HEEXLY** t

  • Criteria for determining number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, Revision 1 August 1977, where the number of tests and failures is determined on a per diesel generator basis.

For the purposes of this test schedule, only valid tests conducted after the license amendment issuance date shall be included in the computation of the "last 20 valid tests."

    • This test frequency shall be maintained until seven consecutive failure free demands have been performed and the number of failures in the last 20 valid demands has been reduced to one or less.

( Amendment No. 94,97 4 - 3 43 l

~- RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards ^ TABLE 4.6-2 ~ ADDITIONAL RELIABILITY ACTIONS No. of failures No. of failures in last 20 in last 100 valid test

  • valid tests Action 3

6 Within 14 days prepare and maintain a report for NRC audit describing the diesel generator reliability improvement program implemented at the site. Minimum requirements for the report are indicated in Note 1 to this table. 5 11 Perform a r'equalification test program for the affected diesel generator. Requalification test program requirements are indicated in Note 2 to this table. l

  • For the purposes of this schedule, only valid tests conducted after l

the license amendmont issuance date shall be included in the computation of the "last 10 valid tests" and "last 100 valid tests." I Amendment No. 9'4,97 4-34k / l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS ( Surveillance Standards NOTE 1 TO TABLE 4.6-2 REPORTING REOUIREMENT As a minimum the Reliability Improvement Program report for NRC audit shall include: a) a summary of all tests (valid and invalid) that occurred within the time period over which the last 20/100 valid tests were performed. b) analysis of failures and determination of root causes of failures c) evaluation of~each of the recommendations of NUREG/CR-0660, "Enhancement of Onsite Emergency Olesel Generator Reliability in-Operating Reactors," with respect to their application to the Plant d) identification of all actions taken or to be taken to 1) correct the root causes of failures defined in b) above and 2) achieve a general improvement of diesel generator reliability e) the schedule for implementation of each action from d) above ( f) an assessment of the existing reliability of electric power to engineered-safety-feature equipment Once a licensee has prepared and maintain an initial report detailing the diesel generator reliability improvement program at his site, as defined above, the licensee need prepare only a supplemental report within 14 days after each failure during a valid demand for so long as the affected diesel generator unit continues to violate the criteria (3/20 or 6/100) for the reliability improvement program remedial action. The supplemental report ~ need only update the failure / demand history for the affected diesel generator unit since the last report for that diesel generator. The supplemental report shall also present an analysis of the failure (s) with a root cause determination, if possible, and shall delineate any further procedural, hardware or operational changes to be incorporated into the site diesel generator improvement program and the schedule for implementation of those changes. In addition to the above, submit a yearly data report on the diesel generator reliability. e ( Amendment! No. N,97 ~ 4-341

e j RANCHO SECO UNIT 1 1 TECHNICAL SPECIFICATIONS Surveillance Standards NOTE 2 TO TABLE 4.6-2 DIESEL GENERATOR REOUALIFICATION PROGRAM (1) Perform seven consecutive successful' demands without a failure within 30 days of diesel generator being restored to operable status from the last failure and fourteen consecutive successful demands without a failure within 75 days of diesel generator being restored to operable status from the last failure. (2) If one failure occurs during the first seven tests in the 1 requalification test program, perform seven successful demands without an additional failure within 30 days of diesel generato'r being restored to operable status after the first failure of the requalification test program and fourteen consecutive successful demarJs without a failure within 75 days of being restored to operable status after the first failure of the requalification test program. (3) If o.no failure occurs'during the second seven tests (tests 8 through

14) of (1) above, perform fourteen consecutive successful demands without an additional failure within 75 days of the failure wl)ich occurred during the requalification testing.

(4) Following the second failure during the requalification test progras declare the diesel generator inoperable and follow the ap'plicable requirements of Specification 3.7.2. The diesel generator shall not be considered operable until the diesel generator has successfully requalified. (5) During requalification testing the diesel generator should not be i' tested more frequently than at 24-hour intervals each. After a diesel genrator'has been successfully requalified, subsequent repeated requalification tests will not be required for that diesel generator under the following conditions: (a) The number of failures in the last 20 valid tes.ts is less than S. (b) The number of failures in the last 100 valid tests is less than 11. (c) In the event that following successful requalification of a diesel generator, the number of failures is still in excess of the remedial l action criteria (a and/or b above) the following exception will be allowed untti the diesel generator is no longer in violation of the remedial action criteria (a and/or b above). l Requalifiction testing will not be required provided th'at after each valid tests the number of failures in the last 20 and/or 100 valid tests has not j increased. Once the diesel generator is no longer in violation of the remedial action criteria above the provisions of those criteria alone will prevail. Amend:nent No. M,97 4-34m ~ _n,--,--w,x-e-,,r,,,-w nwm,r,-.- e- -wn..

Reiesued 3 n5/88 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards. BALt1 The operability of the offsite circuits, the nuclear service 4160V buses, and the nuclear service 480V buses are demonstrated by veriYying correct breaker alignments and indicated power availability. Surveillance 4.6.1,B can only be performed during a refueling shutdown when both diesel generator trains are operable or the core is flooded to 37 feet to ensure the required decay heat removal capability is available. 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, and upon receipt of a safety features actuation signal. The testing frequency specified is intended to identify before it can result in a system failure. The fuel oil supply, starting circuits and controls are continuously monitored and any faults are alarmed and indicated. An abnormal condition in these systems would be signaled without having to place the diesel generato*rs on test. The limiting of the maximum load on the TDI diesel generators A2 and 82. to .less than the qualified load of 3300 kw provides assurance that the crankshafts will stay within the proven limits for high-cycle fatigue cracks. Diesel generators A2 and 82 will be loaded during surveillance testing between 3000 and 3300 KH which provides assurance that the qualified load of 3300 KH will not be exceeded. At least once per refueling outage a diesel generator test will be performed simulating a loss of off-site power in conjunction with a simulated SFAS and l loading of actual loads to the maximum extent possible without damaging plant systems (i.e., use of recirculation flow or manual valving out a i systcm to protect plant components). Precipitous failure of the plant battery is extremely unlikely. The surveillance specified is that which has been demonstrated over the years to provide an indication of a cell becoming unserviceable long before it fails. j REFERENCE (1) IEEE 308 ( I Amendment No. 31,68,74,87,94 4-35 ""--*'-9 w m-

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Soecification (Continued) 4.10.1 C. 2. After reinstallation of the sampled adsorber tray, per C.1: (a) Verify that the charcoal adsorbers remove 199.5 percent of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510 while operating the filter train at a stabilized flow rate of at least 2880 CFM but no greater than 3520 CFM. (b) Verify that the HEPA filter bank removes 199.9 percent of the DOP when tested in-place in accordance with ANSI H510 while opera. ting the filter train at a stabilized flow rate of at least 2880 CFM but no greater than 3520 CFM. l 0. Started on a manual signal and operated for 15 minutes in each 31-day period. E. Demonstrated operable at least once per refueling interval or i once every 18 months, whichever occurs first, by: 1. Verifying that on a high radiation signal, the normal system is gutomatically isolated and that the emergency filtering system operates. I 2. Verifying that or, a toxic gas (ammonia excluded) isolation signal, the system automatically switches into the isolation mode of operation with flow through the HEPA filters and charcoal adsorbar banks (only required if the total quantity of gaseous chlorine in the RESTRICTED AREA exceeds 100 pounds). 3. Verifying that the refrigerant system will maintain the Control Room / Technical Support Center temperature at no more i than 80*F for at least 8 hours. Bases The purpose of the Control Room /TSC Emergency Filtering System is to limit the particulate and gaseous fission products and toxic products to which the Control Room area and Technical Support Center would be subjected during an accidental radioactive or chemical release in or near the Auxiliary Building. The system is designed with two redundant filter trains each of which consists of a moisture separator, a heater, a high efficiency particulate filter, two banks of charcoal filters, a second high efficiency particulate filter and a booster fan to pressurize the Control Room and Technical Support Center with outside air. The prestabilized air flow rate will not exceed 3750 CFH during the 15 minutes preceding stabilized air flow. Amendment No. 39, 70, 91, H, 97 4-41b

Rainued 3/15/88 RANCHO SECO UNIT 1 TECHNICAL. SPECIFICATIONS Surveillance Standards Specification (Continued) Because this system is not normally operated, a periodic test is required to ensure its operability when needed. Monthly testing of this system will show that the system is available for its designed safety action. During this test the system will be observed for unusual or excessive noise or vibration when the fan motors are running. The flow of 1760 cfm makeup air was selected to limit the maximum radiation dose to occupants of the Control Room /TSC in an accident.. For this analysis, each 2-inch charcoal filter was conservatively assumed to provide a DF of 10 for iodine ba' sed on laboratory analysis showing >95 percent removal of radioactive methyl iodide. The aggregate 4-inch charcoal depth of the two beds in series provides an overall iodine DF cf 100. The HEPA filter is assumed to provide a DF of 100 for particulates. Refueling interval testing will verify the methyl iodide removal efficiency of the charcoal and the amount of leakage past the charcoal and HEPA filters a a at least equal to the design values. The filtering system is automatically started and the normal system isolated when tha radiation level or when the Chlorine level increase. l The testing required af ter painting, f' ire or chemical release, is not to be interpreted to include minor touch-up painting, housekeeping chemicals and detergents..or other routine maintenance or housekeeping activities. Amendment No. 39,70,91,94 4-41c 4-M

1 r \\ RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.9 REPORTING REQUIREMENTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted. Startup Report 6.9.1.1 A sumary report of plant startup and power escalation testing shall be submitted following (1) Receipt of an operating license; (2) amendment to the license involving a planned increase in power level; (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier; and (4) modifications that may have significantly altered the nuclear, thermal or hydraulic performance of the plant. The report shall address each of the tests identified in the USAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other comit;nents shall be included in this report. 6.9.1.2 Startup reports shall be submitted within (1) Ninety (90) days following completion of the startup test program; (2) Ninety (90) days following resumption or comencement of commercial power operation; or (3) Hine (9) months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three (3) months until all three events have been completed. 6.9.l.3 Prior to exceeding eight effective full power years of operations, Figures 3.1.2-1, -2, and -3 shall be updated for the next service l 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 i with Specification 6.9.1.4. 6.9.1.4 The updated proposed technical specifications referred to in 6.9.1.3 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 specifications submitted in l accordance with 10 CFR 50, Appendix G. Section V.C. Amendment No. U,87,9V 97 6-12 nan ,~ w w ~ ww- ~- v~~ ~ -~c- -~ ~

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls s Soecial Reoorts 6.9.5 Special reports shall be submitted to the Regional Administrator, Region V Office, within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification: A. A one-time only, "Narrative Summary of Operating Experience" will be submitted to cover the transition period (calendar year 1977). B. A Reactor Building structural integrity report shall be submitted within ninety (90) days of completion of each of the following tests covered by Technical Specification 4.4.2 (the integrated leak rate test is covered in Technical Specification 4.4.1.1). 1. Annual Inspection 2. Tendon Stress Surveillance 3. End Anchorage Concrete Surveillance 4. Liner Plate Surveillance C. In-Service Inspection Program D. Inoperable Accident Monitoring Instrumentation 30 days l' E. Status of Inoperable Fire Protection Equipment (3.5.5) 30 days (3.14.1.2, 3.14.2.2, 3.14.3.2, 3.14.2.3, 3.14.5.2, 3.14.4.2, l 3.14.6.2) i i F. Inoperable Emergency Control Room /TSC j Ventilation Room Filter System ll G. Radioactive Liquid Effluent Oose 30 days (3.17.2) H. Noble Gas Limits 30 days (3.18.2) I. Radiolodine and Particulates 30 days (3.18.3) i L Amendment No. 80, 25, 97 6-12f - -}}

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