ML20197G552
| ML20197G552 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 10/28/1998 |
| From: | CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20197G504 | List: |
| References | |
| PROC-981028, NUDOCS 9812070295 | |
| Download: ML20197G552 (73) | |
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DAVIS-BESSE NUCLEAR POWER STATION UNIT NO.1 TECHNICAL REQUIREMENTS MANUAL
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DAVIS-BESSE i
NUCLEAR i
POWER STATION UNIT NO.1 TECHNICAL l
REQUIREMENTS MANUAL l
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DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 TECHNICAL REQUIREMENTS MANUAL 0V PAGE/ REVISION INDEX E8gg Revision /Date of Revision a
8 10/28/98 a-1 8
10/28/98 a-2 8
10/28S 8 a-3 8
10/28/98 b
8 10/28/98 I
8 10/28/98 II 8
10/28/98 III 8
10/20/98 1-1 1
02/2I/96 1-2 7
08/12/98 1-3 8
10/28/98 gspGUIUN 1-4 8
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02/07/96 00C0EUM
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[hha 3/4 0 '.
1 02/21/96 3/40-2 8
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02/07/96
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1 Cll
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n 3/43-2 8
10/28/98 3/43-3 8
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10/28/98 3/43-6' 8
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DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 TECHNICAL REQUIREMENTS MANUAL t
PAGE/ REVISION INDEX Eggg Revision /Date of Revision 3/4 3-17 8
10/28/98.
3/4 3-18 8
10/28/98 3/4 3-19 8
10/28/98 3/4 3-20 8
10/28/93 3/43-21 8
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05/22/ %
3/49-1 7
08/12/98 O
3/4 9-2 7
08/12/98 3/4 10-1 0
02/07/96 3/4 11-1 0
02/07/96 5-1 5
01/06/97 i
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j a-1 REV. 8 10re8/98 O
1 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 i
TECHNICAL REQUIREMENTS MANUAL i
REVISION /UCN/PAGE INDEX I
I Revision UCN Number ERRE I
96-010T l-1 3/4 0-1 i
3/4 0-2 I
3/434 2
96-018T I
3/48-1 3/48-2 3
96-044T 3/4 8-1 4
96-066T 5-1 5
96-126T 5-1 6
97-011T 1-2 l
l-3 3/4 0-2 3/44-2 97-057T 3/43-1 3/4 3-la 7
97-031T I
1-2 1-3 3/4 9-1 l
3/49-2 8
97-091T I
II l
Ill 1-3 1-4 3/40-2 3/43-1 3/43-2 3/43-3 3/43-4 3/43-5 3/43-6 J
a-2 REV. 8 10/28/98
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1 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 TECHNICAL REQUIREMENTS MANUAL REVISION /UCN/PAGE INDEX Revision UCN Number Pagg 8(cont) 97-091T 3/43-7 3/43-8 3/43-9 3/4 3-10 3/4 3-11 3/4312 3/4 3-13 3/4 3-14 3/4 3-15 3/4 3-16 3/4 3-17 3/4 3-18 3/4 3-19 3/4 3-20 3/4 3-21 3/4 3-22 O
a-3 REV. 8 10/28/98 m-
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i DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 i
TECHNICAL REQUIREMENTS MANUAL O
l TABLE INDEX Table Number Title Esat 3.3-2 REACTOR PROTECTION SYSTEM INSTRUMENTATION 3/43-2 RESPONSE TIMES i
3.3-5 SAFETY FEATURES SYSTEM RESPONSE TIMES 3/43-5 33-7 SEISMIC MONITORING INSTRUMENTATION 3/4 3-18 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION 343-21 3 3-13 STEAM AND FEEDWATER RUPTURE CONTROL SYSTEM 3/4 3-14 RESPONSE TIMES O
1 I
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l DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1
'ECHMICAL REQUIREMENTS MANUAL lO TABLE OF CONTENTS 1
l SECTION PAGE I
1.0 USE AND APPLICATION................................
1-1 l
2.0 S AFETY LI MITS........................................
2-1 i
3.0 LIMITING CONDITIONS FOR OPERATION APPLICABILITY...
3/40-1 4.0 SURVEILLANCE REQUIREMENT APPLICABILITY..........
3/40-1 3/4.1 REACTIVITY CONTROL SYSTEMS......................
3/41-1 3/4.2 POWER DISTRIBUTION LIMITS.........................
3/42-1
-3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION......
3/43-1
! / }-
3/4.3.2 SAFETY SYSTEM INSTRUMENTATION 3.3.2.1 SAFETY FEATURES ACTIJATION SYSTEM.........
3/43-4 3.3.2.2 STEAM AND FEEDWATER RUPTURE CONTROL SYSTEM 3/4 3-13 INSTRUMENTATION l
3/4.3.3 MONITORING INSTRUMENTATION 3.3.3.3 SEISMIC INSTRUMENTATION........
3/4 3-16 j
3.33.4 METEOROLOGICAL INSTRUMENTATION........
3/4 3-20 3/4.4 REACTOR COOLANT SYSTEM 3.4.11 REACTOR COOLANT SYSTEM VENTS..
3/44-1 l
3/4.5 EMERGENCY CORE COOLING SYSTEMS.................
3/45-1 l~
3/4.6 CONTAINMENT SYSTEMS..............
3/46-1 l
l 3/4.7 PLANT S YSTEMS..................................
3/47-1 l
1 REV. 8 10/28/98 O
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I DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 TECHNICAL REQUIREMENTS MANUAL i
(
TABLIE OF CONTENTS (continued) i i
SECTION PAGE 3/4.8 ELECTRICAL SYSTEMS l
3.8.1 A. C. SOURCES-OPERATING.........................
3/48-1 l
3/4.9 REFUELING OPERATIONS 3.9.5 COMMUNICATIONS...............................
3/49-1 l
3/4.10 SPECIAL TEST EXCEPTIONS.............................
3/4 10-1 l
3/4.11 RADIOACTIVE EFFLU~ENTS..............................
3/4 11-1 5.0 ADMINISTRATIVE CONTROLS...........................
5-1 I
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DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 TECHNICAL REQUIREMENTS MANUAL INTRODUCTION a
Based on the NRC's Final Policy Statement on Technical Specification Improvements for nuclear power plants, and 10 CFR 50.36 as amended in Final Rule published in the Federal Register dated July 13, I
1995, certain requirements may be relocated from the Operating License Technical Specifications to other licensee controlled documents. In an effort to centralize the requirements relocated from the Technical Specifications and to ensure the necessary administrative controls are applied to these 4
requirements, the Davis-Besse Technical Requirements Manual (TRM) has been developed.
The TRM provides one location for relocated items in a consistent format. The TRM retains the current Technical Specification numbering for relocated items with one exception. This exception is for the i
BASES section. Instead of being in a separate section, the BASES immediately follows the LIMITING CONDITION FOR OPERATION (LCO) and SURVEILLANCE REQUIREMENTS. 'Ihe TRM contains its own DEFINITIONS which may not be the same as the Operating License Technical Specification i
DEFINITIONS.
REGULATORY STATUS /REOUIREMENTS 2
Although the TRM itselfis not legally binding like the Operating License Technical Specifications, the requirements in the TRM are part of the licensing basis. Furthermore, the TRM is incorporated by i
reference in the Updated Safety Analysis Report (USAR) and is considered to be part of the USAR.
Violations of the TRM requirements should be documented by the PCAQ process.
These controls are in place because the purpose of relocating the requirements for Technical Specifications is not to reduce the level of control on the items but to provide flexibility for change under the 10CFR50.59, Safety Review / Evaluation, process.
Deviations from the TRM will be screened for reportability in accordance with the PCAQ pro.:ess.
l CHANGES TO THE TRM Design modifications, procedure changes, license amendments, etc. have the potential to affect the TRM. If this occurs, the initiating department must follow the administrative controls in NG-NS-00806, 2
i
" Preparation and Control of USAR Changes."
1 lli REV. 8 10/28/98 i
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l 1.0 USE AND APPLICATION
/*
1.1 Definitions The defined terms of this section appear in capitalized type and are applicable throughout the Technical l
Requirements Manual.
Ican Definition e
ACTIONS ACTIONS shall be those additional requirements specified as corollary statements to each principal requirement I
and shall be part of the requirements.
CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment,
[
as necessary, of the channel output such that it responds with necessary range and accuracy to known values of the
]
parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and l
shall include the CHANNEL FUNCTIONAL TEST.
l CHANNEL CALIBRATION may be performed by any l
series of sequential, overlapping or total channel steps such that the entire channel is calibrated.
O CIIANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This l
determination shall include, where possible, comparison of j
the channel indication and/or status with other indications j
and/or status derived from independent instrument channels measuring the same parameter.
CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:
a.
Analog channels - the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.
1 l
- b. Bistable channels - the injection of a simulated 4
signal into the channel sensor to verify OPERABILITY including alarm and/or trip functions.
1-1 REV.1 02/21/96 i
CORE ALTERATION CORE ALTERATION shall be the movement of any l
fuel, sources, or reactivity control components, within I
i the reactor vessel with the vessel head removed and I7 j
fuel in the vessel. Suspension of CORE I
ALTERATIONS shall not preclude completion of I
movement of a component to a safe position.
FREOUENCY NOTA 3ON The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined below.
NOTATION FREQUENCY S
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
W At least once per 7 days.
M At least once per 31 days.
Q At least once per 92 days.
SA At least once per 6 months.*
E At least once per 18 months.*
R At least once per 24 months.*
S/U Prior to each reactor startup.
N/A Not applicable
'6 months is defined to be 184 days,18 months is defined to be 550 days, and 24 months is defined to be 730 days.
1-2 REV. 7 08/12/98
!~
OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be l
OPERABLE or have OPERABILITY when it is capable
' p of performing its specified function (s). Implicit in this d
definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment, that are required for the system, subsystem, train, component or device to perform its function (s), are also capable of performing their related support function (s).
OPERATIONAL MODE An OPERATIONAL MODE shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature as specified below.
MODE TITLE REACTIVITY
% RATED AVERAGE CONDITION, Keff THERMAL POWER
- COOLANT TEMPERATURE 1
POWER OPERATION 2 0.99
>5%
2 280*F 2
STARTUP 2 0.99 s5%
2 280 F 3
HOT STANDBY
<0.99 0
2 280 F 4
110T SHUTDOWN
< 0.99 0
280*F > Tavg > 200 F 5
COLD SHUTDOWN
< 0.99 0
s 200*F 6
REFUELING" s 0.95 0
s 140 F
- Excluding decay heat.
" Reactor vessel head unbolted or removed and fuel in the vessel.
REACTOR PROTECTION The REACTOR PROTECTION SYSTEM RESPONSE TIME shall be l SYSTEM RESPONSE TIME that time interval from when the monitored parameter exceeds its trip l8 setpoint at the channel sensor until power interruption at the control l
rod drive breakers.
1-3 REV.8 10/28/98 i
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REFUELING INTERVAL A R1 FUELING INTERVAL is a period of time s 730 days.
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SAFETY FEATURE RESPONSE The SAFETY FEATURE RESPONSE TIME shall be that time i
TIME interval from when the monitored parameter exceeds its SFAS I
actuation setpoint at the channel sensor until the safety features i
equipment is capable of performing its safety function (i.e., the valves 18 travel to their required positions, pump discharge pressures reach their I required values, etc.). Times shall include diesel generator staning and I sequence loading delays where applicable.
I STEAM AND FEEDWATER The STEAM AND FEEDWATER RUPTURE CONTROL SYSTEM
[
RUPTURE CONTROL SYSTEM RESPONSE TIME shall be that time interval from when the lg RESPONSE TIME monitored parameter exceeds its SFRCS actuation setpoint at the j
channel sensor until the equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).
g l
-Jy 1-4 REV. 8 10/28/98 l
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. -.... _............ =....... _.
i 2.0 SAFETY LIMITS I
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2-1 REV. 0 02/07/96
l APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the specified surveillance 1
interval.
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3/4 0-1 REV.1 02/21/96 l
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i APPLICACILITY 4
JASES i
4.0.2 The provisions of this requirement provide allowable tolerances for performing surveillance i
activities beyond those specified in the nominal surveillance interval. These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations. The phrase "at least" associated with a surveillance frequency does not negate this allowable tolerance value and permits the performance of more frequent surveillance activities.
'Ihe allowab'. tolerance for performing surveillance activities is sufficiently restrictive to ensure that the reliability associated with the surveillance activity is not significantly degraded beyond that obtained from the nominal specified interval. It is not intended that the allowable tolerance be used as a convenience to repeatedly schedule the performance of surveillances at the allowable tolerance limit.
The allowable tolerance for prforming surveillance activities also provides flexibility to accommodate the length of a fuel cycle for surveillances that are specified to be performed 18 at least once each REFUELING INTERVAL. It is the intent that REFUELING INTERVAL surveillances be performed in an OPERATIONAL MODE consistent with safe plant operation.
3/4 0-2 REV. 8 10/28/98
l 3/4.1 REACTIVITY CONTROL SYSTEMS I
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i 3/4.2 POWER DISTRIBUTION LIMITS O
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3/42-1 REV. 0 02/07/96
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l 3/4,3 INSTRUMENTATION 3rd.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LCO 33.1.1
'Ihe Reactor Protection System (RPS) instrumentation channels shall be l
I OPERABLE with REACTOR PROTECTION SYSTEM RESPONSE TIMES as jg shown in TRM Table 3.3-2.
l APPLICABILITY: As shown in Technical Specification Table 3.31.
l8 l
ACTIONS l
CONDITION REQUIRED ACTION COMPLETION TIME I
REACTOR PROTECTION As shown in Technical Specification As shown in Technical l8 SYSTEM RESPONSE TIME Table 3.3-1.
Specification Table 33-1.
l outside limits.
l SURVEILLANCE REQUIREMENT SURVEILLANCE FREQUEW"r 4.3.1.1 REACTOR PROTECTION SYSTEM RESPONSE TIME in accordance with Techn. cal 18 I
(
of each reactor trip function shall be demonstrated in Specification Surveillance I
accordance with Technical Specification Surveillance Requirement 4.3.1.1.3.
I J
Requirement 4.3.1.1.3.
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i TATLE 3.3-2 i
REACTOR PROTECTION SYSTEM INSTRUMENTATION RESPONSE TIMES l
FUNCTIONAL UNIT RESPONSE TIMES" i
(seconds)
- 1. Manual Reactor Trip Not Applicable
- 2. High Flux
- s 0.266
- 3. RC High Temperature Not Applicable
- 4. Flux - AFlux - Flow * - Variable Flow s 1.77
- Constant Flow s 0.266 8
- 5. RC Low Pressure s 0.341
- 6. RC High Pressure s 0.34I ha
- 7. RC Pressure - Temperature - Constant Temperature Not Applicable
- 8. High Flux / Number of Reactor Coolant Pumps On*
s 0.631***
- 9. Containment High Pressure Not Applicable Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
Including sensor (except as noted), RPS instrument delay and the breaker delay.
A 0.24 sec delay time has been assumed for pump monitor.
l 3/43-2 REV. 8 10/28/98 n
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3/4,3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The measurement of response time at the specified frequencies provides assurance that the RPS action function associated with each channel is completed within the time limit assumed in the safety analyses. No g
credit was taken in the analyses for those channels with response times indicated as not applicable.
1 Response time may be demo astrated by any series of sequential, overlapping or total channel test measurements provided tha. such tests demonstrate the total channel response time as defined. Sensor response time verification nay be demonstrated by either 1) in place, onsite or offsite test measurements or
- 2) utilizing replacement a msors with certified response times.
a l
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3/4.3 INSTRUMENTATION i
3i4.3.2 SAFETY SYSTEM INSTRUMENTATION 3.3.2.1 Safety Features Actuation System LCO 3.3.2.1 The Safety Features Actuation System (SFAS) functional units shall be l8 OPERABLE with RESPONSE TIMES as shown in TRM Table 3.3-5.
l l
APPLICABILITY: As shown in Technical Specification Table 3.3-3.
l8 ACTIONS l
CONDITION REQUIRED ACTION COMPLETION TIME SAFETY FEATURES in accordance with Technical in accordance with Technical l8 RESPONSE TIME outside Specification 3.3.2.1.
Specification 3.3.2.1.
l limits.
l SURVEILLANCE REQUIREMENT SURVEILLANCE FREQUENCY 4.3.2.1 The SAFETY FEATURES RESPONSE TIME of each SFAS In accordance with Technical l8
[N function shall be demonstrated in accordance with Technical Specification Surveillance l
Specification Surveillance Requirement 4.3.2.1.3.
Requirement 4.3.2.1.3.
l 3/4 3-4 REV. 8 10/28/98 f~%
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TABLE 3.3-6 SAFETY FEATURES SYSTEM RESPONSE TIMES l
INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS l
l l
- 1. Manual a.
Fans l
1.
Emergency Vent Fan NA 2.
Containment Cooler Fan NA l
b.
ECCS Room NA i
2.
Emergency Ventilation NA 3.
Containment Air Sample NA 4.
DELETED NA l
5.
Penetration Room Purge NA.
c.
Control Room HV & AC Units NA d.
High Pressure injection 1.
High Pressure injection Pumps NA l
2.
High Pressure injection Valves NA e.
Component Cooling Water x
1.
Component Cooling Water Pumps NA 2.
Component Cooling Aux. Equip. Inlet Valves NA 3.
Component Cooling to Air Compressor Valves NA f.
Service Water System 1.
Service Water Pumps NA 2.
Service Water From Component Cooling Heat NA Exchanger Isolation Valves g.
Containment Spray isolation Valves NA h.
Emergency Diesel Generator NA i.
Containment Isolation Valves 1.
Vacuum Relief NA 2.
Normal Sump NA 3.
RCS Letdown Delay Coil Outlet NA 4.
RCS Letdown High Temperature NA I
l 3/4 3-5 REV. 8 10/28/98 i O t
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i TABLE 3.3-6 iContinued)
I SAFETY FEATURES SYSTEM RESPONSE TIMES O
l' INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS i
i.
Containment Isolation Valves (cont'd) 5.
Pressurizer Sample NA 6.
Service Water to Cooling Water NA f
7.
Vent Header NA 3
8.
Drain Tank NA 9.
Core Flood Tank Vent NA
- 10. Core Flood Tank Fill NA l
- 11. Steam Generator Sample NA
- 12. Quench Tank NA
- 13. Emergency Sump NA E
- 14. RCP Seal Return NA 3
- 15. Air Systems NA 16.
N System NA 2
- 17. Quench Tank Sample NA l-
- 18. RCP Sealinlet NA i - g
- 19. Core Flood Tank Sample NA
- 20. RCP Standpipe Demin Water Supply NA 21.
Containment H Dilution inlet NA 2
Containment H Dilution Outlet NA 22.
2 j.
BWST Outlet Valves NA k.
Low Pressure injection 1.
Decay Heat Pumps NA 2.
Low Pressure Injection Valves NA 3.
Decay Heat Pump Suction Valves NA 4.
Decay Heat Cooler Outlet Valves NA 5.
Decay Heat Cooler Bypass Valves NA 1.
Containment Spray Pump NA 1
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3/4 3-6 REV. 8 10/28/98 l
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TABLE 3.3-5 (Continued)
SAFETY FEATURES SYSTEM RESPONSE TIMES
)
INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS m.
Component Cooling Isolation Valves l
1.
Inlet to Containment NA 2.
Outlet from Containment NA 3.
Inlet to CRDM's NA 4.
CRDM Booster Pump Suction NA 5.
Component Cooling from Decay Heat Coolers NA i
2.
Containment Pressure - High i
a.
Fans
- 1. Emergency Vent Fans s 25*
- 2. Containment Cooler Fans s 45*
b.
- 1. ECCS Room 5 75*
- 2. Emergency Ventilation 575*
- 3. Containment Air Sample s 30'
- 4. DELETED
- 5. Penetration Room Purge s75*
c.
Control Room HV &AC Units s!0*
d.
High Pressure injection 1.' High Pressure injection Pumps s30*
- 2. High Pressure injection Valves
$30*
e.
Component Cooling Water
- 1. Component Cooling ' Water Pumps 5180*
- 2. Component Cooling Aux. Equip. Inlet Valves s180*
- 3. Component Cooling to Air Compressor Valves
$180*
f.
Service Water System
- 1. Service Water Pumps
$45*
- 2. Service Water From Component Cooling Heat sNA*
Exchanger Isolation Valves g.
Containment Spray isolation Valves s80*
h.
Emergency Diesel Generator s15*
3/43-7 REV. 8 10/28/98
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TABLE 3.3-5 (Continued)
SAFETY FEATURES SYSTEM RESPONSE Tluss INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 2.
Containment Pressure - High (Continued) 1.
Containment Isolation Valves 1.
Vacuum Relief s 30' 2.
Normal Sump s 25*
3.
RCS Letdown Delay Coil Outlet s30*
4.
RCS Letdown High Temperature s 30' 5.
Pressurizer Sample s48*
6.
Service Water to Cooling Water s 45' 7.
Vent Header s 15' 4
8.
Drain Tank s 15' 9.
Core Flood Tank Vent s 15*
10.
Core Flood Tank Fill s 15*
11.
Steam Generator Sample s 15*
12.
Quench Tank 5 15*
13.
Emergency Sump NA' I
14.
RCP Seat ndum s 45'
_ A 15.
Air System 5 15'
- 16.
N System s 15' 2
17.
Quench Tank Sample s 35*
18.
RCP SealInlet s 17*
19.
Core Flood Tank Sample s 15' 20.
RCP Standpipe Demin Water Supply s 15' 21.
Containment H Dilution Inlet 5 75*
2 22.
Containment H Dilution Outlet -
s75*
2 j.
BWST Outlet Valves NA'
. k.
Low Pressure Injection 1.
Decay Heat Pumps s 30' 2.
Low Pressure injection Valves sNA*
3.
Decay Heat Pump Suction Valves sNA 4.
Decay Heat Cooler Outlet Valves s NA*
5.
Decay Heat Cooler Bypass Valves sNA*
3/4 3-8 REV. 8 10/28/98 l
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TABLE 3.3-5 (Continued) l s,
SAFETY FEATURES SYSTEM RESPONSE TIMES
(\\s) i i
INITIATING SIGNAL AND FUNCTION BESPONSE TIME IN SECONDS 3.
Containment Pressure-High-High a.
Containment Spray Pump s 80*
b.
Component Cooling Isolation Valves I
1.
Inlet to Containment s 25*
2.
Outlet from Containment s25*
3.
Inlet to CRDM's s35' 4.
CRDM Booster Pump Suction 535*
5.
Component Cooling from Decay Heat Cooler sNA*
4.
RCS Pressure-Low a.
Fans 1.
Emergency Vent Fans
$25*
2.
Containment Cooler Fans
$45' S
b.
ECCS Room 575*
2.
Emergency Ventilation s75*
/]
3.
Containment Air Sample 530*
t'
/
4.
DELETED v
5.
Penetration Room Purge 575*
c.
Control Room HV & AC Units s10*
d.
High Pressure injection 1.
High Pressure injection Pumps s30*
2.
High Pressure injection Valves 530*
e.
Component Cooling Water 1.
Component Cooling Water Pumps s180*
2.
Component Cooling Aux. Equipment inlet Valves s180*
3.
Component Cooling to Air Compressor Valves s180*
f.
Service Water System 1.
Service Water Pumps
$45*
2.
Service Water from Component Cooling Heat sNA*
Exchanger Isolation Valves 3/43-9 REV. 8 10/28/98 l
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TABLE 3.3-5 (Continued)
SAFETY FEATURES SYSTEM RESPONSE TIMf4 Ox INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 4.
RCS Pressure-Low (continued) g.
Containment Spray isolation Valves 580*
h.
Emergency Diesel Generator sl5' i.
Containment Isolation Valves 1.
Vacuum Relief 530*
2.
Normal Sump 525' 3.
RCS Letdown Delay Coil Outlet 530' 4.
RCS Letdown High Temperature 530' 5.
Pressurizer Sample 545' 6.
Service Water to Cooling Water 545' 3
7.
Vent Header
$15' 8.
Drain Tank 515' 9.
Core Flood Tank Vent 515*
10.
Core Flood Tank Fill sl5' 11.
Steam Generator Sample 515*
i
's j 12.
Quench Tank 515*
13.
Emergency Sump SNA*
14.
Air Syste as
$15*
N System 515*
15.
2 t
16.
Quench Tank Sample s35*
17.
Core Flood Tank Sample sl5' 18.
RCP Standpipe Demin Water Supply sl5' 19.
Containment H Dilution inlet s75*
2 Containment H Dilution Outlet s75*
20.
2 j.
BWST Outlet Valves NA*
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3/4 3-10 REV. 8 10/28/98
- O t
l TABLE 3.3-6 (Continued)
A' SAFETY FEATURES SYSTEM RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 5.
RCS Pressure-Low-Low a.
Low Pressure injection 1.
Decay Heat Pumps s30*
2.
Low Pressure injection Valves sNA*
3.
Decay Heat Pump Suction Valves sNA*
4.
Decay Heat Cooler Outlet Valves 5:NA*
5.
Decay Heat Cooler Bypass Valves sNA*
b.
Component Cooling Isolation Valves 3
1.
Auxiliary Equipment inlet 590*
2.
Inlet to Air Compressor s90*
3.
Component Cooling from Decay Heat Cooler
$NA*
1 c.
Containment isolation Valves l.
RCP Seal Retum 545*
2.
RCP SealInlet 517' 6.
DELETED i
TABLE NOTATION l
Diesel generator starting and sequence loading delays included when applicable. Response time limit includes movement of valves and attainment of pump or blower discharge pressure.
l l
3/4 3-11 REV. 810/28/98 i
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=-
3/4.3 INSTRUMENTATION
?
BASES 1
3/4.3.2.1 SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION l
The measurement of response time at the specified frequencies provides assurance that the SFAS action l
function associated with each channel is completed within the time limit assumed in the safety analyses. No l
l credit was taken in the analyses for those channels with response times indicated as not applicable.
l8 l
{
Response time may be demonstrated by any series of sequential, overlapping or total channel test l
measurements provided that such tests demonstrate the total channel response time as defined. Sensor l
response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or l
- 2) utilizing replacement sensors with certified response times.
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3/4 3-12 REV 8.10/28/98 10
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l 3/4.3 INSTRUMENTATION i
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3/4.3.2 SAFETYSYSTEMINSTRUMENTATION 3.3.2.2 Steam and Feedwater Rupture Control System Instrumentation l8 LCO 3.3.2.2 The Steam and Feedwater Rup.ture Control System (SFRCS) instrumentation l8 channels shall be OPERABLE with RESPONSE TIMES as shown in TRM l
Table 3.3-13.
l APPLICABILITY: Modes 1,2 and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME SFRCS RESPONSE TIME in accordance with Technical in accordance with Technical l8 outside limits.
Specification 3.3.2.2.
Specification 3.3.2.2.
l SURVEILLANCE REQUIREMENT SURVEILLANCE FREQUENCY O()'
4.3.2.2 'lhe SFRCS RESPONSE TIME of each SFRCS function shall in accordance with Technical l8 be demonstrated in accordance with Technical Specification Specification Surveillance l
Surveillance Requirement 4.3.2.2.3.
Requirement 4.3.2.2.3.
j 3/4 3-13 REV. 8 10/28/98 l
,O l
TABLE 3.3-13 STEAM AND FEEDWATER RUPTURE CONTROL SYSTEM RESPONSE TIMES t
ACTUATED EOUIPMENT RESPONSE TIME IN SECONDS
- 1. Auxiliary Feed Pump s 40
- a. Main Steam Low Pressure Channels s6
- b. Feedwater/ Steam Generator High 5 6.5 Differential Pressure Channels y
- 3. Main Feedwater Valves
- a. Main Control s8
- b. Startup Control s 13
- c. Stop Valve s 16
- 4. Turbine Stop Valves **
s1 bv I
The response time is to be the time elapsed from the monitored variable exceeding the trip setpoint until the l
MSIV is fully closed.
- The response time is to be the time elapsed from the main steam line low pressure trip condition until the TSV is fully closed.
l 3/4 3-14 REV. 8 10/28/98 O
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3/4.3 INSTRUMENTATION l
(~)'T
(._
BASES 3/4.3.2.2 STEAM AND FEEDWATER RUPTURE CONTROL SYSTEM INSTRUMENTATION The measurement of response time at the specified frequencies provides assurance that the SFRCS action l
function associated with each channel ic completed within the time limit assumed in the safety analyses. No l
credit was taken in the analyses for those channels with response times indicated as not applicable.
l l8 Response time may be demonstrated by any series of sequential, overlapping or total channel test l
measurements provided that such tests demonstrate the total channel response time as defined. Sensor l
response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or l
- 2) utilizing replacement sensors with cenified response times.
l l
1 The SFRCS response time for the turbine stop valve closure is based on the combined response times of l
main steam line low pressure sensors, logic cabinet delay for main steam line low pressure signals and l
closure time of the turbine stop valves. This SFRCS response time ensures that the auxiliary feedwater to the l unaffected steam generator will not be isolated due to a SFRCS low pressure trip during a main steam line l
A break accident.
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3/4 3-15 REV. 8 10/28/98 bG f
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3/4,3 INSTRUMENTATION 3i4.3.3 MONITORINGINSTRUMENTATION 3.3.3.3 Seismic Instrumentation l
\\
LCO 3.3.3.3 7he seismic monitoring instrumentation shown in TRM Table 3.3-7 shall be OPERABLE. 18 l
APPLICABILITY: At all times ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restoretheinoperable 30 Days instruments inoperable, instrument to OPERABLE status.
B. Required Action and B.1 Initiate a Potential Condition None associated Completion Time Adverse to Quality Report (PCAQR) of Condition A not met.
if one has not already been initiated.
SURVEILLANCE REQUIREMENT SURVEILLANCE FREQUENCY 4.3.3.3.1.a Perform CHANNEL CHECK for TRM Table 3.3-7 Monthly l8 Items l' and 3".
I 4.3.3.3.1.b Perform a CHANNEL FUNCTIONAL TEST 6 Months for TRM Table 3.3-7 Items I and 3.
l8 l
4.3.3.3.1.c Perform a CHANNEL CALIBRATION for TRM 18 Months l8 Table 3.3-7 Items Ic and Id (outside of containment), 2, and 3.
4.3.3.3.1.d Perform a CHANNEL CALIBRATION for TRM At least once each REFUELING 18 Table 3.3-7 Items la and Ib (inside containment).
INTERVAL (Continued) l I
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3/4 3-16 REV. 8 10/28/98 l bp
SURVEILLANCE REQUIREMENT (CONTINUED)
SURVEILLANCE FREQUENCY
]
rs 4.3.3.3.2 Each of the above seismic monitoring instruments Following a Seismic Event actuated:
a) shall be restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AND b) shall have a CHANNEL CALIBRATION performed within 5 days.
AND c) Each instrument actuated shall be analyzed to determine the magnitude of the vibratory ground motion. Prepare and submit a special report to the Commission within 10 days describing the magnitude, frequency, spxtrum, and resultant effect upon the facility features imponant to safety.
- Except seismic trigger
- *With cabinet room indication m
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l 3/4 3-17 REV. 8 10/28/98 l
TABLE 3,3-7 l
l SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENT i
t INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE 1.
Strong Motion Triaxial Accelerometers l
l a.
Containment Concrete Foundation, Elev. 565
- Ig
]
b.
Containment Interior Secondary Shield Wall, Elev. 653
- Ig I
l' c.
Auxiliary Building Basement l
Floor, Elev. 545
- Ig 1
d.
Station site - Minimum of 300
- Ig i
[
feet from containment vessel within the site boundary
- 2.
Peak Recording Accelerometers I
l a.
Shield Building Top, Minimum
- Ig 1
l Elev. 812 l d b.
Auxiliary Building Roof, Elev.
I 660
- Ig I
c.
Control Room. Elev. 623
- Ig I
l 3.
Seismic Trigger a.
Station site - Minimum of 300 feet from containment 1-10 Hz*
1" i
vessel within the site 0.00$g - 0.02g* "
boundary l
- Minimum Frequency Response Range
" With cabinet room indication j
- " Actuation Range l
3/4 3-18 REV. 810/26/98 i
3/4.3 INSTRUMENTATION N
BASES i
3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event so that the response of those features important to safety may be evaluated. This capability is required to permit comparison of the measured response to that used in the design basis for the facility. This instrumentation is consistent with the recommendations of Regulatory Guide 1.12
" Instrumentation for Eanhquakes," April 1974.
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3/4.3 lNSTRUMENTATION 3i4.3.3 MONITORINGINSTRUMENTATION l
3.3.3.4 MeteorologicalInstrumentation 1
LCO 3.3.3.4 The meteorological monitoring instrumentation channels shown in TRM l8 Table 3.3-8 shall be OPERABLE.
1 APPLICABILITY: At all times ACTIONS f
CONDITION REQUIRED ACTION COMPLETION TIME 1
l A. One or more required A.1 Restore the inoperable channel to 7 Days channels inoperable.
OPERABLE status.
i l
B. Required Action and B.1 initiate a Potential Condition None associated Completion Time Adverse to Quality Report of Condition A not met.
(PCAQR)if one has not already l
been initiated.
O 1 O l
SURVEILLANCE REQUIREMENT SURVEILLANCE FREQUENCY 4.3.3.4.a Perform CHANNEL CHECK for TRM Table 33-8 24 Hours l8 instruments.
4.3.3.4.b Perform a CHANNEL CALIBRATION for TRM 6 Months 18 l
Table 3.3-8 instruments.
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l 3/4320 REV. 8 10/28/98
.O l
IABLE 3 34 f
1 METEOROLOGICAL MONITORING INSTRUMENTATION x
MINIMUM INSTRUMENT LOCATION OPERABLE
(
l.
WIND SPEED L
a.
Nominal Elev.
612 1
b.
Nominal Elev.
827 1
2.
WIND DIRECTION l
l a.
Nominal Elev.
612 I
i b.
Nominal Elev.
827 1
3.
AIR TEMPERATURE-DELTA T a.
Nominal Elev.
827-612 1
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~
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l 3/4 3-21 REV. 8 10/28/98 I
.=
d 3/4.3 INSTRUMENTATION BASES 3/433.4 METEOROLOGICA.L INSTRUMENTATION i
i The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release or radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public. This instrumentation is consistent with the recommendations of Regulatory Guide 1.23 "Onsite Meteorological Programs," February 1972.
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3/4 3-22 REV. 8 10/28/98 1
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3/4.4 REACTOR COOLANT SYSTEM C/ '
3.4.11 Reactor Coolant System Vents LCO 3.4.11 The following Reactor Coolant System vent paths shall be OPERABLE:
)
a.
Reactor Coolant System Loop I with vent path through valves RC 4608A and RC 4608B.
b.
Reactor Coolant System Loop 2 with vent path through valves RC 4610A and RC 4610B.
c.
Pressurizer; with vent path through EITHER valves RClI and RC2A (PORV) OR valves RC 239A and RC 200.
APPLICABILITY: Modes 1,2,3 i
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One Vent Path A.1 Restore the inoperable vent 30 Days Inoperable path to OPERABLE status.
p B. Two Vent Paths B.1 Restore at least one of the inoperable 72 Hours
('
Inoperable vent paths to OPERABLE status.
C. Three Vent Paths C.1 Restore at least two of the inoperable 72 Ilours Inoperable vent paths to OPERABLE status.
D. Required Action and D.1 Be in HOT STANDBY.
6 Hours Associated Completion AND time of Action A E B D.2 Be in HOT SHUTDOWN.
Within the following 30 2 C not met.
Hours 3/4 4-1 REV. 0 02/07/96 J
1 l
SURVEILLANCE REQUIREMENT i
l'~'
i (,,)
SURVEILLANCE FREQUENCY 4.4.11.1 Verify all manual isolation valves in each vent path are At least once each locked in the open position.
REFUELING INTERVAL 6
l l
4.4.11.2 Cycle each valve in the vent path through at least one At least once each complete cycle of full travel from the Control Room.
REFUELING INTERVAL 4.4.113 Verify flow through the reactor coolant vent system vent At least once each l
paths.
REFUELING INTERVAL l
1 l
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(
3/4 4-2 REV,6 02/05/98
3/4.4 REACTOR COOLANT SYSTEM
')
HASES i
3/4.4.11 IIIGII POINT VENTS i
The Reactor Coolant System high point vents are installed per NUREG-0737 item II.B.1 requirements.
The operability of the system ensures capability of venting steam or nonconuensable gas bubbles in the reactor cooling system to restore natural circulation following a small break loss of coolant accident.
I i
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3/44-3 REV. 0 02/07/96
3/4.5 EMERGENCY CORE COOLING SYSTEMS O
THIS PAGE INTENTIONALLY LEFT BLANK 4
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3/4.6 CONTAINMENT SYSTEMS O
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l 3/46-1 REV. 0 02/07/96
3/4.7 PLANT SYSTEMS O
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3/4.8 ELECTRICAL POWER SYSTEMS
/'~'
3/4.8.1 A. C. SOURCES-OPERATING C}
LCO 3.8.1.3 Two separate and independent emergency diesel generators shall be OPERABLE in accordance with Technical Specification 3.8.1.1.
APPLICABILITY: Modes 1,.2,3, and 4, during performance of preplanned maintenance activities.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One emergency diesel A.1 Verify that the Station Blackout Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to l3 generator removed from Diesel Generator (SBODG) has removing the emergency diesel service for preplanned passed the monthly SBODG test generator from service for maintenance.
DB-SC-04271 within the last 30 preplanned maintenance.
days.
AND A.2 Perform SR 4.8.1.3.
Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to l3 removing the emergency diesel generator from service for i) preplanned maintenance and once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter until the emergency diesel generator is returned to OPERABLE.
B. Required Action and B.1 Initiate a Potential Condition None associated Completion Adverse to Quality Report Time of Condition A not (PCAQR).
m et.
SURVEILLANCE REQUIREMENT SURVEILLANCE FREQUENCY 4.8.1.3 Verify the SBODG is capable of connection Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to removing an l3 -
to the essential bus associated with an emergency diesel generator from service emergency diesel generator removed from for preplanned maintenance and once per service for preplanned maintenance.
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter until the emergeacy diesel generator is retumed to OPERABLE.
O 3/4 8-1 REV. 3 06/21/96 V
I
l 3/4.8 ELECTRICAL POWER SYSTEMS i
BASES i
3/4.8.1.3 A. C. SOURCES - OPERATING l
i l
The ACTIONS provide verification that the Alternate A. C. (AAC) power source, the Station Blackout Diesel Generator, is functional and capable of being connected to the safety bus associated with the inoperable Emergency Diesel Generator. These actions are consistent with the NRC criteria for ensuring that the probability of a core damage accident given a Station Blackout event is not significantly increased due to the performance of Emergency Diesel Generator preventive maintenance during power operations.
l These actions are applicable only when an Emergency Diesel Generator becomes inoperable for the performance of preplanned maintenance activities. (Reference NRC Safety Evaluation for License Amendment 206, dated February 26,1996).
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1
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(mj%
j 3/48-2 REV. 2 05/22/96 l
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l 3/4.9 REFUELING OPERATIONS O
3.9.5 Communications LCO 3.9.5 Direct communications shall be maintained between the control room
[7 and personnel at the refueling station.
l APPLICABILITY: During CORE ALTERATIONS.
l7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l
Direct communications between the Suspend CORE ALTERATIONS Immediately l
control room and personnel at the l7 i
refueling station cannot be l
l maintained.
l l
g SURVEILLANCE REQUIREMENT SURVEILLANCE FREQUENCY l
4.9.5 Demonstrate direct communications between the control Within I hour prior to the start of I
room and personnel at the refueling station.
CORE ALTERATIONS 17 AND I
l Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter, I
l l
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l 3/4 9-1 REV. 7 08/12/98 l
3/4.9 REFUELING OPERATIONS i
j BASES 1
1 3/4.9.5 COMMUNICATIONS The requirements for communications capability ensures that refueling station personnel can be l
l promptly informed of significant changes in the facility status or core reactivity condition during l7 l
l
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1 3/4 9-2 REV. 7 08/12/98 O
. _ =.
3/4.10 SPECIAL TEST EXCEPTIONS O
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3/4 10-1 REV. 0 02/07/96 O
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aw-Ai 3/4.11 RADIOACTIVE EFFLUENTS O
O
" "^"' ' " ' ' " ' ' " ^ " ' '""'" ' ^ * "
O 3,4,,-,
aev.o 02,07,ee
~ -.
i 5.0 ADMINISTRATIVE CONTROLS m
FACILITY STAFF OVERTIME Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a nominal,40-hour week while the plant is 5
operating. Selected positions may work up to 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts under a rotating work week schedule, with a nominal 40-hour work week. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed:
An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, a.
excluding shift turnover time.
b.
An individual should not be perrnitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any j
24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period, not more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in a seven day period, all excluding shift turnover time.
c.
A break of at least eight hours should be allowed between work periods, including shift turnover time.
O d.
Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.
I 5-1 REV. 5 01/06/97 k