Text

Errata to Amends 97 & 98 to License DPR-54,correcting Inadvertent Errors in Tech Specs
	ML20151B323
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Site:	Rancho Seco
Issue date:	03/30/1988
From:	; Office of Nuclear Reactor Regulation
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ML20151B322	List: ML20151B320; ML20151B323;
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i RANCHO SECO UNIT 1 itCHNICAl. SPECIFICATIONS Revised 3/30/88 TABLE OF CONTENTS (Continued) e

$ action Ein 3.14.2 Water iGian 3-53 3.14.3 Sorav and serinkler Sntes 3-g 3.14.4 Qb satan 3-g 3.14.5 Fire Hete Statteti 3-57 3.14.6 Fire Barrier Penetration Fire Sealt 3-58 3,15 rad 10ACTfVE Lf00fD EFFLUENT HON!TORING INSTRUw!NTATION 3-60 l 3.16 RADICACTIVE CASEOUS EFFLUENT WONITORING INSTRUwtNTA,I1Qg 3-63 3.17 Lf00fD EFFLUENTS 3-70 3.17,1 Ceacentration 3-70 3.17.2 D211 3-71 3.17.3 4jeuid Heldue Tankt 3-72

, 3.17.4 Lieuid Effluent Radwatte Treatment 3-72a [ ,

3.18 Ca5E005 EFFLUENTS 3-73 3.18.1 Dele Rate 3 73 3.18.2 Dete-Neble Catet 3-74 3.18.3 Dete-fed!re-131. Iodine-133.TritiuE.andRadicactive Materiali in Particulate Fora 3-75 3.18.4 Cateeut Radvatte Treatment

3-78 3.18.5 Cat sternae Tankt 3-79 3.19 DELETED 3.20 DELETED 3.21 SOLID RADIDACTIVE WASTER 3-80 3.22 RADIOLOGICAL ENVIRONWENTAL MON!TORING 3-81 3.23 LAND USE CENSUt 3-87 3.24 EXPLC3fVE CAS MIXTURE 3-89 3.25 EUEL CYCLE Dost 3-90 3.26 IEIEDL A! ORATORY COMPARIs0N PROC 1 TAM 3-52 3.27 IfuCLEAR SERVICE ELEttifCAL Eff LDING EMERCENCY MEAffMC 3-93 AND AIR CONDITIONIMC 3.28.

TDI DIESEL CENERATot CONTROL adoM ft1EMTfat 3-93a VENTILAT10N SY1TDl 3.29 METEOROLOGICAt. n0NITot!NG INSTRUMENTAfl0K 3 94 3.30 HYOROGEN RECOMBINER$ 3 96 8804000167 000330 A endment Ne, 53.05.93 iv PDR ADOCK 05000312 p PDR

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RANCHO SECO LMIT 1 gevised 3/30/88 TECHNICAL SPECIFICATIONS LIST OF TABLES Section Etqe 1.2-1 8PERATI0fML t00ES 1-2c l 2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTIfG LIMITS 2-9 3.0-1 APPLICABILITY OF SPECIFICATIONS 3.0.3 AfD 3.0.4 3-Oa 3.3-1 PRIf%RY COOLANT SYSTEM PRESSliRE ISOLATION VALVES 3-22a 3.5.1-1 INSTRLNENTS OPERATIfG C0f0!TIONS 3-27 3.5.5-1 ACCIDENT MTORIfG INSTRlNENTATION OPERABILITY RE@!REMENTS 3-38e l 3.5.7-1 EMERGENCY SHIJTDOWN INSTRLNENTATION 3-38d l 3.6-1 SAFETY FEATURES CONTAlfNENT ISOLATION VALVES 3-40 3.7-1 VOLTAGE PROTECTION S.YSTEM RELAY TRIP VALUES 3-41a 3.7-2 VOLTAGE PROTECTION SYSTEM LIMITIfG C0f0ITIONS 3-43b ll 3.12-1 SAFETY RELATED HYPAULIC SNUB 8ERS 3-51a-e 3.14-1 FIRE DETECTION INSTRLMENTS FOR SAFETY SYSTEMS 3-55 3.14-2 WATER SUPPRESS:0N ZONES 3-56b,c 1

3.14-3 CARDCN DIOXIDE SUPPRESSION ZONES 3-56d,e 3-14-4 FIRE HOSE STATIONS 3-57a,b 3.15-1 RADI0 ACTIVE LI@!D EFFLUENT.t0NITORIfG INSTRlNENTATION 3-61 3.16-1 RADI0 ACTIVE CASES EFFLUENT MTORIfG INSTRLHENTATION 3-64 3,22-1 RADIOLOGICAL ENVIR0fNENTAL MONITOR!fG PROGRH 3-83 !

3.22-2 REPORTIfG LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS 3-86 IN ENVIR0tNENTAL S.WLES 3.29-1 ETER0 LOGICAL 10NITORIfG INSTRlNENTATION 3-95 4.0 APPLICABILITY OF SPECIFICATIONS 4.0.2 Af0 4.0.3 4-Oa 1 4.1-1 INSTRLNENT SURVEILLNCE RE@IREMENTS 4-3 4.1-2 MINIRN E@lPENT TEST FRE@ENCY 4-8 4.1-3 MINIHN S/WLIfG FRE@ENCY 4-9 4.2-1 CAPSULE ASSENLY WITHDRAWAL SCHEDULE AT DAVIS-BESSE 1 4-12b 4

Junendment No. 26.53.66.68.76.11.88 IX

, 85.87.88.98.97 O

i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Revised 3/30/88 Limiting Conditions for Operktion 3.1.4 IEACTORCOOLANTSYSTEMACTIVITY I igacification l i 3.1.4.1 The specific activity of the reactor coolant due to nuclides with '

half lives longer than 20 minutes shall not exceed 43/E K1crocuries ll 1

per go whenever the reactor is critical. E is the average (mean) aeta and ganna energies per disintegration, in MeV, weighted in i proportion to the measured activity of the radionuclides in i reactor coolant samples.

Action With the specific activity of the reactor coolant qreater than 43/l micrecuries/ gram, be in at least HOT SHUTDOWN with n 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , jl g l The above ps'ecification 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 i 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 untti the operator stops the leakage by reducing the reactor coolant system pressure below the set point of the steam safety valves and isolates the faulty steam generator. The operator can identify a faulty steam generator by using the off-gas monitors on the condenser air ejector 1 lines; thus he can isolate the faulty steam generator within 34 minutes after

the tube break occurred. During that 34 minute period, a maximum of 2740 ftJ of hot reactor coolant will have leaked into the secondary system; '

i this is equivalent to a cold volume of 1980 ft3 i

j The controlling dose for the steam generator tube rupture accident is the j whole-body dose resulting from imersion in the cloud of released activity.

i 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 ren, the l limit in 10 CFR Part 20 for whole body dose in an unrestricted area, i Althcugh only volatile isotopes will be released from the secondary system,

, the following whole-body dose calculation conservatively assumes that all of

! the radioactLvity which enters the secondary system with the reactor coolant is released to the stmosphere. Both the beta and gamma radiation from these j isotopes contribute to the whole-body dose. The gama 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 11 ult

to the potential gama dose. The semi-infinite cloud model is applicable to 4 the beta dose because of the short range of beta radiation in air. It is l

! further assumed that meteorological conditions during the course of the 1 i accidentcorrespondtoPasquillTypeFand0.{meterpersecondwindspeed. l

! resulting in a X/Q value of 8.51 x 10-4 sec/m3 i

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! Amendment No. 81,97,98 3*I ,

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Revised 3/30/88 RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.3 EMERGENCY CORE COOLING, REACTOR BUILDING EMERGENCY COOLING AND .

REACTOR BUILDING SPRAY SYSTEMS Aeolicability All modes from HEATUP-COOLDOWN to P0HER OPEkATION, inclusive.

Obiective l To define the conditions necessary to assure imediate availability of the ;

emergency core cooling, Reactor Building emergency cooling and Reactor ;

Building spray systems.

Scecification 3.3.1 The reactor shall not remain critical, unless the following I conditions are met:

A. Two indeggn l OPERABLEti) withdent eachECCS subsystems subsystem (injection comprised of: systems) shall be11

1. One,0PERABLE(l) high pressure injection pump. 11 I l

. 2. One OPERABLE decay heat removal pump.

3. One OPERABLE decay heat removal cooler.

4. An OPERABLE (1) flow path capable of taking suction from Il the borated water storage tank on safety injection sic.al. The Reactor Building emergency sump isolation 1

valve shall be either manually or remote-manually operable.

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5. One of the two BNST isolation valves shall be open (SFV !

25003 or SFV 25004). This valve may be closed during the l quarterly valve test specified in the Specifications ,I 4.5.1.2A and 4.5.2.2A.

Action 4

a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least H01 SHUTDOHN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 ,

hours. ]

Note (l) When the requirements of Specification 3.2.2.2 LTOP are in effect the HPI flow path will not be OPERABLE. ;

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Amendment No. 4, 73, 97 3-19 l

- RANCHO SECO UNIT 1 Revised 3/30/88 ,

TECHNICAL SPECIFICATIONS ,,

Limiting Conditions for Operation !

3.5.7 IMERGENCY SHUTDONN INSTRUMENTATION i Annlicability

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A11 modes from HOT 5HUTDONN thru PONER OPERATION, inclusive.

! 11tr.jfIeation )

i The emergency 4hutdown instrumentation channels shown in Table 1 3.5.6-1 shall be OPERABLE with readouts displayed external to the !

Control Room. j

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a. With the number of OPERABLE emergency shutdown instrumentation ,

channels less than required by Table 3.5.7-1, restore the ll lI inoperable channel (s) to OPERABLE status within 7 days, or be in HEATUP-COOLDONN tithin the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l 81181 The OPERASILITY of the emergency shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT l SHUTDOHN of the facilit:r from locations outside of the control room. (This {

capability is required in the event control room habitability is lost and is consistent with Genertl Design Criteria 19 of Appendix A to 10 CFR 50.) i 1

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! Ahendment No. 97 3-341

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Revised 3/30/88 !

i RANCH 0 SECO UNIT 1 l '

TECHNICAL SPECIFICATIONS ..

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i Limiting Conditions for Operation [

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SPENT Fulk E l.

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i Annlicability

- 1 Applies to the Spent Fuel Pool Cooling System and Spent Fuel Pool .

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i water level whenever spent fuel is being stored in the pool.

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ch.iective ,

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' To provide for adequate cooling of the Spent Fuel Pool to ensure that le pool temperature is kept low enough to prevent boiling, 1

and u maintain an adequate water level to ensure sufficient I i shielding. I; snacification 3.9.1 One train of the Decay Heat Removal System (DHR$) must be put in l

service to provide afternate cooling for the Spent Fuel Pool if the bulk coolant temperature reaches 140'F and the Spent Fuel Pool Cooling System is inoperable, and as a supplement to the Spent Fuel Pool Cooling System if a maximum temperature of 180'F is exceeded.

1 3.9.2 . If a train pf the DHR$ is being used to provide alternate cooling '

j for the $ pent Fuel Pool, it shall be declared inoperable for.other i pur oses and the provisions,of Technical Specification 3.3.1 shall lTl app y unless the reactor is in Cold Shutdown. l 3.9.3 Use of the DHR$ for Spent Fuel Pool cocling shall be limited to no ante than 100 cumulative hours (wh'en not in Cold Shutdown) in any I

. 12-month period, l

t 3.9.4. Reactor shutdown must be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if the Spent Fuel l Pool bulk coolant temperature reaches 180'F, and the reactor must '

l i be in Cold shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

j o 3.9.5 At least 37 feet of water shall be maintained in the spent fuel j

pool. The water level in the spent fuel pool may be less than 37 feet if the dose rate from the irradiated fuel at the surface of the water is 2.5 mres/hr or less. ,

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'Hith the requirement of Specification 3.9.5 not satisfied, tuspenf all l movement of fuel assemblies and crane operations with loads in the fuel

- storage areas and restore the water level to within its limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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) Amendment No. 84,97 3-464 ,

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Revised 3/30/s8 RANCW SECO UNIT'1

TECHNICAi. SPECIFICATIONS

, Surveillance Staadards

2. The cell - to - cell and terminal connections are clean and are coated with an anti-corrosion material,

3. The total resist.snce of all cell - to - coli and terminal connections is less than or equal to 201 above an established base-line or benchmark value, and

4. The battery charger will supply at least the established current output necessary to re-charge the batt*ry following an emergency discharge in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months er less.

d. At least once per refueling interval, during COLD SHUTD0kN, by verifying that the battery capacity is adequate to supply cnd maintain in OPERABLE status, all of the actual or simulated emergency loads for the design duty cycle 3r load profile when ,

thebatteryissubjectedtoaservicetest.

e. At least once per 60 months, during COLD SHUIDOWN, by verifying that the battery is at least 80% of the manufacturer'r, rating when subjected to a performance dScharge test. This performance discharge test si,y he terformed in lieu of the battery service test tw:,utrev by Daergency Power Sy*.tes Periodic Testing Specificatloa 4.0.4.d. provided thht the performance discharge test is ps.rformed in the "as-found" condition.

Each vital 125 volt DC and vital 120 volt AC bus listed in f.

Specifications 3.7.10 and 61 shall be deteistned OPERABLE and I energized at least once per 7 days by verifying correct

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breaker alignment and indicated power availability with an overall voltage of greater than or equal to 125 volts DC and 120 volts AC, respectiv61y.

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l Amendment No. 97 g-341 l ,,r r . . . . . .

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Revised 3/30/88 ;

RANCHO SECO UNIT 1

/ TECHNICAL. SPECIFICATIONS Administrative Controls LICENSEE EVENT REPORT l

6.9.4.1 The types of events listed below shall be the subject of written ['

reports to the Director of the Regional Office within thirty (30) days of occurrence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form, pursuant to 10 CFR 50.73 and the guidance of NUREG-1022. ,

a. (1) The enmpletion of any nuclear plant shutdown required by '

the plant's Technical Specification; or !

(11) any operation or condition prohibited by the L nt's i Technical Specifications; or (iii) Any deviation from the plant's Technical ;

Specifications authoriz3d pursuant to 10 CFR 50.54(x). l

b. Any event or condition that resulted in the condition of the !

nuclear power plant, including its principal safety barriers, i

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- being seriously degraded, or that resulted in the nuclear power plant being:

(1) In an unanalyzed condition that significantly compromised plant safety; (ii) In a condition that was outside the design basis of the l plant; or j (iii) In a condition not covered by the plant's operating :

and emergency procedures. j

c. any natural phenomenon or other external condition that posed an act"al threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of ,

duties necessary for the safe operation of the nuclear power plant.

d. Any event or condition that resulted in manual or automatic

' actuation of any Engineered Safety Feature (ESF), including the Reactor Protection System (RPS). However, actuation of an ESF, including the RPS, that resulted from and was part of the preplanned sequence during testing or reactor operation need not be reported.

e. any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to:

1. Shut down the reactor and maintain it in a safe shutdown condition; Amendment No. 17,JJ,63,98 6-12e (1)

i Revised 3/30/88 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls LICENSEE EVENT REPORT

2. Remove residual heat;

3. Control the release of radioactive material; or

4. Mitigate the consequences of an accident.

f. Events covered in paragraph 6.9..i.1.e of this section may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant to this paragraph if redundant equipment in the same system was operable and available to perform the required safety function,

g. Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or -two independent trains or channels to become inoperable in a single system designed to:

1. Shut down the reactor and maintain it in a safe shutdown conditlon;

2. Remove residual heat:

3. Control the release of radioactive material; or

4. Hitigate the consequences of an accident.

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h. 1. Any airborne radioactivity release that exceeded 2 times the applicable concentrations of the limits specified in Appendix B, Table II of 10 CFR 20 in unrestricted areas, when averaged over a time period of one hour.

2. Any liquid effluent release that excs aded 2 times the limiting combined Maximum Permissibir. Concentration (MPC)

(see Note 1 of Appendix B to 10 CFR 20) at the point of

. entry into the receiving watar (i.a., unrestricted area) for all radionuclides except tritium and dissolved noblu gases, when averaged over a time period of one hour,

i. Any event that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases,

j. Failura of the pressurizer EHOVs or Primary System Safety Valves.

Amendnent No. 17,49,53,$3,97,98 6-12e (2)

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ML20151B323

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