ML20236U082
| ML20236U082 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 11/25/1987 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20236U070 | List: |
| References | |
| TAC-65256, NUDOCS 8712020268 | |
| Download: ML20236U082 (66) | |
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. ATTACHMENT 2'
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, Revised Technical: Specification'Pages (Inc1Edes all Proposed Amendment 164'-
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS' TABLE OF CONTENTS-Section P_a9ft 1 DEFINITIONS 1-1 1.1 RATED POWER- 1-1 1.2 REACTOR OPERATING CONDITIONS ~ '1 1.2.1 C.old Shutdown 1-1 1.2.2 Eqt ShutdowJl 1-1 1.2.3 Reactor Critical 1-1 1.2.4 Hot Standby 1 q l.2.5 Power Ooeration- '- 1 -1 1.2.6 Refuelina Shutdown 1-1 [ 1.2.7 Refuelino Operation 1-2 1.2.8 Refuelina Interval 1-2 1 1.2.9 Startuo 1-2 < l.2.10 Remain Critical 1-2 1.2.11 Tg 1-2 1.2.12 Heatuo - Cooldown Mode 1-2 164-~ 1.2.13 Action 1-2 l 1.3 OPERABLE l 1.4 PROTECTION INSTRUMENTATION LOGIC 1-2 1.4.1 Instrument Channel 1-2 1.4.2 Reactor Protection System 1-2a 1.4.3 Protection Channel 1-3 1.4.4 Reactor Protection' System Loain 1-3 i 1.4.5 Safety Features-System Loaic- 1-3' 1.4.6 Dearee of Redundanry -l-3 Proposed Amendment No. 164 i-E
-RANCHO SECO UNIT'l. -
TECHNICAL SPECIFICATIONS .l TABLE OF CONTENTS' (Continued)' Sectiori Eagft 1.16 RESTRICTED AREA 1-7 1.17 . SITE BOUNDARY 1-7 1 1.18 DOSE EOUIVALENT I-1'31 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 i 2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTIVE INSTRUMENTATION 2-5' '. 164-- 3 LIMITING CONDITIONS FOR OPERATION 3-0 J
- 3.0 GENERAL LIMITING CONDITIONS FOR OPERATION 3-0 3.1 REACTOR COOLANT SYSTEM 3-1
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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 3-8 3.1.5 Chemistry 3-10 l 1 Proposed Amendment No. 164 ita l-L
-RANCHO SECO UNIT 1 TECHNICAL' SPECIFICATIONS' TABLE OF CONTENTS. (Continued)' ,
'Section PAqe j
.3.1.6 Leakaae 3-12.
l
' Moderator Temperature Coefficient of Reactivity '3-15'
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- 3.1.7 3.1.8- Low Power Physics' Testina Restrictions . 3-15b l 3.1.9 Control Rod Operation 3-16 ;
1 3.2 HIGH PRESSURE INJECTION' CHEMICAL ADDITION.'AND LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEMS 3-17 3.3 EMERGENCY CORE COOLING. REACTOR BUILDING EMERGENCY. ,, COOLING. AND REACTOR BUILDING SPRAY SYSTEMS 3-19 j 3.4 STEAM AND P0HER CONVERSION SYSTEM 3-23 1 ' 3.5 INSTRUMENTATION SYSTEMS 3-25
.3.5.1 Doerational Safety Instrumentation 3-25 3.5.2 Control Rod Groun and Power Distribution Limits 3-31 3.5.3 Safety Features Actuation System Setooints 3 ;
3.5.4 Incore Instrumentation ~3-36 3.5.5 Accident Monitorina Instrumentation 3-38a .,
,1 1644- 3.5.6 Emeroency Shutdown Instrumentation 3-38d
'3-39 I 3.6 REACTOR BUILDING 3.7 @XILIARY 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 HOIST 3-49 3.12 SHOCK SUPPRESSORS (SNUBBERS) :3-51 a 3.13 AIR FILTER SYSTEMS 3-52 3.14 FIRE SUPPRESSION 3-53 .
3.14.1 Instrumentation 3-53 3.14.2 Hater System 3-53 ) 3.14.3 Sorav and Sorinkler Systems 3-56 4 3.14.4 CO2 System 3-56 Amendment No. 164 iii .
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued) 1 Section EA91 1 164- 4 SURVEILLANCE STANDARDS 4-0 1 4.0 GENERAL SURVEILLANCE REQUIREMENTS 4-0 4.1 OPERATIONAL SAFETY REVIEH 4-1 q 1 4.2 SURVEILLANCE OF ASME CODE CLASS 1. 2. AND 3 SYSTEMS 4-10 4.2.1 Reactor Vessel Surveillance Soecimens 4-10 4-11 I 1 4.2.2 Inservice Insoection. 164-~ 4.2.3 Leakaae Surveillance 4-13 l 4.3 TESTjNG FOLLOHING OPENING OF SYSTEM 4-14 ! l l 4.4 REACTOR BUILDING 4-15 4.4.1 Containment Leakage Tests 4-15 4.4.2 Structural Intearity 4-21 4.4.3 livdrogen Purae Svstem 4-25 i 4.5 EMERGENCY CORE COOLING AND REACTOR BUILDING 4-26 COOLING SYSTEM PERIODIC TESTING i 4.5.1 Emeroency Core Coolina System 4-26 l 4.5.2 Reactor Buildina Coolina Systems 4-29 4.5.3 Decav Heat Removal System and Reactor Buildina Sorav , Syltem Leakage 4-32 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTING 4-34 ) 4.7 REACTOR CONTROL R0D SYSTEM TESTS 4-36 1 4.7.1 Control Rod Drive System Functional Tests 4-36 j 1 i 4.7.2 Control Rod Proaram Verification 1 l (Group vs. Core Positions) 4-37 l 4.8 AUXILIARY FEEDWATER PUMP PERIODIC TESTING 4-39 ) l 4.9 REACTIVITY ANOMALIES 4-40 l 4.10 EMERGENCY CONTROL ROOM FILTERING SYSTEM 4-41 Proposed Amendment No. 164 v i
1 RANCH 0SECO UNIT l' TECHNICAL SPECIFICATIONS- ] TABLE OF CONTENTS (Continued) Section- Eagg 4.11 REACTOR BUILDING PURGE EXHAUST FILTERING SYSTEM 4-42. - 4.12' AUXILIARY AND SPENT FUEL' BUILDING FILTER SYSTEMS 4-43: 4'.13 AUGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH 4-44 ENERGY LINES OUTSIDE OF CONTAINMENT 4.14 SHOCK SUPPRESSORS-(SNUBBERS) : 4-47 ' 4.15 RADI0 ACTIVE MATERIALSLSOURCES 4-48 4.16 Reserved 4-49 j 4.17- STEAM GENERATORS 4-51 4.17.1 Steam Generator Samole Selection and Inspection 4-51 4.17.2 Steam Generator Tube Samole Selection and Insoection 4-51 4.17.3 Insoettion Frequencies 4-52 4.17.4 Acceptance Criteria 4-53 4.17.5 Reports 4-54. l 4.17.6 OTSG Auxiliary Feedwater Header Surveillance 4-54 j 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 1 4.19 RADIOACTIVE LIOUID EFFLUENT INSTURMENTATION 4-63 4.20 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION 4-65 4.21 LIOUID EFFLUENTS 4-69 4.21.1 Concentration 4-69 i 4.21.2 Dose Calculations 4-72' i i 4.21.3 Liauid Holduo Tanks 4-73 4.22 GASE0US EFFLUENTS 4-74 4-74 4.22.1 Dose Rate' Proposed Amendment No. 164 164-- vi j j
I RANCHO.SECO UNIT 1 TECHNICAL SPECIFICATIONS l 41 TABLE OF CONTENTS '(Continued)- Section P_aage 4.22.2' Noble Gases 4-77
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4.22.3 Iodine-133. Tritium and Radionuclides in Particulate Form 4-78 l l 4.23 ~ GASEOUS RADHASTE TREATHENT 4-79 4.24 GAS STORAGE TANKS 4-80 , l 4.25 SOLID RADI0 ACTIVE HASTES 4 ] 4.26 RADIOLOGICAL ENVIRONMENTAL MONITORING 4-83 4.27 LAND USE CENSUS 4-86 4.28 EXPLOSIVE GAS MIXTURE 4-87 4.29 FUEL CYCLE DOSE '4-89 :]
.i 4.30 INTERLABORATORY COMPARISION PROGRAM SURVEILLANCE REQUIREMENT 4-90 l
4.31 NUCLEAR SERVICE ELECTRICAL BUILDING EMERGENCY HEATING i t VENTILATION AND AIR CONDITIONING 91 5 DESIGN FEATURES 5-1 5.1 SIIE 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 , 5.3.1 Reactor Core 5-4 l 5.3.2 Reactor Coolant System 5-4 I 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 5-6 5.4.1 New Fuel Insoection and Temocrary Storage Rack- 5-6' 5.4.2 New and Soent Fuel Storage Racks and Failed Fuel Storace Container Rack 5-6 5.4.3 New and Soent Fuel Temocrary Storage 5-6 5.4.4 Soent Fuel Pool and Storage Rack Desian 5-6 - Proposed Amendment No.164 164- - yjj 44 L /
l RANCHO SECO UNIT 1-TECHNICAL SPECIFICATIONS LIST OF. TABLES ~ IAhlst P_itut :
'164-- 1.2-1 OPERATIONAL MODES 1-2c 2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS 2 1 3-0a j
- 164-+- 3. 0- 1 APPLICABILITY OF SPECIFICATIONS 3.0.3 AND 3.0.4 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS 3-27 !
3.5.5-1 ACCIDENT MONITORING INSTRUMENTATION OPERABILITY 3-38b 3 REQUIREMENTS 164-- 3.5.6-1 EMERGENCY SHUTDOWN INSTRUMENTATION 3-38e -l 3.6-1 SAFETY FEATURES CONTAINMENT ISOLATION VALVES 3-40 j 3.7-1 VOLTAGE PROTECTION SYSTEM RELAY TRIP VALUES 3-41a 3.7-2 VOLTAGE PROTECTION SYSTEM LIMITING CONDITIONS 3-41b j 3.12-1 SAFETY.RELATED HYDRAULIC SNUBBERS 3-51a-e j 3.14-1 FIRE DETECTION INSTRUMENTS FOR SAFETY SYSTEMS 3-55, 3.14-2 INSIDE BUILDING FIRE HOSE STATIONS 3-57a. 3.15-1 RADIOACTIVE LIQUID EFFLUENT MONITORING-INSTRUMENTATION 3-61 ., 3.16-1 RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION 3-64 ! 3.22-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 3-83~ l 3.22-2 REPORTING LEVELS FOR RADI0 ACTIVITY 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 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY 4-8 ; 4.1-3 MINIMUM SAMPLING FREQUENCY 4-9 'I 4.2-1 CAPSULE ASSEMBLY HITHDRAHAL SCHEDULE AT DAVIS-BESSE 1 4-12b 4.10-1 ENVIRONMENTAL RADIATION MONITORING PROGRAM 4-42 4.10-2 OPERATIONAL ENVIRONMENTAL RADIATION MONITORING PROGRAM 4-22a 4.14-1 DESIGNATED SAFETY RELATED HYDRAULIC SNUBBERS FUNCTIONALLY 4-47d,e i TESTED ONLY AS REQUIRED BY THE SNUBBER SEAL REPLACEMENT PROGRAM 4.17-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED 4-56 l 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 GASE0US EFFLUENT HONITORING INSTRUMENTATION' 4-66 SURVEILLANCE REQUIREMENTS l Proposed Amendment No. 164 ix 1
RANCHO CECO UNIT 1 l TECHNICAL SPECIFICATIONS
- 1. DEFINITIONS The following terms are defined for uniform interpretation of these specifications.
1.1 RATED POWER i Rated power is a steady reactor core output of 2772 MHt. 164- 1.2 REACTOR OPERATING CONDITIONS (OPERATIONAL 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 1 0.99, and Tavg is no more than 200*F. Pressure is defined by Specification 3.1.2. 1 See Table 1.2-1 4 1.2.2 Hot Shutdown i 164- The reactor is in the hot shutdown condition when keff is 1 0.99, and l Tavg is at or greater than 525'F. See Table 1.2-1 1.2.3 Reactor Critical The reactor is critical when the neutron chain reaction is self-sustaining l 164-~ and keff > 0.99. ; 1.2.4 Hot Standby 1 The reactor is in the hot standby condition when all of the following ] 1644- conditions exist: (See Table 1.2-1) A. Tavg is greater than 525'F. 164-~ B. keff is >0.99. l l C. Indicated neutron power on the power range channels is less than i 2 percent of rated power. 1.2.5 Power Operation The reactor is in a power operation condition when the indicated neutron power is above 2 percent of rated power as indicated on the power range channels. See Table 1.2-1. 1.2.6 Refueling Shutdqwn l 164- The reactor is in the refueling shutdown condition when the reactor core l - is 1 0.95 and the coolant temperature at the decay heat ' reactivity, removal pump suct is no more than 140*F. Pressure is defined by keff, ion 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. l Proposed Amendment No. 164 1-1 l
l i i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS ! l Definitions 1.2.7 Refueling Operation An operation involving a change in core geor.ietry by manipulation of fuel or control rodt when the reactor vessel head is removed. 1.2.8 Refueling Interval
- Time between normal refuelings of the reactor, not to exceed 24 months for i the first refueling and 18 months thereafter without prior approval of the i NRC. l 1.2.9 Startu.a The reactor shall be considered in the startup mode when the shutdown margin ,
is reduced with the intent of going critical. 1.2.10 Remain Critical A technical specification that requires that the reactor shall not remain critical shall mean that an uninterrupted normal hot shutdown procedure will l l be completed d thin 12 hours. 1.2.11 Tg i At operating conditions T is defined as the arithmetic average of the coolanttemperaturesint$ghot and cold legs of the loop with the greater i number of reactor coolant pumps operating, if such a distinction of loops can be made, i 164-- 1.2.12 Heatuo - Cooldown The reactor is in heatup-cooldown when the range of reactor coolant temperature is greater than 200*F and less than 525'F. ! l 164- 1.2.13 Action { Action including time requirements shall be that part of a specification which prescribes remedial measures requir d under designated conditions. l 1.2.14 Leakage ! A. IDENTIFIED LEAKAGE shall be: -!
- a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as i pump seal or valve packing leaks that are captured and conducted !
to a sump or collecting tank, or j
- b. Leakage into the containment atmosphere from sources that are i 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 ,
i Proposed Amendment No. 164 1-2 a
l 1 1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS -l Definitions , 164- c. Reactor coolant system leakage through a steam generator to the , secondary system. B. UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE. C. PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System i component body, pipe wall or vessel wall, j J D. CONTROLLED LEAKAGE shall be that leakage from the reactor coolant pump seals. 1.3 OPERABLE i 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 j 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 ) purpose of satisfying the requirements of its applicable Limiting Condition for Operation provided its redundant system or component is OPERABLE with an OPERABLE normal and emergency power source. 1.4 PROTECTION INSTRUMENTATION LOGIC l 1.4.1 Instrument Channel An instrument channel is the combination of sensor, wires, amplifiers and ! ouput devices which are connected for the purpose of measuring the value of a process variable for the purpose of observation, control and/or protection. An instrument channel may be either analog or digital. ], 1.4.2 Reactor Protection System 1 i The reactor protection system is shown in Figures 7.1-1 and 7.2-2 of the FSAR. It-is that combination of protective channels and associated i 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 ; control protective trip breakers and activating relays or coils. i Proposed Amendment No. 164 ;
1644 TABLE 1.2-1 OPERATIONAL MODES Operational Reactivitiy Coolant Indicated Remarks Mode Condition Temperature Neutron keff Tavg*F Power (% of Rated Power) Power >0.99 >525 22 Operation Hot >0.99 >525 <2 Standby l l Startup >0.99 >525 <2 i <1.00 Hot 10.99 2525 0 Shutdown Heatup - 10.99 >200 0 Cooldown (525 Cold 10.99 1200 0 RCS Pressure as Shutdown defined in 3.1.2. Refueling
- 10.95 1140 at 0 RCS Pressure Shutdown DHR pump as defined suction in 3.1.2.
Refueling Reactor Vessel head removed and Refueling Shutdown conditions operation l
- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
l l Proposed Amendment No. 164 164-1-2c
c q 4- , , RANCHO SECO UNTI 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation i
- 3. LIMITING CONDITIONS FOR OPERATION 164-- 3.0 General Limiting Conditions For Operation 3.0.1 Compliance with the Limiting Conditions.for Operation' contained ,
in the succeeding Specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting. Conditions for Operation, j the associated Action including time requirements shall.be met. i 3.0.2 Noncompliance with a' Specification shall exist when the
. requirements of the Limiting Condition for Operation and.
associated Action including time requirements are not met. l within the specified time' intervals. If the: Limiting Condition.
- for Operation is restored prior to expiration of the specified time intervals, completion of the Action including time requirements is not required. ]
3.0.3 When a Limiting Condition for Operation is not met',' except as ! provided in the associated Action including time requirements, I within 1 hour action shall be initiated to place the unit in a MODE in which the Specification does not apply to placing it, as applicable, in:
- 1. At least HOT STANDBY within the next 6 hours, l
- 2. At least H0T SHUTDOWN within the following 6 hours, and i
- 3. At least COLD SHUTDOWN within the subsequent 24 hours. !
Where corrective measures are completed that permit operation under the Action including time requirements, the Action ; including time requirement may be taken in accordance with the l 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. l 3.0.4 Entry into on OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Limiting Conditions for Operation are not met and the associated Action including time requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL MODE or ' specified condition may be made in accordance with Action including time requirements when conformance to them permits continued operation of the facility for an unlimited period of time. This provision shall not prevent passage through or to OPERATIONAL-MODES as required to comply with Action requirements. Exceptions to these requirements are stated in the individual specifications and Table 3.0-1. Proposed Amendmea+ No. 164
- 3-0 '
a d
' TABLE 3.0-1 Applicability of Specifications 3.0.3,and 3.0.4. .( The "NA" indicatesLthat'the
- provisions of Specification (s) 3.0.3 and/or 3.0.4 are not applicable.to the sections
' identified).
Section. Specification 3.d.3 Specification 3.'0.4 3.1 3.2.1 3.2.2 N/A 3.3 3.4 3.5.1 3.5.2 3.5.3 3.5.4 N/A N/A 3.5.5 ' N/A __ 3.5.6 N/A 3.6 N/A i 3.7 'l i 3.8 N/A 3.9 N/A N/A 3.10 3.11 N/A N/A 3.12 3.13 # 3.14 N/A N/A 3.15 N/A N/A
- 3.16 N/A N/A l
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-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 the reactor is in Refueling Shutdown or Refueling Operation. .164-Proposed Amendmer,t No.164 3-Oa
1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 164-~ 3.1 REACTOR COOLANT SYSTEM Aonlicability 1 164-~ All operational modes. ; Obiective To specify those limiting conditions for operation of the reactor coolant i system which must be met to ensure safe reactor operations. j 3.1.1 OPERATIONAL COMPONENTS l Specification 3.1.1.1 Reactor Coolant Pumps 1 A. Pump codinations permissible for given power levels shall be ! as shown in Table 2.3-1. l I B. The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant. C. Operation at power with two pumps shall be limited to 24 hours 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. 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 subtritical, at least one pressurizer code safety valve shall be operable if all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Boiler and Pressure Vessel Code, Section III. Proposed Amendment No. 164 3-1
i 1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation j 164- Attion In COLD SHUTDOHN with no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity. changes and place an OPERABLE decay heat removal loop into operation in the shutdown cooling mode. When above the hot shutdown condition, with one pressurizer code safety ! valve inoperable, either restore the inoperable valve to OPERABLE status j within 15 minutes or be in a least HOT SHUTDOWN within 6 hours. y 3.1.1.4 Pressurizer Electromatic Relief Valve A. The nominal setpoint of.the pressurizer electromatic relief-vd ve shall be 2450 psig 10 psig except when required for cold overpressure protection. 1644 B. If the EMOV and its associated block valve are not OPERABLE whenever the reactor is in-HOT STANDBY or critical, the i following actions shall be taken:
- 1. With the EMOV inoperable, within 60 minutes either restore the EMOV to OPERABLE status or close the associated block valve and remove power from the block valve; otherwise, be in a least HOT STANDBY within the next 6 hours and in COLD SHUTDOHN within the following 30 hours. The requirement for an operable low pressure setpoint for the EMOV.for LTOP in Specification 3.2.2 is not applicable if
- the EMOV is inoperable.
3.1.1.5 Decay Heat Removal l A. At least two of the coolant loops listed below shall be 164- operable except during fuel loading and refueling. One loop shall be in operation when the coolant average temperature is below 280*F. The one operating coolant loop required need not be in operation for a maximum of one hour provided (1) no operations are permitted that would cai w dilution of the reactor coolant system boron concentrat . 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 1 - _ - - - - - _ - - - - - - - - - - _ _ - - - - - - _ - - . - _ - - - - _ - _ _
i RANCHO SECO UNIT 1 I TECHNICAL SPECIFICATIONS Limiting Conditions for Operation
- 3. Decay Heat Removal Loop (A) ,
I
- 4. Decay Heat Removal Loop (B) )
)
164- B. With less than the above required coolant loops OPERABLE, immediately initiate corrective action to return the required-coolant loops to OPERABLE status as soon as possible; be in COLD SHUTD0HN within 20 hours. 3.1.1.6 Reactor Coolant System High Point Vents A. The vent path on Loop A and vent path on Loop B shall be i operable and closed during power operation. B. The vent path on the pressurizer shall be operable and closed 4 during power operation. 1 1 C. With one of the above reactor coolant system vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the valves in the i inoperable vent path; restore the inoperable vent path.to ' OPERABLE status within 30 days. If the status is not. restored to operable in 30 days, be in HOT STANDBY within 12 hours and in COLD SHUTDOWN within the following 30 hours. , D. With two or more of the above reactor coolant system vent l l paths inoperable; maintain the. inoperable vent paths closed I with power removed from the valve actuators of all the valves-in the inoperable vent paths, and restore at least (two) of-the vent paths to OPERABLE status within 72 hours. If the status is not restored to operable in 72 hours, be in HOT STANDBY within 12 hours and in' COLD SHUTD0HN within the i following 30 hours. l Bases 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 q 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) i
^
i Proposed Amendment No. 164 l : 3-2 1 2
-4 t
ufy
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RANCHO SECO UNIT 1 ; TECHNICAL SPECIFICATIONS l l Limiting Conditions for Operation ) 3.1.2 PRESSURIZATION, HEATUP, AND C00LDOWN LIMITATIONS Specification 3.1.2.1 Inservice 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 eight EFP years of power operations, the reactor coolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with 164- Figure 3.1.2-1 and Figure 3.1.2-2 respectively. The Reactor i Coolant System temperature and pressure shall be determined to be l within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. Heatup and cooldown rates shall not exceed the rates stated on the associated figure. Action ] If heatup and cooldown rates are exceeded, stabilize the temperature and restore the temperature and/or pressure to within the limits within 30 minutes, and perform an engineering evaluation to determine the effects of ; the out-of-limit condition on the structural integrity of the Reactor { Coolant System. Through this evaluation, determine that the Reactor Coolant ' l System remains acceptable for continued operation or be in at least HOT SHUTDOWN within the next 6 hours and reduce RCS Tavg and pressure to less than 200*F and 500 psig, respectively, within the following 30 hours. This action applies to Specifications 3.1.2.4 and 3.1.2.5, below. 3.1.2.3 The secondary side of the steam generator shall not be pressurized above 200 psig if the temperature of the steam generator shell is below 130*F. 3.1.2.4 The pressurizer heatup and cooldown rates shall not exceed 100*F in any 1-hour period. l 3.1.2.5 The spray shall not be used if the temperature difference between the pressurizer and spray fluid is greater than 410*F. Proposed Amendment No. 164 3-3
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.3 HINIMUM CONDITIONS FOR CRITICALITY Specifications 3.1.3.1 The reactor coolant temperature shall be above 525'F except for portions of low power physics testing when the requirements of Specification 3.1.8 shall apply. 3.1.3.2 Reactor coolant temperature shall be above DTT + 10*F. 3.1.3.3 When the reactor coolant temperature is below the minimum temperature specified in 3.1.3.1 above, except for portions of low power physics testing when the requirements of Specification 3.1.8 shall apply, the reactor shall be subcritical by an amount equal to or greater than the calculated reactivity insertion due to depressurization. 164~ Action Hith the reactor subcritical by less than the required amount, immediately initiate and continue boration until the required SHUTDOHN MARGIN is restored. l 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. A_clign 164- Hith the reactor subtritical by less than the required amount, immediately , initiate and continue boration until the required SHUT 00HN MARGIN is restored. 3.1.3.5 Except for physics tests and as limited by 3.5.2.1 and 3.5.2.5, safety rod groups shall be fully withdrawn prior to any other reduction in shutdown margin by deboration or regulating rod I withdrawal during the approach to criticality. Following safety rod withdrawal, the regulating rods shd 11 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 i 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. l Proposed Amendment No. 164 3-6
- a RANCHO SECO UNIT l'.
TECHNICAL SPECIFICATIONS . i Limiting Conditions for Operation-I Bases.'(Continued)- , The potential reactivity insertion due.tof 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.
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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.- from a~ reduction of moderator density. 164-~
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l Proposed Amendment No. 164 ; 3-7 i
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RANCHO SECO UNIT 1 1 TECHNICAL SPECIFICATIONS: !
.i Limiting Conditions for Operation j l
3.1.4 REACTOR COOLANT. SYSTEM ACTIVITY Specification 164-~ 3.1.4.1 The specific activity of the reactor coolant due to nuclides with l half lives longer than 30 minutes shall not exceed 43/E microcuries l per gm whenever the reactor is critical. E is the average (mean) .j beta and gamma energies per disintegration, in MeV, weighted in proportion to the measured activity of the-radionuclides in 164- reactor coolant samples. Action Hith the specific activity of the reactor coolant greater than 43/E i'
- microcuries/ gram, be in at least HOT SHUTDOWN within 6 hours.
Bases The above specification is based'on limiting-the consequences of a postulated accident involving the double-ended rupture of a steam generator tube. The rupture of a steam generator tube. enables reactor coolant and its associated activity to enter the secondary system where volatile isotopes could be discharged to the atmosphere through~ condenser air-ejectors and through steam safety valves (which may lift momentarily). Since the major portion of the activity entering the secondary system is due to noble gases, the bulk of the l 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 thg tube break occurred. During that 34 minute period, a maximum of 2740 164-~ fto of hot reactor coolant will have leaked in l this is equivalent to a cold volume of 1980 ftgo the secondary system;
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The controlling dose for the steam generator tube rupture accident is the j whole-body dose resulting from immersion in the cloud of released activity. To insure that the public is adequately protected, the specific activity of o the reactor coolant will be limited to a value which will insure that the i l 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. 1 l' 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 i 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.gmeterpersecondwindspeed, resulting in a X/Q value of 8.51 x 10-4 sec/m . Proposed Amendment 164 3-8
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.6 LEAKAGE 164- Acolicability Applies whenever the reactor is in HOT SHUTDOWN thru P0HER-OPERATION, inclusive. Objective
- To monitor and limit RCS leakage.
Specification 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,
- d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and l
- e. 16 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2155 t 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 in Specification 3.3.4. Surveillance requirements are provided in Specification 4.5.1.1.B.4 and 5.
Action A. With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT SHUTDOHN within 6 hours and in COLD SHUTD0HN within the following 30 hours. j B. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours or be in at least HOT SHUTDOWN within the next 6 hours and in COLD SHUTD0HN within the following 30 hours. C. During power operation, two reactor coolant leak detection systems of different operating principles shall be in operation, with one of j the two systems sensitive to radioactivity. The systems sensitive to radioactivity may be out-of-service for 48 hours provided'two
- other means are available to detect leakage.
Proposed Amendment No. 164 i 3-12 l
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation BAS 11 Every rehsonable 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 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 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 such leakage wili 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 leaks in the order of drops.per minute through any of the walls of the primary system could be indicative.of materials failure such as by stress 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. The 10 GPM IDENTIFIED LEAKAGE limitation provides allowar:e 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 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.- Proposed Amendment No. 164 3-13
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RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS Limiting Conditions for Operation -] Bases (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. If leakage is to the Reactor Building it may be identified by one or more of l the following methods: l l A. Sumo Levels - All Reactor Building leakage is collected in the l l Reactor Building sumps. These sumps drain by gravity into a 120 I gallon Reactor Building drain accumulation tank. The drain ! accumulation tank is used to measure the drain flow with level l indicators at 20 gallons and 120 gallons. The tank is dumped into the East decay heat removal pump room sump. The frequency of dumping the accumulation tank and time interval between levels are l recorded in the Control Room and are direct measures of the flow i 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 j i l l l Proposed Amendment No. 164 164-- 3-13a
RANCHO SECO UNIT 1 , TECHNICAL SPECIFICATIONS- j Limiting Conditions for Operation 3.1.7 MODERATOR TEMPERATURE-COEFFICIENT OF REACTIVITY 164-~ Acolicability POWER OPERATION i Specification l The moderator temperature coefficient shall not be positive at power levels above 95 percent of rated power.
'l 164- Action Hith the moderator temperature coefficient at a positive value, be in at least HOT SHUTDOWN within 6 hours.
Bases i 1 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 -l percent of rated power the Final Acceptance _ Criteria will not b exceeded l with a positive moderator temperature coefficient of +0.9 x 10 g ak/k/F' ; corrected to 95 percent of rated power. All other accident analyses as s reported in the FSAR have been performed for a range of moderator temperature , coefficients including +0.9 x 10-4 Ak/k/F. I 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 value, the following j 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. 1 Proper corrections will be added for these conditions to provide a conservative moderator coefficient.
l Doppler contribution at hot zero power - During measurement of 2. the isothermal moderator coefficient at hot zero power, the i 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. l 3. Moderator temperature change - The hot zero power measurement l 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.
l Proposed Amendment 164 l 3-15
. . 1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS-Limiting Conditions for Operation 3.1.9 CONTROL R00 OPERATION 164-~ ADolicability All modes above COLD SHUTDOWN 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. 1644 Action t In the event the limits of Specifications 3.1.9.1 or 3.1.9.2 are exceeded, j the center control rod drive mechanism shall be checked for accumulation of j undissolved gases. This shall be performed within 24 hours or be in HOT- i SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following -I
- 30 hours.
1 Bases By maintaining the reactor coolant temperature and pressure as specified above, any dissolved gases in the reactor coolant system are maintained in solution. l 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. The limiting pressure versus temperature curve for dissolved gases is determined by the equilibrium pressure versus temperature curve for the . I dissolved gas concentration of 100 std. cc/ kilogram of. water. The ! equilibrium total pressure is the sum of the partial pressure of the ; dissolved gases plus the partial pressure of water at a given temperature. I The margin of error consists of the maximum pressure difference between the pressure sensing tap and lowest pressure point in the system, the maximum pressure gage error, and the pressure difference due to the maximum temperature gage error. If either the maximum dissolved gas concentration (100 std. cc/ kilogram of-water) is exceeded or the operating pressure falls below the limiting pressure versus temperature curve, the center CRDM should be checked for l accumulation of undissolved gases. Proposed Amendment No. 164 3-16 _ -_ _ ______- _ _ __- - --__ ___ - _ a
1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation i 3.2 HIGH PRESSURE INJECTION. CHEMICAL ADDITION. AND LOH TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEMS ADolicability Specification 3.2.1 applies to the operational status. of the high pressure injection and chemical addition systems. Specification 3.2.2 applies to'the operational status of the Low Temperature Overpressure Protection (LTOP) System when the.RCS temperature falls below 350*F and the RCS is not.open to atmosphere. Specification 3.2.2 is not applicable when the Reactor Vessel. ] head is removed, when any one of the 4 OTSG manways is open, when the , pressurizer heater bundle is removed, or when the pressurizer manway is j removed. Obiective j i Specification 3.2.1 provides for adequate boration under all. operating conditions to assure ability to bring the reactor to a cold shutdown condition. Specification 3.2.2 defines the necessary conditions for preventing an excessive overpressure transient to occur at low temperatures. . j Specification I 3.2.1 The reactor shall not remain critical unless the following conditions are met: 3.2.1.1 Two pumps capable of supplying high pressure injection 164~ are operable (also see Specification 3.3). 3.2.1.2 The borated water storage tank and its. flow path to the j reactor for high pressure injection are operable. 3.2.1.3 A source of concentrated boric acid solution in addition to the borated water storage tank is available and ; operable. This requirement is fulfilled by the . I concentrated boric acid storage tank. This tank shall , contain at least the equivalent of 10,000 gallons of ; I 7,100 ppm boron. System piping and. valves necessary to establish a flow path for high pressure injection shall also be operable and shall have at least the same I temperature as the boric acid storage tank. One i associated boric acid pump is operable. The concentrated boric acid storage tank water shall not be less than 70F, and at least one channel of heat tracing shall be I operable for this tank's associated piping. The-concentrated boric acid storage tank boron concentration shall not exceed 8,500 ppm boron. Proposed Amendment No. 164 3-17
-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' Acolicability 164-~ All modes from HEATUP-COOLDOWN to POWER OPERATION, inclusive. 1 Objective To define the conditions necessary to assure immediate availability of the emergency core cooling, Reactor Building emergency cooling and Reactor Building spray systems.
Specification 3.3.1 The reactor shall not remain critical, unless the following ) conditions are met: ) 164- A. Two independent ECCS subsystems (injection systems) shall be j OPERABLE with each subsystem comprised of: l
- 1. One OPERABLE high pressure injection pump.
l 2. One OPERABLE decay heat removal pump. ) l
- 3. One OPERABLE decay heat removal cooler. j 1
- 4. An OPERABLE flow path capable of taking suction from the borated water storage tank on safety injection signal.
The Reactor Building emergency sump isolation valve shall , be either manually or remote-manually operable. !
- 5. One of the two BHST isolation valves shall be open (SFV 25003 or SFV 25004). This valve may be closed during the :
quarterly valve test specified in the Specifications 4.5.1.2A and 5.4.2.2A. Action
- a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours or be in at least HOT SHUTD0HN within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
l Proposed Amendment No. 164 3-19 l
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164+ . RANCHO SECO UNIT l' l TECHNICAL' SPECIFICATIONS ]
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Limiting Conditions.for 0peration I
- 3. Each reactor. coolant system core flooding system tank shall be
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OPERABLE with:- 4 4
- 1. ' A contained borated water volume between 1010 and 1070 cubic. feet-(or a;1evel between 12.75 and 13.25 feet)'of borated water between.575 Land 625 psig.
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- 2. A boron concentration ~at not_less than 1,800 ppm. boron. 'I 1
- 3. The electrically operated discharge valve from the core 1 flood tank shall' be open when the Reactor Coolant System ]o pressure is greater than 800 psig. The breakers shall be, open and-so tagged.'
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- 4. An OPERABLE pressure instrument. channel. ;
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- 5. The electrically operated vent valve'(HV-26511 and !
HV-26512) from the core flood tank.shall be closed. The i breakers shall be open and so tagged except during normal , venting operations. ] i Action l l a. Hith one core flooding tank inoperable,- except as a result of a closed isolation valve, restore the inoperable tank to OPERABLE status within one hour or be in at least HOT SHUTDOHN within the next 6 hours and in ! COLD SHUTDOWN within the following 24 hours. i
- b. Hith any core. flooding tank inoperable due to the !
isolation valve being closed, either. immediately open the isolation valve or be in at least' HOT SHUTD0HN within one hour and be in COLD SHUTDOHN within the next 24 hours. C. Reactor Building spray system and Reactor Building emergency cooling system. The following combination of system components must be operable: , l 1. Two Reactor Building spray pumps and their associated ; spray headers with a' minimum of 32 percent NaOH solution ' in the spray additive tanks and, 1
- 2. A minimum level of 78 inches of solution shall be available in each spray additive tank.
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- 3. Four emergency cooling units, two with charcoal filter units. There are two cooling units in'each'of two emergency cooling trains (Train A and B).
Proposed Amendment No.164 3-19a
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' RANCHO'SECO UNIT 1 TECHNICALLSPECIFICATIONS Limiting Conditions lfor Operation' 9
164'- L Action i Hith o'ne train of thejabove required containment cooling units-
. inoperable and both containment. spray; systems;0PERABLE, restore >
the-; inoperable train-of cooling units to.0PERABLE status within 7 days or'be in at'least HOT!SHUTDOHN:within the next 6 hours' and'in COLD SHUTDOHN within the following 30 hours. With t'wo trains-of.the above; required cont'ainment cooling' units D inoperable, and both containment' spray systems.0PERABLE,. restore at least;one' train of cooling units ,to OPERABLE status within 72 hours'or'be in at-least HOT'SHUTDOHN within.theLnext-6 hours and:in. COLD SHUTDOHN within<the following.30 hours.
' Restore both above' required trainstofc cooling units to'0PERABLE '
status within 7 days.of initial'1oss or be in'at'least HOT-SHUTDOWN within~the next 6 hours'and.in~ COLD SHUTD0HN within. E the following-30 hours. i Hith one train of the above required containment cooling units - inoperable and.one containment spray system inoperable. restore the inoperable' spray: system to OPERABLE status within 72 hours or be in at least HOT SHUTD0HN within the'next 6 hours and in { COLD SHUTDOHN within the following 30 hours. Restore the:- inoperable train of. containment. cooling units to OPERABLE - status within 7 days of-initial loss or be in at least HOT SHUTDOHN within the next 6 hours'and in COLD SHUTDOHN within the following 30 hours. D. The borated water storage tank shall.'be 0PERABLE with:
- 1. A minimum contained borated water volume of 390,000 gallons.
- 2. A minimum concentration of 1,800 ppm of- boron, and
- 3. A minimum. water temperature of 40*F.
Action l Hith the borated water storage tank inoperable, restore the tank to OPERABLE status within one-hour.or be in at least HOT- , SHUTDOHN within the next 6 hours and in-COLD SHUTDOHN within -J the following 30 hours.
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i i y Proposed Amendment 164' .. 1 3-201 o
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'164- j RANCHO'SECO UNIT 1 1 TECHNICAL SPECIFICATIONS l t
Limiting Conditions for Operation E. Nuclear Service Cooling Water (NSCH) System
- 1. Two NSCH Loops.are OPERABLE.
Action Hith only one NSCH loop OPERABLE, restore at least two loops'to' OPERABLE status within 72 hours or'be in at least HOT SHUTDOWN within the next 6 hours and in COLD SHUTD0HN within the following 30 hours. ; F. Nuclear Service Raw Hater (NSRN) System j
- 1. Two NSRH loops shall be OPERABLE.and each nuclear service spray pond shall be OPERABLE with: )
l 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 in at least HOT SHUTDOWN within 6 hours and COLD SHUTD0HN within the following 30 hours. G. Safety features valves and-interlocks associated with each of the above systems are operable. Inoperable valves shall be placed in the safety features position.- 164-~ 3.3.2 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 164-~ in 3.3.2.2. Valve leakage shall not exceed the amounts l
indicated.
- 2. In the event that integrity of any pressure isolation valve.
specified in Table 3.3-1 cannot be demonstrated, reactor I.- operation may continue,-provided that at least two valves.in each high pressure line having a non-functional valve are in and remain in, the mode corresponding to the isolated condition.(a) 164-~ 3. If Specifications 3.3.2.1 and 3.3.2.2 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 additional 24 hours. Proposed Amendment 164 3 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions' for 0peration j 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 H 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 i the event of a loss-of-coolant accident. Both core flooding, tanks are-required the core.(iat)a single core flood tank has insufficient inventory to reflood i
- 1 The borated water storage tank is used for two purposes:
l A. As a supply of borated water for accident-conditions. I B. As a. supply of borated water for flooding the fuel transfer canal l during refueling operation.(2) ! 390,000 gallons of borated water are supplied for emergency core cooling and Reactor Building spray in the event of a loss--of-core coolant accident. . This amount fulfills requirements for emergency core cooling. The borated water storage tank minimum volume of 390,000 gallons is based on refueling l volume requirements. Heaters maintain the borated water supply at a < tcaperature to prevent freezing. The boron concentration is set at the amount of boron required to maintain the core 1 percent subtritical at 70*F l 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 deenergized. l l 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 l components are available. i The spray system utilizes common suction lines with the decay heat removal j system. If a single train of equipment is removed from either system, the other train must be assured to be operable in each system. Proposed Amendment 164 3-22
yq [znw l t 1 I RANCHO SECO UNIT'1 TECHNICAL SPECIFICATIONS
, Limiting Conditions for Operation 164-See USAR paragraph 9.4.2.3(m) for.the details which establish the maximum
- water temperature of.the Nuclear Service spray ponds at 95'F.
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
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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.
The nuclear service cooling water systein consists of two independent, full capacity 100 percent redundant systems, to ensure continuous heat removal.(3) 164-~ The requirements of Specification 3.3.2 assure that the decay heat removal system will not be overpressurized, resulting in a LOCA that bypasses 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. { 1 i
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l l I I l Proposed Amendment No. 164 . 3-22a l
RdNCHO SECO UNIT 1 i TECHNICAL SPECIFICATIONS' Limiting Conditions for'0peration Table 3' 5.1-1 INSTRUMENTS OPERATING' CONDITIONS (C) (A) (B) Operator Action if Functional Unit Minimum Operable Minimum Degree. ' Conditions of Columns A' Channel s of Redundancy and B Cannot be Met' Reactor Protection System
- 1. Manual pushbutton 1 0. ' Bring to hot shutdown within 12 hours -
- 2. Power range instrument 3(a)- 1(a) Bring to' hot shutdown within 12 hours .
channel
- 3. Intermediate range 1 0- Bring to hot shutdown within 12 ,
' instrument channels >
hours (b)- }l
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- 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 hours instrument channels .f i
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- 6. Pressure-temperature 2 1 Bring to hot shutdown within 12 hours 1 instrument channels
- 7. Flux / imbalance / flow 2 1 Bring to hot shutdown within 12 hours instrument channels (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 hours. (b) When 2 of 4 power range instrument channels are greater than 10 percent full power, hot shutdown is not required. 1' l (c) When 1 of 2 intermediate range instrument channels is greater that 1010 amps,- or. 2 of .4 power -i 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 hours.
I Proposed Amendment No. 164 3-27 a
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' RANCHO SECO UNIT,1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.2 Control Rod Grouo and Power Distribution Limits Acolicability 1
This specification applies to power distribution and. operation of control ]' rods during power operation. Obiective < l 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. j Specification 4 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 margin is less than 1% ak/k then, within one hour, initiate and continue boration until the required shutdown margin is established. l 3.5.2.2 Operation with inoperable rods: A. Operation with more than one inoperable rod as defined in l' Specification 4.7.1 and 4.7.2.3 in the safety or regulating rod banks shall not be permitted. 164~ Action
- a. If a control rod in the regulating and/or safety rod banks is !
declared inoperable in the withdrawn position as defined in i 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 ' shutdown margin. Boration may be initiated to increase the available rod worth either to compensate for the worth of the l 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% ok/k hot shutdown margin exists combining the worth of the inoperable roa-with each of the other rods, the reactor shall be brought to 1644 the hot shutdown condition within 6 hours 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 .
hours 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 hours, 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
1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.4 INCORE INSTRUMENTATION Applicability 164- When the incore instrumentation is used for surveillance of the REACTOR P0HER IMBALANCE or QUADRANT POWER TILT.- f i Obiective To specify the functional and operational requirements of the incore instru- l mentation system. ' Specification 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 shall be operable to assist in the periodic calibration of the out-of-core detectors in regard to the core imbalance trip limits. The detectors shall be arranged as follows and may be a part of both basic arrangements. 3.5.4.1 Axial Imbalance A. Three detectors in each of 3 strings shall lie in the same axial plane with 1 plane in each axial core half. i l B. The axial planes in each core half shall be symmetrical about l the core mid-plane. ; I C. The detector shall not have radial symmetry. ;
)
3.5.4.2 Radial Tilt l A. Two sets of 4 detectors shall lie in each core half. Each set l of 4 shall lie in the same axial plane. The two sets in the I same core half may lie in the same axial plane. B. Detectors in the same plane shall have quarter core radial symmetry. 164- Action Hith less than the specified minimum incore detectors arrangement OPERABLE , do not use incore detectors for the applicable monitoring functions, and
- reduce power to 1 801. of the allowable power.
I Proposed Amendment No. 164 3-36 E_ _ -
7 i RANCHO SECO UNIT 1 '
' TECHNICAL SPECIFICATIONS Limiting Conditions for Operation.
164-* 3.5.6 . EMERGENCY SHUTD0HN INSTRUMENTATION Anolicability All modes from HOT SHUTDOWN thru POWER OPERATION, inclusive. Specification . The emergency shutdown instrumentation channels shown in Table _
-i 3.5.6-1 shall be OPERABLE with readouts displayed external to the Control Room.
Action
- 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 j in HEATUP-COOLDOWN within the next 12 hours.
l Bases ; i The OPERABILITY of the emergency shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT i SHUTDOWN of the facility from locations outside of the control room. (This ! capability is required in the event control room habitability is lost and is i consistent with General Design Criteria 19 of Appendix A to 10 CFR 50.)- 1 l I l Proposed Amendment No. 164 3-38d j i
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS
- Limiting Conditions.for Operation 164-. ~ TABLE 3.5.6 EMERGENCY SHUTD0HN INSTRUMENTATION'(Panel H2SD)
( l Minimum Number j ' of Channels Instrument Operable
- 1. Hide Range'OTSG Level 1/0TSG
- 2. Hide Range OTSG Pressure' 1/0TSG'
- 3. Pressurizer Level 1
- 4. Hide Range. Reactor Coolant Pressure 1
- 5. Wide Range' Reactor Coolant Hot Leg Temperature .1/ loop
- 6. Hide Range Reactor Coolant Cold Leg Temperature 'l/ loop
- 7. Source Range Neutron Flux Indicator
- 1
- 8. Makeup Tank Level 1
*Deenergized and disconnected except when control !
room habitability is. lost. l Proposed Amendment No. 164 ! 3-38e l
)
a
RANCHO SEC0' UNIT'l' j TECHNICAL SPECIFICATIONS ;
-4 Limiting Conditions for Operation . ]
3.6 REACTOR BUILDING Applicability ; All modes from STARTUP thru POWER OPERATION, inclusive.-
~
164-~ Obiective j i 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: 1 A. Reactor coolant pressure is 300 psig or greater. 1 B. Reactor coolant temperature is 200*F or greater. C. Nuclear fuel is in the core. 164- Action Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT SHUTDOWN within the next 6 hours and in COLD
- SHUTDOWN within the following 30 hours.
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 1 by control rod motion or boron dilution whenever the containment , integrity is not intact. I 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 l 3.6-1 shall be OPERABLE with closure times as shown in Table 3.6-1. If, under reactor critical operating conditions an automatic 3 containment isolation valve is determined to be inoperable - the ; other containment isolation valve in the line shall be tested to ! insure operability. If the inoperable valve is not restored within ! 48 hours, the reactor shall be brought to the cold shutdown ; condition within an additional 24 hours or the valve will be placed in a safety features position. Proposed Amendment No. 164 3-39 !
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation-3.9 SPENT FUEL POOL j Applicability 164- tspplies.to the Spent Fuel Pool Cooling System and Spent Fuel Pool j
- water level whenever spent fuel is being stored in the pool.
Obiective To provide for adequate cooling of the Spent fuel Pool.to ensure .j that the pool temperature is kept low enough to prevent boiling, J 164- and to maintain an adequate water level to ensure sufficient
- shielding.
Specification , 3.9.1 One train of the Decay Heat Removal System (DHRS):must be put in service to provide' alternate cooling for the Spent Fuel Pool if the l bulk coolant. temperature reaches 1140*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. I 3.9.4 Reactor shutdown must be initiated within 1 hour if the Spent Fuel Pool bulk coolant temperature reaches 180*F, and the reactor must be in Cold Shutdown within 24 hours. 164- 3.9.5 At least 37 feet of water shall be maintained in the spent fuel ! 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 ] I- With the requirement of Specification 3.9.5 not satisfied, suspend all l movement of fuel assemblies and crane operations with loads in the fuel ,
- storage areas and restore the water level to within its limit within 4 hours.
I Proposed Amendment No. 164 3-46a
l 1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS i Limiting Conditions for Operation RA_SES This specification provides a method to ensure that the Spent Fuel- Pool bulk temperature does not reach the boiling point. The use of a train of the Decay Heat Removal System (DHRS), per Operating Procedure A.21, Section 7.3, provides immediate alternate cooling capability to ensure this. Either train of the DHRS can easily be lined up for Spent Fuel Pool cooling by opening two manual valves (DHS-032 and DHS-055 or 056), one motor operated valve (HV-26047 or 46), and starting the appropriate decay heat pump (P-261A or B). However, since use of the DHRS train for Spent Fuel Pool cooling effectively removes it from its normal service, an operating duration limit of 100 hours per 12-month period is imposed. 164- The arrangement of spent 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 water is maintained over the top of the irradiated fuel assemblies (active fuel) seated in the storage racks. References [1] Licensing Report for High Density Spent Fuel Storage Racks for Rancho Seco. [2] Time to Boil Calculation, Supplement No. 2 to Thermo-Hydraulic Calculations for Rancho Seco Nuclear Station; Report No. TM-661. 1 j Proposed Amendment No. 164 3-46b
l- 1 l l !
' RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 164~- 3.10 SECONDARY SYSTEM ACTIVITY ADD 11cability All modes from HEATUP-COOLDOWN to POWER OPERATION. inclusive.
Obiective To limit the maximum secondary system activity. S Specification Tne reactor shall not remain critical if the iodine 131 activity in the 164~~ secondary side of a steam generator exceeds 0.2 pc/ml. 164- Action i Should the specified value be exceeded, the reactor shall be brought to. HOT
- SHUTDOWN in 6 hours or less and in COLD SHUTDOHN in the following 30 hours.
Bases 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, 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. Cl) The whole body dose is negligible since any noble gases entering the secondary coolant system are continuously vented to the atmosphere by the f condenser air ejector, thus, in the event of a loss of load incident there 1 are only small quantities of these gases which would be released. j 1131'is the significant isotope because of its low MPC.in air and because the other iodine isotopes have shorter half-lives, and therefore, cannot build up to significant concentrations in the secondary coolant, given the 4 limitations on primary system leak rate and technical specification limiting j activity. Onc tenth of the contained iodine is assumed to reach the site l bpyndary, making allowance for plateout and retention in water droplets. , 1831 is assumed t contribute 70 percent of the total thyroid dose based ! on the ratio of I 31 to the total iodine isotopes given in Table 11-3 of l the FSAR. The maximum inhalation dose at the site boundary is then as follows: Dose (rem) - Ci V B DCF-(0.1) X/Q 164-~ Ci- Secondary coolant activity (0.286 pc/ml 1131 ; equivalent) V- Secondary water volume relgasgd to atmosphere (102 m3) B- Breathing rate (3.47 x 10-* m3/sec) X/Q - Ground level release dispersion factor (8.51 x 10-4 , sec/m3) Proposed Amendment No. 164 3-47
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards
- 4. SURVEILLANCE STANDARDS 164~ 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 Operation unless otherwise stated in an individual Surveillance Requirement.
4.0.2 Failt:re to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 1.9, shall constitute. noncompliance with the OPERABILITY requirements for a ; Limiting Condition for Operation. The time limits of the Action ! including time requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The Action including time requirements may be delayed for up to 24 i hours to permit the completion of the surveillance when the allowable outage time limits of the Action including time requirements are less than 24 hours. 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 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 l l Proposed Amendment No. 164 4-0
1644 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 _identi fied) . 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 4.10 4.11 4.12 j 4.13 4.14 _ m,_ -.- # 4.15 4.16 4.17 4.18 4.19 N/A 4.20 N/A 4.21 N/A 4.22 N/A 4.23 4.24 4.25 N/A 4.26 N/A l l 4.27 N/A { 4.28 l 4.29 N/A ,
.j 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 4-0a
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 164-~ ADolicability Applies to items directly related to safety limits and limiting conditions for operation during HEATUP-COOLDOWN thru power operation, inclusive. 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 l Specification 4.1.1 The minimum frequency and type of surveillance required for reactor ) 164- protection system, safety feature actuation system, Process j
- Instrumentation, and Emergency Shutdown instrumentation shall be as !
stated in Table 4.1-1 (see 4.0.1). 4.1.2 Equipment and sampling test shall be performed as detailed in i tables 4.1-2 and 4.1-3. 4 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. fla115 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 l system. Furthermore, such failures are, in many cases, revealed by alarm or l annunciator action. Comparison of output and/or state of independent channels measuring the same variable supplements this type of built-in surveillance. Based on experience in operation of both conventional and nuclear systems, when the unit is in operation, the minimum checking frequency stated is deemed adequate for reactor system instrumentation. l l Proposed Amendment No. 164 4-1
e - > - . 7 RANCHO SECO UNIT 1n
, TECHNICAL-SPECIFICATIONS '
1 s ., t < < Surveillance Standards-
. .' Table 4.1-1, (Continued)!
INSTRUMENT SURVEILLANCE RE0VIREMENTS Channel Description Check Test -Calibrate Remarks
- 42. Reactor Building drain- '
accumulation tank level
.NA" i 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 M' Q
- b. Containment Area Monitors W NA R 164** c. Chlorine Cetector 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 contro1' rods.
Indicator (Primary Detector) M NA R
- 53. EMOV Position Indicator (Backup Detector) M NA R T/C or Acoustic
- 54. EMOV Block Valve Position Indicator M NA R
- 55. Safety Valve Position In-dictator (Primary Detector) M NA R T/C l
- 56. Safety Valve Position In-dictator (Backup Detector)
Acoustic M NA 'R Proposed Amendment No. 164 4-7b l
RANCHO SECO UN2T 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.1-1 (Continued) INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks 164+ 82. Spray Pond Water Temperature D NA R
- 83. Spray Pond Water Level D NA R Emercenev Shutdown Instrumentation
- 84. Wide Range OSTG Level M NA R BS. Wide Range OSTG Pressure Pressurizer Level M NA R
- 86. Wide Range Reactor Coolant Hot Leg Temperature M NA R l 87. Wide Range Reactor Coolant l
Cold Leg Temperature M NA R l
- 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 S - Each Shift M = Monthly P = Prior to each startup if not done previous week
. D = Daily Q = Quarterly R = Once during the refueling interval W = Weekly SY = Semiannual
*Deenergized and disconnected except when control roem habitability is lost.
Proposed Amendment No. 164
- 4-7g i
i
. RANCHO SECO UNIT 1.
TECHNICAL' SPECIFICATIONS ,f Surveillance Standards
' TABLE 4.1-2 MINIMUM EQUIPMENT TEST-FREQUENCY Item Test Freauency
- 1. . Control-rods- Rod drop times of all Each refueling shutdown full length rods
- 2. Control rod Movement of each rod Every'two' weeks.
movement
'3. Pressu'rizer code Setpoint Note 3 safety valves
- 4. Main steam safety Setpoint Note 3 valves
- 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' ! charcoal filter units.
]
- 9. Fire pumps and Functional Monthly I power supplies
- 10. Reactor Building Functional Each refueling 4 isolation trip interval
- 11. Spent fuel Functional Each refueling ;
l cooling system interval prior to l fuel handling l
- 12. Turbine Overspeed Calibration Each refueling Trips interval 1 4
Proposed Amendment No. 164 4-8 L
j b 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, (l) Each refueling Valves gotevisualinspection,-interval
-verify valve not stuck open
'14. Reactor Coolant Functional (jpstof Each refueling interval System High Point each valve >
Vents
- 15. Low Temperature Functional (5) Prior to RCS temperature Overpressure decreasing below 350*F-Protection (EMOV) .
1644- 16. EMOV block valve Functional (6) Quarterly
- 1. Verifying through manual actuation that the valve is fully open with a force of 1 400 lbs. (applied vertically upward). j
- 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 granted 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.
1
- 5. EMOV block valve closed during test. i i 164-~ 6. EMOV closed during test.
l i 3 l Proposed Amendment No. 164 4-8a
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 164- 4.2.3 Leakage Surveillance 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 in HOT SHUTDOWN thru POWER: 0PERATION,~ inclusive,
- b. Monitoring the containment sump (drain accumulator tank) inventory and discharge at least once per day when the reactor is in HOT SHUTDOHN thru POWER OPERATION, inclusive.
- 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 in HOT SHUTD0HN thru POWER OPERATION, inclusive. !
- d. Performance of a Reactor Coolant System water inventory balance at least once per week when the reactor is in HOT SHUTDOWN thru POWER OPERATION, inclusive.
I I i { i 1
- Proposed Amendment No. 164 4-13 l l
i
i ' j RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS' .i Surveillance Standards- i j
-4.4L REACTOR BUILDING 4.4.1 CONTAINMENT LEAKAGE TESTS Anolicability Applies to containment leakage. q i
Obiective To verify that leakage from.the Reactor Building is maintained within allowable limits. Specification 4.4.1.1 Intearated Leakace Rate Tests 4.4.1.1.1 Calculated Peak Pressure Leakage Rate 1644 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 La, where La is 0.10 percent per day of the Reactor Building atmosphere held at that pressure. l
.y J
\
1 I I l 4
- Proposed Amendment No. 164 1 4-15 ;
i
.]
I i RANCHO SECO UNIT l' TECHNICAL SPECIFICATIONS. Surveillance Standards 164-~ 4.4.'1.1.2' ' Conduct of Tests A. The test duration shall.be at least 24' hours unless' experience from at least two prior tests on similar vessels'provides evidence of the adequacy of a shorter ! test duration. ! Test accuracy shall be verified by supplementary means, !
.B.
such as measuring the quantity of air required to return .l
~to the starting point or.by imposing a known leak rate to -{
demonstrate the validity of measurements. j 1 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. 164-~ 4.4.1.1.3 Frequency of Test J 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. j 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. I i 164-~ 4.4.1.1.4 Corrective Action and Retest i If repairs are necessary to meet the criteria'of 4.4.l.1.1, 164- the integrated leak rate test need not be repeated provided J 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.
1 l I Proposed Amendment No. 164 4-16
;e >
l RANCHO.SECO' UNIT 1-TECHNICAL SPECIFICATIONS. Surveillance Standards 164-* 4.4.1.1.5 Report of. Test.Results l E Each integrated leak rate-test will be the subject of.a summary. technical; report which:will include a description of.
' test ~ methods used and assummary.of-local leak. detection ,
tests ~ Sufficient data and analysis shall be included to. !. 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 i significant errors such.as:those. associated with' instrumentation sensitivities 'and ' data scatter.: 1 [ i l l
.i j . i 1
Proposed' Amendment No.jl64
.4-16a ,
i RANCHO SECO UNIT 1 I TECHNICAL SPECIFICATIONS j Surveillance Standards 4.4.1.2 Local Leakage Rate Testi 4.4.1.2.1 Scope of Testing. The local leak rate shall be measured for. each of the following components: ; q (1) Personnel hatch (2) Emergency hatch (3) . Equipment hatch seals . (4) Fuel transfer tube seals 1 (5) Fuel transfer tube shroud bellows . J (6) Reactor. Building normal sump drain line (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 (27) Demineralized water (28) Service air (29) Core Flood Tank fill and nitrogen supply (30) Core Flood Tank drain and sample
- 4 i
(31) Steam Generator drain (32) Auxiliary steam (33) Component cooling water inlet I (34) Component cooling water outlet J (35) CRD cooling water return Exemption is required for testing inboard isolation valves in the. j reverse direction. This is~necessary to preclude draining the Core Flood Tanks. 1 i Proposed Amendment No. 164 4-17
4 1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS
. Surveillance Standards. 1 4.4.1.2.2 Conduct of Tests
~
164- The containment leakage rates shallibe demonstrated at:the-specified test schedule and shall be determined in-conformance -) with the criteria specified in Appendix J of 10 CFR 50 using i 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 atmosphere held at that ,
- pressure for a minimum of 24 hours. 1 l
4.4.1.2.3 Test Frequency'
]
Local leak detection tests shall be performed at a' frequency of at least each refueling interval, except that: q (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 that 52 psig semi-annually. (c) The personnel and emergency hatches' inner and ! outer door 0-ring seals shall be tested within 72 hours 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 1 0.5 psig. (Appendix J of 10CFR50 is exempted)
(1) The leak rate (Lt) established at the reduced 164-- pressure of 9.5 1 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 components and the total must meet the criterion of 4.4.1.2.2. Proposed Amendment No. 164 4-18
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 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 prior 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. 164-+ (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 I 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 the containment and are locked, sealed or otherwise , secured in the closed position. These penetrations - shall be verified closed during each COLD SHUTDOHN , except that such verification need not be performed more !
- often than once per 92 days.
q l l l l l 1 Proposed Amendment No. 164 I 4-18a { I
- ~
RANCHO SECO UNIT 1 , TECHNICAL SPECIFICATIONS ] Surveillance Standards j 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 1 plate separators,
- 2. The pilot cell specific gravity, corrected to 77*F and .
full electrolyte level, is no greater than 0.010 below i 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
- 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 -
than 0.010 below the average value of all the connected cells, l S. The voltage of each connected cell is greater than 2.07 volts, and l
- 6. The total battery terminal voltage is greater than or !
equal to 125 volts DC when on float charge. j c. At least every re-fueling interval by verifying that: )
- 1. The cells, cell plates and battery racks show no visual i 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, 1
- Proposed Amendment No. 164 4-35e J
1 l l RANCHO SECO UNIT 1 i TECHNICAL SPECIFICATIONS ] 1 Surveillance Standards ! 164- 2. The cell - to - cell and terminal connections are clean and are coated with an anti-corrosion material,
- 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 current output necessary to re-charge the battery following an emergency discharge in 8 hours or less, i
- d. At least once per refueling interval, during COLD SHUTDOWN, by l
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 manufacturer's rating when subjected to a performance 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.
- 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 overall voltage of greater than or equal to 125 volts DC and
- 120 volts AC, respectively.
I Proposed Amendment No. 164 4-35f i _________a
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 l 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.
l 6.9.2 Environmental Reoorts 6.9.2.1 Annual Radiological Reports Annual reports covering the activities of the unit, as described below, for the previous calendar year shall be sumitted prior to l 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 i 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 Ooeratina Reoort l 6.9.2.2.1 Routine radiological environmental operating reports covering I the operation of the unit during the previous calendar. year i shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality. Proposed Amendment No. 164 164-~ 6-12a
' ll
- i RANCHO SECO UNIT l'
-TECHNICAL' SPECIFICATIONS Ad'minist'rative' Controls REPORTING REQUIREMENTS (Continued) 6.9.2.2.2 The annual radiological environmental operating reports sh'all.
include summaries, interpretations, and statistical; evaluation of the results of the radiological environmental surveillance activities for the report period,lincludi.ng.a' comparison with preoperational studies, ' operational controls (as appropriate), ' and previous environmental surveillance reports, and:an.- assessment"of.the observed impacts of the plant operation'on the environment. The reports shall ~also include the~results. of the land use censuses. If; harmful effects or' evidence of irreversible damage are detected,by. the monitoring, the report shall-provide'an analysis of the problem and a planned course of action to alleviate the problem. l The annual radiological envirohmentalLoperating 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: ~ E 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. I The reports shall also' include the-following: a summary-description of the radiological environmental monitoring program; including sampling methods for each sample type, size and physical characteristics.of each sample type, sample-preparation 1 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 '. 2 9 . l 6.9.2.3 Semiannual Radioactive' Effluent-Release Report. I Routine radioactive effluent release. reports covering the operation of the unit during the previous six months of operation shall be a 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 'l criticality. i I i i l- Proposed Amendment No. 164-j . .- 164~ 6-12b
, RANCHO SECO UNIT'l' TECHNICAL SPECIFICATI0NSD 1 Administrative Controls.
REPORTING REQUIREMENTS; (Continued) , 6.9.2.3.1 :The radioactive effluent" release reports shall include;a summary of the quantities of radioactive liqbid and gaseous . effluents and solid waste'.'r' leased from the unit-as outlined in'. Regulatory Guide,1.21 >"HeapJring, Evaluating,'and Reporting Radioactivity in~ Solid' Haste;'and Releases of Radioactive Materials"in. Liquid and Gaseous Effluents from Light-Hater-Cooled Nuclear. Power Plants,";with datarsummarized on a. cuarterly basis, following :the format of Appendix .B _thereof. The radioactive effluent release reports shall include the . release!of gaseous effluents during each quarter, as outlined in Regulatory Guide l.21,. with the data summarized on a quarterly basis, following the format:of /.ppendix B thereof. A summary of meteorological' conditions duringlthe 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.theiradiation 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.beLincluded'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 rele'ase reports shall include an assessment of radiation doses from the. radioactive l_iquid and gaseous effluents released from the unit during each calendar' i quarter, as. outline in' Regulatory Guide l'.21. . The releases of. effluents shall be used for determining ~the gaseous pathwayJ doses. The~ assessment of radiation doses shall.be performed in accordance with the Offsite' Dose Calculation Manual-(ODCM).- ' The radioactive effluent' release reports shal.11 include any-
- changes. to the. PROCESS CONTROL PROGRAM .(PCP): or (ODCM): made .
~
during the reporting period,' as provided in. Specifications'
'6.14 and 6.15.
Proposed Amendment No' 164 164 - 6-12 c ' ~ K o m' ,
Lt' ;,L y. ;
. RANCHO.SECO UNIT 1r ..
TECHNICAL SPECIFICATIONS-
' Administra'tive 'Contirols MONTHLY REPORT 6.9.3 Routine reports of operating' statistics, including narrative summary of: operating and shutdown experience',Lof. lifts of the: .
' Primary. System Safety Valves or EMOVs,'of. major. safety; related. > '
maintenance, and tabulations of facility changesktests or-experiments required pursuant to 10'CFR 50.59(b)', shall-be; submitted , 4 on a monthly basis to the Office of Management'Information and' . ' Program Control,'U. S. Nuclear Regulatory' Commission,-Hashington, 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, including corrective actions and measures.to prevent recurrence, shall be reported to the NRC as^ Licensee Event 1 Reports. .. Supplemental' reports - may.be required to fully describe finalL. resolution:of occurrence. In case ofl corrected or supplemental ' reports, a License Event: Report. shall be: completed and reference sha11Lbe made to the original report date, pursuant to-the requirements of 10 CFR.50.73.- ? l
~
6.9.~4.1 'The types of events; listed below shall.be the subject.of. written reports to the Director 'of the Regional Officeiwithin thirty (30) days of occurrence of the-event. .The written :. report sha11' include, . , as a minimum, a completed copy of a-licensee' event report. form, pursuant to 10.CFR 50.73 and the guidance of NUREG-1022.1
- a. (i) The completion of any nuclear plant shutdown: required by the plant's Technical Specifications; or'
~
(ii) Any operation or condition' prohibited by the plant's ! Technical Specifications; or - (iii) Any deviation from the plant's. Technical Specifications. authorized pursuantito 10 CFR 50.54(x).
- b. Any event or con'ditiont' hat resulted in the' condition of the nuclear power. plant, . including .its principal ' safety barriers, being seriously degraded,.or that resulted in the nuclear power; plant being: ..
~
(1) In an unanalyzed condition"that significantly compromised-plant safety; 3 (ii) In a condition.that was outside the design' basis ~ofLthe plant; or (iii) In a condition not covered by:the' plant's' operating?and-emergency procedures. < Proposed Amendment No.=164
- 164-- 6-12d~
l
. j l
J RANCHO SECO UNIT'l I TECHNICAL SPECIFICATIONS Administrative Controls LICENSEE EVENT REPORT - (Conti nued)'
~
- c. Any natura1' phenomenon ~or other' external condition that posed i an actual. threat to the safety.of the nuclear power plant or- j significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power l plant.
- d. Any event or condition that resulted in manual or.automaticj
. actuation of any Engineered Safety Feature ~(ESF), including the Reactor ~ Protection System (RPS). However, actuation of an i
ESF, including the RPS, that resulted from and was part of the ]
-preplanned sequence.during testing or reactor operation need- "
not be reported. ;
- e. Any. event or condition that alone could have prevented the fulfillment of the safety function of structures or systems '{
-that.are needed to:- !
( l
- 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. Mitigate the consequences of an accident.
-l '
- 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 i' 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
- 4. Mitigate the consequences of.an accident.
Proposed Amendment.No. 164 164-- 6-12e
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS. Administrative Controls 1 LICENSEE EVENT REPORT (Continued) i l
- h. 1. Any airborne radioactivity release that exceeded 2 times j the applicable concentrations 'of the limits specified in l Appendix B, Table II of 10 CFR 20 in unrestricted areas, j 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- j entry into the receiving water (i.e., unrestricted area) l
- for all radionuclides except tritium and dissolved noble i gases, when averaged over a time period of one hour.
I
- 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 releases, or radioactive releases.
- j. Failure of the pressurizer EMOVs or Primary System Safety Valves.
Soecial Reports 6.9.5 Special reports shall 6)e 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). , i B. A Reactor Building Structural integrity report shall be submitted within ninety (90) days of completion of each of the , following tests covered by Technical Specification 4.4.2 (the integrated leak rate test is covered in Technical Specification . 4.4.1.1). l
- 1. Annual Inspection
)
- 2. Tendon Stress Surveillance
- 3. End Anchorage Concrete Surveillance
- 4. Liner Plate Surveillance Proposed Amendment No. 164 3 164~ 6-12f )
l .___-m._.. -_____ _m_.._______m.______,_m_ _ q
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls C. Inservice Inspection Program D. Inoperable Accident Monitoring Instrumentation 30 days (3.5.5) E. Status of Inoperable Fire Protection Equipment F. Inoperable Emergency Control Room /TSC Ventilation Room Filter System G. Radioactive Liquid Effluent Dose 30 days (3.17.2) i i H. Noble Gas Limits 30 days (3.18.2) I. Radioiodine and Particulate 30 days (3.18.3) J. Gaseous Radwaste Treatment 30 days (3.19) K. Radiological Monitoring Program 30 days (3.22) o L. Monitoring Point Substitutions 30 days (3.22) H. Deleted N. Fuel Cycle Dose 30 days (3.25) O. Deleted P. Steam Gerierator Tube Inspection 30 days (4.17.5) Proposed Amendment No. 164 164-*- 6-12g
- u}}