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| document type = TECHNICAL SPECIFICATIONS, TECHNICAL SPECIFICATIONS & TEST REPORTS
| document type = TECHNICAL SPECIFICATIONS, TECHNICAL SPECIFICATIONS & TEST REPORTS
| page count = 52
| page count = 52
| project = TAC:49198
| stage = Request
}}
}}


=Text=
=Text=
{{#Wiki_filter:3 s-Attachment 1 Revised Technical Specifications for Fire Protection System Clean Water Supply Revised Pages:       216b' 216c 216d 216k Reference 1)   Letter from J. M. Pilant to D. B. Vassallo dated June 28, 1982, " Fire Protection Rule 10CFR50, Appendix R" As discussed in Sections 1.4 and 7.0 of Reference 1, the District is providing a Clean Water Fire Protection System for CNS which upgrades the existing system that takes suction from the Missouri River. This change is not an NRC requirement but is being performed with direction from the CNS insurance company.
{{#Wiki_filter:s-3 Revised Technical Specifications for Fire Protection System Clean Water Supply Revised Pages:
The electric and diesel fire pumps will be separate and independent in the modified system and the requirements of 10CFR50 Appendix R3 and Branch Technical Position 9.5-1 Appendix A will be met.
216b' 216c 216d 216k Reference 1)
1 8212030162 821124 PDR ADOCK 05000298 P               PDR
Letter from J. M. Pilant to D. B. Vassallo dated June 28, 1982, " Fire Protection Rule 10CFR50, Appendix R" As discussed in Sections 1.4 and 7.0 of Reference 1, the District is providing a Clean Water Fire Protection System for CNS which upgrades the existing system that takes suction from the Missouri River.
This change is not an NRC requirement but is being performed with direction from the CNS insurance company.
The electric and diesel fire pumps will be separate and independent in the modified system and the requirements of and Branch Technical Position 9.5-1 10CFR50 Appendix R3 Appendix A will be met.
1 8212030162 821124 PDR ADOCK 05000298 P
PDR


e LIMITING CONDITIONS FOR OPERATION               ' SURVEILLANCE REQUIREMENTS 3.14 FIRE DETFCTION SYSTEM                       4.14 FIRE DETECTION SYSTEM APPLICABILITY                                   APPLICABILITY Applies to the operational status of the         Applies to the operational status of the Fire Detection System.                           Fire Detection System.
e LIMITING CONDITIONS FOR OPERATION
' SURVEILLANCE REQUIREMENTS 3.14 FIRE DETFCTION SYSTEM 4.14 FIRE DETECTION SYSTEM APPLICABILITY APPLICABILITY Applies to the operational status of the Applies to the operational status of the Fire Detection System.
Fire Detection System.
OBJECTIVE To assure continuous automatic surveillance throughout the Main Plant.
OBJECTIVE To assure continuous automatic surveillance throughout the Main Plant.
SPECIFICATIONS                                     SPECIFICATIONS A. The Fire Detection System instumen-       A. Each detector on Table 3.14 shall be tation for each fire detection zone               demonstrated operabic every 6 months shown in Table 3.14 shall be operable.           by performance of a channel functional test.
SPECIFICATIONS SPECIFICATIONS A.
B. With one or more of the fire detection     B. The NFPA Code 72.D Class P. supervised instrument (s) shown in Table 3.14               circuits supervision associated with inoperabic:                                       the detector alarms of each of the above required fire detection
The Fire Detection System instumen-A.
: 1. L'ithin I hour establish a fire               instruments shall be demonstrated watch patrol to inspect the                   OPERABLE at least once per 6 months.
Each detector on Table 3.14 shall be tation for each fire detection zone demonstrated operabic every 6 months shown in Table 3.14 shall be operable.
by performance of a channel functional test.
B.
With one or more of the fire detection B.
The NFPA Code 72.D Class P. supervised instrument (s) shown in Table 3.14 circuits supervision associated with inoperabic:
the detector alarms of each of the above required fire detection
: 1. L'ithin I hour establish a fire instruments shall be demonstrated watch patrol to inspect the OPERABLE at least once per 6 months.
zone (s) with the inoperable instru-ment (s) at Icast once per hour, and
zone (s) with the inoperable instru-ment (s) at Icast once per hour, and
: 2. Restore the inoperable instrument (s) to OPERABLE status within 14 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.7.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plar.s and schedule for re-storing the instrument (s) to OPERABLE status.
: 2. Restore the inoperable instrument (s) to OPERABLE status within 14 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.7.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plar.s and schedule for re-storing the instrument (s) to OPERABLE status.
3.15 FIRE SUPPRESSION WATER SYSTEM               4.15 FIRE SUPPRESSION WATER SYSTEM APPLICABILITY                                   APPLICABILITY Applies to the availability of water for         Applies to the availability of water fire fighting purposes.                           for fire fightiag purposes.
3.15 FIRE SUPPRESSION WATER SYSTEM 4.15 FIRE SUPPRESSION WATER SYSTEM APPLICABILITY APPLICABILITY Applies to the availability of water for Applies to the availability of water fire fighting purposes.
for fire fightiag purposes.
OBJ ECTIVE To assure a continuous operable water supply for fire fighting systems from 2 fire pumps.
OBJ ECTIVE To assure a continuous operable water supply for fire fighting systems from 2 fire pumps.
                                              -216b-
-216b-


LIMITING CONDITIONS FOR OPERATION                 SURVEILLANCE REQUIREMENTE 3.15 (cont'd)                                   4.15 (cont'd)
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTE 4.15 (cont'd) 3.15 (cont'd)
SPECIFICATIONS                                   SPECIFICATIONS A. The fire suppression water system shall     A. The Fire Suppression Water Supply be OPERABLE with:                               System shall be demonstrated operable:
SPECIFICATIONS SPECIFICATIONS A.
: 1. Two fire pumps, each with a capacity       1. At least once per 31 days by of at least 2000 gpm, with their                 starting each pump on a stag-g                                                          gered start-up basis and operating discharge aligned to the fire it for:
The Fire Suppression Water Supply A.
The fire suppression water system shall be OPERABLE with:
System shall be demonstrated operable:
1.
At least once per 31 days by 1.
Two fire pumps, each with a capacity g
of at least 2000 gpm, with their starting each pump on a stag-discharge aligned to the fire gered start-up basis and operating it for:
suppression header.
suppression header.
: 2. An OPERABLE flow path capable of                 a) A minimum of 15 minutes taking suction from either of two                     f r a diesel engine-driven 500,000 gallon water storage tanks                   fire pump, and or the Missouri River and transferring the water through                  b) A minimum of 7 minutes for distribution piping with OPERABLE                    an electrical motor-driven sectionalizing control or isolation                   fire pump.
a) A minimum of 15 minutes 2.
valves to the yard hydrant valves and the front valve ahead of the            2. At least once per 31 days by verifying that each valve water flow alarm device on each sprinkler, hose standpipe or spray               (manual, power operated or automatic) in the flow path system riser.
An OPERABLE flow path capable of taking suction from either of two f r a diesel engine-driven 500,000 gallon water storage tanks fire pump, and or the Missouri River and b) A minimum of 7 minutes for transferring the water through an electrical motor-driven distribution piping with OPERABLE sectionalizing control or isolation fire pump.
that is not locked, sealed or therwise secured in position, B. If the requirement of 3.15.A cannot be is in its correct position.
valves to the yard hydrant valves 2.
met, restore the inoperable equipnent to OPERABLE status within 7 days or prepare and submit a Special Report to          3. At least once per 12 months by cycling each testable valve in the Commission pursuant to Specifica-the flow path through at least tion 6.7.2 within the next 30 days m'e e mplete cycle of full outlining the plans and procedures to travel, be used to provide for the loss of redundancy in this system.
At least once per 31 days by and the front valve ahead of the water flow alarm device on each verifying that each valve sprinkler, hose standpipe or spray (manual, power operated or automatic) in the flow path system riser.
: 4. At least once per 18 months by C. With the fire suppression system in-performing a system functional operable:                                            test whfch includes simulated automatic actuation of the system throuFh out its operating
that is not locked, sealed or B.
: 1. Establish a backup fire suppression sequence, and-water system wfthin 24 hours, and a) Verifying that each auto-
If the requirement of 3.15.A cannot be therwise secured in position, is in its correct position.
: 2. Submit a Special Report in accordance matic valve in the flow path with Specification 6.7.2; actuates to its correct p siti n n a test signal, a) By telephone wfthin 24 hours, and b) Verifying that each pump b) In writing no later than the first working day following the developes at least 2000 gpm with at least 110 psi, event, outlining the action taken, the cause of the inoperability and the plans and schedule for restor-ing the system to OPERABLE status.
met, restore the inoperable equipnent to OPERABLE status within 7 days or 3.
                                              -216c-
At least once per 12 months by prepare and submit a Special Report to cycling each testable valve in the Commission pursuant to Specifica-the flow path through at least tion 6.7.2 within the next 30 days m'e e mplete cycle of full outlining the plans and procedures to
: travel, be used to provide for the loss of redundancy in this system.
4.
At least once per 18 months by performing a system functional C.
With the fire suppression system in-test whfch includes simulated operable:
automatic actuation of the system throuF out its operating h
1.
Establish a backup fire suppression sequence, and-water system wfthin 24 hours, and a) Verifying that each auto-2.
Submit a Special Report in accordance matic valve in the flow path with Specification 6.7.2; actuates to its correct p siti n n a test signal, a) By telephone wfthin 24 hours, and b) Verifying that each pump b) In writing no later than the developes at least 2000 gpm with first working day following the at least 110 psi, event, outlining the action taken, the cause of the inoperability and the plans and schedule for restor-ing the system to OPERABLE status.
-216c-


O LIMITING CONDITIONS FOR OPERATION                             SURVEILLANCE REQUIREMENTS 4.15 (cont'd) c)   Cycling each valve in the flow path that is not testable during pinnt operation through at least one complete cycle or full travel, and d)   Verifying that each high pres-sure pump starts-(sequentially) to maintain the fire suppres-sion water system pressure
O LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.15 (cont'd) c)
                                                                                        > 65 psig.
Cycling each valve in the flow path that is not testable during pinnt operation through at least one complete cycle or full travel, and d)
: 5. At least once per 3 years by performing a flow test of the.
Verifying that each high pres-sure pump starts-(sequentially) to maintain the fire suppres-sion water system pressure
> 65 psig.
5.
At least once per 3 years by performing a flow test of the.
system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Editien, published by the National Fire Protection Association.
system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Editien, published by the National Fire Protection Association.
: 6. The fire pump diesel engine shall be demonstrated OPERABLE:
6.
a)   At least once per 31 days by verifying;
The fire pump diesel engine shall be demonstrated OPERABLE:
: 1)   The fuel storage tank contains at least 150*
a)
gallons of fuel, and 4         l,p}
At least once per 31 days by verifying; 1)
: 2)   The diesel starts from ambient conditions and operates for at least 15 l                                                                                             minutes.
The fuel storage tank l,p}
b)   At least once per 92 days by
contains at least 150*
;                                                                                      verifying that a sample of i                                                                                       diesel fuel from the fuel storage tank, obtained in
gallons of fuel, and 4 2)
!                                                                                      accordance with ASTM-D270-65, l                                                                                       is within the acceptable j                                                                                       limits specified in Table 1 of ASTM-D975-74 for viscosity water content and sediment.
The diesel starts from ambient conditions and operates for at least 15 l
I                                                                                c)-   At least once per 18 months by:
minutes.
!                                                                                      1)   Subjecting the diesel to an l
b)
inspection in accordance with procedures prepared in con-junction with its manufactur-er's recommendations for the class of service, and
At least once per 92 days by verifying that a sample of i
                                                                            *This number shall become 250 gallons when the clean water fire protection system becomes operable.
diesel fuel from the fuel storage tank, obtained in accordance with ASTM-D270-65, l
is within the acceptable j
limits specified in Table 1 of ASTM-D975-74 for viscosity water content and sediment.
c)-
At least once per 18 months I
by:
1)
Subjecting the diesel to an inspection in accordance with l
procedures prepared in con-junction with its manufactur-er's recommendations for the class of service, and
*This number shall become 250 gallons when the clean water fire protection system becomes operable.
l
l
                                                                        -216d-
-216d-


INSTRUMENT LOCATION                                                                       INSTRUMENT ID NO.
INSTRUMENT LOCATION INSTRUMENT ID NO.
                                                                  -2           Control Room                                                                         FP-SD-17-1 4
-2 Control Room FP-SD-17-1 FP-SD-17-2 4
FP-SD-17-2 FP-SD-17-3 l
FP-SD-17-3 l
3     Cable Spreading Room.                                                               FP-SD-16-1 FP-SD-16-2 FP-SD-16-3 FP-SD-16-4 FP-SD-16-5' FP-SD-16-6 Cable Expansion Room                                                                 FP-SD-16-7 FP-SD-16-8 4       Switchgear. Rooms DC Switchgear Rooms                                                                 FP-SD-15-2 FP-SD-15-3 Critical Switchgear Room                                                             FP-SD-22-1 FP-SD-22-2 5       Station Battery Rooms                                                               FP-SD-15-1
3 Cable Spreading Room.
:                                                                                                                                                                FP-SD-15-4 FP-SD-15-1A FP-SD-15-4A Diesel Generator Rooms
FP-SD-16-1 FP-SD-16-2 FP-SD-16-3 FP-SD-16-4 FP-SD-16-5' FP-SD-16-6 Cable Expansion Room FP-SD-16-7 FP-SD-16-8 4
                                                                                                ~
Switchgear. Rooms DC Switchgear Rooms FP-SD-15-2 FP-SD-15-3 Critical Switchgear Room FP-SD-22-1 FP-SD-22-2 5
6                                                                                            FP-SD-10-1 i
Station Battery Rooms FP-SD-15-1 FP-SD-15-4 FP-SD-15-1A FP-SD-15-4A 6
FP-SD-10-2
Diesel Generator Rooms FP-SD-10-1
.                                                                                                                                                                  FP-SD-10-3 FP-SD-10-4 CO2-SD-DG-1A CO2-SD-DG-1B CO2-SD-DG-lc CO2-SD-DC-lD CO2-SD-DG-2A CO2-SD-DG-2B CO2-SD-DG-2C CO2-SD-DG-2D 7         Diesel Fuel Storage Rooms                                                           CO2-TD-DG-1A
~
!                                                                                                                                                                  CO2-TD-DG-1B 8         Safety Related Equipment not in Reactor Building
i FP-SD-10-2 FP-SD-10-3 FP-SD-10-4 CO2-SD-DG-1A CO2-SD-DG-1B CO2-SD-DG-lc CO2-SD-DC-lD CO2-SD-DG-2A CO2-SD-DG-2B CO2-SD-DG-2C CO2-SD-DG-2D 7
: i.                                                                           RHR Service Water Booster Pumps                                                     FP-SD-14-3 FP-SD-14-1 Emergency Condensate Storage Tanks Service Water Pumps                                                                 FP-FD-32-1                       l FP-FD-32-2 9         Auxiliary Relay Room & Reactor Protection System Rooms Auxiliary Relay Room                                                                 FP-SD-15-9
Diesel Fuel Storage Rooms CO2-TD-DG-1A CO2-TD-DG-1B 8
                                                                            -Reactor Protection System Room 1A                                                   FP-SD-15-7 Reactor Protection System Room IB                                                   FP-SD-15-8
Safety Related Equipment not in Reactor Building i.
                                                                                                                                  -216k-
RHR Service Water Booster Pumps FP-SD-14-3 Emergency Condensate Storage Tanks FP-SD-14-1 Service Water Pumps FP-FD-32-1 l
FP-FD-32-2 9
Auxiliary Relay Room & Reactor Protection System Rooms Auxiliary Relay Room FP-SD-15-9
-Reactor Protection System Room 1A FP-SD-15-7 Reactor Protection System Room IB FP-SD-15-8
-216k-


i g
i g
Attachment 2 Revised Technical Specifications for Scram Discharge Volume Modifications Revised Pages:       28 33 40 61 The original 12" Scram Discharge Instrument Volume (SDIV) was set to initiate a scram at a level in the volume corresponding to
Revised Technical Specifications for Scram Discharge Volume Modifications Revised Pages:
                    < 3G gallons. The 36 gallons was based on an availability consideration giving the operator 20 minutes to respond to an inadvertently closed drain valve assuming each control rod leaked 50 cc/ minute. There are now two instrument volumes of approximately 22 gallons each, one for each group of hydraulic control units in the reactor building.           Each group has approximately one-half of the hydraulic control units.
28 33 40 61 The original 12" Scram Discharge Instrument Volume (SDIV) was set to initiate a scram at a level in the volume corresponding to
The new instrument volumes initiate alarms, rod blocks, and scrams at specified levels rather than volumes.               A level transmitter or level switch measures level rather than volume.
< 3G gallons.
The surveillance program to provide functional checks of the SDIV level instrumentation is provided.           Station procedures provide for periodic verification of the correlation between level and volume. An SDV not drained alarm has been established at
The 36 gallons was based on an availability consideration giving the operator 20 minutes to respond to an inadvertently closed drain valve assuming each control rod leaked 50 cc/ minute.
                    < 114 inches. The references for all levels are the center lines of the lower instrument tap on each SDIV. The scram level for each instrument volume assures an adequate scram discharge volume exists so that all control rods can insert fully.         It should be noted that CNS now has larger scram discharge volumes (excluding the instrument volumes) than existed before this modification.
There are now two instrument volumes of approximately 22 gallons each, one for each group of hydraulic control units in the reactor building.
Each group has approximately one-half of the hydraulic control units.
The new instrument volumes initiate alarms, rod blocks, and scrams at specified levels rather than volumes.
A level transmitter or level switch measures level rather than volume.
The surveillance program to provide functional checks of the SDIV level instrumentation is provided.
Station procedures provide for periodic verification of the correlation between level and volume.
An SDV not drained alarm has been established at
< 114 inches.
The references for all levels are the center lines of the lower instrument tap on each SDIV.
The scram level for each instrument volume assures an adequate scram discharge volume exists so that all control rods can insert fully.
It should be noted that CNS now has larger scram discharge volumes (excluding the instrument volumes) than existed before this modification.
1 I
1 I
f
f
        .  ._. .-          - . ~ .        . _ - _ .- _  .
-. ~.


COOPER EUCLFAR STATION TABLE 3.1.1-                                                     ,
COOPER EUCLFAR STATION TABLE 3.1.1-REACTOR PROTECTION SYSTEM INSTRUMENTATION REQUIREMENTS Minimum Number Action Requir.
REACTOR PROTECTION SYSTEM INSTRUMENTATION REQUIREMENTS Minimum Number   Action Requir.
Applicability Conditions of Operable When Equipmen Reactor Protection Mode Switch Position Trip Level Channels Per Operability i.
Applicability Conditions                                   of Operable     When Equipmen Reactor Protection                     Mode Switch Position                 Trip Level           Channels Per     Operability i.
System Trip Function Shutdown Startup Refuel Run Setting Trip Systems (1) Not Assured (
System Trip Function                 Shutdown   Startup   Refuel   Run           Setting             Trip Systems (1) Not Assured (
Mode Switch in Shutdown X(7)
Mode Switch in Shutdown                 X(7)         X         X             X                               1                 A Manual Scram                           X(7)         X         X             X                               1-               A IRM (17)                               X(7)         X         X       (5)         1 120/125 of in-         3                 A High Flux                                                                       dicated scale Inoperative                                 X         X       (5)                                   3                 A APRM (17)
X X
High Flux (Flow biased)                                               X     1 (0.66W+54%)   FRP     2               A or C (14)       MFLPD High Flux                       X(7)       X(9)     X(9)     (16)         1 15% Rated Power                         A or C Inoperative                               X(9)     X(9)             X           (13)               2               A or C Downscale                                   (11)                       X(12) > 2.5% of indi-           2               ~ A or C cated scale High Reactor Pressure                             X(9)     X(10)           X     1 1045 psig               2                 A NBI-PS-55 A,B,C, & D High Drywell Pressure                             X(9)(8) X(8)               X     1 2 psig                 2               A or D PC-PS-12 A,B,C, & D Reactor Low Water Level                             X         X             X     > + 12.5 in. indi-       2               A or D NBI-LIS-101 A,B,C, & D                                                         cated level Scram Discharge Volume                               X       X(2)             X     1 92 inches               3                 A High Water Level CRD-LS-231 A & B CRD-LS-234 A & B CRD-LT-231 C & D CRD-LT-234 C & D
X 1
                                                            ~'           - _ -_
A Manual Scram X(7)
X X
X 1-A IRM (17)
X(7)
X X
(5) 1 120/125 of in-3 A
High Flux dicated scale Inoperative X
X (5) 3 A
APRM (17)
High Flux (Flow biased)
X 1 (0.66W+54%)
FRP 2
A or C (14)
MFLPD High Flux X(7)
X(9)
X(9)
(16) 1 15% Rated Power A or C Inoperative X(9)
X(9)
X (13) 2 A or C Downscale (11)
X(12)
> 2.5% of indi-2
~ A or C cated scale High Reactor Pressure X(9)
X(10)
X 1 1045 psig 2
A NBI-PS-55 A,B,C, & D High Drywell Pressure X(9)(8) X(8)
X 1 2 psig 2
A or D PC-PS-12 A,B,C, & D Reactor Low Water Level X
X X
> + 12.5 in. indi-2 A or D NBI-LIS-101 A,B,C, & D cated level Scram Discharge Volume X
X(2)
X 1 92 inches 3
A High Water Level CRD-LS-231 A & B CRD-LS-234 A & B CRD-LT-231 C & D CRD-LT-234 C & D
~'


COOPER NUCLFAR STATION TABLC 4.1.1 (Page 2)
COOPER NUCLFAR STATION TABLC 4.1.1 (Page 2)
REACTOR PROTECTION SYSTEM (SCRAM INSTRUMENTATION) FUNCTIONAL TESTS.
REACTOR PROTECTION SYSTEM (SCRAM INSTRUMENTATION) FUNCTIONAL TESTS.
MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTR. AND CONTROL CIRCUITS Instrument Channel                 Group (2)         Functional Test             Minimum Frequency (3)
MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTR. AND CONTROL CIRCUITS Instrument Channel Group (2)
High Water Level in Scram Discharge             A         Trip Channel and Alarm             Once/3 Months       '
Functional Test Minimum Frequency (3)
Volume CRD-LS-231 A & B CRD-LS-234 A & B CRD-LT-231 C & D CRD-LT-234 C & D Main Steam Line liigh Radiation                 B         Trip Channel and Alarm (4)         Once/ Week RMP-RM-251 A,B,C, & D
High Water Level in Scram Discharge A
, Main Steam Line Isolation Valve                 A         Trip Channel and Alarm             Once/ Month (1) g Closure MS-LMS-86 A,B,C, & D i   MS-LMS-80 A,B,C, & D Turbine Control Valve Fast Closure               A         Trip Channel and Alarm             Once/ Month (1)
Trip Channel and Alarm Once/3 Months Volume CRD-LS-231 A & B CRD-LS-234 A & B CRD-LT-231 C & D CRD-LT-234 C & D Main Steam Line liigh Radiation B
IGF-63/0PC -1,2,3,4 Turbine First Stage Pressure                     A         Trip Channel and Alarm             Once/3 Months Permissive MS-PS-14 A,B,C, & D Turbine Stop Valve Closure                       A         Trip Channel and Alarm             Once/ Month (1)
Trip Channel and Alarm (4)
Once/ Week RMP-RM-251 A,B,C, & D Main Steam Line Isolation Valve A
Trip Channel and Alarm Once/ Month (1)
,g Closure MS-LMS-86 A,B,C, & D i
MS-LMS-80 A,B,C, & D Turbine Control Valve Fast Closure A
Trip Channel and Alarm Once/ Month (1)
IGF-63/0PC -1,2,3,4 Turbine First Stage Pressure A
Trip Channel and Alarm Once/3 Months Permissive MS-PS-14 A,B,C, & D Turbine Stop Valve Closure A
Trip Channel and Alarm Once/ Month (1)
SVOS-1 (1), SVOS-1 (2)
SVOS-1 (1), SVOS-1 (2)
SVOS-2 (1), SVOS-2 (2)
SVOS-2 (1), SVOS-2 (2)
Reactor Pressure Permissive                     A         Trip Channel and Alarm             Once/3 Months NBI-PS-51 A B.C & D
Reactor Pressure Permissive A
Trip Channel and Alarm Once/3 Months NBI-PS-51 A B.C & D


LIMITING CONDITIONS FOR OPERATION                 SURVEILLANCE REQUIREMENTS 3.1   BASES (cont'd.)                             4.1 BASES (cont'd.)
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.1 BASES (cont'd.)
against short reactor periods in                 revealed only on test. Therefore, these ranges.                                    it is necessary to test them periodi-cally.
3.1 BASES (cont'd.)
The control rod drive scram system is designed so that all of the water             A study was conducted of the instru-which is discharged from the reactor               mentation channels included in the by a scram can he accommodated in the             Group (B) devices to calculate their discharge piping. The scram discharge              " unsafe" failure rates. The analog volume accommodates in excess of 36               devices (sensors and amp 11ifers) gallons of water and is the low point             are predicted to have an unsafe
against short reactor periods in revealed only on test. Therefore, it is necessary to test them periodi-these ranges.
                                                                                                -6 in the piping. No credit was taken                fc1]ure rate of less than 20 X 10 for this volume in the design                      failures / hour. The bi-stable trip of the discharge piping as concerns               circuits are predicted to have an the amount of water which must be                 unsafe 2 X 10
cally.
_failurerateof3cssthan failures / hour. Consider-accommodated during a scram.
The control rod drive scram system is designed so that all of the water A study was conducted of the instru-mentation channels included in the which is discharged from the reactor by a scram can he accommodated in the Group (B) devices to calculate their
" unsafe" failure rates. The analog discharge piping. The scram discharge volume accommodates in excess of 36 devices (sensors and amp 11ifers) gallons of water and is the low point are predicted to have an unsafe
-6 fc1]ure rate of less than 20 X 10 in the piping.
No credit was taken failures / hour. The bi-stable trip for this volume in the design of the discharge piping as concerns circuits are predicted to have an the amount of water which must be unsafe _failurerateof3cssthan 2 X 10 failures / hour. Consider-accommodated during a scram.
ing the two hour monitoring interva]
ing the two hour monitoring interva]
During normal operation the dis-                   f r the analog devices as assumed charge volume is enpty; however,                   bove, and a weekly test interval should it fill with water, the water               f r the hi-stable trip circuits, discharged to the piping from the                 the design reliability goal of reactor could not be accommodated which            0.99999 is attained with ample margin.
During normal operation the dis-f r the analog devices as assumed charge volume is enpty; however, bove, and a weekly test interval should it fill with water, the water f r the hi-stable trip circuits, discharged to the piping from the the design reliability goal of 0.99999 is attained with ample margin.
would result in slow scram times or partial control rod insertion. To pre-             The bi-stabic devices are monitored clude this occurrence, diverse indi-               during plant operation to record their failure history and establish a test cation (two level switches and two level transmitters for each discharge             interval using the curve of Figure volume) has been provided in the                    4.1.1. There are numerous identical bi-stable devices used throughout instrument volumes which alarm and                                                        ,
reactor could not be accommodated which would result in slow scram times or The bi-stabic devices are monitored partial control rod insertion. To pre-clude this occurrence, diverse indi-during plant operation to record their cation (two level switches and two failure history and establish a test level transmitters for each discharge interval using the curve of Figure 4.1.1.
scram the reactor when the volume of               the plant 's instrunentation systen.
There are numerous identical volume) has been provided in the instrument volumes which alarm and bi-stable devices used throughout scram the reactor when the volume of the plant 's instrunentation systen.
Therefore, significant dota on the water reaches 92 inches. As indicated above, there is sufficient volume in                failure rates for the bi-stable devices the piping to acco nmodate the scram               should be accumulated rapidly.
water reaches 92 inches. As indicated Therefore, significant dota on the failure rates for the bi-stable devices above, there is sufficient volume in the piping to acco nmodate the scram should be accumulated rapidly.
without impairment of the scram times The frequency of calibration of the or amount of insertion of the control rodo. This function shuts the reactor APPP Flow Biasing Network has been established as each refueling out-down while sufficient volume remains               age. The flow bicsing network is to accommodate the discharged water functionally tested at least once and precludes the situation in which Per month and, in addition, cross a scram would be required but not be calibration checks of the flow abic to perform its function adequately, input to the flow biasing network can be made during the functional A source range monitor (SRM) system is also provided to supply additional                 test by direct meter reading. There neutron level information during start-           are several instruments which must up but has no scram functions (refer-             be calibrated and it will take sev-ence paragraph VII.S.4 FSAR). Thus,               eral days to perform the calibration the IRM and APRM are required in the                 f the entire network. While the calibra+1on is being performed, a
without impairment of the scram times or amount of insertion of the control The frequency of calibration of the rodo. This function shuts the reactor APPP Flow Biasing Network has been established as each refueling out-down while sufficient volume remains age. The flow bicsing network is to accommodate the discharged water functionally tested at least once and precludes the situation in which Per month and, in addition, cross a scram would be required but not be calibration checks of the flow abic to perform its function adequately, input to the flow biasing network can be made during the functional A source range monitor (SRM) system is also provided to supply additional test by direct meter reading. There are several instruments which must neutron level information during start-be calibrated and it will take sev-up but has no scram functions (refer-eral days to perform the calibration ence paragraph VII.S.4 FSAR). Thus, f the entire network. While the the IRM and APRM are required in the
      " Refuel" and " Start / Hot Standby" modes.
" Refuel" and " Start / Hot Standby" modes.
In the power range the APRM system provides required protection (refer-
calibra+1on is being performed, a In the power range the APRM system provides required protection (refer-m._.~ _ _. _. _
    ._    _        . m ._ .~ _ _ . _ . _ .
TABLE 3.2.C CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION Minimum Number of Function Trip Level Setting Operable Instrument Channels / Trip System (5)
TABLE 3.2.C CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION                                           .
APRM Upscale (Flow Bias) ji (0.66W + 427) FRP (2) 2(1)
Minimum Number of Function                                           Trip Level Setting                         Operable Instrument Channels / Trip System (5)
AFRP Upscale (Stertup)
APRM Upscale (Flow Bias)                           ji (0.66W + 427) FRP     (2)                                 2(1)
< 12%
AFRP Upscale (Stertup)                             < 12%           _ MFLPD                                     2(1)
_ MFLPD 2(1)
APRM Downscale (9)                                 3,2.5%                                                       2(1) i     APRM Inoperative                                   (10b)                                                       2(1)
APRM Downscale (9) 3,2.5%
FEM Upscale (Flow Bias)                           j[ (0.66W + 40%) (2)                                         1 RBM Downscale (9)                                 2; 2.5%                                                     1
2(1) i APRM Inoperative (10b) 2(1)
,!    RBM Inoperative                                     (10c)                                                       1 i
FEM Upscale (Flow Bias) j[ (0.66W + 40%) (2) 1 RBM Downscale (9) 2; 2.5%
IRM Upscale (8)                                   ji 108/125 of Full Scale                                     3(1)
1 RBM Inoperative (10c) 1 i
IEK Downscale (3)(8)                               2;2.5%                                                       3(1)
IRM Upscale (8) ji 108/125 of Full Scale 3(1)
IPM Detector Not Full In (8)                                                                                   3(1).
IEK Downscale (3)(8) 2;2.5%
1' IRM Inoperative (8)                                 (10a)                                                       3(1) 5
3(1)
{     SRM Upscale (8)                                   ;E 1 x 10 counts /Second                                     1(1)(6)
IPM Detector Not Full In (8) 3(1).
}     SRP Detector Not Full In (4)(8)                     (3 100 cps)                                                 1(1)(6)
1 IRM Inoperative (8)
SRM Inoperative (8)                                 (10a)                                                       1(1)(6)
(10a) 3(1) 5
Flow Bias Comparator                               j[10% Difference In Recirc. Flows                           1 Flow Bias Upscale /Inop.                           ;l 110% Recirc. Flow                                         1 SRM Downscale (8)(7)                               3,3 Counts /Second (11)                                     1(1)(6)
{
SDV Water Level High                               ;[46 inches                                                 1(12)_
SRM Upscale (8)
;E 1 x 10 counts /Second 1(1)(6) 7
}
SRP Detector Not Full In (4)(8)
(3 100 cps) 1(1)(6)
SRM Inoperative (8)
(10a) 1(1)(6)
Flow Bias Comparator j[10% Difference In Recirc. Flows 1
Flow Bias Upscale /Inop.
;l 110% Recirc. Flow 1
SRM Downscale (8)(7) 3,3 Counts /Second (11) 1(1)(6)
SDV Water Level High
;[46 inches 1(12)_
CRD-231E, 234E
CRD-231E, 234E


Attachment 3 Revised Technical Specification for HPS Power Monitoring System Revised Pages:       193 195 197 199 Reference 1)   Letter from D. B. Vassallo to J. M. Pilant dated May 4, 1982, " Reactor Protection System (RPS)
Revised Technical Specification for HPS Power Monitoring System Revised Pages:
Power Monitoring System Design Modification"
193 195 197 199 Reference 1)
: 2)   Letter from D. B. Vassallo to J. M. P11 ant dated July 8,1982, same subject During the Spring 1982 refueling outage, eight Class IE Electrical Protection Assemblics (EPA's) were installed in the RPS power monitoring system. This change was in response to the concern that the original RPS was not seismically qualified and could degrade during a seismic event. The proposed Technical Specifications are in accordance with General Electric verified time delays as required in Reference 1 and the model Technical Specification of Reference 2. Please note that exception is taken to the Model Technical Specification surveillance requirement of a
Letter from D. B. Vassallo to J. M. Pilant dated May 4, 1982, " Reactor Protection System (RPS)
    " channel functional" test every six months. This would require deenergizing each half of the RPS system either during the test       -
Power Monitoring System Design Modification" 2)
or transfer to the alternate supply.       This puts unwanted transients on critical equipment (especially the Main Steam Line Radiation Monitors) and induces an unnecessary risk of a plant scram. The 18-month test frequency proposed for the functional test and channel calibration is consistent with other Technical Specifications for electrical breakers in essential systems.
Letter from D. B. Vassallo to J. M. P11 ant dated July 8,1982, same subject During the Spring 1982 refueling outage, eight Class IE Electrical Protection Assemblics (EPA's) were installed in the RPS power monitoring system.
This change was in response to the concern that the original RPS was not seismically qualified and could degrade during a seismic event.
The proposed Technical Specifications are in accordance with General Electric verified time delays as required in Reference 1 and the model Technical Specification of Reference 2.
Please note that exception is taken to the Model Technical Specification surveillance requirement of a
" channel functional" test every six months.
This would require deenergizing each half of the RPS system either during the test or transfer to the alternate supply.
This puts unwanted transients on critical equipment (especially the Main Steam Line Radiation Monitors) and induces an unnecessary risk of a plant scram.
The 18-month test frequency proposed for the functional test and channel calibration is consistent with other Technical Specifications for electrical breakers in essential systems.


LIMITING CONDITIONS FOR OPERATION                 SURVETLLANCE REQUIREMENTS 4.9 AUXILIARY ELECTl:1 CAL SYSTEM 3.9 AUXILIARY ELECTRICA1. SYS'lEM Applicability:
LIMITING CONDITIONS FOR OPERATION SURVETLLANCE REQUIREMENTS 4.9 AUXILIARY ELECTl:1 CAL SYSTEM 3.9 AUXILIARY ELECTRICA1. SYS'lEM Applicability:
Applicability:
Applicability:
Applies to the periodic testing Applies to the auxiliary electrical requirements of the auxiliary power system.
Applies to the periodic testing Applies to the auxiliary electrical requirements of the auxiliary power system.
Line 155: Line 274:
Specification:
Specification:
Specification:
Specification:
A.           ary & Meal huhment A. Auxiliary Electrical Equipment
A.
: 1. mergency Buses Undervoltage The reactor shall not be made criti-Relays cal-from a Cold Shutdown Condition unless all of the following condi-
ary & Meal huhment A.
: a. Loss of voltage relays tions are satisfied:
Auxiliary Electrical Equipment 1.
Once every 18 months, loss
mergency Buses Undervoltage The reactor shall not be made criti-Relays cal-from a Cold Shutdown Condition unless all of the following condi-a.
: 1. Both off-site sources (345 KV and                   f vo tage on emergency 69 KV) and the startup transformer and emergency transformer are avail-               buses is simulated to able and capable of automatically demonstrate the load shed-ding fr m emergency buses supplying power to the 4160 Volt                   and the automatic start emergency buses IF and IG.                         of diesel generators.
Loss of voltage relays tions are satisfied:
: 2. Both diesel generators shall be Undervoltage relays operable and there shall be a mini-b.
Once every 18 months, loss 1.
mum of 45,000 gal. of diesel fuel in Once every 18 months, low the fuel oil storage tanks.
Both off-site sources (345 KV and f vo tage on emergency 69 KV) and the startup transformer buses is simulated to and emergency transformer are avail-demonstrate the load shed-able and capable of automatically ding fr m emergency buses supplying power to the 4160 Volt and the automatic start emergency buses IF and IG.
voltage on emergency buses is simulated to demonstrate
of diesel generators.
: 3. The 4160V critical buses IF and IG                 disconnection of the emer-and the 480V critical buses IF and IG are tnergized, gency buses from the offsite power source. The under-v Itage relays shall be
2.
: a. The loss of voltage relays and calibrated once every 18 their auxiliary relays are                   m nds, operable.
Both diesel generators shall be b.
The undervoltage relays and      2. Diesel Generators b.
Undervoltage relays operable and there shall be a mini-mum of 45,000 gal. of diesel fuel in Once every 18 months, low the fuel oil storage tanks.
their auxiliary relays are
voltage on emergency buses is simulated to demonstrate 3.
: a. Each diesel-generator shall be started operable,                                   manually and loaded to not less than 4      The four unit 125V/250V batteries and              35% of rated load for no less than 2 their chargers shall be operable.
The 4160V critical buses IF and IG disconnection of the emer-and the 480V critical buses IF and IG gency buses from the offsite are tnergized, power source. The under-v Itage relays shall be a.
hours once each month to demonstrate operational readiness.
The loss of voltage relays and calibrated once every 18 their auxiliary relays are m nds, operable.
: 5.   .The power monitoring system for the inservice RPS MG set or alternate source shall be operable.
2.
                                                  -193-j.
Diesel Generators b.
The undervoltage relays and their auxiliary relays are Each diesel-generator shall be started a.
: operable, manually and loaded to not less than 35% of rated load for no less than 2 4
The four unit 125V/250V batteries and hours once each month to demonstrate their chargers shall be operable.
operational readiness.
5.
.The power monitoring system for the inservice RPS MG set or alternate source shall be operable.
-193-j.


  ~
~
LIMITING CONDITIONS FOR OPERATION               SURVEILLANCE REOUIREMENTF i 3.9.A                                           4.9.A (cont'd.)
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTF i
cell and overall battery voltage shall be measured and logged,
3.9.A 4.9.A (cont'd.)
: b. Every three conths the measurements shall be mede of the voltage of each cell to nearest 0.1 Volt, specific gravity of each cell, and tenperature of every sixth cell. These measure-ments shall be logged.
cell and overall battery voltage shall be measured and logged, b.
: c. Once each operating cycle, the stated batteries shall be subjected to a rated load discharge test. The specific gravity and voltage of each cell shall be determined after the discharge and logged.
Every three conths the measurements shall be mede of the voltage of each cell to nearest 0.1 Volt, specific gravity of each cell, and tenperature of every sixth cell. These measure-ments shall be logged.
B. Operation with Inoperable Equipment         4. Power Monitoring System for RPS System
c.
,    Whenever the reactor is in Run Mode or           The above specified RPS power monitor-Startup Mode with the reactor not in a           ing system instrunentation shall be Cold Condition, the availability of               deternined operable:
Once each operating cycle, the stated batteries shall be subjected to a rated load discharge test.
electric power shall be as specified in 3.9.A, except as specified in 3.9.B.1.       a. At least once per operating cycle by demonstratirg the operability
The specific gravity and voltage of each cell shall be determined after the discharge and logged.
: 1. From and after the date incoming power                 of over-voltage, under-voltage is not availabic from a startup or emer-             and under-frequency protective gency transformer, continued reactor                   instrumentation by performance of   ,
B. Operation with Inoperable Equipment 4.
operation is permissible under this                   a channel calibration including condition for seven days. At the end                   simulated autematic actuation of of this period, provided the second                     the protective relays, tripping source of incoming power has not been                 logic and output circuit breakers made immediately available, the NRC                   and verifying the following set-must be notified of the event and the                 points.
Power Monitoring System for RPS System Whenever the reactor is in Run Mode or The above specified RPS power monitor-Startup Mode with the reactor not in a ing system instrunentation shall be Cold Condition, the availability of deternined operable:
electric power shall be as specified in 3.9.A, except as specified in 3.9.B.1.
a.
At least once per operating cycle by demonstratirg the operability
: 1. From and after the date incoming power of over-voltage, under-voltage is not availabic from a startup or emer-and under-frequency protective gency transformer, continued reactor instrumentation by performance of operation is permissible under this a channel calibration including condition for seven days. At the end simulated autematic actuation of of this period, provided the second the protective relays, tripping source of incoming power has not been logic and output circuit breakers made immediately available, the NRC and verifying the following set-must be notified of the event and the points.
plan to restore this second source.
plan to restore this second source.
During this period, the two diesel gencr-               1. Over-voltage < 132 VAC, with ators and associated critical buses must                     tine delay < 2 sec.
During this period, the two diesel gencr-1.
Over-voltage < 132 VAC, with ators and associated critical buses must tine delay < 2 sec.
be deaanstrated to be operable.
be deaanstrated to be operable.
: 2. Under-voltage > 108 VAC with
2.
: 2. From and after the date that incoming                       time delay < 2 sec.
Under-voltage > 108 VAC with
power is not available from both start-up and emergency transformers, continued                 3. Undcr-frequency > 57 Hz. with operation is permissible, provided the                       time delay < 2 sec.
: 2. From and after the date that incoming time delay < 2 sec.
power is not available from both start-up and emergency transformers, continued 3.
Undcr-frequency > 57 Hz. with operation is permissible, provided the time delay < 2 sec.
two diesel generators and associated critical buses are demonstrated to be
two diesel generators and associated critical buses are demonstrated to be
                                            -195-
-195-


LIMITING CONDITIONS FOR OPERATION                       SURVEILLANCE REQUIREMENTS l3.9.B.5(cont'd.)                                         4.9.B From and after the date that one of the 125 or 250 volt hattery systems is made or found to be inoperable for any reason, continued reactor operation is permissible during the succeeding ten days within electrical safety considera-tions, provided repair work is initiated in the most expeditious manner to return the failed component to an operable state, and Specifications 3.5.A.5 and 3.5.F are satisfied. The NRC shall be notified within 24 hours of the situation, the precautions to be taken during this period and the plans to return the failed components to an operable state.
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS l3.9.B.5(cont'd.)
: 6. With one RPS. electric power monitoring channel for an inservice RPS MG set or alternate power supply inoperable, restore the inoperable channel to operable status within 72 hours or remove the associated RPS MG set or alternate power supply from service.
4.9.B From and after the date that one of the 125 or 250 volt hattery systems is made or found to be inoperable for any reason, continued reactor operation is permissible during the succeeding ten days within electrical safety considera-tions, provided repair work is initiated in the most expeditious manner to return the failed component to an operable state, and Specifications 3.5.A.5 and 3.5.F are satisfied. The NRC shall be notified within 24 hours of the situation, the precautions to be taken during this period and the plans to return the failed components to an operable state.
: 7. With both RPS electric power monitoring channels for an inservice RPS MG set or alternate power supply inoperable, restore at least one to operable status within 30 minutes or remove the associated RPS MG set or alternate power supply from service.
6.
                                                  -197-
With one RPS. electric power monitoring channel for an inservice RPS MG set or alternate power supply inoperable, restore the inoperable channel to operable status within 72 hours or remove the associated RPS MG set or alternate power supply from service.
7.
With both RPS electric power monitoring channels for an inservice RPS MG set or alternate power supply inoperable, restore at least one to operable status within 30 minutes or remove the associated RPS MG set or alternate power supply from service.
-197-


4.9 BASES The monthly test of the diesel generator is_ conducted to check for equipment failures and deterioration. Testing is conducted up to equilibrium operating             'i conditions to demonstrate proper. operation at these conditions. The diesel generator will be manually started, synchronf red and connected to the bus i
4.9 BASES The monthly test of the diesel generator is_ conducted to check for equipment failures and deterioration. Testing is conducted up to equilibrium operating
and load picked up. The diesel generator should be' loaded to at least 35T of rated load to prevent fouling of the engine.         It le expected that the-diesel generator will be run for at least two hours. Diesel generator experience                   ,
'i conditions to demonstrate proper. operation at these conditions. The diesel generator will be manually started, synchronf red and connected to the bus i
d
and load picked up.
          'at other generating. stations indicates that the testing frequency.is adequate
The diesel generator should be' loaded to at least 35T of rated load to prevent fouling of the engine.
!          -and provides a_high. reliability-of operation should the system be required.
It le expected that the-diesel generator will be run for at least two hours. Diesel generator experience d
Each diesel' generator has two air compressors and two air receivers for i           starting. It .f s expected that the air compressors wf]1 run only infrequently.
'at other generating. stations indicates that the testing frequency.is adequate
3          Duringlthe monthly check of the diesel generator, each receiver in each' set-of receivers will be drawn down below;the point at which the corresponding compressor eutomatically starts to check operation and the ability of the compressors to recharge the receivers.
-and provides a_high. reliability-of operation should the system be required.
The diesel generator fuel consumption rate at full load is approximately 275-gallons per hour. Thus, the monthly load test of the diesel generators will test the operation and the ability of the. fuel oil transfer pumps to refill                 l the dey tank and will check the operation of these pumps from the energency                 '
Each diesel' generator has two air compressors and two air receivers for i
i        source.
starting.
i
It.f s expected that the air compressors wf]1 run only infrequently.
!          The rest of the diesel generator during the refueling outage will be core l~
Duringlthe monthly check of the diesel generator, each receiver in each' set-3 of receivers will be drawn down below;the point at which the corresponding compressor eutomatically starts to check operation and the ability of the compressors to recharge the receivers.
comprehensive in that it will functionally test the system; i.e, it will check diesel generator starting and closure of diesel generator breaker and sequencing of load on the diesel generator. The diesel generator will be started by simulatien.of a loss-of-coolant accident. In addition, en i           underveltage condition will be imposed to simulate a loss of of f-site power.
The diesel generator fuel consumption rate at full load is approximately 275-gallons per hour. Thus, the monthly load test of the diesel generators will test the operation and the ability of the. fuel oil transfer pumps to refill l
Periodic tests between refueling outages verify the ability 'of the diesel generator to run at full load and the core and containment cooling pumps to deliver full flow. Periodic testing of the various_ components, plus a func-
the dey tank and will check the operation of these pumps from the energency i
.          tional test once-a-cycle, is sufficient to maintain adequate' reliability.
source.
i The rest of the diesel generator during the refueling outage will be core l
comprehensive in that it will functionally test the system; i.e, it will
~
check diesel generator starting and closure of diesel generator breaker and sequencing of load on the diesel generator. The diesel generator will be started by simulatien.of a loss-of-coolant accident.
In addition, en i
underveltage condition will be imposed to simulate a loss of of f-site power.
Periodic tests between refueling outages verify the ability 'of the diesel generator to run at full load and the core and containment cooling pumps to deliver full flow. Periodic testing of the various_ components, plus a func-tional test once-a-cycle, is sufficient to maintain adequate' reliability.
Although station batteries will deteriorate with time, utility experience indicates there is almost no possibility of precipitous failure. The type of surveillance described in this specification is that which has been demonstrated over the years to provide an indication of a cell becoming irregular or unserviceable long before it becomes a failure. _In addition, the checks described also provide adequate indication that the batteries have the speci-fied ampere-hour capability.'
Although station batteries will deteriorate with time, utility experience indicates there is almost no possibility of precipitous failure. The type of surveillance described in this specification is that which has been demonstrated over the years to provide an indication of a cell becoming irregular or unserviceable long before it becomes a failure. _In addition, the checks described also provide adequate indication that the batteries have the speci-fied ampere-hour capability.'
The diesel fuel oil quality must be checked to ensure proper operation of the diesel generators. Water content should be mininized because water in the fuel could centribute to excessive demage to the diesel engine.
The diesel fuel oil quality must be checked to ensure proper operation of the diesel generators. Water content should be mininized because water in the fuel could centribute to excessive demage to the diesel engine.
When it is determined that some auxiliary electrical equipment is out of service, the increased surveillance required in Section 4.5.F is deemed adequate to provide assurance that the remaining cauipment will be operable.                                                   ,
When it is determined that some auxiliary electrical equipment is out of service, the increased surveillance required in Section 4.5.F is deemed adequate to provide assurance that the remaining cauipment will be operable.
The Reactor Protection Systen (RPS) is equipped with a seismically qualified, Class 1E power monitoring system. This system consists of eight Electrical Protection Assemblies (EPA) which isolate the power sources from the RPS if the input voltage and frequency are not within limits specified for safe system operation.         Isolation of RPS power causes that RPS division'to_ fail safe.
The Reactor Protection Systen (RPS) is equipped with a seismically qualified, Class 1E power monitoring system. This system consists of eight Electrical Protection Assemblies (EPA) which isolate the power sources from the RPS if the input voltage and frequency are not within limits specified for safe system operation.
                                                      -RR9-
Isolation of RPS power causes that RPS division'to_ fail safe.
-RR9-


Attachment 4 Revised Technical Specification for Plant Staff Working Hours Revised Page:           226 Generic Letter 82-12 dated June 15, 1982, stated:
Revised Technical Specification for Plant Staff Working Hours Revised Page:
                            "Our letter of February 8, 1982, requested that you take action as necessary to revise the administrative section of your technical specifications to assure that your plant administrative procedures follow the revised working hour guidelines, including a provision for documentation of authorized deviations which should be availabic for NRC review. You should review your past actions to assure that they are consistent with the attached revised policy statement.
226 Generic Letter 82-12 dated June 15, 1982, stated:
"Our letter of February 8, 1982, requested that you take action as necessary to revise the administrative section of your technical specifications to assure that your plant administrative procedures follow the revised working hour guidelines, including a provision for documentation of authorized deviations which should be availabic for NRC review.
You should review your past actions to assure that they are consistent with the attached revised policy statement.
Note that the revised guidelines are to be incorporated by October 1,1982."
Note that the revised guidelines are to be incorporated by October 1,1982."
In discussions with the Staff, the District was directed to revise the Technical Specifications to state that working hours will be controlled in accordance with a CNS Station Operating Procedure.     This proposed change is attached.
In discussions with the Staff, the District was directed to revise the Technical Specifications to state that working hours will be controlled in accordance with a CNS Station Operating Procedure.
This proposed change is attached.


                                                                                            .            u j .a
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                                                                                                                                      ~
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j, '
6.3       Statien Optreting Procedures                   ,
6.3 Statien Optreting Procedures E
                                                                                                                                '    E
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: 6. 3.1 -
: 6. 3.1 -   Station personnel shall be provided detailed written procedures to bey '
Station personnel shall be provided detailed written procedures to bey '
usedforoperationandmaintenanceofsystem'componentsandhysdems that could have an effect on nuclear safety.j                             -
usedforoperationandmaintenanceofsystem'componentsandhysdems that could have an effect on nuclear safety.j
                                                                                                        .f [   4, x,
.f [
6.3.2     Written integrated and system procedures and instructions including applicable check off lists shall be provided and adhered to for the following:
4, x,
            'A.     Normal startup, operation, shutdown and fue handling operadio's                                     n of the station including all systems and c'omponents                 '
6.3.2 Written integrated and system procedures and instructions including applicable check off lists shall be provided and adhered to for the following:
involving nuclear safety.                                             '
'A.
Normal startup, operation, shutdown and fue handling operadio's n
of the station including all systems and c'omponents involving nuclear safety.
s t
s t
B. Actions to be taken to correct specific and forseen potential or actual malfunctions of safety related systems or. components including responses to alarms, primary system Jgaks and abnormal reactivity changes.
B.
t               >
Actions to be taken to correct specific and forseen potential or actual malfunctions of safety related systems or. components including responses to alarms, primary system Jgaks and abnormal reactivity changes.
C. Emergency conditions involving possible or actual releases of radio-active materials.                 y D.     Implementing procedures of the Sectrrity Plan and the Emergency Plan.
t C.
E. Implementing procedures for che fire protection program.
Emergency conditions involving possible or actual releases of radio-active materials.
                                                                              \
y D.
F.     Administrative procedures for shift overtime.                           '
Implementing procedures of the Sectrrity Plan and the Emergency Plan.
6.3.3     The following maintenance and test procedures.will be provided to satisfy routine inspection, preventive maintenance programs, and operating ' license requirements.
E.
i A.     Routine testing of Engineered Safeguards and eqbipment a , required by the facility License and the Technical Specifications.
Implementing procedures for che fire protection program.
B.     Routine testing of standby and redundant equipment.
\\
C.     Preventiveorcorrectivemaintenanceofplantequipmentandsysthms                                  y that could have an effect on nuclear safety.
F.
D.     Calibration and preventive maintenance of instrumentation that could ,
Administrative procedures for shift overtime.
6.3.3 The following maintenance and test procedures.will be provided to satisfy routine inspection, preventive maintenance programs, and operating ' license requirements.
i A.
Routine testing of Engineered Safeguards and eqbipment a, required by the facility License and the Technical Specifications.
B.
Routine testing of standby and redundant equipment.
Preventiveorcorrectivemaintenanceofplantequipmentandsysthms C.
y that could have an effect on nuclear safety.
D.
Calibration and preventive maintenance of instrumentation that could,
af fect the nuclear safety oflh'e plant.
af fect the nuclear safety oflh'e plant.
l                                                             ..
l E.
Special testing of equipment for proposed changes to operatEonal E.
Special testing of equipment for proposed changes to operatEonal procedures or proposed system design changes.
procedures or proposed system design changes.
6.3.4 Radiation control procedures shall be naintained and made available to all station personnel. These procedares shall(show permissible radiation exposure, and shall be consistent with the ' requirements of 10 CFR 20.
6.3.4     Radiation control procedures shall be naintained and made available to all station personnel. These procedares shall(show permissible radiation exposure, and shall be consistent with the ' requirements of 10 CFR 20.
i.
i.
Y.
I
I
                                                              ;'                 Y.
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q i
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                                                                                                                                  -a
-a
                                                                                                                                  \
\\
                                                            ./
./
I
I
;                                                    -226-
-226-
{           -                          , ._.
{


Attachment 5
(
(
Revised Technical Specification for SRAB Duties Revised Page:       222 In the recent Amendment 80 to the License , in Specification 6.2.1.B, the word " approved" was changed to
Revised Technical Specification for SRAB Duties Revised Page:
              " approve" as an apparent typographical error. As a lecent I&E inspection report pointed out, this minor alteration actually changed the duties which SRAB must perform.         The word
222 In the recent Amendment 80 to the License,
              " approve" is being changed back to " approved" because it is now clear that this word was not a typographical error as previously thought.
in Specification 6.2.1.B, the word " approved" was changed to
" approve" as an apparent typographical error.
As a lecent I&E inspection report pointed out, this minor alteration actually changed the duties which SRAB must perform.
The word
" approve" is being changed back to " approved" because it is now clear that this word was not a typographical error as previously thought.
1 i
1 i
i i, ^
i i, ^
Line 273: Line 434:


5.2 (cont'd) tary material reviewed; copies of the minutes shall be for-warded to the Chairman of the NPPD Safety Review and Audit Board and the Division Manager of Power Operations within one month.
5.2 (cont'd) tary material reviewed; copies of the minutes shall be for-warded to the Chairman of the NPPD Safety Review and Audit Board and the Division Manager of Power Operations within one month.
: 7. Procedures:
7.
Procedures:
Written administrative procedures for Committee operation shall be prepared and maintained describing the method for submission and content of presentations to the committee, provisions for use of subcommittees, review and approval by members of written Committee evaluations and recommendations, dissemination of minutes, and such other matters as may be appropriate.
Written administrative procedures for Committee operation shall be prepared and maintained describing the method for submission and content of presentations to the committee, provisions for use of subcommittees, review and approval by members of written Committee evaluations and recommendations, dissemination of minutes, and such other matters as may be appropriate.
B. NPPD Safety Review and Audit Board.
B.
The board must: verify that operation of the plant is consistent with company policy and rules, approved operating procedures and             l operating license provisions; review safety related plant changes, proposed tests and procedures; verify that unusual events are promptly investigated and corrected in a manner which reduces the probability of recurrence of such events; and detect trends which may not be apparent to a day-to-day observer.
NPPD Safety Review and Audit Board.
The board must: verify that operation of the plant is consistent with company policy and rules, approved operating procedures and l
operating license provisions; review safety related plant changes, proposed tests and procedures; verify that unusual events are promptly investigated and corrected in a manner which reduces the probability of recurrence of such events; and detect trends which may not be apparent to a day-to-day observer.
Audits of selected aspects of plant operation shall be performed with a frequency commensurate with their safety significance and in such i
Audits of selected aspects of plant operation shall be performed with a frequency commensurate with their safety significance and in such i
a manner as to assure that an audit of all nuclear safety related activities is completed within a period of two years. Periodic review of the audit programs should be performed by the Board at least twice a year to assure that such audits are being accomplished in accordance with requirements of Technical Specifications. The audits shall be performed in accordance with appropriate written instructions or procedures and should include verification of compliance with inter-nal rules, procedures (for example, normal, of f-normal, emergency, op-erating, maintenance, surveillance, test and radiation control proce-dures and the emergency and security plans), regulations involving nuclear safety and operating license provisions; training, qualification and performance of operating staff; and corrective actions following abnormal occurrences or unusual events. A representative portion of procedures and records of the activities performed during the audit period shall be audited and, in addition, observations of perfor-mance of operating and maintenance activities shall be included.
a manner as to assure that an audit of all nuclear safety related activities is completed within a period of two years.
Written reports of such audits shall be reviewed at a scheduled meeting of the Board and by appropriate members of management including those having responsibility in the area audited. Follow-up action, including reaudit of deficient areas, shall be taken when indicated.
Periodic review of the audit programs should be performed by the Board at least twice a year to assure that such audits are being accomplished in accordance with requirements of Technical Specifications. The audits shall be performed in accordance with appropriate written instructions or procedures and should include verification of compliance with inter-nal rules, procedures (for example, normal, of f-normal, emergency, op-erating, maintenance, surveillance, test and radiation control proce-dures and the emergency and security plans), regulations involving nuclear safety and operating license provisions; training, qualification and performance of operating staff; and corrective actions following abnormal occurrences or unusual events. A representative portion of procedures and records of the activities performed during the audit period shall be audited and, in addition, observations of perfor-mance of operating and maintenance activities shall be included.
Written reports of such audits shall be reviewed at a scheduled meeting of the Board and by appropriate members of management including those having responsibility in the area audited.
Follow-up action, including reaudit of deficient areas, shall be taken when indicated.
In addition to the above, the Safety Review and Audit Board will audit the facility fire protection and its implementing procedures at least once every 24 months.
In addition to the above, the Safety Review and Audit Board will audit the facility fire protection and its implementing procedures at least once every 24 months.
                                                  -222-
-222-


Attachment 6 Revised Technical Specification for Listing of Snubbers Revised Pages:       137a
Revised Technical Specification for Listing of Snubbers Revised Pages:
                                                .137b 137e 137f-137m The current Technical Specifications for Cooper Nuclear Station lists snubbers under three different categories on Tables 3.6.1, 3.6.2. , and 3.6.3.
137a
Nebraska Public Power District requests a revision to the Technical Specifications as shown on the attached pages.         In addition to revising the existing tables, this request will add a new table, Table 3.6.4, which lists Inaccessible Safety Related Hydraulic Shock Suppressors (Snubbers).
.137b 137e 137f-137m The current Technical Specifications for Cooper Nuclear Station lists snubbers under three different categories on Tables 3.6.1, 3.6.2., and 3.6.3.
Nebraska Public Power District requests a revision to the Technical Specifications as shown on the attached pages.
In addition to revising the existing tables, this request will add a new table, Table 3.6.4, which lists Inaccessible Safety Related Hydraulic Shock Suppressors (Snubbers).
This request is made for reasons as follows:
This request is made for reasons as follows:
(a)   In order to add a new category (table).
(a)
In order to add a new category (table).
(b) To have a listing that is compatible with what is contained in the District's computer system and to facilitate data withdrawals and entries.
(b) To have a listing that is compatible with what is contained in the District's computer system and to facilitate data withdrawals and entries.
(c) Some snubber listings have been added/ deleted as a result of modifications to Terus Attached Piping.
(c) Some snubber listings have been added/ deleted as a result of modifications to Terus Attached Piping.
Line 293: Line 462:
This change will provide revised tables that are functionally superior to the existing tables.
This change will provide revised tables that are functionally superior to the existing tables.


T   +-
T
+-
[
[
LIMITING CONDITION FOR OPERATION                   SURVEILLANCE REQUIREMENT 3.6.H         Shock Suppressors (Snubbers)         4.6.H Shock Suppressors (Snubbers)
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.H Shock Suppressors (Snubbers) 4.6.H Shock Suppressors (Snubbers) 1.
: 1. During all modes of operation             The following surveillance require-except Cold Shutdown and Refuel,         ments apply to all snubbers listed all safety related snubbers shall         in Tables 3.6.1, 3.6.2, 3.6.3, and be operable except as noted in           3.6.4.                                       l 3.6.H.2 through 3.6.H.5 below.
During all modes of operation The following surveillance require-except Cold Shutdown and Refuel, ments apply to all snubbers listed all safety related snubbers shall in Tables 3.6.1, 3.6.2, 3.6.3, and be operable except as noted in 3.6.4.
: 1. All snubbers shall be visually
l 3.6.H.2 through 3.6.H.5 below.
: 2. The snubbers listed in Tables                   inspected in accordance with 3.6.1, 3.6.2, 3.6.3, and 3.6.4                 the following schedule:
1.
are required to protect the primary coolant system or other safety             Number of Snubbers         Next Required related systems or components.           Found Inoperable           Inspection All others are therefore exempt           During Inspection           Interval from these specifications,               or During Inspection Interval
All snubbers shall be visually 2.
: 3. With one or more snubbers in-operable, within 72 hours re-             0                     18 months f; 25%
The snubbers listed in Tables inspected in accordance with 3.6.1, 3.6.2, 3.6.3, and 3.6.4 the following schedule:
place or restore the inoper-               1                     12 months f; 25%
are required to protect the primary coolant system or other safety Number of Snubbers Next Required related systems or components.
able snubber (s) to OPERABLE               2                     6 months + 25%
Found Inoperable Inspection All others are therefore exempt During Inspection Interval from these specifications, or During Inspection Interval 3.
status and perform an engi-                 3, 4                 124 days   [25%
With one or more snubbers in-operable, within 72 hours re-0 18 months f; 25%
neering evaluation per                     5,6,7                 62 days   f;25%
place or restore the inoper-1 12 months f; 25%
Specification 4.6.H.4 on                   8 or more             31 days   j; 25%         '
able snubber (s) to OPERABLE 2
the supported component or declare the supported system or                 The required inspection interval subsystem inoperable and follow                 shall not be lengthened more the appropriate ACTION state-                   than one step at a time.
6 months + 25%
status and perform an engi-3, 4 124 days
[25%
neering evaluation per 5,6,7 62 days f;25%
Specification 4.6.H.4 on 8 or more 31 days j; 25%
the supported component or declare the supported system or The required inspection interval subsystem inoperable and follow shall not be lengthened more the appropriate ACTION state-than one step at a time.
ment for that system.
ment for that system.
Snubbers may be categorized in
Snubbers may be categorized in 4.
: 4. If a snubber is determined to be                 groups, " accessible" or "inac-inoperable while the reactor is                 cessible" based on their acces-
If a snubber is determined to be groups, " accessible" or "inac-inoperable while the reactor is cessible" based on their acces-in the shutdown or refuel mode, sibility for inspection during the snubber shall be made oper-reactor operation and by type, able or replaced prior to reactor hydraulic or mechanical. These
'                                                                    sibility for inspection during in the shutdown or refuel mode, the snubber shall be made oper-                 reactor operation and by type,           ^
^
able or replaced prior to reactor               hydraulic or mechanical. These startup.                                         four groups may be inspected independently according to the
startup.
: 5. Snubbers may be added to, removed,               above schedule.
four groups may be inspected independently according to the 5.
or substituted for, by analysis, from safety related systems with-           2. Visual Inspection Acceptance out prior License Amendment to                   Criteria Tables 3.6.1, 3.6.2, 3.6.3, and 3.6.4, provided that a revision                 Visual inspections shall verify l
Snubbers may be added to, removed, above schedule.
l                  to these tables is included with                 (1) that there are no visible-l                   a subsequent License Amendment                   indications of damage or impair-request.                                         ed OPERABILITY, (2) attachments to the foundation or supporting l
or substituted for, by analysis, from safety related systems with-2.
Visual Inspection Acceptance out prior License Amendment to Criteria Tables 3.6.1, 3.6.2, 3.6.3, and l
3.6.4, provided that a revision Visual inspections shall verify l
to these tables is included with (1) that there are no visible-l a subsequent License Amendment indications of damage or impair-request.
ed OPERABILITY, (2) attachments to the foundation or supporting l
l
l
                                                    -137a-
-137a-


LIMITING CONDITION FOR OPERATION       SURVEILLANCE REQUIREMENT 4.6.H   Shock Suppressors (Snubbers)
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 4.6.H Shock Suppressors (Snubbers)
(cont'd) structure are secure. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE ior the purpose of establishing the next visual inspection interval, providing that (1) the cause of the rejection is clearly estab-lished and remedied for that particular snubber and for other snubbers that may be generically susceptible; or (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE per Specifi-cations 4.6 H.6 or 4.6.H.7 as applicable. However, when the fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be determined in-operable and cannot be determined OPERABLE via functional testing for the purpose of establishing the next visual inspection inter-val. All snubbers connected to an inoperable common hydraulic fluid reservoir shall be counted as inoperabic snubbers.
(cont'd) structure are secure.
: 3. At least once per-18 months dur-
Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE ior the purpose of establishing the next visual inspection interval, providing that (1) the cause of the rejection is clearly estab-lished and remedied for that particular snubber and for other snubbers that may be generically susceptible; or (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE per Specifi-cations 4.6 H.6 or 4.6.H.7 as applicable. However, when the fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be determined in-operable and cannot be determined OPERABLE via functional testing for the purpose of establishing the next visual inspection inter-val. All snubbers connected to an inoperable common hydraulic fluid reservoir shall be counted as inoperabic snubbers.
                                                    .ing shutdown, a representative sample, 10% of the total of each type of snubber in use in the plant, shall be functionally tested either in place or in a bench test. For each snubber that does not meet the func-tional test acceptance criteria of Specification 4.6.H.5 or 4.6.H.6, an additional 10% of that type of snubber shall be function-ally tested.
3.
: 4. The representative sample select-ed for functional testing shall include various configuration, operating environments and the range of size and capacity of snubbers. Tables 3.6.1, 3.6.2, 3.6.3, and 3.6.4 may be used     l jointly or separately as the basis for the sampling plan.
At least once per-18 months dur-
                                  -137b-
.ing shutdown, a representative sample, 10% of the total of each type of snubber in use in the plant, shall be functionally tested either in place or in a bench test.
For each snubber that does not meet the func-tional test acceptance criteria of Specification 4.6.H.5 or 4.6.H.6, an additional 10% of that type of snubber shall be function-ally tested.
4.
The representative sample select-ed for functional testing shall include various configuration, operating environments and the range of size and capacity of snubbers. Tables 3.6.1, 3.6.2, 3.6.3, and 3.6.4 may be used l
jointly or separately as the basis for the sampling plan.
-137b-


  ' LIMITING CONDITION FOR OPERATION       SURVEILLANCE REQUIREMENT 4.6.H Shock Suppressors (Snubbers)
' LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 4.6.H Shock Suppressors (Snubbers)
(cont'd)
(cont'd)
Concurrent with the first in-service visual inspection and at least once per 18 months thereafter, the installation and maintenance records of each snubber listed in Tables 3.6.1, 3.6.2, 3.6.3, and-3.6.4   l shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review. If the indicated ser-vice life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be reevaluated or the snubber shall be replaced or recondi-tioned so as to extend its service life beyond the date of the next scheduled service life review. This reevaluation, replacement or reconditioning shall be indicated in the records.
Concurrent with the first in-service visual inspection and at least once per 18 months thereafter, the installation and maintenance records of each snubber listed in Tables 3.6.1, 3.6.2, 3.6.3, and-3.6.4 l
  'e
shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review.
                                      -137e-
If the indicated ser-vice life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be reevaluated or the snubber shall be replaced or recondi-tioned so as to extend its service life beyond the date of the next scheduled service life review. This reevaluation, replacement or reconditioning shall be indicated in the records.
'e
-137e-


Table 3.6.1 ACCESSIBLE SAFETY RELATED HYDRAULIC SHOCK SUPPRESSORS (SNUBBERS)
Table 3.6.1 ACCESSIBLE SAFETY RELATED HYDRAULIC SHOCK SUPPRESSORS (SNUBBERS)
Snubber                                               Location CS-SNUB-(CS-SI)                                         R-903-SE CS-SNUB-(CS-SIO)                                       R-931-NE CS-SNUB-(CS-Sil)                                       R-931-NE CS-SNUB-(CS-S2)                                         R-903-SE CS-SNUB-(CS-S3)                                         R-931-SE CS-SNUB-(CS-S6)                                         R-881-SE QUAD CS-SNUB-(CS-S7)                                         R-881-SE QUAD CS-SNUB-(CS-VE7)                                       R-881-SE QUAD HPCI-SNUB-(HP-Sil)                                     R-859-HPCI RM HPCI-SNUB-(HP-SIS)                                     R-859-HPCI RM llPCI-SNUB-(HP-S18)                                     R-859-SW QUAD HPCI-SNUB-(HP-S18A)                                     R-859-HPCI RM HPCI-SNUB-(HP-S22A)                                     R-859-llPCI RM llPCI-SNUB-(HP-S4)                                     R-859-SW QUAD HPCI-SNUB-(RF-S3)                                       R-859-IIPCI RM HPCI-SNUB-(RF-S4)                                       R-881-SW TORUS HPCI-SNUB-(RF-SS)                                       R-881-SW TORUS MS-SNUB-(BS-SI)                                         R-881-SW TORUS MS-SNUB-(BS-Sil3A)                                     R-881-NW TORUS MS-SNUB-(BS-Sil3B)                                     R-881-NW TORUS MS-SNUB-(BS-Sll6A)                                     R-881-NW TORUS MS-SNUB-(BS-5116B)                                     R-881-NW TORUS MS-SNUB-(BS-S125A)                                     R-881-SW TORUS MS-SNUB-(BS-S125B)                                     R-881-SW TORUS MS-SNUB-(BS-S2)                                         R-881-SW TORUS MS-SNUB-(BS-S3)                                         R-881-NW TORUS MS-SNUB-(BS-S4)                                         R-903-A RPR llX RM MS-SNUB-(BS-SS)                                       R-903-A RHR llX RM MS-SNUB-(MS-SI)                                         R-859-HPCI RM MS-SNUB-(MS-SIO)                                       R-881-SW TORUS MS-SNUB-(MS-Sil)                                       R-881-SW TORUS MS-SNUB-(MS-SillA)                                     R-903-A RHR HX RM MS-SNUB-(MS-SilA)                                       R-881-SW TORUS MS-SNUB-(MS-S12)                                       R-881-SW TORUS MS-SNUB-(MS-S12A)                                       R-881-SW TORUS MS-SNUB-(MS-S13)                                       R-903-B RHR 11X RM MS-SNUB-(MS-S13A)                                       R-903-B RHR HX FM MS-SNUB-(MS-S13B)                                     R-903-B RHR HX RM MS-SNUB-(MS-S14)                                       R-903-B RHR HX RM MS-SNUB-(MS-SIS)                                       R-931-B RHR HX RM PS-SNUB-(MS-SI5A)                                     R-931-B RilR HX RM MS-SNUB-(MS-S16A)                                     R-881-NW TORUS MS-SNUB-(MS-S16B)                                     R-881-NW TORUS MS-SNUB-(MS-S17)                                       R-903-A RilR llX RM MS-SNUB-(MS-S18)                                       R-903-A RilR HX RM MS-SNUB-(MS-S19)                                       R-903-A RHR llX RM MS-SNUB-(MS-S2)                                       R-859-ilPCI RM MS-SNUB-(MS-S20)                                       R-931-A RHR HX FM MS-SNUB-(MS-S20A)                                     R-931-A RHR HX RM MS-SNUB-(MS-S23)                                       R-881-NE TORUS MS-SNUB-(MS-S24)                                       R-881-NE TORUS MS-SNUB-(MS-S25)                                       R-859-NE QUAD MS-SNUB-(MS-S26)                                       R-859-NE QUAD
Snubber Location CS-SNUB-(CS-SI)
                                      -137f-
R-903-SE CS-SNUB-(CS-SIO)
R-931-NE CS-SNUB-(CS-Sil)
R-931-NE CS-SNUB-(CS-S2)
R-903-SE CS-SNUB-(CS-S3)
R-931-SE CS-SNUB-(CS-S6)
R-881-SE QUAD CS-SNUB-(CS-S7)
R-881-SE QUAD CS-SNUB-(CS-VE7)
R-881-SE QUAD HPCI-SNUB-(HP-Sil)
R-859-HPCI RM HPCI-SNUB-(HP-SIS)
R-859-HPCI RM llPCI-SNUB-(HP-S18)
R-859-SW QUAD HPCI-SNUB-(HP-S18A)
R-859-HPCI RM HPCI-SNUB-(HP-S22A)
R-859-llPCI RM llPCI-SNUB-(HP-S4)
R-859-SW QUAD HPCI-SNUB-(RF-S3)
R-859-IIPCI RM HPCI-SNUB-(RF-S4)
R-881-SW TORUS HPCI-SNUB-(RF-SS)
R-881-SW TORUS MS-SNUB-(BS-SI)
R-881-SW TORUS MS-SNUB-(BS-Sil3A)
R-881-NW TORUS MS-SNUB-(BS-Sil3B)
R-881-NW TORUS MS-SNUB-(BS-Sll6A)
R-881-NW TORUS MS-SNUB-(BS-5116B)
R-881-NW TORUS MS-SNUB-(BS-S125A)
R-881-SW TORUS MS-SNUB-(BS-S125B)
R-881-SW TORUS MS-SNUB-(BS-S2)
R-881-SW TORUS MS-SNUB-(BS-S3)
R-881-NW TORUS MS-SNUB-(BS-S4)
R-903-A RPR llX RM MS-SNUB-(BS-SS)
R-903-A RHR llX RM MS-SNUB-(MS-SI)
R-859-HPCI RM MS-SNUB-(MS-SIO)
R-881-SW TORUS MS-SNUB-(MS-Sil)
R-881-SW TORUS MS-SNUB-(MS-SillA)
R-903-A RHR HX RM MS-SNUB-(MS-SilA)
R-881-SW TORUS MS-SNUB-(MS-S12)
R-881-SW TORUS MS-SNUB-(MS-S12A)
R-881-SW TORUS MS-SNUB-(MS-S13)
R-903-B RHR 11X RM MS-SNUB-(MS-S13A)
R-903-B RHR HX FM MS-SNUB-(MS-S13B)
R-903-B RHR HX RM MS-SNUB-(MS-S14)
R-903-B RHR HX RM MS-SNUB-(MS-SIS)
R-931-B RHR HX RM PS-SNUB-(MS-SI5A)
R-931-B RilR HX RM MS-SNUB-(MS-S16A)
R-881-NW TORUS MS-SNUB-(MS-S16B)
R-881-NW TORUS MS-SNUB-(MS-S17)
R-903-A RilR llX RM MS-SNUB-(MS-S18)
R-903-A RilR HX RM MS-SNUB-(MS-S19)
R-903-A RHR llX RM MS-SNUB-(MS-S2)
R-859-ilPCI RM MS-SNUB-(MS-S20)
R-931-A RHR HX FM MS-SNUB-(MS-S20A)
R-931-A RHR HX RM MS-SNUB-(MS-S23)
R-881-NE TORUS MS-SNUB-(MS-S24)
R-881-NE TORUS MS-SNUB-(MS-S25)
R-859-NE QUAD MS-SNUB-(MS-S26)
R-859-NE QUAD
-137f-


Table 3.6.1 ACCESSIBLE SAFETY RELATED HYDRAULIC SHOCK SUPPRESSORS (SNUBBERS) (cont'd)
Table 3.6.1 ACCESSIBLE SAFETY RELATED HYDRAULIC SHOCK SUPPRESSORS (SNUBBERS) (cont'd)
Snubber                                           Location MS-SNUB-(MS-S3)                                     R-859-HPCI RM HS-SNUB-(MS-S4)                                     R-859-HPCI RM MS-SNUB-(MS-S75)                                   R-931-A RilR HX RM MS-SNUB-(MS-S76)                                   R-931-B RHR HX RM MS-SNUB-(MS-S7A)                                   R-859-UPCI RM MS-SNUB-(HS-S7B)                                   R-859-HPCI RM MS-SNUB-(MS-S8)                                     R-881-SW TORUS RCIC-SNUB-(RF-SI)                                   R-881-NE QUAD RCIC-SNUB-(RF-SIA)                                 R-881-NE QUAD RCIC-SNUB-(RF-S45C)                                 R-881-NE QUAD RCIC-SNUB-(RF-S45D)                                 R-881-NE QUAD RCIC-SNUB-(RF-S46A)                                 R-881-NE QUAD RCIC-SNUB-(RF-SSIA)                                 R-881-NE TORUS       ,
Snubber Location MS-SNUB-(MS-S3)
RCIC-SNUB-(RF-SSIB)                                 R-881-NE TORUS REC-SNUB-(RCC-S20)                                 R-931-NW REC-SNUB-(RCC-S21)                                 R-931-hv REC-SNUB-(RCC-S22)                                 R-931-SW I
R-859-HPCI RM HS-SNUB-(MS-S4)
RFC-SNUB-(RCC-S3)                                   R-931-NE REC-SNUB-(RCC-S4)                                   R-931-NE RF-SNUB-(RF-S2)                                     R-881-NE TORUS RF-SNUB-(RF-S6)                                     R-881-SE TORUS RHR-SNUB-(RH-S103A)                                 R-859-SW QUAD RHR-SNUB-(Ril-S107A)                               R-859-NW OPAD RHR-SNUB-(RE-S20)                                   R-903-INJ V RM RilR-SNUB-(RH-S21)                                 R-903-INJ V RM RHR-SNUB-(RH-S22)                                   R-881-hv TORUS RHR-SNUB-(RH-S23)                                   R-881-NW TORUS RHR-SNUB-(RE-S24)                                   R-881-NW TORUS I
R-859-HPCI RM MS-SNUB-(MS-S75)
RHR-SNUB-(RH-S25)                                   R-903-NU RHR-SNUB-(RH-S25A)                                 R-903-NW RHR-SNUB-(RF-S26)                                   R-903-NW RHR-SNUB-(RH-S27A)                                 R-931-A RHR HX RM RHR-SNUB-(RH-S29)'                                 P-903-INJ V RM RilR-ShTB-(RH-S30A)                                 R-881-SW TORUS RHR-SNUB-(RH-S30B)                                 R-881-SW TORUS RHR-SNUB-(RH-S32)                                   R-881-SW TORUS RHR-SNUB-(RH-S33D)                                 R-8SI-NW TORUS RHR-SNUB-(RH-S34)                                   R-903-SW l       RHR-SNUB-(RH-S35)                                   R-903-B RHR llX RM RilR-SNUB-(RH-S36)                                 R-903-B RHR llX RM RER-SNUB-(RH-S37)                                   R-903-B RHR HX RM RHR-SNUB-(RH-S38)                                   R-903-B RHR llX RM RHR-SNUB-(RH-S39)                                   R-903-R RIIR HX RM RHR-SNUS-(RH-S40)                                   R-903-B RilR HX RM RPR-SNUB-(RH-S41)                                   R-859-SW OUAD RHR-SNUB-(RH-S42)                                   R-859-SW QUAD RHR-SNUB-(RH-S43)                                   R-881-SW TORUS RHR-SNUB-(RH-S44)                                   R-881-SW QUAD RHR-SNUB-(RH-S45)                                   R-881-SW QUAD RHR-SNUB-(RH-S48)                                   R-881-NW QUAD RHR-SNUB-(RH-S49)                                   R-881-NW QUAD RHR-SNUB-(RH-SSI)                                   P-903-A RHR HX RM RHR-SNUB-(RH-SS2)                                   R-903-A RHR HX RM
R-931-A RilR HX RM MS-SNUB-(MS-S76)
                                        -137g-
R-931-B RHR HX RM MS-SNUB-(MS-S7A)
R-859-UPCI RM MS-SNUB-(HS-S7B)
R-859-HPCI RM MS-SNUB-(MS-S8)
R-881-SW TORUS RCIC-SNUB-(RF-SI)
R-881-NE QUAD RCIC-SNUB-(RF-SIA)
R-881-NE QUAD RCIC-SNUB-(RF-S45C)
R-881-NE QUAD RCIC-SNUB-(RF-S45D)
R-881-NE QUAD RCIC-SNUB-(RF-S46A)
R-881-NE QUAD RCIC-SNUB-(RF-SSIA)
R-881-NE TORUS RCIC-SNUB-(RF-SSIB)
R-881-NE TORUS REC-SNUB-(RCC-S20)
R-931-NW REC-SNUB-(RCC-S21)
R-931-hv REC-SNUB-(RCC-S22)
R-931-SW I
RFC-SNUB-(RCC-S3)
R-931-NE REC-SNUB-(RCC-S4)
R-931-NE RF-SNUB-(RF-S2)
R-881-NE TORUS RF-SNUB-(RF-S6)
R-881-SE TORUS RHR-SNUB-(RH-S103A)
R-859-SW QUAD RHR-SNUB-(Ril-S107A)
R-859-NW OPAD RHR-SNUB-(RE-S20)
R-903-INJ V RM RilR-SNUB-(RH-S21)
R-903-INJ V RM RHR-SNUB-(RH-S22)
R-881-hv TORUS RHR-SNUB-(RH-S23)
R-881-NW TORUS RHR-SNUB-(RE-S24)
R-881-NW TORUS I
RHR-SNUB-(RH-S25)
R-903-NU RHR-SNUB-(RH-S25A)
R-903-NW RHR-SNUB-(RF-S26)
R-903-NW RHR-SNUB-(RH-S27A)
R-931-A RHR HX RM RHR-SNUB-(RH-S29)'
P-903-INJ V RM RilR-ShTB-(RH-S30A)
R-881-SW TORUS RHR-SNUB-(RH-S30B)
R-881-SW TORUS RHR-SNUB-(RH-S32)
R-881-SW TORUS RHR-SNUB-(RH-S33D)
R-8SI-NW TORUS RHR-SNUB-(RH-S34)
R-903-SW l
RHR-SNUB-(RH-S35)
R-903-B RHR llX RM RilR-SNUB-(RH-S36)
R-903-B RHR llX RM RER-SNUB-(RH-S37)
R-903-B RHR HX RM RHR-SNUB-(RH-S38)
R-903-B RHR llX RM RHR-SNUB-(RH-S39)
R-903-R RIIR HX RM RHR-SNUS-(RH-S40)
R-903-B RilR HX RM RPR-SNUB-(RH-S41)
R-859-SW OUAD RHR-SNUB-(RH-S42)
R-859-SW QUAD RHR-SNUB-(RH-S43)
R-881-SW TORUS RHR-SNUB-(RH-S44)
R-881-SW QUAD RHR-SNUB-(RH-S45)
R-881-SW QUAD RHR-SNUB-(RH-S48)
R-881-NW QUAD RHR-SNUB-(RH-S49)
R-881-NW QUAD RHR-SNUB-(RH-SSI)
P-903-A RHR HX RM RHR-SNUB-(RH-SS2)
R-903-A RHR HX RM
-137g-


Table 3.6.1 ACCESSIBLE SAFETY RELATED HYDRAULIC SHOCK SUPPRESSORS (SNUBBERS) (cont'd)
Table 3.6.1 ACCESSIBLE SAFETY RELATED HYDRAULIC SHOCK SUPPRESSORS (SNUBBERS) (cont'd)
Snubber                                           Location RHR-SNUB-(RH-S54)                                   R-859-NW QUAD RFR-SNUB-(RH-S55)                                   R-859-NW QUAD RHR-SNUB-(RH-S56)                                   R-903-A RilR HX RM RHR-SNUB-(RH-S57)                                   R-903-A RilR HX RM RHR-SNUB-(RH-S59)                                   R-881-NW TORUS RHR-SNUB-(RH-S65)                                   R-881-SW QUAD RHR-SNUB-(RH-566)                                   R-903-INJ V RM RilR-SNUB-(RH-S76A)                                 R-881-SW TORUS RHR-SNUB-(RH-S76B)                                 R-881-SW TORUS RHR-SNUB-(RH-S77)                                   R-881-SW TORUS RHR-SNUB-(RH-S78A)                                 R-881-NW TORUS RIIR-SNUB-(RH-S78B)                                 R-881-NW TORUS RHR-SNUB-(Ril-S80)                                 R-881-NW QUAD RHR-SNUB-(Ril-S96A)                                 R-903-NW RiiR-SNUB-(RH-S98)                                 R-881-NW QUAD RWCU-SNUB-(CU-S89)                                 R-881-SE TORUS SW-SNUB-(SW-il23A)                                 IS-SWP RM SW-SNUB-(SW-H23D)                                   IS-SWP RM SW-SNUB-(SW-il23E)                                 IS-SWP RM SW-SNUB-(SW-H2311)                                 IS-SWP RM 4
Snubber Location RHR-SNUB-(RH-S54)
                                    -137h-
R-859-NW QUAD RFR-SNUB-(RH-S55)
R-859-NW QUAD RHR-SNUB-(RH-S56)
R-903-A RilR HX RM RHR-SNUB-(RH-S57)
R-903-A RilR HX RM RHR-SNUB-(RH-S59)
R-881-NW TORUS RHR-SNUB-(RH-S65)
R-881-SW QUAD RHR-SNUB-(RH-566)
R-903-INJ V RM RilR-SNUB-(RH-S76A)
R-881-SW TORUS RHR-SNUB-(RH-S76B)
R-881-SW TORUS RHR-SNUB-(RH-S77)
R-881-SW TORUS RHR-SNUB-(RH-S78A)
R-881-NW TORUS RIIR-SNUB-(RH-S78B)
R-881-NW TORUS RHR-SNUB-(Ril-S80)
R-881-NW QUAD RHR-SNUB-(Ril-S96A)
R-903-NW RiiR-SNUB-(RH-S98)
R-881-NW QUAD RWCU-SNUB-(CU-S89)
R-881-SE TORUS SW-SNUB-(SW-il23A)
IS-SWP RM SW-SNUB-(SW-H23D)
IS-SWP RM SW-SNUB-(SW-il23E)
IS-SWP RM SW-SNUB-(SW-H2311)
IS-SWP RM 4
-137h-


Table 3.6.2 ACCESSIBLE SAFETY RELATED MECHANICAL SHOCK SUPPRESSORS (SNUBBERS)
Table 3.6.2 ACCESSIBLE SAFETY RELATED MECHANICAL SHOCK SUPPRESSORS (SNUBBERS)
Snubber                                               Location MS-SFUB-(MS-S149B)                                       R-903-STM TUNNEL MS-SNUB-(MS-S16)                                         R-881-NW TORUS MS-SNUB-(MS-S9A)                                         R-881-SW TORUS MS-SNUB-(MS-S9B)                                         R-881-SW TORUS RCIC-SNUB-(RF-S51C)                                       R-881-NE TORUS-RHR-SNUB-(RH-S58)                                         R-903-A RHR HX RM SGT-SNUB-(PSSP-40)                                       R-881-SW TORUS SGT-SNUB-(PSSP-74)                                       R-881-SW TORUS SW-SNUB-(SW-H23B)                                         IS-SWP RM SW-SNUB-(SW-H23C)                                         IS-SWP RM SW-SNUB-(SW-H23F)                                 g       IS-SWP RM SW-SNUB-(SW-H23G)                                         IS-SWP PJi 4
Snubber Location MS-SFUB-(MS-S149B)
R-903-STM TUNNEL MS-SNUB-(MS-S16)
R-881-NW TORUS MS-SNUB-(MS-S9A)
R-881-SW TORUS MS-SNUB-(MS-S9B)
R-881-SW TORUS RCIC-SNUB-(RF-S51C)
R-881-NE TORUS-RHR-SNUB-(RH-S58)
R-903-A RHR HX RM SGT-SNUB-(PSSP-40)
R-881-SW TORUS SGT-SNUB-(PSSP-74)
R-881-SW TORUS SW-SNUB-(SW-H23B)
IS-SWP RM SW-SNUB-(SW-H23C)
IS-SWP RM SW-SNUB-(SW-H23F) g IS-SWP RM SW-SNUB-(SW-H23G)
IS-SWP PJi 4
b 4
b 4
i l
i l
l l
l l
l
l
!                                                            -1371-
-1371-


Table 3.6.3 INACCESSIBLE SAFETY RELATED MECHANICAL SHOCK SUPPRESSORS (SNUBBERS)
Table 3.6.3 INACCESSIBLE SAFETY RELATED MECHANICAL SHOCK SUPPRESSORS (SNUBBERS)
Snubber                                               Location CS-SNUB-(CS-S4)                                       DW-934 CS-SNUB-(CS-SS)                                       DW-934 CS-SNUB-(CS-S8)                                       DW-934 CS-SNUB-(CS-S9)                                       DW-934 MS-SNUB-(MS-S21)                                       DW-901 l.
Snubber Location CS-SNUB-(CS-S4)
MS-SNUB-(MS-S22)                                       DW-901 MS-SNUB-(MS-S63)                                       DW-921             i MS-SNUB-(SS-A2)                                         DW-921 MS-SNUB-(SS-A3)                                         DW-921 MS-SNUB-(SS-B2)                                         DW-921 MS-SNUB-(SS-B3)                                         DW-921 MS-SNUB-(SS-C2)                                         DW-921 MS-SNUB-(SS-C3)                                         DW-921 MS-SNUB-(SS-D2)                                         DW-921 MS-SNUB-(SS-D3)                                         DW-921 MS-SNUB-(VR-55-23-X)                                   DW-901
DW-934 CS-SNUB-(CS-SS)
      !!S-SNUB-(VR-55-26-Z)                                   DW-901           i MS-SNUB-(VR-55-9-Y)                                     DW-901           l I
DW-934 CS-SNUB-(CS-S8)
MS-SNUB-(VR-55-9-Z)                                     DW-901 I
DW-934 CS-SNUB-(CS-S9)
MS-SNUB-(VR-56-12-Y)                                   DW-901 MS-SNUB-(VR-56-24-X)                                     DW-901 MS-SNUB-(VR-58-12-Y)                                     DW-921 MS-SNUB-(VR-59-7-X)                                     DW-921 MS-SNUB-(VR-59-7-Z)                                     DW-901 MS-SNUB-(VR-60-7-X)                                     DW-921 MS-SNUB-(VR-60-7-Z)                                     DW-901         -
DW-934 MS-SNUB-(MS-S21)
MS-SNUB-(VR 17-X)                                   DW-901 MS-SNUB-(VR-61-8-X)                                     DW-901 MS-SNUB-(VR-61-8-Z)                                     DW-921 MS-SNUB-(VR-62-17-X)                                     DW-901 MS-SNUB-(VR-62-8-X)                                     DW-901 MS-SNUB-(VR-62-8-Z)                                     DW-921 MS-SNUB-(VR-H61D)                                       DW-888 MS-SNUB-(VR-H62B)                                       DW-888 MS-SNUB-(VR-H62C)                                       DW-888 MS-SNUB-(VR-H63B)                                       DW-888
DW-901 l
;    MS-SNUB-(VR-H63C)                                       DW-888 MS-SNUB-(VR-H64D)                                       DW-888 MS-SNUB-(VR-SI)                                         DW-901 MS-SNUB-(VR-SIO)                                         DW-901 MS-SNUB-(VR-SI1)                                         DW-921 MS-SNUB-(VR-S12)                                         DW-901
MS-SNUB-(MS-S22)
      !!S-SNUL-(VR-S 14 )                                     DW-888 MS-SNUB-(VR-S2)                                         DW-901 MS-SimB-(VR-S20)                                         DW-921 MS-SNUB-(VR-S21)                                         DW-921 MS-SNUB-(VR-S22)                                         DW-901 MS-SNUB-(VR-S23A)                                       DW-901 MS-SNUB-(VR-S23B)                                       DW-901 MS-SNUB-(VR-S24A)                                       DW-901 MS-SNUB-(VR-S24B)                                       DW-901 MS-SNUB-(VR-S25)                                         DW-901
DW-901 MS-SNUB-(MS-S63)
                                        -137j-
DW-921 i
MS-SNUB-(SS-A2)
DW-921 MS-SNUB-(SS-A3)
DW-921 MS-SNUB-(SS-B2)
DW-921 MS-SNUB-(SS-B3)
DW-921 MS-SNUB-(SS-C2)
DW-921 MS-SNUB-(SS-C3)
DW-921 MS-SNUB-(SS-D2)
DW-921 MS-SNUB-(SS-D3)
DW-921 MS-SNUB-(VR-55-23-X)
DW-901
!!S-SNUB-(VR-55-26-Z)
DW-901 i
MS-SNUB-(VR-55-9-Y)
DW-901 l
I MS-SNUB-(VR-55-9-Z)
DW-901 MS-SNUB-(VR-56-12-Y)
DW-901 I
MS-SNUB-(VR-56-24-X)
DW-901 MS-SNUB-(VR-58-12-Y)
DW-921 MS-SNUB-(VR-59-7-X)
DW-921 MS-SNUB-(VR-59-7-Z)
DW-901 MS-SNUB-(VR-60-7-X)
DW-921 MS-SNUB-(VR-60-7-Z)
DW-901 MS-SNUB-(VR 17-X)
DW-901 MS-SNUB-(VR-61-8-X)
DW-901 MS-SNUB-(VR-61-8-Z)
DW-921 MS-SNUB-(VR-62-17-X)
DW-901 MS-SNUB-(VR-62-8-X)
DW-901 MS-SNUB-(VR-62-8-Z)
DW-921 MS-SNUB-(VR-H61D)
DW-888 MS-SNUB-(VR-H62B)
DW-888 MS-SNUB-(VR-H62C)
DW-888 MS-SNUB-(VR-H63B)
DW-888 MS-SNUB-(VR-H63C)
DW-888 MS-SNUB-(VR-H64D)
DW-888 MS-SNUB-(VR-SI)
DW-901 MS-SNUB-(VR-SIO)
DW-901 MS-SNUB-(VR-SI1)
DW-921 MS-SNUB-(VR-S12)
DW-901
!!S-SNUL-(VR-S 14 )
DW-888 MS-SNUB-(VR-S2)
DW-901 MS-SimB-(VR-S20)
DW-921 MS-SNUB-(VR-S21)
DW-921 MS-SNUB-(VR-S22)
DW-901 MS-SNUB-(VR-S23A)
DW-901 MS-SNUB-(VR-S23B)
DW-901 MS-SNUB-(VR-S24A)
DW-901 MS-SNUB-(VR-S24B)
DW-901 MS-SNUB-(VR-S25)
DW-901
-137j-


2   3   4   L4<&+ 2 L4 Table 3.6.3 INACCESSIBLE SAFETY RELATED MECHANICAL SHOCK SUPPRESSORS (SNUBBERS) (cont'd)
2 3
            ~ Snubber                                           Location MS-SNUB-(VR-S26).                                     DW-888 MS-SNUB-(VR-S27)                                     DW-901
4 L4<&+
    !!S-SNUB-(VR-S3)                                     DW-888 MS-SNUB-(VR-S30)                                     DW-921                     l MS-SNUB-(VR-S31)                                     DW-921 MS-SNUB-(VR-S32)                                     DW-888 MS-SNUB-(VR-S4)-                                     DV-901 MS-SNUB-(VR-S40)                                     DW-921 MS-SNUB-(VR-S41)-                                     DW-921 MS-SNUB-(VR-S42A)                                     DW-921 MS-SNUB-(VR-S42B)                                     DW-921 MS-SNUB-(VR-S43)                                     DW-888 MS-SNUB-(VR-S50A)                                     DW-921 MS-SNI'B-(VR-S50B)                                   DW-921 MS-SNUB-(VR-SSI)                                     DW-888
2 L4 Table 3.6.3 INACCESSIBLE SAFETY RELATED MECHANICAL SHOCK SUPPRESSORS (SNUBBERS) (cont'd)
    !!S-SNUB-(VR-SS A)                                   DW-901 ItS-SNUB-(V R-SS B)                                   DW-901 MS-SNUB-(VR-S6)                                       DW-901 MS-SNUB-(VR-S60)                                     DP-921 MS-SNUB-(VR-S61)                                     DW-921 MS-SNUB-(VR-S62A)                                     DW-921 MS-SNUB-(VR-S62B)                                     DW-921 MS-SNUB-(VR-S63)                                     DW-921 MS-SNUB-(VR-570A)                                     DW-901 MS-SNUB-(VR-S70B)                                     DW-901 MS-SNUB-(VR-S71A)                                     DW-901 MS-SNUB-(VP-S71B)                                     DW-901 MS-SNUB-(VR-S72)                                     DW-901 MS-SNUB-(VR-S73)                                     DW-901 MS-SNUB-(VR-S74)                                     DW-901 MS-SNUB-(VR-S7A)                                     DW-888 MS-SNUB-(VR-S7B)                                     DW-888 MS-SNUB-(VR-S8)                                       DW-888 MS-SNUB-(VR-S80)                                     DW-901
~ Snubber Location MS-SNUB-(VR-S26).
    -MS-SNUB-(VR-S81)                                     DW-901 MS-SNUB-(VR-S82)                                     DW-901 MS-SNUB-(VF-S83A)                                     DW-901 MS-SNUB-(VR-S83B)                                     DW-901 MS-SNUB-(VR-S84)                                     DW-901 MS-SNUB-(VR-585)                                     DW-901 MS-SNUB-(VR-S86A)                                     DW-901 MS-SNUB-(VR-S86B)                                   DW-901 MS-SNUB-(VR-S87A)                                     DW-888 MS-SNUB-(VR-S87B)                                   DW-888 MS-SNUB-(VR-S88)                                     DW-888 RF-SNUB-(RF-S10)                                     DW-921 RF-SNUB-(RF-S11)                                     DW-921 RF-SNUB-(RF-S12)                                     DW-921 RF-SNUB-(RF-S13)                                     DW-921 R F-SNUB-(RF-S 14)                                   DW-921 RF-SNUB-(RF-S15)                                     DW-921 RF-SNUB-(RF-S16)                                     DW-921
DW-888 MS-SNUB-(VR-S27)
                                      -137k-
DW-901
!!S-SNUB-(VR-S3)
DW-888 MS-SNUB-(VR-S30)
DW-921 l
MS-SNUB-(VR-S31)
DW-921 MS-SNUB-(VR-S32)
DW-888 MS-SNUB-(VR-S4)-
DV-901 MS-SNUB-(VR-S40)
DW-921 MS-SNUB-(VR-S41)-
DW-921 MS-SNUB-(VR-S42A)
DW-921 MS-SNUB-(VR-S42B)
DW-921 MS-SNUB-(VR-S43)
DW-888 MS-SNUB-(VR-S50A)
DW-921 MS-SNI'B-(VR-S50B)
DW-921 MS-SNUB-(VR-SSI)
DW-888
!!S-SNUB-(VR-SS A)
DW-901 ItS-SNUB-(V R-SS B)
DW-901 MS-SNUB-(VR-S6)
DW-901 MS-SNUB-(VR-S60)
DP-921 MS-SNUB-(VR-S61)
DW-921 MS-SNUB-(VR-S62A)
DW-921 MS-SNUB-(VR-S62B)
DW-921 MS-SNUB-(VR-S63)
DW-921 MS-SNUB-(VR-570A)
DW-901 MS-SNUB-(VR-S70B)
DW-901 MS-SNUB-(VR-S71A)
DW-901 MS-SNUB-(VP-S71B)
DW-901 MS-SNUB-(VR-S72)
DW-901 MS-SNUB-(VR-S73)
DW-901 MS-SNUB-(VR-S74)
DW-901 MS-SNUB-(VR-S7A)
DW-888 MS-SNUB-(VR-S7B)
DW-888 MS-SNUB-(VR-S8)
DW-888 MS-SNUB-(VR-S80)
DW-901
-MS-SNUB-(VR-S81)
DW-901 MS-SNUB-(VR-S82)
DW-901 MS-SNUB-(VF-S83A)
DW-901 MS-SNUB-(VR-S83B)
DW-901 MS-SNUB-(VR-S84)
DW-901 MS-SNUB-(VR-585)
DW-901 MS-SNUB-(VR-S86A)
DW-901 MS-SNUB-(VR-S86B)
DW-901 MS-SNUB-(VR-S87A)
DW-888 MS-SNUB-(VR-S87B)
DW-888 MS-SNUB-(VR-S88)
DW-888 RF-SNUB-(RF-S10)
DW-921 RF-SNUB-(RF-S11)
DW-921 RF-SNUB-(RF-S12)
DW-921 RF-SNUB-(RF-S13)
DW-921 R F-SNUB-(RF-S 14)
DW-921 RF-SNUB-(RF-S15)
DW-921 RF-SNUB-(RF-S16)
DW-921
-137k-


I T1blo 3.6.3 INACCESSIBLE SAFETY RELATED MECHANICAL SHOCK SUPPRESSORS (SNUBBERS) (cont'd)
I T1blo 3.6.3 INACCESSIBLE SAFETY RELATED MECHANICAL SHOCK SUPPRESSORS (SNUBBERS) (cont'd)
Snubber                                           Location RF-SNUB-(RF-S17)                                     DW-921 RF-SNUB-(RF-S18)                                     DW-921 RF-SNUB-(RF-S19)                                     DW-921 RF-SNUB-(RF-S8)                                     DW-921                   ,
Snubber Location RF-SNUB-(RF-S17)
RF-SNUB-(RF-S9)                                     DW-921                   '
DW-921 RF-SNUB-(RF-S18)
RIIR-SNUB-(RH-S10)                                   DW-901                   I RHR-SNUB-(Ril-Sil)                                   DW-901 RHR-SNUB-(RH-S13)                                   DW-921 RHR-SNUB-(RH-S14)                                   DW-921                   i RHR-SNUB-(RH-SIS)                                   DW-921                   !
DW-921 RF-SNUB-(RF-S19)
RHR-SNUB-(RH-S16)                                   DW-901 RiiR-SNUB-(RH-S17)                                   DW-901                   i RHR-SNUB-(Rif-S18)                                   DW-901                   l RHR-SNUB-(RH-S19)                                   DW-901                   i RHR-SNUB-(RH-53)                                     DW-FLG AREA               }
DW-921 RF-SNUB-(RF-S8)
DW-921 RF-SNUB-(RF-S9)
DW-921 I
RIIR-SNUB-(RH-S10)
DW-901 RHR-SNUB-(Ril-Sil)
DW-901 RHR-SNUB-(RH-S13)
DW-921 RHR-SNUB-(RH-S14)
DW-921 i
RHR-SNUB-(RH-SIS)
DW-921 RHR-SNUB-(RH-S16)
DW-901 RiiR-SNUB-(RH-S17)
DW-901 i
RHR-SNUB-(Rif-S18)
DW-901 l
RHR-SNUB-(RH-S19)
DW-901 i
RHR-SNUB-(RH-53)
DW-FLG AREA
}
RilR-SNUB-(RH-S4)
RilR-SNUB-(RH-S4)
DW-FLG AREA RHR-SNUB-(Ril-SS)                                   DW-921 RHR-SNUB-(RH-S6)                                     DW-921                   '
DW-FLG AREA RHR-SNUB-(Ril-SS)
RHR-SNUB-(RH-S67)                                   DW-901 RHR-SNUB-(Ril-S68)                                   DW-901 RifR-SNUB-(RH-S69A)                                 DW-901 RHR-SNUB-(RH-S69B)                                   DW-901                   ,
DW-921 RHR-SNUB-(RH-S6)
RHR-SNUB-(Ril-S7)                                   DW-921                   .
DW-921 RHR-SNUB-(RH-S67)
RHR-SNUB-(RH-S70)                                   DW-901                   !
DW-901 RHR-SNUB-(Ril-S68)
RHR-SNUB-(Rll-S71)                                   DW-901 RilR-SNUB-(RH-S72)                                   DW-901                   !
DW-901 RifR-SNUB-(RH-S69A)
RilR-SNUB-(RH-S72A)                                 DW-901 RHR-SNUB-(RH-S73)                                   DW-901 RHR-SNUB-(RH-S8A)                                   DW-901 RHR-SNUB-(RH-S8B)                                   DW-901 RHR-SNUB-(RH-S9)                                     DW-901 RR-SNUB-(SS-1A)                                     DW-888 RR-SNUB-(SS-1B)                                     DW-888                   i RR-SNUB-(SS-2A)                                     DW-888 RR-SNUB-(SS-2B)                                     DW-888                   l RR-SNUB-(SS-3A1)                                     DW-901                   ;
DW-901 RHR-SNUB-(RH-S69B)
RR-SNUB-(SS-3A2)                                     DW-901                   i RR-SNUB-(SS-3BI)                                     DW-901                   i RR-SNUB-(SS-3B2)                                     DW-901                   i RR-SNUB-(SS-4A)                                     DW-901                   i RR-SNUB-(SS-4B)                                     DW-901 RR-SNUB-(SS-SA)                                     DW-888 RR-SNUE-(SS-5B)                                     DW-888 RR-SNUB-(SS-8A1)                                     DW-901 RR-SNUB-(SS-8A2)                                     DW-901 RWCU-SNUB-(CU-S3A)                                   DW-921 RWCU-SNUB-(CU-S3B)                                   DW-921
DW-901 RHR-SNUB-(Ril-S7)
                                    -1371-l
DW-921 RHR-SNUB-(RH-S70)
DW-901 RHR-SNUB-(Rll-S71)
DW-901 RilR-SNUB-(RH-S72)
DW-901 RilR-SNUB-(RH-S72A)
DW-901 RHR-SNUB-(RH-S73)
DW-901 RHR-SNUB-(RH-S8A)
DW-901 RHR-SNUB-(RH-S8B)
DW-901 RHR-SNUB-(RH-S9)
DW-901 RR-SNUB-(SS-1A)
DW-888 RR-SNUB-(SS-1B)
DW-888 i
RR-SNUB-(SS-2A)
DW-888 RR-SNUB-(SS-2B)
DW-888 l
RR-SNUB-(SS-3A1)
DW-901 RR-SNUB-(SS-3A2)
DW-901 i
RR-SNUB-(SS-3BI)
DW-901 i
RR-SNUB-(SS-3B2)
DW-901 i
RR-SNUB-(SS-4A)
DW-901 i
RR-SNUB-(SS-4B)
DW-901 RR-SNUB-(SS-SA)
DW-888 RR-SNUE-(SS-5B)
DW-888 RR-SNUB-(SS-8A1)
DW-901 RR-SNUB-(SS-8A2)
DW-901 RWCU-SNUB-(CU-S3A)
DW-921 RWCU-SNUB-(CU-S3B)
DW-921
-1371-l


Table 3.6.4 INACCESSIBLE SAFETY RELATED HYDRAULIC SHOCK SUPPRESSORS (SNUBBERS)
Table 3.6.4 INACCESSIBLE SAFETY RELATED HYDRAULIC SHOCK SUPPRESSORS (SNUBBERS)
Snubber                                             Location RR-SNUB-(SS-7Al)                                       DW-901 RR-SNUB-(SS-7A2)                                       DW-901 RR-SNUB-(SS-7BI)                                       DW-901 RR-SNUB-(SS-7B2)                                       DW-901 1
Snubber Location RR-SNUB-(SS-7Al)
DW-901 RR-SNUB-(SS-7A2)
DW-901 RR-SNUB-(SS-7BI)
DW-901 RR-SNUB-(SS-7B2)
DW-901 1
'i
'i
                                      -137m-L
-137m-L


Attachment 7 Revised Technical Specification for SBGT Testing Requirement Revised Pages:       165 182 183 215 215d 215e References 1) ANSI N510-1980
Revised Technical Specification for SBGT Testing Requirement Revised Pages:
165 182 183 215 215d 215e References 1) ANSI N510-1980
: 2) NRC Inspection Report 50-298/82-02, Item 10.a(2)
: 2) NRC Inspection Report 50-298/82-02, Item 10.a(2)
: 3) Regulatory Guide 1.52 The current Technical Spe.-ifications for Cooper Nuclear Station lists requirements for the testing of charcoal filters in the Standby Gas Treatment System and the Control Room Ventilation System.
: 3) Regulatory Guide 1.52 The current Technical Spe.-ifications for Cooper Nuclear Station lists requirements for the testing of charcoal filters in the Standby Gas Treatment System and the Control Room Ventilation System.
Nebraska Public Power District requests a revision to the Technical Specifications as shown on the attached pages.           This request is made in order to bring testing criteria for filters, as contained in the Technical Specifications, into line with current industry standards and guidance (References 1 and 3). In relation to this request, Messrs. L. Wilborn and             B. Murray conducted a routine inspection (Reference 2) on January 11-15, 1982. Their inspection report also recommended this change in Item 10.a(2).
Nebraska Public Power District requests a revision to the Technical Specifications as shown on the attached pages.
This request is made in order to bring testing criteria for filters, as contained in the Technical Specifications, into line with current industry standards and guidance (References 1 and 3).
In relation to this request, Messrs. L. Wilborn and B. Murray conducted a routine inspection (Reference 2) on January 11-15, 1982.
Their inspection report also recommended this change in Item 10.a(2).
l e
l e
                                                , - , ,  - - - -      - , -  --- .e m---
7
.e m---


1 LkMITINGCONDITIONSFOROPERATION                 SURVEILLANCE REOUIREMENTS 3.7. (cont'd.)~                               4.7 (cont'd.)
1 LkMITINGCONDITIONSFOROPERATION SURVEILLANCE REOUIREMENTS 3.7. (cont'd.)~
B. Standby Gas Treatment System           B. Standby Cas Treatment System
4.7 (cont'd.)
: 1. Except as specified in 3.7.B.3 below,   1. At least ence per operating cycle the both circuits of the standby gas treat-       following conditions shall be demon-ment system and the diesel generators         strated, required for operation of such circuits shall be operable at all times when     a. Pressure drop across the combined HEPA secondary containment integrity is           filters and charcoal adsorber banks is required.                                     less than 6 inches of water at the sys-tem design flow rate.
B.
: b. Inlet heater input is capable of reduc-ing R.H. from 100 to 70% R.H.
Standby Gas Treatment System B.
2.a. The results of the in-place cold DOP     2.a. The tests and sample analysis of Speci-and halogenated hydrocarbon tests at         fication 3.7.B.2 shall be performed at design flows on HEPA filters and char-       least once per year for standby service coal adsorber banks shall show 199%           or after every 720 hours of system oper-DOP removal and 199% halogenated hydro-       ation and following significant painting, carbon removal.                               fire or chemien1 release ir any ventila-tion zone communicating with the system.
Standby Cas Treatment System 1.
: b. The results of laboratory carbon sample b. Cold DOP testing shall be performed after analysis shall show 199% radioactive         cach complete or partial replacement of methyl iodide removal at a velocity           the HEPA filter bank or af ter any struc-within20percengof actual system de-           tural maintenance on the system housing.
Except as specified in 3.7.B.3 below, 1.
At least ence per operating cycle the both circuits of the standby gas treat-following conditions shall be demon-ment system and the diesel generators
: strated, required for operation of such circuits shall be operable at all times when a.
Pressure drop across the combined HEPA secondary containment integrity is filters and charcoal adsorber banks is required.
less than 6 inches of water at the sys-tem design flow rate.
b.
Inlet heater input is capable of reduc-ing R.H. from 100 to 70% R.H.
2.a. The results of the in-place cold DOP 2.a. The tests and sample analysis of Speci-and halogenated hydrocarbon tests at fication 3.7.B.2 shall be performed at design flows on HEPA filters and char-least once per year for standby service coal adsorber banks shall show 199%
or after every 720 hours of system oper-DOP removal and 199% halogenated hydro-ation and following significant painting, carbon removal.
fire or chemien1 release ir any ventila-tion zone communicating with the system.
b.
The results of laboratory carbon sample b.
Cold DOP testing shall be performed after analysis shall show 199% radioactive cach complete or partial replacement of methyl iodide removal at a velocity the HEPA filter bank or af ter any struc-within20percengof actual system de-tural maintenance on the system housing.
sign, 11.75 mg/m inlet methyl iodide concentration, 370% R.H. and
sign, 11.75 mg/m inlet methyl iodide concentration, 370% R.H. and
_30 F.
_30 F.
: c. Fans shall be shown to operate within   c. Halogenated hydrocarbon testing shall be
c.
        +10% design flow.                             performed af ter each complete er partial replacement of the charcoal adsorber bark or af ter any structural maintenance on the system housing.
Fans shall be shown to operate within c.
: d. Each circuit shall be operated with the heaters on at least 10 hours every month.
Halogenated hydrocarbon testing shall be
I                                                 e. Test sealing of gaskets for housing doors I                                                     downstream of the HEPA filters and char-coal adsorbers shall be performed at, i                                                     and in conformance with, each test per-formed for compliance with Specification 4.7.B.2.a and Specification 3.7.B.2.a.
+10% design flow.
: 3. From and after the date that one cir-   3. System drains where present shall be in-cuit of the stanWoy gas treatment sys-       spected quarterly for adequte water 1cvel tem is made or iound to be inoperable         in loop-seals.
performed af ter each complete er partial replacement of the charcoal adsorber bark or af ter any structural maintenance on the system housing.
for any reason, reactor operation or l        fuel handling is permissible only during l       the succeeding seven days unicss such circuit is sooner made operabic, pro-vided that during such seven days all active components of the other standby gas treatment circuit shall be operable, t
d.
                                              -165-
Each circuit shall be operated with the heaters on at least 10 hours every month.
I e.
Test sealing of gaskets for housing doors I
downstream of the HEPA filters and char-coal adsorbers shall be performed at, i
and in conformance with, each test per-formed for compliance with Specification 4.7.B.2.a and Specification 3.7.B.2.a.
3.
From and after the date that one cir-3.
System drains where present shall be in-cuit of the stanWoy gas treatment sys-spected quarterly for adequte water 1cvel tem is made or iound to be inoperable in loop-seals.
for any reason, reactor operation or fuel handling is permissible only during l
l the succeeding seven days unicss such circuit is sooner made operabic, pro-vided that during such seven days all active components of the other standby gas treatment circuit shall be operable,
-165-t


3.7.B & 3.7.C BASES (cont'd)
3.7.B & 3.7.C BASES (cont'd)
High efficiency particulate absolute (HEPA) filters are installed before and after the charcoal adsorbers to minimize potential release of particulates to the environment and to prevent clogging of the iodine adsorbers.
High efficiency particulate absolute (HEPA) filters are installed before and after the charcoal adsorbers to minimize potential release of particulates to the environment and to prevent clogging of the iodine adsorbers.
The charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment. The in-place test results should indicate a system leak tightness of less than I percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates. The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 99 percent for expected accident conditions. If the efficiencies of the HEPA filters and charcoal adsorbers are as specified, the                 l resulting doses will be less than the 10 CFR 100 guidelines for the accidents analyzed. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.
The charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment. The in-place test results should indicate a system leak tightness of less than I percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates. The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 99 percent for expected accident conditions.
Only one of the two standby gas treatment systems is needed to cleanup the reactor building atmosphere upon containment isolation. If one system is found to be inoperable, there is no immediate threat to the containment system performance and reactor operation or refueling operation may continue while repairs are being made. If neither circuit is operable, the plant is brought to a condition where the standby gas treatment system is not required.
If the efficiencies l
4.7.B & 4.7.C     BASES Standby Cas Treatment System and Secondary Containment Initiating reactor building isolation and operation of the standby gas treatment system to maintain at least a 1/4 inch of water vacuum within the secondary containmcat provides an adequate test of the operation of the reactor building isolation valves, leak tightness of the reactor building and performance of the standby gas treatment system. Functionally testing the initiating sensors and associated trip channels demonstrates the capability for automatic actuation. Performing these tests prior to re-fueling will demonstrate secondary containment capability prior to the time the primary containment is opened for refueling. Periodic testing gives sufficient confidence of reactor building integrity and standby gas treat-ment system performance capability.
of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the 10 CFR 100 guidelines for the accidents analyzed. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.
Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. A 7.8 kw heater is capable of maintaining relative humidity below 70%. Heater capacity and pressure drop should be determined at least once per operating cycle to show system performance capability.
Only one of the two standby gas treatment systems is needed to cleanup the reactor building atmosphere upon containment isolation.
The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant shall be performed in accordance with ANSI N510-1980. The test cannisters that are installed with the adsorber trays should be used for the charcoal adsorber efficiency test. Each sample should be at least two inches in diameter and a length equal to the thickness of the bed. If test results are unacceptable, all adsorbent in the system shall be replaced
If one system is found to be inoperable, there is no immediate threat to the containment system performance and reactor operation or refueling operation may continue while repairs are being made.
                                            -182-
If neither circuit is operable, the plant is brought to a condition where the standby gas treatment system is not required.
4.7.B & 4.7.C BASES Standby Cas Treatment System and Secondary Containment Initiating reactor building isolation and operation of the standby gas treatment system to maintain at least a 1/4 inch of water vacuum within the secondary containmcat provides an adequate test of the operation of the reactor building isolation valves, leak tightness of the reactor building and performance of the standby gas treatment system.
Functionally testing the initiating sensors and associated trip channels demonstrates the capability for automatic actuation.
Performing these tests prior to re-fueling will demonstrate secondary containment capability prior to the time the primary containment is opened for refueling. Periodic testing gives sufficient confidence of reactor building integrity and standby gas treat-ment system performance capability.
Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. A 7.8 kw heater is capable of maintaining relative humidity below 70%.
Heater capacity and pressure drop should be determined at least once per operating cycle to show system performance capability.
The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant shall be performed in accordance with ANSI N510-1980. The test cannisters that are installed with the adsorber trays should be used for the charcoal adsorber efficiency test.
Each sample should be at least two inches in diameter and a length equal to the thickness of the bed.
If test results are unacceptable, all adsorbent in the system shall be replaced
-182-


4.7.B & 4.7.C' BASES with an~ adsorbent qualified according to Table 1 of Regulatory Guide 1.52.
4.7.B & 4.7.C' BASES with an~ adsorbent qualified according to Table 1 of Regulatory Guide 1.52.
The replacement tray for the adsorber tray removed for the test should meet the same adsorbent quality. -Tests of the HEPA filters with DOF aerosol shall be performed in accordance to ANSI N510-1980. Any filters found defective
The replacement tray for the adsorber tray removed for the test should meet the same adsorbent quality. -Tests of the HEPA filters with DOF aerosol shall be performed in accordance to ANSI N510-1980. Any filters found defective l
        'shall be replaced with filters qualified pursuant to Regulatory Position       l C.3.d. of Regulatory Guide 1.52.
'shall be replaced with filters qualified pursuant to Regulatory Position C.3.d. of Regulatory Guide 1.52.
All elements of the heater should be. demonstrated to be functional and operable during the test of heater capacity. Operation of the heaters will prevent moisture buildup in the filters and adsorber system.
All elements of the heater should be. demonstrated to be functional and operable during the test of heater capacity. Operation of the heaters will prevent moisture buildup in the filters and adsorber system.
With doors closed and fan in operation, DOP aerosol shall be sprayed externally along the full linear periphery of each respective door to check the gasket seal. Any detection of DOP in the fan exhaust shall be considered an unacceptable test result and the gaskets repaired and test repeated.
With doors closed and fan in operation, DOP aerosol shall be sprayed externally along the full linear periphery of each respective door to check the gasket seal. Any detection of DOP in the fan exhaust shall be considered an unacceptable test result and the gaskets repaired and test repeated.
If system drains are present in the filter /adsorber banks, loop-seals must be used with adequate water level to prevent by-pass Icakage from 4
If system drains are present in the filter /adsorber banks, loop-seals must be used with adequate water level to prevent by-pass Icakage from the banks.
the banks.
4 If significant painting, fire or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for. operational use.
If significant painting, fire or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for. operational use. The determination of significance shall be made by the operator on duty at the time of the incident. Knowledgeable staff members should be consulted prior to making this determination.
The determination of significance shall be made by the operator on duty at the time of the incident. Knowledgeable staff members should be consulted prior to making this determination.
Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system performance capability.     If one standby gas treatment system is inoperable, the other system must be tested daily. This substantiates the availability of the operable system and thus reactor operation or refueling operation can continue for a limited period of time.
Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system performance capability.
3.7.D & 4.7.D   BASES Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary con-tainment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize j       the potential leakage paths from the containment in the event of a loss-of-coolant accident.
If one standby gas treatment system is inoperable, the other system must be tested daily. This substantiates the availability of the operable system and thus reactor operation or refueling operation can continue for a limited period of time.
3.7.D & 4.7.D BASES Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary con-tainment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize j
the potential leakage paths from the containment in the event of a loss-of-coolant accident.
The maximum closure times for the automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.
The maximum closure times for the automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.
These valves are highly reliable, have a low service requirement, and most are normally closed. The initiating sensors and associated trip channels are also checked to demonstrate the capability for automatic isolation.
These valves are highly reliable, have a low service requirement, and most are normally closed. The initiating sensors and associated trip channels are also checked to demonstrate the capability for automatic isolation.
The test interval of once per operating cycle for automatic initiation i
The test interval of once per operating cycle for automatic initiation i
l
l
                                                  -183-
-183-


l LIMITING CONDITIONS FOR OPERATION               SURVEILLANCE REQUIREMENTS 4.12 Additional Safety Related Plant 3.12 Additional Safety Related Plant Capabilities Capabilities Applicability:
l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.12 Additional Safety Related Plant 3.12 Additional Safety Related Plant Capabilities Capabilities Applicability:
Applicability:
Applicability:
Applies to the operating status of the     APPlics to the surveillance require-main control room ventilation system,       ments for the main control room venti-lation system, the reactor building the reactor building closed cooling closed cooling water system and the water system and the service water service water system which are required system.
APPlics to the surveillance require-Applies to the operating status of the ments for the main control room venti-main control room ventilation system, lation system, the reactor building the reactor building closed cooling closed cooling water system and the water system and the service water service water system which are required system.
by the corresponding Limiting conditions for Operation.
by the corresponding Limiting conditions for Operation.
Objective:
Objective:
Objective:
Objective:
To assure the availability of the main     To verify that operability or availa-bility under conditions for which these control room ventilation system, the capabilities are an essential response reactor building closed cooling water to station abnormalities.
To assure the availability of the main To verify that operability or availa-bility under conditions for which these control room ventilation system, the capabilities are an essential response reactor building closed cooling water to station abnormalities.
system and the service water system upon the conditions for which the capability is an essential response to station abnormalities.
system and the service water system upon the conditions for which the capability is an essential response to station abnormalities.
A. Main Control Room Ventilation A. Main Control Roem Ventilation
A.
: 1. Except as specified in Specification
Main Control Roem Ventilation A.
: 1. At least once per operating cycle, the 3.12.A.3 below, the control room air       Pressure drop across the combined HEPA filters and charcoal absorber banks treatment system, the diesel shall be demonstrated te be less than generators required for operation of 6 inches of water at system design flow this system and the nain control room rate.
Main Control Room Ventilation 1.
At least once per operating cycle, the 1.
Except as specified in Specification Pressure drop across the combined HEPA 3.12.A.3 below, the control room air filters and charcoal absorber banks treatment system, the diesel shall be demonstrated te be less than generators required for operation of 6 inches of water at system design flow this system and the nain control room rate.
air radiation monitor shall be oper-able at all times when containment integrity is required.
air radiation monitor shall be oper-able at all times when containment integrity is required.
2.a. The tests and sample analysis of 2.a. The results of the in-place cold DOP Specification 3.12. A.2 shall be performed and halogenated hydrocarbon tests at least once per year for stardby service at design flows on HEPA filters r after every 720 hours of system and charcoal absorber banks shall peration and following significant paint-show > 99% C0P removal and > 99%
2.a. The tests and sample analysis of 2.a. The results of the in-place cold DOP Specification 3.12. A.2 shall be performed and halogenated hydrocarbon tests at least once per year for stardby service at design flows on HEPA filters r after every 720 hours of system and charcoal absorber banks shall show > 99% C0P removal and > 99%
halogenated hydrocarbon removal,            ing, fire r chemical release in any ventilation zone communicating with the system,
peration and following significant paint-ing, fire r chemical release in any halogenated hydrocarbon removal, ventilation zone communicating with the
: b. The results of laboratory carben       b. Cold DOP testing shall be perforned fter each complete or partial replace-sample analysis shall show > 99%
: system, b.
ment of the HEPA filter bank or afcer radioactive methyl iodide removal any structural maintenance on the system at a velocity witgin20%ofsystem
Cold DOP testing shall be perforned b.
                    ~
The results of laboratory carben fter each complete or partial replace-sample analysis shall show > 99%
h using.
ment of the HEPA filter bank or afcer radioactive methyl iodide removal any structural maintenance on the system at a velocity witgin20%ofsystem h using.
~
design, 1.75 mg/m inlet iodide concentration, > 95% R.H. and
design, 1.75 mg/m inlet iodide concentration, > 95% R.H. and
                        ~
<30*F.
        <30*F.
~
: c. Fans shall be shown to operate with-   c. Mal genated hydrocarbon testing shall in + 10% design flow.                       be performed after each complete or partial replacement of the charcoal absorber bank or af ter any structural maintenance on the system housing.
c.
                                              -215-
Mal genated hydrocarbon testing shall c.
Fans shall be shown to operate with-in + 10% design flow.
be performed after each complete or partial replacement of the charcoal absorber bank or af ter any structural maintenance on the system housing.
-215-


3.12 BASES A. Main Control Room Ventilation System The control room ventilation system is designed to filter the control room atmosphere for intake air and/or for recirculation during control room isolation conditions. The system is designed to automatically start upon control room isolation and to maintain the control room pressure to the design positive pressure so that all leakage should be out leakage.
3.12 BASES A.
Main Control Room Ventilation System The control room ventilation system is designed to filter the control room atmosphere for intake air and/or for recirculation during control room isolation conditions. The system is designed to automatically start upon control room isolation and to maintain the control room pressure to the design positive pressure so that all leakage should be out leakage.
High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential intake of radioiodine to the control room. The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates.
High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential intake of radioiodine to the control room. The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates.
The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 99 percent for expected accident conditions. If the efficiencies of the HEPA filters and charcoal       l adsorbers are as specified, the resulting doses will be less than the allowable levels stated in Criterion 19 of the General Design Criteria for Nuclear Power Plants, Appendix A to 10 CFR Part 50. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.
The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 99 percent for expected l
If the system is found to be inoperable, there is not immediate threat to the control room and reactor operation or refueling operation may continue for a limited period of time while repairs are being made. If the system cannot be repaired within seven days, the reactor is shutdown and brought to cold shutdown within 24 hours, or refueling operations are terminated.
accident conditions.
B. Reactor Building Closed Cooling Water System The reactor building closed cooling water system has two pumps and one heat exchanger in each of two loops. Each loop is capable of supplying the cooling requirements of the essential services following design accident conditions with only one pump in either loop.
If the efficiencies of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the allowable levels stated in Criterion 19 of the General Design Criteria for Nuclear Power Plants, Appendix A to 10 CFR Part 50.
Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.
If the system is found to be inoperable, there is not immediate threat to the control room and reactor operation or refueling operation may continue for a limited period of time while repairs are being made.
If the system cannot be repaired within seven days, the reactor is shutdown and brought to cold shutdown within 24 hours, or refueling operations are terminated.
B.
Reactor Building Closed Cooling Water System The reactor building closed cooling water system has two pumps and one heat exchanger in each of two loops. Each loop is capable of supplying the cooling requirements of the essential services following design accident conditions with only one pump in either loop.
The system hes additional flexibility provided by the capability of inter-connection of the two loops and the backup water supply to the critical loop by the service water system. This flexibility and the need for only one pump in one loop to meet the design accident requirements justifies the 30 day repair time during normal operation and the reduced requirements during head-off operations requiring the availability of LPCI or the core spray-systems.
The system hes additional flexibility provided by the capability of inter-connection of the two loops and the backup water supply to the critical loop by the service water system. This flexibility and the need for only one pump in one loop to meet the design accident requirements justifies the 30 day repair time during normal operation and the reduced requirements during head-off operations requiring the availability of LPCI or the core spray-systems.
C. Service Water System The service water system consists of four vertical service water pumps located in the intake structure, and associated strainers, piping, valving and instrumentation. The pumps discharge to a common header from which independent piping supplies two Seismic Class I cooling water loops and one turbine building loop. Automatic valving is provided to shutoft all supply to the turbine building loop on drop in header pressure thus assuring supply to the Seismic Class I loops each of which feeds one diesel generator, two RHR service water booster pumps, one control room basement fan coil unit and one RBCCW
C.
                                              -215d-
Service Water System The service water system consists of four vertical service water pumps located in the intake structure, and associated strainers, piping, valving and instrumentation. The pumps discharge to a common header from which independent piping supplies two Seismic Class I cooling water loops and one turbine building loop. Automatic valving is provided to shutoft all supply to the turbine building loop on drop in header pressure thus assuring supply to the Seismic Class I loops each of which feeds one diesel generator, two RHR service water booster pumps, one control room basement fan coil unit and one RBCCW
-215d-


3.12 BASES (cont'd) heat exchanger. Valves are included in the common discharge header to permit the Seismic Class I service water system to be operated as two independent loops. The heat exchangers are valved such that they can be individually backwashed without interrupting system operation.
3.12 BASES (cont'd) heat exchanger. Valves are included in the common discharge header to permit the Seismic Class I service water system to be operated as two independent loops. The heat exchangers are valved such that they can be individually backwashed without interrupting system operation.
During normal operation two or three pumps will be required. Three pumps are used for a normal shutdown.
During normal operation two or three pumps will be required. Three pumps are used for a normal shutdown.
The loss of all a-c power will trip all operating service water pumps.
The loss of all a-c power will trip all operating service water pumps.
The automatic emergency diesel generator start system and emergency equipment starting sequence vf]1 then start one selected service water pump in 30-40 seconds. In the meantime, the drop in service water header pressure will close the turbine building cooling water isolation valve guaranteeing supply to the reactor building, the control room basement, and the diesel generators from the one service water pump.
The automatic emergency diesel generator start system and emergency equipment starting sequence vf]1 then start one selected service water pump in 30-40 seconds.
In the meantime, the drop in service water header pressure will close the turbine building cooling water isolation valve guaranteeing supply to the reactor building, the control room basement, and the diesel generators from the one service water pump.
Due to the redundance of pumps and the requirement of only one to meet the accident requirements, the 30 day repair time is justified.
Due to the redundance of pumps and the requirement of only one to meet the accident requirements, the 30 day repair time is justified.
D. Battery Room Ventilation The temperature rise and hydrogen buildup in the battery rooms without adequate ventilation is such that continuous safe operation of equipment in these rooms cannot be assured.
D.
4.12 BASES A. Main Control Room Ventilation System Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate wf]1 indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Pressure drop should be determined at least once per operating cycle to show system performance capability.
Battery Room Ventilation The temperature rise and hydrogen buildup in the battery rooms without adequate ventilation is such that continuous safe operation of equipment in these rooms cannot be assured.
4.12 BASES A.
Main Control Room Ventilation System Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate wf]1 indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Pressure drop should be determined at least once per operating cycle to show system performance capability.
Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant should be performed in accordance with ANSI N510-1980.
Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant should be performed in accordance with ANSI N510-1980.
l The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. The test cannisters that are installed with the adsorber trays should be used for the charcoal adsorber efficiency test. Each sample should be at least twc inches in diameter and a length equal to the thickness of the bed. If test results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent qualified according to Table 1 of Regulatory Cuide 1.52.
l The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. The test cannisters that are installed with the adsorber trays should be used for the charcoal adsorber efficiency test.
The rcplacement tray for the absorber tray removed for the test should meet the same adsorbent quality. Tests of the HEPA filters with DOP aerosol shall be performed in accordance to ANSI N510-1980. Any HEPA filters found           l defective shall be replaced with filters qualified pursuant to Regulatory Position C.3.d of Regulatory Guide 1.52.
Each sample should be at least twc inches in diameter and a length equal to the thickness of the bed.
If test results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent qualified according to Table 1 of Regulatory Cuide 1.52.
The rcplacement tray for the absorber tray removed for the test should meet the same adsorbent quality. Tests of the HEPA filters with DOP aerosol shall be performed in accordance to ANSI N510-1980. Any HEPA filters found l
defective shall be replaced with filters qualified pursuant to Regulatory Position C.3.d of Regulatory Guide 1.52.
Operation of the system for 10 hours every month will demonstrate operability of the filters and adsorber system and remove excessive moisture built up on the adsorber.
Operation of the system for 10 hours every month will demonstrate operability of the filters and adsorber system and remove excessive moisture built up on the adsorber.
                                              -215e-
-215e-


4 i
4 i
Attachment 8 Revised Technical Specification for.
Revised Technical Specification for.
:                                    10CFR50 Appendix J Testing Revised Pages:                                           159                       176 160                       177
10CFR50 Appendix J Testing Revised Pages:
:                                                                                                161                       178 162                       184 162a                     Deleted:   15Sa 163                                 178a 164                                   183a 168 173 Reference 1)         Letter from D. G. Eisenhut to J. M. Pilant dated 4
159 176 160 177 161 178 162 184 162a Deleted:
September 3, 1982, " Exemption to Appendix J to 10CFR Part 50 and Safety Evaluation Report" L
15Sa 163 178a 164 183a 168 173 Reference 1)
This change is in response to Reference 1.                                                                 The feedwater check valves will be tested with air or nitrogen as required.
Letter from D. G. Eisenhut to J. M. Pilant dated September 3, 1982, " Exemption to Appendix J to 4
The requested Technical Specifications for the containment air locks are proposed.                       It should be noted that the SER of
10CFR Part 50 and Safety Evaluation Report" L
,          Reference 1 stated that " test performance requires shutting down the reactor and opening the equipment hatch in order to install a
This change is in response to Reference 1.
!            strongback on the inner airlock door                                                           . . ." It is true that the strongback has to be attached to the                                                           inside door of the contain-
The feedwater check valves will be tested with air or nitrogen as required.
!            ment airlock in order to pressurize                                                           the airlock to accident
The requested Technical Specifications for the containment air locks are proposed.
,            pressure (Pa), but the drywell does                                                           not have to be entered to install the strongback because the strongback is stored inside the containment airlock. To attach the strongback during full power operation will expose personnel to radiation levels of approximately 500 mr/hr.                                   Total exposure for this testing is estimated to be approximately 2 rem, which the District feels is
It should be noted that the SER of Reference 1 stated that " test performance requires shutting down the reactor and opening the equipment hatch in order to install a strongback on the inner airlock door..." It is true that the strongback has to be attached to the inside door of the contain-ment airlock in order to pressurize the airlock to accident pressure (Pa), but the drywell does not have to be entered to install the strongback because the strongback is stored inside the containment airlock.
;            excessive. Even though it is not necessary to enter the drywell
To attach the strongback during full power operation will expose personnel to radiation levels of approximately 500 mr/hr.
Total exposure for this testing is estimated to be approximately 2 rem, which the District feels is excessive.
Even though it is not necessary to enter the drywell to do the containment airlock test at Pa, the District still
+
+
to do the containment airlock test at Pa, the District still i            followed the NRC recommendation in the proposed Technical Specification.
followed the NRC recommendation in the proposed Technical i
s The District will calculate a new correlation of reduced pressure leakage rates to full pressure leakage rate for the bellows leakage test and the containment airlock test.                                                                 The calculation I.
Specification.
will be based on the Franklin Research Center Technical j           Evaluation Report, Appendix A, Procedure 'B'. This calculation i           will be added to the appropriate procedure for the local leak rate tests.
The District will calculate a new correlation of reduced pressure s
leakage rates to full pressure leakage rate for the bellows leakage test and the containment airlock test.
The calculation I.
will be based on the Franklin Research Center Technical j
Evaluation Report, Appendix A, Procedure 'B'.
This calculation i
will be added to the appropriate procedure for the local leak rate tests.
I
I
                , . - - . - - - -  ,e - , . . - - . , . . - - - . - . . . , . , _ - - , _ - - , - - - - - ,.,-- ,,-- .
,e


Since resolution of the above three issues concludes the Staff's review of CNS as regards Appendix J, minor forinat changes are being proposed for Section 3/4.7 of the Technical, Specifications which should make this section easier to utilize.
Since resolution of the above three issues concludes the Staff's review of CNS as regards Appendix J, minor forinat changes are being proposed for Section 3/4.7 of the Technical, Specifications which should make this section easier to utilize.
Line 510: Line 1,058:
I
I


LIMITING CONDITIONS FOR OPERATION               SURVEILLANCE REOUIREMENTS 3.7 CONTAINMENT SYSTEMS                         4.7   CONTAINMENT SYSTEMS Applicability:                                 Applicability:
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 4.7 CONTAINMENT SYSTEMS 3.7 CONTAINMENT SYSTEMS Applicability:
Applies to the operating status of             Applies to the primary and secondary the primary and secondary contain-             e ntainment integrity.
Applicability:
Applies to the operating status of Applies to the primary and secondary the primary and secondary contain-e ntainment integrity.
ment systems.
ment systems.
Objective:                                       Objective:
Objective:
To assure the integrity of the pri-             To verify the integrity of the primary mary and secondary containment systems,         and secondary containment.
Objective:
Specification:                                   Specffication:
To assure the integrity of the pri-To verify the integrity of the primary mary and secondary containment systems, and secondary containment.
A. Primary Containment                     A.     Primary Containment Suppression Pool                          1. Suppression Pool l 1.                                                                                             l At any time that the nuclear system         a. The suppression pool water level is pressuris:ed above atmospheric               and temperature shall be checked pressure or work is being done                   nce Per day, which has the potential to drain the vessel, the suppression pool            b. Whenever there is indication of water volume and temperature shall               relief valve operation or testing be maintained within the following              which adds heat to the suppression limits except as specified in                   p   1, the pool temperature shs]1 3.7.A.2. and 3.5.F.5.                           be continual 13 monitored anc also observed and logged every 5
Specification:
: a. Minimum water volume - 87,650 ft 3            minutes until the heat addition is terminated.
Specffication:
3
A.
: b. Maximum water volume - 91,000 f t
Primary Containment A.
: c. Whenever there is indication of
Primary Containment 1.
: c. Maximum suppression pool temperature           relief valve operation with the during normal power operation - 90 F.           temperature of the suppression For 45 days, commencing July 16, 1981,         pool reaching 160 F or more and the suppression pool temperature me"           the primary coolant system pres-be increased to 95 whenever the                 sure greater than 200 psig, an river water temperature is such that           external visual examination of the peol temperature cannot he main-           the suppression chamber shall toined below 90 F.                             be conducted before resuming
Suppression Pool l
: d. During testing which adds heat to               P wer Peration.
l 1.
the suppression pool, the water temperature shall not exceed 10 F           d. A visual inspection of the above the normal power operation                 suppression chatt.ber interior, limit specified in c. above. In                 including water line regions, connection with such testing, the               shall be made at each major pool temperature must he reduced to             refueling outage, below the normal power operation limit specified in c. above within 24 hours.
Suppression Pool At any time that the nuclear system
: e. The reactor shall be scrammed from any operating condition if the pool temperature reaches 110 F. Power operation shall not be resumed until the pool temperature is reduced below the normal power operation limit specified in c.
: a. The suppression pool water level is pressuris:ed above atmospheric and temperature shall be checked pressure or work is being done nce Per day, which has the potential to drain
: b. Whenever there is indication of the vessel, the suppression pool water volume and temperature shall relief valve operation or testing which adds heat to the suppression be maintained within the following limits except as specified in p
1, the pool temperature shs]1 be continual 1 monitored anc also 3
3.7.A.2. and 3.5.F.5.
observed and logged every 5 3
a.
Minimum water volume - 87,650 ft minutes until the heat addition is terminated.
3 b.
Maximum water volume - 91,000 f t
: c. Whenever there is indication of c.
Maximum suppression pool temperature relief valve operation with the during normal power operation - 90 F.
temperature of the suppression For 45 days, commencing July 16, 1981, pool reaching 160 F or more and the suppression pool temperature me" the primary coolant system pres-be increased to 95 whenever the sure greater than 200 psig, an river water temperature is such that external visual examination of the peol temperature cannot he main-the suppression chamber shall toined below 90 F.
be conducted before resuming d.
During testing which adds heat to P wer Peration.
the suppression pool, the water temperature shall not exceed 10 F
: d. A visual inspection of the above the normal power operation suppression chatt.ber interior, limit specified in c. above.
In including water line regions, connection with such testing, the shall be made at each major pool temperature must he reduced to refueling outage, below the normal power operation limit specified in c. above within 24 hours.
e.
The reactor shall be scrammed from any operating condition if the pool temperature reaches 110 F.
Power operation shall not be resumed until the pool temperature is reduced below the normal power operation limit specified in c.
above.
above.
                                                -159-
-159-


LIMITING CONDITIONS FOR OPERATION             SURVEILLANCE REQUIREMENTS 3.7.A.1 (cont'd.)                             4.7.A (cont'd.)
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A.1 (cont'd.)
: f. During reactor isolation conditions,     2. Leak Rate Testing the reactor pressure vessel shall be depressurized to less than 200           a. Integrated leak rate teste (ILRT's) psig at normal cooldown rates if               shall be performed to verify primary the pool temperature reaches 120 F.           containment integrity. Primary con-tainment integrity is confirned if the
4.7.A (cont'd.)
: 2. Containment Integrity                         leakage rate does not exceed the equivalent of 0.635 percent of the Primary containment integrity shall           Primary containment volume per 24 hours at 58 psig, he traintained at all times when the reactor is critical or when the reactor water temperature is above         b. Integrated leak rate tests raay be per-212*F and fuel is in the reactor               f rmed at either 58 psig or 29 psig, the vessel except while performing "open           leakage rate test period, extending to vessel" physics tests at power levels         24 hours of retained internal pressure.
2.
Leak Rate Testing
: f. During reactor isolation conditions, the reactor pressure vessel shall be depressurized to less than 200
: a. Integrated leak rate teste (ILRT's) psig at normal cooldown rates if shall be performed to verify primary the pool temperature reaches 120 F.
containment integrity.
Primary con-tainment integrity is confirned if the 2.
Containment Integrity leakage rate does not exceed the equivalent of 0.635 percent of the Primary containment integrity shall Primary containment volume per 24 hours he traintained at all times when the at 58 psig, reactor is critical or when the reactor water temperature is above
: b. Integrated leak rate tests raay be per-212*F and fuel is in the reactor f rmed at either 58 psig or 29 psig, the vessel except while performing "open leakage rate test period, extending to vessel" physics tests at power levels 24 hours of retained internal pressure.
If it can be demonstrated to the satis-not to exceed 5 MW(t).
If it can be demonstrated to the satis-not to exceed 5 MW(t).
faction of those responsible for the acceptance of the containment structure that the leakage rate can be accurately determined during a shorter test peried, the agreed-upon shorter period may be used.
faction of those responsible for the acceptance of the containment structure that the leakage rate can be accurately determined during a shorter test peried, the agreed-upon shorter period may be used.
Prior to initial operation, integrated leak, rate tests must be performed at 58 and 29 psig (with the 29 psig test being performed prior to the 58 psig test) to establish the allowable leak rate, L (in percent of containment volume per 2k hours) at 29 psig as the lesser of the following values:
Prior to initial operation, integrated leak, rate tests must be performed at 58 and 29 psig (with the 29 psig test being performed prior to the 58 psig test) to establish the allowable leak rate, L (in percent of containment volume per 2k hours) at 29 psig as the lesser of the following values:
(L, is 0.635 percent)
(L, is 0.635 percent)
L = 0.635 I'tm b
L = 0.635 I'tm bam l
am l                                                        for 'tm < 0.7 b
for 'tm < 0.7 bam where Itm = measured ILR at 29 psig Lam = measured ILR at 58 psig, and L
am where Itm = measured ILR at 29 psig L
~ 1.0 tm <
am = measured ILR at 58 psig, and
am L = 0.635 P t
                                                              ~ 1.0 tm <
Pa
L am L = 0.635 P t
-160-
P a
                                              -160-


LIMITING CONDITIONS FOR OPFRATION     SURVEILLANCE RFOUIREMFUTS 4.7.A.2.b. (cont'd.)
LIMITING CONDITIONS FOR OPFRATION SURVEILLANCE RFOUIREMFUTS 4.7.A.2.b. (cont'd.)
l3.7.A(cont'd.)                                                                         .
l3.7.A(cont'd.)
where P, = peak accident pressure, 58 psia P = appropriately mecsured test pres" sures (psia) for 'tm > 0.7 am
where P, = peak accident pressure, 58 psia P = appropriately mecsured test pres" sures (psia) for 'tm > 0.7 am c.
: c. The ILRT's shall be performed at the fc31owing minimum frequency:
The ILRT's shall be performed at the fc31owing minimum frequency:
: 1. Prior to initial urft operation.
1.
: 2. At approximately three and one-third year interva]n re that any ten-year interval would include four ILRT's. These intervals may be extended up to eight months if recessary to coincide with refueling outage.
Prior to initial urft operation.
: d.  {he measured leakage retes, I'{:m and am, p' hall be Jess than 0.75 ,t and 0.75 a for the reduced pressure tests and peak pressure test respectively.
2.
: e. Except for the initial ILRT, all ILRT's shall be performed without any pre-liminary leak detection surveys and leak repaire immediately prior to the test. If an ILRT has to be ter-minated due to excessive leakage through identified leakage paths, the leakage through such paths r. hall be determined by a local leakage test and recorded. After repairs are made another ILRT shall be conducted.
At approximately three and one-third year interva]n re that any ten-year interval would include four ILRT's.
These intervals may be extended up to eight months if recessary to coincide with refueling outage.
{he measured leakage retes, I'{:m and d.
am, p' hall be Jess than 0.75
,t and 0.75 a for the reduced pressure tests and peak pressure test respectively.
e.
Except for the initial ILRT, all ILRT's shall be performed without any pre-liminary leak detection surveys and leak repaire immediately prior to the test.
If an ILRT has to be ter-minated due to excessive leakage through identified leakage paths, the leakage through such paths r. hall be determined by a local leakage test and recorded. After repairs are made another ILRT shall be conducted.
If an ILET is completed but the acceptance criteria of Specificatien 4.7.A.2.d is not satisfied and repairs are necessary, the ILRT need rot he 4
If an ILET is completed but the acceptance criteria of Specificatien 4.7.A.2.d is not satisfied and repairs are necessary, the ILRT need rot he 4
                                          -161-
-161-
                      ~ _
~


LIMITING CONDITIONS FOR OPIRATION SURVEILLANCE REQUIREMENTS
LIMITING CONDITIONS FOR OPIRATION SURVEILLANCE REQUIREMENTS
  '3.7.A (Cont'd)                   4.7.A.2.e (cont'd)                           l repeated provided locally measered leakage reductions, achieved by re-pairs, reduce the containment's overall measured leakage rate suf-ficiently to meet the acceptance criteria.
'3.7.A (Cont'd) 4.7.A.2.e (cont'd) l repeated provided locally measered leakage reductions, achieved by re-pairs, reduce the containment's overall measured leakage rate suf-ficiently to meet the acceptance criteria.
: f. Local Leak Rate Tests
f.
Local Leak Rate Tests
: 1. With the exceptions specified below, local leak rate tests (LLRT's) shall be performed on the primary containment testable penetrations and isolation valves at a pressure of 58 psig during each reactor shut-down for refueling, or other conven-ient intervals, but in no case at intervals greater than two years.
: 1. With the exceptions specified below, local leak rate tests (LLRT's) shall be performed on the primary containment testable penetrations and isolation valves at a pressure of 58 psig during each reactor shut-down for refueling, or other conven-ient intervals, but in no case at intervals greater than two years.
Table 3.7.2 specifies testable penetrations with double 0-ring seals, Table 3.7.3 specifies test-able penetrations with testable bellows, and Table 3.7.4 specifies primary containment testable isolation valves. The test dur-ation of all valves and penetra-tions shall be of sufficient length to determine repeatable results.
Table 3.7.2 specifies testable penetrations with double 0-ring seals, Table 3.7.3 specifies test-able penetrations with testable bellows, and Table 3.7.4 specifies primary containment testable isolation valves. The test dur-ation of all valves and penetra-tions shall be of sufficient length to determine repeatable results.
The total acceptable leakage for all valves and penetrations other than the MSIV's is 0.60 La.
The total acceptable leakage for all valves and penetrations other than the MSIV's is 0.60 La.
: 2. Bolted double-gasket seals (Table 3.7.2) shall be tested af ter each opening and during each reactor shutdown for refueling, or other convenient intervals but in no case at intervals greater than two years.
: 2. Bolted double-gasket seals (Table 3.7.2) shall be tested af ter each opening and during each reactor shutdown for refueling, or other convenient intervals but in no case at intervals greater than two years.
: 3. The main steam isolation valves (MSIV's) shall be tested at a pres-sure of 29 psig. If a total leak-age rate of 11.5 scf/hr for any one MSIV is exceeded, repairs and retest shall be performed to correct the condition. This is an exemption to Appendix J of 10CFR50.
: 3. The main steam isolation valves (MSIV's) shall be tested at a pres-sure of 29 psig.
                                      -162-
If a total leak-age rate of 11.5 scf/hr for any one MSIV is exceeded, repairs and retest shall be performed to correct the condition. This is an exemption to Appendix J of 10CFR50.
-162-


9 l
9 l
LIMITING CONDITIONS FOR OPERATION                                               SURVEILLANCE REQUIREMENTS 3.7.A (Cont'd)                         '
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A (Cont'd) 4.7.A.2.F (cont'd)
4.7.A.2.F (cont'd)                   ~%
~%
                                                ^
: 4. Main steam line and feedwater line
: 4. Main steam line and feedwater line
                                      /                                                               expansion bellows as specified in i
/
Table 3.7.3 shall be tested by pressuririr.g'between the laminations of the belinws I at a pressure of 5 psig.
expansion bellows as specified in
^
i Table 3.7.3 shall be tested by pressuririr.g'between the laminations I
of the belinws at a pressure of 5 psig.
This is an exemption to Appendix J
This is an exemption to Appendix J
                                            ),                                                         of 10CFR50.             _
),
                                                                    ^
of 10CFR50.
                                      ; '.* Y                                              5. The personnel airlock shall be tested at 58 psig at- ' intervals no longer than 'six months. This testing may+be extended to the next
^
                                    ',                                                                refueling outage (not t'o exceed.24 months) provided that there have been no tirlock openings since the last successful test at 58 psig.                   In the event the personnel airlock is not opened between refueling outages, it shall be leak checked at 3 psig at intervals no longer than six months.
: 5. The personnel airlock shall be
Within three days of opening (or avery three days during pericds of frequent optning) when containment integrity is' required, test the .
; '. Y tested at 58 psig at- ' intervals no longer than 'six months. This testing may+be extended to the next refueling outage (not t'o exceed.24 months) provided that there have been no tirlock openings since the last successful test at 58 psig.
In the event the personnel airlock is not opened between refueling outages, it shall be leak checked at 3 psig at intervals no longer than six months.
Within three days of opening (or avery three days during pericds of frequent optning) when containment integrity is' required, test the.
personnel airlock at 3 psig. This is an exemption to Appendix J.of 100FR50.
personnel airlock at 3 psig. This is an exemption to Appendix J.of 100FR50.
o
o g.
: g.             ' Continuous' Leak Rate Monit'or i hen the primary cc' ntainment . is ,
' Continuous' Leak Rate Monit'or I,
I, j      ,I inerted the containment sh'all be '                   .
i hen the primary cc' tainment. is,
e     -
n inerted the containment sh'all be '
j
,I e
continueusly ronitored for grose
continueusly ronitored for grose
                          /                           J'                           .
/
leakage by retiew of the. inerting system makeup r'equirements. This monitoring system may be;taken out of service for *aaintenance but shall b'e returned to service as soon as p"g a c ticab le.
J' leakage by retiew of the. inerting system makeup r'equirements. This monitoring system may be;taken out of service for *aaintenance but shall b'e returned to service as soon as p" a c ticab le.
                                                                                              ,                                                              a .
g a
: h.               Dryuell Surfacen i                                                                     >
h.
j                               The interior surfaces of,the drywell and torus shall be visually inspected each operating cycle for evidence of torus corrosion or Icakage.
Dryuell Surfacen i
                                                                  ,-            w                                                                       ,
j The interior surfaces of,the drywell and torus shall be visually inspected each operating cycle for evidence of torus corrosion or Icakage.
f                                                             ,
w f
                                                                                      'n                                   p
'n p
                                                                                                                                                      ,9 e
,9
                                                                                  'l N 3,                                       V  !'
'l e
                                                                                                                                  ~
N V
1.'\ '
3, 1.'\\ '
~
e y
e y
e .                                       '.          /
e.
                        ,    ;,                                                                          ,e                         <
/
                                                                                                                                              , -s
,e
                ,,              f                                    -162a-                                                     ,
- s f
-162a-


l LIMITING CONDITIONS FOR OPERATION                 SURVEILLANCE REQUIREMENT 5
l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 5
        ]
]
3.7.A (cont'd.)                                   4.7.A (cont'd.)
3.7.A (cont'd.)
i
4.7.A (cont'd.)
: 3. Pressure Suppression Chamber -             3. Pressurc Suppression Chamber -
i 3.
                    ' Reactor Building Vacuum Breakerg               Reactor Building Vacuum Breakers
Pressure Suppression Chamber -
: a.     Except as specified in 3.7.A.3.b           a. The pressure suppression chamber-reactor
3.
              ,  s below, two pressure suppression                   building vacuum breakers and associated chamber-reactor building vacuum                 instrumentation, including set points breakers shall be operable at all               shall be checked for proper operation times when primary containment in-               every three months.
Pressurc Suppression Chamber -
' Reactor Building Vacuum Breakerg Reactor Building Vacuum Breakers a.
Except as specified in 3.7.A.3.b a.
The pressure suppression chamber-reactor s below, two pressure suppression building vacuum breakers and associated chamber-reactor building vacuum instrumentation, including set points breakers shall be operable at all shall be checked for proper operation times when primary containment in-every three months.
tegrity is required. The set point of the differential pressure instru-mentation which actuates the pressure suppresrion chamber-reactor building air actuated vacuum breakers shall be 0.5 psid. The self actuated vacuur breakers shall open fully when subjected to a force equivalent to 0.5 psid actinF on the valve disc.
tegrity is required. The set point of the differential pressure instru-mentation which actuates the pressure suppresrion chamber-reactor building air actuated vacuum breakers shall be 0.5 psid. The self actuated vacuur breakers shall open fully when subjected to a force equivalent to 0.5 psid actinF on the valve disc.
: b.     From and after the date that one of         b. During each refueling outage each the pressure suppression chamber-               vacuum breaker shall be tested to reactor building vacuum breakers is             determine that the force required made'or found to be inoperable for               to open the vacuum breaker does not any reason, the vacuum M eaker switch             exceed the force specified in shall be secured in S.e closed position         Specifications 3.7.A.3.a and each and reactor operation is permissible           vacuum breaker shall be inspected only during the succeeding seven days           and verified to meet design unless such vacuum breaker is sooner             requirements, made operable, provided that the repair procedure does not violate primary containment integrity.
b.
4 .'   Drywell-Pressure Suppression Chamber       4. Drywell-Pressure Suppression Chamber Vacuum Breakers                                 Vacuum Breakers
From and after the date that one of b.
: a. When primary containment is required,       a. Each drywell-suppression chamber vacuum all drywell-suppression chamber vac-             breaker shall be exercised through an
During each refueling outage each the pressure suppression chamber-vacuum breaker shall be tested to reactor building vacuum breakers is determine that the force required made'or found to be inoperable for to open the vacuum breaker does not any reason, the vacuum M eaker switch exceed the force specified in shall be secured in S.e closed position Specifications 3.7.A.3.a and each and reactor operation is permissible vacuum breaker shall be inspected only during the succeeding seven days and verified to meet design unless such vacuum breaker is sooner requirements, made operable, provided that the repair procedure does not violate primary containment integrity.
                    .uum breakers shall be operable at the           opening-closing cycle every 30 days.
4.'
Drywell-Pressure Suppression Chamber 4.
Drywell-Pressure Suppression Chamber Vacuum Breakers Vacuum Breakers a.
Each drywell-suppression chamber vacuum a.
When primary containment is required, all drywell-suppression chamber vac-breaker shall be exercised through an
.uum breakers shall be operable at the opening-closing cycle every 30 days.
0.5 psid setpoint and positioned in the fully closed position as indicated by the position indicating system except during testing and except as specified
0.5 psid setpoint and positioned in the fully closed position as indicated by the position indicating system except during testing and except as specified
                    - in 3.7. A.4.b and .c below.
- in 3.7. A.4.b and.c below.
            -b.     Three drywe]I-suppression chamber           b. When it is determined that a vacuum
-b.
    '                                                                breaker valve is inoperable for opening vacuum breakers may be determined to be inoperabic f'or opening pro-               at a time when operability is required vided they are secured in the fully             all other vacuum breaker valves shall closed position or that the require-             be exercised immediately and every 15 ment of 3.7.A.4.c is demonstrated to             days thereafter until the inoperah3e be met.                                         valve has been returned to normal service.
Three drywe]I-suppression chamber b.
When it is determined that a vacuum vacuum breakers may be determined breaker valve is inoperable for opening to be inoperabic f' r opening pro-at a time when operability is required o
vided they are secured in the fully all other vacuum breaker valves shall closed position or that the require-be exercised immediately and every 15 ment of 3.7.A.4.c is demonstrated to days thereafter until the inoperah3e be met.
valve has been returned to normal service.
m
m
                                                            -163-
-163-


LIMITING CONDITIONS FOR OPERATION               SURVEILLANCE REQUIREMENTS 3.7.A.4 (cont'd.)                               4.7.A.4 (cont'd.)                             l
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A.4 (cont'd.)
: c. The total leakage between the dry-       c. Once each operating cycle, each vacuum well and suppression chamber shall             breaker valve shall be visually in-be less than the equivalent Jeakage           sPected to insure proper maintenance through a 1" dianeter orifice.                 and operation of the position frdication switch. The differential pressure set-point shall be verified.
4.7.A.4 (cont'd.)
: d. If specifications 3.7.A.4.a. b or c,     d. Prior to reactor startup after each cannot be met, the situation shall             refueling, a leak test of the drywell be corrected within 24 hours or the             to suppression chamber structure reactor will be placed in a cold               shall be conducted to demonstrate sht/ S T. condition within the sub-             that the requirerent of 3.7 A.4.c sequent 24 hours.                               I8 **E*
l c.
: 5. Oxygen concentration                      5. Oxygen Concentration
Once each operating cycle, each vacuum c.
: a. After completion of the startup test       a. The primary containment oxygen con-program and demonstration of plant             centration shall be measured and electrical output, the primay con-             recorded at least twice weekly.
The total leakage between the dry-well and suppression chamber shall breaker valve shall be visually in-be less than the equivalent Jeakage sPected to insure proper maintenance through a 1" dianeter orifice.
and operation of the position frdication switch. The differential pressure set-point shall be verified.
d.
Prior to reactor startup after each d.
If specifications 3.7.A.4.a. b or c, cannot be met, the situation shall refueling, a leak test of the drywell be corrected within 24 hours or the to suppression chamber structure shall be conducted to demonstrate reactor will be placed in a cold sht/ S T. condition within the sub-that the requirerent of 3.7 A.4.c sequent 24 hours.
I8 **E*
5.
Oxygen Concentration 5.
Oxygen concentration a.
The primary containment oxygen con-a.
After completion of the startup test centration shall be measured and program and demonstration of plant electrical output, the primay con-recorded at least twice weekly.
tainment atmosphere shall be reduced to less than 4% oxygen with nitrogen gas during reactor power operation with reactor coolant pressure above 100 psig, except as specified in 3.7.A.5.b.
tainment atmosphere shall be reduced to less than 4% oxygen with nitrogen gas during reactor power operation with reactor coolant pressure above 100 psig, except as specified in 3.7.A.5.b.
: b. Within the 24-hour period subsequent       b. The quantity of liquid nitrogen in to placing the reactor in the Run mode         the liquid nitrogen storage tank sha.11 following a shutdown, the containment         be determined twice per week when the atmosphet e oxygen concentration shall         v luce requirements of 3.7.A.5.c are be reduced to less than 4% by volume           I" *ff*C'*
b.
and maintained in this condition.
The quantity of liquid nitrogen in b.
Within the 24-hour period subsequent to placing the reactor in the Run mode the liquid nitrogen storage tank sha.11 following a shutdown, the containment be determined twice per week when the atmosphet e oxygen concentration shall v luce requirements of 3.7.A.5.c are I" *ff*C'*
be reduced to less than 4% by volume and maintained in this condition.
De-inerting may commence 24 hours prior to a shutdown.
De-inerting may commence 24 hours prior to a shutdown.
: c. When the containment atmosphere oxygen concentration is required to be less than 4%, the minimum quantity of liquid i
When the containment atmosphere oxygen c.
nitrogen in the liquid nitrogen storage tank shall be 500 gallons.
concentration is required to be less than 4%, the minimum quantity of liquid nitrogen in the liquid nitrogen storage i
tank shall be 500 gallons.
i l
i l
l   d. If the specifications of 3.7.A 5.a thre e cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition within 24 hours.
l d.
l l   e. The specifications of 3.7.A.5.a thru d are not applicabic during a 48 hour l
If the specifications of 3.7.A 5.a thre e cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition within 24 hours.
l l
e.
The specifications of 3.7.A.5.a thru d are not applicabic during a 48 hour l
continuous period between the dates of March 22, 1982 and March 25, 1982.
continuous period between the dates of March 22, 1982 and March 25, 1982.
                                                -164-
-164-


COOPER NUCLEAR STATION TABLE 3.7.1 (Page 1)
COOPER NUCLEAR STATION TABLE 3.7.1 (Page 1)
PRIMARY CONTAINMENT ISOLATION VALVES Number of Power     Maximum                               Action On Operated Valves     Operating               Normal         Initiating Valve & Steam                           Inboard       Outboard     Time (Sec) (1)         Position (2)   Signal (3)
PRIMARY CONTAINMENT ISOLATION VALVES Number of Power Maximum Action On Operated Valves Operating Normal Initiating Valve & Steam Inboard Outboard Time (Sec) (1)
,      Main Steam Isolation Valves MS-AO A,B,C, & D                                           4                       3<T<5                   0             GC MS-AO A,B,C, & D                                                      4            3<T<5                    0            CC
Position (2)
Signal (3)
Main Steam Isolation Valves MS-AO A,B,C, & D 4
3<T<5 0
GC
]
]
i Drywell Floor Drain Iso. Valves                                 1         1             15                       0             GC RW-AO-82, RW-AC-83 Drywell Equipment Drain                                         1         1             15                       0             GC
MS-AO A,B,C, & D 4
* 8 Iso. Valves RW-AO-94, RW-AO-95
3<T<5 0
  $i Main Steam Line Drain                                           1         1             30                       C             SC l     Valves MS-MO-74, MS-MO-77 I
CC i
Reactor Water Sample Valves                                     1         1             15                       0             GC l
Drywell Floor Drain Iso. Valves 1
RRV-740AV, RRV-741AV                                                                                                                       l-l l     Reactor Water Cleanup System                                   1         1             60                       0             GC'
1 15 0
!      Iso. Valves RWCU-MO-15, RWCU-MO-18 l
GC RW-AO-82, RW-AC-83 Drywell Equipment Drain 1
2 RHR Reactor Head Spray                                         1         1             60                       C             SC l     Iso. Valves RHR-MO-32. RHR-MO-33 RHR Suction Cooling Iso.                                       1         1             40                       C             SC
1 15 0
:      Valve RHR-MO-17. RHR-MO-18
GC 8
!      RHR Discharge to Radwaste                                       1         1             20                       C             SC Iso. Valves RHR-MO-57, RHR-MO-67 l
Iso. Valves RW-AO-94, RW-AO-95 i
1 Suppression Chamber Purge &                                               2             15                       C             SC Vent PC-245AV, PC-230MV Suppression Chamber N     2 Supply                                       2             15                       C             SC PC-237AV, PC-233MV                                                                                                                           !
Main Steam Line Drain 1
i
1 30 C
SC l
Valves MS-MO-74, MS-MO-77 I
I Reactor Water Sample Valves 1
1 15 0
GC l-l RRV-740AV, RRV-741AV l
l Reactor Water Cleanup System 1
1 60 0
GC' t
Iso. Valves RWCU-MO-15, RWCU-MO-18 l
2 RHR Reactor Head Spray 1
1 60 C
SC l
Iso. Valves RHR-MO-32. RHR-MO-33 I
RHR Suction Cooling Iso.
1 1
40 C
SC Valve RHR-MO-17. RHR-MO-18 RHR Discharge to Radwaste 1
1 20 C
SC l
Iso. Valves RHR-MO-57, RHR-MO-67 y
1 Suppression Chamber Purge &
2 15 C
SC I
Vent PC-245AV, PC-230MV Suppression Chamber N Supply 2
15 C
SC I
2 PC-237AV, PC-233MV i


TABLE 3.7.4 PRIMARY CONTAINMENT TESTABLE ISOLATION VALVES TEST PEN. NO.                     VALVE NUMBERS                                   MEDIA X-7A     MS-AO-80A and MS-AO-86A, Main Steam Isolation Valves                 Air X-7B     MS-AO-80B and MS-AO-86B, Main Steam Isolation Valves                 Air X-7C     MS-AO-80C and MS-AO-86C, Main Steam Isolation Valves                 Air X-7D     MS-AO-80D and MS-AO-86D, Main Steam Isolation valves                 Air X-8       MS-M0-74 and MS-MO-77, Main Steam Line Drain                         Air X-9A     RF-15CV and RF-16CV, Feedwater Check Valves                           Air X-9A     RCIC-AO-22. RCIC-MO-17, and RWCU-15CV, RCIC/RWCU Connection to Feedwater                                               Air X-9B     RF-13CV and RF-14CV, Feedwater Check Valves                           Air X-9B     HPCI-AO-18 and HPCI-MO-57 HPCI Connection to Feedwater               Air X-10     RCIC-MO-15 and RCIC-MO-16, RCIC Steam Line                           Air X-11     HPCI-MO-15 and HPCI-M0-16, RPCI Steam Line                           Air X-12     RHR-MO-17 and RHR-MO-18, RHR Suction Cooling                         Air X-13A     RHR-MO-25A and RHR-M0-27A, RHR Supply to RPV                         Air X-13B     RHR-MO-25B and RHR-M0-27B, RHR Supply to RPV                         Air X-14     RWCU-MO-15 and RWCU-M0-18. Inlet to RWCU System                     Air X-16A     CS-MO-11A and CS-MO-12A, Core Spray to RPV                           Air X-16B     CS-M0-11B and CS-M0-12B, Core Spray to RPV                           Air
TABLE 3.7.4 PRIMARY CONTAINMENT TESTABLE ISOLATION VALVES TEST PEN. NO.
,      X-17     RHR-MO-32 and RHR-M0-33, RPV Head Spray                             Air X-18     RW-732AV and RW-733AV, Crywell Equipment Sump Discharge             Air X-19     RW-765AV and RV ,'66AV, Drywell Floor Drain Sump Discharge           Air
VALVE NUMBERS MEDIA X-7A MS-AO-80A and MS-AO-86A, Main Steam Isolation Valves Air X-7B MS-AO-80B and MS-AO-86B, Main Steam Isolation Valves Air X-7C MS-AO-80C and MS-AO-86C, Main Steam Isolation Valves Air X-7D MS-AO-80D and MS-AO-86D, Main Steam Isolation valves Air X-8 MS-M0-74 and MS-MO-77, Main Steam Line Drain Air X-9A RF-15CV and RF-16CV, Feedwater Check Valves Air X-9A RCIC-AO-22. RCIC-MO-17, and RWCU-15CV, RCIC/RWCU Connection to Feedwater Air X-9B RF-13CV and RF-14CV, Feedwater Check Valves Air X-9B HPCI-AO-18 and HPCI-MO-57 HPCI Connection to Feedwater Air X-10 RCIC-MO-15 and RCIC-MO-16, RCIC Steam Line Air X-11 HPCI-MO-15 and HPCI-M0-16, RPCI Steam Line Air X-12 RHR-MO-17 and RHR-MO-18, RHR Suction Cooling Air X-13A RHR-MO-25A and RHR-M0-27A, RHR Supply to RPV Air X-13B RHR-MO-25B and RHR-M0-27B, RHR Supply to RPV Air X-14 RWCU-MO-15 and RWCU-M0-18. Inlet to RWCU System Air X-16A CS-MO-11A and CS-MO-12A, Core Spray to RPV Air X-16B CS-M0-11B and CS-M0-12B, Core Spray to RPV Air X-17 RHR-MO-32 and RHR-M0-33, RPV Head Spray Air X-18 RW-732AV and RW-733AV, Crywell Equipment Sump Discharge Air X-19 RW-765AV and RV,'66AV, Drywell Floor Drain Sump Discharge Air X-25 PC-232MV and PC-238AV, Purge and Vent Supply to Drywell Air X-25 ACAD-1305MV and ACAD-1306MV, supply to Drywell Air X-26 PC-231MV and PC-246AV, Purge and Vent Exhaust -
    . X-25     PC-232MV and PC-238AV, Purge and Vent Supply to Drywell             Air X-25     ACAD-1305MV and ACAD-1306MV, supply to Drywell                       Air X-26     PC-231MV and PC-246AV, Purge and Vent Exhaust -             nrywell Air X-26     ACAD-1310MV, Bleed f rom Drywell                                     Air I
nrywell Air X-26 ACAD-1310MV, Bleed f rom Drywell Air I
f
f
                                              -173-
-173-


3.7.A & 4.7.A   BASES Prinary Containment The integrity of the primary containment and operation of the core standby cooling system in combination, limit the off-site doses to values less than those-suggested in 10CFR100 in the event of a break in the primary system piping. Thus, centainment integrity is specified whenever the potentia]
3.7.A & 4.7.A BASES Prinary Containment The integrity of the primary containment and operation of the core standby cooling system in combination, limit the off-site doses to values less than those-suggested in 10CFR100 in the event of a break in the primary system piping. Thus, centainment integrity is specified whenever the potentia]
for violation of the primary reactor system integrity exists. Concern about such a violation exists whenever the reactor is critical and above atmos-pheric pressure. An exception is made to this requirement during initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required. There will be no pressure on the system at this time, thus greatly reducing the chances of a pipe break. The reactor nay be taken critical during this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring. Procedures and the Rod Worth Minimizer would limit control worth such that a rod drop would not rceult in any fuel damage. In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment system, which shall be oper-ational during this time, offer a sufficient barrier to keep off-site doses well below 10CFR10G limits.
for violation of the primary reactor system integrity exists. Concern about such a violation exists whenever the reactor is critical and above atmos-pheric pressure. An exception is made to this requirement during initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required. There will be no pressure on the system at this time, thus greatly reducing the chances of a pipe break. The reactor nay be taken critical during this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring. Procedures and the Rod Worth Minimizer would limit control worth such that a rod drop would not rceult in any fuel damage.
In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment system, which shall be oper-ational during this time, offer a sufficient barrier to keep off-site doses well below 10CFR10G limits.
The pressure suppression pool water provides the heat sink for the reactor primary system energy release fo]Iowing a postulated rupture of the system.
The pressure suppression pool water provides the heat sink for the reactor primary system energy release fo]Iowing a postulated rupture of the system.
The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1035 psig. Since all of the gases in the drywell are purged into the precoure suppression chamber air space during a loss-of-coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber (water and air) was obtained by con-sidering that the total volume of reactor coolant to be condensed is diccharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.
The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1035 psig.
Since all of the gases in the drywell are purged into the precoure suppression chamber air space during a loss-of-coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber (water and air) was obtained by con-sidering that the total volume of reactor coolant to be condensed is diccharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.
Using the mininum or maximum water volumes given in the specification, con-tainment pressurc during the design basis accident isapproximately38psjg which is below the maximum of 62 psig. Maximum water volume of 91,000 ft regults in a downcomer submergence of 5' and the minimum volume of 87,650 ft results in a submergence approximately 12 inches less. The majority of the Eodega tests were run with a submerged length of 4 feet and with complete condensation. Thus, with respect to downconer submergence, this specification is adequate. The maximum temperature at the end of blowdown tested during the Humbolt Bay and Bodega Bay tests was 170*F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures ebove 170*F.
Using the mininum or maximum water volumes given in the specification, con-tainment pressurc during the design basis accident isapproximately38psjg which is below the maximum of 62 psig. Maximum water volume of 91,000 ft regults in a downcomer submergence of 5' and the minimum volume of 87,650 ft results in a submergence approximately 12 inches less. The majority of the Eodega tests were run with a submerged length of 4 feet and with complete condensation. Thus, with respect to downconer submergence, this specification is adequate. The maximum temperature at the end of blowdown tested during the Humbolt Bay and Bodega Bay tests was 170*F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures ebove 170*F.
Sheuld it be necessary to drai- he suppression chamber, this should only
Sheuld it be necessary to drai-he suppression chamber, this should only
                                                -176-                                   l
-176-l


b.7.A&4.7.A BASES (cont'd) be done when there is r.o requirement for core standby cooling systems opera-bility as explained in bases 3.5 F.
b.7.A&4.7.A BASES (cont'd) be done when there is r.o requirement for core standby cooling systems opera-bility as explained in bases 3.5 F.
Experimental data indicates that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 160*F during any period of relief valve operation with sonic conditions at the discharge exit.
Experimental data indicates that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 160*F during any period of relief valve operation with sonic conditions at the discharge exit.
Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regine of potentially high suppression chamber loadings.
Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regine of potentially high suppression chamber loadings.
In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open. This action would include: (1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor chutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separatcd from that of the stuck-open relief valve to assure nixing and uniformity of energy insertion to the pool.
In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open. This action would include:
(1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor chutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separatcd from that of the stuck-open relief valve to assure nixing and uniformity of energy insertion to the pool.
Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally change very slowly and monitoring these para-meters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be continually monitored and frequently Icgged during periods of significant heat addition, the temperature trends will be closely fo)) owed so that appropriate action can be taken. The requirement for an external visual examination following any event where potentially high loadings could occur providos assurance that no significant damage was encountered.
Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally change very slowly and monitoring these para-meters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be continually monitored and frequently Icgged during periods of significant heat addition, the temperature trends will be closely fo)) owed so that appropriate action can be taken. The requirement for an external visual examination following any event where potentially high loadings could occur providos assurance that no significant damage was encountered.
Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to he the points of highest stress.
Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to he the points of highest stress.
Inerting Safety Guide 7 assumptions for Metal-k'ater reaction result in hydrogen concentration in excess of the Safety Guide 7 flammability limit. By keeping the oxygen concentration less than 4% by volume the requirements of Safety l         Guide 7 are satisfied.
Inerting Safety Guide 7 assumptions for Metal-k'ater reaction result in hydrogen concentration in excess of the Safety Guide 7 flammability limit. By keeping the oxygen concentration less than 4% by volume the requirements of Safety l
i The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is much more probable than the occurrence of the loss-ef-coolant accident upon which the specified oFygen concentration limit is based. permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety offered without significantly reducing the margin of safety. Thus, to preclude the possibility l         of starting the reactor and operating for extended periods of time with significant leaks in the primary system, leak inspections are scheduled l         during periods when the primary system is at or near rat (c' rpc ratits ter..p-l         erature and pressure. The 24-hour period to provide inerting is judged to bc sufficient to perform the leak inspection and establish the required oxygen I         concentration.
Guide 7 are satisfied.
i The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is much more probable than the occurrence of the loss-ef-coolant accident upon which the specified oFygen concentration limit is based. permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety offered without significantly reducing the margin of safety. Thus, to preclude the possibility l
of starting the reactor and operating for extended periods of time with significant leaks in the primary system, leak inspections are scheduled l
during periods when the primary system is at or near rat (c' rpc ratits ter..p-l erature and pressure. The 24-hour period to provide inerting is judged to bc sufficient to perform the leak inspection and establish the required oxygen I
concentration.
l
l
!                                                -177-
-177-


3.7.A & 4.7.A BASES (cont'd)
3.7.A & 4.7.A BASES (cont'd)
The primary containment is normally slightly pressurized during periods of reactor operation. Nitrogen used for inerting could leak out of the contain-ment but air could not-leak in to increase oxygen concentration. Once the
The primary containment is normally slightly pressurized during periods of reactor operation. Nitrogen used for inerting could leak out of the contain-ment but air could not-leak in to increase oxygen concentration. Once the containment is filled with nitrogen to the required concentration, no moni-toring of oxygen concentration is necessary. However, at least twice a week the oxygen concentration will be determined as added assurance.
  ,                  containment is filled with nitrogen to the required concentration, no moni-toring of oxygen concentration is necessary. However, at least twice a week the oxygen concentration will be determined as added assurance.
The 500 gallon conservative limit on the nitrogen storage tank assures that adequate time is available to get the tank refilled assuming normal plant operation. The estimated maximum makeup rate is 1500 SCFD which would require about 160 gallons for a 10 day makeup requirement. The normal leak rate should be about 200 SCFD.
The 500 gallon conservative limit on the nitrogen storage tank assures that adequate time is available to get the tank refilled assuming normal plant operation. The estimated maximum makeup rate is 1500 SCFD which would require about 160 gallons for a 10 day makeup requirement. The normal leak rate should be about 200 SCFD.
Vacuum Relief The purpose of the vacuum relief valves is to equalize the pressure between the J
Vacuum Relief The purpose of the vacuum relief valves is to equalize the pressure between the J
                                                                  -178-                               l
-178-l


  '3.7.D & 4.7.D BASES (cont'd) results in a failure probability of 1.1 x 10-     that a line will not isolate.
'3.7.D & 4.7.D BASES (cont'd) results in a failure probability of 1.1 x 10-that a line will not isolate.
More frequent testing for valve operability results in a greater assurance that the valve will be operable when needed.
More frequent testing for valve operability results in a greater assurance that the valve will be operable when needed.
In order to assure that the doses that may result from a steam line break do not exceed the 10CFR100 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steam line isolation valves. Analyses indicate that fuel rod cladding perforations would be avoided for main steam valve closure times, including instrument delay, as long as 10.5 seconds.
In order to assure that the doses that may result from a steam line break do not exceed the 10CFR100 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steam line isolation valves. Analyses indicate that fuel rod cladding perforations would be avoided for main steam valve closure times, including instrument delay, as long as 10.5 seconds.
The primary containment is penetrated by several small diameter instrument lines connected to the reactor coolant system. Each instrument line contains a 0.25 inch restricting orifice inside the primary containment and an excess flow check valve outside the primary containment. A program for periodic testing and examination of the excess flow check valves is performed as follows:
The primary containment is penetrated by several small diameter instrument lines connected to the reactor coolant system. Each instrument line contains a 0.25 inch restricting orifice inside the primary containment and an excess flow check valve outside the primary containment. A program for periodic testing and examination of the excess flow check valves is performed as follows:
: 1.     Vessel at pressure sufficient to actuate valves. This could be at time of vessel hydro following a refueling outage.
1.
: 2.     Isolate sensing line from its instrument at the instrument manifold.
Vessel at pressure sufficient to actuate valves. This could be at time of vessel hydro following a refueling outage.
: 3. Provide means for observing and collecting the instrument drain or vent valve flow.
2.
: 4. Open vent or drain valve.
Isolate sensing line from its instrument at the instrument manifold.
: a. Observe flow cessation and any leakage rate.
3.
: b. Reset valve after test completion.
Provide means for observing and collecting the instrument drain or vent valve flow.
: 5. The head seal leak detection line cannot be tested in this manner. This valve will not be exposed to primary system pressure except under unlikely conditions of seal failure where it could be partially pressurized to reactor pressure. Any leakage path is restricted at the source and there-fore this valve need not be tested. This valve is in a sensing line that is not safety related.
4.
: 6. Valves will be accepted if a m'arked decrease in flow rate is observed and the leakage rate is acceptable.
Open vent or drain valve.
3.7.E       Bases In conjunction with the Mark I Containment Short Term Program, a plant unique analysis was performed as described in the licensee's i
Observe flow cessation and any leakage rate.
a.
b.
Reset valve after test completion.
5.
The head seal leak detection line cannot be tested in this manner. This valve will not be exposed to primary system pressure except under unlikely conditions of seal failure where it could be partially pressurized to reactor pressure. Any leakage path is restricted at the source and there-fore this valve need not be tested. This valve is in a sensing line that is not safety related.
6.
Valves will be accepted if a m'arked decrease in flow rate is observed and the leakage rate is acceptable.
3.7.E Bases In conjunction with the Mark I Containment Short Term Program, a plant unique analysis was performed as described in the licensee's i
letter of October 4,1976, which demonstrated a factor of safety of at least two for the weakest element in the suppression chamber support system and attached piping. The maintenance of drywell-suppression chamber differential pressure of 1.0 psid and a suppression chamber water level corresponding to a downcomer submergence range of three to four feet will assure the integrity of the suppression chamber when subjected to post-LOCA suppression pool hydrodynamic forces.
letter of October 4,1976, which demonstrated a factor of safety of at least two for the weakest element in the suppression chamber support system and attached piping. The maintenance of drywell-suppression chamber differential pressure of 1.0 psid and a suppression chamber water level corresponding to a downcomer submergence range of three to four feet will assure the integrity of the suppression chamber when subjected to post-LOCA suppression pool hydrodynamic forces.
                                                  -184-
-184-
_-}}
_}}

Latest revision as of 19:12, 20 December 2024

Proposed Changes to Tech Specs Assuring Continuous Operable Water Supply for Fire Fighting Sys from Two Fire Pumps
ML20028B697
Person / Time
Site: Cooper Entergy icon.png
Issue date: 11/24/1982
From:
NEBRASKA PUBLIC POWER DISTRICT
To:
Shared Package
ML20028B696 List:
References
TAC-49198, NUDOCS 8212030162
Download: ML20028B697 (52)
v • d • e


Text

s-3 Revised Technical Specifications for Fire Protection System Clean Water Supply Revised Pages:

216b' 216c 216d 216k Reference 1)

Letter from J. M. Pilant to D. B. Vassallo dated June 28, 1982, " Fire Protection Rule 10CFR50, Appendix R" As discussed in Sections 1.4 and 7.0 of Reference 1, the District is providing a Clean Water Fire Protection System for CNS which upgrades the existing system that takes suction from the Missouri River.

This change is not an NRC requirement but is being performed with direction from the CNS insurance company.

The electric and diesel fire pumps will be separate and independent in the modified system and the requirements of and Branch Technical Position 9.5-1 10CFR50 Appendix R3 Appendix A will be met.

1 8212030162 821124 PDR ADOCK 05000298 P

PDR

e LIMITING CONDITIONS FOR OPERATION

' SURVEILLANCE REQUIREMENTS 3.14 FIRE DETFCTION SYSTEM 4.14 FIRE DETECTION SYSTEM APPLICABILITY APPLICABILITY Applies to the operational status of the Applies to the operational status of the Fire Detection System.

Fire Detection System.

OBJECTIVE To assure continuous automatic surveillance throughout the Main Plant.

SPECIFICATIONS SPECIFICATIONS A.

The Fire Detection System instumen-A.

Each detector on Table 3.14 shall be tation for each fire detection zone demonstrated operabic every 6 months shown in Table 3.14 shall be operable.

by performance of a channel functional test.

B.

With one or more of the fire detection B.

The NFPA Code 72.D Class P. supervised instrument (s) shown in Table 3.14 circuits supervision associated with inoperabic:

the detector alarms of each of the above required fire detection

1. L'ithin I hour establish a fire instruments shall be demonstrated watch patrol to inspect the OPERABLE at least once per 6 months.

zone (s) with the inoperable instru-ment (s) at Icast once per hour, and

2. Restore the inoperable instrument (s) to OPERABLE status within 14 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.7.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plar.s and schedule for re-storing the instrument (s) to OPERABLE status.

3.15 FIRE SUPPRESSION WATER SYSTEM 4.15 FIRE SUPPRESSION WATER SYSTEM APPLICABILITY APPLICABILITY Applies to the availability of water for Applies to the availability of water fire fighting purposes.

for fire fightiag purposes.

OBJ ECTIVE To assure a continuous operable water supply for fire fighting systems from 2 fire pumps.

-216b-

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTE 4.15 (cont'd) 3.15 (cont'd)

SPECIFICATIONS SPECIFICATIONS A.

The Fire Suppression Water Supply A.

The fire suppression water system shall be OPERABLE with:

System shall be demonstrated operable:

1.

At least once per 31 days by 1.

Two fire pumps, each with a capacity g

of at least 2000 gpm, with their starting each pump on a stag-discharge aligned to the fire gered start-up basis and operating it for:

suppression header.

a) A minimum of 15 minutes 2.

An OPERABLE flow path capable of taking suction from either of two f r a diesel engine-driven 500,000 gallon water storage tanks fire pump, and or the Missouri River and b) A minimum of 7 minutes for transferring the water through an electrical motor-driven distribution piping with OPERABLE sectionalizing control or isolation fire pump.

valves to the yard hydrant valves 2.

At least once per 31 days by and the front valve ahead of the water flow alarm device on each verifying that each valve sprinkler, hose standpipe or spray (manual, power operated or automatic) in the flow path system riser.

that is not locked, sealed or B.

If the requirement of 3.15.A cannot be therwise secured in position, is in its correct position.

met, restore the inoperable equipnent to OPERABLE status within 7 days or 3.

At least once per 12 months by prepare and submit a Special Report to cycling each testable valve in the Commission pursuant to Specifica-the flow path through at least tion 6.7.2 within the next 30 days m'e e mplete cycle of full outlining the plans and procedures to

travel, be used to provide for the loss of redundancy in this system.

4.

At least once per 18 months by performing a system functional C.

With the fire suppression system in-test whfch includes simulated operable:

automatic actuation of the system throuF out its operating h

1.

Establish a backup fire suppression sequence, and-water system wfthin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and a) Verifying that each auto-2.

Submit a Special Report in accordance matic valve in the flow path with Specification 6.7.2; actuates to its correct p siti n n a test signal, a) By telephone wfthin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and b) Verifying that each pump b) In writing no later than the developes at least 2000 gpm with first working day following the at least 110 psi, event, outlining the action taken, the cause of the inoperability and the plans and schedule for restor-ing the system to OPERABLE status.

-216c-

O LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.15 (cont'd) c)

Cycling each valve in the flow path that is not testable during pinnt operation through at least one complete cycle or full travel, and d)

Verifying that each high pres-sure pump starts-(sequentially) to maintain the fire suppres-sion water system pressure

> 65 psig.

5.

At least once per 3 years by performing a flow test of the.

system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Editien, published by the National Fire Protection Association.

6.

The fire pump diesel engine shall be demonstrated OPERABLE:

a)

At least once per 31 days by verifying; 1)

The fuel storage tank l,p}

contains at least 150*

gallons of fuel, and 4 2)

The diesel starts from ambient conditions and operates for at least 15 l

minutes.

b)

At least once per 92 days by verifying that a sample of i

diesel fuel from the fuel storage tank, obtained in accordance with ASTM-D270-65, l

is within the acceptable j

limits specified in Table 1 of ASTM-D975-74 for viscosity water content and sediment.

c)-

At least once per 18 months I

by:

1)

Subjecting the diesel to an inspection in accordance with l

procedures prepared in con-junction with its manufactur-er's recommendations for the class of service, and

  • This number shall become 250 gallons when the clean water fire protection system becomes operable.

l

-216d-

INSTRUMENT LOCATION INSTRUMENT ID NO.

-2 Control Room FP-SD-17-1 FP-SD-17-2 4

FP-SD-17-3 l

3 Cable Spreading Room.

FP-SD-16-1 FP-SD-16-2 FP-SD-16-3 FP-SD-16-4 FP-SD-16-5' FP-SD-16-6 Cable Expansion Room FP-SD-16-7 FP-SD-16-8 4

Switchgear. Rooms DC Switchgear Rooms FP-SD-15-2 FP-SD-15-3 Critical Switchgear Room FP-SD-22-1 FP-SD-22-2 5

Station Battery Rooms FP-SD-15-1 FP-SD-15-4 FP-SD-15-1A FP-SD-15-4A 6

Diesel Generator Rooms FP-SD-10-1

~

i FP-SD-10-2 FP-SD-10-3 FP-SD-10-4 CO2-SD-DG-1A CO2-SD-DG-1B CO2-SD-DG-lc CO2-SD-DC-lD CO2-SD-DG-2A CO2-SD-DG-2B CO2-SD-DG-2C CO2-SD-DG-2D 7

Diesel Fuel Storage Rooms CO2-TD-DG-1A CO2-TD-DG-1B 8

Safety Related Equipment not in Reactor Building i.

RHR Service Water Booster Pumps FP-SD-14-3 Emergency Condensate Storage Tanks FP-SD-14-1 Service Water Pumps FP-FD-32-1 l

FP-FD-32-2 9

Auxiliary Relay Room & Reactor Protection System Rooms Auxiliary Relay Room FP-SD-15-9

-Reactor Protection System Room 1A FP-SD-15-7 Reactor Protection System Room IB FP-SD-15-8

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Revised Technical Specifications for Scram Discharge Volume Modifications Revised Pages:

28 33 40 61 The original 12" Scram Discharge Instrument Volume (SDIV) was set to initiate a scram at a level in the volume corresponding to

< 3G gallons.

The 36 gallons was based on an availability consideration giving the operator 20 minutes to respond to an inadvertently closed drain valve assuming each control rod leaked 50 cc/ minute.

There are now two instrument volumes of approximately 22 gallons each, one for each group of hydraulic control units in the reactor building.

Each group has approximately one-half of the hydraulic control units.

The new instrument volumes initiate alarms, rod blocks, and scrams at specified levels rather than volumes.

A level transmitter or level switch measures level rather than volume.

The surveillance program to provide functional checks of the SDIV level instrumentation is provided.

Station procedures provide for periodic verification of the correlation between level and volume.

An SDV not drained alarm has been established at

< 114 inches.

The references for all levels are the center lines of the lower instrument tap on each SDIV.

The scram level for each instrument volume assures an adequate scram discharge volume exists so that all control rods can insert fully.

It should be noted that CNS now has larger scram discharge volumes (excluding the instrument volumes) than existed before this modification.

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COOPER EUCLFAR STATION TABLE 3.1.1-REACTOR PROTECTION SYSTEM INSTRUMENTATION REQUIREMENTS Minimum Number Action Requir.

Applicability Conditions of Operable When Equipmen Reactor Protection Mode Switch Position Trip Level Channels Per Operability i.

System Trip Function Shutdown Startup Refuel Run Setting Trip Systems (1) Not Assured (

Mode Switch in Shutdown X(7)

X X

X 1

A Manual Scram X(7)

X X

X 1-A IRM (17)

X(7)

X X

(5) 1 120/125 of in-3 A

High Flux dicated scale Inoperative X

X (5) 3 A

APRM (17)

High Flux (Flow biased)

X 1 (0.66W+54%)

FRP 2

A or C (14)

MFLPD High Flux X(7)

X(9)

X(9)

(16) 1 15% Rated Power A or C Inoperative X(9)

X(9)

X (13) 2 A or C Downscale (11)

X(12)

> 2.5% of indi-2

~ A or C cated scale High Reactor Pressure X(9)

X(10)

X 1 1045 psig 2

A NBI-PS-55 A,B,C, & D High Drywell Pressure X(9)(8) X(8)

X 1 2 psig 2

A or D PC-PS-12 A,B,C, & D Reactor Low Water Level X

X X

> + 12.5 in. indi-2 A or D NBI-LIS-101 A,B,C, & D cated level Scram Discharge Volume X

X(2)

X 1 92 inches 3

A High Water Level CRD-LS-231 A & B CRD-LS-234 A & B CRD-LT-231 C & D CRD-LT-234 C & D

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COOPER NUCLFAR STATION TABLC 4.1.1 (Page 2)

REACTOR PROTECTION SYSTEM (SCRAM INSTRUMENTATION) FUNCTIONAL TESTS.

MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTR. AND CONTROL CIRCUITS Instrument Channel Group (2)

Functional Test Minimum Frequency (3)

High Water Level in Scram Discharge A

Trip Channel and Alarm Once/3 Months Volume CRD-LS-231 A & B CRD-LS-234 A & B CRD-LT-231 C & D CRD-LT-234 C & D Main Steam Line liigh Radiation B

Trip Channel and Alarm (4)

Once/ Week RMP-RM-251 A,B,C, & D Main Steam Line Isolation Valve A

Trip Channel and Alarm Once/ Month (1)

,g Closure MS-LMS-86 A,B,C, & D i

MS-LMS-80 A,B,C, & D Turbine Control Valve Fast Closure A

Trip Channel and Alarm Once/ Month (1)

IGF-63/0PC -1,2,3,4 Turbine First Stage Pressure A

Trip Channel and Alarm Once/3 Months Permissive MS-PS-14 A,B,C, & D Turbine Stop Valve Closure A

Trip Channel and Alarm Once/ Month (1)

SVOS-1 (1), SVOS-1 (2)

SVOS-2 (1), SVOS-2 (2)

Reactor Pressure Permissive A

Trip Channel and Alarm Once/3 Months NBI-PS-51 A B.C & D

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.1 BASES (cont'd.)

3.1 BASES (cont'd.)

against short reactor periods in revealed only on test. Therefore, it is necessary to test them periodi-these ranges.

cally.

The control rod drive scram system is designed so that all of the water A study was conducted of the instru-mentation channels included in the which is discharged from the reactor by a scram can he accommodated in the Group (B) devices to calculate their

" unsafe" failure rates. The analog discharge piping. The scram discharge volume accommodates in excess of 36 devices (sensors and amp 11ifers) gallons of water and is the low point are predicted to have an unsafe

-6 fc1]ure rate of less than 20 X 10 in the piping.

No credit was taken failures / hour. The bi-stable trip for this volume in the design of the discharge piping as concerns circuits are predicted to have an the amount of water which must be unsafe _failurerateof3cssthan 2 X 10 failures / hour. Consider-accommodated during a scram.

ing the two hour monitoring interva]

During normal operation the dis-f r the analog devices as assumed charge volume is enpty; however, bove, and a weekly test interval should it fill with water, the water f r the hi-stable trip circuits, discharged to the piping from the the design reliability goal of 0.99999 is attained with ample margin.

reactor could not be accommodated which would result in slow scram times or The bi-stabic devices are monitored partial control rod insertion. To pre-clude this occurrence, diverse indi-during plant operation to record their cation (two level switches and two failure history and establish a test level transmitters for each discharge interval using the curve of Figure 4.1.1.

There are numerous identical volume) has been provided in the instrument volumes which alarm and bi-stable devices used throughout scram the reactor when the volume of the plant 's instrunentation systen.

water reaches 92 inches. As indicated Therefore, significant dota on the failure rates for the bi-stable devices above, there is sufficient volume in the piping to acco nmodate the scram should be accumulated rapidly.

without impairment of the scram times or amount of insertion of the control The frequency of calibration of the rodo. This function shuts the reactor APPP Flow Biasing Network has been established as each refueling out-down while sufficient volume remains age. The flow bicsing network is to accommodate the discharged water functionally tested at least once and precludes the situation in which Per month and, in addition, cross a scram would be required but not be calibration checks of the flow abic to perform its function adequately, input to the flow biasing network can be made during the functional A source range monitor (SRM) system is also provided to supply additional test by direct meter reading. There are several instruments which must neutron level information during start-be calibrated and it will take sev-up but has no scram functions (refer-eral days to perform the calibration ence paragraph VII.S.4 FSAR). Thus, f the entire network. While the the IRM and APRM are required in the

" Refuel" and " Start / Hot Standby" modes.

calibra+1on is being performed, a In the power range the APRM system provides required protection (refer-m._.~ _ _. _. _

TABLE 3.2.C CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION Minimum Number of Function Trip Level Setting Operable Instrument Channels / Trip System (5)

APRM Upscale (Flow Bias) ji (0.66W + 427) FRP (2) 2(1)

AFRP Upscale (Stertup)

< 12%

_ MFLPD 2(1)

APRM Downscale (9) 3,2.5%

2(1) i APRM Inoperative (10b) 2(1)

FEM Upscale (Flow Bias) j[ (0.66W + 40%) (2) 1 RBM Downscale (9) 2; 2.5%

1 RBM Inoperative (10c) 1 i

IRM Upscale (8) ji 108/125 of Full Scale 3(1)

IEK Downscale (3)(8) 2;2.5%

3(1)

IPM Detector Not Full In (8) 3(1).

1 IRM Inoperative (8)

(10a) 3(1) 5

{

SRM Upscale (8)

E 1 x 10 counts /Second 1(1)(6) 7

}

SRP Detector Not Full In (4)(8)

(3 100 cps) 1(1)(6)

SRM Inoperative (8)

(10a) 1(1)(6)

Flow Bias Comparator j[10% Difference In Recirc. Flows 1

Flow Bias Upscale /Inop.

l 110% Recirc. Flow 1

SRM Downscale (8)(7) 3,3 Counts /Second (11) 1(1)(6)

SDV Water Level High

[46 inches 1(12)_

CRD-231E, 234E

Revised Technical Specification for HPS Power Monitoring System Revised Pages:

193 195 197 199 Reference 1)

Letter from D. B. Vassallo to J. M. Pilant dated May 4, 1982, " Reactor Protection System (RPS)

Power Monitoring System Design Modification" 2)

Letter from D. B. Vassallo to J. M. P11 ant dated July 8,1982, same subject During the Spring 1982 refueling outage, eight Class IE Electrical Protection Assemblics (EPA's) were installed in the RPS power monitoring system.

This change was in response to the concern that the original RPS was not seismically qualified and could degrade during a seismic event.

The proposed Technical Specifications are in accordance with General Electric verified time delays as required in Reference 1 and the model Technical Specification of Reference 2.

Please note that exception is taken to the Model Technical Specification surveillance requirement of a

" channel functional" test every six months.

This would require deenergizing each half of the RPS system either during the test or transfer to the alternate supply.

This puts unwanted transients on critical equipment (especially the Main Steam Line Radiation Monitors) and induces an unnecessary risk of a plant scram.

The 18-month test frequency proposed for the functional test and channel calibration is consistent with other Technical Specifications for electrical breakers in essential systems.

LIMITING CONDITIONS FOR OPERATION SURVETLLANCE REQUIREMENTS 4.9 AUXILIARY ELECTl:1 CAL SYSTEM 3.9 AUXILIARY ELECTRICA1. SYS'lEM Applicability:

Applicability:

Applies to the periodic testing Applies to the auxiliary electrical requirements of the auxiliary power system.

electrical systems.

Objective:

Objective:

Verify the operability of the auxiliary To assure an adequate supply of elec-electrical system.

trical power for operation of those systems required for safety.

Specification:

Specification:

A.

ary & Meal huhment A.

Auxiliary Electrical Equipment 1.

mergency Buses Undervoltage The reactor shall not be made criti-Relays cal-from a Cold Shutdown Condition unless all of the following condi-a.

Loss of voltage relays tions are satisfied:

Once every 18 months, loss 1.

Both off-site sources (345 KV and f vo tage on emergency 69 KV) and the startup transformer buses is simulated to and emergency transformer are avail-demonstrate the load shed-able and capable of automatically ding fr m emergency buses supplying power to the 4160 Volt and the automatic start emergency buses IF and IG.

of diesel generators.

2.

Both diesel generators shall be b.

Undervoltage relays operable and there shall be a mini-mum of 45,000 gal. of diesel fuel in Once every 18 months, low the fuel oil storage tanks.

voltage on emergency buses is simulated to demonstrate 3.

The 4160V critical buses IF and IG disconnection of the emer-and the 480V critical buses IF and IG gency buses from the offsite are tnergized, power source. The under-v Itage relays shall be a.

The loss of voltage relays and calibrated once every 18 their auxiliary relays are m nds, operable.

2.

Diesel Generators b.

The undervoltage relays and their auxiliary relays are Each diesel-generator shall be started a.

operable, manually and loaded to not less than 35% of rated load for no less than 2 4

The four unit 125V/250V batteries and hours once each month to demonstrate their chargers shall be operable.

operational readiness.

5.

.The power monitoring system for the inservice RPS MG set or alternate source shall be operable.

-193-j.

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTF i

3.9.A 4.9.A (cont'd.)

cell and overall battery voltage shall be measured and logged, b.

Every three conths the measurements shall be mede of the voltage of each cell to nearest 0.1 Volt, specific gravity of each cell, and tenperature of every sixth cell. These measure-ments shall be logged.

c.

Once each operating cycle, the stated batteries shall be subjected to a rated load discharge test.

The specific gravity and voltage of each cell shall be determined after the discharge and logged.

B. Operation with Inoperable Equipment 4.

Power Monitoring System for RPS System Whenever the reactor is in Run Mode or The above specified RPS power monitor-Startup Mode with the reactor not in a ing system instrunentation shall be Cold Condition, the availability of deternined operable:

electric power shall be as specified in 3.9.A, except as specified in 3.9.B.1.

a.

At least once per operating cycle by demonstratirg the operability

1. From and after the date incoming power of over-voltage, under-voltage is not availabic from a startup or emer-and under-frequency protective gency transformer, continued reactor instrumentation by performance of operation is permissible under this a channel calibration including condition for seven days. At the end simulated autematic actuation of of this period, provided the second the protective relays, tripping source of incoming power has not been logic and output circuit breakers made immediately available, the NRC and verifying the following set-must be notified of the event and the points.

plan to restore this second source.

During this period, the two diesel gencr-1.

Over-voltage < 132 VAC, with ators and associated critical buses must tine delay < 2 sec.

be deaanstrated to be operable.

2.

Under-voltage > 108 VAC with

2. From and after the date that incoming time delay < 2 sec.

power is not available from both start-up and emergency transformers, continued 3.

Undcr-frequency > 57 Hz. with operation is permissible, provided the time delay < 2 sec.

two diesel generators and associated critical buses are demonstrated to be

-195-

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS l3.9.B.5(cont'd.)

4.9.B From and after the date that one of the 125 or 250 volt hattery systems is made or found to be inoperable for any reason, continued reactor operation is permissible during the succeeding ten days within electrical safety considera-tions, provided repair work is initiated in the most expeditious manner to return the failed component to an operable state, and Specifications 3.5.A.5 and 3.5.F are satisfied. The NRC shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the situation, the precautions to be taken during this period and the plans to return the failed components to an operable state.

6.

With one RPS. electric power monitoring channel for an inservice RPS MG set or alternate power supply inoperable, restore the inoperable channel to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or remove the associated RPS MG set or alternate power supply from service.

7.

With both RPS electric power monitoring channels for an inservice RPS MG set or alternate power supply inoperable, restore at least one to operable status within 30 minutes or remove the associated RPS MG set or alternate power supply from service.

-197-

4.9 BASES The monthly test of the diesel generator is_ conducted to check for equipment failures and deterioration. Testing is conducted up to equilibrium operating

'i conditions to demonstrate proper. operation at these conditions. The diesel generator will be manually started, synchronf red and connected to the bus i

and load picked up.

The diesel generator should be' loaded to at least 35T of rated load to prevent fouling of the engine.

It le expected that the-diesel generator will be run for at least two hours. Diesel generator experience d

'at other generating. stations indicates that the testing frequency.is adequate

-and provides a_high. reliability-of operation should the system be required.

Each diesel' generator has two air compressors and two air receivers for i

starting.

It.f s expected that the air compressors wf]1 run only infrequently.

Duringlthe monthly check of the diesel generator, each receiver in each' set-3 of receivers will be drawn down below;the point at which the corresponding compressor eutomatically starts to check operation and the ability of the compressors to recharge the receivers.

The diesel generator fuel consumption rate at full load is approximately 275-gallons per hour. Thus, the monthly load test of the diesel generators will test the operation and the ability of the. fuel oil transfer pumps to refill l

the dey tank and will check the operation of these pumps from the energency i

source.

i The rest of the diesel generator during the refueling outage will be core l

comprehensive in that it will functionally test the system; i.e, it will

~

check diesel generator starting and closure of diesel generator breaker and sequencing of load on the diesel generator. The diesel generator will be started by simulatien.of a loss-of-coolant accident.

In addition, en i

underveltage condition will be imposed to simulate a loss of of f-site power.

Periodic tests between refueling outages verify the ability 'of the diesel generator to run at full load and the core and containment cooling pumps to deliver full flow. Periodic testing of the various_ components, plus a func-tional test once-a-cycle, is sufficient to maintain adequate' reliability.

Although station batteries will deteriorate with time, utility experience indicates there is almost no possibility of precipitous failure. The type of surveillance described in this specification is that which has been demonstrated over the years to provide an indication of a cell becoming irregular or unserviceable long before it becomes a failure. _In addition, the checks described also provide adequate indication that the batteries have the speci-fied ampere-hour capability.'

The diesel fuel oil quality must be checked to ensure proper operation of the diesel generators. Water content should be mininized because water in the fuel could centribute to excessive demage to the diesel engine.

When it is determined that some auxiliary electrical equipment is out of service, the increased surveillance required in Section 4.5.F is deemed adequate to provide assurance that the remaining cauipment will be operable.

The Reactor Protection Systen (RPS) is equipped with a seismically qualified, Class 1E power monitoring system. This system consists of eight Electrical Protection Assemblies (EPA) which isolate the power sources from the RPS if the input voltage and frequency are not within limits specified for safe system operation.

Isolation of RPS power causes that RPS division'to_ fail safe.

-RR9-

Revised Technical Specification for Plant Staff Working Hours Revised Page:

226 Generic Letter 82-12 dated June 15, 1982, stated:

"Our letter of February 8, 1982, requested that you take action as necessary to revise the administrative section of your technical specifications to assure that your plant administrative procedures follow the revised working hour guidelines, including a provision for documentation of authorized deviations which should be availabic for NRC review.

You should review your past actions to assure that they are consistent with the attached revised policy statement.

Note that the revised guidelines are to be incorporated by October 1,1982."

In discussions with the Staff, the District was directed to revise the Technical Specifications to state that working hours will be controlled in accordance with a CNS Station Operating Procedure.

This proposed change is attached.

u j.a

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6.3 Statien Optreting Procedures E

^

6. 3.1 -

Station personnel shall be provided detailed written procedures to bey '

usedforoperationandmaintenanceofsystem'componentsandhysdems that could have an effect on nuclear safety.j

.f [

4, x,

6.3.2 Written integrated and system procedures and instructions including applicable check off lists shall be provided and adhered to for the following:

'A.

Normal startup, operation, shutdown and fue handling operadio's n

of the station including all systems and c'omponents involving nuclear safety.

s t

B.

Actions to be taken to correct specific and forseen potential or actual malfunctions of safety related systems or. components including responses to alarms, primary system Jgaks and abnormal reactivity changes.

t C.

Emergency conditions involving possible or actual releases of radio-active materials.

y D.

Implementing procedures of the Sectrrity Plan and the Emergency Plan.

E.

Implementing procedures for che fire protection program.

\\

F.

Administrative procedures for shift overtime.

6.3.3 The following maintenance and test procedures.will be provided to satisfy routine inspection, preventive maintenance programs, and operating ' license requirements.

i A.

Routine testing of Engineered Safeguards and eqbipment a, required by the facility License and the Technical Specifications.

B.

Routine testing of standby and redundant equipment.

Preventiveorcorrectivemaintenanceofplantequipmentandsysthms C.

y that could have an effect on nuclear safety.

D.

Calibration and preventive maintenance of instrumentation that could,

af fect the nuclear safety oflh'e plant.

l E.

Special testing of equipment for proposed changes to operatEonal procedures or proposed system design changes.

6.3.4 Radiation control procedures shall be naintained and made available to all station personnel. These procedares shall(show permissible radiation exposure, and shall be consistent with the ' requirements of 10 CFR 20.

i.

Y.

I

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i

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./

I

-226-

{

(

Revised Technical Specification for SRAB Duties Revised Page:

222 In the recent Amendment 80 to the License,

in Specification 6.2.1.B, the word " approved" was changed to

" approve" as an apparent typographical error.

As a lecent I&E inspection report pointed out, this minor alteration actually changed the duties which SRAB must perform.

The word

" approve" is being changed back to " approved" because it is now clear that this word was not a typographical error as previously thought.

1 i

i i, ^

P

5.2 (cont'd) tary material reviewed; copies of the minutes shall be for-warded to the Chairman of the NPPD Safety Review and Audit Board and the Division Manager of Power Operations within one month.

7.

Procedures:

Written administrative procedures for Committee operation shall be prepared and maintained describing the method for submission and content of presentations to the committee, provisions for use of subcommittees, review and approval by members of written Committee evaluations and recommendations, dissemination of minutes, and such other matters as may be appropriate.

B.

NPPD Safety Review and Audit Board.

The board must: verify that operation of the plant is consistent with company policy and rules, approved operating procedures and l

operating license provisions; review safety related plant changes, proposed tests and procedures; verify that unusual events are promptly investigated and corrected in a manner which reduces the probability of recurrence of such events; and detect trends which may not be apparent to a day-to-day observer.

Audits of selected aspects of plant operation shall be performed with a frequency commensurate with their safety significance and in such i

a manner as to assure that an audit of all nuclear safety related activities is completed within a period of two years.

Periodic review of the audit programs should be performed by the Board at least twice a year to assure that such audits are being accomplished in accordance with requirements of Technical Specifications. The audits shall be performed in accordance with appropriate written instructions or procedures and should include verification of compliance with inter-nal rules, procedures (for example, normal, of f-normal, emergency, op-erating, maintenance, surveillance, test and radiation control proce-dures and the emergency and security plans), regulations involving nuclear safety and operating license provisions; training, qualification and performance of operating staff; and corrective actions following abnormal occurrences or unusual events. A representative portion of procedures and records of the activities performed during the audit period shall be audited and, in addition, observations of perfor-mance of operating and maintenance activities shall be included.

Written reports of such audits shall be reviewed at a scheduled meeting of the Board and by appropriate members of management including those having responsibility in the area audited.

Follow-up action, including reaudit of deficient areas, shall be taken when indicated.

In addition to the above, the Safety Review and Audit Board will audit the facility fire protection and its implementing procedures at least once every 24 months.

-222-

Revised Technical Specification for Listing of Snubbers Revised Pages:

137a

.137b 137e 137f-137m The current Technical Specifications for Cooper Nuclear Station lists snubbers under three different categories on Tables 3.6.1, 3.6.2., and 3.6.3.

Nebraska Public Power District requests a revision to the Technical Specifications as shown on the attached pages.

In addition to revising the existing tables, this request will add a new table, Table 3.6.4, which lists Inaccessible Safety Related Hydraulic Shock Suppressors (Snubbers).

This request is made for reasons as follows:

(a)

In order to add a new category (table).

(b) To have a listing that is compatible with what is contained in the District's computer system and to facilitate data withdrawals and entries.

(c) Some snubber listings have been added/ deleted as a result of modifications to Terus Attached Piping.

(d) Locations given in the tables were not entirely accurate and needed to be more specific.

This change will provide revised tables that are functionally superior to the existing tables.

T

+-

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.H Shock Suppressors (Snubbers) 4.6.H Shock Suppressors (Snubbers) 1.

During all modes of operation The following surveillance require-except Cold Shutdown and Refuel, ments apply to all snubbers listed all safety related snubbers shall in Tables 3.6.1, 3.6.2, 3.6.3, and be operable except as noted in 3.6.4.

l 3.6.H.2 through 3.6.H.5 below.

1.

All snubbers shall be visually 2.

The snubbers listed in Tables inspected in accordance with 3.6.1, 3.6.2, 3.6.3, and 3.6.4 the following schedule:

are required to protect the primary coolant system or other safety Number of Snubbers Next Required related systems or components.

Found Inoperable Inspection All others are therefore exempt During Inspection Interval from these specifications, or During Inspection Interval 3.

With one or more snubbers in-operable, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> re-0 18 months f; 25%

place or restore the inoper-1 12 months f; 25%

able snubber (s) to OPERABLE 2

6 months + 25%

status and perform an engi-3, 4 124 days

[25%

neering evaluation per 5,6,7 62 days f;25%

Specification 4.6.H.4 on 8 or more 31 days j; 25%

the supported component or declare the supported system or The required inspection interval subsystem inoperable and follow shall not be lengthened more the appropriate ACTION state-than one step at a time.

ment for that system.

Snubbers may be categorized in 4.

If a snubber is determined to be groups, " accessible" or "inac-inoperable while the reactor is cessible" based on their acces-in the shutdown or refuel mode, sibility for inspection during the snubber shall be made oper-reactor operation and by type, able or replaced prior to reactor hydraulic or mechanical. These

^

startup.

four groups may be inspected independently according to the 5.

Snubbers may be added to, removed, above schedule.

or substituted for, by analysis, from safety related systems with-2.

Visual Inspection Acceptance out prior License Amendment to Criteria Tables 3.6.1, 3.6.2, 3.6.3, and l

3.6.4, provided that a revision Visual inspections shall verify l

to these tables is included with (1) that there are no visible-l a subsequent License Amendment indications of damage or impair-request.

ed OPERABILITY, (2) attachments to the foundation or supporting l

l

-137a-

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 4.6.H Shock Suppressors (Snubbers)

(cont'd) structure are secure.

Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE ior the purpose of establishing the next visual inspection interval, providing that (1) the cause of the rejection is clearly estab-lished and remedied for that particular snubber and for other snubbers that may be generically susceptible; or (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE per Specifi-cations 4.6 H.6 or 4.6.H.7 as applicable. However, when the fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be determined in-operable and cannot be determined OPERABLE via functional testing for the purpose of establishing the next visual inspection inter-val. All snubbers connected to an inoperable common hydraulic fluid reservoir shall be counted as inoperabic snubbers.

3.

At least once per-18 months dur-

.ing shutdown, a representative sample, 10% of the total of each type of snubber in use in the plant, shall be functionally tested either in place or in a bench test.

For each snubber that does not meet the func-tional test acceptance criteria of Specification 4.6.H.5 or 4.6.H.6, an additional 10% of that type of snubber shall be function-ally tested.

4.

The representative sample select-ed for functional testing shall include various configuration, operating environments and the range of size and capacity of snubbers. Tables 3.6.1, 3.6.2, 3.6.3, and 3.6.4 may be used l

jointly or separately as the basis for the sampling plan.

-137b-

' LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 4.6.H Shock Suppressors (Snubbers)

(cont'd)

Concurrent with the first in-service visual inspection and at least once per 18 months thereafter, the installation and maintenance records of each snubber listed in Tables 3.6.1, 3.6.2, 3.6.3, and-3.6.4 l

shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review.

If the indicated ser-vice life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be reevaluated or the snubber shall be replaced or recondi-tioned so as to extend its service life beyond the date of the next scheduled service life review. This reevaluation, replacement or reconditioning shall be indicated in the records.

'e

-137e-

Table 3.6.1 ACCESSIBLE SAFETY RELATED HYDRAULIC SHOCK SUPPRESSORS (SNUBBERS)

Snubber Location CS-SNUB-(CS-SI)

R-903-SE CS-SNUB-(CS-SIO)

R-931-NE CS-SNUB-(CS-Sil)

R-931-NE CS-SNUB-(CS-S2)

R-903-SE CS-SNUB-(CS-S3)

R-931-SE CS-SNUB-(CS-S6)

R-881-SE QUAD CS-SNUB-(CS-S7)

R-881-SE QUAD CS-SNUB-(CS-VE7)

R-881-SE QUAD HPCI-SNUB-(HP-Sil)

R-859-HPCI RM HPCI-SNUB-(HP-SIS)

R-859-HPCI RM llPCI-SNUB-(HP-S18)

R-859-SW QUAD HPCI-SNUB-(HP-S18A)

R-859-HPCI RM HPCI-SNUB-(HP-S22A)

R-859-llPCI RM llPCI-SNUB-(HP-S4)

R-859-SW QUAD HPCI-SNUB-(RF-S3)

R-859-IIPCI RM HPCI-SNUB-(RF-S4)

R-881-SW TORUS HPCI-SNUB-(RF-SS)

R-881-SW TORUS MS-SNUB-(BS-SI)

R-881-SW TORUS MS-SNUB-(BS-Sil3A)

R-881-NW TORUS MS-SNUB-(BS-Sil3B)

R-881-NW TORUS MS-SNUB-(BS-Sll6A)

R-881-NW TORUS MS-SNUB-(BS-5116B)

R-881-NW TORUS MS-SNUB-(BS-S125A)

R-881-SW TORUS MS-SNUB-(BS-S125B)

R-881-SW TORUS MS-SNUB-(BS-S2)

R-881-SW TORUS MS-SNUB-(BS-S3)

R-881-NW TORUS MS-SNUB-(BS-S4)

R-903-A RPR llX RM MS-SNUB-(BS-SS)

R-903-A RHR llX RM MS-SNUB-(MS-SI)

R-859-HPCI RM MS-SNUB-(MS-SIO)

R-881-SW TORUS MS-SNUB-(MS-Sil)

R-881-SW TORUS MS-SNUB-(MS-SillA)

R-903-A RHR HX RM MS-SNUB-(MS-SilA)

R-881-SW TORUS MS-SNUB-(MS-S12)

R-881-SW TORUS MS-SNUB-(MS-S12A)

R-881-SW TORUS MS-SNUB-(MS-S13)

R-903-B RHR 11X RM MS-SNUB-(MS-S13A)

R-903-B RHR HX FM MS-SNUB-(MS-S13B)

R-903-B RHR HX RM MS-SNUB-(MS-S14)

R-903-B RHR HX RM MS-SNUB-(MS-SIS)

R-931-B RHR HX RM PS-SNUB-(MS-SI5A)

R-931-B RilR HX RM MS-SNUB-(MS-S16A)

R-881-NW TORUS MS-SNUB-(MS-S16B)

R-881-NW TORUS MS-SNUB-(MS-S17)

R-903-A RilR llX RM MS-SNUB-(MS-S18)

R-903-A RilR HX RM MS-SNUB-(MS-S19)

R-903-A RHR llX RM MS-SNUB-(MS-S2)

R-859-ilPCI RM MS-SNUB-(MS-S20)

R-931-A RHR HX FM MS-SNUB-(MS-S20A)

R-931-A RHR HX RM MS-SNUB-(MS-S23)

R-881-NE TORUS MS-SNUB-(MS-S24)

R-881-NE TORUS MS-SNUB-(MS-S25)

R-859-NE QUAD MS-SNUB-(MS-S26)

R-859-NE QUAD

-137f-

Table 3.6.1 ACCESSIBLE SAFETY RELATED HYDRAULIC SHOCK SUPPRESSORS (SNUBBERS) (cont'd)

Snubber Location MS-SNUB-(MS-S3)

R-859-HPCI RM HS-SNUB-(MS-S4)

R-859-HPCI RM MS-SNUB-(MS-S75)

R-931-A RilR HX RM MS-SNUB-(MS-S76)

R-931-B RHR HX RM MS-SNUB-(MS-S7A)

R-859-UPCI RM MS-SNUB-(HS-S7B)

R-859-HPCI RM MS-SNUB-(MS-S8)

R-881-SW TORUS RCIC-SNUB-(RF-SI)

R-881-NE QUAD RCIC-SNUB-(RF-SIA)

R-881-NE QUAD RCIC-SNUB-(RF-S45C)

R-881-NE QUAD RCIC-SNUB-(RF-S45D)

R-881-NE QUAD RCIC-SNUB-(RF-S46A)

R-881-NE QUAD RCIC-SNUB-(RF-SSIA)

R-881-NE TORUS RCIC-SNUB-(RF-SSIB)

R-881-NE TORUS REC-SNUB-(RCC-S20)

R-931-NW REC-SNUB-(RCC-S21)

R-931-hv REC-SNUB-(RCC-S22)

R-931-SW I

RFC-SNUB-(RCC-S3)

R-931-NE REC-SNUB-(RCC-S4)

R-931-NE RF-SNUB-(RF-S2)

R-881-NE TORUS RF-SNUB-(RF-S6)

R-881-SE TORUS RHR-SNUB-(RH-S103A)

R-859-SW QUAD RHR-SNUB-(Ril-S107A)

R-859-NW OPAD RHR-SNUB-(RE-S20)

R-903-INJ V RM RilR-SNUB-(RH-S21)

R-903-INJ V RM RHR-SNUB-(RH-S22)

R-881-hv TORUS RHR-SNUB-(RH-S23)

R-881-NW TORUS RHR-SNUB-(RE-S24)

R-881-NW TORUS I

RHR-SNUB-(RH-S25)

R-903-NU RHR-SNUB-(RH-S25A)

R-903-NW RHR-SNUB-(RF-S26)

R-903-NW RHR-SNUB-(RH-S27A)

R-931-A RHR HX RM RHR-SNUB-(RH-S29)'

P-903-INJ V RM RilR-ShTB-(RH-S30A)

R-881-SW TORUS RHR-SNUB-(RH-S30B)

R-881-SW TORUS RHR-SNUB-(RH-S32)

R-881-SW TORUS RHR-SNUB-(RH-S33D)

R-8SI-NW TORUS RHR-SNUB-(RH-S34)

R-903-SW l

RHR-SNUB-(RH-S35)

R-903-B RHR llX RM RilR-SNUB-(RH-S36)

R-903-B RHR llX RM RER-SNUB-(RH-S37)

R-903-B RHR HX RM RHR-SNUB-(RH-S38)

R-903-B RHR llX RM RHR-SNUB-(RH-S39)

R-903-R RIIR HX RM RHR-SNUS-(RH-S40)

R-903-B RilR HX RM RPR-SNUB-(RH-S41)

R-859-SW OUAD RHR-SNUB-(RH-S42)

R-859-SW QUAD RHR-SNUB-(RH-S43)

R-881-SW TORUS RHR-SNUB-(RH-S44)

R-881-SW QUAD RHR-SNUB-(RH-S45)

R-881-SW QUAD RHR-SNUB-(RH-S48)

R-881-NW QUAD RHR-SNUB-(RH-S49)

R-881-NW QUAD RHR-SNUB-(RH-SSI)

P-903-A RHR HX RM RHR-SNUB-(RH-SS2)

R-903-A RHR HX RM

-137g-

Table 3.6.1 ACCESSIBLE SAFETY RELATED HYDRAULIC SHOCK SUPPRESSORS (SNUBBERS) (cont'd)

Snubber Location RHR-SNUB-(RH-S54)

R-859-NW QUAD RFR-SNUB-(RH-S55)

R-859-NW QUAD RHR-SNUB-(RH-S56)

R-903-A RilR HX RM RHR-SNUB-(RH-S57)

R-903-A RilR HX RM RHR-SNUB-(RH-S59)

R-881-NW TORUS RHR-SNUB-(RH-S65)

R-881-SW QUAD RHR-SNUB-(RH-566)

R-903-INJ V RM RilR-SNUB-(RH-S76A)

R-881-SW TORUS RHR-SNUB-(RH-S76B)

R-881-SW TORUS RHR-SNUB-(RH-S77)

R-881-SW TORUS RHR-SNUB-(RH-S78A)

R-881-NW TORUS RIIR-SNUB-(RH-S78B)

R-881-NW TORUS RHR-SNUB-(Ril-S80)

R-881-NW QUAD RHR-SNUB-(Ril-S96A)

R-903-NW RiiR-SNUB-(RH-S98)

R-881-NW QUAD RWCU-SNUB-(CU-S89)

R-881-SE TORUS SW-SNUB-(SW-il23A)

IS-SWP RM SW-SNUB-(SW-H23D)

IS-SWP RM SW-SNUB-(SW-il23E)

IS-SWP RM SW-SNUB-(SW-H2311)

IS-SWP RM 4

-137h-

Table 3.6.2 ACCESSIBLE SAFETY RELATED MECHANICAL SHOCK SUPPRESSORS (SNUBBERS)

Snubber Location MS-SFUB-(MS-S149B)

R-903-STM TUNNEL MS-SNUB-(MS-S16)

R-881-NW TORUS MS-SNUB-(MS-S9A)

R-881-SW TORUS MS-SNUB-(MS-S9B)

R-881-SW TORUS RCIC-SNUB-(RF-S51C)

R-881-NE TORUS-RHR-SNUB-(RH-S58)

R-903-A RHR HX RM SGT-SNUB-(PSSP-40)

R-881-SW TORUS SGT-SNUB-(PSSP-74)

R-881-SW TORUS SW-SNUB-(SW-H23B)

IS-SWP RM SW-SNUB-(SW-H23C)

IS-SWP RM SW-SNUB-(SW-H23F) g IS-SWP RM SW-SNUB-(SW-H23G)

IS-SWP PJi 4

b 4

i l

l l

l

-1371-

Table 3.6.3 INACCESSIBLE SAFETY RELATED MECHANICAL SHOCK SUPPRESSORS (SNUBBERS)

Snubber Location CS-SNUB-(CS-S4)

DW-934 CS-SNUB-(CS-SS)

DW-934 CS-SNUB-(CS-S8)

DW-934 CS-SNUB-(CS-S9)

DW-934 MS-SNUB-(MS-S21)

DW-901 l

MS-SNUB-(MS-S22)

DW-901 MS-SNUB-(MS-S63)

DW-921 i

MS-SNUB-(SS-A2)

DW-921 MS-SNUB-(SS-A3)

DW-921 MS-SNUB-(SS-B2)

DW-921 MS-SNUB-(SS-B3)

DW-921 MS-SNUB-(SS-C2)

DW-921 MS-SNUB-(SS-C3)

DW-921 MS-SNUB-(SS-D2)

DW-921 MS-SNUB-(SS-D3)

DW-921 MS-SNUB-(VR-55-23-X)

DW-901

!!S-SNUB-(VR-55-26-Z)

DW-901 i

MS-SNUB-(VR-55-9-Y)

DW-901 l

I MS-SNUB-(VR-55-9-Z)

DW-901 MS-SNUB-(VR-56-12-Y)

DW-901 I

MS-SNUB-(VR-56-24-X)

DW-901 MS-SNUB-(VR-58-12-Y)

DW-921 MS-SNUB-(VR-59-7-X)

DW-921 MS-SNUB-(VR-59-7-Z)

DW-901 MS-SNUB-(VR-60-7-X)

DW-921 MS-SNUB-(VR-60-7-Z)

DW-901 MS-SNUB-(VR 17-X)

DW-901 MS-SNUB-(VR-61-8-X)

DW-901 MS-SNUB-(VR-61-8-Z)

DW-921 MS-SNUB-(VR-62-17-X)

DW-901 MS-SNUB-(VR-62-8-X)

DW-901 MS-SNUB-(VR-62-8-Z)

DW-921 MS-SNUB-(VR-H61D)

DW-888 MS-SNUB-(VR-H62B)

DW-888 MS-SNUB-(VR-H62C)

DW-888 MS-SNUB-(VR-H63B)

DW-888 MS-SNUB-(VR-H63C)

DW-888 MS-SNUB-(VR-H64D)

DW-888 MS-SNUB-(VR-SI)

DW-901 MS-SNUB-(VR-SIO)

DW-901 MS-SNUB-(VR-SI1)

DW-921 MS-SNUB-(VR-S12)

DW-901

!!S-SNUL-(VR-S 14 )

DW-888 MS-SNUB-(VR-S2)

DW-901 MS-SimB-(VR-S20)

DW-921 MS-SNUB-(VR-S21)

DW-921 MS-SNUB-(VR-S22)

DW-901 MS-SNUB-(VR-S23A)

DW-901 MS-SNUB-(VR-S23B)

DW-901 MS-SNUB-(VR-S24A)

DW-901 MS-SNUB-(VR-S24B)

DW-901 MS-SNUB-(VR-S25)

DW-901

-137j-

2 3

4 L4<&+

2 L4 Table 3.6.3 INACCESSIBLE SAFETY RELATED MECHANICAL SHOCK SUPPRESSORS (SNUBBERS) (cont'd)

~ Snubber Location MS-SNUB-(VR-S26).

DW-888 MS-SNUB-(VR-S27)

DW-901

!!S-SNUB-(VR-S3)

DW-888 MS-SNUB-(VR-S30)

DW-921 l

MS-SNUB-(VR-S31)

DW-921 MS-SNUB-(VR-S32)

DW-888 MS-SNUB-(VR-S4)-

DV-901 MS-SNUB-(VR-S40)

DW-921 MS-SNUB-(VR-S41)-

DW-921 MS-SNUB-(VR-S42A)

DW-921 MS-SNUB-(VR-S42B)

DW-921 MS-SNUB-(VR-S43)

DW-888 MS-SNUB-(VR-S50A)

DW-921 MS-SNI'B-(VR-S50B)

DW-921 MS-SNUB-(VR-SSI)

DW-888

!!S-SNUB-(VR-SS A)

DW-901 ItS-SNUB-(V R-SS B)

DW-901 MS-SNUB-(VR-S6)

DW-901 MS-SNUB-(VR-S60)

DP-921 MS-SNUB-(VR-S61)

DW-921 MS-SNUB-(VR-S62A)

DW-921 MS-SNUB-(VR-S62B)

DW-921 MS-SNUB-(VR-S63)

DW-921 MS-SNUB-(VR-570A)

DW-901 MS-SNUB-(VR-S70B)

DW-901 MS-SNUB-(VR-S71A)

DW-901 MS-SNUB-(VP-S71B)

DW-901 MS-SNUB-(VR-S72)

DW-901 MS-SNUB-(VR-S73)

DW-901 MS-SNUB-(VR-S74)

DW-901 MS-SNUB-(VR-S7A)

DW-888 MS-SNUB-(VR-S7B)

DW-888 MS-SNUB-(VR-S8)

DW-888 MS-SNUB-(VR-S80)

DW-901

-MS-SNUB-(VR-S81)

DW-901 MS-SNUB-(VR-S82)

DW-901 MS-SNUB-(VF-S83A)

DW-901 MS-SNUB-(VR-S83B)

DW-901 MS-SNUB-(VR-S84)

DW-901 MS-SNUB-(VR-585)

DW-901 MS-SNUB-(VR-S86A)

DW-901 MS-SNUB-(VR-S86B)

DW-901 MS-SNUB-(VR-S87A)

DW-888 MS-SNUB-(VR-S87B)

DW-888 MS-SNUB-(VR-S88)

DW-888 RF-SNUB-(RF-S10)

DW-921 RF-SNUB-(RF-S11)

DW-921 RF-SNUB-(RF-S12)

DW-921 RF-SNUB-(RF-S13)

DW-921 R F-SNUB-(RF-S 14)

DW-921 RF-SNUB-(RF-S15)

DW-921 RF-SNUB-(RF-S16)

DW-921

-137k-

I T1blo 3.6.3 INACCESSIBLE SAFETY RELATED MECHANICAL SHOCK SUPPRESSORS (SNUBBERS) (cont'd)

Snubber Location RF-SNUB-(RF-S17)

DW-921 RF-SNUB-(RF-S18)

DW-921 RF-SNUB-(RF-S19)

DW-921 RF-SNUB-(RF-S8)

DW-921 RF-SNUB-(RF-S9)

DW-921 I

RIIR-SNUB-(RH-S10)

DW-901 RHR-SNUB-(Ril-Sil)

DW-901 RHR-SNUB-(RH-S13)

DW-921 RHR-SNUB-(RH-S14)

DW-921 i

RHR-SNUB-(RH-SIS)

DW-921 RHR-SNUB-(RH-S16)

DW-901 RiiR-SNUB-(RH-S17)

DW-901 i

RHR-SNUB-(Rif-S18)

DW-901 l

RHR-SNUB-(RH-S19)

DW-901 i

RHR-SNUB-(RH-53)

DW-FLG AREA

}

RilR-SNUB-(RH-S4)

DW-FLG AREA RHR-SNUB-(Ril-SS)

DW-921 RHR-SNUB-(RH-S6)

DW-921 RHR-SNUB-(RH-S67)

DW-901 RHR-SNUB-(Ril-S68)

DW-901 RifR-SNUB-(RH-S69A)

DW-901 RHR-SNUB-(RH-S69B)

DW-901 RHR-SNUB-(Ril-S7)

DW-921 RHR-SNUB-(RH-S70)

DW-901 RHR-SNUB-(Rll-S71)

DW-901 RilR-SNUB-(RH-S72)

DW-901 RilR-SNUB-(RH-S72A)

DW-901 RHR-SNUB-(RH-S73)

DW-901 RHR-SNUB-(RH-S8A)

DW-901 RHR-SNUB-(RH-S8B)

DW-901 RHR-SNUB-(RH-S9)

DW-901 RR-SNUB-(SS-1A)

DW-888 RR-SNUB-(SS-1B)

DW-888 i

RR-SNUB-(SS-2A)

DW-888 RR-SNUB-(SS-2B)

DW-888 l

RR-SNUB-(SS-3A1)

DW-901 RR-SNUB-(SS-3A2)

DW-901 i

RR-SNUB-(SS-3BI)

DW-901 i

RR-SNUB-(SS-3B2)

DW-901 i

RR-SNUB-(SS-4A)

DW-901 i

RR-SNUB-(SS-4B)

DW-901 RR-SNUB-(SS-SA)

DW-888 RR-SNUE-(SS-5B)

DW-888 RR-SNUB-(SS-8A1)

DW-901 RR-SNUB-(SS-8A2)

DW-901 RWCU-SNUB-(CU-S3A)

DW-921 RWCU-SNUB-(CU-S3B)

DW-921

-1371-l

Table 3.6.4 INACCESSIBLE SAFETY RELATED HYDRAULIC SHOCK SUPPRESSORS (SNUBBERS)

Snubber Location RR-SNUB-(SS-7Al)

DW-901 RR-SNUB-(SS-7A2)

DW-901 RR-SNUB-(SS-7BI)

DW-901 RR-SNUB-(SS-7B2)

DW-901 1

'i

-137m-L

Revised Technical Specification for SBGT Testing Requirement Revised Pages:

165 182 183 215 215d 215e References 1) ANSI N510-1980

2) NRC Inspection Report 50-298/82-02, Item 10.a(2)
3) Regulatory Guide 1.52 The current Technical Spe.-ifications for Cooper Nuclear Station lists requirements for the testing of charcoal filters in the Standby Gas Treatment System and the Control Room Ventilation System.

Nebraska Public Power District requests a revision to the Technical Specifications as shown on the attached pages.

This request is made in order to bring testing criteria for filters, as contained in the Technical Specifications, into line with current industry standards and guidance (References 1 and 3).

In relation to this request, Messrs. L. Wilborn and B. Murray conducted a routine inspection (Reference 2) on January 11-15, 1982.

Their inspection report also recommended this change in Item 10.a(2).

l e

7

.e m---

1 LkMITINGCONDITIONSFOROPERATION SURVEILLANCE REOUIREMENTS 3.7. (cont'd.)~

4.7 (cont'd.)

B.

Standby Gas Treatment System B.

Standby Cas Treatment System 1.

Except as specified in 3.7.B.3 below, 1.

At least ence per operating cycle the both circuits of the standby gas treat-following conditions shall be demon-ment system and the diesel generators

strated, required for operation of such circuits shall be operable at all times when a.

Pressure drop across the combined HEPA secondary containment integrity is filters and charcoal adsorber banks is required.

less than 6 inches of water at the sys-tem design flow rate.

b.

Inlet heater input is capable of reduc-ing R.H. from 100 to 70% R.H.

2.a. The results of the in-place cold DOP 2.a. The tests and sample analysis of Speci-and halogenated hydrocarbon tests at fication 3.7.B.2 shall be performed at design flows on HEPA filters and char-least once per year for standby service coal adsorber banks shall show 199%

or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system oper-DOP removal and 199% halogenated hydro-ation and following significant painting, carbon removal.

fire or chemien1 release ir any ventila-tion zone communicating with the system.

b.

The results of laboratory carbon sample b.

Cold DOP testing shall be performed after analysis shall show 199% radioactive cach complete or partial replacement of methyl iodide removal at a velocity the HEPA filter bank or af ter any struc-within20percengof actual system de-tural maintenance on the system housing.

sign, 11.75 mg/m inlet methyl iodide concentration, 370% R.H. and

_30 F.

c.

Fans shall be shown to operate within c.

Halogenated hydrocarbon testing shall be

+10% design flow.

performed af ter each complete er partial replacement of the charcoal adsorber bark or af ter any structural maintenance on the system housing.

d.

Each circuit shall be operated with the heaters on at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month.

I e.

Test sealing of gaskets for housing doors I

downstream of the HEPA filters and char-coal adsorbers shall be performed at, i

and in conformance with, each test per-formed for compliance with Specification 4.7.B.2.a and Specification 3.7.B.2.a.

3.

From and after the date that one cir-3.

System drains where present shall be in-cuit of the stanWoy gas treatment sys-spected quarterly for adequte water 1cvel tem is made or iound to be inoperable in loop-seals.

for any reason, reactor operation or fuel handling is permissible only during l

l the succeeding seven days unicss such circuit is sooner made operabic, pro-vided that during such seven days all active components of the other standby gas treatment circuit shall be operable,

-165-t

3.7.B & 3.7.C BASES (cont'd)

High efficiency particulate absolute (HEPA) filters are installed before and after the charcoal adsorbers to minimize potential release of particulates to the environment and to prevent clogging of the iodine adsorbers.

The charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment. The in-place test results should indicate a system leak tightness of less than I percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates. The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 99 percent for expected accident conditions.

If the efficiencies l

of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the 10 CFR 100 guidelines for the accidents analyzed. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.

Only one of the two standby gas treatment systems is needed to cleanup the reactor building atmosphere upon containment isolation.

If one system is found to be inoperable, there is no immediate threat to the containment system performance and reactor operation or refueling operation may continue while repairs are being made.

If neither circuit is operable, the plant is brought to a condition where the standby gas treatment system is not required.

4.7.B & 4.7.C BASES Standby Cas Treatment System and Secondary Containment Initiating reactor building isolation and operation of the standby gas treatment system to maintain at least a 1/4 inch of water vacuum within the secondary containmcat provides an adequate test of the operation of the reactor building isolation valves, leak tightness of the reactor building and performance of the standby gas treatment system.

Functionally testing the initiating sensors and associated trip channels demonstrates the capability for automatic actuation.

Performing these tests prior to re-fueling will demonstrate secondary containment capability prior to the time the primary containment is opened for refueling. Periodic testing gives sufficient confidence of reactor building integrity and standby gas treat-ment system performance capability.

Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. A 7.8 kw heater is capable of maintaining relative humidity below 70%.

Heater capacity and pressure drop should be determined at least once per operating cycle to show system performance capability.

The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant shall be performed in accordance with ANSI N510-1980. The test cannisters that are installed with the adsorber trays should be used for the charcoal adsorber efficiency test.

Each sample should be at least two inches in diameter and a length equal to the thickness of the bed.

If test results are unacceptable, all adsorbent in the system shall be replaced

-182-

4.7.B & 4.7.C' BASES with an~ adsorbent qualified according to Table 1 of Regulatory Guide 1.52.

The replacement tray for the adsorber tray removed for the test should meet the same adsorbent quality. -Tests of the HEPA filters with DOF aerosol shall be performed in accordance to ANSI N510-1980. Any filters found defective l

'shall be replaced with filters qualified pursuant to Regulatory Position C.3.d. of Regulatory Guide 1.52.

All elements of the heater should be. demonstrated to be functional and operable during the test of heater capacity. Operation of the heaters will prevent moisture buildup in the filters and adsorber system.

With doors closed and fan in operation, DOP aerosol shall be sprayed externally along the full linear periphery of each respective door to check the gasket seal. Any detection of DOP in the fan exhaust shall be considered an unacceptable test result and the gaskets repaired and test repeated.

If system drains are present in the filter /adsorber banks, loop-seals must be used with adequate water level to prevent by-pass Icakage from the banks.

4 If significant painting, fire or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for. operational use.

The determination of significance shall be made by the operator on duty at the time of the incident. Knowledgeable staff members should be consulted prior to making this determination.

Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system performance capability.

If one standby gas treatment system is inoperable, the other system must be tested daily. This substantiates the availability of the operable system and thus reactor operation or refueling operation can continue for a limited period of time.

3.7.D & 4.7.D BASES Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary con-tainment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize j

the potential leakage paths from the containment in the event of a loss-of-coolant accident.

The maximum closure times for the automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.

These valves are highly reliable, have a low service requirement, and most are normally closed. The initiating sensors and associated trip channels are also checked to demonstrate the capability for automatic isolation.

The test interval of once per operating cycle for automatic initiation i

l

-183-

l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.12 Additional Safety Related Plant 3.12 Additional Safety Related Plant Capabilities Capabilities Applicability:

Applicability:

APPlics to the surveillance require-Applies to the operating status of the ments for the main control room venti-main control room ventilation system, lation system, the reactor building the reactor building closed cooling closed cooling water system and the water system and the service water service water system which are required system.

by the corresponding Limiting conditions for Operation.

Objective:

Objective:

To assure the availability of the main To verify that operability or availa-bility under conditions for which these control room ventilation system, the capabilities are an essential response reactor building closed cooling water to station abnormalities.

system and the service water system upon the conditions for which the capability is an essential response to station abnormalities.

A.

Main Control Roem Ventilation A.

Main Control Room Ventilation 1.

At least once per operating cycle, the 1.

Except as specified in Specification Pressure drop across the combined HEPA 3.12.A.3 below, the control room air filters and charcoal absorber banks treatment system, the diesel shall be demonstrated te be less than generators required for operation of 6 inches of water at system design flow this system and the nain control room rate.

air radiation monitor shall be oper-able at all times when containment integrity is required.

2.a. The tests and sample analysis of 2.a. The results of the in-place cold DOP Specification 3.12. A.2 shall be performed and halogenated hydrocarbon tests at least once per year for stardby service at design flows on HEPA filters r after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system and charcoal absorber banks shall show > 99% C0P removal and > 99%

peration and following significant paint-ing, fire r chemical release in any halogenated hydrocarbon removal, ventilation zone communicating with the

system, b.

Cold DOP testing shall be perforned b.

The results of laboratory carben fter each complete or partial replace-sample analysis shall show > 99%

ment of the HEPA filter bank or afcer radioactive methyl iodide removal any structural maintenance on the system at a velocity witgin20%ofsystem h using.

~

design, 1.75 mg/m inlet iodide concentration, > 95% R.H. and

<30*F.

~

c.

Mal genated hydrocarbon testing shall c.

Fans shall be shown to operate with-in + 10% design flow.

be performed after each complete or partial replacement of the charcoal absorber bank or af ter any structural maintenance on the system housing.

-215-

3.12 BASES A.

Main Control Room Ventilation System The control room ventilation system is designed to filter the control room atmosphere for intake air and/or for recirculation during control room isolation conditions. The system is designed to automatically start upon control room isolation and to maintain the control room pressure to the design positive pressure so that all leakage should be out leakage.

High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential intake of radioiodine to the control room. The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates.

The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 99 percent for expected l

accident conditions.

If the efficiencies of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the allowable levels stated in Criterion 19 of the General Design Criteria for Nuclear Power Plants, Appendix A to 10 CFR Part 50.

Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.

If the system is found to be inoperable, there is not immediate threat to the control room and reactor operation or refueling operation may continue for a limited period of time while repairs are being made.

If the system cannot be repaired within seven days, the reactor is shutdown and brought to cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or refueling operations are terminated.

B.

Reactor Building Closed Cooling Water System The reactor building closed cooling water system has two pumps and one heat exchanger in each of two loops. Each loop is capable of supplying the cooling requirements of the essential services following design accident conditions with only one pump in either loop.

The system hes additional flexibility provided by the capability of inter-connection of the two loops and the backup water supply to the critical loop by the service water system. This flexibility and the need for only one pump in one loop to meet the design accident requirements justifies the 30 day repair time during normal operation and the reduced requirements during head-off operations requiring the availability of LPCI or the core spray-systems.

C.

Service Water System The service water system consists of four vertical service water pumps located in the intake structure, and associated strainers, piping, valving and instrumentation. The pumps discharge to a common header from which independent piping supplies two Seismic Class I cooling water loops and one turbine building loop. Automatic valving is provided to shutoft all supply to the turbine building loop on drop in header pressure thus assuring supply to the Seismic Class I loops each of which feeds one diesel generator, two RHR service water booster pumps, one control room basement fan coil unit and one RBCCW

-215d-

3.12 BASES (cont'd) heat exchanger. Valves are included in the common discharge header to permit the Seismic Class I service water system to be operated as two independent loops. The heat exchangers are valved such that they can be individually backwashed without interrupting system operation.

During normal operation two or three pumps will be required. Three pumps are used for a normal shutdown.

The loss of all a-c power will trip all operating service water pumps.

The automatic emergency diesel generator start system and emergency equipment starting sequence vf]1 then start one selected service water pump in 30-40 seconds.

In the meantime, the drop in service water header pressure will close the turbine building cooling water isolation valve guaranteeing supply to the reactor building, the control room basement, and the diesel generators from the one service water pump.

Due to the redundance of pumps and the requirement of only one to meet the accident requirements, the 30 day repair time is justified.

D.

Battery Room Ventilation The temperature rise and hydrogen buildup in the battery rooms without adequate ventilation is such that continuous safe operation of equipment in these rooms cannot be assured.

4.12 BASES A.

Main Control Room Ventilation System Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate wf]1 indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Pressure drop should be determined at least once per operating cycle to show system performance capability.

Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant should be performed in accordance with ANSI N510-1980.

l The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. The test cannisters that are installed with the adsorber trays should be used for the charcoal adsorber efficiency test.

Each sample should be at least twc inches in diameter and a length equal to the thickness of the bed.

If test results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent qualified according to Table 1 of Regulatory Cuide 1.52.

The rcplacement tray for the absorber tray removed for the test should meet the same adsorbent quality. Tests of the HEPA filters with DOP aerosol shall be performed in accordance to ANSI N510-1980. Any HEPA filters found l

defective shall be replaced with filters qualified pursuant to Regulatory Position C.3.d of Regulatory Guide 1.52.

Operation of the system for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month will demonstrate operability of the filters and adsorber system and remove excessive moisture built up on the adsorber.

-215e-

4 i

Revised Technical Specification for.

10CFR50 Appendix J Testing Revised Pages:

159 176 160 177 161 178 162 184 162a Deleted:

15Sa 163 178a 164 183a 168 173 Reference 1)

Letter from D. G. Eisenhut to J. M. Pilant dated September 3, 1982, " Exemption to Appendix J to 4

10CFR Part 50 and Safety Evaluation Report" L

This change is in response to Reference 1.

The feedwater check valves will be tested with air or nitrogen as required.

The requested Technical Specifications for the containment air locks are proposed.

It should be noted that the SER of Reference 1 stated that " test performance requires shutting down the reactor and opening the equipment hatch in order to install a strongback on the inner airlock door..." It is true that the strongback has to be attached to the inside door of the contain-ment airlock in order to pressurize the airlock to accident pressure (Pa), but the drywell does not have to be entered to install the strongback because the strongback is stored inside the containment airlock.

To attach the strongback during full power operation will expose personnel to radiation levels of approximately 500 mr/hr.

Total exposure for this testing is estimated to be approximately 2 rem, which the District feels is excessive.

Even though it is not necessary to enter the drywell to do the containment airlock test at Pa, the District still

+

followed the NRC recommendation in the proposed Technical i

Specification.

The District will calculate a new correlation of reduced pressure s

leakage rates to full pressure leakage rate for the bellows leakage test and the containment airlock test.

The calculation I.

will be based on the Franklin Research Center Technical j

Evaluation Report, Appendix A, Procedure 'B'.

This calculation i

will be added to the appropriate procedure for the local leak rate tests.

I

,e

Since resolution of the above three issues concludes the Staff's review of CNS as regards Appendix J, minor forinat changes are being proposed for Section 3/4.7 of the Technical, Specifications which should make this section easier to utilize.

The action on the initiating signal for the Reactor Water Sampic Valves in Table 3.7.1 (page 168) is being changed since the valves actually go closed on a signal.

I 1

I

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 4.7 CONTAINMENT SYSTEMS 3.7 CONTAINMENT SYSTEMS Applicability:

Applicability:

Applies to the operating status of Applies to the primary and secondary the primary and secondary contain-e ntainment integrity.

ment systems.

Objective:

Objective:

To assure the integrity of the pri-To verify the integrity of the primary mary and secondary containment systems, and secondary containment.

Specification:

Specffication:

A.

Primary Containment A.

Primary Containment 1.

Suppression Pool l

l 1.

Suppression Pool At any time that the nuclear system

a. The suppression pool water level is pressuris:ed above atmospheric and temperature shall be checked pressure or work is being done nce Per day, which has the potential to drain
b. Whenever there is indication of the vessel, the suppression pool water volume and temperature shall relief valve operation or testing which adds heat to the suppression be maintained within the following limits except as specified in p

1, the pool temperature shs]1 be continual 1 monitored anc also 3

3.7.A.2. and 3.5.F.5.

observed and logged every 5 3

a.

Minimum water volume - 87,650 ft minutes until the heat addition is terminated.

3 b.

Maximum water volume - 91,000 f t

c. Whenever there is indication of c.

Maximum suppression pool temperature relief valve operation with the during normal power operation - 90 F.

temperature of the suppression For 45 days, commencing July 16, 1981, pool reaching 160 F or more and the suppression pool temperature me" the primary coolant system pres-be increased to 95 whenever the sure greater than 200 psig, an river water temperature is such that external visual examination of the peol temperature cannot he main-the suppression chamber shall toined below 90 F.

be conducted before resuming d.

During testing which adds heat to P wer Peration.

the suppression pool, the water temperature shall not exceed 10 F

d. A visual inspection of the above the normal power operation suppression chatt.ber interior, limit specified in c. above.

In including water line regions, connection with such testing, the shall be made at each major pool temperature must he reduced to refueling outage, below the normal power operation limit specified in c. above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

e.

The reactor shall be scrammed from any operating condition if the pool temperature reaches 110 F.

Power operation shall not be resumed until the pool temperature is reduced below the normal power operation limit specified in c.

above.

-159-

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A.1 (cont'd.)

4.7.A (cont'd.)

2.

Leak Rate Testing

f. During reactor isolation conditions, the reactor pressure vessel shall be depressurized to less than 200
a. Integrated leak rate teste (ILRT's) psig at normal cooldown rates if shall be performed to verify primary the pool temperature reaches 120 F.

containment integrity.

Primary con-tainment integrity is confirned if the 2.

Containment Integrity leakage rate does not exceed the equivalent of 0.635 percent of the Primary containment integrity shall Primary containment volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> he traintained at all times when the at 58 psig, reactor is critical or when the reactor water temperature is above

b. Integrated leak rate tests raay be per-212*F and fuel is in the reactor f rmed at either 58 psig or 29 psig, the vessel except while performing "open leakage rate test period, extending to vessel" physics tests at power levels 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of retained internal pressure.

If it can be demonstrated to the satis-not to exceed 5 MW(t).

faction of those responsible for the acceptance of the containment structure that the leakage rate can be accurately determined during a shorter test peried, the agreed-upon shorter period may be used.

Prior to initial operation, integrated leak, rate tests must be performed at 58 and 29 psig (with the 29 psig test being performed prior to the 58 psig test) to establish the allowable leak rate, L (in percent of containment volume per 2k hours) at 29 psig as the lesser of the following values:

(L, is 0.635 percent)

L = 0.635 I'tm bam l

for 'tm < 0.7 bam where Itm = measured ILR at 29 psig Lam = measured ILR at 58 psig, and L

~ 1.0 tm <

am L = 0.635 P t

Pa

-160-

LIMITING CONDITIONS FOR OPFRATION SURVEILLANCE RFOUIREMFUTS 4.7.A.2.b. (cont'd.)

l3.7.A(cont'd.)

where P, = peak accident pressure, 58 psia P = appropriately mecsured test pres" sures (psia) for 'tm > 0.7 am c.

The ILRT's shall be performed at the fc31owing minimum frequency:

1.

Prior to initial urft operation.

2.

At approximately three and one-third year interva]n re that any ten-year interval would include four ILRT's.

These intervals may be extended up to eight months if recessary to coincide with refueling outage.

{he measured leakage retes, I'{:m and d.

am, p' hall be Jess than 0.75

,t and 0.75 a for the reduced pressure tests and peak pressure test respectively.

e.

Except for the initial ILRT, all ILRT's shall be performed without any pre-liminary leak detection surveys and leak repaire immediately prior to the test.

If an ILRT has to be ter-minated due to excessive leakage through identified leakage paths, the leakage through such paths r. hall be determined by a local leakage test and recorded. After repairs are made another ILRT shall be conducted.

If an ILET is completed but the acceptance criteria of Specificatien 4.7.A.2.d is not satisfied and repairs are necessary, the ILRT need rot he 4

-161-

~

LIMITING CONDITIONS FOR OPIRATION SURVEILLANCE REQUIREMENTS

'3.7.A (Cont'd) 4.7.A.2.e (cont'd) l repeated provided locally measered leakage reductions, achieved by re-pairs, reduce the containment's overall measured leakage rate suf-ficiently to meet the acceptance criteria.

f.

Local Leak Rate Tests

1. With the exceptions specified below, local leak rate tests (LLRT's) shall be performed on the primary containment testable penetrations and isolation valves at a pressure of 58 psig during each reactor shut-down for refueling, or other conven-ient intervals, but in no case at intervals greater than two years.

Table 3.7.2 specifies testable penetrations with double 0-ring seals, Table 3.7.3 specifies test-able penetrations with testable bellows, and Table 3.7.4 specifies primary containment testable isolation valves. The test dur-ation of all valves and penetra-tions shall be of sufficient length to determine repeatable results.

The total acceptable leakage for all valves and penetrations other than the MSIV's is 0.60 La.

2. Bolted double-gasket seals (Table 3.7.2) shall be tested af ter each opening and during each reactor shutdown for refueling, or other convenient intervals but in no case at intervals greater than two years.
3. The main steam isolation valves (MSIV's) shall be tested at a pres-sure of 29 psig.

If a total leak-age rate of 11.5 scf/hr for any one MSIV is exceeded, repairs and retest shall be performed to correct the condition. This is an exemption to Appendix J of 10CFR50.

-162-

9 l

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A (Cont'd) 4.7.A.2.F (cont'd)

~%

4. Main steam line and feedwater line

/

expansion bellows as specified in

^

i Table 3.7.3 shall be tested by pressuririr.g'between the laminations I

of the belinws at a pressure of 5 psig.

This is an exemption to Appendix J

),

of 10CFR50.

^

5. The personnel airlock shall be
'. Y tested at 58 psig at- ' intervals no longer than 'six months. This testing may+be extended to the next refueling outage (not t'o exceed.24 months) provided that there have been no tirlock openings since the last successful test at 58 psig.

In the event the personnel airlock is not opened between refueling outages, it shall be leak checked at 3 psig at intervals no longer than six months.

Within three days of opening (or avery three days during pericds of frequent optning) when containment integrity is' required, test the.

personnel airlock at 3 psig. This is an exemption to Appendix J.of 100FR50.

o g.

' Continuous' Leak Rate Monit'or I,

i hen the primary cc' tainment. is,

n inerted the containment sh'all be '

j

,I e

continueusly ronitored for grose

/

J' leakage by retiew of the. inerting system makeup r'equirements. This monitoring system may be;taken out of service for *aaintenance but shall b'e returned to service as soon as p" a c ticab le.

g a

h.

Dryuell Surfacen i

j The interior surfaces of,the drywell and torus shall be visually inspected each operating cycle for evidence of torus corrosion or Icakage.

w f

'n p

,9

'l e

N V

3, 1.'\\ '

~

e y

e.

/

,e

- s f

-162a-

l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 5

]

3.7.A (cont'd.)

4.7.A (cont'd.)

i 3.

Pressure Suppression Chamber -

3.

Pressurc Suppression Chamber -

' Reactor Building Vacuum Breakerg Reactor Building Vacuum Breakers a.

Except as specified in 3.7.A.3.b a.

The pressure suppression chamber-reactor s below, two pressure suppression building vacuum breakers and associated chamber-reactor building vacuum instrumentation, including set points breakers shall be operable at all shall be checked for proper operation times when primary containment in-every three months.

tegrity is required. The set point of the differential pressure instru-mentation which actuates the pressure suppresrion chamber-reactor building air actuated vacuum breakers shall be 0.5 psid. The self actuated vacuur breakers shall open fully when subjected to a force equivalent to 0.5 psid actinF on the valve disc.

b.

From and after the date that one of b.

During each refueling outage each the pressure suppression chamber-vacuum breaker shall be tested to reactor building vacuum breakers is determine that the force required made'or found to be inoperable for to open the vacuum breaker does not any reason, the vacuum M eaker switch exceed the force specified in shall be secured in S.e closed position Specifications 3.7.A.3.a and each and reactor operation is permissible vacuum breaker shall be inspected only during the succeeding seven days and verified to meet design unless such vacuum breaker is sooner requirements, made operable, provided that the repair procedure does not violate primary containment integrity.

4.'

Drywell-Pressure Suppression Chamber 4.

Drywell-Pressure Suppression Chamber Vacuum Breakers Vacuum Breakers a.

Each drywell-suppression chamber vacuum a.

When primary containment is required, all drywell-suppression chamber vac-breaker shall be exercised through an

.uum breakers shall be operable at the opening-closing cycle every 30 days.

0.5 psid setpoint and positioned in the fully closed position as indicated by the position indicating system except during testing and except as specified

- in 3.7. A.4.b and.c below.

-b.

Three drywe]I-suppression chamber b.

When it is determined that a vacuum vacuum breakers may be determined breaker valve is inoperable for opening to be inoperabic f' r opening pro-at a time when operability is required o

vided they are secured in the fully all other vacuum breaker valves shall closed position or that the require-be exercised immediately and every 15 ment of 3.7.A.4.c is demonstrated to days thereafter until the inoperah3e be met.

valve has been returned to normal service.

m

-163-

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A.4 (cont'd.)

4.7.A.4 (cont'd.)

l c.

Once each operating cycle, each vacuum c.

The total leakage between the dry-well and suppression chamber shall breaker valve shall be visually in-be less than the equivalent Jeakage sPected to insure proper maintenance through a 1" dianeter orifice.

and operation of the position frdication switch. The differential pressure set-point shall be verified.

d.

Prior to reactor startup after each d.

If specifications 3.7.A.4.a. b or c, cannot be met, the situation shall refueling, a leak test of the drywell be corrected within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the to suppression chamber structure shall be conducted to demonstrate reactor will be placed in a cold sht/ S T. condition within the sub-that the requirerent of 3.7 A.4.c sequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

I8 **E*

5.

Oxygen Concentration 5.

Oxygen concentration a.

The primary containment oxygen con-a.

After completion of the startup test centration shall be measured and program and demonstration of plant electrical output, the primay con-recorded at least twice weekly.

tainment atmosphere shall be reduced to less than 4% oxygen with nitrogen gas during reactor power operation with reactor coolant pressure above 100 psig, except as specified in 3.7.A.5.b.

b.

The quantity of liquid nitrogen in b.

Within the 24-hour period subsequent to placing the reactor in the Run mode the liquid nitrogen storage tank sha.11 following a shutdown, the containment be determined twice per week when the atmosphet e oxygen concentration shall v luce requirements of 3.7.A.5.c are I" *ff*C'*

be reduced to less than 4% by volume and maintained in this condition.

De-inerting may commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a shutdown.

When the containment atmosphere oxygen c.

concentration is required to be less than 4%, the minimum quantity of liquid nitrogen in the liquid nitrogen storage i

tank shall be 500 gallons.

i l

l d.

If the specifications of 3.7.A 5.a thre e cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l l

e.

The specifications of 3.7.A.5.a thru d are not applicabic during a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> l

continuous period between the dates of March 22, 1982 and March 25, 1982.

-164-

COOPER NUCLEAR STATION TABLE 3.7.1 (Page 1)

PRIMARY CONTAINMENT ISOLATION VALVES Number of Power Maximum Action On Operated Valves Operating Normal Initiating Valve & Steam Inboard Outboard Time (Sec) (1)

Position (2)

Signal (3)

Main Steam Isolation Valves MS-AO A,B,C, & D 4

3<T<5 0

GC

]

MS-AO A,B,C, & D 4

3<T<5 0

CC i

Drywell Floor Drain Iso. Valves 1

1 15 0

GC RW-AO-82, RW-AC-83 Drywell Equipment Drain 1

1 15 0

GC 8

Iso. Valves RW-AO-94, RW-AO-95 i

Main Steam Line Drain 1

1 30 C

SC l

Valves MS-MO-74, MS-MO-77 I

I Reactor Water Sample Valves 1

1 15 0

GC l-l RRV-740AV, RRV-741AV l

l Reactor Water Cleanup System 1

1 60 0

GC' t

Iso. Valves RWCU-MO-15, RWCU-MO-18 l

2 RHR Reactor Head Spray 1

1 60 C

SC l

Iso. Valves RHR-MO-32. RHR-MO-33 I

RHR Suction Cooling Iso.

1 1

40 C

SC Valve RHR-MO-17. RHR-MO-18 RHR Discharge to Radwaste 1

1 20 C

SC l

Iso. Valves RHR-MO-57, RHR-MO-67 y

1 Suppression Chamber Purge &

2 15 C

SC I

Vent PC-245AV, PC-230MV Suppression Chamber N Supply 2

15 C

SC I

2 PC-237AV, PC-233MV i

TABLE 3.7.4 PRIMARY CONTAINMENT TESTABLE ISOLATION VALVES TEST PEN. NO.

VALVE NUMBERS MEDIA X-7A MS-AO-80A and MS-AO-86A, Main Steam Isolation Valves Air X-7B MS-AO-80B and MS-AO-86B, Main Steam Isolation Valves Air X-7C MS-AO-80C and MS-AO-86C, Main Steam Isolation Valves Air X-7D MS-AO-80D and MS-AO-86D, Main Steam Isolation valves Air X-8 MS-M0-74 and MS-MO-77, Main Steam Line Drain Air X-9A RF-15CV and RF-16CV, Feedwater Check Valves Air X-9A RCIC-AO-22. RCIC-MO-17, and RWCU-15CV, RCIC/RWCU Connection to Feedwater Air X-9B RF-13CV and RF-14CV, Feedwater Check Valves Air X-9B HPCI-AO-18 and HPCI-MO-57 HPCI Connection to Feedwater Air X-10 RCIC-MO-15 and RCIC-MO-16, RCIC Steam Line Air X-11 HPCI-MO-15 and HPCI-M0-16, RPCI Steam Line Air X-12 RHR-MO-17 and RHR-MO-18, RHR Suction Cooling Air X-13A RHR-MO-25A and RHR-M0-27A, RHR Supply to RPV Air X-13B RHR-MO-25B and RHR-M0-27B, RHR Supply to RPV Air X-14 RWCU-MO-15 and RWCU-M0-18. Inlet to RWCU System Air X-16A CS-MO-11A and CS-MO-12A, Core Spray to RPV Air X-16B CS-M0-11B and CS-M0-12B, Core Spray to RPV Air X-17 RHR-MO-32 and RHR-M0-33, RPV Head Spray Air X-18 RW-732AV and RW-733AV, Crywell Equipment Sump Discharge Air X-19 RW-765AV and RV,'66AV, Drywell Floor Drain Sump Discharge Air X-25 PC-232MV and PC-238AV, Purge and Vent Supply to Drywell Air X-25 ACAD-1305MV and ACAD-1306MV, supply to Drywell Air X-26 PC-231MV and PC-246AV, Purge and Vent Exhaust -

nrywell Air X-26 ACAD-1310MV, Bleed f rom Drywell Air I

f

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3.7.A & 4.7.A BASES Prinary Containment The integrity of the primary containment and operation of the core standby cooling system in combination, limit the off-site doses to values less than those-suggested in 10CFR100 in the event of a break in the primary system piping. Thus, centainment integrity is specified whenever the potentia]

for violation of the primary reactor system integrity exists. Concern about such a violation exists whenever the reactor is critical and above atmos-pheric pressure. An exception is made to this requirement during initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required. There will be no pressure on the system at this time, thus greatly reducing the chances of a pipe break. The reactor nay be taken critical during this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring. Procedures and the Rod Worth Minimizer would limit control worth such that a rod drop would not rceult in any fuel damage.

In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment system, which shall be oper-ational during this time, offer a sufficient barrier to keep off-site doses well below 10CFR10G limits.

The pressure suppression pool water provides the heat sink for the reactor primary system energy release fo]Iowing a postulated rupture of the system.

The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1035 psig.

Since all of the gases in the drywell are purged into the precoure suppression chamber air space during a loss-of-coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber (water and air) was obtained by con-sidering that the total volume of reactor coolant to be condensed is diccharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

Using the mininum or maximum water volumes given in the specification, con-tainment pressurc during the design basis accident isapproximately38psjg which is below the maximum of 62 psig. Maximum water volume of 91,000 ft regults in a downcomer submergence of 5' and the minimum volume of 87,650 ft results in a submergence approximately 12 inches less. The majority of the Eodega tests were run with a submerged length of 4 feet and with complete condensation. Thus, with respect to downconer submergence, this specification is adequate. The maximum temperature at the end of blowdown tested during the Humbolt Bay and Bodega Bay tests was 170*F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures ebove 170*F.

Sheuld it be necessary to drai-he suppression chamber, this should only

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b.7.A&4.7.A BASES (cont'd) be done when there is r.o requirement for core standby cooling systems opera-bility as explained in bases 3.5 F.

Experimental data indicates that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 160*F during any period of relief valve operation with sonic conditions at the discharge exit.

Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regine of potentially high suppression chamber loadings.

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open. This action would include:

(1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor chutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separatcd from that of the stuck-open relief valve to assure nixing and uniformity of energy insertion to the pool.

Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally change very slowly and monitoring these para-meters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be continually monitored and frequently Icgged during periods of significant heat addition, the temperature trends will be closely fo)) owed so that appropriate action can be taken. The requirement for an external visual examination following any event where potentially high loadings could occur providos assurance that no significant damage was encountered.

Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to he the points of highest stress.

Inerting Safety Guide 7 assumptions for Metal-k'ater reaction result in hydrogen concentration in excess of the Safety Guide 7 flammability limit. By keeping the oxygen concentration less than 4% by volume the requirements of Safety l

Guide 7 are satisfied.

i The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is much more probable than the occurrence of the loss-ef-coolant accident upon which the specified oFygen concentration limit is based. permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety offered without significantly reducing the margin of safety. Thus, to preclude the possibility l

of starting the reactor and operating for extended periods of time with significant leaks in the primary system, leak inspections are scheduled l

during periods when the primary system is at or near rat (c' rpc ratits ter..p-l erature and pressure. The 24-hour period to provide inerting is judged to bc sufficient to perform the leak inspection and establish the required oxygen I

concentration.

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3.7.A & 4.7.A BASES (cont'd)

The primary containment is normally slightly pressurized during periods of reactor operation. Nitrogen used for inerting could leak out of the contain-ment but air could not-leak in to increase oxygen concentration. Once the containment is filled with nitrogen to the required concentration, no moni-toring of oxygen concentration is necessary. However, at least twice a week the oxygen concentration will be determined as added assurance.

The 500 gallon conservative limit on the nitrogen storage tank assures that adequate time is available to get the tank refilled assuming normal plant operation. The estimated maximum makeup rate is 1500 SCFD which would require about 160 gallons for a 10 day makeup requirement. The normal leak rate should be about 200 SCFD.

Vacuum Relief The purpose of the vacuum relief valves is to equalize the pressure between the J

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'3.7.D & 4.7.D BASES (cont'd) results in a failure probability of 1.1 x 10-that a line will not isolate.

More frequent testing for valve operability results in a greater assurance that the valve will be operable when needed.

In order to assure that the doses that may result from a steam line break do not exceed the 10CFR100 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steam line isolation valves. Analyses indicate that fuel rod cladding perforations would be avoided for main steam valve closure times, including instrument delay, as long as 10.5 seconds.

The primary containment is penetrated by several small diameter instrument lines connected to the reactor coolant system. Each instrument line contains a 0.25 inch restricting orifice inside the primary containment and an excess flow check valve outside the primary containment. A program for periodic testing and examination of the excess flow check valves is performed as follows:

1.

Vessel at pressure sufficient to actuate valves. This could be at time of vessel hydro following a refueling outage.

2.

Isolate sensing line from its instrument at the instrument manifold.

3.

Provide means for observing and collecting the instrument drain or vent valve flow.

4.

Open vent or drain valve.

Observe flow cessation and any leakage rate.

a.

b.

Reset valve after test completion.

5.

The head seal leak detection line cannot be tested in this manner. This valve will not be exposed to primary system pressure except under unlikely conditions of seal failure where it could be partially pressurized to reactor pressure. Any leakage path is restricted at the source and there-fore this valve need not be tested. This valve is in a sensing line that is not safety related.

6.

Valves will be accepted if a m'arked decrease in flow rate is observed and the leakage rate is acceptable.

3.7.E Bases In conjunction with the Mark I Containment Short Term Program, a plant unique analysis was performed as described in the licensee's i

letter of October 4,1976, which demonstrated a factor of safety of at least two for the weakest element in the suppression chamber support system and attached piping. The maintenance of drywell-suppression chamber differential pressure of 1.0 psid and a suppression chamber water level corresponding to a downcomer submergence range of three to four feet will assure the integrity of the suppression chamber when subjected to post-LOCA suppression pool hydrodynamic forces.

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