Text

Proposed Tech Specs Re Comparison of Plant Tech Specs to STS
	ML20235Q511
Person / Time
Site:	Rancho Seco
Issue date:	10/01/1987
From:	; SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	;
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ML20235Q480	List: ML20235Q478; ML20235Q511;
References
	TAC-63030, TAC-65256, NUDOCS 8710070612
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4 ATTACHMENT I RANCHO SECO TECHNICAL SPECIFICATIONS (Pages Affected by Proposed Amendment No. 164) i 8710070612 871001 PDR ADOCK 05000312 P

PDR

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'l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS j

TABLE OF CONTENTS Section Eagg i

DEFINITIONS 1-1 1.1 RATED P0HFR 1-1 1.2 REACTOR OPERATING.CQNDITIONS 1-1 i

1.2.1 Cold Shutdown 1-1 j

1.2.2 Hot Shutdown 1-1 1.2.3 Reactor Critical

' l-1 l

l 1.2.4 liot Standby 1-1 1.2.5 Power Ooeration 1-1 1.2.6 Bafgelina Shutdown 1-1 1

l 1.2.7 Refuelina 00eration 1-2 1.2.8 Refuelina Interval 1-2 l

1.2.9 Startup 1-2 1.2.10 Remain Critical 1-2 1.2.11 T

1-2 g

1.2.12 Heatuo - Cooldown Mode 1-2 164-~

1.2.13 Action 1-2 1.3 OPERABLE 1-2 1.4 PROTECTION INSTRUMENTATION LOGIC 1-2 1.4.1 Instrument Channel 1-2 1.4.2 Reattpr Protcction System 1-2a 1.4.3 Protection 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 Proposed Amendment No. 164 i

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS.

1 TABLE OF CONTENTS (Continued)

Section EAga-1.16 RESTRICTED AREA 1-7 1.17.

SITE B0UNDARY 1-7 1.18 00SE EOUIVALENT I-131 1-7 1.19 MEMBER (S) 0F THE'PUBLIC 1-7 2

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS' 2-l' 2.1 SAFETY LIMITS. REACTOR CORE 2-1 2.2 SAFETY LIMITS. REACTOR SYSTEM PRESSURE 2-4 2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTIVE INSTRUMENTATION. 2-5 164-3 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 Operational Comoonents 3-1 3.1.2 Pressurization. Heatuo. and Cooldown Limitations 3-3 3.1.3 Minimum Conditions for Criticality 3-6 3.1.4 Reactor Coolant System Activity 3 3.1.5 Chemistry 3-10 l

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l Proposed Amendment No. 164 ita

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)

Section PAqa 3.1.6 Lukitqg 3-12 3.1.7 Moderator Temperature Coefficient of Reactivity 3-15 3.1.8 Low Power Physics Testina Restrictions 3-15b 3.1.9 Control Rod Operation 3-16 3.2 HIGH PRESSDRE INJECTION. CHEMICAL ADDITION. AND LOH TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEtiS 3-17 1

3.3 EMERGENCY CORE C00 LING.,_REAC_ TOR _DDILDING EMERGENCY C00LLE AND REACTOR BUILDING SPRAY SYSTEMS 3-19 3.4 STEAM AND POWER CONVERSION SYSTEM 3-23 1

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3.5 INSTRUMENTATION SYSTEMS 3-25 j

j 3.5.1 Operational Safety Instrumentation 3-25 j

i 3.5.2 Control Rod Grouo and Power Distribution Limits 3-31 3.5.3 Safety Features Actuation System Setooints 3-34 4

3.5.4 Incore Instrumentation 3-36 l

l 3.5.5 Accident Monitorina Instrumentation 3-38a

)

164-+-

3.5.6 Emeraency Shutdown Instrumentation 3-38d i

3.6 REACTOR BUILDING 3-39 i

3.7 6SXILIARY ELECTRICAL SYSTEMS 3-41 3.8 FUEL 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 H0IST 3-49 3.12 SHOCK SUPPRESSORS (SNUBBERS) 3-51 3.13 AIR FILTER SYSTEMS 3-52 3.14 D RE SUPPRESSION 3-53 3.14.1 IDitIumentation 3-51 i

3.14.2 Water System 3-53 3.14.3 Soray and Sorinkler Systems 3-56 3.14.4 CO2_ System 3-56 Amendment No. 164 iii L

' RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS q

TABLE.0F CONTENTS- (Continued)-

Section Eage 164-4 SURVEILLANCE STANDARDS

.4,

4.0 GENERAL SURVEILLANCE REQUIREMENTS 4-0 4.1 OPERATIONAL SAFETY REVikd 4-1 I

4.2 SURVEILLANCE OF'ASME CODE CLASS 1. 2. AND 3 SYSTEMS 4-10 4.2.1 Reactor Vessel Surveillance Specimens 4-10 4.2.2 Inservice Insoection 4-11 164-~

4.2.3 Leakaae Surveillance 4-13 4.3 TESTING FOLLOWING OPENING OF SYSTEM 4-14 4.4 REACTOR BUILDlEG 4-15 4.4.1 Containment Leakaae Tests 4-15 4.4.2 Structural Intearity 4-21 4.4.3 Hydroaen Purae System 4-25 4.5 EMERGENCY CORE COOLING AND REACTOR BUILDING 4-26 COOLING SYSTEM PERIODIC TESTING 4.5.1 Emeroency Core Coolina System 4-26 4.5.2 Reactor Buildino Coolina Systems 4-29 4.5.3 Decav Heat Removal System and Reactor Buildina Sorav.

System Leakaae 4-32 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTING 4-34 4.7 REAC]QR CONTROL ROD SYSTEM TESTS 4-36 4.7 1 Coatrol Rod Drive System Functional Tests 4-36 4.7.2 Caatrol Rod Proaram Verification (Group vs. Core Positions) 4-37 4.8 AUXILIARY FEEDHATER PUMP PERIODIC TESTING 4-39 4.9 REACTIVITY ANOMALIES 4-40 4.10 EMERGENCY CONTROL ROOM FILTERING SYSTEM 4-41 Proposed Amendment No. 164 v

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)

Section P_ age 4.11 REACTOR BUILDING PURGE EXHAUST FILTERING SYSTEM 4-42 4.12 AUXILIARY AND SPENT FUEL BUILDING FILTER SYSTEMS 4-43 4.13

&VGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH 4-44 ENERGY LINES OUTSIDE OF CONTAINMENT 4.14 SHOCK SUPPRESSORS (SNUBBERS) 4-47 l

4.15 RADI0 ACTIVE MATERIALS SOURCES 4-48 4.16

' Reserved 4-49 4.17 STEAM GENERATORS 4-51 4.17.1 Steam Generator Samole Selection and Instantion 4-51 4.17.2 Steam Generator Tube Samole Selection and Insoection 4-51 4.17.3 Inspection Frequencies 4-52 4.17.4 Angsptance Criteria 4-53 4.17.5 Reports 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 RADI0 ACTIVE LIOUID EFFLUENT INSTURMENTATION 4-63 4.20 RADIOACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION 4-65 4.21 LIOUID EFFLUENTS 4-69 4.21.1 Concentration 4-69 4.21.2 Dose Calculatio.nl 4-72 4.21.3 Liauid Holdup Tanks 4-73 4.22 GASEOUS EFFLUENTS 4-74 4.22.1 Dose Rate 4-74 Proposed Amendment No. 164 164--

vi

l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)

Section P_ asst 4.22.2 Noble Gases 4-77 I

4.22.3 lodine-133. Tritium and Radionuclides in Particulate Form

_4-78 4.23 GASEOUS RADHASTE TREATMENT 4-79 4.24 GAS STORAGE TANKS 4-80 4.25 SOLID RADI0 ACTIVE WASTES 4-81 4.26 ILA.DIOLOGICAL ENVIRONMENTAL MONITORING 4-83 4.27 LAND USE CENSUS 4-86 4.25 EXPLOSIVE GAS MIXTURE 4-87 i

4.29 BJEL CYCLE DOSE 4-89 8.30 INTERLABORATORY COMPARISION PROGRAM SURVEILLANCE RE0UIREMERI 4-90 4.31 NUCLEAR SERVICE ELECTRICAL B1!I1 DING EMERGENCY HEATING VENTILATION AND AIR CONDITIONING 4-91 5

DESIGN FEATURES 5-1 5.1 SJJ1 5-1 5.2 CONTAINMENT 5-2 5.2.1 Reactor Buildina 5-2 5.2.2 Reactor Buildina Isolation System 5-3 5.3 REACTOR 5-4 i

5.3.1 fleactor Core 5-4 5.3.2 Reactor Coolant System 5-4 5.4 NEH AND SPENT FUEL STORAGE FACILITIES 5-6 l

5.4.1 New Fuel Insoection and Temocrary Storace Rack 5-6 1

l 5.4.2 N_gw and Soent Fuel Storace Racks and Failed Fuel l

Storaae Container Rack 5-6 l

5.4.3 New and Soent Fuel Temocrary Storage 5-6 5.4.4 Soent Fuel Pool and Storaae Rack Desian 5-6 1

Proposed Amendment No. 164 164--

vii

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS i

LIST OF TABLES lablf!

PELE 2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS 2-9 164-~

3.0-1 APPLICABILITY OF SPECIFICATIONS'3.0.3 AND 3.0.4 3-0a 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS 3-27 3.5.5-1 ACCIDENT MONITORING INSTRUMENTATION OPERABILITY 3-38b REQUIREMENTS 164-*

3.5.6-1 EMERGENCY SHUTDOWN INSTRUMENTATION 3-38e 3.6-1 SAFETY FEATURES CONTAINMENT ISOLATION VALVES 3-40 3.7-1 VOLTAGE PROTECTION SYSTEM RELAY TRIP VALUES 3-41a i

3.7-2 VOLTAGE PROTECTION SYSTEM LIMITING CONDITIONS 3-41b

]

3.12-1 SAFETY RELATED HYDRAULIC SNUBBERS 3-51a-e l

3.14-1 FIRE DETECTION INSTRUMENTS FOR SAFETY SYSTEMS 3-55

]

3.14-2 INSIDE BUILDING FIRE HOSE STATIONS 3-Da l

3.15-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3-61

)

3.16-1 RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION 3-04 3.22-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 3-83 l

3.22-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS 3-86 IN ENVIRONMENTAL SAMPLES 164-~

4.0-1 APPLICABILITY OF SPECIFICATIONS 4.0.2 AND 4.0.3 4-0a 4.1-1 INSTRUMENT SURVEILLANCE REQUIREMENTS 4-3 1

4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY 4-8 4.1-3 HINIMUM SAMPLING FREQUENCY 4-9 4.2-1 CAPSULE ASSEMBLY HITHDRAHAL SCHEDULE AT DAVIS-BESSE 1 4-12b 4.10-1 ENVIRONMENTAL RADIATION MONITORING PROGRAM 4-42 l

4.10-2 OPERATIONAL ENVIRONMENTAL RADIATION MONITORING PROGRAM 4-22a 4.14-1 DESIGNATED SAFETY RELATED HYDRAULIC SNUBBERS FUNCTIONALLY 4-47d,e TESTED ONLY AS REQUIRED BY THE SNUBBER SEAL REPLACEMENT PROGRAM 4.17-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED 4-56 DURING INSERVICE INSPECTION 4.17-2A STEAM GENERATOR TUBE INSPECTION 4-57 4.17-2B STEAM GENERATOR TUBE INSPECTION (SPECIAL LIMITED AREA) 4-57a 4.17-3 OTSG AUXILIARY FEEDHATER HEATER SURVEILLANCE 4-57b,c 4.19-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 4-64 SURVEILLANCE REQUIREMENTS 4.20-1 RADI0 ACTIVE GASEQUS EFFLUENT MONITORING INSTRUMENTATION 4-66 l-SURVEILLANCE REQUIREMENTS Proposed Amendment No. 164 ix L_.

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 1.

DEFINITIONS

'The following terms are defined for uniform interpretation of these specifications.

1.1 RATED P0HER Rated power is a steady reactor core output of 2772 MHt.

1 164-1.2~

REACTOR OPERATING CONDITIONS 10PERATIONAL MODE OR MODE)

An operational mode or mode shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.2-1, 1.2.1 Cold Shutdown 164-The reactor is in the cold shutdown condition when keff is 10.99, and i

Tavg is no more than 200*F.

Pressure is defined by Specification 3.1.2.

See Table 1.2-1 1.2.2 Hot Shutdown 164-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.

4 1.2.3 Reactor Critical The reactor is critical when the neutron chain reaction is self-sustaining 164-~

and keff 1 0.99.

1.2.4 Hot Standby The reactor is in the hot standby condition when all of the following 164-~

conditions exist:

(See Table 1.2-1)

A.

Tavg is greater than 525*F.

164--

B.

keff is >0.99.

i C.

Indicated neutron power on the power range channels is less than 2 percent of rated power.

1.2.5 Power Ooeration i

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 Refuelina Shutdown The reactor is in the refueling shutdown condition when, even with 'all-rods 164--

removed, the reactor core reactivity, keff, is 1 0.99 and the cociant temperature at the decay heat removal pump suction is no more.than 140*F.

Pressure is defined by Specification 3.1.2.

A refueling shutdown refers to a shutdown to replace or rearrange all or a portion of the fuel assemblies -

164-~

and/or control rods.

See Table 1.2-1.

Proposed Amendment No. 164 1-1

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS i

l Definitions l

1.2.7 Refuelina Operation An operation involving a change in core geometry by manipulation of fuel or i

I control rods when the reactor vessel head is removed.

1.2.8 Refueling _Interva1

Time betweer, normal refuelings of the reactor, not to exceed 24 months for l

the first refueling and 18 months thereafter without prior approval of the i

NRC.

j 1.2.9 Startuo j

1 The reactor shall be considered in the startup mode when the shutdown margin l

is reduced with the intent of going critical.

l 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 a

1 be completed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1.2.11 T yg g

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

l 164~-

1.2.12 ik3_tuo - Cooldoy_a The reactor is in heatup-cooldown when the range of reactor coolant temperature is greater than 200*F and less than 525'F.

164-1.2.13 Action Action (time) requirements shall be that part of a specification which prescribes remedial measures required under designated conditions.

1.2.14 Leakage A.

IDENTIFIED LEAKAGE shall be:

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a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks 15at 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 Proposed Amendment No.164 1-2

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RANCHO SECO UNIT 1 i

TECHNICAL SPECIFICATIONS Definitions 164-c.

Reactor coolant system leakage through a steam generator to the secondary system.

j B.

UNIDENTIFIEDLEAKAGEshallbeallleakagewhichisnotIDENTIFIEDj 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 seal water flow 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 is inoperable or its emergency power source is inoperable, it may be considered OPERABLE for the 3

purpose of satisfying the requirements of its applicable Limiting Condition i l for Operation provided its redundant system or component is OPERABLE with an.

OPERABLE normal and emergency power source.

1.4 PROTECTION IN_STRUMENTATION LOGIC 1.4.1 Instrument Channel An instrument channel is the combination of sensor, wires, amplifiers and ouput devices which are connected for the purpose of measuring the value of I

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 The reactor protection system is shoe in Figures 7.1-1 and 7.2-2 of the FSAR.

It-is that combination of protective channels and associated 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 drive s

control protective trip breakers and activating relays or coils, Proposed Amendment No. 164 1-2a iNj

164-TABLE 1.2-1 OPERATIONAL MODES Operational Reactivitiy Coolant Indicated

' Remarks Mode Condition Temperature Neutron "

keff Tavg*F Power (% of'. '-

Rated Power-

>0.99

>525 12 Operation i

Hot

>0.99

>525

<2 Standby Startup

>0.99

>525

<2

<1.00 Hot 50.99 152.[

0 Shutdown 1

Heatup -

10.99

>200 0

At least one Cooldown

<525 RCP/0TSG with Tavg>280*F Cold 10.99 1200 0

RCS Pressure as Shutdown

.i defined.

f f

Refueling

10.99 1140 at 0

RCS Pressure I

Shutdown assuming all DHR pump as defined rods out suction in 3.1.2 Refueling Reactor Vessel head removed and Refueling Shutdown conditions operation

Fuel in the reactbr vessel with the vessel head closure bolts less than i

fully tensioned or with the head removed.

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Pro;Vsed Amendment No.164 1-2dn

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% :ri RANCHO SECO UNTI 1 I

TECHNICAL SPECIFICATIONS-Limiting Conditions for[ Operation I

J 3.

LIMITING CONDITHlNS FOR OPERATION p

164-3.0 General Limiting' Conditions For. Operation

)

3.0.1 Compliance with the Limiting Conditions for Operation contained i

in the succeeding Specifications is required during the I

OPERATIONAL MODES cr other coditions specified therein; except I

that upon failure to meet the Limiting Conditions 'for Operation,.

the associated Action (time) requirements shalb be met.

_4 1

3.0.2 Noncc.apliance with a Specification shall exist when the

)

4 requirements of the. Limiting Condition for Operati6n and j

associated Action (time) requirements are not met within'the 1

specified time intervals.

If the Limiting Condition'for Opfation is restored prior to expiration of the specified time intervals, completion of the Action (time) requirements is not

)

requireo.

1

,3.0.3 When a Limiting Condition for Operation is not met, except as J

provided in the associated Action (time) requirements, within 1

]

hour action shall be initiated to place the unit in a MODE in s

which the Sp6cification does not apply to placing it, as l

applicable, in:

1 i

1.

At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 2.

At least HOT SHUTD0HN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and 3.

At least COLD SHUTD0HN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the Action (time) requirements, the Action (time) 1 requirement may be taken in accordance with.the specified time limits as measured from the time of failure-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 on OPERATIONAL MODE or other. specified condition shall not be made when the conditions for the Limiting i

l-Conditions for Operation are not met and the associated Action (time) requires a shutdovn if they are not met within a specified time interval. fEntry into an OPERATIONAL H0DE or specified condition may be made in accordance with Action (time) requirements when conformance to them permits continued operation of the facility for:an ur. limited period of. time.

This provision shall not prevent passage through OPERATIONAL MODES as required to comply with Action requirements.

Exceptions to these requirements are stated in the individual specifications and Table 3.0-1.

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l Proposed Amendment No. 164 3-0 l

1

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1644, d

TABLE 3.0-1 Applicability of Specifications 3.0.3 and 3.0.4.

(The "NA". indicates that the l

provisions of Specification (s) 3.0.3 and/or 3.0.4'are not applicable to the sections l

l Lidentified).

Section Specification 3.0.3 Specification 3.0.4 3.1 3.2.1 3.2.2 N/A 3.3 l

3.4

)

3.5.1 N/A 3.5.2 3.5.3 j

3.5.4 N/A N/A 3.5.5 N/A 3.5.6 j

3.6 N/A 3.7 3.8 N/A 3.9 l

3.10 3.11 N/A N/A 3.12 3.13 3.14 N/A N/A 3.15 N/A N/A

~1 3d6 N/A N/A i

3.17 N/A N/A 3.18 N/A N/A 3.19 N/A N/A 3.20 N/A N/A 3.21 N/A N/A 3.22 N/A N/A-3.23 N/A N/A g

3.24 N/A N/A 3.25 N/A N/A 3.26 N/A N/A

The provisions of Specification 3.0.3 are not applicable when l

the reactor is in Refueling Shutdown or Refueling Operation.

164-Proposed Amendment No. 164 l

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RANCHO SECO UNIT.1 1

TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 164-+ =

j 3.1 REACTOR COOLANT SYSTEM Applicability All modes from HEATUP-C00LD0HN to POWER OPERATION,. inclusive.

'Dbiettive

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-To specify those liiniting conditions for. operation of the reactor coolant system which must be met to ensure safe reactor operations ~.

3.1.1 OPERATIONAL COMPONENTS Specification

-]

3.1.1.1 Reactor Coolant Pumps A.

Pump combinations permissible for given power levels shall-be as shown in Table 2.3-1.'

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I B.

The boron concentration in the reactor coolant system shall j

not be reduced unless'at.least one reactor coolant pump or one J

decay heat removal pump is circulating reactor coolant.

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C.

Operation at power with two pumps shall be limited to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

']

in any 30 day _ period.

164~

D.

At least 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.

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

I i

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 i

safety valve shall be operable if all reactor coolant system I

openings are closed, except for hydrostatic tests in accordance with ASME Boiler ~ and Pressure Vessel Code, Section

'i III.

Proposed Amendment No. 164 3-1

I RANCHO SECO UNIT 1

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TECHNICAL SPECIFICATIONS

.)

Limiting Conditions for Operation

'J 164-ACJ; ion 1

In COLD SHUTDOWN with no pressurizer code safety valve OPERABLE, immediately f

suspend all operations involving positive reactivity changes and place an 1

OPERABLE decay heat removal loop into operation in the shutdown cooling mode.

1 When above the hot shutdown condition, with one pressurizer code safety j

valve inoperable, either restore the inoperable valve to OPERABLE status j

within 15 minutes 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 required for cold overpressure protection.

4 164-B.

If the EH0V and its associated block valve are not OPERABLE whenever the reactor is in HOT STANDBY.or critical, the-following actions shall be taken:

a.

With the EH0V inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .either restore the EH0V to OPERABLE status or close the associated block l

valve and remove power from the block valve; otherwise, i

be in a least HOT STANDBY within the'next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s-and in-l COLD SHUTDOHN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .. The j

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 164-operable with one loop in operation when the coolant average I

temperature is below 280'F, except during fuel loading and refueling. One of the two loops required need not be in operation for a maximum of one hour provided (1) no operations 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.

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, Proposed Amendment No. 164 3-la

RANCHO SECO UNIT 1 j

TECHNICAL SPECIFICATIONS

]

Limiting Conditions for Operation I

3.

Decay Heat Removal Loop (A) 4.

Decay Heat Removal Loop (B) 164~

B.

With less than the above required coolant loops 0PERABLE, immediately initiate corrective action to return the required

)

coolant loops to OPERABLE status as soon as possible; be in l

COLD SHUTDOWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

3.1.1.6 Reactor Coolant System High Point Vents A.

The vent path on Loop A and vent path on Loop B shall be operable and closed during power operation.

B.

The vent path on the pressurizer shall be operable and closed i

during power operation.

i C.

With one of the above reactor coolant system vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vr.nt path is maintained closed with power removed from the valve actuator of all the valves in the J

inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days.

If the status is not restored

'j to operable in 30 days, be in HOT STANDBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and i

in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

D.

With two or more of the above reactor coolant system vent paths inoperable; maintain the inoperable vent paths closed I

with power removed from the valve actuators of all the valves i

in the inoperable vent paths, and restore at least (two) of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

If the status is not restored to operable in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , be in HOT STANDBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

fktlel A reactor coolant pump or decay heat removal pump is required to be in operation 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.

(1) l Proposed Amendment No. 164 3-2

RANCHO SECO UNIT-1 TECHNICAL SPECIFICATIONS

~

Limiting Conditions for Operation 3.1.2

, PRESSURIZATION, HEATUP, AND C00LDOWN LIMITATIONS Specification 1

3.1.2.1 Jnservice Leak and Hydrostatic Tests:

Pressure temperature limits for the' first eight EFP years 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 eig' ht 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 164*

Figure 3.1.2-1 and Figure 3.1.2-2 respectively.

The Reactor Coolant System temperature and pressure shall be determined to~be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydostatic testing operations. Heatup and cooldown rates shall not exceed the rates stated on the associated figure.

Action If heatup and cooldown rates are exceeded, stabilize the temperature and restore the temperature and/or pressure to within the limits as soon as practical, 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 SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce RCS Tavg and pressure to less than 200*F and 500 psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ..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 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 i

the pressurizer and spray fluid is greater than 410*F.

]

lu i

i l

)

l Proposed Amendment No. 164

)

l-3-3 l

_J

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATI0NS' 1

Limiting Conditions.for Operation 3.1.3 MINIMUM CONDITIONS FOR CRITICALITY Speci fications -

q 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

'i Specification 3.1.8 shall apply.

j 3.1.3.2 Reactor coolant temperature.shall be above DTT + 10*F.

j 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 i

power physics testing when the requirements of Specification 3.1.8 shall apply, the reactor shall' be subcritical by an amount equal to l

or greater than the calculated reactivity insertion due to j

i depressurization.

164*

Action j

Hith the reactor subtritical by less than the required amount, immediately initiate and continue boration until the required SHUTDOHN MARGIN is

restored, i

3.1.3.4 The reactor shall be maintained subtritical 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.

164-Hith the reactor subcritical by less than the ' required amount, immediately initiate and continue boration until the required SHUTDOHN MARGIN is j

restored.

1 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 margi'n 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.

Bases 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 temperatures 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.

Proposed Amendment No. 164 3-6

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation-1 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.

I 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 magnitude of a power excursion resulting j

from a reduction of moderator density.

l Proposed Amendment No.164 3-7

RANCHO SECO UNIT 1 TECHNICAL SPECIFICAT10NS Limiting Conditions for Operation 3.1.4 REACTOR COOLANT SYSTEM ACTIVITY Specification 164-~

3.1.4.1 The specific activity of the reactor coolant due to nuclides with 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 1644 reactor coolant samples.

Action g

Hith the specific activity of the reactor coolant greater than 43/E microcuries/ gram, be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , j

Bases The above specification is based on limiting the consequences of a postulated accident involving the double-ended rupture of a s. team generator tube.

The rupture of a steam generator tube enables reactor coolant and its associated

-l 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 isolates 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 the tube break occurred. During that 34 minute period, a maximum of 2740 ft3ofhotreactorcoolantwillhaveleakedingothesecondarysystem; 164-~

this is equivalent to a cold volume of 1980 ft.

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 i

the reactor coolant will be limited to a value'which will insure that the i

164-~

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 on 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 accidentcorrespondtoPasquillTypeFand0.Qmeterpersecondwindspeed, resulting in a X/Q value of 8.51 x 10-4 sec/mo.

Proposed Amendment 164 3-8 i

i

RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Limiting Conditions for Operation j

3.1.6 LEAKAGE 164-golicability Applies whenever the reactor is critical or in HOT SHUTD0HN, Obiective To monitor and limit RCS leakage.

Speci fication l

164-3.1.6.1 Reactor Coolant System leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, c.

1 GPM total primary-to-secondary leakage through the steam generators and 500 gallons per day through the tubes of any one generator, j

I d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and e.

16 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure I

of 2155 z 10 psig.

f.

1 GPM leakage from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.3-1, which shall be included in IDENTIFIED LEAKAGE. Specific requirement for Reactor Coolant System Pressure Isolation Valves are provided l

in Specification 3.3.4.

Surveillance requirements are provided in Specification 4.5.1.1.B.4 and 5.

Action A.

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTD0HN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOHN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1 B.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE B0UNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOHN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTD0HN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

If reactor shutdown is required, the rate of shutdown and the condition of shutdown shall be determined by a safety evaluation l

for each case and justified in writing as soon thereafter as practicable.

C.

During power operation, two reactor coolant leak detection systems of different operating principles shall be in operation, with one of the two systems sensitive to radioactivity. The systems sensitive to radioactivity may be out-of-service for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months provided two other means are available to detect leakage.

1 Proposed Amendment No. 164 3-12

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation EAlfi Every reasonable effort will be made to reduce reactor coolant leakage including evaporative losses (which may be on the order of 0.5 gpm) to the lowest possible rate and at least below 1 gpm in order to prevent a large leak from masking the presence of a smaller leak.

Evaporative losses i

identified during startup testing of 0.5 gpm are not considered part of the 1 gpm unidentified leakage. Water inventory balances, radiation monitoring equipment, boric acid crystalline deposits, and physical inspections can

]

disclose reactor coolant leaks. Any leak of radioactive fluid, whether from i

the reactor coolant system primary boundary or not can be a serious problem

(

with respect to in-plant radioactivity contamination and cleanup or it could develop into a still more serious problem; therefore, first indications of 4

such leakage will be followed up as soon as practicable.

164--

Although the specified leakage rates are acceptable from a dose point of view, especially if they are to closed systems, it must be recognized that j

leaks in the order of drops per minute through any of the walls of the J

primary system could be indicative of materials failure such as by stress j

corrosion cracking.

If depressurization, isolation and/or other safety 164-measures are not taken promptly, these small breaks could develop into much larger leaks, possibly into a gross pipe rupture. Therefore, a shutdown requirement is imposed.

i The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The total steam generator tube leakage limit of 1 GPM for all steam generators ensures that the dosage contribution from tube leakage will be limited to a small fraction of Part 100 limits in the event of either a q

steam generator tube rupture or steam line break.

The 1 GPM limit is consistent with the assumptions used in the analysis of these accidents.

The 1 gpm leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable because it may be indicative of an impending gross failure of the pressure boundary.

Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

1 Proposed Amendment No. 164 3-13

_._ ]

RANCHO SECO UNIT 1-TECHNICAL SPECIFICATIONS' Limiting Conditions for Operation BA121 (Continued)

When the source of leakage has been identified, the situation can be evaluated to determine if operation can safely continue.

This evaluation will be performed by the Operating-Staff and will be documented in writing and approved by the Unit Operations Superintendent. Under these conditions,.

an allowable reactor coolant system leakage rate of 10 gpm has been-established.

This explained leakage rate of 10 gpm is also.well within the capacity of one high pressure injection pump'and makeup would be available even under the loss of off-site power condition.

1 If. leakage.is to the Reactor Building it may be identified by one or more of the following methods:

A.

Sutno Levels - All Reactor Building leakage.is collected in the Reactor Building sumps.

These sumps drain by gravity into a 120 gallon Reactor Building drain accumulation' tank.

The drain accumulation tank is used to measure the drain flow with level indicators at 20 gallons and 120 gallons.

The tank is dumped into the East decay heat removal pump room sump. :The frequency of y

dumping the accumulation tank and time interval between levels are recorded in the-Control Room and are direct measures of the flow rate.

Depending on the level at which the tank is dumped..the time to confirm a 1 'gpm leak is between 40 minutes and 120 minutes.

l l

Proposed Amendment No. 164 164~~.

3-13a i

RANCHO SECO UNIT:1 1

l TECHNICAL SPECIFICATIONS Limiting Conditions. for Operation -

1 3.1.7' MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY-i i

164- -

Applicability POWER OPERATION-j d

Specification

.The moderator temperature coefficient shall not be positive at power levels above 95 percent of rated-power.

-164~

Action Hith the moderator temperature coefficient'at a positive value, be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Bases A non-positive moderator coefficient at power levels above 95 percent of rated power is specified such that the maximum clad temperatures will not exceed the Final Acceptance Criteria based on LOCA analyses.

Below 95-percent of rated power the Final Acceptance Criteria will not bq exceeded with a positive moderator temperature coefficient of +0.9 x 10-9 Ak/k/F corrected to 95 percent of rated power.

All other accident analyses as reported in the FSAR have been performed for a range of moderator temperature coefficients including +0.9 x 10-4 Ak/k/F.

The experimental value of the moderator coefficient will be corrected to obtain the hot full power moderator coefficient. When the hot zero-power value is corrected to obtain the 95 percent power vclue,-the following corrections will be applied:

1.

Uncertainty in isothermal measurement - The measured moderator temperature coefficient will contain uncertainty owing to the AT of the base and perturbed conditions and the uncertainty in the reactivity measurement.

Proper corrections will:be added for these conditions to provide a conservative moderator coefficient.

2.

Doppler contribution at hot zero power - During measurement of the isothermal moderator coefficient at hot zero power, the j

fuel temperature will increase by the same amount as the moderator.

The measured temperature coefficient must therefore be increased to obtain a pure moderator temperature coefficient.

3.

Moderator temperature change - The hot zero power measurement must be corrected for the difference in water temperature at zero power (532*F) and.15 percent power (582*F). Above this power, the average moderator temperature remains 582*F.

4.

Fuel temperature interaction (power effect) - The moderator coefficient must be adjusted to account for the interaction of an average moderator temperature with increasing fuel temperatures as power increases. ' Adjust the moderator coefficient at 15 percent power to the coefficient at any-power. level above 15 percent.

Proposed Amendment 164 3-15

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation l

3.1.9 CONTROL R00 OPERATION 164--

Aeolicability All modes above COLD SHUTDOWN l

Specification 3.1.9.1 The concentration of dissolved gases in the reactor coolant shall be limited to 100 std. cc/ kilogram of water at the reactor vessel outlet temperature.

3.1.9.2 Allowable combinations of pressure and temperature for control rod operation shall be to the left of and above the limiting pressure versus temperature curve for a dissolved gas concentration of 100 std. cc/ kilogram of water as shown in Figure 3.1.9-1.

R 164-Action l

In the event the limits of Specifications 3.1.9.1 or 3.1.9.2 are exceeded, the center control rod drive mechanism shall be checked for accumulation of undissolved gases. This shall be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following i

30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Bases By maintaining the reactor coolant temperature and pressure as specified I

above, any dissolved gases in the reactor coolant system are maintained in solution.

Although the dissolved gas concentration is expected to be approximately 20-40 std. cc/ kilogram of water, the dissolved gas concentration is conservatively assumed to be 100 std. cc/ kilogram of water at the reactor vessel outlet temperature.

l The limiting pressure versus temperature curve for dissolved gases is determined by the equilibrium pressure versus temperature curve for the dissolved gas concentration of 100 std. cc/ kilogram of water.

The equilibrium total pressure is the sum of the partial pressure of the i

dissolved gases plus the partial pressure of water at a given temperature, j

The margin of error consists of the maximum pressure difference between the pressure sensing tap and lowest pressure point in the system, the maximum j

pressure gage error, and the pressure difference due to the maximum temperature gage error.

l If either the maximum dissolved gas concentration (100 std. cc/ kilogram of l

water) is exceeded or the operating pressure falls below the limiting l

)

pressure versus temperature curve, the center CRDM should be checked for j

accumulation of undissolved gases.

l Proposed Amendment No. 164 3-16 1

l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation

]

3.3 EMERGENCY CORE COOLING,. REACTOR BUILDING EMERGENCY COOLING AND REACTOR BUILDING SPRAY SYSTEMS 1

Applicability 164-+ -

All modes from HEATUP-C00LDOWN to POWER OPERATION, inclusive.

Objective To define the conditions necessary to assure immediate availability of the emergency core cooling, Reactor Building emergency cooling and Reactor-j

~

Building spray systems.

i Specification j

3.3.1 The reactor shall not remain critical, unless the following q

conditions are met:

1 A.

Injection System 1.

The borted water storage tank shall contain a minimum of 390,000 gallons of water having a minimum concentration of 1,800 ppm boron at a temperature not less than 40*F.

The 3

manual valves on the discharge line from the borated water J

storage tank shall be locked open.

l 2.

Two out of three high pressure injection pumps shall be operable.

l 3.

Two safety features actuated decay heat removal pumps shall operable.

4.

Both decay heat removal coolers shall be operable.

5.

Two BHST level instrument channels shall be operable.

6.

The Reactor Building emergency sump isolation valve shall be either manually or remote-manually operable.

7.

One of the two BHST isolation valves shall be open (SFV 25003 or SFV 25004).

This valve may be closed during tho quarterly valve test specified in the Specifications 4.5.1.2A and 4.5.2.2A.

B.

Core Flooding System 1

The two core floodidng tanks shall each contain 1040 1 30 ft3 of borated water at 600 25 psig.

2.

Core flooding tank boron concentration shall not be less than 1,800 ppm boron.

3.

The electrically operated discharge valves for the core t

flood tanks shall be open. The breakers shall be open and so tagged.

Pro wsed Amendment No.164 3-19

)

RANCHO SECO UNIT 1 TECHNICAL' SPECIFICATIONS h

Limiting Conditions for Operation j

i 4.

Two core flood tank pressure instrument channels'shall be 1

operable (one per tank minimum).

.l 5.

The electrically operated vent valves (HV-26511 and HV-26512) from the core flood tanks shall be closed. The-i breakers shall be open and so tagged except during normal-1' venting operations.

ll C.

Reactor Building spray system and Reactor Building emergency cooling system.

The following combination of system components must be operable:

1.

Two Reactor Building spray pumps and their associated spray headers with a minimum of 32 percent Na0H solution in the spray additive tanks and, 2.

A minimum level of 78 inches of. solution shall be available in each spray additive tank.

3.

Four emergency cooling units, two with charcoal-filter.

164~

units.

There are two cooling units in each of two emergency cooling trains (Train A and B).

Action Hith one train of the above required containment cooling units inoperable and both containment spray systems OPERABLE, restore the inoperable train of cooling units to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTD0HN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

With two trains of the above required containment cooling units inoperable, i

and both containment spray-systems OPERABLE, restore at least one train of cooling units to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOHN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOHN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Restore both above required trains of_ cooling units to OPERABLE i

status within 7 days of initial loss or be in at least HOT SHUTD0HN within j

the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOHN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

]

)

Hith one train of the above required containment cooling units inoperable I

and one containment spray system inoperable, restore the inoperable spray i

system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOHN-within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOHN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

]

Restore the inoperable train of containment cooling units to OPERABLE status l

within 7 days of initial loss or be in at least HOT SHUTDOHN within the next j

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOHN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

i Proposed Amendment 164 3-20 J

RANCHO SECO UNIT'l TECHNICAL' SPECIFICATIONS' Limiting Conditions for Operation 164-D.

Nuclear Service Cooling Hater (NSCH) System 1.

Two NSCH Loops are operable.

611100 Hith only one NSCH loop OPERABLE, restore at least two loops to j

OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next'6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1 E.

Nuclear Service Raw Hater (NSRH) System

.1.

Two NSRH loops shall be OPERABLE and each nuclear' service spray pond shall be OPERABLE with; a.

A minimum water level of 5'4", and b.

A maximum water temperature of 95'F.

Action With the requirements of the above specification not satisfied, be a

in at least HOT SHUTD0HN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within i

the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

F.

Safety features valves e.nd interlocks associated with each of

+-

the above systems are operable.

Inoperable valves shall be placed in the safety features position.

I 3.3.2 Maintenance shall be allowed during power operation on any component (s) in the high pressure, low pressure, nuclear service cooling and raw water cooling, Reactor Building spray, or Reactor Building emergency. cooling systems, the core flooding system pressure instrument channels or BHST level channels, which will not degrade safety features system A or B below the level of performance with the single subsystem removed from service.

In the context of this specification, a safety features system consists of the following subsystems:

high pressure injection, low pressure

)

injection, Reactor Building emergency air cooling, Reactor Building i

spray, diesel generator, nuclear service cooling water and nuclear service raw water.

If the system being repaired is not restored to meet t'ne requirements of specification 3.3.1 within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

If the requirements of specification 3.3.1 are not met within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the reactor shall be placed in a cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Proposed Amendment No. 164 3-21

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.3.3 Prior to initiating maintenance on any of the component (s), the I

duplicate (redundant) components shall be verified operable by.

j checking that the surveillance test for the components (s) has been successfully completed and will remain in effect for the duration of the maintenance period.

Inservice testing'per specification 4.2.2.1 shall not be performed on any component (s) whose duplicate (redundant) component (s) has (have) been declared inoperable or is out of service for any reason.

i 3.3.4 During power operation, hot standby, hot shutdown or startup conditions, the primary coolant system pressure isolation valves shall be functional as follows:

1.

-All pressure isolation valves listed in Table 3.3-1 shall be functional as a pressure isolation device, except as specified in 3.3.4.2.

Valve leakage shall not exceed the amounts j

indicated.

1 J

2.

In the event that integrity of any pressure isolation valve specified in Table 3.3-1 cannot be demonstrated, reactor operation may continue, provided that at least two valves in each high pressure line having a non-functional valve are in i

I and remain in, the mode corresponding to the isolated i

condition.(a)

)

3.

If Specifications 3.3.4.1 and 3.3.4.Z cannot be met, a shutdown shall be initiated, the reactor shall not remain critical and shall be brought to a cold shutdown condition within an i

l additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l 3.3.5 Should the redundant equipment or system as specified in Section 3.3.3 become inoperable, as defined in Specification 1.3, the reactor shall not remain critical and be placed in cold shutdown condition within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

1 Proposed Amendment 164 3-21a

RANCHO SECO UNIT 1 4

TECHNICAL SPECIFICATIONS i

y Limiting Conditions for Operation Bases The requirements of Specification 3.3.1 assure that, before the reactor can be made critical, adequate safety features are operable.

Two high pressure injection pumps and two decay heat removal pumps are specified. However, only one of each is necessary to supply emergency coolant to the reactor in the event of a loss-of-coolant accident.

Both core flooding tanks are required gp)a single core flood tank has insufficient inventory to reflood the core.t' The borated water storage tank is used for two purposes:

A.

As a supply of borated water for accident conditions.

Asasupplyofboratedwater{orfloodingthefueltransfercanal B.

during refueling operation.(2 390,000 gallons of borated water are supplied for emergency core cooling and j

Reactor Building spray in the event of a loss-of-core coolant accident.

q This amount fulfills requirements for emergency core cooling. The borated water storage tank minimum volume of 390,000 gallons is based on refueling volume requirements. Heaters maintain the borated water supply at a temperature to prevent freezing.

The boron concentration is set at the amount of boron required to maintain the core 1 percent subcritical at 70*F without any control rods in the core.

This concentration is 1585 ppm boron while the minimum value specified in the tanks is 1,800 ppm boron.

(a)

Motor operated valves shall be placed in the closed position and power supplies deenergired.

The requirement that one BHST isolation valve shall be open assures a static head to the injection pump not lined up to the makeup tank.

The post accident Reactor Building cooling may be accomplished by two spray units or by a combination of two emergency cooling units and one spray unit.

The specified requirements assure that the required post accident components are available.

The spray system utilizes common suction lines with the decay heat removal system.

If a single train of equipment is removed from either system, the other train must be assured to be operabl,e in each system.

Proposed Amendment 164 3-22 l

l

RANCHO SECO UNIT-1

' TECHNICAL SPECIFICATIONS

' Limiting Conditions for Operation When'the reactor is critical, maintenance is allowed per Specification 3.3.2'

~

provided requirements in Specification 3.3.3 are met which assure operability of the duplicate components. Operability'of the specified components shall be based on the' results of testing as required by Technical Specification j

4.5.

The maintenance period of up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable if the

- operability of equipment redundant to that removed from service is demonstrated immediately subsequent to removal. -The basis of acceptability 1

is a low likelihood of failure within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months following such demonstration.

1644 See USAR paragraph 9.4.2.3(m) for the details which establish the maximum I

water temperature of the Nuclear Service spray ponds at 95*F.

l In the event that the need for emergency core cooling should occur, functioning of one train (one high pressure injection pump, one decay heat removal pump and both core flooding tanks) will protect the core and in the 1

event of a main coolant loop severance, limit the peak clad temperature to less than 2,200*F and the metal-water reaction to less than 1 percent of the clad.

l The nuclear service cooling water system consists of two independent, full capacity (3)100 percent redundant systems, to ensure continuous heat removal.

The requirements of Specification 3.3.4 assure that the decay heat removal system will not be overpressurized, resulting in a LOCA that bypasses i

I containment. Two in-series check valves function as a pressure isolation barrier between the high pressure reactor coolant system and the lower.

pressure decay heat removal system extending beyond containment. Valve leakage limits provide assurance that the valves are performing their intended isolation function.

Proposed Amendment.No. 164 3-22a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for. Operation Table 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS (C')

(A)

(B)

Operator Action if l

Functional Unit Minimum Operable Minimum Degree Conditions of Columns A l

Channel s of Redundancy and B Cannot be Met

{

Reactor Protection System 1.

Manual pushbutton 1

0 Bring to hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 2.

Power r&nge instrument 3(a) 1(a)

Bring to hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months channel 3.

Intermediate ra.'ge 1

.0 Bring to hot shutdown within 12 instrument channels hours (b) i 4.

Source range instrument 1

0 Bring to hot shutdown within 12 164

channels hours (c)(d) 5.

Reactor coolant temperature 2

1 Bring to hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months instrument channels 6.

Pressure-temperature 2

1 Bring to hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months instrument channels 7.

Flux / imbalance / flow 2

1 Bring to hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months j

instrument channels l

(a)

For channel testing, calibration or maintenance the minimum number of operable channels may be two and a degree of redundancy of one for a maximum of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3 (b)

When 2 of 4 power range instrument channels are greater than 10 percent full power, hot shutdown is J

not required.

(c)

When 1 of 2 intermediate range instrument channels is greater that 1010 amps, or 2 of 4 power range instrument channels are greater than 10 percent full power, hot shutdown is not required.

164-'

(d)

Verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.5.2.1 within one hour and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , l

l l

1 Proposed Amendment No. 164 3-27 1

1)

RANCHO SECO UNIT 1 TECHNICAL-SPECIFICATIONS Limiting Conditions for Operation l

3.5.2 Control Rod GrouD and' Power Distribution Limits Acolicability This specification applies to power distribution and operation of control-rods during power operation.

Obiective To assure an acceptable core power distribution during power. operation,.to set a limit on potential reactivity insertion from a hypothetical control rod ejection, and to assure core subcriticality after a reactor trip.

S Sp3cification 3.5.2.1 The available shutdown margin shall be not less than 1% Ak/k with the highest worth control rod fully withdrawn.~

If the shutdown j

margin is less than 1% Ak/k then, within one hour, initiate and 1

continue boration until the required shutdown margin is established.

3.5.2.2. Operation with inoperable rods:

A.

Operation with more than one inoperable rod as defined in Specification 4.7.1 and 4.7.2.3 in the safety or regulating j

rod banks shall not be permitted.

j 164-~

Action

]

a.

If a control rod in the regulating and/or safety rod banks is i

declared inoperable in the withdrawn position as defined in j

Specification paragraph 4.7.1.1 and 4.7.1.3, an evaluation shall be i

initiated immediately to verify ~the existence of 1% Ak/k hot i

shutdown margin.

Boration may be initiated to increase the available rod worth either to compensate for the worth of the inoperable rod or until the regulating banks are fully' withdrawn, whichever occurs first.

b.

If within one hour of determination of an inoperable rod as defined in Specification 4.7.1, it is not determined that a 1% Ak/k hot shutdown margin exists combining the worth of the inoperable rod with each of the other rods, the reactor shall be brought to 164-the hot shutdown condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months unless this margin is established.

c.

Following the determination of an inoperable rod as defined in Specification 4.7.1, all rods shall be exercised by a' movement until indication is noted but not exceeding 2 inches within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and exercised weekly until the rod problem is solved.-

d.

If a control rod in the regulating or safety rod groups is 164-declared inoperable per 4.7.1.2, power shall be. reduced within one hour to 60% of the thermal power allowable for the reactor coolant pump combination, and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the Nuclear Power-trip setpoint shall be reduced to less than or equal to 70% of the thermal power allowable for the reactor coolant pump combination.

Proposed Amendment No. 164 3-31

l RANCHO SECO UNIT 1 TECHNICAL' SPECIFICATIONS Limiting Conditions for Operation 3.5.4-INCORE INSTRUMENTATION Applicability

~164-Hnen the incore instrumentation is used for surveillance of the REACTOR -

POWER IMBALANCE or QUADRANT POWER TILT.

.j Obiective l

Te specify the functional and operational requirements of the incore instru-mutation system.

Specification j

Above 80 percent of operating power determined by the reactor coolant pump 164-~

combination, reference Table 2.3-1, at least 23 individual incore detectors l

shall be operable to assist in the periodic calibration of the out-of-core,

l detectors in regard to the core imbalance trip limits.

The detectors shall 1

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

'i 3.5.4.1 Axial Imbalance j

A.

Three detectors in each of 3 strings shall lie in the same i

axial plane with 1 plane in each axial core half.

B.

The axial planes in each core half shall be symmetrical about the core mid-plane.

C.

The detector shall not have radial symmetry.

3.5.4.2 Radial Tilt A.

Two sets of 4 detectors shall lie in each' core half.

Each set of 4 shall lie in the same axial plane.

The two sets in the same core half may lie in the same axial plane.

I B.

Detectors in the same plane shall have quarter core radial l

symmetry.

l 164-Action J

Hith less than the specified minimum incore detectors arrangement OPERABLE,

do not use incore detectors for the applicable monitoring functions, and l

reduce power to 1 801. of the allowable power.

]

I i

Proposed Amendment No. 164 3-36 l

j

- 1 RANCHO SECO UNIT 1-TECHNICAL SPECIFICATI0NS' i

Limiting' Conditions for Operation 164--

3.5.6 EMERGENCY SHUTDOWN INSTRUMENTATION L

Aeolicability At all' times when the reactor is critical.

Specification The emergency shutdown instrumentation channels shown in Table

-1 3.5.6-1 shall be OPERABLE with readouts displayed external to.the Control Room.

Action 1

a.

With the number of OPERABLE emergency shutdown instrumentation

~

channels less than required by Table 3.5.6-1, restore the inoperable channel (s) to OPERABLE status within-7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l, M.e1 The OPERABILITY of the emergency shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT SHUTDOHN of the facility from locations outside of the control room.

(This capability is required in the event control room habitability is-lost and is consistent with General Design Criteria 19 of Appendix A to.10 CFR 50.)

4 i

I Proposed Amendment No. 164-3<38d x

y

RANCHO SECO UNIT 1

. TECHNICAL SPECIFICATIONS 4-Limiting Conditions for Operation 164-TABLE 3.5.6-1 EMERGENCY SHUTDOHN INSTRUMENTATION (Panel H2SD)

' Minimum Number of Channels

___In s t rument Operable

1. Hide Range OSTG Level 1/0STG

2. Hide Range OSTG Pressure 1/OSTG

3. Pressurizer Level 1

4. Hide Range Reactor Coolant Pressure 1

5. Hide Range Reactor Coolant Hot _ Leg Temperature-1/ loop

6. nide Range Reactor Coolant Cold Leg Temperature 1/ loop

7. Source Range Neutron Flux Indicator

1

8. Makeup Tank Level 1

i

Deenergized and disconnected except when control room habitability is lost.

i

.l Proposed Amendment No. 164 3-38e

RANCHO SECO UNIT 1-TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.6 REACTOR BUILDING epolicability 164-~

Applies to the containment when the reactor core reactivity,'kert, is 20.99.

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:

I 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.

164*

A_ttlRH Hithout primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT SHUTDOWN 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 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

3.6.3 Positive reactivity insertions which would result in the reactor being subcritical 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 Reactor Building internal pressure exceeds 1.5 psig or vacuum exceeds -1.5 psig.

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

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

If, under reactor critical. operating conditions an automatic l

containment isolation valve is determined to be inoperable, the other containment isolation valve in the line shall be tested to i

insure operability.

If the inoperable valve is not restored within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the reactor shall be brought to the cold' shutdown condition within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the valve will be placed in a safety features position.

Proposed Amendment No. 164 3-39

I' RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation' j

^

3.9 SPENT FUEL POOL applicability 164-Applies to the Spent Fuel Pool Cooling System and Spent Fuel Pool water level whenever spent fuel is being stored in the pool.

QbitCliEft l

To provide for adequate cooling of the Spent Fuel Pool to ensure that the pool temperature is kept low enough to prevent boiling, 164-and to maintain an adequate water level to ensure sufficient shielding.

Soecifi ation 1

3.9.1 One train of the Decay Heat Pemoval Systerc (DHRS) must be put in service to provide alternate 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.

3.9.2 If a train of the DHRS is being used to provide alternate cooling for the Spent Fuel Pool, it shall be declared inoperable for other purposes and the provisions of Technical Specification 3.3.2 shall apply unless the reactor is in Cold Shutdown.

3.9.3 Use of the DHRS for Spent Fuel Pool cooling shall be limited to no more than 100 cumulative hours (when not in Cold Shutdown) in any 12-month period.

3.9.4 Reactor shutdown must be initiated within I hour if the Spent Fuel Pool bulk coolant temperature reaches,180*F, and the reactor must be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .,

164-3.9.5 At least 37 feet of water shall be maintained in the spent foal 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 mrem /hr or less; Action With the requirement of Specification 3.9.5 not satisfied, suspend all 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 .

i

/

Proposed Amendment No. 164 3-46a

j 7 f

RANCHO SECO UNIT 1 7

l/

TECHNICAL SPECIFICATIONS

[

Limiting ConditionM[for Operation

! t

\\

1 l

BASIS

. g a

3:

2>

This specif; cation provides a method to. ensure that the Spent Fuel Podi' bulk

-temperature'daes not each theLboiling point. The use of a train of the Decay HeatsRemoval System (DHPS),'per Operating Procedure A.21, Section 7.3, providesfAsnedsate alternate cooling ctpabilityLto ensure this.

Either tr? w of t,hk f)h4$ can easily be lined 'up for Spent fuel Pool' cooling' by opening two l

manue vdives (DHS-032 and DHS-055 or'056), one motor operated valve :

1 (HV-26@7 cr:46), and startingnthe appropriate decay heat pump (P-26'lA or B).

However, 23nce use ds' th) DHRS t' rain for Spent Fuel Pool cooling effectively removes it'from its nor' mal service, an operating duration limit of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months per 12-month period is. imposed.

164-The arrangement o s ent fuel storage racks provide a minimum of 23 feet of-water shielding over stored fuel assemblies to limit radiation at the surface of the water to no more than 2.5 mrem /hr during the storage period.

37 feet of water in the spent fuel pool ensures that at least 23 feet of" waf er is maintained over the top of the irradiated fuel assemblies (active t

fuel) seated in the storage racks.

References 3

[1]

LicensingReportfarHighDensitySpentFuelStorageRacksIor Rancho Seco.

[2]

Time to Boil Calculation, Supplement No. 2 to Thermo-Hydraulic Calculations for Rancho Seco Nuclear Station; Report No. TM-661.

?

i.

f t

i e

Proposed Amendment No. 164 3-46b

%%u

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation l

164--

3.10 SECONDARY SYSTEM ACTIVITY Applicability All modus from HEATUP-COOLDOWN to POWER OPERATION, inclusive.

Objective To limit the maximum secondary system activity.

Specification i

The reactor shall not remain critical if the iodine 131actividinthe 164-~

secondary side of a steam generator exceeds 0.2 pc/ml.

164-Action 1

Should the specified value be exceeded, the reactor shall be brought to HOT SHUTDOWN in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or less and in COLD SHUTDOWN in the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Bains for the purpose of determining a maximum allowable secondary coolant activity, the activity contained in the mass released following a loss of load accident is considered. As stated in FSAR paragraph 14.1.2.8.3, 1

224,000 pounds of water are released to the atmosphere via the relief valves. A site boundary dose limit of 1.5 rem is used. This is the recommended annual dose limit to,the thyroid for general population.

(I)

The whole body dose is negligible since any noble gases entering the secondary coolant system are continuously vented to the atmosphere by the condenser air ejector, thus, in the event of a loss of load incident there are only small quantities of these gases which would be released.

1 131 is the significant isotope because of its low MPC in air and because the other iodine isotopes have shorter half-lives, 'and therefore, cannot I

build up to significant concentrations.in the secondary coolant, given the limitations on primary system leak rate and technical specification limiting activity. One-tenth of the contained iodine it assumed to reach the site bpyndary, making allowance for plateout and retention in water droplets.

1831 is assumed t 3Sontribute 70 percent of the total thyroid dose based l

on the ratio of I to the, total iodine isotopes given in Table 11-3 of the FSAR.

i The maximum inhalation dose at the site boundary !s then as follows:

Dose (rem)-CiVBUIF-(0.1)X/Q 164-~

Ci-Secondary coolant activity (0.286 sc/ml 1131 equivalent)

V-Secondary sater volume relgasgdcto atmosphere (102.m3) 8-Bresthing rate (3.47 x 10-* m3/sec)

Grou,ng)levelreleasedispersionfactor(8.51x10-4 X/Q -

sec/m Proposed Amendment No. 164 i

3-47 u

1. -

s RANCHO SECO UNIT'1 TECHNICAL SPECIFICATIONS Surveillance Standards.

)

i

'4.

' SURVEILLANCE STANDARDS i

1644 4.0 General Surveillance Requirements 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for.0peration unless otherwise stated-in an individual Surveillance Requirement.

4.0.2 Failure to perform a Surv'ei11ance Requirement within the allowed-surveillance interval, defined by Specification 1.9,'shall constitute noncompliance with the OPERABILITY requirements for.a H'

Limiting Condition for Operation.

The time limits of the Action' (time) requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The Action (time) requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the completion of the surveillance when the allowable outage time limits y

of the Action (time) requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ._

Exceptions to these requirements are stated in the individual specifications and Table 4.0-1.

4.0.3 Entry into an OPERATIONAL MODE or other specified condition shall l

not be made unless the-Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified.

Exceptions to these requirements are stated in the individual specifications and Table 4.0-1.

i Proposed Amendment No. 164 i

4-0

164-TABLE 4.0-1 Applicability of Specifications 4.0.2 and 4.0.3.

(The "NA" indicates that the provisions of Specification (s) 4.0.2 and/or 4.0.3 are not applicable to the sections identified).

Section Specification 4.0.2 Specification 4.0.3 4.1 4.2 4.3 4.4 N/A N/A 4.5 4.6 4.7 4.8 4.9 4.10 4.11 4.12 4.13 4.14 4.15 4.16 4.17 4.18 4.19 N/A 4,20 N/A 4.21 N/A 4 12 N/A 4.23 4.24 4.25 N/A 4.26 N/A 4.27 N/A 4 28 1

4.29 N/A 4.30 4.31

The provisions of Specification 4.0,2 are not applicable to the Subsequent Visual Inspection Period.

164-Proposed Amendment No. 164 4-Oa

RANCHO SECO UNIT 1 TECHNICAL-SPECIFICATIONS Surveillance Standards.

164--

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.

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

4.1 OPERATIONAL SAFETY REVIEH Specification 4.1.1 The minimum frequency and type of surveillance required for. reactor 164-protection system, safety feature protection system, Process Instrumentation, and Emergency Shutdown instrumentation 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 effective full power days using the incore instrumentation detector system.

flani Check Failures such as blown instrument fuses, defective indicators, faulted' amplifiers which result in " upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system.

Furthermore, such failures are, in many cases, revealed by alarm or annunciator action. Comparison of output and/or state of independent channels measuring the same variable supplements this type of built-in.

u 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.

Proposed Amendment No. 164 4-1 4

1

RANCHO SECO UNIT 1 TECHNICAL SPECIF8 CATIONS Surveillance Standards Table 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks

42. Reactor Building drain accumulation tank level NA NA R

43. Incore neutron detectors M(1)

NA NA (1) Check functioning, including functioning of computer read-out and/or recorder readout.

44.

a.

Process and area radi-ation monitoring system W H

Q b.

Containment Area Monitors W

NA R

164**

c.

Chlorine Detector W

M R

45.

Emergency plant radiation Instruments M(1)

NA R

(1) Battery check 46.

Environmental air monitors M(1)

NA R

(1) Check functioning 47.

Strong motion accelerometer Q(1)

NA R

(1) Battery check

48. Auxiliary Feedwater Start Circuit a.

Phase Imbalance /Under-power RCP S

NA R

b.

Low Main Feedwater Pressure NA M

R

49. Pressurizer Water Level M

NA R

50. Auxiliary Feedwater Flow Rate M

NA R

164*

51. Spent Fuel Pool Level W(1)

NA R

(1)

Daily during refueling when moving fuel or

52. EMOV Power Position control rods.

Indicator (Primary Detector)

M NA R

53. EMOV Position Indicator j

(Backup Detector)

M NA R

j T/C or Acoustic j

54. EHOV Block Valve Position Indicator M

NA R

55. Safety Valve Position In-dictator (Primary Detector) M NA R

T/C

56. Safety Valve Position In-dictator (Backup Detector)

Acoustic M

NA R

Proposed Amendment No. 164 4-7b l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS l

Surveillance Standards Table 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Descriofdon Check Test Calibrate Remarks

)

164*

82. Spray Pond Water Temperature D

NA R

83. Spray Pond Water Level D

NA R

ime.cgtnev Shutdown Instrumentation i

84. Wide Range OSTG Level M

NA R

85. Wide Range OSTG Pressure Pressurizer Level M

NA R

l

86. Wide Range Reactor Coolant Hot Leg Temperature M

NA R

87. Wide Range Reactor Coolant Cold Leg Temperature H

NA R

88. Wide Range Reactor Coolant Pressure M

NA R

89. Source Range Neutron Flux Indicator" NA NA R

90. Makeup Tank Level M

NA R

1 S - Each Shift M = Monthly P = Prior to each startup if not done previous week l

\\

D = Daily Q = Quarterly R = Once during the refueling interval W = Weekly SY = Semiannual

Deenergized and disconnected sxcept when control room habitability is lost.

l l

Proposed Amendment No. 164 4-7g

. __ __O

J RANCHO SECO UNIT 1,

TECHNICAL SPECIFICATIONS Surveillance St'ndards a

TABLE 4.1-2 s

MINIMUM EQUIPMENT. TEST FREQUENCY Item Test Freauency

~1.

. Control rods Rod drop times of all Each refueling shutdown L

full length rods i

2.

LControl rod Movement of'each rod Every two weeks j

movement-3.

Pressurizer code Setpoint Note 3 safety valves

]

4.

Main steam safety' Setpoint Note 3' valves q

'i 5.

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

6.

Turbine steam stop Movement of each valve Monthly valves 164~

7.

Reactor coolant Leakage See Specification system 4.2.3.1 8.

Charcoal and high Charcoal and HEPA Each refueling interval efficiency filters filter for iodine and at any time work on and particulate filters could' alter removal efficiencies.

their integrity.

DOP test on HEPA filters.

Freon test on 1

charcoal filter units.

1 l

9.

Fire pumps and Functional Monthly I

power supplies i.i I

10.

Reactor Building Functional

~ Each refueling isolation trip interval 11.

Spent fuel Functional Each refueling cooling system interval prior to

~~

1 fuel handling 12.

Turbine Overspeed Calibration Each refueling Trips' interval I

i

)

Proposed Amendment No. 164

'4-8 1

i

' RANCHO SECO UNIT.1 TECHNICAL SPECIFICATIONS

' Surveillance Standards, TABLE 4.1-2.- (Continued)

MINIMUM EQUIPMENT TEST FREQUENCY 164*-

Item Test Freauency 13.

Internals Vent Manual actuation, (1)

Each' refueling interval.

gotevisual' inspection, Valves verify valve not stuck open Functional Each refueling interval-eachvalve(jpstof_

14.

Reactor Coolant System High Point Vents 15.

Low Temperature Functional (5)

Prior to RCS. temperature Overpressure

. decreasing below 350'F-Protection (EMOV) 164-~

16.

EMOV block valve Fun'ctional (6)

Quarterly.

1.

Verifying thrcugh manual actuation that the valve is~ fully open with a force of 1 400 lbs.-(applied vertically upward).

2.

Check visually accessible surfaces to evaluate observed surface irregularities.

3.

Tested in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as-required by 10 CFR 50, Section 50.55a(g), except where specific written relief. has been grante6 by the NRC pursuant to 10 CFR 50 Section 50.55a(g)(6)(1).

4.

Cycle each valve in the vent path through at least one complete cycle of full travel from the control room and verify the flow of gas through the system vent path. Verify all manual isolation valves in each vent path are locked in the open position.

5.

EMOV block valve closed during test.

164-~

6.

EMOV closed during test.

l Proposed Amendment No.164 4-8a

=_

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards i

164-4.2.3 Leakaae Surveillance I

4.2.3.1 Reactor Coolant System leakage shall be demonstrated to be within each of the limits of Specification 3.1.6.1 by:

a.

Monitoring the containment atmosphere particulate radioactivity monitor by verifying the monitor is indicating below the alarm setpoint at least once per day when the reactor is critical or in HOT SHUTDOWN.

b.

Monitoring the containment sump (drain accumulator tank) inventory and discharge at least once per day when the reactor is critical or in HOT SHUTDOWN.

c.

Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2155 10 psig at least once per 31 days when the reactor is critical or in HOT SHUTDOWN.

d.

Performance of a Reactor Coolant System water inventory balance at least once per week'when the reactor is critical or in HOT SHUTDOWN.

l 9:

Proposed Amendment No. 164 4-13

)

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS

]

Surveillance Standards 4.4 REACTOR BUILDING

4. 4.1 -

CONTAINMENT ~ LEAKAGE TESTS Anolicability.

Applies to containment leakage.

Qhjective To. verify that' leakage from the Reactor Building is maintained within 1

allowable limits.

8 Sp1cification 4.4.1.1 Integrated Leakaae Rate Tests 4.4.1.1.1 Calculated Peak Pressure Leakage Rate 164*

The containment. leakage rates shall be demonstrated at the specified test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR.50 using the methods and provisions of ANSI N45.4 - (1972), and a test pressure, Pa, of 52 psig with an acceptance criterion of 75 percent of the Reactor Building atmosphere held at that pressure.

1

^!.

i l

Proposed Amendment No. 164 4-15

--_ -_ _ _ a

I RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 164-~

4.4.1.1.2 Conduct of Tests A.

The test duration shall be at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless experience from at least two prior tests on similar vessels provides evidence of the adequacy of a shorter test duration, i

B.

Test accuracy shall be verified by supplementary means, such as measuring the quantity of air required to return j

to the starting point or by imposing a known leak rate to demonstrate the validity of measurements.

C.

Closure of containment isolation valves for the purpose of the test shall be accomplished by the means provided for normal operation of the valves without preliminary

)

exercises or adjustment.

l 164-~

4.4.1.1.3 Frequency of Test After the initial preoperational leakage rate. test, two integrated leakage rate tests shall be performed at approximately equal intervals between each major shutdown for inservice inspection to be performed at 10 year intervals.

In addition, an integrated test shall be performed at each 10 year interval, coinciding with the inservice inspection shutdown.

The test shall coincide with a shutdown for major fuel reloading.

1 164-~

4.4.1.1.4 Corrective Action and Retest If repairs are necessary to meet the criteria of 4.4.1.1.1, 164-the integrated leak rate test need not be repeated provided local leakage rate measurements are made before and after repair to demonstrate that the leakage rate reduction achieved by repairs reduces the overall measured integrated leak rate to an acceptable value.

~

l l

l Proposed Amendment No. 164 4-16

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS:

Surveillance Standards

164-4.4.1.1.5 Report of Test Results Each integrated leak rate test will be th'e. subject.of a summary technical report which will include a description'of-test methods used and a summary of local leak' detection tests. Sufficient data and analysis shall be included too.

o show that.a stabilized leak rate was attained and to. identify all significant required correction factors such as those.

associated with humidity and barometric' pressure, and all significant errors such as.'those associated with..

. instrumentation sensitivities and data scatter.

l l

j I

I Proposed Amendment No. 164-4-16a i

I

'I

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.4.1.2 Lqcal Leakage Rate Tests 4.4.1.2.1 Scope of Testing The local leak rate shall be measured for each of the U

following components:

-l (1)

Personnel. hatch (2)

Emergency hatch (3)

Equipment hatch seals (4)

Fuel transfer tube seals (5)

Fuel transfer tube-shroud bellows l

(6)

Reactor Building normal sump drain line I

(7)

Reactor coolant pump seal water outlet line (8)

Reactor coolant pump seal inlet line (9)

Reactor Building equalizing.line (10)

Decay Heat Removal-inlet lines (11)

Reactor Building spray inlet lines (12)

High pressure injection lines (13)

Electrical penetrations (14)

Reactor Building purge inlet line (15)

Reactor Building purge outlet line (16)

Reactor Building atmosphere sample lines (17)

- Letdown to purification demineralized line (18)

Pressurizer. relief tank gas sample line (19)

Reactor coolant system vent header (20)

Pressurizer relief tank nitrogen supply line (21)

Pressurizer sample line (22)

Reactor coolant drain tank header (23)

Reactor Building hydrogen sample line (24)

H2 recombiner penetration 164-(25)

CRD cooling' water supply (26)

Reactor Building nitrogen supply header l

(27)

Demineralized water 1

(28)

Service air.

(29)

Core Flood Tank fill and nitrogen. supply (30)

Core Flood Tank drain and sample *

(31)

Steam Generator drain l

(32)

Auxiliary steam (33)

Component cooling water inlet (34)

Component cooling water outlet-(35)

CRD cooling water return Exemption is required for testing inboard isolation valves in the reverse direction..This is necessary to preclude draining the Core Flood Tanks.

Proposed Amendment No. 164 4-17

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.4.1.2.2 Conduct of Tests 164-The containment leakage rates shall be demonstrated at the specified test schedule and shall be determined in conformance with the criteria specified in Appendix J of.10 CFR 50 using the methods and provisions of ANSI N45.4-1972 and a test pressure, Pa, of 52 psig with'an acceptance criterion of 0.06 weight percent of the Reactor Building j

atmosphere held at that pressure for a minimum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.1.2.3 Test Frequency Local leak detection tests shall be performed at a frequency i

of at least each refueling interval, except that:

(a) The equipment hatch and fuel transfer tube seals shall be additionally tested after each opening.

This testing shall be done prior to containment integrity being required by Specification 3.6.1.

(b) The personnel and emergency hatches shall be tested between the inner and outer doors at a pressure not less 1

that 52 psig semi-annually.

(c) The persor.nel and emergency hatches' inner and outer door 0-ring seals shall be tested within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each opening when containment integrity is required in Specification 3.6.1.

Test pressure for the 164-personnel and emergency hatches' 0-ring seals shall be 9.5 0.5 psig. (Appendix J of 10CFR50 is exempted) l

{l1 The leak rate (L ) established at the reduced t

164--

pressure of 9.5 0.5 psig shall be extrapolated to the leak rate (La) that will occur at the i

calculated peak containment pressure of 52 psig using the following formula:

La - 5.2 Lt (2) The extrapolated leak rate (La) will be added to the local leak rates established for the other 4

components and the total must meet the criterion of 4.4.1.2.2.

l l

Proposed Amendment No. 164 4-18

..I RANCHO SECO UNIT 1 TECHNICAL SPECIF1 CATIONS Surveillance Standards -

(d) The Containment purge and. equalizing valves shall be tested at least once every 6 months.

(e) The Containment purge valves shall'be. tested prior to the initial purge on each cold shutdown and prio'r to reaching hot shutdown during heatup for a return to -

operation. A test conducted.for this section may be applied to satisfy the requirement for a 6-month test of section (d) above if it is conducted within that

]

interval. -If the equalizing valves are not tested with the purge valves under this section, their 6-month test requirement must~still be met.

1644 (f). At least once per 31 days by verifying that all penetrations not capable of being closed by OPERABLE containment. automatic isolation valves and required to be closed during accident conditions.are closed by.

valves, blind flanges, or deactivated automatic valves secured in their positions.

Exceptions to this are those valves listed in Table 3.6-1, and any other valves, blind flanges, and deactivated automatic valves which are located inside q

the containment and are locked, sealed or otherwise secured in the closed position.

These penetrations 1

I shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

i

~

i I

i Proposed Amendment ~No. 164 l

4-18a 1

1 RANCHO SECO UNIT.1 i

TECHNICAL SPECIFICATIONS Surveillance Standards 164-4.6.4 Each 125 volt DC battery and battery charger shall be demonstrated OPERABLE:

a.

At least weekly by verifying that:

1.

The electrolyte level of the pilot cell is above the plate separators, 2.

The pilot cell specific gravity, corrected to 77'F and full electrolyte level, is no greater than 0.010 below the average corrected value of all connected cells taken from the previous month, 3.

The pilot cell voltage is greater than 2.07 volts, and i

4.

The total. battery terminal voltage is greater than or equal to 125 V DC when on float charge.

b.

At least monthly by verifying that:

1.

The electrolyte level of each cell is above the plate separators, 2.

The electrolyte temperatures in a representative sample of cells consisting of at least every sixth cell are within 5'F, 3.

The average specific gravity of all connected cells, corrected to 77*F and full electrolyte level, is greater than or equal to 1.200, 4.

The minimum specific gravity, corrected to 77'F and full electrolyte level, of each connected cell is no greater I

than 0.010 below the average value of all the connected I

cells, 5.

The voltage of each connected cell is greater than 2.07 volts, and 6.

The total battery terminal voltage is greater than or equal to 125 volts DC when on float charge.

c.

At least every re-fueling interval by verifying that:

1.

The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration, and spaces between cells and between each end cell and its battery rack are within the required seismic design tolerances, Proposed Amendment No. 164 4-35e

RANCHO SECO UNIT 1 TECHNICAL. SPECIFICATIONS I

f Surveillance Standards 164-2.

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

i 3.

The total resistance of all cell - to - cell and terminal connections is less than or equal to 20% above an established base-line or benchmark value, and 4.

The battery charger will supply at least the established i

current output necessary to re-charge the battery following an emergency discharge in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or less.

d.

At least once per refueling interval, during COLD SHUTDOWN, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status, all of the actual or simulated emergency loads for the design duty cycle or load profile when the battery is subjected to a service test.

e.

At least once per 60 months, during COLD SHUTDOWN, by verifying that the battery is at least 80% of the i

manufacturer's rating when subjected to a performance l

l discharge test.

This performance discharge test may be performed in lieu of the battery 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.

l 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 energized at least once per 7 days by verifying correct breaker alignment and indicated power availability with an

)

i overall voltage of greater than or equal to 125 volts DC and j

l 120 volts AC, respectively.

l 1

l l

Proposed Amendment No. 164 4-35f

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.9 REPORTING REQUIREMENTS (Continued) i 164-6.9.1.3 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 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 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 accordance with 10 CFR 50, Appendix'G,Section V.C.

6.9.2 Environmental Recorts 6.9.2.1 ennual Radiological Reports Annual reports covering the activities of the unit, as described below, for the previous calendar year shall be sumitted prior to March 1 of each year following initial criticality.

Reports required on an annual basis shall include:

A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man rem exposure, according to work and job functions, *e.g.,

reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.

The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements.

Small exposures, totaling less than 20*/. of the individual total dose, need not be accounted for.

In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

6.9.2.2 Annual Radioliaical Environmental Ooerittina Reoort 6.9.2.2.1 Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.

The initial report shall be submitted prior to May 1 of the year following i

initial criticality.

Proposed Amendment No. 164 164--

6-12a

o RANCHO SECO UNIT 1 I

TECHNICAL SPECIFICATIONS Administrative Controls REPORTING REQUIREMENTS (Continued) 6.9.2.2.2 The annual radiological environmental operating reports shall include summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the report period, including a comparison with j

preoperational studies, operational controls (as appropriate),

and previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on 1

the environment.

The reports shall also include the results of the land use censuses.

If harmful effects or evidence of i

irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course j

of action to alleviate the problem.

The annual radiological environmental operating reports shall include summarized and tabulated results in the format of Table 6.9-1, of all radiological environmental samples taken during the report period.

In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the following:

a summary

]

description of the radiological environmental monitoring f'

program; including sampling methods for each sample type, size and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equipment used; a map of all sampling locations keyed to a table giving distances and directions from one reactor; the result of land use censuses, and the results of licensee participation in the Interlab Comparison Program.

The annual report shall also include information related to Specification 4.29.

6.9.2.3 Semiannual Radioactive Effluent Release Reoort Routine radioactive effluent release reports covering the operation of the unit during the previous six months of operation shall be submitted within 60 days after January 1 and July 1 of each year.

The period of the first report shall begin with the date of initial criticality.

Proposed Amendment No. 164 164-6-12b

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls REPORTING REQUIREMENTS (Continued) i 6.9.2.3.1 The radioactive effluent release reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guido 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Hastes and Releases of j

Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," with data summarized j

on a quarterly basis, following the format of Appendix B j

thereof.

The radioactive effluent release reports shall include the release of gaseous effluents during each quarter, as outlined in Regulatory Guide 1.21, with the data summarized on a quarterly basis, following the format of Appendix B thereof.

A summary of meteorological conditions during the release of gaseous effluents will be retained on-site for two years.

In addition, any changes to the Offsite Dose Calculation Manual will be submitted with the Semiannual Radioactive Effluent Release Report.

The radioactive effluent release reports shall include an assessment of the radiation doses from radioactive effluents to individuals due to their activities inside the site boundary during the report period. All assumptions used in making these assessments (e.g., specific activity, exposure time, and location) shall be included in these reports.

The radioactive effluent release reports shall include the following information for all unplanned releases to unrestricted areas of radioactive materials in gaseous and liquid effluents:

a.

A description of the event and equipment involved.

b.

Cause(s) for the unplanned release.

c.

Actions taken to prevent recurrence.

d.

Consequences of the unplanned release.

The radioactive effluent release reports shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter, as outline in Regulatory Guide 1.21.

The releases of effluents shall be used for determining the gaseous pathway doses.

The assessment of radiation doses shall be performed in accordance with the Offsite Dose Calculation Manual (ODCM).

l The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) or (00CM) made i

during the reporting period, as provided in Specifications j

6.14 and 6.15.

164-Proposed Amendment No. 164 6-12c 4

y RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS

)

Administrative Controls tiONTHLY REPORT l

6.9.3 Routine reports of operating statistics, including narrative summary of operating and shutdown experience, of lifts of the Primary System Safety Valves or EHOVs, of major safety. related maintenance, and tabulations of facility changes, tests or experiments required pursuant to 10 CFR 50.59(b), shall be j

submitted on a monthly basis to the Office of Management Information and Program Control, U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, with a copy to the Regional

)

Office, postmarked no later than the 15th day of each month following the calendar month covered by the report.

LICENSEE EVENT REPORT 6.9.4 The LICENSEE EVENT REPORTS of Specification 6.9.4.1 below, f

including corrective actions and measures to prevent recurrence, shall be reported to the NRC as Licensee Event Reports, j

Supplemental reports may be required to fully describe final resolution of occurrence.

In case of corrected or supplemental reports, a License Event Report shall be completed and reference shall be made to the original report date, pursuant to the requirements of 10 CFR 50.73.

)

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, I

as a minimum, a completed copy of a licensee event report form, I

pursuant to 10 CFR 50.73 and the guidance of NUREG-1022.

a.

(i)

The completion of any nuclear plant shutdown required by the plant's Technical Specifications; or 1

(ii) Any operation or condition prohibited by the plant's Technical Specifications; or j

(iii) Any deviation from the plant's Technical Specifications authorized pursuant to 10 CFR 50.54(x).

1 l

b.

Any event or condition that resulted in the condition of the j

nuclear power plant, including its principal safety barriers, j

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

(i)

In an unanalyzed condition that significantly compromised plant safety; (ii)

In a condition that was outside the design basis of the plant; or (iii) In a condition not covered by the plant's operating and emergency procedures.

l Proposed Amendment No. 164 1

164~

6-12d

]

i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS t

Administrative Controls i

' LICENSEE EVENT REPORT (Continued) c.

Any natural phenomenon or other external condition 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.

d.

Any event or condition that resulted in manual or automatic l

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 1

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; 2.

Remove residual heat; 3.

Control the release of radioactive material; or 4.

Hitigate the consequences of an accident.

f.

Events covered in paragraph 6.9.4.1.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 condition; 2.

Remove residual heat; 3.

Control the release of radioactive material; or j

t 4.

Hitigate the consequences of an accident.

l Proposed Amendment No. 164 164--

6-12e

-]

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls

.L_LCENSEE EVENT REPORT (Continued) 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 exceeded 2 times the limiting combined Maximum Permissible Concentration (MPC)

(see Note 1 of Appendix B to 10 CFR 20) at the point of entry into the receiving water (i.e., unrestricted area) for all radionuclides except tritium and dissolved noble gases, when averaged over a time period of one hour.

l 1.

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 J

releases, or radioactive releases.

j.

Failure of the pressurizer EMOVs or Primary System Safety Valves.

q Special Reoorti 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).

l 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 Speci fication 4.4.1.1).

1.

Annual Inspection 2.

Tendon Stress Surveillance 3.

End Anchorage Concrete Surveillance I

l 4.

Liner Plate Surveillance l

Proposed Amendment No. 164 j

i 64~

6-12f

RANCHO SECO'UNITL1

-l -

'r TECHNICAL SPECIFICATIONS

Administrative Controls i

l C.

Inservice Inspection Program j

D..

Inoperable' Accident Monitoring Instrumentation 3-0 days

' I (3.5.5)

E.

Status of Inoperable Fire Protection Equipment q

0 F.

' Inoperable Emergency Control Room /TSC.

i Ventilation Room Filter System.

G.

Radioactive Liquid Effluent Dose 30 days (3.17.2)

H.

Noble Gas Limits 30 days-(3.18.2) 1

.I.

Radiciodine and Particulate 13i0 days (3.18.3)-

J.

Gaseous Radwaste Treatment 30 days (3.19)

K.

Radiological Monitoring Program -

30 days (3.22).

L.

Monitoring Point. Substitutions 30 days (3.22)

M.

Deleted N.

Fuel Cycle Dose 30 days (3.25) 0.

Deleted P.

Steam Generator Tube inspection 30 days (4.17.5) r

]

.i i

J

-a d

Proposed Amendment No.-164'

'164--

6-12g q

l

. i

l FACILITY CHANGE SAFETY ANALYSIS ATTACHMENT II Proposed Amendment No. 164 Page 1 of 16 DESCRIPTION:

Proposed Amendment No. 164 makes various modifications and addi-tions to Rancho Seco Generating Station Technical Specifications (see Enclosure 1 for specific changes / additions).

f REASON FOR CHANGES:

]

Purpose Proposed Amendment No. 164 incorporates the most significant of j

the. findings from the " Technical Review Report Evaluation of-Rancho Seco Nuclear Generating Station Technical Specifications".

This review was performed for the NRC and was based upon the fol-lowing criteria:

1)

Technical Specification provisions include definitive or quantitative performance criteria, 4

2)

Action Statements adequately address each key element j

of the respective Limiting Condition for Operation, 1

3)

Surveillance requirements are provided and are adequate to demonstrate operability, and 4)

Rancho Seco Technical Specifications. provisions are not I

substantially less conservative or comprehensive than current regulatory requirements (including Standard Technical Specifications).

Of the approximately 190 comments generated by this review, the District and the NRC agreed on Rancho Seco implementing the items

]

addressed by this proposed amendment.

EVALUATION AND BASIS FOR SAFETY FINDINGS:

Systems, Subsystems. Comoonents Affected Proposed Amendment No. 164 makes changes to multiple sections of Technical Specifications which impact various safety-related sys-tems.

These impacted systems include:

1) reactor coolant sys-tem, including reactor coolant pumps and pressurizer electromatic rotor operated valve (EMOV), 2) control rod drive system, 3) de-cay heat removal system, and 4) reactor coolant system leak de-

)

tection system.

1

FACILITY CHANGE SAFETY ANALYSIS ATTACHMENT II Proposed Amendment No. 164 Page 2 of 16 1

Safety Functions of Affected Systems / Components I

k The reactor coolant system (RCS) provides heat removal capability to the reactor during all modes of plant operation.

Addition-1 ally, the reactor coolant system serves as a barrier to prevent i

radionuclides in the reactor coolant from reaching the atmosphere (USAR section 4. 2.1.1).

The reactor coolant pumps circulate coolant through the reactor, OTSGs, and interconnecting piping.

Each of the four reactor coolant pumps take a suction on the two OTSGs through separate suction lines.

The pumps discharge into

{

lines leading to the reactor vessel.

j i

The reactor coolant system pressurizer establishes and maintains the reactor coolant system pressure within prescribed limits and i

provides a steam surge chamber and a water reserve to accommodate l

reactor coolant density changes during operation (USAR section 4.2.2.3).

Overpressure protection consists of two code safety valves and one electromatic relief valve.

The pressurizer electromatic relief valve (EMOV) is mounted on a separate nozzle on the top head of the pressurizer.

The main valve operation is controlled by the opening or closing of a j

pilot valve that causes unbalanced forces to exist on the main valve disc.

The electromatic relief valve setpoint was established to be less than the safety valve setpoint, but high enough to prevent operation of the valve during transients.

No safety function is required of the EMOV.

Its capacity is based on proper sequencing of the normal control functions of the ICS turbine bypass system, turbine system, and feedwater system l

during the normal transients.

(USAR section 4.2.4.2, Pressurizer l

Electromatic Relief Valve.)

The control rod drive mechanism (CRDM) positions the control rod within the reactor core and indicates the location of the control rod with respect to the reactor core.

The speed at which the control rod is inserted or withdrawn from the core is consistent with the reactivity change requirements during reactor operation.

For conditions that require a rapid shutdown of the reactor, the shim safety rod drive mechanism releases the Control Rod Assembly l

(CRA) and supporting CRDM components permitting the CRA to move j

by gravity into the core.

The reactivity is reduced during such a rod insertion at a rate sufficient to control the core under any operating transient or accident condition.

(USAR section 3.2.4.3, Control Rod Drive Mechanisms.)

The decay heat removal system removes decay heat from the core k

l and sensible heat from the reactor coolant system.

The system l

also provides auxiliary spray to the pressurizer for complete j

i depressurization, maintaining the reactor coolant temperature l

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FACILITY CHANGE SAFETY ANALYSIS AsfACEMENT II Proposed Amendment No. 164 Page 3 of 16

{

u during refueling, and provides a meansJfor filling and draining

]

the fuel transfer canal.

In the event of a loss-of-coolant accident, the system injects borated water intolthe reactor vessel for long-term emergency cooling.

(USAR section 9.5, Decay Heat Removal, and section 6.2, Emergency Core Cooling System.)

The RCS leak detection system has the safety function of preven-ting Loss of Coolant accidents by early detection and warning.

As amplified by the Bases of Technical Specification section 3.1.6, "Every reasonable effort will be made to reduce reactor coolant leakage including evaporative losses to the lowest pos-sible rate and at least below 1 gpm in order to prevent a large leak from masking the presence of a smaller one..."

RCS leak detection is based on three separate and diverse methods:

1) sump level, 2) radioactivity, and 3) RCS inventory calculation (USAR section 4.2.3.7 and Technical Specification section 3.1.6).

Effects on Safety Functions / Analysis of Iffects on Safety Functions z

The items modified by Proposed Amendment No. 164 (as listed in En-closure 1) are addressed individually belcu:

1)

Add a shutdown requirement if the RCS activity exceeds 43/E.

Present Technical Specifications limit the total fission product activity of the reactor coolant due to nuclides with half-lives longer than 30 minutes to 43/E microcuries per gram whenever the reactor is critical.

(E is the average beta and gamma energies per disintegration, in MeV.)

This limit on coolant activity ensures that the public is ade-quately protected and that the whole-body annual dose at the site boundary will not exceed 0.5 Rem.

However, no action is stipulated if the RCS activity exceeded this limit.

This modification adds an ACTION statement to section 3.1.4.1 of Rancho Seco Technical Specifications.

The proposed state-ment directs that, if reactor coolant specific activity

(" specific activity" replaces the existing " total fission product activity") exceeds 43/E microcuries/ gram, the plant must be in HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

This specification closely corresponds to section 3.4.9 (Specific Activity) of Standard Technical Specifications.

2)

Add a new specificat' ion to address " Ultimate Heat Sink".

Two new specifications are added to Rancho Seco Technical Specification (section 3.3.1.D., Nuclear Service Cooling Water System, and section 3.3.1.E, Nuclear Service Raw Water Cooling System).

These sections address the operability of

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I FACI.LITY CHANGE SAFETY ANALYSIS ATTACHMENT II Proposed Amendment No. 164 Page 4 of 16

'the Nuclear Service Cooling Water system loops and the i

Nuclear Service Raw Water system spray ponds.

Section j

3.3.1.E. requires a minimum Raw Water spray pond water level of 5'4" and a maximum spray pond water temperature of 95 degrees F.

The. basis for the maximum water temperature of 95 degrees F is discussed in paragraph 9.4.2.3(m) of the USAR and states, in part:

"The existing nuclear service raw water outlet temperature of 95 degrees F...is a temperature which can readily be maintained by the spray ponds even un-der conditions which limit heat dissipation, and therefore

)

95 degrees F was the upper limit for cooling water tempera-ture specified to component manufacturers".

Each of these added sections contain an ACTION statement which provides direction in the event the specification is not met.

These requirements for operability closely correspond to the requirements stated in sections.3.7.4 (Service Water System) and 3.7.5 (Ultimate Heat Sink) of Standard Technical Speci-fications.

4 Spray pond water temperature instrumentation and spray pond 1

level instrumentation (line items 82 and 63) are added to 1

Table 4.1-1, Instrument Surveillance Requirements.

This ensures that the instruments will be demonstrated as i

functioning correctly to determine the operability of the j

spray pond.

3)

Rewrito the RCS Leakage specification to include time and

)

action criteria for each statement.

l These changes are also addressed in paragraph #21 in this Safety Analysis section.

This modification rewrites the RCS leakage specification (present Technical Specifications sections 3.1.6.1 through 3.1.6;10).

The rewritten specification (proposed section 3.1.6.1, paragraphs a. through f.) delineates the limits on leakage from the reactor coolant system and unidentified leakage.

The associated ACTION statements specify the ac-tion to be taken if leakage limits are e::ceeded for:

pressure boundary leakage (ACTION statement a.), and other-than-pressure boundary leakage (ACTION statement b.).

ACTION statement c. specifies the operability requirements for reactor coolant leak detection systems.

The Bases for this specification explain that any pressure boundary leakage is unacceptable in that it may be indicative of a materials failure leading to a gross breach of RCS l

integrity.

This modification is based on Standard Technical Sp,acifications section 3.4.7.2 (Ossrational Leakage).

Proposed ACTION statement c. is dxisting Technical I

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y-J FACILITY CHANGE SAFETY ANALYSIS ATTACkMENT II I

Proposed Amendment No. 164 Page 5 of 16 k

Specification section 3.1.6.7.

a Surveillance Standard 4.2.3 (Leakage Surveillance) is added to Rancho Seco Technical Specifications.

This section specifies the method by which RCS leakage will be demonstra-ted to be within the limits specified in section 3.1.6'1.-

These requirements are based on the Surveillance Require-ments listed in Standard Technical Specifications section 4.4.7.2.1.

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Additionally, a definition of " Leakage" 10 added to Tech-nical Specifications in proposed section 2.'2.1(.

t 4)

Add a time 34mit for jnoperable codtrol rod limits to reduce power.

E' ACTION statements (a. through d.) are added to Rancho Seco Technical specification ae'Jtion 3.5.2.2 A.

to specify the actions to be taken for opcration with inoperable control rods.

Operability of the control rods is necessary to 1) ensure that acceptable power distribution limits are main-tained and 2) ensure that minimum shutdann nargin is, main-tained.

/

In ACTION statement b., a time limit of 6'Nours is estab-itshed to bring the reactor to hot shutdows if the determin-ation is made that a 1%

k/k hot shutdown margin does not exist.

In statement d.,

a new, one-hour time limit'is es-tablished to reduce power to 60% of the thermal power allow-able for the reactor coolant pump combination if a control rod is declared inoperable.

(only the tiws linit is added; the direction to reduce power because.of an inoperable con-trol rod presently existsfin the current Technical Specifi-cations.)

Additionally, 4. requirement is added in statement

d. to reduce the Nuclear Power trip setpoint to equal to or less than 70% of the thermal power allowable for the exist-ing reactor coolant pump combination within the next four hours.

l These requirements conform to section' 3.1.3.1.c;3.d) of a

Standard Technical Specifications.

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i 5)

Add statement for restoration / evaluation if heatup/cooldown limits are exceeded.

These changes are also addressed in paragraph #13 in this safety Analysis section.

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i FACILITY CHANGE SAFETY ANALYSIS ATTACHMENT II Proposed Amendment No. 164 Page 6 of 16 i

This modification adds requirements to Rancho Seco Technical i

Specification section 3.1.2.2 (Heatup Cooldown), to verify l

that RCS temperature and pressure are within limits at least once per 30 minutes during evolutions such as heatup, cool-down, and hydrostatic testing.

The modification also adds an ACTION statement to this section that describes the ac-tions to be taken if heatup or cooldown rates are exceeded.

These limits and actions help reduce the effects of thermal gradients on the reactor vessel wall.

l 1

These Technical Specification modifications closely corres-l pond to the ACTION statement in section 3.4.10.1 (Pressure /

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Temperature Limits) of Standard Technical Specifications.

]

6)

Add sp2cific values and tolerances and acceptance criteria j

for battery testing.

~

This modification deletes existing requirements for the testing of the 125v DC system (section 4.6.4 of existing Technical Specifications) and substitutes specific values, tolerances, and acceptance criteria for battery testing.

The new requirements specify surveillance at weekly, month-ly, and refueling intervals to demonstrate the operability j

of the 125 volt DC battery and battery charger.

These criteria are based on Surveillance Requirements sec-tions 4.8.2.3.2 and 4.8.2.4.1 of Standard Technical Specifi-cations.

i 7)

Add requirement to close the EMOV block valve if the EMOV becomes inoperable.

Section 3.1.1.4 (Pressurizer Electromatic Relief Valve) of j

the present Rancho Seco Technical Specifications, specifies l

the nominal setpoint of the pressurizer electronatic relief

(

valve (EMOV).

This modification adds a requirement for op-erability of the EMOV, and specifies an action if the EMOV l

becomes inoperable.

Operation (and therefore, operability)

J of the EMOV minimizes the undesirable opening of the pres-surizer code safety valves.

The action added as part of this modification (3.1.1.4.B.a.)

conforms to section 3.4.5 (Relief Valve) of Standard Tech-nical Specifications.

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FACILITY CHANGE SAFETY ANALYSIS ATTACHMENT II J

Proposed Amendment No. 164 Page 7 of 16 I

1 8)

Add requirement that an RCP be in operation during operation j

greater than 280 degrees F.

1 This modification adds an additional requirement to Rancho 1

Seco Technical Specifications (3.1.1.1.D. ) which specifies

]

that one reactor coolant pump be in operation whenever RCS j

temperature exceeds 280 degrees F.

The 280-degree F figure approaches the upper temperature limit for operability of the Decay Heat Removal System, as stated in the Bases for Technical Specification section 3.1, Reactor Coolant System

("The decay heat removal system suction piping is designed for 300 degrees F and 300 psig; thus, the system can remove decay heat when the reactor coolant system is below this temperature".)

The operation of one reactor coolant pump provides sufficient heat removal capability to remove decay heat and provides adequate flow to ensure mixing, prevent stratification, and produce gradual reactivity changes j

during boron concentration reductions.

]

This requirement is based on Standard Technical Specifica-tion section 3.4.1.2.b.

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9)

Add a requirement that a decay heat pump be in operation continuously when RCPs are not running (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action).

Rancho Seco Technical Specification section 3.1.1.5 (Decay Heat Removal), specifies that two of the listed coolant loops be operable when RCS average temperature is less than 280 degrees F.

(The listed coolant loops are RCS loop A, RCS loop B, DHR loop A, and DHR-loop B.)

This modification adds a requirement to section 3.1.1.5.A. that one of the four loops be in operation continuously to remove decay heat when RCS average temperature is less than 280 degrees F.

A one-hour exception to this requirement is permitted, pro-vided that no operations take place which could dilute the boron concentration, and core outlet temperature is main-tained at least 10 degrees less than saturation temperature.

This modification is based on the requirements of Standard Technical Specifications section 3.4.1.3.

10)

Add to definitions " Policy" statements, i.e.,

change of i

modes to be completed within specified time limits and LCO l

restrictions to a mode where Technical Specifications do not apply.

This modification adds a definition of " Action" (1.2.13) to the list of definitions (existing Technical Specification section 1.).

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FACILITY CHANGE SAFETY ANALYSIS ATTACHMENT II Proposed Amendment No. 164 Page 8 of 16 Additionally, this modification adds " General Limiting Con-ditions for Operation" to section 3.0 of Rancho Seco Tech-nical Specifications (proposed sections 3.0.1, 3.0.2, 3.0.3, and 3.0.4).

These LCOs provide direction on general compli-ance with Technical Specifications, including:

3.0.1) an overall statement on the relationship between Specifications and ACTION requirements, 3.0.2) a definition of non-compliance, 3.0.3) clarification of time limits for ACTION requirements, and 3.0.4) limitations on MODE changes when LCOs are not met.

(NOTE:

For clarification purposes, the word " time" is added to ACTION in sections 3.0.1 through j

3.0.4.)

l Time limitations on ACTION requirements are added by this j

modification throughout section 3 of Rancho Seco Technical Specifications.

If a section 3 specification has no time limit on the ACTION requirement, the 1-hour limit specified by proposed section 3.0.3 is judged acceptable, and section 3.0.3 applies.

Section 3.0.3 delineates the action to be taken for circumstances not directly provided for in the AC-TION statements and whose occurrence would violate the in-tent of the specification.

i Section 3.0.4 allows entry into an operational mode when i

conditions for Limiting Conditions for Operation are not

(

l met, but when conformance with the resulting ACTION require-l ments permits continued operation of the facility for an un-j limited period of time.

Entry into an operational mode will I

not be made when conditions for Limiting Conditions for Op-eration are not met and the associated ACTION requires a i

shutdown.

However, this does not prevent the entry into an operational mode as required to comply with an ACTION state-ment.

These sections (3.0.1 through 3.0.4) closely cor-i respond to standard Technical Specifications sections 3.0.1 through 3.0.4, respectively.

Table 3.0-1 is added to specify those Technical Specifica-i tion sections for which 3.0.3 and/or 3.0.4 do not apply.

Additionally, this modification adds " General Surveillance Requirements" to section 4.0 of Rancho Seco Technical Speci-fications (proposed sections 4.0.1, 4.0.2, and 4.0.3).

These sections provide direction on general compliance with surveillance requirements, including:

4.0.1) when surveil-lance requirements shall be met, 4.0.2) the impact of fail-ing to meet surveillance requirements, and 4.0.3) limita-tions on MODE changes when surveillance requirements are not met.

These additions (4.0.1 through 4.0.3) closely corres-pond to Standard Technical Specifications sections 4.0.1,

l FACILITY CHANGE SAFETY ANALYSIS ATTACHMENT II Proposed Amendment No. 164 Page 9 of 16 4.0.3, and 4.0.4, respectively.

Table 4,0-1 is added to specify those Technical Specifica-l tion sections for which 4.0.2 and/or 4.0.3 do not apply.

11)

Decrease shutdown criteria fer " Secondary System Activity" from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to HOT SHUTDOWN to HOT STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN in the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Section 3.10 (Secondary System Activity) of Rancho Seco Technical Specifications limits iodine-131 activity in the secondary side of the OTSGs while the reactor is critical.

This modification deletes the present " Applicability" state-ment, and substitutes an " Applicability" of:

"All modes i

from HEATUP-COOLDOWN to POWER OPERATION, inclusive."

The j

modification also adds an ACTION statement which specifies actions to be taken (and their time requirements) if the io-dine level exceeds its limit.

The time limits being added conform to those given in Stand-i l

ard Technical Specification section 3.7.1.4.

12)

Add a surveillance requirement for testing the EMOV block valve.

l This modification adds a surveillance requirement for the l

l EMOV block valve to Rancho Seco Technical Specification Table 4.1-2 (Minimum Equipment Test Frequency).

The re-quirement being added (item 16) directs that the valve be 1

functionally tested on a Quarterly basis.

A footnote speci-1 fies that the EMOV be fully closed during the test.

13)

Add a surveillance requirement for logging RCS temperature at 30 minute intervals during heatup and cooldown transi-tions.

These changes are also addressed in paragraph #5 in this j

Safety Analysis section.

This modification adds requirements to Rancho Seco Technical Specification section 3.1.2.2 (Heatup Cooldown), to verify I

that RCS temperature and pressure are within limits at least once per 30 minutes during evolutions such as heatup, cool-down, and hydrostatic testing.

During heatup (and other evolutions), the thermal gradients in the reactor vessel wall produce thermal stresses.

During evolutions which cause thermal changes within the reactor vessel, the chance of exceeding pressure / temperature limitations is increased.

More frequent observation of RCS temperature and pressure

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FACILITY CHANGE BAFETY ANALYSIS ATTACHMENT II Proposed Amendment No. 164' Page 10 of 16 I

will minimize the potential of violating these limits.

This requirement is based on Standard Technical Specifications surveillance Requirements, section 4.4.10.1.1.-

An ACTION statement is also added to Technical Specification

=section 3.1.2.2 which stipulates the actions to be taken in the event that.the heatup and.cooldown rates are exceeded.

This ACTION-statement, which applies to all Heatup-Cooldown

[

sections of Technical Specifications (3.1.2.2 through 3.1.2.5) is based on the ACTION statement in Standard Tech-nical Specifications section 3.4.10.1.

14)

Add a surveillance requirement that valves outside the con-tainment be checked once per 31 days.

This modification adds a new requirement to existing Rancho Seco Surveillance Standard 4.4.1.2.5 (proposed section 4.4.1.2.3).

This proposed new requirement (4.4.1.2.3.f) directs the verification that all penetrations not capable of being closed by operable containment automatic isolation-l valves but required to be closed, are closed by other meth-ods (valves, blank flanges, etc.).

This verification will take place once per 31 days.

The new requirement also lists exceptions to this:

valves listed in Table 3.6-1 (Safety Features Containment Isolation Valves), blind flanges, and deactivated automatic valves.

This requirement conforms to Standard Technical Specifica-tions section 4.6.1.1.a.

j 15)

Add a surveillance to verify shutdown margin within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter upon loss of all source range instrumentation.

This modification adds a footnote to Rancho Seco Technical Specification Table 3.5.1-1 (Instruments Operating Condi-tions), item 4 - source range instrument channels.

This footnote states: " Verify compliance with the SHUTDOWN MARGIN requirements of specification 3.5.2.1 within one hour and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ."

The operability of the'incore de-tectors ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core.

If less than the minimum number of source range channels are operable, more frequent verifications of shutdown margin are needed.

i This footnote is based on Standard Technical Specifications Table 3.3-1 (Reactor Protection System Instrumentation), 'AC-l TION statement number 6.

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FACILITY CHANGE SAFETY ANALYSIS ATTACHMENT II Proposed Amendment No. 164 Page 11 of 16

)

i 16)

Add a specification for detectors on CRHVAC (STS 3/4.3.3.7).

Addressed in Proposed Amendment No. 161, CR/TSC Essential Air Filtering System.

17)

Add an LCO to go to decay heat cooling if both pressurizer code safeties become inoperable.

Rancho Seco Technical Specifications section 3.1.1.3 (Pres-surizer Safety valves) addresses the operability require-ments for the pressurizer code safety valves.

This section 3

presently specifies the limitations on operation with only i

one safety valve, but does not give any action requirement when safety valves are not operable (one or both).

This modification adds two ACTION statements to section 3.1.1.3.

The first states that, when in COLD SHUTDOWN, with j

no operable pressurizer code safeties, an operable decay heat removal loop is to be placed into operation in the 6

shutdown cooling mode.

This paragraph conforms to section 3.4.2 of Standard Technical Specifications. The second paragraph of this ACTION statement directs that, when above the HOT SHUTDOWN condition, if either pressurizer code safety becomes inoperable and cannot be made operable with 15 minutes, the plant is to be in HOT SHUTDOWN in six hours.

This paragraph is based on section 3.4.3 of Standard Tech-nical Specifications.

i 18)

Add remote shutdown monitoring instrumentation.

j This modification adds section 3.5.6 (Emergency Shutdown In-strumentation) and corresponding Table 3.5.6-1 (Emergency Shutdown Instrumentation - Panel H2SD) to Rancho Seco Tech-nical Specifications.

This section, specifies the require-ments for the operability of emergency shutdown instruments-tion channels.

Table 3.5.6-1 lists the required operable channels.

The requirement for emergency shutdown instrumen-tation ensures a capability to shut down the facility and maintain it in HOT SHUTDOWN from locations external to the control room, consistent with GDC 19 of 10 CFR 50, Appendix A.

The specification and ACTION statement given in section 3.5.6 closely conform to section 3.3.3.5 and Table 3.3-9 of i

Standard Technical Specifications.

The instrumentation listed in proposed Table 3.5.6-1 (Emergency Shutdown Instrumentation - Panel H2SD) are added (items 84 through 90) to existing Table 4.1-1 (Instrument Surveillance Requirements) of Rancho Seco Technical 1

FACILITY CHANGE SAFETY ANALYSIS ATTACHMENT II Proposed Amendment No. 164 Page 12 of 16 Specifications.

This ensures the operability requirements of proposed section 3.5.6 are met.

19)

Specify > 23 feet of water above irradiated fuel in storage racks.

This modification adds section 3.9.5 to specification 3.9 (Spent Fuel Pool) of Rancho Seco Technical Specifications.

This proposed specification requires a minimum of 37 feet of water in the spent fuel pool, or a depth of water which maintains a surface dose rate of <2.5 mrem /hr.

This depth of water corresponds to 23 feet of water above the irradia-ted fuel assemblies (active fuel) seated in the storage racks.

If this requirement is not met, the proposed ACTION statement requires that all movement of fuel assemblies be suspended and water level returned within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

This ensures that sufficient water is available to limit radiation at the surface of the water to no more than 2.5 mrem /hr.

This modification closely conforms to section 3.9.11 of Standard Technical Specifications.

Additionally, this modification replaces existing item 51 (Reactor Coolant System Subcooling Margin Monitor) of Tech-nical Specification Table 4.1-1 (Instrument Surveillance Re-quirements) with Spent Fuel Pool Level monitor.

(Subcooling Margin Monitor is duplicated as existing Item 65 of the table.)

A note is added in the " Remarks" column of proposed item 51 which calls for a daily check of Spent Fuel Pool level during refueling or when moving fuel or control rods.

20)

Table 3.5.1-1, Process Instrumentation Item 9.

Delete Ac-tion wording:

"At cold shutdown or refueling ---- ".

Addressed in Proposed Amendment No. 159, Safety Parameter Display System.

21)

Revise RCS Leakage (Specification 3.1.6), ILRT (Specifica-tion 4.4.1.1), and LLRT (Specification 4.4.1.2) to better conform to STS.

These changes are also discussed in paragraph #3 in this Safety Analysis section.

This modification completely revises Rancho Seco Technical Specification section 3.1.6 (Leakage).

Rancho Seco Tech-nical Specification section 3.1.6.1 is revised, and sections 3.1.6.2 through 3.1.6.10 are deleted.

In their place, Tech-nical Specification sections 3.1.6.1.a. through f. are in-serted, along with corresponding ACTION statements a.,

b.,

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FACILITY CHANGE SAFETY ANALYSIS ATTACHMENT II Proposed Amendment No. 164 Page 13 of 16 and c.

(Proposed ACTION statement c. is existing Technical Specification section 3.1.6.7.)

This modification closely conforms to the requirements of Standard Technical Specifi-cations section 3.4.7.2 (Operational Leakage).

Section 4.2.3 (Leakage Surveillance) is added to Rancho Seco Technical Specifications.

This section discusses the meth-ods to be used to demonstrate that Reactor Coolant System leakage is within the limits of revised specification 3.1.6.1.

This section is based on Standard Technical Speci-fication section 4.4.7.2.1.

Section 4.4.1 (Containment Leakage Tests) of the existing Technical Specifications is extensively revised.

Existing specification 4.4.1.1.1 (Calculated Peak Pressure Leakage Rate) is replaced with Standard Technical Specification i

section 4.6.1.2.

It defines the schedule and acceptance i

criteria for Integrated Leak Rate Tests.

Existing sections 4.4.1.1.2 and 4.4.1.1.5, which discuss reduced-pressure testing, are deleted, since Rancho Seco does not conduct reduced-pressure testing.

Eleven penetrations are added to 4.4.1.2.1 (Scope of Testing).

Several sections are deleted due to redundancy with existing or revised Technical Specification sections or CFR requirements.

Existing specification 4.4.1.2.2 (Conduct of Tests) is replaced with the equivalent of Standard Technical Specifications section 4.6.1.2, combined with the acceptance criteria stated in i

existing section 4.4.1.2.3 (which is deleted).

Existing specification 4.4.1.2.5 is renumbered to 4.4.1.2.3 (also see paragraph #14 in this Safety Analysis section).

Other changes are made throughout this section (4.4.1.2) to more l

closely conform to various sections of Standard Technical Specifications.

22)

Resolve item of Master List regarding lack of Action State-ments for Reactor Building Emergency Cooling.

This modification adds ACTION statements to Rancho Seco i

Technical Specifications section 3.3.1.C.

The revised sec-tion specifies actions in the following cases:

With one train of containment cooling units inoperable and both containment spray systems

operable, With two trains of containment cooling units inoperable and both containment spray systems operable, and With one train of containment cooling units inoperable and one containment spray system inoperable.

FACILITY CHANGE SAFETY ANALYSIS ATTACHMENT II Proposed Amendment No. 164 Page 14 of 16 These requirements on the operability of the Reactor Build-ing Emergency Cooling ensure that post-accident cooling of the containment atmosphere is provided.

i This modification closely conforms to the wording of Stan-dard Technical Specifications section 3.6.2.3.

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)

Other general modifications to Technical Specifications are made in this Proposed Amendment:

The Table of contents is modified as necessary to incor-porate the changes generated by this Proposed Amendment.

The definitions of operational nodes (existing sections 1.2.1 through 1.2.7) are expanded to more closely conform to those definitions given in Standard Technical Specifica-tions.

Additionally, proposed Table 1.2-1 (Operational Modes) is added to more graphically show the relationships between plant operating modes.

An ACTION statement is added to section 3.1.7 (Moderator Temperature coefficient of Reactivity) which places the plant in HOT SHUTDOWN if the moderator temperature co-efficient becomes positive above 95% power.

An ACTION statement is added to section 3.6 (Reactor Building), which specifies the action to be taken if con-tainment integrity is lost.

If integrity cannot be res-tored within one hour, the plant is to be in HOT SHUTDOWN in six hours and in COLD SHUTDOWN within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

This requirement conforms to Standard Technical Specifications i

section 3.6.1.1.

In various places throughout section 3 of Technical Specifi-cations, the ACTION statements are separated from the body of the requirements, and labeled as ACTION statements (in addition to those already discussed).

This more closely corresponds to the format used in Standard Technical Spec-ifications.

ACTION is defined in section 1.2.13 as " ACTION j

(Time) Requirements".

This definition encompasses those existing remedial measures which are stated in Technical Specification sections but not called out in formal ACTION steps.

In various places throughout section 3 of Technical Specifi-cations, the Applicability statements are clarified or added in places where none presently exist.

This more closely corresponds to the format used in Standard Technical Speci-fications.

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FACILITY CHANGE SAFETY ANALYSIS ATTACHMENT II Proposed Amendment No. 164 Page 15 of 16 In various places throughout Technical Specifications, editorial changes are made to correct errors or clarify the text.

These changes do not involve changes in intent.

Existing specifications 3.1.2.6 and 3.1.2.7 are deleted, and their text moved to proponed sections 6.9.1.3 and 6.9.1.4, respectively.

These sections are moved as they address reporting requirements and are not considered Limiting 1

Conditions for Operation.

i l

The page numbers given in Proposed Amendment No. 164 are not i

final, due to the fact that multiple Proposed Amendments are l

in the review and approval process.

These numbers will be l

finalized when the Proposed Amendments in question are ap-J proved by the NRC.

1 Summary Proposed Amendment No. 164 makes various modifications and addi~

tions to Rancho Seco Generating Station Technical Specifications (see Enclosure 1 for specific changes / additions).

The proposed amendment incorporates the most significant findings from the

" Technical Review Report Evaluation of Rancho Seco Nuclear Generating Station Technical Specifications".

Of the approxi-mately 190 comments generated by this review, the District and the NRC agreed on Rancho Seco implementing the items addressed by this proposed amendment.

l Proposed Amendment No. 164 makes changes to multiple sections of l

Technical Specifications which impact various plant safety rela-i ted systems.

These impacted systems include:

1) reactor coolant system, including reactor coolant pumps and pressurizer electro-matic motor operated valve (EMOV), 2) control rod drive system,

3) decay heat removal system, and 4) reactor coolant system leak l

detection system.

These proposed modifications to Technical Specifications do not change the facility as described in Licensing Basis Documents; there are no changes to the USAR from Proposed Amendment No. 164.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously eval-uated in the safety analysis report will not be increased because all of the changes to Rancho Seco Technical Specification to be implemented by Proposed Amendment No. 164 closely conform to, or are based on, Standard Technical Specifications.

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i FACILITY CHANGE SAFETY ANALYSIS ATTACHMENT II Proposed Amendment No. 164 Page 16 of116.

The possibility for an accident or malfunction of a'different type than any evaluated previously in the safety analysis report j

will not be created because the changes to Technical Specifica-l tions implemented by Proposed Amendment No. 164 do not change the functionality of any safety. system.

The margin of safety as defined in the basis for any Technical Specifications will not be reduced because all of the changes to Rancho Seco Technical Specification to be implemented by Proposed' Amendment No. 164 closely conform to, or are based on, Standard i

Technical Specifications..

I Therefore, Proposed Amendment No. 164 does not involve an-Unre-viewed Safety Question.

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ENCLOSURE 1 HRC_ COMMENTS REGARDING RANCH 0_SECO TECHNICAL SPECIFICATIONS 6ND SMUD PROPOSED RESOLUTIONS 1.

Add a shutdown requirement if the RCS activity exceeds 43/E.

Changed the term " Total fission product activity" to " Specific activity",

and correspondingly added " ACTION" sentence to' Spec. 3.1.4.1.

2.

Add a new Tech. Spec. to address " Ultimate Heat Sink".

Added new Tech. Spec. 3.3.1.0. and 3.3.1.E.

Section 3.3.1.D addresses NSCH operability; 3.3.1.E specifies spray pond water level and temperature. Added items 82 and 83 to Table 4.1-1.

3.

Rewrite the RCS Leakage Specification to include time and action criteria l

for each statement.

Added new Tech. Spec. 3.1.6.1 per STS 3.4.7.2 including ACTION statements

a. and b.

Made existing RSTS 3.1.6.7 into ACTION statement c.

Also added new Tech. Spec. 4.2.3 on Leakage Surveillance per STS 4.4.7.2.

4.

Add a time limit for inoperable control rod limits to reduce power.

Added ACTION statements to Spec. 3.5.2.2A.

In ACTION statement d., the statement also included re-setting the trip setpoint, per STS Spec.

j 3.1.3.1.

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5.

Add statement for restoration / evaluation if heatup/cooldown limits are i

exceeded.

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Added action statement per STS 3.4.10.1 to RSTS 3.1.2.2..

6.

Add specific values and tolerances and acceptance criteria for battery

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testing.

Deleted existing Spec. 4.6.4; replaced with new one per STS 4.8.2.3.2 and 4.8.2.4.1.

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7.

Add a requirement to close the EMOV block valve if the EMOV becomes i

inoperable.

Added Spec. 3.1.1.4B per STS 3.4.5.

8.

Add a requirement that an RCP be in operation during operation greater than 280*F.

Added Spec. 3.1.1.10.

9.

Add a requirement that a decay heat pump be in operation continuously when RCPs are not running (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action).

Modified Spec. 3.1.1.5A accordingly.

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10. Add to definitions " Policy" statements, i.e., change.of modes to be completed within specified time limits and LCO restrictions to a mode where Tech. Specs. do not apply.

1 a.

Spec. 1.2, added "(OPERATIONAL MODE - MODE)".to REACTOR OPERATING CONDITIONS.

b.

Added Spec. 1.2.13 "Acilsn".

c.

Added Spec. 3.0 " General Limiting Conditions for Operation."

d.

Added Specs. 3.0.1, 3.0.2, 3.0.3 and 3.0.4.

The first three are from the STS; Spec 3.0.4 is as revised by NRC's Generic Letter 87-09, June 4, 1987.

e.

Added action requirements throughout Section 3 except where the 1-hour action of Spec. 3.0.3 was judged acceptable, i.e., if a Section 3 spec. has no specific action required, Spec. 3.0.3 applies.

f.

Added Spec. 4.0 " General Surveillance Requirements."

11. Decrease shutdown criteria for " Secondary System Activity" from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

to HOT SHUTDONN to HOT STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and cold shutdown in the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The above action requirement, equivalent to STS 3.7.1.4, was incorporated in Spec. 3.10.

12. Add a surveillance requirement for testing the EMOV block valve.

Added Item 16, EMOV block valve to Table 4.1-2, calling for functional testing, quarterly, with the EMOV closed during the test.

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13. Add a surveillance requirement for logging RCS temperature at 30 minute intervals during heatup and cooldown transitions.

Added the equivalent of STS 3.4.10 to Spec. 3.1.2.2.

14. Add a surveillance requirement that valves outside the containment be checle.ed once per 31 days.

Added Spec. 4.4.1.2.5(f), equivalent to STS 4.6.1.1.

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15. Add a surveillance to verify shutdown margin within I hour and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter upon loss of all tource range instrumentation.

Added action note (d) to Table 3.5.1-1, Item 4 which states:

" Verify I

compliance with the SHUTDOWN MARGIN requirements of Spec. 3.5.2.1 within one hour and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

16. Add a specification for detectors on CRHVAC (STS 3/4.3.3.7).

Specification 4.10.1.E.2 submitted in Proposed Amendment No. 161, 6/29/87 calls for verification of the toxic (chlorine) gas detector isolation-signal. Definition 1.3, Item (2) specifies the operability.

Page 2 of 4

17. Add an LCO to go to' decay heat cooling if both pressurizer code safeties become inoperable.

Added new Spec. 3.1.1.3 action statements equivalent'to STS 3.4.2 and' 3.4.3.

18. Add remote shutdown monitoring instrumentation.

Added new Spec. 3.5.6, Emergency Shutdown Instrumentation with Table 3.5.6-1, equivalent to STS 3.3.3.5 and Table 3.3-9.

Also, added Emergency Shutdown Instrumentation items 84 through 90 to Table ~4.1-1, equivalent to STS Table 4.3-6.

19.

Specify 223 feet of water above. irradiated fuel in storage racks.

Added new Spec. 3.9.5 equivalent to STS 3.9.11, and replaced Item 51 Table 4.1-1, Subcooling Margin Monitor, which was redundant with Table 4.1-1, Item 65, with Spent Fuel Pool level. Added a note to Item 51 calling for daily check during refueling when moving fuel or control rods.

20.

Table 3.5.1-1, Process Instrumentation Item 9.

Delete Action wording:

"At cold shutdown or refueling Above word deletions being made in Proposed Amendment No. 159, SPDS which is still in preparation.

21.

Revise RCS Leakage (Spec. 3.1.6), ILRT (Spec. 4.4.1.1), and LLRT (Spec.

4.4.1.2) to better conform to STS.

Spec. 3.1.6.1 was revised, equivalent to STS 3.4.7.2, thus making Specs.

3.1.6.2 through 3.1.6.8 unnecessary (redundant with the action requirements of the revised Spec. 3.1.6.1), and therefore' deleted.

Applicable paragraphs of the STS Bases (3/4.4.7.2) were added.

Added new Spec. 4.2.3, Leakage Surveillance, equivalent to STS 4.4.7.2.1.

Revised Spec. 4.4.1.1 as follows:

a.

Replaced Spec. 4.4.1.1.1 with the equivalent of STS 4.6.1.2.

b.

Deleted Specs. 4.4.1.1.2 and 4.4.1.1.5 which discuss testing at reduced pressure.

Reduced pressure testing is not practiced at Rancho Seco.

c.

Added eleven penetration to the LLRT Scope of Testing, Spec.

t 4.4.1.2.1.

d.

Replaced Spec. 4.4.1.2.2 with the equivalent of STS 4.6.1.2, and combined it with Spec. 4.4.1.2.3.

e.

Deleted Spec. 4.4.1.2.4 which is an LC0 action statement covered in the action requirements of Spec. 3.1.6.

f.

Renumbered Spec. 4.4.1.25 as Spec. 4.4.1.2.3, and as committed by LER 85-027 revised the 10 psig test pressure to show it as maximum (9.5 % 0.5 psig).

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Also added Specification 1.2.1.14 (Leakage).

22.

Resolve Item No. 31 of Master List regarding lack of action statements for Reactor Building Emergency Cooling.

Add the equivalent of STS 3.6.2.3 action statement to Spec. 3.3.1C.

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ATTACHMENT III, NO SIGNIFICANT HALARDS CONSIDERATION In February and March, 1987, an evaluation of.the Rancho Seco Technical Specifications was conducted.by Parameter, Inc., under contract to the NRC, Region V.

The evaluation consisted of a comparison, for adequate safety content, of the Rancho Seco Technical Specifications (RSTS) with NUREG-0103, Rev. 4,- the Standard Technical Specifications-(STS).

The evaluation resulted in approximately 190 conditions in which the RSTS did not have equivalent specifications with the STS.

In subsequent meetings between the' District and the NRC on June 11, 1987, 17 of these findings were.

identified as requiring District review and resolution prior to plant restart.

Since then, specific NRC requests and voluntary District action has increased the "near-term" list to the 22 items contained in this Proposed. Amendment No. 164.

The proposed amendment makes changes to multiple sections of the RSTS which l

affect various plant safety related systems.

These affected systems include:

1) reactor coolant system, including reactor coolant pumps and pressurizer.

electromatic motor operated valve (EMOV), 2) control rod drive system, 3) decay.

heat removal system, and 4) reactor coolant leak detection system.

1 The District has reviewed the proposed changes against each of the criterion of 10 CFR 50.92 and has reached the following conclusions:

1 1.

The probability of occurrence or the consequences of an accident or i

malfunction of equipment important to safety previously evaluated in the safety analysis report will not be increased because all of the changes to Rancho Seco Technical Specification to be implemented by Proposed Amendment No. 164 closely conform to, or are based on, Standard Technical Specifications.

2.

The possibility of an accident or malfunction'of a different type than any evaluated previously in the: safety analysis report will not be created because the changes to lechnical Specifications implemented by Proposed Amendment No. 164 do not change the functionality of any safety system.

3.

The margin of safety as defined in the basis for any Technical Specifications will not be reduced because all of the changes to Rancho l

Seco Technical Specification to be implemented by Proposed Amendment No. 164 closely conform to, or are based on, Standard Technical Specifications.

On the basis of the above, the District concludes that the proposed changes embodied in Proposed-Amendment No. 164 are enhancements to the RSTS which do not constitute any significant hazard to the public, and in no way endangers the public's health and safety.

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ATTACHMENT IV Previously submitted Proposed Amendments have been affected by the addition of Standard Technical Specification sections 3.0 and 4.0.

When these Proposed Amendments are approved, Tables 3.0-1 and 4.0-1, of Proposed Amendment No. 164 will need to be revised.

The following is a listing of the Proposed Amendments affected and the revisions required to Tables 3.0-1 and 4.0-1 of Proposed Amendment No. 164.

Eroposed Amendment No.

Discussion (Section numbers refer to the sections in the specific Pro-posed Amendment.)

147 Add Sections 3.27 and 3.28 to Table 3.0-1.

Add Section 4.32 to Table 4.0-1.

151 Add Section 4.14.f. to Table 4.0-1.

Add "NA" for Section 4.14.f.

under both 4.0.2 and 4.0.3 on Table 4.0-1.

152 Add Section 3.5.6 to Table 3.0-1 158 Add Section 3.29 to Table 3.0-1.

Add Section 4.33 to Table 4.0-1.

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160 Add Section 3.29 to Table 3.0-1.

Add "NA" for Section 3.29 under both 3.0.3 and 3.0.4 on Table 3.0-1.

l Add Section 4.34 to Table 4.0-1.

Add "NA" for Section 4.34 under both 4.0.2 and 4.0.3 on Table 4.0-1.

Page 1 of 1

v t e Letter Sequence Other
TAC:63030, (Approved, Closed) TAC:65256, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance... Supplement, Supplement, Supplement, Supplement Administration Meeting Questions Answers Results Approval, Approval, Approval, Approval, Approval, Approval Other: A11745, Evaluation of DSR-48 Emergency Diesel Generator Crankshafts at Rancho Seco Nuclear Generating Station, ML20126J866, ML20126J973, ML20135E867, ML20149D631, ML20149F003, ML20196E677, ML20196E697, ML20206H162, ML20212B420, ML20214K863, ML20215G661, ML20234C529, ML20234C558, ML20234D524, ML20234D615, ML20234D649, ML20235M565, ML20235Q511, ML20236L543, ML20236N639, ML20236N713, ML20236S088, ML20237B265, ML20237B279
MONTHYEAR1987-12-01 [Table View]

ML20235Q511

Text

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