ML20088A640
| ML20088A640 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 04/09/1984 |
| From: | Schnell D UNION ELECTRIC CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| ULNRC-792, NUDOCS 8404120330 | |
| Download: ML20088A640 (185) | |
Text
_ _
UNION ELECTRIC COM PANY 1901 GRATIOT STREET ST. Louis, MISSOURI 4dl 9,
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Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Denton:
ULNRC-792 DOCKET NUMBER 50-483 1
CALLAWAY PLANT, UNIT 1 CALLAWAY TECHNICAL SPECTFICATIONS
References:
- 1) D. G. Eisenhut letter to D. F. Schnell dated March 8, 1984
- 2) ULNRC-787 dated April 5, 1984 Attachments: 1). Summary Listing of Attached Specifications
- 2) Specifications Resolved Since Final Draft
' 3)' Specifications'Being Appealed
- 4) Specifications Which Have Not Been Resolved Reference 1 transmitted the Callaway Technical Specifications in final draft form.
Union Electric has reviewed this final-draft and has included as attachments to this letter those changes which are necessary so that the Callaway Technical ~ Specifications accurately reflect the plant design and operating program.
Attachment'1 is a summary listing of the specifications contained in Attachments'2, 3 and 4. contains those specifications which have been revised since the final draft and for which UE and NRC have reached mutual agreement on the wording and substance. contains those speci-fications which are being appealed as noted in Reference 2.
, contains those specifications which have not been resolved.
Tables 3.3-10 and 3.3-11 include, as item 19,
-Reactor Coolant Radiation Level Instrumentation.
These are footnoted as not needed until startup following the first
-refueling.
These instruments are not in.the Callaway design.
The SNUPPS submittal.on Regulatory Guide 1.97, Revision 2 was made by SLNRC 82-031 dated July 6, 1982 and subsequently incorporated in SNUPPS FSAR Revision 10.in September, 1982.
The review of this-information is still~in_ progress and a license condition will assure proper resolution.
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8404120330 840409
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ULNRC-792 Mr. Harold R. Denton Page 2 The second unresolved specification (3/4 5.5) concerns the boron concentration in the Boron Injection Tank (BIT).
A proposed specification change with justification was submitted by SLNRC 84-0046 dated March 20, 1984.
This is under review by NRC.
The Callaway Technical Specifications reflect an acceptance criteria of 12 seconds for the diesel generator start time from ambient conditions.
FSAR text changes which support this will be provided by letter prior to fuel load and will be incorporated in the next SNUPPS FSAR revision.
Except as noted herein, in my judgement, the Callaway Technical Specifications accurately reflect the plant design and_ operating program as described in the FSAR and other information on our docket.
Very truly yours, Donald F. Schnell DFS/cfs Attachments cc: J. J. Holonich
STATE OF MISSOURI )
)
Donald F.
Schnell, of lawful age, being first duly sworn upon oath says that he is Vice President-Nuclear and an officer of Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are-true and correct to the best of his
- knowledge, information and_ belief.
m By Donald F.
Schnell Vice President Nuclear h2dd day of 1984!
SUBSCRIBED and sworn'to before me this 1
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Glenn L. Koester Vice President Operations Kansas Gas & Electric P.O. Box 208 Wichita, Kansas 67201 Donald T. McPhee Vice President Kansas City Power and Light Company 1330 Baltimore Avenue Kansas City, Missouri 64141 Gerald Charnoff, Esq.
Shaw, Pittman, Potts'& Trowbridge 1800 M. Street, N.W.
Washington, D.C.
20036 Nicholas A. Petrick Executive Director SNUPPS 5 Choke Cherry Road Rockville, Maryland 20850 John H. Neisler Callaway Resident Office U.S. Nuclear Regulatory Commission RR$1 Steedman, Missouri 65077 William Forney Division of Projects and Resident Programs, Chief, Section lA U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Bruce Little Callaway Resident Office L
U.S. Nuclear Regulatory Commission RRil-1 Steedman, Missouri 65077
Page 1 of 3 ATTACIU4ENT 1 Summary Listing of Attached Specifications Item Page Agree Open Issue 1
III x
Editorial 2
IV x
Editorial 3
VII x
Editorial 4
IX x
Awaiting NRC Review Deletion of BIT 5
X x
Editorial 6
XI x
Editorial 7
XV x
Awaiting NRC Review Deletion of BIT 8
2-2 x
Editorial "f,#"
x(p2-7)
Incorporation of Callaway Setpoints 9
x 10 B2-5 x
Editorial 11 3/4 1-1 x
Editorial 12 3/4 1-4 x
Editorial 13 3/4 1-6 x
Editorial 14 3/4 1-7 x
Awaiting NRC Review Boration Systems 15 3/4.1-10 x
Editorial 16 3/4.1-11 x
Awaiting NRC Review Eoration Systems 17 3/4 1-12 x
Awaiting NRC Review Boration Systems 18 3/4 1-13 x
Awaiting NRC Review Boration Systems 19 3/4 1-14 x
Editorial Boration Systems 20 3/4 1-15 x
Editorial 3/4 1-13 21 thru 3/4 1-24 x
Numbering changes due to boration systems 22 3/4 1-15 x
Editorial 23 3/4 1-21 x
Editorial 24 3/4 1-22 x
Editorial 25 3/4 2-1 x
Editorial 26 3/4 2-8 x
Flow Uncertainties 27 3/4 2-9 x
Flow Uncertainties 28 3/4 2-10
?
Flow Uncertainties 29 3/4 2-11 x
Editorial 30 3/4 3-5 x
Editorial 31 3/4 3-10 x
Editorial 32 3/4 3-13 x
Inclusion of new action statement 33 3/4 3-14 x
Editorial i
34 3/4 3-17 x
Editorial l
35 3/4 3-19 x
Inclusion of solid state load sequencer 36 3/4 3-21 x
Inclusion of action statement for SSLS 3f3-22 x
Inclusion of Callaway Setpoints 36 Inclusionofsolidstateloadsekuencer 37 3/4 3-27 x
38 3/4 3-30 x
Editorial l
39 3/4 3-31 x
Editorial 40 3/4 3-32 x
Editorial 3
3 33 41 x
Editorial 3-37 42 3/4.3-37 x
Inclusion of solid state load sequencer 43 3/4 3-29 x
Radiation Monitoring Instrumentation setpoints e-
-n v
ATTACHMENT 1 Page 2 of I Item Page.
Agree Open Issue
~
l
'44 3/4 3-40 x
Radiation monitoring instrumentation setpoints 45 3/4 3-44 x
Seismic switch trip setpoints
'46 3/4 3-45 x
Exception to analog channel operational test L47 -
3/4 3-52 x
Editorial 48 x
Reg Guide 1.97 instrumentation thru 3/ }56 49
-3/4 3-61 x
Inclusion of ESF transforners/ Editorial 50 3/4 3-64 x
Editorial 51' 3/4.3-66 x
Editorial 52 3/4 3-67 x
Changes to clarify note 2 53 3/4 3-69 x
Inclusion of instrumentation 3.b thru 3.e
' 54 -
.3/4 3-70 x
Editorial 55 3/4 3-71 x
Editorial 56
-3/4 3-72 x
Inclusion of instrumentation 3.b thru 3.e 57 3/4 3-73 x
Editorial 58 3/4 3-74 x
Changes to clarify note 3 59 3/4 3-76 x
Inclusion of Turbine overspeed protection program 60 3/4 4-1 x
Editorial 61 3/4 4 x Editorial 62 3/4 4-l'3 x
Deletion of " Condition IV" from C.3 and C.4 63 3/4 4-14 x
Editorial 64 3/4 4-18 x
Editorial 65 3/4 4-26 x
Editorial 66 3/4 4-30 x
Editorial 67 3/4 4-31 x
Editorial 68 3/4 4-34 x
RHR suction relief valves for cold over-3/4 4-35 pressure protection appeal issue 69 3/4 5-3'
.x Editorial 70' 3/4 5-5 x
Editorial 71 3/4 5-6 x
RHR System flowrate
'72 3/4 5-8 x
Editorial 73
'3/4 5-9 x
Editorial 74
-3/4 5-10_
x Deletion of boron injection tank spec
-75.
3/4 5 x Editorial based.on Item 74 76
-3/4 6-2 x
' Editorial 77 3/4 6-3 x
Verbal agreement reached new words need review 78 3/4 6-5 x
UE verifying revised numbers 79 3/4 6-6 x
Revised containment internal pressure setpoints 3/4 6-8 80 x
Containment vessel structural integrity - appeal thru 6-10
'81-3/4 6-11 x
Increase. cont ventilation limit to 2000 hrs.
82 3/4 6-12 x
Increase limit to.05 La
.83 3/4_6-16 x
Containment isolation valves - appeal 84 3/4 6-20 x-Editorial
'85 3/4 6-21 x.
Editorial 86 3/4 6-22' x
Editorial 87 3/4 6-26 x
Editorial 88 3/4'6-29 x
Inclusion of breathing air supply valves 89
'3/4 6-30 x
Revision of hydrogen analyzer surveillance 90-3/4_6-31 x
Deletion of hydrogen purge
'91.
3/4 7-7 x
Editorial
- 3/4.7-8 x-Editorial
,.,-..-_.,,,,-_ _ -~ - -
ATTACHMENT 1 Page 3 of 3 Item Page*
Agree Open Issue 93 3/4 7-15
.x Editorial 94 3/4 7-27 x
Increasing volume of fire suppression tanks 95 3/4 7-30 x
Inclusion of ESF transformers in specification 96 3/4 7-33 x
Editorial 97 3/4 7-34 x
Editorial 98 3/4 7-35 x
Editorial' 99 3/4 7-36 x
Editorial 100 3/4 7-37 x'
Editorial 101 3/4 7-38 x
Editorial 102 3/4 8-2 x
Moved proposed surveillances to other specs 103 3/4 8-3 x
Change voltage valves / include new fuel spec
.104 3/4 8-4 x
Inclusion of new fuel oil spec 105
_3/4 8-5 x'
Change voltage values 106
. 3/.4. 8.6
.,.x
. Editor.ial..
107 3/4 8-7 Added statement for DG Start counting 108 3/4 8-8 x
RHR suction relief issue - appeal 109 3/4 8-10 x
Editorial-110 3/4 8-12 x
RHR suction relief issue - appeal 111 3/4 8 x Editorial and changes to values in thru 3/4 8-24 Table 3.8-1 112 3/4 9-15 x
Change to incorporate SER requirement 113 3/4 11-4 x
change to note 6 114-3/4 11-7 x
Added demin vessels to Loo 115 3/4 11-9 x
Editorial 116 3/4 11-11 x.
Editorial 117 3/4 11-16 x
Change 31 days to 7 days for surveillance 118 3/4 12-4 x
Editorial 119 3/4.12-8 x
Change to note 7 120 3/4 12-14 x
Editorial
'121 B3/4 1-2 x
Boration system under review by NRC
~121 B3/4 1-3 x
Boration system under review by NRC
'122 B3/4 2-4 x
Rod bow penalty 123 B3/4 2-5 x
RCS Flow Rate - Ventury Fouling 124 B3/4 4-3 x
Clarification on S/G Turg inspection 125 B3/4 4-15 x-RHR suction relief valves - appeal 126 B3/4 5-1 x
Editorial 127' B3/4 5-2 x
Deletion of BIT under review by NRC 128-B3/4 6-2 x
containment internal pressure 129 B3/4_6 x containment purge limit increase to 2000 hrs.
130 B3/4 6-4 x
Editorial 131' B3/4 12-2 x
Editorial 132 5-1 x
Editorial 133 6-1 x
' Editorial
-134 6-3,4 x
organizational chart changes 135' 6-6 x
Editor'ial 136 6-9 x
Turbine overspeed prote: tion program
'137 6-13 x-Temporary change review 138 6-15 x
. Turbine overspeed-protection program / editorial 139 6-16 x
Turbine overspeed protection program
- 140.'
6-19 x
Editorial 4
i ATTACIDIENT 2 Specifications Resolved Since Final Draft A
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111111:1r INDEX 5AFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1,, SAFETY LIMITS 2.1,1 REACTOR C0RE................................................
2-1 2.1.2 REACTOR COO LANT SYSTEM PRESSURE............... :.............
2-1 f
FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION..
2-2 FIGURE 2.1-2 (BLANK)..............................................
2-3 2.2 LIMITING SAFETY SYSTEM SETTINGS i
2.2.1. REACTOR TRIP SYSTEM INSTRUMENTATION +RtP-SETP0INTS.... '......
2-4 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS....
. 2-5 BASES l
I SECTION PAGE L
2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................
B 2-1 2.1.2 ~ REACTOR COOLANT SYSTEM PRESSURE.............................
B 2-2 1
l 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS...............
B 2-3 l:
i l
L
- CALLAWiv
'J'iIT '?
III 1
L
DRET INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued)
TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION.....................................
3/4 3-63 TABLE 4.3-8 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................
3/4 3-65 Radioactive Gaseous Effluent Monitoring Instrumentation..
3/4 3-67 TABLE 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION.....................................
3/4 3-68 TABLE 4.3-9 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................
3/4 3-71 4
3/4. 3. TURBINE OVERSPEED PROTECTION.........................
3/4 3-74 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation..............................
'3/4 4-1 Hot Standby..............................................
3/4 4-2 Hot Shutdown.............................................
3/4 4-3 Cold Shutdown - Loops Filled.............................
3/4 4-5 Cold Shutdown - Loops Not Fi11ed.........................
3/4 4-6 3/4.4.2 SAFETY VALVES Shutdown...............................................
3/4 4-7 0perating..............................................
3/4 4-8 3/4.4.3 PRESSURI2ER..............................................
3/4 4-9 3/4.4.4 RELIEF VALVES............................................
3/4 4-10 3/4.4.5 STEAM GENERATORS.........................................
3/4 4-11 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION.........................
3/4 4-16 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION.......................
3/4 4-17 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.......,........................
3/4 4-18 Operational Leakage......................................
3/4 4-19 CALLAWAY - UNI 1
VII
DRL7 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE PLANT SYSTEMS (Continued)
TABLE 3.7-3 STEAM LINE SAFETY VALVES PER L00P.....................
3/4 7-3 Auxiliary Feedwater System...............................
3/4 7-4 Condensate Storage Tank..................................
3/4 7-6 Specific Activity........................................
3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM...............................
3/4 7-8 Main Steam Line Isolation Valves.........................
3/4 7-9 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION..........
3/4 7-10 3/4.7.3 COMPONENT COOLING WATER SYSTEM...........................
3/4 7-11
-3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM...........................
3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK.......................................
3/4 7-13 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM................
3/4 7-14 3/4.7.7 EMERGENCY EXHAUST SYSTEM.................................
3/4 7-17 3/4.7.8 SNUBBERS.................................................
3/4 7-19 FIGURE 4.7-1 SAMPLING PLAN 2) FOR SNUBBER FUNCTIONAL TEST.........
3/4 7-24 3/4.7.9 SEALED-SOURCE CONTAMINATION..............................
3/4 7-25 3/4.7.10 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System............................
3/4 7-27 Spray and/or Sprinkler Systems...........................
3/4 7-30 Halon Systems............................................
3/4 7-32
-Fire Hose Stations.......................................
3/4 7-33 A.
TAB LE 3. 7 F I RE H0S E STATIONS....................................
3/4 7-34 3/4.7.11 FIRE BARRIER PENETRATIONS................................
3/4 7-36 3/4.7.12 AREA TEMPERATURE MONITORING..............................
3/4 7-37 e
CALLAWAv - UNIT !
X g
-e
DRIH INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE PLANT SYSTEMS (Continued) 5 TABLE 3.7-& AREA TEMPERATURE M0NITORING...........................
3/4 7-38 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 0perating................................................
3/4 8-1 TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE........................
3/4 8-7 Shutdown.................................................
3/4 8-8 3/4.8.2 D.C. SOURCES 0perating................................................
3/4 8-9 TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS.....................
3/4 8-11 Shutdown.................................................
3/4 8-12 3/4.8.3 ONSITE POWER DISTRIBUTION 0perating...........................'.....................
3/4 8-13 Shutdown.................................................
3/4 8-15 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent Protective Devices.....................................
3/4 8-16 TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES..................................
3/4 8-18 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION......................................
3/4 9-1 3/4.9.2 INSTRUMENTATION..........................................
3/4 9-2 3/4.9.3 DECAY TIME...............................................
3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS........................
3/4 9-4 3/4.9.S COMMUNICATIONS...........................................
3/4 9-5 CALLAWAY - UNIT 1 XI
DRUT (10ACCEPTAfM.E.
OPEtt.AT104 l
F 660 N
~
32400 PSIA w
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-+-
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g yw 2250 PSIA N
i c
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N w2000 PSIA
-'c-e
-'c'.1
'n%
'> 620 N
^.';'.
x
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'w "N
= i-i :
A m x\\
'1860 PSIA h
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. x 600
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. v 1:. '.'.
- rW 680 1 v s
1
,,A 560
'1
/
O 02 0.4 0.6 0.8 1.0 1.2 ACCEPTABLE FRACTION OF RATED THERMAL POWER-OPEr2AnoO FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - F.00R LOOPS IN OPERATION CALLAWAY - UNIT 1 2-2
TABfE 2.2-1 P
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E
4 TOTAL SENSOR ERROR
's FUNCTIONAL UNIT ALLOWANCE (TA)
Z Q
TRIP SETPOINT
' ALLOWABLE VALUE E
1.
Manual Reactor Trip N.A.
N.' A.
N.A.
N.A.
N.A.
G.
2.
Power Range, Neutron Flux
- a. High Setpoint
- 7. 5 4.56 0
$109% of RTP*
1112.3% of RTP*
i
- b. Low Setpoint 8.3 4.56 0
125% of RTP*
$28.3% of RTP*
3.
Power Range, Neutron Flux, 2.4 0.5 0
14% of RTP* with 16.3% of RTP* with High Positive Rate a time constant a time constant l
->2 seconds
>2 seconds J
4.
Power Range, Neutron Flux, 2.4 0.5 0
14% of RTP* with 16.3% of RTP* with
)
1 High Negative Rate a time constant a time constant E
>2 seconds
>2 seconds to 4
5.
Intermediate Range, 17.0
- 8. 41 '.
0 125% of RTP*
135.3% of RTP*
i Neutron Flux 6.
Source Range, Neutron Flux 17.0 10.01 0
$105 cps
$1.6 x 10s cps I
i 7.
Overtemperature AT 6.1
, 2.76 1.8 See Note 1 See Note 2 8.
Overpower AT 4.6 1.3 1.2 See Note 3 See Note 4 l.
9.
Pressurizer Pressure-Low 5.0 2.21 2.0
>1885 psig
>1874 psig...
i 10.
Pressurizer Pressure-High 7.5 4.96
- 1. 0 12385 psig 12400 psig 11.
Pressurizer Water Level-High b.0 2.18 2.0 192% of instrument.
193.8% of instrument' span span 12.
Reactor Coolant Flow-Low 2.5 1.0 1.5
>90% of loop
>89.2% of loop design flow **
design flow **
l
- RTP = RATED THERMAL POWER
- Loop design flow = 95,700 gpm
9 TABLE 2.2-1 (Continued)
E E
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 4
TOTAL SENSOR ERROR c
. FUNCTIONAL UNIT ALLOWANCE (TA)
Z M
TRIP SETPOINT ALLOWABLE VALUE z
Z
- 13. Steam Generator Water 23.5 21.18 2.0
>23.5% of narrow
>22.0% of narrow s
level Low-Low range instrument range instrument span span:
14.
Undervoltage - Reactor 7.7 1c33 0
>10584 Volts A.C
>10356 Volts A.C.
Coolant Ptaps a
15.
Underfrequency - Reactor 3.3 0
0
>57.2 Hz
>57.1 Hz
]
Coolant Pumps 16.
Turbine Trip a.
Low Fluid Oil Pressure N.A.
N.A.
N.A.
>598.94 psig
>539.42 psig
~
T b.
Turbine Stop Valve N.A.
N.A.
N.A.
>1% open
>1% open Closure 17.
Safety Injection Input-N.A.
N.A.
N.A.
N.A.
N.A.
from ESF g
4 k
- 4 i
i i
I i.
l t
}
z-
Q TABLE 2.2-1 (Continued)
F REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 4
TOTAL SENSOR ERROR g
FUNCTIONAL UNIT ALLOWANCE (TA)
Z_
Q TRIP SETPOINT ALLOWABLE VALUE Q
18.
Reactor Trip System Interlocks g
a.
Intermediate Range N.A.
N.A.
N.A.
>1 x 10 10 amps
>6 x 10 21 amps Neutron Flux, P-6 b.
Low Power Reactor Trips
7##
Block, P-7 1
1)
P-10 input N.A.
N.A.
N.A.
10% of RTP*
2.4% of RTP*
2)
P-13 input N.A.
N.A.
N.A.
$10% of RTP*
$12.4% of RTP*
Turbine Impulse Turbine Impulse u4 Pressure Equivalent Pressure Equivalent c.
Power Range Neutron N.A.
N. A..
N.A.
148% of RTP*
151.3% of RTP*
Flux, P-8 d.
Power Range Neutron N.A.
N.A.
N.A.
150% of RTP*
153.3% of RTP^
Flux, P-9 fa.tz. f
[10%ofRTP*
>6.TX of RTP*
e.
Power Range Neutron N.A.
N.A.
N.A.
Flux, P-10 A
f.
Turbine Impulse Chamber N.A.
N.A.
N.A.
510% of RTP*
512.4% of.RTP* Turbine Pressure, P-13 Turbine Impulse Impulse Pressure Pressure Equivalent Equivalent 19.
Reactor Trip Breakers
'N.A.
N.A.
N.A' N.A.
N.A.
20.
Automatic Trip and Interlock N.A.
N.A.
N.A.
N.A.
N.A.
Logic
^RTP = RATED TilERMAL POWER
TABLE 2.2-1...ontinuzdl a
TABLE NOTATIONS F
ry 2:
f (AI))
(1 + TsS) [T (21 + TsS)'
T'i],,+ K (P - P') -
1 OVERTEMPERATURE AT 1
I +
)
'=
3
'r,,,
NOTE 1:
E.
- *'S) 1
-K2 AT ((}1,+ T 5) (1 + T 5) < AT8:(K1 v
3 2
Measured AT by RTD Manifold Instrumentation; < 1"i I '"F 1 u
y t~
=
AT Lead-lag compensator on measured AT; Where:-
s
- I'b-
=
...for AT, it=8s, 1+T5 2
Time constants utilized in lead-lag compensator
=
r ti,12 T2 = 3 s; c
=, Lag compensator on measured AT; ts a
1 T t3 = 0 s; y
z3 Time constant utilized in the lag compensator for A,
=
I Indicated AT at RATED THERMAL POWERE T3 f.
'?
=
ce AT, *
- r,,
1.10;
=
K 3 0.0137/*F; T
The function generated by.the lead-lag compensator for
=
avg K2 r
1,,
3+*E
=
dynamic compensation; for,T,yg, t. = 28.s, 1 + tsS Time constants utilized in the lead-lag' coinpensator i.
i.
i,
=
is T4, ts = 4 s; s.,
3,,
Average temperature, 'F; P
=
T
!!ag compensator on measured Tavg; 1 -
=
- r p,
.<r,;p
-1 + tsS lag compensator, is = 0 s; Time constant utilized in the measured T,yg i,
=
ts 0
8
q Q
TABLE 2.2-1 (Continued) k TABLE NOTATIONS (Continued) a NOTE 1:
(Continued)
(,
C
- 5 T'
'< 588.5'F (Nominal T at' RATED THERMAL POWER);
4 y
avg H
K3 0.000671;
=
Pressurizer pressure, psig; P
=
P' 2235 psig (Nominal RCS operating pressure);
=
Laplace transform operator, s 1;
~
j S
=
i and f (AI) is'a function of the indicated difference between top and bottom detectors of the t
l power-range neutron ion chambers; wi.th gains to be selected based on measured instrument
! 7 response during plant STARTUP tests such that:
(i)
For q 9 between -35% and + 7%, fg(AI) = 0, where q and q are percent RATED THERMAL
~
t b
t b
j POWER in the top and bottom halves of the core respectively, and qt*Ab is total THERMAL j
POWER in percent of RATED THERMAL POWER; l
(ii)
For each percent that the magnitude of q
~9 exceeds -35%, the AT Trip Setpoint shall t
b l
be automatically reduced by 1.26% of its value at RATED THERMAL POWER; and (iii)
For each percert that the magnitude of qt ~9 exceeds +7%, the AT Trip Setpoint shall b
be automatically reduced by 1.05% of its value at RATED THERMAL POWER.
j i
NOTE 2:
The channel's maximum Trip Setpoint shall'not exceed its computed Trip Setpoint by more than 2.8%
l of AT span.
l l
TABLE 2.2-1 (Continued)
.n, TABLE' NOTA' LIONS (dontinue'd)
.5
...t>
g NOTE.,3:
OVERPOWER.AT
-3 y[
-(1 + r35) 1 AT, (K,
Ks 1 t5 1
y AT 2(AI)}
tsS
,I.- Es.[I, 1 tsS 4
7 4
Fd Where:
AT
= Measbred AT by RTD Manifold Instrumentation; 1+tS Lead-lag compensator on measured AT;
=
1+T S2 Time constants utilized in lead-lag compensator for AT,
=
ti, T2
'tz, = 8 s., T2 = 3 s; 1
Lag compensator on measured AT;
=
'?
1 + tas wo Time constant utilized in'.the lag compensator for AT, r3 = 0 s;
=
T3 AT, Indicated AT at RATED THERMAL POWER;
=
1.085; K
=
4 0.02/*F for increasing average temperature and 0 for decreasing average K
=
3 temperature; b
1}'t3 The function generated by the rate-lag compensator for T,yg dynamic 4
=
7 compensation; Tige constant utilized in the rate-lag compensator for T,yg,17 = 10 s;
=
t7 1
~
y, Lag compensator on measured T,yg,
=
Is Time constant util'ized in the measured T,yg lag compensator, to = 0 s;
=
d f
9 TABLE.2.2-1 (Continued)
I~
E TABLE NOTATIONS (Continued)
E J '
NOTE 3:
(Continued) c-5-
0.00128/*F for T > T" and K = 0 for T $ T";
K.
=.
w f
Aver. age Temperature, 'F; T
=
i T"
Indicated T,yg at RATED THERMAL POWER (Calibration temperature for AT
=
instrumentation, 5 588.5*F);
I Laplace transform operator, s 1; and S
=
f (AI) 0 for all AI.
=
2 c
ty w*
NOTE 4:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by.more than 2.8% of AT span.
4 l
T 8.(
e
- l
=~-.
I.l LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor j{$1$ to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at about.10s counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active. 1 Overtemperature AT-The Overtemperature AT trip provides core protection to prevent DN8 ~ for all combinations of pressure, pcwer, coolant temperature, and axial power distribution, provided that the transient is. slow with respect to piping transit delays frcm the core to the temperature detectors (about 4 seconds), and pressure is within the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced. changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribu-l. tion. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater'than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is. automatically reduced l according-to the notations in Table 2.2-1. Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT trip, l and provides a backup to the High Neutron Flux trip. The Setpoint is auto-j. matically varied with: (1) coolant temperature.to correct'for temperature l-induced. changes in density an'd heat capacity of water, and (2) rate of change L of temperature for dynamic compensation for piping delays from the core to the j' loop temperature detectors, to ensure that the allowable heat generation ' rate (kW/ft) is not exceeded. The Overpower AT trip provides protection l to mitigate ^the consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases." [ CALLAWY - UNU 1 3 2-5 i
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T,y >200 F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% Ak/k for four loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUT 00WN MARGIN less than 1.3% ak/k, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm baron or equivalent until the required SHUTDOWN MARGIN .is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.3% Ak/k: a. Within 1 hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased [ allowance for the withdrawn worth of the immovable or untrippable control rod (s); b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1 at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6; When in MODE 2 with K,ff less than 1, within 4 hours prior to c. achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6; d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specifica-tion 4.1.1.1.le. below, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and Spec 6scation
- See Special Test Exception ^3.10.1.
CALLAWAY - UNIT 1 3/4 1-1 -.. -. -. - ~
M 8 Li L REACTIVITY CONTROL SYSTEMS H00ERATOR TEMPERATURE COEFFICIENT ' LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be: Less positive than 0 Ak/k/*F for the all rods withdrawn, beginning a. of cycle life (80L), hot zero THERMAL POWER condition; <M6 and b. Less negative than -4.1 x 10 4 Ak/k' F for the all rods withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition.
- APPLICABILITY:
Specification 3.1.1.3a. - MODES 1 and 2*#. Specification 3.1.1.3b. - MODES 1, 2, and 3#. ACTION: With the MTC more positive than the limit of Specification 3.1.1.3a. a. above, operation in MODES 1 and 2 may proceed provided: 1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than 0 ak/k/ F within 24 hours or be in HOT STANDBY within the next 6 hours. These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6; 2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and 3. A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive HTC to within its limit for the all rods withdrawn condition. L b. With the MTC more negative than the limit of Specification 3.1.1.3b. above, be in HOT SHUTDOWN within 12 hours. l
- With Keff greater than or egual to 1.
spgica,on l
- See Special Test Exception ^3.10.3.
CALLAWAY - UNIT 1 3/4 1-4 b
REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION The Reactor Coolant System lowest operating loop temper'ature (T"V9) 3.1.1.4 shall be greater than or equal to 551 F. APPLICABILITY: MODES 1 and 2#*. ACTION: With a Reactor Coolant System operating. loop temperature (T,yg) less than 551*F, restore T,yg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS 4.1.1.4 The' Reactor Coolant System temperature (T"V9) shall be determined to be greater than or equal to 551*F: a. Within 15 minutes prior to. achieving reactor criticality, and b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T,yg is less than 561*F with the T,yg-Tref Deviation Alarm not reset. I
- With Keff greater than or eo.ual to 1.
,s cc4c tion e
- See Special Test Exception ^3.10.3.
CALLAWAY - UNIT I 3/4 1-6
REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two 'centrifegal charging pumps shall be OPERABLE. PPLICABILITY: MODES 1, 2, and 3
- ACTI'O'N:
hith'onlyonecentriftsilchargingpumpOPERABLE,restoreatleasttwocentrif-ugal charging pumps te.e)ERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUT 00WN MARGIN equivalent to at least 1% Ak/k at 200 F within the next i hours; restore at least two charging pumps to OPERABLE status within the next. 7 ; days or be in COLD SHUTDOWN within the next 30 hours. m ~:: ~.-~~...:. ::: 2T:r ':= 2::. 22:L... SURVEILLANCE REOUIREMENTS 4.1.2.4 At least'tvo centrifugal charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flow, that the pump develops a differential. pressure of greater than or' equal to 2400 psid when tested pursuant to Specification 4.0.5. g 4 t
- The provisions of Specifications 3.0.4 and74.0.4 are not appficable for entry into MODE 3 for the centrifugal charging pump declared inoperable ' pursuant to Specification 4.1.2.3.2 provided^that centrifugal charging,oump is restored to operable status within 4 hours or prior to the temperature of one or more of the RCS co N lags-exceeding 375 F.
g. CALLAWAY - UNIT 1 3/4 1-10 ~.,
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the following target band (flux difference units) about the target flux difference: a. t 5% for core average accumulated burnup of less than or equal to 3000 MWD /MTU; and b. + 3%, -12% for core average accumulated burnup of greater than 3000 MWD /MTV. The indicated AFD may deviate outside the above required target band at greater than or equal to 50% but less than 90% of RATED THERMAL POWER provided the indicated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cumulated penalty deviation time does not exceed 1 hour during the previous 24 hours. The indicated AFD may deviate outside the above required target band at greater than 15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed 1 hour during the previous 24 hours. APPLICABILITY: MODE 1, above 15% of RATED THERMAL POWER *. ACTION: a. With the indicated AFD outside of the above required target band and with THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes, either: 1. Restore the indicated AFD to within the above required target band limits, or 2. Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER. b. With the indicated AFD outside of the above required target band for more than 1. hour of cumulative penalty deviation time during the previous 24 hours or outside the Acceptable Operation Limits of Figure 3.2-1 and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER, reduce: 1. THERMAL POWER to less than 50% of RATE THERMAL POWER within 30 minutes, and 2. The Power Range Neutron Flu Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
- See Special Test Exception 3.10.2.
- Surveillance testing of the Power Range Neutron Flux channel may be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the Acceptable Operation Limits of Fi.gure 3.2-1.
A total of 16 hours operation may be accumulated with the AFD outside of the above required target band during testing without penalty d riation. CnLLAWAY - UNIT 1 3/4 2-1
POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R shall be maintained within the region of allowable operation shown on Figure 3.2-3 for four loop operation: Where: N F a. R = 1.49 [1.0 + 2 (1.0 - P)] ' b* P _ THERMAL POWER , and RATED THERMAL POWER + Fh=MeasuredvaluesofF obtained by using the mcvable incore N c. detectors to obtain a power distribution map. The measured valuesofFhshallbeusedtocalculateRsinceFigure3.2-3 includes penalties for = detected fccdsater vcaturi fesling of-0.1'.' and T-ee-measurement uncertainties of.'.9% for flow and 4% for incore measurement of F APPLICABILITY: MODE 1. _ ACTION: With the combination of RCS total flow : ate and R outside the region of acceptable operation shown on Figure 3.2-3: a. Vithin 2 hours either: l. Restore the combination of. RCS total flow rate and R to within the above limits, or 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours. CALLAWAf - UNIT 1 3/4'2-8 e w a
1 PENALTIES Oi E."^' FOR Ui40CTECTED TECOWi.-TER I VENTUI.l IOULir4C At40 MEASUREMENT UNCERTAINTIES OF 2.0% FOR FLOW AND 4.0% FOR INCORE MEASUREMENT OF Fay ARE INCLUDED I y IN THIS FIGURE = 48 46 AC'C5P' TABLE UNXdC'$' TABLN P 22; ! OPERATION OPERATION REGION REGION g ...y. - -t ;
- 2:
st -2 1+ -it C l0!..__ _. e g a. 2 :- h lSi Q' G 42 -(1.02,6,41,.68) p m 3 = = p-- - 9 ./ u. .._...t.. 4 40 l w =w=. O I i. b _. - + - - -
- (1.00,39.08) u)
= oe =_:n = < = ue... 38 E" = _. _. - c;..___
- 27. ;* *--
36 ~ .... =. ..... :2 4- = u x;1 34 0.90 0.95 1.00 1.05 1.10 R = fan /1.49 [1.0 + 0.2(1.0-P)] FIGURE 3.2-3 RCS TOTAL FLOW RATE VERSUS R FOUR LOOPS IN OPERATION CALLAWAY - U GT 1 3/4 2-9
l a I}f POWER DISTRIBUTION LIMITS f F" 9} U 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER.*
- ~
ACTION: 'a..$With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09: 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour y until either: 2 -- L ~ T _ a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or ~"~ b) THERMAL POWER is reduced to less than 50% of RATED THERMAL _1. POWER. 2.- Within 2 hours either: ~ ~ a) Reduce the QUADRANT POWER TILT RATIO to within its limit, or ~ ~b) Reduce THERMAL POWER at least 3% from RATED. THERMAL POWER -~ - - - - - -- for each 1% of indicated QUADRANT POWER TILT RATIO in ~ ~ ~ ~ ~ ~ excess of 1 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours. 3. Verify that the QUADRANT POWER TILT RATIO is within its limit ~" ~ within 24 hours after exceeding the limit or reduce THERMAL ______ POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER . ithin the next 4 hours, and w 4. Identify and correct the cause of the out-of-limit condition _.__ prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL power may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER. S sedicalion P
- See Special Test Exception ^3.10.2.
CALLAWAY - UNIT 1 3/4 2-11 C
ABLE 3.3-1 (Continued) t TABLE NOTATIONS / T Only if the Reactor Trip System breakers (happen to be)pn the closed position e Control Rod Drive System is capabTe of roa withdrawal.
- Values left bla~nk pending NRC approval of three loop operation.
- The provisions of Specification 3.0.4 are not applicable.
- Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
- Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
- Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6 hours. ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied: a. The inoperable channel is placed in the tripped condition within 1 hour, b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours for surveillance testing of other channels per Specification 4.3.1.1, and c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours per Specification 4.2.4.2. ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level: ( a. Below the P-6 (Intermediate Range Neutron Flux interlock) Set' point, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint; and b. Above the P-6 (Intermediate Range Neutron Flux interlock) Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL DOWER above 10% of RATED THERMAL POWER. s, 3-5 CALLAWAY, UNIT 1
t 1 . TABLE 4.3-1 (Continued) TABLE NOTATIONS b c osed and the Con-hthe Reactor Trip System breaker -nod Drive System is capable of rod wiw.u.awa.. trui
- Below P-6 (Intermediate Range Neutron Flux interlock) Setpoint.
- Below P-10 (Low Setpoint Power Range Neutron Flux interlock) Setpoint.
- Below P-6 (Intermediate Range Neutron Flux interlock) Setpoint.
(1)' If not performed in previous 7 days. (2): Comparison of calorimetric to excore power indication above 15% of RATED' THERMAL POWER. Adjust excore channel gains consistent with calorimetric . power if absolute difference is greater than 2%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1. g-(3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are hot applicable for entry into MODE 2 or 1. (4) Neutron detectors may be excluded from CHANNEL CALIBRATION. .(5) Detector plateau curves shall be obtained, evaluated and compared to manufacturer's data. For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification'4.0.4 are not applicable for entry into MODE 2 or 1. ~ Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The (6) provisions of Specification 4.0.4 are not applicable for entry into ~" ~ MODE 2 or 1. (7) Each train shall be tested at least every 62 days on a STAGGERED TEST ^ BASIS. With power greater than or equal to the interlock Setpoint the required (8) ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the inter-lock is in the required state by observing the permissive annunciator window. Monthly surveillance in MODES 3*, 4*, and 5* shall also include-verification (9) that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window. l-Monthly surveillance shall include verification of the Boron Dilution Alarm Setpoint of less than or equal to an increase of twice the count rate within a 10-minute period. (10) Setpoint verification is.not applicable. (11) At least once per 18 months and following maintenance'or adjustment of the Reactor trip breakers, the TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verification of the Undervoltage and Shunt trips. (12) At least once per 18 months during shutdown, verify that on a simulated Boron Dilution Doubling test signal the normal CVCS discharge valves will close and the centrifugal charging pumps suction valves from the RWST l will open within 30 seconds. I (13) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate. CALLAWAY - UNIT 1 3/4 3-12
INSTRUMENTATION 3/4.3.2 ENGINEEREDSAFETYiEATURESACTUATIONSYSTEMINSTRUMENTATION LIMITING CONDITION FOR OPERATION / 3.3.2 The Engineered Safety Features Actuation System (ESFA3) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE-with their Trip b Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5. APPLICABILITY: As shown in Table 3.3-3. l- ' ACTION:
- a.
With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4 adjust the Setpoint consistent with the Trip Setpoint value. [b... ith an ESFAS Instrumentation or Interlock Trip Setpoint less W conservative than the value shown in the Allowable Values column of ' ' Table 3.3-4, either: 1.' Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4 and determine within 12 hours that Equation 2.2-1 was satisfied for the affected channel, or ~
- 2..
Declare the channel inoperable and apply the applicable ACTION '~r statement requirements of Table 3.3.3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent 17 with the Trip Setpoint value. _ Equation 2.2-1 Z + R + S 5,TA Where: Z =-The value from Column Z of Table 3.3-4 for the affected
- channel, R = The "as measured" value (in percent span) of rack error for the affected channel,
~~ = Either the "as measured" value (in percent span') of the S ~' . sensor error, or the value from Column S (Sensor Error) of Table 3.3-4 for the affected channel, and TA.= The value from Column TA (Total Allowance) of Table 3.3-4 for the affected channel. D c._ With an ESFAS instrumentation channel or interlock inoperable, take the j -ACTION shown in Table 3.3-3. !~ j SURVEILLANCE REQUIREMENTS 4.3.2l1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by the performance of:the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2. 4.3'.2.2 'The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested atleast once per N times 18 months w' tre N is the total number of redundant chan-L nels.in a s'pecific ESFAS function as shown in the " Total No. cf Channels" Column ~ of Table 3.3-3. CALLAWAY -/ UNIT 1 3/4 3-13 L_
n TABLc 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION n ? E MINIMUM g CHANNELS CHANNELS APPLICABLE TOTAL NO. FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 'E 1. Safety Injection (Reac' tor Trip, Phase "A" Isolation () g ~ Feedwater Isolation, Compo-nent Cooling Water,-Turbine Trip, Auxiliary Feedwater-Motor-Driven Pump, Emergency Diesel Generator Operation, Contain-ment Cooling, and Essential Service Water Operation) a. Manual Initiation 2 1 2 1, 2,.3, 4 18 b. Automatic Actuation 2 1 2 1, 2, 3,'4 14 m} Logic and Actuation J, Relays (SSPS) 'c. Containment 3 2 2 1, 2, 3_ 15* Pressure-High-1 d. Pressurizer 4 2 3 1, 2, 3# 19* Pressure - Low e. Steam Line Pressure-3/ steam lin.e 2/ steam'1ine 2/ steam line 1, 2, 3## 15* Low any steam' line
- 4
~, 2. Containment, Spray 2 pair 1 pair 2 pair 1,2,3,4 18 a. Manual Initiation operated simultaneously b. Automatic Actuation 2 1 -2 1,2,3,4 14 Logic and Actitation Relays (SSPS) t c. Containment Pressure-4 2 l 3 1, 2, 3 16 High-3
TABLE 3.3-3 (C:ntinued) ENGINEERED SAFETY FEATURES ACTUATION' SYSTEM INSTRUMENTATION n> MINIMUM h TOTAL NO. CHANNELS CHANNELS ' APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION Feedwater Isolatip&Trbine TriD 5. 4 a. Automatic Actuation 2 1 2 1, 2 21 g Logic and Actuation Relay (SSPS) b. Steam Generator 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2 19* Water Level-in any oper-in each oper-l High-High ating stm gen. ating stm. gen. c. Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements. 6. Auxiliary Feedwater a. Manual Initiation 3(1/ pump) 1/ pump' 1/ pump 1,2,3 24 b. Automatic Actuation Logic 2 1 2 1,2,3 21 and Actuation Relays (SSPS) t i c. Automatic Actuation Logic and Actuation Relays 2 1,2,3 21 .(80P ESFAS) 2 1 d. Steam Generator Water \\ Level-Low-Low i
- 1) Start Motor-i l
Driven Pumps 4/stm. gen. 2/stm. gen. 3/stm. gen. 1,2,3 19* in any opera-in each ting stm. operating a gen. stm. gen. I e
TABLE 3.3-3 d ntinu"d) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION O P E MINIMUM Q TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES , ACTION C5 8. Loss of Power .-4 a. 4 kV Bus Undervoltage 4/ Bus 2/ Bus 3/ Bus 1, 2, 3, 4 19* y -Loss of Voltage b. 4 kV Bus Undervoltage 4/ Bus 2/ Bus 3/ Bus 1,2,3,4 19" -Grid Degraded Voltage 9. Control Room Isolation a. Manual Initiation 2 1 2 All 18 b. Automatic Actuation 2 1 2 All 14 O R Logic and Actuation M Relays (SSPS) Y T g c. Automatic Actuation Logic H and Actuation Relays (B0P ESFAS) 2 1 2 All 14 d. Phase "A" Isolation See Item 3.a. above for all Phase "A" Isolation initiating functions and . requirements. 10. Engineered Safety Features Actuation System Interlocks d. Pressurizer Pressure, 3 2 2 1,2,3 20 P-11 b. Reactor Trip, P-4 4-2/ Train 2/ Train 2/ Train 1, 2, 3 22 ll. $Ni S Souk dMt/ 2-l/%i A 1/%iA Q.-Q%w l,2,Q 4 2$ , w q- .n--- -., nn.m,w. w, amg p e.,.
H TABLE 3.3-3 (Continued) ACTION STATEMENTS (Continued) ACTION 18 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. ACTION 19 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied: a. The inoperable channel is placed in the tripped condition f s within 1 hour, and b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours for surveillance testing of other channels per Specification ~, 4.3.2.1. ACTION 20 - With less than the Minimum Number of Channels OPERABLE, within 1 hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.O.3. ) ACTION 21 - Wi.th the number of OPERABLE Channels one less than the Minimum ~ Channels OPERABLE requirement, be in at least HOT STANDBY i within 6 hours and in at least HOT SHUTDOWN within the following 6 hours; however, one channel may be bypassed for up to 2 hours i for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE. ACTION 22 - With the number of OPERABLE channels one less than the Total ~ Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within ~ 6 hours and in at least HOT SHUTDOWN within the following 6 hours. ACTION 23 - With the number of OPERABLE channels one less than the Total T Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or declare the associated valve I inoperable and take the ACTION required by Specification 3.7.1.5. ACTION 24 - With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirement, declare the affected auxiliary feedwater pump inoperable and take the ACTION required by Specification 3.7.1.2. &CTINW 9GS4'c Y,ts w l 0 A DW(q fy l I i l CALLAWAY - UNIT 1 3A 3-21 l k
- Action 25 - With the number of OPERABLE channels one less than the mininum channels OPERABLE requirement, declare the affected diesel generator and'off-site power source inoperable and take the ACTION required by Specification 3.8.1.1. 4 6 9 4 e b 9 - 3/4 ' 5 - 21 A
~. ~ {~N ,e - ,m l CAe oxy h. TABLE 3.3-4 5: g ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR TRIP ALLOWABLE E FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (S) SETPOINT VALUE ~ P 1. Safety Injection r (Reactor Trip, Phase "A" Isolation, Feedwater p i Isolation, Turbine Trip, Component Cooling Water, i, Auxiliary Feedwater - 1 ' ' Motor 9 riven Pump, Emergency Diesel Generator Operation, Containment 4 Cooling, and Essential i Service Water Operation) 4 a. Manual Initiation N.A. N.A. N.A. N.A. N.A. 'if N b. Automatic Actuation Logic and Actuation O i Relays (SSPS) N.A. N.A. N.A. N.A. N.A. 2 !} c. Containment Pressure - 3.6 2.0 4.6 ~1"1 HIgh-1 3,4-0.71 -h4 $ 3.5 psig 1-M psig ---( i ii d. Pressurizer Pressure -
- 18. (,
14.41 2.0 1834 Low 43re-4GrM
- 4. 5-
> 1849 psig > 4839-psig i i ! e. Steam Line Pressure - 19.(o 14.6f 2.0 Gl6 57/ i. Low 44-e 40-M-M > -585-psig > 564-psig* 2. Containment Spray l a. Manual Initiation N.A. N.A. N.A. N.A. N.A. I b. Automatic Actuation Logic and Actuation Relays (SSPS) N.A. N.A. N.A. N.A. N.A. i
TABLE 3.3-4'(Continued) 9 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOI P E4 SENSOR TRIP ALLOWABLE TOTAL ,i ALLOWANCE (TA) Z_ ERROR (S) SETPOINT VALUE FUNCTIONAL UNIT c b 2. Containment Spray (Continued) w -x c. Containment Pressure- .' 3 0.71 2.0 5 27.0 psig i 28.3 sig Nigh-3 3. Containment Isolation s a. Phase "A" Isolation
- 1) Manual Initiation N.A.
N.A. N.A. N.A. N.A. R
- 2) Automatic Actuation Logic and Actuation T
Relays (SSPS) H.A. N.A. N.A. N.A. N.A. U See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.
- 3) Safety Injection b.
Phase "B" Isolation
- 1) Manual Initiation N.A.
N.A. N.A. N.A. N.A.
- 2) Automatic Actuation Logic and Actuation Relays (SSPS)
N.A. N.A. N.A. N.A. 'N.A. .4 .y Pressure-High-3 4.3 .0.71 2.0 1 27.0 psig i 28. psig i
- 3) Containment J
c. Containment Purge Isolation j j
- 1) Manual Initiation N.A.
N.A. N.A. N.A. H.A. I
- 2) Automatic Actuation Logic and Actuation Relays (SSPS)
N.A. N.A. N.A. N.A. N.A. ~m :. : 2;-
n a> ~ 9 . TABLE 3.3-4 (Continued) " 4 n;,i m i r- ' Lt T ' 'r~- E ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 4 TOTAL SENSOR TRIP ALLOWABLE 5 FUNCTIONAL UNIT ALLOWANCE (TA) ERROR (S) SETPOINT VALUE Z y 3. Containment Isolation (Continued)
- 3) Automatic Actuation Logic and Actuation Relays (BOP ESFAS)
N.A. N.A. N.A. N.A. N.A. 4) Phase "A" See Item 3.a. above for all Phase "A" Isolation Trip Setpoints and Allowable Isolation
- Values, 4.
Steam Line Isolation w1 a. Manual Initiation N.A. N.A. N.A. N.A. N.A. w4 b. Automatic Actuation Logic and Actuation Relays (SSPS) N.A. N.A. N.A. N.A. N.A. c. Containment Pressure-High-2 4.3 0.71 2.0 1 17.0 psig 5 18.3 psig d. Steam Line Pressure- ,_ 571 psig* Low 19.6 14.81 2.0 > 615 psig .4 \\ e. Steam Line Pressure Negative Rate - High 3.0 0.5 0 1 -100 $ -124 psi /s** " psi /s i 5. Feedwater Isolation & Turbine Trip a. Automatic Actuation Logic and Actuation ,N.A. N.A N.A. N.A. N.A. Relays (SSPS)
r~-- - -~- TABLE 3.3-4 (Cont'inued) n. 2 F ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP'SETPOINTS ER TOTAL SENSOR TRIP ALLOWABLE e FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (S) SETPOINT .VALUE e Z 5. Feedwater Isolation (Continued) g - b. Steam Generator Water - Level-High-High 5' 2.18 2.0 $ 78% of i 79.8% of narrow range narrow range instrument instrument span span ] s c. Safety, Injection See Item 1. above for all Safety Inject W detpoints-and-Allowable Values. 6. Auxiliary Feedwater a. Manual Initiation N.A. N.A. N.A. N.A. N.A. M b. Automatic Actuation ~ Logic and Actuation Relays (SSPS) N.A. N.A. N.A. N.A. N.A. , c. Automatic Actuation Logic and Actuation Relays (BOP ESFAS) N.A. N.A. N.A. N.A. N.A. d. Steam Generator Water
- .4 Level-Low-Low T
- 1) Start Motor-Driven Pumps 2j3.5 21.18 2.0 1 23.5% of 1 22.0% of narrow range narrow range instrument instrument span span
- 2) Start Turbine-Driven Pump 23.5
.21.18 2.0 1 23.5% of 1 22.0% of narrow range narrow range instrument instrument span span
r /
- n..
~ ~ ' STABLE 3.3'-4' (Continued) n. 2 F
- ' ENGINEERED' SAFETY FEATURES ACTUATION SYST$M INSTROMENTATION TRIP SETPOINTS E
22 ' TRIP 'ALLOWADLE TOTAL SENSOR 'e -I f " FUNCTIONAL' UNIT ' ALLOWANCE (TA) Z_ - ERROR (S) SETPOINT VALUE z U 6 Auxiliary Feedwater (Continued) w e. Safety Injection-Start Motor-Driven Pumps See' Item 1. above for all Safety Injection Trip'Setpoints and Allowable Values. f. Loss-of-Offsite Power-4 1 Start Turbine-Driven 3 Pump N.A. N. A.
- N.~ A. ' i N.A.
N.A. g.' Trip of All Main j R Feedwater Pumps-1 Start Motor-Driven l T Pumps cN.A. N.A. N.A. N.A. N.A. y ~,, 1 h. Auxiliary Feedwater l Pump Suction Pressure-Low (Transfer to ESW) N.A. N.A. N.A. 123.4 121.72 i 7. Automatic Switchover to Containment Sump K a. Automatic Actuation 1 l Logic and Actuation .4 Relays (SSPS) N.A. N.A. N.A. N.A. N.A. s i b. RWST Level-Low-Low 3.4 . 1~. 21 2.0 ~$36% ->35.2U ,.,D Coincident with Safety Injection ee Item.1.nabove for Safety Injection-Trip
- tpoints-and-All a e Values.
] 8. Loss of Power l a. 4 kV Undervoltage i -Loss of Voltage N.A.
- N.A.
N.A. 83V (120V Bus) 83+0,-8.3V (120V Bus) w/ls delay w/1+0.2,-0.5s delay e r 7 s
fx.) Q' 0Muy AL N TABLE 3.3-4 (Crntinued) n. 2{ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR TRIP ALLOWABLE g FUNCTIONAL UNIT ALLOWANCE (TA) Z, ERROR (S) SETPOINT VALUE Z 8. Loss of Power (Continued) b. 4 kV Undervoltage -Grid Degraded Voltage N.A. H.A. N.A. 104.5V 104.5+2.6,-0V (120V Bus) (120V Bus) w/119s delay w/119 11.6s delay 9. Control Room Isolation 'a. Manual Initiation N.A. N.A. N.A. N.A. N.A. R b. Automatic Actuation N.A. N.A. H.A. H.A. N.A. Logic and Actuation ,Y Relays (SSPS) A ~ c. Automatic Actuation Logic and Actuation Relays (B0P ESFAS) N.A. N.A. N.A. N.A. N.A. d. Phase "A" Isolation See Item 3.a. above for all Phase "A" Isolation Trip Setpoints and Allowable Values.
- 10. Engineered Safety Features Actuation System Interlocks a.
Pressurizer Pressure, P-11 N.A. N.A. N.A. 5 1970 psig i 1981 psig b. Reactor Trip, P-4 N.A. H.A. N.A. N.A. H.A. l], Sd(d S}be. Load NA NA-Nb N Wk s4a-ca l y _w-- ,,3,-,,,, m..,. 4 ma p.,,,y n p sy~_ _ nvr,e p --
TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 3. Pressurizer Pressure-Low } a;.: Safety Injection (ECCS) $ 29(1)/12(4)
- 1) Reactor Trip
$2 I
- 2) Feedwater Isolation
<7 h2(5) i - 3) Phase "A" Isolation
- 4) Auxiliary Feedwater 5 60
[
- 5) Essential Service Water 5 60(1)
- 6) Containment Cooling
$ 60(1) I ~ I
- 7) Component' Cooling Water N.A.
5 14(6)
- 8) Start Diesel Generators
- 9) Turbine Trip N.A.
4. Steam Line Pressure-Low a. Safety Injection (ECCS) $ 24(3)/12(4) , 1) Reactor Trip (from SI) $2 2) Feedwater Isolation <7 h2(5) 3) Phase "A" Isolation 4) Auxiliary Feedwater $ 60 5) Essential Service Water < 60(1) e i 6) Containment Cooling 60(1) 7) Component Cooling Water N.A. 5 14(6) 8) Start Diesel Generators 9) Turbine Trip N.A. ~ b. Steam'Line Isolation 57 L i I'_. . -~. CALLAWAY - UNIT 1 3/4 3-30
l J i I TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES -i INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS i' 5. Containment Pressure--High-3 a. Containment Spray 5 32(1)/20(2) lb. Phase "B" Isolation 5 31.5 6. Containment Pressure--High-2 I Steam Line Isolation S7 i 7. Steam Line Pressure-Negative Rate-High Steam Line Isolation N.A. 8. Steam Generator Water Level--High-High '( a. Feedwater Isolation S7 b. Turbine Trip $ 2.5 I 9. Steam Generator Water Level - Low-Low a. Start Motor-Driven Auxiliary 1 Feedwater Pumps 5 60 i b. Start Turbine-Driven Auxiliary Feedwater Pump $ 60 10. Loss-of-Offsite Power Start Turbine-Driven Auxiliary Feedwater Pump N.A. 11. Trip of All Main Feedwater Pumps Start Motor-Driven Auxiliary Fee'dwater Pumps N.A. s l } CALLAWAY - UNIT 1 3/4 3-31 =,.-.
TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 12. Auxiliary Feedwater Pump Suction Pressure-Low Transfer to Essential Service Water N.A. 13. RWST Level-Low-Low Coincident with Safety Injection Automatic Switchover to Containment < 60 ~ Sump 14. Loss'of Power a. 4 kV Bus Undervoltage- < 14 ~ Loss of Voltage / b. 4 kV Bus Undervoltage- < 144 ~ Grid Degraded Voltage 15. Phase "A" Isolation a. Control Room Isolation N.A. b. Containment Purge Isolation < 2(5) TABLE NOTATIONS 1 (1) Diesel generator star, ting and sequence loading delays included. (2) Diesel generator starting delay not included. Offsite power available. (3) Diesel generator starting and sequence loading delay included. RHR pumps not included. (4) Diesel generator starting and sequence loading delays not included. Offsite power available. RHR pumps not included. (5) Does'not include valve closure time. (6) Includes time for diesel to reach full speed. 4 g CALLAWAY - UNIT 1 3/4 3-32 ~. - - ~. -..,--,,~,,,,_.,,_-.,,--,-,-.-,--.,.,,_,,-_~n ,y ,.c r,-- - - - .v,,--
TABLE 4.3-2 hh ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION !h SURVEILLANCE REQUIREMENTS TRIP c: 2E ANALOG ACTUATING MODES Il CHANNEL DEVICE MASTER SLAVE FOR WHICN ~ FUNCTIONAL UNIT CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANC bd CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRES 1. Safety Injection (Reactor Trip, Phase "A" Isolation, Feedwater Isolation, Turbine Trip, Component Cooling Water, Auxiliary Feedwater-5 Motor-Driven Pump. Emergency Diesel Generator Operation, Containment Cooling, and Essential Service R Water Operation) ^ a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4 df 1,2,3,4 b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Logic and Actuation Relays (SSPS) c. Containment Pressure-S R M N.A. N.A. N.A. N.A. 1,2,3 High-1 d. Pressurizer Pressure-S R M N.A. N.A. N.A. N.A. " 1,2,3 ~ Low e. Steam Line Pressure-S R M N.A. N.A. N.A. N.A. 1,2,3
- 4 Low 2.
Containment Spray N.A. N.A. N. 1,2,3,4 a. Manual Initiation N.A. N.A. N.A. R b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 3) 1, 2, 3, 4 Logic and Actuation ' Relays (SSPS) c. Containment Pressure-S R M N.A. N.A. N.A. N.A. 1,2,3 High-3 s ?
g TABLE 4.3-2 (Continued) ~ S E ENGINEERED SAFETY' FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS e-TRIP ANALOG ACTUATING MODES c5 CHANNEL DEVICE MASTER SLAVE . FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANO FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST . TEST TEST IS REQUIRED 3.' Containment Isolation a. Phase "A" Isolation.
- 1) Manual Initiation N.A.
N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4
- 2) Automatic Actuation N.A N.A.
N.A. N.A. M(1) M(1) (3 1; 2, 3, 4 Logic and Actuation Relays (SSPS)
- 3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
T b. Phase "B" Isolation
- 1) Manual Initiation N.A.
N.A. N.A. R N.A. N.A. N.A. I', 2, 3, 4
- 2) Automatic Actuation N.A.
N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays (SSPS)
- 3) Containment 5
R M N.A. N.A. N.A. N.A. 1, 2, 3 Pressure-High-3 c. Containment Purge Isolation
- 1) Manual Initiation N.A.
N.A. N.A. R N.A. N.A. N. A;* 1, 2, 3, 4
- 2) Automatic Actuation N.A.
N.A. N.A. N.A. M(1) M(1) (3) 1, 2, 3, 4 Logic and Actuation Relays (SSPS)
- 3) Autcmatic Actuation Logic and Actuation'
.N.A. N.A. 1,2,3,4 'lelays (BOP ESFAS) N.A. N.A. N.A. N.A. M(2)(i)
- 4) Phase "A" See Item 3.a. above,for all Phase "A" Isolation Surveillance Requirements.
Isolation
~, 4 TABLE 4.3-2 (Continued) c3 p: 5: ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION t SURVEILLANCE REQUIREMENTS 32 TRIP c: 25 ANALOG ACTUATING MODES -d CHANNEL DEVICE MASTER SLAVE
- FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLAN(
~ FUNCTIONAL UNIT CHECK - CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIREI 4. Steam Line Isolation a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,.2, 3 b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2,i3 Logic and Actuation j Relays (SSPS) i c. Containment Pressure-S R M N.A. N.A. N.A. N.A. 1, 2, 3 u, l 3s High-2 l d. Steam Line Pressure-S R M N.A. N.A. N.A. N.A. 1,2,3 b' Low t i e. Steam Line Pressure-S R M N.A. N.A. N.A. N.A. 3, 4 l Negative Rate-High 5. Feedwater Isolation & Turbine Trip a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q(3) 1, 2 i Logic and Actuation Relay b. Steam Generator Water S R M N.A. N.A. N.A. N.A.4 1, 2 i Level-High-High c. Safety Injection See Iteq 1. above for all Safety Injection Surveillance Requirements. j 6. Auxiliary Feedwater j
- a.. Manual Initiation N.A.
N.A. N.A. R. N.A. N.A. N.A. 1,2,3 b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q l', 2,"3 Logic and Actuation ) Relays (SSPS) i
f ~ TABLE 4.3-2 (Continuedl ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION. r-h SURVEILLANCE REQUIREMENTS TRIP s-ANALOG ACTUATING MODES c O CIIANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANC FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED 6.- Auxiliary Feedwater (Continued) c. Automatic Actuation Logic and Actuation Relays (BOP ESFAS) N.A. N.A. N.A. N.A. M(2)G) N.A. N.A. 1, 2,.3 d. Steam Generator Water S R M N.A. N.A. N.A. N.A. 1, 2, 3 Level-Low-Low w e. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements. t. Loss-of-Offsite Power N.A. R N.A. M N.A. N.A. N.A. 1,2,3 w N.A. 1, 2 0 g. Trip of All Main N.A. N.A. N.A R N.A. N.A. Feedwater Pumps , h. Auxiliary Feedwater S R M N.A. N.A. N.A. N.A. 1, 2, 3 Pump Suction Pressure-Low 7. Automatic Switchover to Containment Sump a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) 1, 2, 3, 4 Logic and Actuation .4 i Relays (SSPS) b. RWST Level - Low-Low S R H N.A. N.A. N.A. N.A. 1,2,3,4 Coincident With i Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements. 8. Loss of Power a. 4 kV Undervoltage-N.A. R N.A. M N.A. N.A. N.A. 1~, 2, 3, 4 Loss of Voltage
TABLE 4.3-2 (Continued) F ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ~ ( f TRIP ANALOG ACTUATING MODES Z CHANNEL DEVICE MASTER SLAVE FOR WHICH w CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY ~ RELAY SURVEILLANT FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIREI 8. Loss of Power (Continued) b. 4 kV Undervoltage-N.A. R N.A. M N.A. N.A. N.A. 1,2,3,4 Grid Degraded Voltage 9. Control Room Isolation a. Manual Initiation N. ~A. N.A. N.A. R N.A. N.A. N.A. All b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q(3) All logic and Actuation u d, Relays (SSPS) w c. Automatic Actuation Logic and Actuation Relays (BOP ESFAS) N.A. N.A. N.A. N.A. M(2)(i) N.A. N.A. All d. Phase "A" Isolation See Item 3.a. above for all Phase "A" Isolation Surveillance Requirements. 10. Engineered Safety Features Actuation System Interlocks a. Pressurizer Pressure, N.A. R M N.A. N.A. N.A. N.A. 1,2,3 P-11 b. Reactor Trip, P-4 N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3 N.A. M A. - TABL bTATIONS A. %M M. A. EA. 1 2,y 11. Sclict State. lced Sequew.c,r 3 (1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. 2 L.Cpntinuity check may be ex_q1uded_from the ACTUATION LOGIC TEST. ~ ~~ .(3), Eyceit~Reliys K602, 'K620, K622, K624, K630, K740, aiid K741, wiiiEII~iha11 be telited at lea 5Conce per 18 months during refueling and during each COLD SHUTDOWN exceeding 24 hours unless they have been tested within the previous 90 days. ~.. _ _. _.
TABLE 3.3-6 c3> h RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS ~ MINIMUM CHANNELS CHANNELS APPLICABLE-ALARM / TRIP Ei FUNCTIONAL UNIT TO TRIP / ALARM OPERABLE MODES SETPOINT ACTION ~d 1. Containment r, a. Containment Atmosphere Gaseous Radioactivity- '1 2 All 26 High (GT-RE-31 & 32) b. Gaseous Radioactivity-RCS Leakage Detection N.A. 1 1,2,3,4 N.A. 29 (GT-RE-31 & 32) c. Particulate N.A. I 1,2,3,4 N.A. 29 Radioactivity-u, 3 RCS Leakage Detection (GT-RE-31 & 32) u, 2. Fuel Building a. Fuel Building Exhaust-30 Gaseous Radioactivity-1 2 High (GG-RE-27 & 28) b.* Criticality-High < 15 mR/h 28 Radiation Level 1 2 (50-RE-37 & 38) d.4 3. Control Room Air Intake-Gaseous Radioactivity-High 1 2 All 27 (GK-RE-04 & 05) 9
s TABLE 3.3-6 (Continued) TABLE NOTATIONS
- With fuel in the fuel storage areas or fuel building.
- With irradiated fuel I,n the fuel storage areas or fuel building.
3
- Trip Setpoint concentration value (pCi/cm ) is to be established such that the_ actual submersion dose rate w uld not exceed 2 mR/hMacntE17oom.
- Trip Setpoint concentrat~ ion value GJCi7cm ) is to~ be established such that 8
.the actual submersion dose rate would not exceed 4 mR/h in the fuelsbuilding. 3
- Trip Setpoint concentration value (pCi/cm ) is to be established such that the actual submersion dose rate would not exceed 9 mR/hsin the containment building.
The Setpoint value may be increased up to the equivalentglimits of Specification 3.11.2.1 in accordance with the nethodology and parameters { in the OD,CM during containment purge or vent provided the Setpoint value.does not exceed the max,fmum concentration activity in the containment' determined by the sample analysis performed prior to each release in accordance with Tab.e 4.11-2. ) s ACTION STATEMENTS ACTION 26 - With less than the Minirum Channels OPERABLE requirement, operation may continue provided the containment purge valves are maintained closed. 4 s ACTION 27 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, within 1 hour isolates.the Control Room Errargency Ventilation System and initiate operation of the Control Room Emergency Ventilation System in the recirculation mode. ACTION 28 - With less than the Minimum Channels OPERABLE requirement, operation may contique for up to 30 days provided an appropriate portable continuous monitor with the same Alarm Setpoint'is provided in the' fuel area. sRestore the inoperable monitors to CPERABLE status within 30 days or suspend p11' operations involving fuel movement l in the. fuel building. ACTION 29 - Must satisfy the ACTION requirement for Specification 3.4.6.1. \\ i ACTION 30 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, within 1 hour isolate the Fuel Building Ventilation System and initiate operation of the Emergency Exhaust System to majntain the fuel bailding at a negative pressure. s, s s s 8 8 W e a O 'g t CALLAWAY - UNIT 1 3/4 3-40 b
~ TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS' ( ,~ ANALOG ~ 4 CHANNEL CHANNEL CHANNEL OPERATIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST 1. Triaxial Peak Recording Accelerographs a. Radwaste Base Slab N.A. R N. A. b. Control Room N.A. R N.A. c. ESW Pump Facility N.A. R N.A. ~d. Ctmt Structure N.A. R N. A.
- e.
Auxiliary Bldg. SI Pump Suction N.A. R N.A. f
- f.
SGB Piping N.A. R N.A. .g. SGB Support N.A. R N. A. 2.
- Triaxial Time History and Response Spectrum Recording System, Monitoring the Following Accelerometers (Active) a.
Ctmt. Base Slab M R SA b. _Ctmt. Oper. Floor M R c. Reactor Support M R SA** d. Aux. Bldg. Base Slab M R SA** e. Aux. Bldg. Control Room Air Filters M R SA** 'f. Fre~e Field M R SA** 3. Triaxial Response-Spectrum Recorder (Passive) Ctmt. Base Slab N.A. R N.A.* 4. Triaxial Seismic Switches a. OBE Ctmt. Base Slab M R SA b. SSE Ctmt. Base Slab M R SA c. OBE Ctmt. Oper. F1. M R SA .d. SSE Ctmt. Oper. F1. M R SA e. System Trigger M .R SA
- Checking at the Main Control Board Annunciators for contact closure output in the Control Room shall be performed at least once per 184 days.
- The-B1 stable Trip Setpoint need not be determined during the performance of an ANALOG CHANNEL OPERATIONAL TEST.
0 D CALLAWAY, UNIT 1 3/4 3-45
INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION ^l LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be 0?ERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION: x, a. With the number of OPERABLE accident monitoring instrumentation channels less than the-Total Number of Channels shown in ~ ' ' Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours. b. With the number of OPERABLE accident monitoring instrumentation-channels except the reactor coolant radiation level monitor and the unit vent - high range noble gas monitor less than the Minimum Channels"0PERABLE requirements of Tabic 3.3-10, restore ~ the inoperable channel (s) to OPERABLE status.within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours. .edf7 With the number of OPERABLE channels for the reactor coolant [ radiation level monitor or the unit vent - high ringe. noble gas monitor less than the Minimum Channels OPERABLE requirements of~ Table 3.3-10, restore the inoperable channel.to-OPERABLE status within 7 days, or be in at least HOT SKUTDOWN within the next 12 hours < d. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE. REQUIREMENTS i l-4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencie's.shown in Table 4.3-7. t 4 ( CALLAWAY - UNIT 1 3/4 3-52 LL
TABLE 3.3-11 (Continued) FIRE 0ETECTION INSTRUMENTS TOTAL NUMBER OF INSTRUMENTS
- INSTRUMENT LOCATION ZONE HEAT FLAME SMOKE (x/y)
(x/y) (x/y) 6202-Elec. Equipment Rm. 601 3/0 6203-Air Handling Equip. Rm. 601 3/0 6301-Fuel Bldg. 2047'6" Gen. Fir. 602 2/0 6303-Fuel Bldg. Exh. Filt. Absorb. Rm. A 601 2/0 6304-Fuel Bldg. Exh. Filt. Absorb. Rm. B 601 2/0 -North ESW Pumphouse 002 3/0 -South ESW Pumphouse 001 3/0 -ESW Cooling Tower 001 1/0 -ESW Cooling Tower 002 1/0 -ESF Transkmer XN6cl Olb e/g -Ese Transfermer xN00E on c/e TABLE NOTATIONS
- (x/y): x is number of Function A (early warning fire detection and notification only) instruments.
y is number of Function 8 (actuation of fire suppression systems and early warning and notification) instruments.
- The fire detection instruments located within the containment are not required to be OPERABLE during the performance of Type A containment leakage rate tests.
(1) Zone is associated with a Halon protected space. Each space has two separate detection circuits (zones). One zone, in its entirety, needs to remain (operable] uc, (2) Line-type heat detector. CALLAWAY - UNIT 1 3/4 3-61
TABLE 3.3-12 9 F: RADIDACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION E' 3E MINIMUM 1 CHANNELS e I ~ e INSTRUMENT OPERABLE ACTION 5 -d 1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release a. Liquid Radwaste Discharge Monitor (HB-RE-18) 1 31 b. Steam Generator Blowdown Discharge Monitor (BM-RE-52) 1 32 c. Turbine Building Drain Monitor (LE-RE-S9) 1 32 d. ' Secondary Liquid Waste System Monitor (HF-RE-45) 1 33 2. Flow Rate Measurement Devices u, D u, a. Liquid Radwaste Discharge Line E*
- 1) Waste Monitor Tank A Di,scharge Line 1
34
- 2) Waste Monitor Tank B Discharge Line 1
34 b. Steam Generator Blowdown Discharge Line j 1 34 c. Secondary Liquid Waste System Discharge Line 1 34 Combin,ed Cooling Tower Blowdown Line and Bypass Flow N, 1 34 ,, g o.. t ,s - y e m
TABLE '.3-8 4 n 2 F RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E 2E ANALOG ~ 3 CHANNEL g ( *. c CHANNEL SOURCE CHANNEL OPERATIONAL { IA?TRUMENT CHECK CHECK CALIBRATION TEST 1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release a. Liquid Radwaste Discharge Monitor (HB-RE-18) D P R(2) Q(1) b. Steam Generator Blowdown Discharge Monitor D M R(2) Q(1) (BM-RE-52) c. Turbine Building Drain Monitor (LE-RE-59) D M R(2) Q(1) ,s s d. Secondary Liquid Waste System D P R(2) Q(1) Y Monitor (HF-RE-45) 2. Flow Rate Measurement Devices a. Liquid Radwaste Discharge Line D(3) N.A. R N.A. b. Steam Generator Blowdown Discharge Line D(3) N.A. R N.A. c. Secondary Liquid Waste System Discharge D(3) N.A. R N.A. Line d. Combined Cooling Tower Blowdown Line and D(3) N.A. R N.A. Bypass Flow t i ,. R * =
- 1 f
y
5. TABLE 4.3-b(Continued) TABLE NOTATIONS g (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate tnat auto *atic m ~ isolation of this pathway and control room alarm annunciation occur as appropriate if any of the following conditions exists: ~ a. Instrument indicates measured levels above the Alarm / Trip Setpoint ~ (isolation and alarm), or b. Circuit failure (alarm only), or 1 - c. Instrument indicates a downscale failure (alarm only), or' ' d. Instrument controls not set in operate mode (alarm only). [ (2) $heinitialCHANNELCALIBRATIONshallbeperformedusingoneormoreof the reference (gas or liquid and solid) standards certified by the ' National Bureau of Standards (NBS) or using standards that have been
- obtained from suppliers that participate in measurement assurance activi-
- ties with NBS.
These standards shall permit calibrating the system over its intended range of energy, measurement range, and establish monitor response to a solid calibration source. For subsequent CHANNEL CALIBRATION, NBS traceable standard (gas, liquid, or solid) may be used; or a gas, liquid, or solid source that has been, calibrated by relating it to equip-ment that was previously (within 30 days) calibrated by the same geometry _ and type of source standard traceable to NBS.
- (3)- CHANNEL CHECK shall consist of verifying indication of flow during periods of release.
CHANNEL CHECK shall be made at least o,nce per 24 hours on days on which continuous, periodic, or batch releases are made. l l t CALLA' JAY , UNIT 1 3/4 3-67
- o~
TABLE 3.3-13
- p I
RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION E% MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION y C ~5 1. WASTE GAS HOLDUP SYSTEM Explosive Gas ] Monitoring System a. Hydrogen Monitors 1/recombiner 44 2/recombiner 42 b. Oxygen Monitor 2. Unit Vent System a. Noble Gas Activity Monitor-l 1 40 Providing Alarm (GT-RE-21) w i 'E b. Iodine Sampler 1 43 w E c. Particulate Sampler 1 43 d. Flow Rate Monitor 1 39 e.- Sampler Flow Rate Monitor 1-39 i 3. Containment Purge System j a. Noble Gas Activity Monitor - Providing Alarm and Automatic Termination of Release (GT-RE-22, GT-RE-33) 1 41 3 ( 'b. Iodine Sampler 1 43 i c. Particulate Sampler 1 43 d. Flow Rate "Oriter 2 N. A.
- 45 e.
Sampler Flow Rate Monitor 1 39 4 ?) w y
4 -TABLE 3.3-13 (Continued) i TABLE NOTATIONS l At all times. y
- During WASTE GAS HOLD UP SYSTEM operation.
ACTION STATEMENTS ACTION 38 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank (s) may be' released to the environment for up to 14 days provided that prior to initiating the release: a. At least two independent samples of the tank's contents are analyzed, and b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup. Otherwise, suspend release of radioactive effluents via this pathway. ' ACTION 39 - With the number of channels OPERABLE less than required by the - Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated based on fan status and operating curves or actual measurements at least once per 4 hours. ACTION 40 - With the number.of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this , 1 pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours and these samples are analyzed for radioactivity within 24 hours. . ACTION 41 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway. ACTION 42 - With the. Outlet Oxygen Monitor channel inoperable, operation of the system may continue provided grab samples are taken nd analyzed at least once per 24 hours. With both oxygen channel or oth the inlet oxygen and inlet hydrogen monitors inoperable, susp6nd oxygen supply to the recombiner. Addition of waste gas to the system may l continue provided grab samples are taken and analyzed at least once L per 4 hours during degassing operations and at least once per 24 hours during other operations. ACTION 43 - With the number of channels OPERABLE less than required by the j L Minimum Channels OPERABLE requircment, effluent releases via the affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as l required in Table 4.11-2. (- [ ACTION 44 - With the number of channels OPERABLE one less than required by the L Minimum Channels OPERABLE requirement, suspend oxygen supply to the recombiner. J ' ACTION 45 - Flow rate for this system shall be be. sed on fan status and operating . curves or actual measurements. CALLAWAY - UNIT 1 3/4 3-71 [
.s. TAB'LE 4.3-9 9 { RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 'c = 4 ANALOG i. CHANNEL MODES FOR WHICH 'c-CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE 5 INSTRUMENT CHECK CHECK CALIBRATION TEST IS REQUIRED 1. WASTE GAS HOLDUP SYSTEM Explosive Gas Monitoring System a. Inlet Hydrogen Monitor. D N.A. Q(4) M b. Outlet Hydrogen Monitor D N.A. Q(4) M c. Inlet Oxygen Monitor D N.A. Q(5) M d. Outlet Oxygen Monitor D N.A. Q(6) M s w 1 ?. Unit Vent System Y a. Noble Gas Activity Monitor D M R(3) Q(2) d Providing Alarm (GT-RE-21) ~ Iodine Sampler W N.A. N.A. N. A. b. c. Particulate Sampler W N.A. N.A. N.A. ^ ~ d. Flow Rate Monitor D N.'. R(7) Q A e. Sampler Flow Rate Monitor D .N.A. R Q 4 i 3. Containment Purge System a. Noble Gas Activity Monitor - i Providing Alarm and Automatic Termination of Release D P R(3) Q(1) (GT-RE-22, GT-RE-33) i Iodine Sampler W N.A. N.A. N.A. c. Particulate Sampler W N.A. N.A. N.A.
- ~
d. F1ow Rate."M.. iter -B N. A. N.A. R(7) -Q N. A. i e. Sampler Flow Rate Monitor 'D N.A. R Q
i TABLE 4.3-9 (Continued) ~ TABLE NOTATIONS [ ~ At al1 times. y
- During WASTE GAS HOLDUP SYSTEM operation.
(1) TheANALOGCHANNELOPERATIONALTESishallalsodemonstrate-thatautomatic isolation of this pathway and controliroom alarm annunciation occur as . appropriate if any of the follo. wing conditions exists: a. Instrument indicates measured levels above the Alarm / Trip,Setpoint-(isolation and alarm), or b'. Circuit failure (alarm only), or c. Instrument indicates a downscale failure (alarm only), or 'd. Instrument controls not set in operate mode (alarm only). T (2)~ The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of tha following conditions exists: ^ Instrument indicates measured levels aboye the Alarm Setpoint, or a. b. Circuit failure, or c.'~ Initrumentindicatesadownscalefailure,or 3 ~ I 1
- d. % nstrument controls not set in operate mode.
N (3) The initial CHANNEL CALIBRATION shall be performed using one or more of ~ .i the reference (gas or liquid and solid) standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participat.e in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy, measurement range, and establish monitor response to a solid _ calibration source. For subsequent CHANNEL CALIBRATION, NBS traceable ~ standard (gas, liquid, or solid) may be used; or a gas, liquid, or solid source that has been calibrated by relating it to equipment that was previously (within 30 days) by the same geometry and type of source traceable to NBS. (4) The CHANNEL CALIBRATION shall include the use of standard gas samples .. containing a nominal: = = a. One volume percent hydrogen, balance nitrogen, and b. Four volume percent hydrogen, balance nitrogen. i'
- ~
J ~ 5 P CALLAWAY , UNIT'l ~~ 3/4 3-74
INSTRUMENTATION 3/4.3.4 TURBINE OVERSPEED PROTECTION ' LIMITING CONDITION FOR OPERATION a 3.3.4 AileastoneTurbineOverspeedProtectionSystemshallbeOPERABLE. APPLICABILITY: MODES 1, 2^, and 3*. ACTION: .a. With one stop valve or one governor valve per high pressure turbine steam line inoperable and/or with one reheat stop valve or one reheat intercept valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours, or close at least one valve in the affected steam . lines, or isolate the turbine from the steam supply within the next 6 hours. b. With the above required Turbine Overspeed Protection System otherwise inoperable, within 6 hours isolate the turbine from the steam supply. SURVEILLANCE REQUIREMENTS 4._3. 4.1 The provisions of Specification 4.0.4 are not applicable. ~ _ _. -. - - 4.3.4.2 The above required Turbine Overspeed Protection System shall be maintained, calibrated, tested, and inspected in accordance with the Callaway Plant's Turbine Overspeed Protection Reliability Program. Adherence to this + program shall demonstrate OPERABILITY of this system. The program and any revisions should be reviewed and approved in accordance with Specification 6.5.1.6d.o Revisions shall be made in accordance with the provisions of 10 CFR 50.59. j - / -e r ~ ~ l F
- Not applicable in MODE 2 or 3 with all main steam line isolation valves and l
associated bypass valves in the closed position and all other steam flow i paths to the turbine isolated. t CALLAWAY - UNIT 1 3/4.3-76 A
nFT 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 All reactor coolant loops shall be in operation. APPLICABILITY: MODES 1 and 2.* ACTION: With less than the above required reactor coolant loops in operation, be in at least HOT STANDBY within 6 hours. SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours. l' l l I l l Spec 4cabon
- See Special Test Exception ^3.10.4.
t CALLAWAY - UNIT 1 3/4 4-1
y REACTOR COOLANT SYSTEM HOT SHUTDOWN i l LIMITING CONDITION FOR OPERATION 3.4.1.3 At least two of the loops listed below shall be OPERABLE and at least one of these loops shall be in operation:* a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,** b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,** c. Reactor Coolant Loop C and its assnciated steam generator and reactor coolant pump,** d. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump,** e. RHR Loop A, and f. RHR Loop B. APPLICABILITY: MODE 4. ACTION: a. With less than the above required reactor coolant and/or RHR loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is an RHR loop, be in COLD SHUTDOWN within 24 hours. b. With no reactor coolant or RHR loop in operation, suspend all i operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to l return the required coolant loop to operation.
- All reactor coolant pumps and RHR pumps may be deenergized for up to 1. hour provided:
(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
- A reactor coolant pump shall not be started unless uMec: the secondary water temperature of each steam generator is less than 50 F above each of the Reactor Coolant System cold leg temperatures.
CALLAWAY - UNIT 1 3/4 4-3
e REACTOR-COOLANT SYSTEM
- \\
SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Inspection Frecuencies - The above required inservice inspections of ,' steam generator tubes shall be performed at the following frequencies-a. The first. inservice inspection shall be performed after 6 Effective Full Power Months but within'24 calendar months of initial criticality. Subsequent inservice inspections shall be performed at intervals of not'less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no. additional degradation has occurred, the inspection interval may be extended to a maximum of once pe.r 40 months; b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month interv.als fall in Category C-3, the inspection frequt. icy shall be increased to at least once per 20 mcnths. The increase in inspection frequency shall apply until the subsequent Specification 4.4.5.3a.; the inteinspections satisfy the criteria of' rval may then:b.e' extended to a maximum of once per 40 months; and 2- - b-c. Additional', unscheduled inservice inspections shall be performed on,. each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following condition : B 1) Reactor-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3~.4.6.2, or ~ 2) A seismic occurrence greater than the'0perating Basis Earthquake, or 3) A loss-of-coolant accident requiring actuation of the Engineered ' Safety Features, or ~ 4) _ A main steam line or feedwater line break. ~ 9 I e 6 e 6 .CALLAWAY
- UNIT 1-3/4 4-13 9
) ~ REACTOR COOLANT SYSTEM k SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria As used 'in this specification: a.- ~ Imperfection means an exception to the di[nensions, finish or 1) contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of ..the nominal-tube wall thickness, if detectable, may be considered as imperfections; 1 2) Degradation means a service-induced' cracking, wastage, wear or ,y . general corrosion occurring on either inside or outside of a tube; 3) Deoraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness cau. d by degradation; 4) % Degradation means the percentage of the tube wall thickness j affected or removed by degradation; ~ > ~ 5) Defect means an imperfection of such, severity that it exceeds the plugging limit. A' tube containing a defect is defective; 6). Plugging Limit means the imperfection depth at or beyon ich the tube shall be rem'oved from service and is equal to 48% of the nominal tube wall thickness; 7) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, e loss-of-coolant accident, or a steam line or'feedwater line break as
- pecified in Specification 4.4.5.3c., above; 8)
Tube Inspection means an inspection of the steam generato'r tube . from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg;.and CALLAWAY -< UNIT 1 3/4 4-14 tw w a,-w e e ---e- = -,. -., 3r ..+-m.,.er.--- %~gs---r-r-y+r~e,,---n,y-+v-t- -we--'. =y-w y-v
~~ REACTOR COOLANT SYSTEM ^ -3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR CPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE: a. The Containment Atmosphere Particulate Radioactivity Monitoring
- System, b.
The Containment Normal Sump Level Measurement System, and c. Either the Containment Air Cooler Condensate Flow Rate or the . Containment Atmosphere Gaseous Radioactivity Monitoring System. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With only two of the above required Leakage Detection Systems OPERABLE, operation may continue for up to 30 days provided grab samples of the contain- ~ ment atmosphere are obtained and analyzed for gaseous and particulate radio-acitivity or a gamma isotopic analysis of the containment atmosphere is performed using the Post Accident Sampling System at least once per 24 hours when the required Gaseous or Particulate Radioactivity Monitoring System is inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.4.6.1. The Leakage Detection Systems shall be demonstrated OPERABLE by: a. Containment Atmosphere Gaseous and Particulate Monitoring System-performance of CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, b. Containment Normal Sump Level Measurement System performance of CHANNEL CALIBRATION at least once per 18 months, and c. Containment Air Cooler Condensate Flow Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months. CALLAWAY - UNIT 1 3/4 4-18 .e e ow - em e w e em.me - N e we v v y
L i REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION ACTION (Continued) MODES 1, 2, 3, 4, and 5: .With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E microCuries per gram of gross radioactivity, perform the sampling and analysis requirements of Item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is. restored to within its limits. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days with a copy to the Director, Nuclear Reactor Regulation, Attention: Chief, Core Performance Branch, and Chief, Accident Evaluation Branch, U.S'. Nuclear Regulatory Commission, Washington, D.C. 20555. This report shall contain the results of the specific activity analyses together with the following information: a. Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded; b. Results of: (1) the last isotopic analysis for radioiodine performed prior to exceeding the limit, (2) analysis while limit was exceeded, and (3) one analysis after the radiolodine was reduced to less than the limit, including for each isotopic analysis, the date and time of sampling and the radiofodine concentration; c. Clean-up flow history starting 48 hours prior to the first sample in which the limit was exceeded; d. History of degassing operations, if any, starting 48 hours prior to the first sample in which the limit was exceeded; and e. The time duration when the specific activity of the reactor coolant exceeded 1 microcurie per gram DOSE EQUIVALENT I-131. SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4. CALLAWAY - UNIT 1 3/4 4-26
MATERIAL PROPERTY B ASIS COPPER CONTENT . C C.N C : " " " '"/ U/- A S S U M E D TO B E 0.10 WT% ~ ' RT INITIAL
- 50'F uoy RT AFTER 16 EFPY :1/4T,110"F uoy 3/4T. 87'F h
CURVE APPLICABLE FOR HEATUP RATES UP TO 60*F/HR AND 100*F/HR FOR THE SERVICE PERIOD UP TO 7 EFPY AND CONTAINS MARGINS OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS. l LEAK TEST f f f f f LIMIT r \\l ] l f l -1 N J f f f CRITICALITY LIMIT 1 i I I - B AS ED O N 60*/ H R --- f' f ! h HE ATUP CURVE g I il I ? I ( l l ~~~ b i (Li ) 8 I i I [ / [ / CRITICALITY LIMIT E [/ k, HEATUP CURVE BASED ON 100*F/HR-- / c. l // ) 5 / / / / / / / / f [f i/ ~ _[ HEATUP_ / CURVES' f 7 / [ _ CRITICALITY LIMIT BASED ON INSERVICE HYDROSTATIC TEST M TEMPERATURE (255'F) FOR THE -- ir 60'F/HR __j SERVICE PERIOD UP TO *6 EFPY 100*F/ H R e'.sv4 ee m. O O 100 200 300 400 500 ' L-INDICATED AVERAGE TEMPERATURE (DEG. F) FIGURE 3.4-2 REACTOR COOLANT SYSTEM "EATUP LIMITATIONS APPLICABLE UP TO 7 EFPY 'CALLAWAY ~ - UNIT 'l 3/4 4-30
i I/' 3000 [ MATERIAL PROPERTY BASIS COPPER CONTENT dO"^trJi^7;'J:L'/) ASSUMED TFBEO.10 WT% RT INITIAL
- 50*F NOT RT AFTER 7 EFPY :1/4T,110'F NDT 3/4T, 87'F J
CURVE APPLICABLE FOR COOLDOWN RATES / UP TO 100'F/HR FOR THE SERVICE PERIOD UP-TO 7 EFPY AND CONTAINS MARGINS OF 10*F f AND 60 PSIG FOR POSSIBLE INSTRUMENT j ERRORS. I I Hil 1111 I "Il l Ill / 2000 ), 5 H / I / 8 1 / w / = 8 l } .. g 7 .? Y E l / ~ i 1000 Y l R 7 r l. COLDOWN RATES l ('F/H ) g/g 20 - fW l 100-l l 0 o 100 200 300 400 500 l. INDICATED TEMPERATURE (DEG. F) l: l l l FIGURE 3.4-3 i REACTOR COOLANT SYSTEM C00LOOWN LIMITATIONS APPLICABLE UP TO 7 EFPY l CALLAWAY'- UNIT 1 3/4 4-31
EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T,yg > 350 F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of: a. One OPERABLE centrifugal charging pump, ' b. ' One OP' ERA 8LE Safety Infection pump, ~ ~ . c. One OPERABLE RHR heat exchanger, --d.- - One OPERABLE RHR pump, and e. ~An OPERABLE flow path capable of taking suction from the refueling - ---water storage tank on a Safety Injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation. APPLICABILITY: MODES 1, 2, and 3.* e f ACTION: ~ a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following e - 6 hours. O ' b ~. In'the event the ECCS is actuated and injects water into the Reactor ^ Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ- ~ ~ ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this - - -Special Report whenever its value exceeds 0.70. [ smq. 4^The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into M006 3 for the centrifugal charging pump and the Safety Injection pumps declared inoperable pursuant to Specification 4.5.3.2 provided the centrifugal charging pump and the Safety Injection pumps are restored to OPERABLE status within 4 hours or prior to the temperature of one or more of the RCS cold legs exceeding 375 F",
- ,. ~
~ CALLAWAY - UNIT 1 3/4 5-3
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) - 2) A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion. e. At least once per 18 months, during shutdown, by: 1) Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection test signal and/or on Automatic Switchover to Containment Sump from RWST Level-Low-Low coincident with Safety Injection test signal; and 2) Verifying that each of the following pumps start automatically 4 upon receipt of a Safety Injection actuation test signal: a) Centrifugal charging pump, b) Safety Injection pump, and c) RHR pump. f. By verifying that each of the following pumps develops the required -~- -differential pressure on recirculation flow when tested pursuant to -Specification 4.0.5: 1) Centrifugal charging pump 1 2400 psid, 2) Safety Injection pump 1 1445 psid, and 3) RHR pump 1 165 psid. g. By verifying the correct position of each mechanical position stop for the following ECCS throttle valves: 1) Withir4! 4 hours fo'11owing completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and 2) At least once per 18 months. HPSI System CVCS System Valve Numbers Valve Numbers EMV095 EMV109 BGV-198 ~ ~.. EMV096 EMV110 BGV-199 3"' ENV097 ,EMV089 BGV-200 EMV098 EMV090 BGV-201 EMV107 EMV091 BGV-202 27/108 EMV092 CALLAWAY - UNIT 1 3/4 5-5 W-
S l ~ EMERGENCY CORE COOLING SYSTEMS S SURVEILLANCE REQUIREMENTS (Continued) h. By performing a flow balance test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that: 1) For. centrifugal charging pump lines, with a single pump running: a) The sum of the injection line flow -rates, excluding the highest flow rate, is greater than or equal to 346 gpm, and b) The total pump flow rate is less than or equal.to 550 gpm. 2) For Safety Injection pump lines, with a single pump running: a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 462 gpm,' and b) The total pump flow rate is less than or equal to 650 gpm. By performing a flow test, during shutdown, following completion of ' ' ') i. modifications'to the RHR subsystems that alter the subsystem flow characteristics and verifying that for RHR pump lines, with a single pump running: 1) The sum of the injection line flow rates is greater than or equal to 3800 gpm, and 12 ) The total pump flow rate is less than or equal to 5500 gpm. / ] 9 s O 9 e 9 e 9 e O CALLAWAY -< UNIT 1 3/4 5-6 -w -,y- -v
~ h EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable requirements of Specification 4.5.2. 4.5.3.2- ' All centrifugal charging pumps and Safety Infection pumps, except the above required OPERABLE pumps, shall be demonstrated inoperable
- by verifying that the motor circuit breakers are secured in the open position within 4 hours after. entering MOBE 4 from M00E 3 or prior to the temperature of one or more of the RCS cold legs decreasing below 325 Fand at least once per 31 days thereafter.
l w 4 '~ i l-
- An inoperable pump may be energized for testing or for filling accumulators provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.
l CALLI.WAY - UNIT 1 3/4 5-8
EMERGENCY CORE COOLING SYST' EMS 1 200 F 3/4.5.4 ECCS' SUBSYSTEMS - T,y ' LIMITING CONDITION FOR OPERATION ~ 3.'5.4 All Safety Injection pumps shall be inoperable. APPLICABILITY: MODE 5 and MODE 6 with the reactor vessel head on. ACTION: With a Safety Injection numn_0PERABLE, restore all Safety Injection pumps to hn..inoperablestatus(Eithin4hoursy) SURVEILLANCE RE0VIREMENTS 4.5.4 All Safety Injection pumps 'shall be demonstrated inoperable *.by verifying that the motor circuit breakers are secured in the open position at least once .per 31 dayse i
- An. inoperable pump may be energized for testing!or for filling accumulators provided the discharge at the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.
CALLA'r.'AY -< UNIT 1 3/4 5-9
4 4 CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE 1 LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to: . a. - An overall integrated leakage rate of: 1) Less than or equal to L, 0.20% by weight of the containment air per 24 hours at P,,848 psig, or 2) Less than or equal to L, 0.14% by weight of the containment airper24hoursatarkducedpressureofP,24psig. t A combined leakage rate of less than h qual td)0.60 L, hen pressuriz for all b. penetrations and valves subject to Type is ano c tests, w to P,, 48 psig. APPLICABILITY: MODES 1, 2, 3, and 4. ' AC. TION: With either the measured overall integrated containment leakage rate , exceeding 0.75 L or 0.75 L, as applicable, or the measured combined leakage rate for all pen $trations a id valves subject to Types B and C tests exceeding 0.60 L restore the overall integrated leakage rate to less than 0.75 L or + less t$a,n 0.75 L, as applicable, and the combined leakage rate for all hene-t trations subject to Type B and C tests to less than 0.60 L, prior to increasing ~the Reactor Coolant System temperature above 200*F. ~ SURVEILLANCE REQUIREMENTS 1 ,4.6.1.2 The containment leakage rates shall be demonstrated rt the following test schedule and shall be determined in conformance with the criteria speci-fied in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI N45.4-1972: i Three Type A tests (Overall Integrated Containment Leakage Rate) a. shall be conducted at 40 1 10 month intervals during shutdown at a pressure not less than either P, 48 psig, or at P, 24 psig,.during each 10 year service period. T$ethirdtestofeabhsetshallbe conducted during the shutdown for the 10 year plant inservice inspection; 'CALLAWAY - UNIT 1 3/4 6-2 t
P I CONTAINMENT SYSTEMS INTERNAL PRESSURE ~iMITINGCONDITIONFOROPERATION L 3.6.1.4 Primary containment internal pressure shall be mai,ntained between +1L and -1 psig. + 1.5 - o.3 APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the containment internal pressure outside of the limits above, restore the internal pressure to within the limits within 1 hour or be in at least HOT 3 STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. .___ u SURVEILLANCE REQUIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to be -within the limits at least once per 12 hours. l I e i l t l CALLAWAY'- UNIT-1 3/4 6-6
CONTAINMENT SYSTEMS P t' CONTAINMENT VENTILATION SYSTEM T ', LIMITING CONDITION FOR OPERATION 3.6.1.7 Each containment purge supply and exhaust isolation valve shall be OPERABLE and: a. Each 36-inc'h containment shutdown purge supply and exhaust isolation valve shall be closed and blank flanged, and . b. The.18-inch containment mini purge _ supply and exhaust isolation valve (s) may be open for up tq]f600 hours)during a calendar year. ~ APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With a 36-inch containment purge supply and/or exhaust isolation' a. valve open or not blank flanged, close and/or. blank flange that valve or isolate the penetration (s) within 4 hours, otherwise be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN ~ within the following 30 hours. b. With the 18-inch containment mini purge supply and/or exhaust isolation valve (s) open for more than 2000 hours during.a calendar year, close the open 18-inch valve (s) or isolate the penetration (s) within 4 hours, otherwise be in at least' HOT STANDBY within the next 6 hours, and in COLD SHUTDOWN within the following 30 hours. i c. With a containment purge supply and/or exhaust isolation valve (s) having a measured-leakage rate in excess of the limits of Specifications 4.6.1.7.2 and/or 4.6.1.7.4, restore the inoperable valve (s) to OPERABLE status within 24 hours, otherwise be 4n at least HOT' STANDBY within the next 6 hours, and in COLD SHUTDOWN within the following 30 hours. i l. l 4 ? O CALLAWAY'- UNIT 1 3/4 6-11 ,,wge- --e-~-q+,- <g- .,---,-<w,-. - - - - - - -.,.e, g,-- - + -,,, - ,-,---.-,..,,,,,,,-,,,,--r,- ,--,,-r,-,,,,,-,,,_,,,-,,,,,,n-,,,...~,--n-------. w
4 4 CONTAINMENT SYSTEMS ( SURVEILLANCE REQUIREMENTS Is 4.6.1.7.1 Each 36-inch containinent. shutdown purge supply and exhaust' isolation valve (s)*'shall be verified blank flanged and ' closed at 1 east once per 31 days. 4.6.1.7.2 Each 36-inch containment shutdown purge supply and exhaust isolation valve and its associated blank flange,shall be ~1eak.testad a't least once per 24 months and following each reinstal'lation of the'31ank flange when pressurized to P, 48 psig, and verifying that when the measured leakage ' rate for these valv$s and flanges, including stem leakage, is added to the leakage' rates determined pursuant to Specification 4.6.1.2d for all o.t.her Type B and C penetrations, the combined leakcge rate issless than O(6,0 LN,.- y 3 A 4.6.1.7.3 The cumulative time that all 18-inch containment mini purge supply ~ and exhaust isolation valves have bde~n open during 5 calenday year shall be determined at least once per 7 days. 5 s. 4.6.1.7.4 At least once per 3.inonths each 18-inch contajnmSnt mini purge supply and exhaust isolation valve with resilient materini seals shall be dem strated OPERABLE by verifying that the measured leakage rate is less than . 05 L, J en pressurized to P,. x-
- ' (.N, 5
~ .r.. + ~ w.,_ s q se s ~ g x x 1 s-g' ' N .\\l 5 ~ 'u ' s t, x <,,- ( s 4 ( s e ~. l - ~ _,1 x, '*Except valves and flanges which are located inside containment > These valves shall be verified to be closed'with their blank flanges' installed prior to entry into MODE 4 fo119 wing each COLD SHUTDOWN. A s 3/4 6-12 CALLAWAY, UNIT <1 \\ 3 Q %[, g s ...-......- -O.
DR/H = TABLE 3.6-1 (Continued) CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds) 1. Phase "A" Isolation (active) - (Continued) P-101 GS HV-31 Sample Line to CTMT A,C 5 Atmos Monitor P-101 GS HV-32 Sample Line to CTMT A,C 5 Atmon Monitor S P-97 GS HV-33 Hydrogen Sample Return A,C 5 From PASS P-97 GS HV-34 Hydrogen Sample Return A,C 5 From PASS F-99 GS HV-36 Sample Line to CTMT A,C 5 Atmos Monitor P-99 GS HV-37 Sample Line to CTMT A,C 5 Atmos Monitor P-56 GS HV-38 Sample Return CTMT A,C 5 Atmos Monitor P-56 GS HV-39 Sample Return CTMT A,C 5 Atmos Monitor P-44 HB HV-7126 RCDT Vent Inside CTMT C 10 P-26 HB HV-7136 RCDT Pumps Disch Hdr C 10 Outside CTMT Iso P-44 HB HV-7150 RCDT Vent Outside C 10 CTMT P-26 HB HV-7176 RCDT Pumps Disch Hrd C 10 Inside CTMT Iso P-30 KA FV-29 Reactor Bldg Instr Air C 5 SupplyOutsideCTMTIso P-32 LF FV-95 CTMT Normal Sumps to C 30 Floor Drain Tank Inside CTMT Iso CALLAWAY - UNIT 1 3/4 6-20
TABLE 3.6-1 (Continued) CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME _PENFTRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds) 1. Phase "A" Isolation (active) - (Continued) P-S2 LF FV-96 CTMT Normal Sumps to C 4 Floor Drain Tank Outside CTMT Iso P-93 SJ HV-5 PZR/RCS Liouid Sample C 5 Inner CTMT Iso P-93 SJ HV-6 PZR/RCS Liquid Sample C 5 Outer CTMT Iso P-69 SJ HV-12 PZR Vapor Sample Inner, C 5 CTMT Iso Outer P-69 SJ HV-13 PZRLVapor Semple -Ineee-C 5 CTMT Iso c P-95 - SJ HV-18 Act,umulator Sample C 5 Inner CTMT Iso; P-95 SJ HV-19 Accumulator Sample C 5 Outer CTMTiIso P-93 SJ HV-127 PZR/RCS Liquid Sample C 5 Outer CTMT 1so P-64 SJ HV-128 PZR/RCS Liquid A,C 5 Sample Inner CTMT Iso P-64 SJ HV-129 PZR/RCS Liquid A,C 5 Sample' Outer'CTMT Iso.. P-64 " SJ HV-130 PZR/RCS Liqui A,C 5 Sample Outer CTMT Iso Valve r. P-57 SJ HV-131 PASS Discharge to A,C 5 RCDT P-57' SJ HV-132 s PASS Discharge to A', C 5 RCDT 2. Phase "A" Isolation (passive)* P-58, EM H"- %88 Accumulator Tank Fill C N.A. Line Iso Valve 'May be opened on'an intermittent basis under administrative control. CALLAWAY - UtlIT 1 3/4 6-21, ^ ' .f i 3 \\ ^
iABLE3.6-1(Continued) u CONTAINMENT ISOLATION VALVES MAXIMUM i TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds) ~2. Phase "A" Isolation (passive)* - (Continued) P-16 EN HV-01 CTMT Recirc Sump to CTMT A N.A. . Spray Pump A Iso P-13 EN HV-07 CTMT Recire Sump to CTMT A N.A. Spray Pump B Iso St%o P.45 .EP HV-800-CTMT Nitrogen Supply C N.A. -Iso Valve P.65 .GS HV-20 Hydrogen Purge Inner C N.A. CTMT Iso P 65 .GS.HV-21 Hydrogen Purge Outer C N.A. CTMT Iso P-67 KC HV-253 Fire Protection System C N.A. Hdr Outer CTMT Iso '3.. Phase "B"_ Isolation (active) .P-74 EG HV-58 CCW'to RCS Iso C 30 'p-75 EG HV-59 CCW Return From C 30 RCS Iso P-75 EG HV.CCW Return From C 30 RCS Iso P-76 EG HV-61 CCW Return From C 30 ~RCS Iso 5-76. EG HV-62 CCW Return From C 30 RCS Iso l
- 4.
Containment Purge Isolation (actiive) V-161 GT HZ-4 CTMT Mini-Purge C 3 Supply Outside CTMT Iso -V-161 GT HZ-5 CTMT Mini-Purge C 3 Supply Inside CTMT Iso 3 ay be opened on an intermittent basis under administrative control. M l CALLAWAY - UNIT 1 3/4 6-2? L
DRUT TABLE 3.6-1 (Continued) CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds) 6. Remote Manual - (Continued) P-14 EJ HV-8811B CTMT Recirc Sump to A N. A., RHR Pump B Sucion P-21 EJ HV-8840 RHR Hot Leg Recirc A 'N.A. Iso Valve P-87 EM HV-8802A* SI Pump A Disch Hot A N.A. Leg Iso Valve P-48 EM HV-8802B* SI Pump B Disch Hot A N.A. Leg Iso Valve P-49 EM HV-8835 SI Pumps Disch to A. N.A. Cold Leg Iso Valve P-89 EN 41V-6 CTMT Spray Pump A A N.A. Disch. Iso Valve arge. P-66 EN HV-12 CTMT Spray Pump B A N.A. Discharge Iso Valve 7. Active for SIS P-80 BG HV-8105 CVCS Charging Line C N.A. P-88 EM HV-8801A Boron Injection to A N.A. RCS Cold Legs P-88 EM HV-8801B Boron Injection to A N.A. RCS Cold Legs 8. Hand-Operated and Check Valves P-41 BB V-118 RCP A Seal C N.A. Water Supply P-22 BB V-148 RCP B Seal C N.A. Water Supply P-39 BB V-178 RCP C Seal C N.A. Water Supply P-40' BB V-208 RCP 0 Seal C N.A. Water Supply ^These valves were assumed to be m'osed during the accident analysis and are normally closed but may be opened on an intermittent basis under administrative control. CALLAWAY - UNIT 1 3/4 6-26 1
TABLE 3.6-1 (Continuedl t CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED ~ (Seconds) 8. Hand-Operated and Check Valves - (Continued) P-66 EN V-017. CTMT Spray Pump B A N.A. to CTMT Spray Nozzles P-45 EP V-046 Accumulator Nitrogen C N.A. Supply Line P-43 HD V-016 -Auxiliary Steam to C N.A. Decon System P'-4 3 HD V-017 Auxiliary Steam to C N.A. Decon System P-63 -KA V-039 Rx Bldg Service Air C N.A. Supply ~ ~ ~ P-63 KA V-118 Rx Bldg Service Air C-N.A. Supply P-30 KA V-204 Rx Bldg Instrument C N.A. Air Supply P-98 KB V-001 Breathing Air Supply C N.A. to Rx Bldg. P-98 KB V-002 Breathing Air Supply C N.A. l P-67 KC V-478 Fire Protection C N.A. Supply to RX Bldg. l P-57 SJ V-111 Liquid Sample from A,C N.A. PASS to RCDT l CALLAWAY, UNIT 1 3/4 6-29
? ~ CONTAINMENT SYSTEMS ( 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN ANALYZERS y LIMITING CONDITION FOR OPERATION ~ 3.6.4.1 Two independent containment hydrogen analyzers shall be OPERABLE. APPLICABILITY: MODES 1 and 2. - - - ACTION: - With one containment hydrogen analyzer inoperable, restore the inoperable - - analyzer to OPERABLE status within 30 days er be in at least HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS ^ 4.6.4.1 - Each containment hydrogen analyzer shall be demonstrated OPERABLE,by r . _. - the performance of an ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and at least once per 31 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gas containing ten volume percent hydrogen, balance nitrogen. g N CALLAWAY - UNIT 1 3/4 6-30 4 e: ,--r
- +--, - -,,
y,-y ., - -,..-,y--.-r.w,~..--y,4 ,,,w,y-,,,.,-
O CONTAINMENT SYSTEMS HYDROGEN CONTROL SYSTEMS ?- = LIMITING CONDITION FOR OPERATION ~3.6.4'2 A Hydrogen Control tem-shall_ba_0PERABLE with two independent ydroae g ecombiner Svste @% pf MPFLich51LITY: MODES 1 and 2 ' ACTION': With one of the two independent Hydrogen Recombiner Systems inoperable, restore the inoperable Hydrogen Recombiner System to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.6.4.2 Each Hydrogen Recombiner System shall bc demonstrated OPERABLE: a. At least once per 6 months by verifying, during a Hydrogen Recombiner System functional test, that the heater air temperature increases to .T.1._.J ::._ greater than or equal to 1150*F within 5 hours; and- ' ~ ~ ~~ b. At least once per 18 months by: 1) Performing a CHANNEL CALIBRATION of all Hydrogen Recombiner System instrumentation and control circuits, 2)' Verifying through a visual examination that there is no evidence of abnormal conditions within the Hydrogen Recombiner System enclosure (i.e., loose wiring or structural connections, deposits of. foreign materials, etc.), and 3) Verifying the integrity of all heater electrical circuits by performing a resistance to ground test following the above required functional test. The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms. = bW ~. n. s. } CALLAWAY - UNIT 1 3/4 6-31
PLANT SYSTEMS SPECIFIC' ACTIVITY LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the Secondary Coolant System shall be less than or equal to 0.1 microcurie / gram DOSE EQUIVALENT I-131. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the specific activity of the Secondary Coolant System greater than 0.he-microcurie / gram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.1.4 The specific activity of the Secondary Coolant System shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-1. l i t CALLAWAY - UNIT 1 3/4 7-7 e
DRUT TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY 1. Gross Radioactivity At least once per 72 hour's. Determination 2. Isotopic Analysis for DOSE a) Once per 31 days, whenever the EQUIVALENT I-131 Concentration gross radioactivity determi-nation indicates concentrations greater than 10% of the allow-able limit for radioiodines. b) Once per 6. months, whenever the gross radioactivity r 4 determination indicates ,~ concentrations less than or equal to 10% of the allowable limit for radiciodines. o CALLAWAY - UNIT 1 3/4 7-8 -s-w
c. ' PLANT SYSTEMS SURVEILLANCE REQUIREMENTS '(Continued)- c.- At least once per 18 months, or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, oi- (2) following ' painting,. fire or chemical release in any ventilation, zone communicating with the system by,: ~ 1) Verifying that the Control Room Emergency Ventilation System satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1% and uses the test procedure guidance in Regulatory Positions C.5.a. C.5.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 2000 cfm i 10% for the Filtration System and 2000 cfm i 10% for the Pressurization System with 500 cfm i 10% going through the Pressurization System filter adsorber unit; j 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regul.atory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%; and 3) Verifying a system flow rate of 2LJO cfm + 10% for the Filtration System and 2000 cfm i 10% for the Pre:st rization System with 500 cfm 10% going through the Pressurization System filter adsorber unit during system' operation when tested in accordance with ANSI N510-1975. 5' d._ After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal, that a laboratory analysis of a represen-tative carbon sample obtained in accordance with Regulatory Position C.6.biof Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%; e. At least once per 18 months by: 1) Verifying.that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 5.4 inches Water Gauge while operating the' system at a flow rate of 2000 cfm + 10% for the Filtration System and 500 cfm i 10% for the Pressurization System filter adsorber unit; i 2) Verifying that on a Control Room Ventilation Isolation test signa, .the system automatically switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks; 3) Verifying that the system maintains the control room at a r positive pressure of greater than or equal to 1/8 inch Water Gauge at less than or equal to a pressurization flow of 400 cfm I. relative to adjacent areas'during system operation; and 4) Verifying that the Pressurization System filter adsorber unit F .. 3ters dissipate 15 1 2 kW in the Pressurization System when i tested in accordance with ANSI N510-1975. l l: CALLAWAY, UNIT 1_ 3/4 7-15 l L
PLANT SYSTEMS 3/4.7.10 FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.10.1 The Fire Suppression Water System shall be OPERABLE with: a. At least two fire suppression pumps, each with a capacity of 1500 gpm, - with their discharge aligned to the fire suppression header; b. Two separate water mnniv tanks, each with a minimum level,of 6 1 feet (260 g gallons) and c. An OPERABLE flow path capable of taking suction from both fire water storage tanks and transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves, the last. valve ahead of the water flow alarm device on each sprinkler or hose standpipe, and the last valve ahead of the deluge valve on each Deluge or Spray System required to be OPERABLE per Specifications 3.7.10.2, and 3.7.10.4. ~ APPLICABILITY: At all times. ACTION: + a. With one of the two required pomps and/or one water supply inoperable, restore the inoperable equipment to OPERABLE status within 7 days or provide an alternate backup pump or supply. The provisions of Speci-fications 3.0.3 and 3.0.4 are not applicable. b. With the Fire Suppression Water System otherwise inoperable establish a backup Fire Suppression Water System within 24 hours. i SURVEILLANCE REQUIREMENTS 4.7.10.1.1 The Fire Suppression Water System shall be demonstrated OPERABLE: a. At least once per 7 days by verif
- g the water level in each f, ire water storage tank exceeds 1 feet (260,000 gallonsQ b.
..At least once per 31 days on a STAGGERED TEST BASIS by starting the [~ electric motor-driven pump and operating it for at least 15 minutes &, on recirculation flow, c. At least once per 31 days by verifying that each valve (manual, power- ' ope' rated,'or automatic) in the flow path is in its correct position, ( i CALLAWAY - UNIT 1 3/4 7-27 i
DRIFi PLANT SYSTEMS SPRAY AND/0R SPRINKLER SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.10.2 The following Spray and/or Sprinkler Systems shall be OPERABLE: a. Wet Pipe Sprinkler Systems Building Elevation Area Protected Auxiliary 2000 North Electric Cable Chase Auxiliary 1988/2000 South Electric Cable Chase Control 1974 - 2073 Vertical Electrical Chases Control 1974 Pipe Space and Tank Room Control 1992 Cable Area Above Access Control b. Pre-Action Sprinkler Systems Building Elevation Area Protected Auxiliary 1974 Cable Trays
- Auxiliary 2000 Cable Trays
- Auxiliary 2026 Cable Trays
- Control 2032 Lower Cable Spreading Room Control 2073 Upper Cable Penetration Area Reactor 2026 North Cable Penetration Area Reactor 2026 South Cable Penetration Area Diesel Gen. (E) 2000 East Diesel Generator Room Diesel Gen. (W) 2000 West Diesel Generator Room c.
Water Sprays Systems Building Elevation Area Protected Auxiliary 2000 Auxiliary Feedwater Pump Turbine ggs prgggg.g grg gggg. a r APPLICABILITY: Whenever equipment protected by the Spray / Sprinkler System is' required to be OPERABLE. ACTION: With one or more of the above required Spray and/or Sprinkler Systems a. inoperable, within 1 hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol. b. The provisions of Specifications 3.0.3 and 3.0.4.are not applicable. SURVEILLANCE REQUIREMENTS 4.7.10.2 Each of the above required Spray and/or Sprinkler Systems shall be demonstrated OPERABLE: a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path is in its correct position;
- Areas contain redundant systems or components which could be damaged.
CALLAWAY - UNIT 1 3/4 7-30
. - ~_- - l DIMFT L PLANT SYSTEMS FIRE HOSE STATIONS 4 LIMITING CONDITION FOR OPERATION 4 .3.7.10.4' The fire hose stations given'in-Table 3.7-F shall be OPERABLE. l APPLICABILITY: Whenever equipment in the areas' protected by the fire hose stations-is required to be OPERABLE. ACTION: ' 4 l = a. .With one or more of the fire hose stations given in Table 3.7-5 f' . inoperable, provide equivalent capacity backup hose protection to the unprotected area from the spare hose connection on the adjacent OPERABLE standpipe. If two standpipe hose connections are not available at the adjacent OPERA 8LE hose station (s), provide gated i wye (s) to ensure continued OPERABILITY of the affected hose station. Where it can be demonstrated that the physical routing of the backup - hose would result in a recognizable hazard to operating technicians, plant equipment, or.the hose itself, or would require the blocking open of a fire door, the hose shall be stored at the point of origin and properly identified as to its intended use. The above action i. 'shall be accomplished within 1 hour if the inoperable fire hose is the primary means of fire suppresrion; otherwise route the addi-i tional hose within 24 hours. - 4 b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. l SURVEILLANCE REQUIREMENTS-4 4.7.10.4 Each of the" fire hose stations given in Table 3.7-5 shall be ' demonstrated OPERABLE: a. At least once per 31 days, by a visual inspection of the fire hose stations accessible _during plant ' operations to assure all required
- equipment is at the station; b.
'At least once per 18 months, by: t 1) Visual inspection of the' stations not accessible during plant l operations to assure all required equipment is at the station, 2) Removing the hose for inspection'and reracliing, and 3) - Inspecting all gaskets and replacing any degraded gaskets in the couplings. j. c. At least once per 3 years, by: 1) - Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage, and j -2) Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above maximu'm fire main operating pressure, whichever is greater. y e ~ ~CALLAWAY - UNIT 1 3/4 7-33 P ,,-+,,w-- ---,r.g-,,. - U-.* r,, .--,-,,---.,--w,r.--w,w,r,-p. ,we.--~..+r.e-,em e -y m -+ _ we s E-
DRlfT 4 TABLE 3.7-2F FIRE HOSE STATIONS BUILDING ELEVATION AREA HOSE RACK Auxiliary 1974 1122 KC-HR-051 Auxiliary 1974 1122 KC-HR-047 Auxiliary 1974 1120 KC-HR-031 Auxiliary 1974 1120 KC-HR-025# Auxiliary 1974 1101 KC-HR-023# Auxiliary 1974 1101 KC-HR-040 Auxiliary 1974 1101 KC-HR-042 Auxiliary 1988 1201 KC-HR-024 Auxiliary 2000 1329 KC-HR-111 Auxiliary 2000 1320 KC-HR-048 Auxiliary 2000 1320 KC-HR-046# Auxiliary 2000 1314 KC-HR-030 Auxiliary 2000 1321 KC-HR-029# Auxiliary 2000 1301 KC-HR-035# Auxiliary 2000 1301 KC-HR-039 Auxiliary 2000 1301 KC-HR-041# Auxiliary 2026 1408 KC-HR-049 Auxiliary 2026 1408 KC-HR-044 Auxiliary 2026 1408 KC-HR-032# Auxiliary 2026 1408 KC-HR-026# Auxiliary 2026 1401 KC-HR-034 Auxiliary 2026 1403 KC-HR-037# __._.------ - Auxil ia ry 2047 1506 KC-HR-050 ^ Auxiliary 2047 1513 KC-HR-043 '. Auxiliary 2047 1506 KC-HR-045 Auxiliary 2047 1501 KC-HR-038 Auxiliary 2047 1504 KC-HR-033 Auxiliary 2047 1502 KC-HR-027 Auxiliary 2064 1119 KC-HR-028# Control 1974 3101 KC-HR-002# Control 1974 3101 KC-HR-014# Control 1984 3204 KC-HR-015# Control 1984 3221 KC-HR-001# Control 2000 3301 KC-HR-004# Control 2000 3301 KC-HR-017# Control 2000 3302 KC-HR-016# Control' 2016 3401 KC-HR-005 Control 2016-3401 KC-HR-019 Control 2016 3401 KC-HR-018 e CALLAWAY - UNIT 1 3/4 7-34
^ DRIT' 4 TABLE 3.7-5 (Continued) FIRE HOSE STATIONS / BUILDING ELEVATION AREA HOSE RACK Control 2032 3501 KC-HR-006# Control 2032 3501 KC-HR-020# Control 2047 3604 KC-HR-007 Control 2047 3616 KC-HR-021 Control 2073 3801 KC-HR-008# Control 2073 3801 KC-HR-022# Reactor 2000 2201 KC-HR-120* Reactor 2000 2201 KC-HR-131* Reactor 2000 2201 KC-HR-124* Reactor 2000 2201 KC-HR-129* Reactor 2026 N.A. KC-HR-121* Reactor 2026 N.A. KC-HR-132*# Reactor 2026 N.A. KC-HR-125* Reactor 2026 N.A. KC-HR-130* Reactor 2047 N.A. KC-HR-128* Reactor 2047 N.A. KC-HR-122^ Reactor 2047 N.A. KC-HR-126* Reactor 2068 N.A. KC-HR-123* Reactor 2068 N.A. KC-HR-127* Fuel 2000 6102 KC-HR-142# Fuel 2000 6102 KC-HR-054# Fuel 2000 6102 KC-HR-143 Fuel 2000 6104 KC-HR-057 Fuel 2026 6201 KC-HR-133 Fuel 2026 6203 KC-HR-052 Fuel 2047 6301 KC-HR-055# Fuel 2047 6302 KC-HR-056# Fuel 2047 6301 KC-HR-053# ESW 2000 N.A. KC-HR-140 ESV 2000 N.A. KC-HR-141' TABLE NOTATIONS 4
- Secondary means of_ fire suppression to Water Sprays / Deluge or Halon System.
- Fire hose for station to be stored external to Reactor Building.
CALLAWAY - UNIT 1 3/4 7-35
1 PLANT SYSTEMS 3/4.7.12 AREA' TEMPERATURE MONITORING LIMITING CONDITION FOR OPERATION 5 3.7.12 The temperature limit of each area given in Table 3.7-fi shall not be exceeded for more than 8 hours or by more than 30 F. APPLICABILITY: Whenever the equipment in an affected area is required to be OPERABLE. ACTION: 5 Withoneor)moreareasexceedingthetemperaturelimit(s)shownin a. Table 3.7-E for more than 8 hours, prepare and submit to the Commis-sion within 30 days, pursuant to Specification 6.9.2, a Special Report that provides a record of the cumulative time and the amount by which the temperature in the affected area (s) exceeded the limit (s) and an analysis to demonstrate the continued OPERABILITY of the affected equipment. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. 5-b. Withoneor)moreareasexceedingthetemperaturelimit(s)shownin Table 3.7-If by more than 30 F, prepare and submit a Special Report as required by ACTION a. above and within 4 hours either restore the area (s) to within the temperature limit (s) or" declare the equipment in the affected area (s) inoperable. SURVEILLANCE REQUIREMENTS 5 4.7.12 The temperature in each of the areas shown in Table 3.7-6 shall be determined to be within its limit at least once per 12 hours. CALLAWAY - UNIT 1 3/4 7-37
DRlFT . 5 TABLE 3.7-( AREA TEMPERATURE MONITORING MAXIMUM TEMPERATURE AREA LIMIT ( F) 1. ESW Pump Room A 119 2. ESW Pump Room B 119 3. Auxiliary Feedwater Pump Room A 119 4. ' Auxiliary Feedwater Pump Room B 119 5. Turbine-Oriven Auxiliary Feedwater Pump Room 147 6. ESF Switchgear Room I 87 7. ESF Switchgear Room II 87 8. RHR Pump Room A 119 9. RHR Pump Room B 119 10. CTMT Spray Pump Room A 119 11. CTMT Spray Pump Room B 119 12. Safety Injection Pump Room A 119 13. Safety Injection Pump Room B 119 14. Centrifugal Charging Pump Room A 119 15. Centrifugal Charging Pump Room B 119 16. Electrical Penetration Room A 101 17. Electrical Penetration Room B 101 18. Component Cooling Water Room A 119 19. Component Cooling Water Room B 119' 20. Diesel Generator Room A 119 21. Diesel Generator Room B 119 22. Control Room 84 CALLAVAY - UN!T 1 3/* 7-30
_ ~. a ELECTRICAL' POWER SYSTEMS i F LIMITING CONDITION FOR OPERATION . ACTION (Continued) f 2. When in MODE 1, 2, or 3, the steam-driven auxiliary feedwater pump is OPERABLE. If these conditions are not satisfied within 2 hours be in at least 3 . HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. d. With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diese1' generators by performing Specification 4.8.1.1.3a.4) within 1 hour and at least once per 8 hours thereafter, unless the diesel generators are already operat-ing; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours. With only one offsite source restored, restore at least two offsite circuits to OPERABLE status within 72 hours from time of initial loss or be in at least HOT STANDBY within the next 4 6 hours and in COLD SHUTDOWN within the following 30 hours, e. With two of the above required diesel generators inoperable, i demonstrate the OPERABILITY of two offsite A.C. circuits by perform-ing Specification 4.8.1.1.1 within 1 hour and at least once per 4' 8 hours thereafter; restore at least one of the inoperable diesel I generators to OPERABLE status within 2 hours or be in at least NOT t ' STAND 8Y within the next 6 hours and in COLD SHUTDOWN within the' following 30 hours. Restore at least two diesel generators to 2 OPERABLE status within 72 hours from time of initial loss or be in .least HOT. STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS i 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the onsite Class IE distribution system shall be: a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and 5. 0;m;;;tr ted OPERACLC in eccecdence with the OPERACILITY er the
- pplic:ble Fir: Det::tien Instrgmentatien (Specificetien 3.0.3.7)
- nd th: pplic:ble fire Suppressica Systems (0pecificetion 0.7.10)
-fe-the ESF tr:n:fer;;re, XN001 end XN002. 4.S.1.1.2 The :: lid state 1 cad ;;quencer 1;gic ; hell be d;;sastrated OPERACLE 4 -b" perfer *ng :n ACTU/ TION LOCIC TEST :nd a " ASTER RELAY TEST st 1: st cace pee-01 dey; end ; SLA"C RELAY TEST st 1:::4 nce per 92 d y;. i i i ~ CALLAWAY - UNIT 1 3/4 8-2 ,an- ,.w.--+--.-,. -,nn.---n.,,--..,,.- ..m,,...nm.,. _,-,- .-.-,..~ ~ ,-a.e-.- ,,.rm,--,-,.- n-e
l ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2 4.8.1.1.) Each diesel generator shall be demonstrated OPERABLE: a. In accordance with the frequency specified in Table 4.8-1 on a STAGGERED TEST BASIS by: 1) Verifying the fuel level in the day tank, 2) Verifying the fuel level in the fuel storage tank, 3)- Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day tank, 4) Verifying the diesel starts from ambient condition and accele-rates to at least 514 rpm in less than or equal to 1 - conds.* The generator voltage and frequency shall be 4000 _ 320 olts and 60 + 1.2 Hz within 12 seconds
- after the start s al.
The diesel generator shall be started for this test by using one of the following signals: a) Manual, or b) Simulated loss-of-offsite pow r by itself, or c) Safety Injection test signal. 5) Verifying the genera pt ds synchronizad, landad to areater than or equal to 6201 kKin less than or equal to 60 seconds, operates with a load greater than or equal to bzul kW for at st 60 minutes, and 6)- Verifying the diesel generator is aligned to provide standby power to the associated emergency busses. b. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to 1 hour by checking for and removing accumulated water from the day tanks; c. At least once per 31 days by checking for and removing accumulated water from the fuel oil storage tanks; I d. By sampling new fuel oil in accordance with ASTM-D4057 prior to ~ ~ ~ ~ ~~ ~~ addition to storage tanks and: 1) By verifying in accordance with the tests specified in ASTM-D975-81 prior to addition to the storage tanks that the sample has:
- These die" del generator starts from ambient conditiens shall be performed only once per. 184 days in these surveillance tests and all other engine starts for' *
.the purpose of this surveillance testing shall be preceded by an engine prelu(p., period and/or other warmup procedures recommended by the manufacturer so that the mechanical stress and wear on the diesel engine is minimized. CALLAWAY - UNIT 1 3/4 8-3
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 1 l a) An API Gravity of within 0.3 degrees at 60*F, or a specific gravity of within 0.0016 at 60/60*F, when compared to the supplier's certificate, or an absolute specific gravity at 60/60'F of greater than or equal to 0.83 but less than or equal to 0.89, or an API gravity of greater than or equal to 27 degrees but less than or equal to 39 degrees; b) A kinematic viscosity at 40*C of greater than or equal to i 1.9 centistokes, but less than or equal to 4.1 centistokes, _ if gravity was not determined by comparison with the supplier's certification; l c) A flash point equal to or greater than 125'F; and L d) A clear and bright appearance with proper color when tested in accordance with ASTM-D4176-82. 4 2) By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM-D975-81 are met . when tested in accordance with ASTM-0975-81 except that the analysis for sulfur may be performed in accordance with ASTM-D1552-79 or ASTM-D2622-82. e. At least once every 31 days by obtaining a sample of fuel oil in accordance with ASTM-D2276-78, and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTM-D2276-78, Method A; f. At least once per 18 months, during shutdown, by: 1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service; 1 2) Verifying the diesel generator capability to reject a load of greater than or equal to 1352 kW (ESW pump) while maintaining voltage at 4000 1-320 volts and frequency at 60 1 5.4 Hz; j 3) Verifying the diesel generator capability to reject a load of 1: 6201 kW without tripping. The generator voltage shall not _.- exceed 4784 volts during and following the load rejection; 4) Simulating a loss-of-offsite power by itself, and: a) Verifying deenergization of the emergency busses and load shedding from the emergency busses, and ~ b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 12 seconds, energizes the auto-connected " ~ ~ ~ ' shutdown loads through the shutdown sequencer and operates for greater than or eq'ual to 5 minutes while its generator CAI.LAWAY - UNIT 1 3/4 8-4
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) is loaded with the shutdown loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4000 320 volts and 60 1 1.2 Hz during this test. 5) Verifying that on a Safety Injection test signal without loss-of-offsite power, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes; and the offsite power source energizes the auto-connected emergency (accident) load through the LOCA sequencer. The generator voltage and frequency shall be 4000 320 volts and 60 1 1.2 Hz within 12 seconds after the auto-start signal; the generator steady-state generator voltage and frequency shall be maintained within these limits during this test; 6) Simulating a loss-of-offsite power in conjunction with a Safety Injection test signal, and a) Verifying deenergization of the emergency busses and load shedding from the emergency busses; b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 12 seconds, energizes the auto-connected emer-gency (accident) loads through the LOCA sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with emergency loads. After energiza-tion, the steady-state voltage and frequency of the emer-gency busses shall be maintained at 4000 320 volts and 60 1.2 Hz during this test; and c) Verifying that all automatic diesel generator trips, except high jacket coolant temperature, engine overspeed, low lube oil pressure, high crankcase pressure, start failure relay, and generator differential, are automatically bypassed upon loss of voltage on the emergency bus concurrent with a Safety Injection Actuation signal. 7) Verifying the diesel generator operates for at least 24 hours. During the first 2 hours of this test, the diesel generator shall be loaded to greater than or equal to 6821 kW and during the remaining 22 hours of this test, the diesel generator shall be loaded to greater than or equal to 6201 kW. The generator voltage and frequency shall be 4000 1 320 volts and 60 + 1.2,-3 Hz within 12 seconds after the start signal; the steady-state generator voltage and frequency shall be maintained within 4000 + 320 volts and 60 + 1.2 Hz during this test. Within 5 minutes after complet-ing this 24-hour test, perform Specification 4.8.1.1.2f.6)b)*;
- If Specification 4.8.1.1.2f.6)b) is not sat,isfactorily completed, it is not necessary to repeat the preceding 24-hour test.
Instead the diesel generator may be operated at 6201 kW for 1 hour or until operating temperature has stabilized. CALLAWAY - UNIT 1 3/4 8-5 -c-
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4 - 8) Verifying that the auto-connected loads to each diesel generator do not exceed 6635 kW; 9) Verifying the diesel generator's capability to: a) Synchronize with the offsite power source while the generator is. loaded with its emergency loads upon a simulated restora-tion of offsite power, b) Transfer its loads to the offsite power source, and c) Be restored to its standby status. 10) Verifying that with the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injection signal overrides the test mode by: (1) returning the diesel generator to standby operation, and (2) automatically energizing the emergency loads with offsite power; 11) Verifying tnat the fuel transfer pamp transfers fuel from each fuel storage tank to the day tank of each diesel via the installed cross-connection lines; and 12) Verifying that the automatic LOCA and shutdown sequence timer is OPERABLE with the interval between each load block within i 10% of its design interval. g. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting both diesel generators simultaneously, during shutdown, and verifying that both diesel generators accelerate to at least 514 rpm in less than or equal to 12 seconds; and h. At least once per 10 years by: l 1) Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite solution, and 2) Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME l Code at a test pressure equal to 110% of the system design ~ pressure. 4.8.1. 4 Reports - All diesel generator failures, valid or nonvalid, shall be repo ed in a Special Report to the Commission pursuant to Specifica-tion 6.9.2 within 30 days. Reports of diesel generator failures shall include. l the information recommended in Regulatory Position C.3.b of Regulatory i = l Guide 1.108, Revision 1, August 1977., If the number of failures in the last l 100 valid rests (on a per nuclear, unit basis) is greater than or equal to 7, j the rep, ort shall ue supplemented to include the additional information recommended in Regulatory Pos' tion C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. l CALLAWAY - UNIT 1 3/4 8-6 . m. -.W
e TABLE 4.8-1 7 6 DIESEL GENERATOR TEST SCHEDULE y NUMBER OF FAILURES IN LAST 100 VALID TESTS
- TEST FREQUENCY
<1 At least once per 31 days 2 At least once per 14 days .3 At least once per 7 days >4 At least once per 3 days + ~ e 3
- e
- (
- Criteria for determining number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, Revision 1, August 1977. where the last 100 tests are determined on a per
~ earJ0it basiff For the purpose of this schedule, only valid tests ~ ~ conducted after the completion of the preoperational test requirements of Regulatory Guide 1.108, Revision 1, August 1977, shall be included in the computation of the "Last 100 Valid Tests." ~ E e t v. 6 e CALLAWAY - UNIT 1 3/4 8-7 t e _,m....,_...__._.._____m_.
ELECTRICAL ~ POWER SYSTEMS SURVEILLANCE' REQUIREMENTS (Continued) b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 110 volts, or battery overcharge with battery terminal voltage above 150 volts, by verifying that: 1) The parameters in Table 4.8-2 meet the Category B limits, 2) There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10 s ohm, and 3). The' average electrolyte temperature of at least every sixth cell is above 60 F. c. At least once per 18 months by verifying that:
- 1) _ The' cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration,
- 2).The cell-to-cell and terminal connections are clean, tight, and coated with anticorrosion material, 3)
The resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10 s ohm, and
- 4),The battery charger will supply at least 300 amperes at 134 volts' for at least 1 hour..
d. At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual emergency loads for 200 minutes when the _ _ _ _. battery is subject to a battery service test; At least once per 60 months, during shutdown, by verifying that the ~ .e. battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. This performance discharge test may be. performed in lieu of the battery service test required by Specification 4.8.2.ld.; and Ace, pu (,o %A gyq(, f. At least once per 18 months during shutdown, by giving performance ^ discharge tests of battery capacity to any battery that shows signs' of degradation or has reached 85% of the service life expected for' ~ the' application. Degradation is indicated when the battery. capacity . drops more than 10% of rated capacity from its average on previous _ erformance tests, or is below 90% of the manufacturer's rating. p 3/4 8-10 CALLAWAY.- UNIT 1
9 TABLE 3.8-1 F l[ CONTAINMENT PENETRATION CONDUCTOR _c ,1 OVERCURRENT PROTECTIVE DEVICES E I,, "d BREAKER PROTECTIVE DEVICE TRIP RESPONSE TIME AT POWERED NUMBER AND LOCATION SETPOINT MAX. SHORT CIRCUIT EQUIPMENT (Amperes) (Sec/ Cycles) 13.8-kV SWITCHGEAR ~ Primary (P) 252PA0107 3600 (50)/ 0.1 Reactor Coolant Pump 372 (51) & 840 (51) DPBB01A P-25?PA0108 360 (50)/372 (51) 0.1 Reactor Coolant Pump j' 840 (51) DPBB01B T P-252PA0205 360 (50)/372(51) 0.1 Reactor Coolant Pump 40 (51) DPBB01C P-252PA0204 360 (50/372 (51) 0.1 Reactor Coolant Pump 40 (51) DPBB01D 480-V LOAD CENTER P-52NG0304 1200 (Inst.) 0.05 Hy;rogen Recombiner B-52NG0301 4320 (5.T.) 0.b SGS01A P-52NG404 1200 (Ir.st.) 0.05 Hydrogen Recombiner B-52NG0401 4320 (5.T.) 0.5 SG501B P-52PG2102 375 (Inst.) 0.025 Pressurizer Backup Through 52PG2112 Heater 8-350 A Fuse N.A. a. i f-y e
i i ~ 9 TABLE 3.8-1(Contibued) CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT PROTECTIVE DEVICES i c-5 BREAKER PROTECTIVE DEVICE TRIP., RESPONSE TIME AT POWERED ~ NUMBER AND LOCATION SETPOINT MAX. SHORT CIRCUIT EQUIPMENT (Amperes) (Sec/ Cycles) 480-V LOAD CENTER (Continued) P-52PG2202 375 (Inst.) 0.025 Pressurizer Backup Through 52PG2212 Heaters B-350 A Fuse N.A. P-52NG01TAF1 2000 (Inst.) 0.016 Containment Cool'r e R B-52NG0108 44,00-(S.T. 0.19 DSGN01A 2 oo T P-52NG03TAF 2000 (Inst.) 0.016 Containment Cooler g U B-52NG0305 g (S.T.) y 0.19 DSGN01C P-52NG02TAF1 2000 (Inst.) 0.016 Containment Cooler B-52NG0208 -2400- (S. T. ) 0.19 DSGN01B 21oo H P-52NG04TAF1 2000 (Inst.) 0.016 Containment Cooler B-52NG0405 N (S.T. ) 0.19 DSGN01D 270o 480 V MOTOR CONTROL CENTER P-52NG01BDF3 75 (Inst.) 0.016 RHR Loop Inlet Iso B-40A Fuse N.A. Viv EJHV8701B P-52NG02BHR2 10 (Inst.) 0.016 ESW from Ctmt Air B-ISA Fuse N.A. Coolers Iso Viv EFHV46 P-52NG02BDF2 45 (Inst.) .016 CCW to Ctmt Iso Viv B-40A Fuse N.A. EGHV60
D El TABLE 3.8-1 (Continued) P CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES E BREAKER ~* PROTECTIVE DEVICE TRIP RESPONSE TIME AT POWERED NUMBER AND LOCATION SETPOINT MAX. SHORT CIRCUIT EQUIPMENT (Amperes) (Sec/ Cycles) 480 V HOTOR CONTROL CENTER (Continued) P-52NG01BGF3 565 (Inst.) 0.016 Accum$latorIsoViv B-60A Fuse N.A. EPHV8808A P-52NG01BGF2 565 (Inst.) 0.016 Accumulator Iso Viv B-60A Fuse N.A. EPHV8808C R P-52NG010FF2 10 (Inst.) 0.016 Ctat Air to Aux Bldg i B-15A Fuse N.A. ESF Filter Iso Viv GSHV20 D React 'ldg Discharge - g;y ' P-52NG01BBR2 29 (Inst.) 0.016 B B-ISA Fuse N.A. Iso Vlve LFFV95 g;3, (()f P-52NG02BBF3 75 (Inst.) 0.016 RHR Loop Inlet B-40A Fuse N.A. Iso Viv BBPV87028 P-52NG02BCF2 75 (Inst.) 0.016 RHR Loop Inlet B-40A Fuse N.A. Iso Viv BBPV8702A P-52NG02BHF3 10 (Inst.) 0.016 ESW to Ctmt Air B-15A Fuse N.A. Coolers Iso Viv EFHV34 P-52NGG1BCF2 10 (Inst.) 0.016 ESW to Ctmt Air B-15A Fuse N.A. Coolers Iso Viv EFHV33 n P-52NG01B0F2 10 (Inst.) 0.016 ESW from Ctmt Air B-15A Fuse N.A. Coolers Iso Vlv EFHV45 e 5 e
9 TABLE 3.8-1-(Continued) r-CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES c '5 BREAKER PROTECTIVE DEVICE TRIP RESPONSE TIME AT POWERED NUMBER ANO LOCATION SETPOINT MAX. SHORT CIRCUIT EQUIPMENT ~ ~ (Amperes) (Sec/ Cycles) 'g s 489 V MOTOR CONTROL CENTER (Continued) . / / i ~r P-52NG01BEF2 75 (Inst.) 0.016 RHR Loop Inlet Iso Viv B-40A Fuse N.A. EJHV8701A 1 J l. P-52NG03CDF4 29 (Inst.) 0.016 JRCP' Thermal Bari f er B-15A Fuse N.A. CCW Iso Valwe BBHV13 w 'H
- Q
,, e, < # / P-52NG03CHF1 29 (I m t.'), 0.016 RCP Thermal Barrier
- 4 m
^ .,N.A. CCW_ Iso Viv BBHV14 B-15A Fuse 7,
- N
,/ ~ O.016 . React'or, Cavity Cooling l P,-52PG19NAF4,, ' 400 (Inst. ) 7 /' 'f i N.A. ' Fan ?CGNO2Af l ,B-100A F6se g y 1 'I260(hst'.N '0.016 CCat Atmosp'heric Control f P-52PG19M0F3 j' ~ h 'r g, ~5y'stei.. Fan DCGR01A _/, N.A. i B-60A Fusa) RCP'~k ' Space Heater-l /T-52PG19NGF2' ,675 yst.) 0.017 B-40 Fus'e N.A. 1 f., J P-52PG19NGF3' 67 (Ihst'.) 20.017' RCP B Space Heater ' ~, i B-40 Fuse / ti. A. - g 1 .c i P-52PG19NEF1 170.(Inst.) 0.016 RCP A,0il 'Lif t Pump B-40A Fuse, N.A. r P-52PG19NGR3 170 (Inst.) 0.016 RCP B Dil Lift Pump i-B-40A Fuse N.A. 4 l P-52PG19NFF1 22 (Inst.) - 0.016 Ctat Normal Sump B-15A Fuse N.A. Punp DPLF05A j .. ~ ~_ p 4* y e
y TA8LE 3.8-1 (Continued) ~ h -CONTAINMENT PENETRATION CONDUCTOR 4 ( -DVERCURRENT PROTECTIVE DEVICES 4 e3 BREAKER PROTECTIVE DEVICE
- TRIP RESPONSE TIME AT POWERED NUMBER AND LOCATION SETPOINT MAX. SHORT CIRCUIT EQUIPMENT j
(Amperes) (Sec/ Cycles) i j 480 V MOTOR CONTROL CENTER (Cont.inued) P-52PG19NFF2 22 (Inst.) 0.016 Ctat Normal Sump B-15A Fuse N.A. pump DPLF05C I P-52PG19NAF2 84 (Inst.) 0.016 Instrumnt Tunnel B-40A Fuse N.A. Sump Pump DPLF07A I P-52NG03CBF4 29 (Inst. 0.016 RCP Thermal Barrier CCW em I E B-15A Fuse N.A. Iso Viv BBHV15 jw l P-52NG03 2 29 (Inst.) 0.016 RCP Thermal Barrier l B-15A F N.f CCW Iso Viv BBHV16 i l P-52PG20NBFS 320 (Inst.) 0.016 Reactor Cavity Cooling B-100A Fuse N.A. Fan DCGN02B P-52PG20NFF4 260 (Inst.) 0.016 Ctat Atmospheric Control I B-60A Fuse N.A. System Fan OCGRO1B 44 l i P-52PG20NBF1 675 ( nst.) 0.017 RCP C Space Heater i 'D-40A Fuse N.A. i 1 s l P-52PG20NCF1 675 Inst.) 0.017 RCP D Space Heater i B-40A Fuse N.A. 1 l P-52PG20NFF3 170 (Inst.) 0.016 RCP C Oil Lift Pump l B-40A Fuse N.A. i j /
~, h$ TABLE 3.8-1 (Co'ntinued) CONTAINMENT PENETRATION CONDUCTOR 3 s '( e OVERCURRENT PROTECTIVE DEVICES C - 5 BREAKER PROTECTIVE DEVICE TRIP RESPONSE TIME AT'. POWEREO .t NUMBER AND LOCATION SETPOINT MAX. SHORT CIRCUIT EQUIPMENT (Amperes) (Sec/ Cycles) '4 480 V MOTOR CONTROL CENTER (Continued) P-52PG20NFF2 170 (Inst.) 0.016 RCP D 011 Lift Pump B-40A Fuse N.A. P-52PG20NER2 22 (Inst.)' O.016 Ctmt Normal Sump B-15A Fuse M.A. Pump DPLF05B 2 ss* P-52PG20NGF4 22 (Inst.) 0.016 .Ctmt Normal Sump [ B-15A Fuse , N.A. -Pump;DPLF050 ~ P-52PG20NDR2 84 (Inst.) 0.016 Instrument Tunnel B-40A Fuse _ N.A. Sump DPLF07B P-52PG1904 1440 (Inst.) b.03 Polar Crane B-600A Fuse N.A. IlKE13 CRDM CONTROL ROD DRIVE POWER N.A. Gripper Coils (106 fused P-10A Fuse N.A. ci,rcuits) B-30A Fuse N.A. Lift Coils (53 fused P-50A Fuse N.A. circuits) B-150A Fuse r (50) . Protective Relay Instantaneous Unit (51) - Protective Relay I,nverse Time Unit Inst. - Instantaneous Protection 5.T. - Short Time Protection ~
REFUELING OPERATIONS 3/4.9.12-SPENT FUEL ASSEMBLY STORAGE s LIMITING CONDITION FOR OPERATION 3.9.12 Spent fuel assemblies stored in Region 2 shall be subject to the ~ following conditions: a. The. combination of initial enrichment and cumulative exposure shall be within the acceptable domain of Figure 3.9-1, and b. No spent fuel assemblies shall be placed in Region 2, nor shall any storage location be changed in designation from being in Region 1 to being in Region 2, while refueling operations are in progress. APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel pool. ' ACTION: a. With the requirements of the above specification not satisfied, suspend all other movement of fuel assemblies and crane operations with loads in the fuel storage areas and move the non-complying fuel assemblies to Region 1. Until these requirements of the above 4.; specification are satisfied, boron concentration of the spent fuel pool shall be verified to be greater than or equal to 2000 ppm at least once per 8 hours. b:. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. i SURVEILLANCE REQUIREMENTS 4.9.12 The bur up of each spent fuel assembly stored in Region 2 shall be ascertained b$ analysis of its burnup histo , prior to storage in Region 2. A complete record of such analysis shall be kept for the time period that the spent fuel assembly remains in Region 2 of he spent fuel pool. ed inepmamR wnfid 3 CALLAWAY ', UNIT 1 3/4 9-15 ,y y
'l I i: i TABLE 4.11-1 (Continued) j TABLE NOTATIONS (Continued) o -(3)The principal gamma emitters for which the LLD specification applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, 2n-65, Ho-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that f only these nuclides are to be consiccred. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semianr:ual Radioactive Effluent Release Report pursuant to Specification 6.9.1.7, in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974. (4)A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative Prior to analysis, all samples taken for the of the liquids released. composite shall be thoroughly mixed in order for the composite samples to be representative of the effluent release. (5)A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release. (6) Samples shall be taken at the initiation of effluent flow and at least once To be represent.-- per 24 hours thereafter while the release is occurring.tive of the liqu [ The ratio of sample volume to effluent effluent stream discharge volume. discharge volume shall be maintained constant for all samples taken for the composite sample. e i CALLAWAY.- UNIT I 3/4 11-4
RADI0 ACTIVE EFFLUENTS V LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each of the following unprotected outdoor tanks shall be limited to less than or equal to 150 Curies, excluding tritium and dissolved or entrained noble gases:
- a.
Reactor Makeup Water Storage Tank, b. Refueling Water Storage Tank, c. Condensate Storage Tank, and d. Outside temporary tanks, excluding demineralizer vessels and the liner being used to solidify radioactive waste. l APPLICABILITY: At all times. ACTION: a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours reduce the tank contents to within the limit, and describe the events leading to n) this condition in the next Semiannual Radioactive Effluent Release y Report, pursuant to Specification 6.9.1.7. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a . representative sample of the tank's contents at least once per 7 days when radioactive materials are being added and within 7 days following any addition of radioactive material to the tank. l l CALLAWAY - UNIT 1 3/4 11-7 l
TABLE 4'.11-2 p. { F RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM N ,e MINIMUM LOWER LIMIT OF e SAMPLING ANALYSIS TYPE OF DETECTION (LLD)(1) g GASEOUS RELEASE TYPE FREQUENCY FREQUENCY ACTIVITY ANALYSIS (pCi/ml) Principal Gamma Emitters (2) 1x10~4 1. Waste Gas Decay Each Tank Each ank f Tank Grab Sample 2. ContaingtPurge Each PURGE (3) Ea h PURGE (3) Principal Gamma Emitters (2) 1x10-4 or Vent gample rab -6 M H-3 (oxide) 1x10 3. Unit Vent M(3),(4) M(3) Principal Gamma Emitters (2) ~4 1x10 Grab R Sample M( ) H-3 (oxide) 1x10-6 4. Spent Fuel Building 11(5) M Principal Gamma Emitters (2) 1x10~4 T Exhaust Grab Sample I) -6 M H-3 (oxide) 1x10 S. Radwaste Building M Principal Gamma Emitters (2) 1x10-4 Vent Grab Sample M 6. All Release Types Continuous (6) y(7) 1-131 1x10-12 as listed in 1., Charcoal -10 2., 3., 4., and Sample I-333 1x10
- 5. above Continuous (6) p(7)
Principal Gamma Emitters (2) 1x10.}1 ~ Particulate Sample Continuous (i) M Gross Alpha 1x10 6 -11 Composite Particulate Sample Continuous (6) Q Sr-89, Sr-90 1x10-11 Composite Particulate' Sample 4 9
k TdBLE4.11-2(Continued) TABLE NOTATIONS (Continued) y ~(2)The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, , Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141, and Ce-144 in iodine and particulate releases. This list does not mean that only these nuclides are to be considered. Other gamma peaks that.are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.7, in the format outlined in Regulatory Guide 1.21,
- f.
. Appendix B, Revision-1, June 1974. (3) Sampling.and analysis shall also be performed following shutdown, startup, a or a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within 1 hour period. (4) Tritium grab samples shall be takep an lyzed)t least once per 24 hours when the refueling canal is flooded.
- 15) Tritium grab samples shall be taker (and analy' zed)tc.least once per 7 days sce-from the ventilation exhaust fro th~e spent fuel pnn1'araa. whenever spent fuel is in the spent fuel poo Grab samples need to'be taken only wnen
.(-- spent fuel is.1n the sonnt fuel pool. j (6 he ratio of the sample flow rate to the sampled. stream flow rate shall be known for the time period covered by each dose or dose rate calculation ~ made in accordece with Specifications 3.11.2.1, 3.11.2.2, and 3.11.2.3. (7) Samples shall be changed at least once per 7 days and analyses shall be completed within'48 hours after changing, or after removal from sampler. Sampling shall also be performed at least 'once per 24 hours for at least 7 days following each shutdown, STARTUP or THERMAL POWER change exceeding 15% of RATED THERM'AL POWER within a 1-hour period and analyses shall be com-pleted within 48 hours of changing. When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement does not apply if: (1) analysis shows that the DOSE ~ . EQUIVALENT I-131 concentration in the reactor coolant has not increased more than a factor of 3, and (2) the noble gas monitor shows that effluent (I activity has not increased more than a factor of 3. { ^ ~ ~ - - - - ~ ~ - ~ _ 1 CALLAWAY , UNIT 1 3/4 11-11
RADI0 ACTIVE EFFLUENTS ( GAS STORAGE TANKS s.. Y LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to 2.5 x 10s Curies of nob?e gases (considered as Xe-133 equivalent). APPLICABILITY: At all times. ACTION: With the quantity of radioactive material in any gas storage tank a. exceeding the above limit, inmediately suspend all additions of radioactive material to the tank and, within 48 hours, reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.7. ~ b. The provisions of Specifications,3.0.3 and 3.0.4 are not applicable. . i,,, - SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas sto e day t.n_k_shall be determined to be within the above limit at least once per when radioactive materials are being added ana within 7 cays foisowi6g any, addition of radioactive material to the tank. E 0 l:!' h ,,t lC. lI* -I t l CALLAWAY'- U!!IT 1 3/4 11-16 .~ e ..n ..4
c- ..o ..........-...~.- ll., A
- TABLE 3.12-1 (Con?.inued) 9 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM r-T NUMBER OF
~ f REPRESENTATIVE EXP0'SURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY N) COLLECTION FREQUENCY OF ANALYSIS AND/OR SAMPLE SAMPLE LOCATIONS i'i
- 2. Airborne Radioiodine and Samples from five loc'ations:
' Continuous sampler Radioiodine Cannister: -e Particulates operation with sample I-131 analysis weekly. w Three samples from close to the collection weekly, or three SITE BOUNDARY locations, more frequently if i in different sectors, of the required by dust Particulate Sampler: f j highest calculated annual average loading. Gross beta radioactivity ground level D/Q. analysisfollogg filter change; and l gamma isotopic analysis (5) One sample from the vicinity l of a community having the highest of composite (by R' calculated annual average ground-location) quarterly. level D/Q. w T One sample from a control location, as for example 15 to 30 km distant and in direction.gvalentwind the least
- 3. Waterborne Gamma isotopic analys s(5)
- a. Surface (0)
One sample upstream. [Compositesamp ver ~.', 'l-month period monthly. Composite for One sample downstream. tritium analysis of composite sam;il'd.(by s location) quarterly. l1 I-131 analysis on each
- b. Drinking One sample of each of one to Compos.ite sample g composite when the dose three of the nearest water over 2-week period j
j supplies within 10 miles when I-131 analysis calculated for the consump-downstream that could be is performed, monthly tion of the water is g ater .I affected by its discharge. composite otherwise. than 1 mrem per year. "Com- 'ositeforgrossbetaag p One sample from a control ; gamma isotopic analyses f monthly. Composite for location. ." j tritium analysis qua *1rly.
[ TABLE 3.12-1 (Continued) TABLE NOTATIONS (Continued) y (4) Airborne particulate sample filters shall be analyzed for" gross beta - radioactivity 24 hours or more after sampling to allow for radon and
- thoron daughter decay.
If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples. (5) Gamma isotopic analysis means the. identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents . from the facility. i 4. (6) The'" upstream sample" shall be taken at a distance beyond significant influence of the discharge. The " downstream" sample shall be taken in an area beyond but near the mixing zone. (7) In this program composite sample aliquots shall be collected at time, intervals that are very short (e.g., hourly) relative to the compositing i period (e.g., monthly) in order to assure obtaining a representative. (,, sample. , ;,..v... 2 (8) Groundwater samples shall be taken when this source i's tapped for drinking ,~ or, irrigation purposes in areas where the hydraulic gradient or recharge ,...(- properties are suitable for contamination. L' (9) The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM. (10')' IV harvest occurs more than once a year, sampling shall be performed If harve.t occurs continuously, sampling during each discrete harvest. c shall be monthly. Attention shall be paid, to including samples of tuberous and root food products. e J- -s e [t e 4 l;
- ]
CALLAWAY.- UNIT 1 3/4 12-8 V.
DRIH RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION i 3.12.3 Analyses shall be performed on radioactive materials suppl ^ed as part of an Interlaboratory Comparison Program that has been approved by the Commission. APPLICABILITY: At all times. ACTION: a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pu'rsuant to Specification 6.9.1.6. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REOUIREMENTS 4.12.3 The Interlaboratory Comparison Program shall be described in the 00CM. ' A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiologics1 Environmental Operating Report pursuant to Specification 6.9.1.6. CALLAWAY - UNIT 1 3/4 12-14
POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR E'NTHALPY RISE HOT CHANNEL FACTOR (Continued) 3.~ The control rod insertion limits of Specification 3.1.3.6 are maintained; and-4. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits. F will be maintained within its limits provided Conditions 1; through 1 H
- 4. above are maintained.
As noted on Figure 3.2-3, RCS flow rate and F may AH 4, be " traded off" against one another (i.e., a low measured RCS flow rate is acceptable if the measured F is also low) to ensure that the calculated DNBR H g will not be below the design DNBR value. TherelaxationofFhasafunction of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. R as calculated in Specification 3.2.3 and used.in Figure 3.2-3, accounts fof F less than or equal to 1.49. This,value is 'used..in.the various accident x,,. H analyses where F influences parameters o'ther than DNBR,-le'.g., peak clad g . temperature, and thus is the maximum "as measured" value allowed.- ('; Fuel rod bowing reduces the value of DNB ratio. Credit is available to ffset this reduction in the generic margin. The generic design margins, ' totaling 9.1% DNBR, completely offset any rod bow penalties. This margin includes the following: i 1) Design. limit DNBR of 1.30 vs. 1.28, 2) Grid Spacing (K ) f 0.046 vs. 0.059, s ~ l 3) Thermal Diffusion Coefficient of 0.038 vs. 0.059, 4) ' DNBR Multipler of 0.86 vs. 0.88, and 5) Pitch reduction. The applicable values of rod bow penalties are referenced in the FSAR. l When nF measurement is taken, an allowance for both experimental error q and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance. ~ 6 t s CALLAWAY - UNIT 1 ~B 3/4 2-4 r
~ 3 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is ba' sed on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveil-lance.of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can b'e taken. DUnscheduled inservice inspections are performed on each steam generator following: (1) primary to secondary tube leaks; (2) a seismic occurrence greater than the Operating Basis. Earthquake; and (3) a loss-of-coolant accident requiring actuation of the Engineered Safety Features, which for this Specification is defined to be a break greater-than that equivalent to the-severance of a 1" inside diameter pipe, or, for a main steamline or feedline, ~ :- a break greater than that equivalent to a steam generator safety valve failing open; to ensure that steam generator tubes retairisufficient integrity for contin'ued operation. Transients less severe than'these do not require inspections because the resulting stresses are well within the stress criteria I 3 l established by Regulatory Guide 1.121, which unplugged steam generator tubes ~ l must be capable of withstanding. . The plant is expected to be operated in a manner such that the secondary -coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may l likely result in stress corrosion cracking. The extent of cracking _during i plant operation would be limited by the limitation of steam generator tube l 1eakage between the' Reactor Coolant System and the Secondary Coolant System .(reactor-to secondary leakage = 500 gallons per day per steam generator). Cracks having a reactor-to-secondary leakage les's than this limit during I ' operation will have an adequate margin of safety to withstand the loads imposed l during normal. operation and by postulated accidents. Operating plants have [ demonstrated that reactor-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator i i blowdown. Leakage in excess of this limit will require plant shutdown and an i-unscheduled inspection, during which the leaking tubes will be located mod plugged. [ Wastage-type. defects are unlikely with proper chemistry treatment of the secondary coolant. -However, even if a defect should develop in service, it L will be found during scheduled inservice steam generator tube examinations. 3-Plugging will be re ired for all tubes with imperfections exceeding the i. plugging limit of of the tube nominal wall thickness. Steam generator ~ ) .CALLAWAY - UNIT 1 ~ B 3/4 4-3 A
DRlH 3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures. The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met. The accumulator power operated isolation valves are considered to be " operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed autc atically whenever permissive conditions are not met. In addition, as these accumulator isolation valves fai.1 to meet single failure criteria, removal of power to the valves is required. The limits for operation wi.th an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak' cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a MODE where this capability is not required. The requirement to verify accumulator isolation valves shut with power removed from the valve operator when the pressurizer is solid ensures the accumulators will not inject water and cause a pressure transient when the l Reactor Coolant System is on solid plant pressure control. l sis.s.4 3/4.5.2 and 3/4. 5.3 ^ ECCS SUBSYSTEMS f l The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period. With the RCS temperature belcw 350 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity con?"on of the reactor and the limited core cooling requirements. CALLAWAY - UNIT 1 B 3/4 5-1 --+-m -m- .e*.m,.. e e'*
CONTAINMENT SYSTEMS }. BASES 3/4 6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 3.0 psig, and (2) the contain7ent peak pressure does not exceed the design pressure of 60 psig during steam line break conditions. The maximum peak pressure expecrea to be obtained from a steam line break , event is 48 psig. The limit of 1.5 psig for initial positive contai,nment pressure will limit the total pressure to 49.5 psig, which is less than design pressure and is consistent with the safety analyses. 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial temperature condition assumed in the safety analysis for a steam line break accident. Measurements shall be made at all listed locations, whether by fixed or portable instruments, prior to determining the average air temperature. 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY ^ This limitation ensures that the structural integrity of the containment vessel will be maintained in accordance with safety analysis requirements for the life of the facility. Structural integrity is required to ensure that the l containment will withstand the maximum pressure of 50 psig in the event of a steam line break accident. The measurement of containment tendon lift-off force, the tensile tests of the tendon wires or strands, the visual examination of tendons, anchorages and exposed interior and exterior surfaces of the --containment, and the Type A leakage test are sufficient to demonstrate this capability. The Surveillance Requirements for demonstrating the containment's structural integrity are in compliance with the recammendations of Regulatory Guide 1.35, -" Inservice Surveillance of Ungrouted Tendnns in Prestressed Concrete Containment Structures," January 1976, and propose'd Regulatory Guide 1.35.1, " Determining Prestressing Forces for Inspection of Prestressed Concrete Containments," ~ April 1979. The required Special Reports from any engineering evaluation of containmen7t abnormaltiet shall include a description of the tendon condition, the conditioni a of the concrete (especially at tendon anchorages), the inspection procedure, ~~ the tolerances on cracking,- the resufts of the engineering evaluation and the 3 corrective actions taken. CALLAWAY - UNIT 1 B 3/4 6-2 k - ~.. .. r.- ,,-,_m__ .m___..,__.,-_,,_,~.
CONTAINMENT SYSTEMS ( BASES-3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM j The 36-inch containment purge supply and exhaust isolation valves are required to,be closed and blank flanged during plant operations since these valves have not been demonstrated capable,of closing during a LOCA or steam line break accident. Maintaining these valves closed and blank flanged during plant operation ensures that excessive quantities of radioactive material will not be released via the Containment Purge System. To provide assurance that the 36-inch containment valves cannot be inadvertently opened, the valves are. blank flanged. The use of. the containment mini purge lines is restricted to the 18-inch f purge supply and exhaust isolation valves since, unlike the 36-inch valves, the 18-inch valves are capable of closing during a LOCA or steam line break accident. 4 Therefore,_ the SITE BOUNDARY dose guideline values of 10 CFR Part 100 would not be exceeded in the event of a cident during containment purging 6peration. Operation will be limited t 2000 hours during a calendar year. The total time -the Containment Purge (vent) m isolation values may be open during MODES 1, 2, 3, and 4 in a calendar year is a function of anticipated need and opera.ing t experience. Only safety-related reasons; e.g., containment pressure control or the reduction of airborne radioactivity to facilitate personnel access for. surveillance and tenance activities, should be used to support additional tim'e reauests. Only safety-related reasons should De.use.d to justify'the opening of these isolation valves during MODES 1, 2,. 3,-and 4 in any calendar year regardless of the. allowable hours. Leakage integrity tests with a maximum allowable leakage rate for containment purge supply and exhaust supply valves will provide early indica-tion of resilient material seal degradation and will allow opportunity for repair before gross leakage failures could develop. The 0.60 L leakage limit ofSpecification3.6.1.2b.shallnotbeexceededwhenthe.leaka$eratesdeter. mined by the. leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests. I 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray System ensures that containment L depressurization and cooling capability will be available in the event of a LOCA or steam line break. The pressure reduction and resultant lower contain-ment leakage rate are consistent with the assumptions used in the safety l. analyses. The Containment Spray System and the Containment Cooling System are redundant to each other in providing post-accident cooling of the Containment However, the Containment Spray System also provides a mechanism atmosphere. for removing iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable spray system to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment. 3/4.6.2.2 SPRAY ADD'ITIVE SYSTEM The OPERABILITY of the Spray Additive' System ensures that' sufficient.NaOH ( is added.to the containment spray in +.he event of a LOCA. The limits on NaOH ~ volume and concentration ensure a ph salue of between 8.5 and 11.0 for the L CALLAWAY - UNIT 1 B 3/4 6-3 r L
e-l CONTAINMENT SYSTEMS BASES _e.. ~ SPRAY ADDITIVE SYSTEM (Continued) solution ~ recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained solution volume limit includes an allowance for solution not usable because of tank discharge line location or other physical characteristics. The eductor flow test of 52 gpm with RWST water is equivalent to 40 gpm NaOH solution. These assumptions are consistent with the iodine removal efficiency assumed in the safety analyses. 3/4.6.2.3 CONTAINMENT COOLING SYSTp 1 The OPERABILITY of the Containment Cooling System ensures that: (1) the containment air temperature will be maintained within limits during' normal operation, and (2) adequate heat removal capacity is available when operated in conjunction with the Containment Spray Systems during post-LOCA conditions. ~The Containment Cooling System and the Containment Spray System are redundant to each othar in providing post-accident cooling of the Containment atmosphere. As a result of this redundancy in cooling capability, the allowable .out-of-service time requirements for the Containment Cooling System have been appropriately adjusted. However, the allowable out-of-service time require-ments for the Containment Spray System have'been maintained consistent with that assigned other inoperable ESF equipment since the Containment Spray System also provides a mechanism for removing iodine from the containment atmosphere. i 3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 thru 57 of Appendix A to 10 CFR Part 50. Containment isolation within c the time limits specified for those isolation valves designed to close auto-matica11y ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA. 3/4.6.4 COMBUSTIBLE GAS CONTROL l The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to L* ~ maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. Either recombiner unit (or the Purge System) is capable of controlling the expected hydrogen generation associated with: (1) zirconium-water reactions, (2) radiolytic decomposition of water, and (3) corros3on of metals within containment. The Hydrogen Pur'ge Subsystem discharges directly to the Emergency Exhaust System. Operation of the Emergency - ExhaustcSystem with the heaters operating for at least 10 continuous hours in a-31-day period is sufficient to reduce the buildup of moisture on the adsorbers *
- and HEPA filters. These hydrogen control-systems are consistent with the
-recommendations of Regulatory Guide 1.7, "cnn+ val af Combustible Gas _ Concentrations in Containment FollowingMoss-of-Coolant Accident," Revision 2] (Iovember1978 i,- CAL'LAWAY - UNIT 1- ' B 3/4 G-4
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program given in the ODCM are made if required by the results of this census..The best information from the door-to-door survey, from aerial survey, or from consulting with local agricul-tural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to.10 CFR Part 50. Restricting the census to gar-dens of greater than 50 m provides assurance that significant exposure pathways 2 via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: (1) 20% of the garden was used for growing broad leaf vegetation (i.e., similar 2 to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m, 3/4.12/.'3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accu-racy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environ-mental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50. CALLAWAY - UNIT 1 B 3/4 12-2 3- - -3 --.. ~..-, --.--
DRFT 5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The Exclusion Area shall be as shown in Figure 5.1-1. LOW POPULATION ZONE 5.1.2 The Low Population Zone shall be as shown in Figure 5.1-2. MAPS DEFINING UNRESTRICTED AREAS AND SITE 90UNDARY FOR RADIOACTIVE GASEOUS ANO LIQUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as defini-tion of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be as shown in Figures 5.1-3 and 5.1-4. The definition of UNRESTRICTED AREA used in implementing the Radiological Effluent Technical Specifications has been expanded over that in 10 CFR 20.3(a)(17). The UNRESTRICTED AREA boundary may coincide with the Exclusion (fenced) Area boundary, as defined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas over water bodies. The concept of UNRESTRICTED AREAS, established atorbeyondtheSITEBOUNDARY,isutilizedintheLj$(({ggC$$p{T(QhpFgg OM$((j)$tokeeplevelsofradioactivematerialsinliquidandgaseous efflQenus as low as is reasonably achievable, pursuant to 10 CFR 50.36a. 5.2 CONTAINMENT CONFIGURATION
- 5. 2.1-The containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:
a. Nominal inside diameter = 140 feet. b. Nominal inside height = 205 feet. c. Nominal thickness of concrete walls = 4 feet. d. Nominal thickness of concrete dome = 3 feet. 4 e. Nominal thickness of concrete base slab = 10 feet, f. Nominal thickness of steel liner = 0.25 inch. g. Net-free volume = 2.5 x 108 cubic feet. DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment building is designed and shall be maintained for a maximum internal pressure of 60 psig and a temperature of 320 F. CALLAWAY - UNIT 1. 5-1 2 m =- - -..
ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 ~The Manager, Callaway Plant, shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. 6.1.2 The Shift Supervisce (or during his absence from the control room, a designated individual) shall be responsible for the control room. command function. A management directive to this effect, signed by the Vice President-Nuclear shall be reissued to all station personnel on an annual basis.
- 6. 2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for unit management and technical support shall be as shown in Figure 6.2-1.
UNIT STAFF 6.2.2 The Unit organization shal.1 be as shown in Figure 6.2-2 and: Each on duty shift shall be composed of at l' east the minimum shift a. crew composition shown in Table 6.2-1; b. At least one licensed Operator shall be in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, 3 or 4, at least one licensed Senior Operator shall be in the control room; c. An individual from the Health Physics organization #, qualified in radiation protection procedures, shall be on site when fuel is in the reactor; d. All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation;- A site Fire Brigade of at least five members # shall be ma'ntained e. onsite at all times. The Fire Brigade shall not include the Shift l Supervisor, and the ther members of the minicum shift crew necessary for safe hutdown cf the unit and any personnel required for other essential functions during a fire emergency; and {.x
- May be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence provided immediate action is taken to fill the required positions.
CALLAWAY - UNIT 1 6-1
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s I r b MANAGER 9JPERINTENDENT PERSONNEL DEVELOPMENT '1 ASST. MANAGER ASST OPERATIONS B TE MAINTENANCE SF I I I I I I I SUPT.* SUPT." SUPERVISOR 3UPERINTENDEN$ SUPT.* SUPT. OPERATIONS IaC MAINTENANCE STARTUP ENGINEERING EDU NG ASST. SUPT. OPERATIONS c; 9l OPERATING T SUPEfMSORS-Rb FC UNIT. REACTOR . OPERATORS EQUIPMENT OPERATORS RA TECl
- DEPARTMENT HEAD t
i 1 UNI
f.. P ADVISOR TO MANAGER _________J_. l l ASST. MANAGER OUALITY l ASSURANCE I I I OA STAFF MANAGER l ^b* ^^ SEE AS ST. MANAGER FIG. G.2-1 l MATERIALS SERVICES RVICES I I I i I I I I I PT SUPV. SUPT. SUPT. SUPT. SUPT. HE RADWASTE CHEMISTRY TRAINING COMPLIANCE SECURITY PHYSICS ~ l IUPV. SUPV. Q/C
- ALTH HEALTH lYSICS PHYSICS-SUPV.
- HNICAL OPERATION D/ CHEM RAD / CHEM O/C REMAN FOREMAN STAFF
-u ~. AMRTJ RE .h D/ CHEM RAD / CHEM 18 ASST. TECHSASST. Also Available On Aperture Card FIGURE 6.2-2 T ORGANIZATION 8 40 412 0 3 3 0 -bl
ADMINISTRATIVE CONTROLS c 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG) T + FUNCTION G.2.3.1 The ISEG.shall function to examine plant operating characteristics, NRCissuances,industryadvisories,QEPORTABLEEVENTS)andothersources of plant design and operating cxperience information, including plants of similar design, which may indicate areas for improving plant safety. The ISEG shall make detailed recommendations for revised procedures, equipment modifica- ~ tions, maintenance activities, operations activities or other means of improving plant safety to the Manager, Nuclear Safety and Emergency Preparedness and the Manager, Callaway Plant. COMPOSITION 6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time engineers located on site. Each shall have a bachelor's degree in engineering i or related science and at least 2 years professional level experience in his field. RESPONSIBILITIES The ISEG shall be responsible fo maintaining:surfeillance of plant 6.2.3.3 r activities to provide independent verification
- that thess activities are j...(.
performed correctly and that human errors are reduced as much as practical. i ' RECORDS' ~6.2.3.,4 Records of activities performed by the ISEG shall be prepared, maintained, and forwarded each calendar month to the Manager, Nuclear Safety and Emergency Preparedness and the Manager, Callaway Plant. 6.2.4 SHIFT TECHNICAL ADVISOR l The Shift Technical. Advisor (STA)** shall provide technical support to the Shift . Supervisor in the areas of thermal hydraulics, reactor engineering and plant i .~ analysis with regard to the safe operation of the unit. i
- 6. 3 UNIT STAFF QUALIFICATIONS l
6.3.1 Each menber of the unit staff shall meet or exceed the minimum L qualifications of ANSI /ANS 3.1-1978, except for the Superintendent, Health ~, Physics, who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, for a Radiation Protection Manager. The licensed Operators and Senior Operators shall also meet or exceed the minimum qualifications l ~of the supplemental requirements specified in Sections A and C of Enclosure 1 l l of the March 28, 1980 NRC letter to all licensees. l
- Not responsible for sign-off function.
- The STA position shall be manned in 110 DES 1, 2, 3 and 4 unless the Shift Supervisor or the individual with a Senior Operator license meets-the quali-L L
fications for the STA as required by the NRC.; CALLAWAY,, UNIT 1 6-6 t' .*---e-a,- r ,,,e w g ,.7 m ,__,,,7 g_
ADMINISTRATIVE CONTROLS ) -[ RESPONSIBILITIES (Continued) Review of Unit operations to detect potential hazards to nuclear m. safety; Investigations or analysis of special subjects as requested by the n. Chairman of the,NSRB; and Review of Unit Turbine Overspeed Protection Reliability Program and o. revisions thereto. 6.5.1.7 The ORC shall: Recommend in writing to the Manager, Callaway Plant approval or a. disapproval of items considered under Specifications 6.5.1.6a. ,g through e., i., j., k., 1., and o. above; b. Render determinations in writing with regard to whether or not each ~ item considered under Specifications 6.5.1.6b. through e., and m., above, constitutes an unreviewed safety question; and c. Provide written notification within fa hours to the Vice President-Nuclear'and the Nuclear Safety Review Board of disagreement betw' en e the ORC and the Manager, Callaway Plant; however, the Manager, Callaway Plant shall have responsibility for. resolution of such .disagreenents pursuant to Specification 6.1.1 above. i -.? ' RECORDS 6.5.1.8 The ORC shall maintain written minutes of each ORC meeting that, - at a minimum document the results of all ORC activities performed under the .responsi i y prov sions of these Technical Specifications. Copies shall be provide th e President-Nuclear and the Nuclear Safety Review Board. 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB) FUNCTION L .6.5.2.1 The NSRB shall function to provide independent review and audit of J designated activities in the areas of: i. a. Nuclear power plant operations, b. Nuclear engineering, c. Chemistry and radiochemistry, [ d. Metallurgy, i l:, e. Instrumentation and control, 1-f. Radiological-safety, g. Mechanical. and electrical engineering, and I h. Qtality assurance practices. The NSRB shall report to as.4 advise the Vice President-Nuclear on those areas of responsibility stated in Specifications 6.5.2.8 and 6.5.2.9. CALLAWAY - UNIT 1 6-9 lt L .m.-, 4o m
~ ADMINISTRATIVE CONTROLS 6.5.3 TECHNICAL REVIEW AND CONTROL ACTIVITIES e 6.5.3.1 Activities which affect nuclear safety shall be conducted as follows: Procedures required by Specification 6.8 and other procedures which a. affect plant nuclear safety, and changes :. hereto, shall be prepared, reviewed and approved. Each such procedure or procedure change shall be reviewed by a qualified individual / group other than the -individual / group _which prepared the procedure or procedure change, but who may be from the same organization as the individual / group .d. which prepared the procedure or procedure-change. Procedures other than Administrative Procedures shall be approved by the appropriate ~ i Department Head as designated in writing by the Manager, Callaway ~ Plant. The Manager, Callaway Plant, shall approve Administrative Procedures, Security Plan implementing proce.dures and Radiological Emergency Response Plan implementing procedures. Temporary changes to procedures which do not change the intent of the approved proce-dures shall be approved for implementation by two members of the plant staff, at least one of whom holds a Senior Operator license, ~ and documented.. The temporary ch= shal1 be approved by the -original approval authority ithin 14 d ys of implementation. _For changes to pr.ocedures which ma e a change in intent of.the 1 ,~ a' proved procedures, the person authorized above to approve the ' (. p procedure shall approve the change prior to implementation; ^ b. Proposed changes or modifications to plant nuclear safety-related structures, systems and co'mponents shall be reviewed as designated by the Manager, Callaway Plant. Each such modification shall be reviewed by a qualified individual / group other than the individual / -group which-designed the modification, but who may be from the same organization as the individual / group which designed the modifica-tions. Proposed modifications to plant nuclear safety-reiated structures, systems and components shall be approved prior to implementation by the Manager, Callaway Plant; Proposed tests and experiments _which affect plant nuclear safety and c. are'not addressed in the. Final Safety Analysis Report or Technical Specifications shall be prepared, reviewed,'and approved. Each such. test or experiment shall be reviewed by a qualified individual / group .othe'r than the individual / group which prepared the proposed test or experiment. Proposed test and experiments shall be approved ~ before implementation by the Manager, Callasay Plant; 1 CALLAWAY ' UNIT 1 6-13 7--.
~ t..
- ADMINISTRATIVE CONTROL 5
^ (t t (, SAFETY LIMIT VI'OLA TON, (Cont [inued) p The Safety Liniit Violation Report shall be submitted to the c. Commission, tt9 NSRB and the Vice President-Nuclear within 14 days of the violation; and Critical operation of the unit shall not be resumed until authorized d. by the Commission. s 6.'8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained-covering the activities referenced below: The applicable procedures recommended in Appendix d, of' Regulatory a. Guide 1.33, Revision 2, February 1978; ( b. The emergency operating procedures required to implement the ~ to NUREG-0737 as stated . requirements nf NUREG-0737 and Supplemen QnSection7.1ofGenericLeLLerNo.82, Plant Security Plan implementation; c. d. Radiological Emergency Response Plan. implementation; PROCESS CONTROL PROGRAM implementation, e. f, 0FFSITE DOSE. CALCULATION ANUA mplementation, 1 s Quality Assurance Progra ffor effluent and environmental m,onitoring, g. s _h. TurbineOverspeedProtectionReliabilityProgram.) ~ and 16.8.2 Each procedure and administrative policy of Specification'6.8.1 above, and changes thereto, including temporary changes shall be reviewed prior to implementation as set forth in Specification 6.5 above x The plant Adininistrative Procedures and changes-thereto shall be reviewed 6.8.3 in accordance with Specification 6.5.1.6 and approved in accordance with The associated icplementing procedures and c.hanges Specification 6.5.3'.1. thereto shall be reviewed and approved in accordance with Specification 6.5.3.1. '6.8.4 The following programs shall be established, implemented, 'and maintained: a cto olant Sources Outside Containment .A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the recirculation portion of the Containment Spray System, Safety Injection System, Chemical and Volume Control System, and RHR System. The program shall include the following: 1) Preventive maintenance and periodic visual insp.ection /. requirements, and + CALLAWAY - UNIT 1 6-15 ~ e-ew y-s--- p -..w-, ,c.. --9 ,,,.,,,.,.p,.,,,, ,_,7,_. --c
- ar-g-weetr-g gvpp ww n
g, t ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 2) Integrated leak test requirements for each system at refueling cycle intervals or less. b. In-Plant Radiation Monitorino A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following: 11) Training of personnal, 2) Procedures for monitoring, and .3) Provisions for maintenance of sampling and analysis equipment. c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube' degradation. This program shall include: ~ 1) Identification ~of a sampling schedule for the critical variables and contro1' points for these variables, g. wh 2) Identification of the procedures used to measure the values of ^ the critical variables, 3) Identification _of psocess sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser ~in-leakage,
- 4) ~ Procedures for the recording and management of data,
-5) Procedures defining corrective action for all off-control poi 7t chemistry conditions, and 6) A proceddre identifying: (a) the authority responsible for the interpreta"on of the data, and (b) the sequence and timing of administrative events required to initiate corrective action. l d. Post-accident Samoling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant l-gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the following: 1) Training of personnel, g 2) Procedures for sampling and analysis, and l 3) Provisions for maintenance of sampling and analysis equipment. e. Turbine, Overs,yd Wetection Ren;2b'.l.*L Pe.m a CALLAWAY - UNIT 1 6-16 (sec, fo%em3 page) ~ L __
e. Turbine Overspeed Protection Reliability Program A progran'to increase the assurance that the turbine overspeed -protection ' system functions, if challenged,and to assure structural integrity of turbine components which could result in missile generation in the event of an actual overspeed occurrence. The program shall include the following:
- 1) Periodic testing and inspection requirements,
- 2) Specification of test and inspection intervals, and
- 3) Administrative restrictions and procedural guidance for program implenentation such as: record keeping; reporting, evaluation and disposition of discrepancies; review and approval of revisions to the program; and authorization (s) required to deviate from the program guidelines.
I O 9 I 6-16a
ORAFT ADMINISTRATIVE CONTROLS ~ SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT
- 6.9.1.7 Routine Semiannual Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year.
The period of the first report shall begin with the date of initial criticality. The Semiannual Radioactive Effluent-Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, i " Measuring, Evaluating,' and Reporting Radioactfvity in Solid Wastes and ,l Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. For solid wastes, the fonsat for Table 3 in Appendix B shall be supplemented with three additional categories: class of solid waste (as defined by 10 CFR Part 60), type of co~ntainer (e.g., LSA, Type A, Type B, Large Quantity), and SOLIDICATION agent or absorbent (e.g., cement, urea formaldehyde). The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability.** This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive i liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figures 5.1-3 and 5.1-4) durina the renort oeriod kskrecal All assumptions used in making these assessments, i.e.,^ specific activity, ~ 4wa exposure time and location, shall be included in these reports. The meteoro ag.yj 8 logical conditions concurrent with the time of release of radioactive materials in gaseous affluents,.as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (00CM). The Semiannual Radioactive Effluent Release Report to be submitted within 60 days.after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from Reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power
- A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall
.specify the releases of radioactive material from each unit.
- In lieu of s9bmission with the Semiannual Radioactive Effluent Release Report, the licensee ~has tne option of retaining this summary of required meteorological data on site in a tile that shall be provided to the NRC upon request.
CALLAWAY - UNIT 1 6-19 {
i 4 3 3 i 5 4 t 5 k <1 l ATTACIIMENT 3_ Specifications Which are Being Appealed ? l I w. A I 9 W-- _,,_,y -r we.
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A',........ -W. y!:. ':fi %- .7^....' d]-l.. '%..> l a. '.. &.. *,., 'i M,.;lugg l.:L. g nyt .~ s t. - ...c. . hhk q., i;;. ' '? ' REACTOR CChi. thT ShSTEM ' $1. L ., L,.. I s. ,t..d..., ;...4-;^.0VERPRESSURE PROTECTION SYSTEMS ? Ny .*,.s..,.,'..>. .. :. ~ ; r......w. t. .+ - e-u -p - m y. w - y:..n.-.. c.. ::....,..:.;.:..>,. 'r. d.,, '. %:;*a.! *.. J ..e. 7.:. LIMITING CONDITION FOR OPERATION ~ ~ ....n... gLU M.. ! :-:M... b.Kf, -.a.*.~.: ' a.. _. ..,. v. +:-:W.n.., f.:s. 4
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s 7 .z ..h..IN M.k;kM NkM $M swh feNek urdons eac.k teid[.[.. f/ w.7 =,.y.ai.h.:,%: ", ..; :.: i wu C &...%.m::&t 12..o,&..,d cQ:fGoa%. l ef, n t or. a-T ..... w: m.y ' ,-n e 314.9'3.m At'::least one af sthe following Overpressure Protection Systems shall ~*'R 171; ~ .:,... i.i: ^' - be OPERABLE::. "7 Yt.h.;h:MlT:^r'~y: f,=.w,W f:-O %.if?.T. r.s 4' ..r.M. 4:=[ ~
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.:}.?[1l&gg,.FMJexc. power-operated relief valves ( ..'f;.' e.' Tig $ (. $ Twa . - s%;p eed the Timit established iri Figure 3.4-4,.or --12tW :s ' A v.R.:'.ib'i ? fi??.h G.G.M M Q li2.'?M'ie. W :- ..q~. r. g;*w C.p....M.The React.or CooTant System (RCS) depressurized with an RCS vent.of '. J. '.17 s... .W:.....,H.n. +:.O ... e m 2.../,n.5.,.%c-greater than or. equal ; to.2. square. inches. - s s . j;,.a... rJ 4 .y> .. i.;,y .g . +m.. a %.. %.., %y.. P y;. p; ;, 7 % ' APPLICABILITY: MODE 13 when the temperature of any RCS cold leg is less than
- -3._...
or equal to. 368*F,. MODES 4 and 5,. and MODE 6 with the reactor vessel head on. ~ .-: m n-s.u-::i;;;g w.:x: e,. -hoo?c1?O is orb TML ~ ... ex.M ACTIONi *.-e%. @ M,r.L9di. M / "2:...43:@ y ~N3 one.N.@T "E*MS . w h rend untoes.. ~ y: ' V. '~ " M 'M.'<.%ip.M.gi$y# t ith one PORVAinoperable' either restore the S:p:dMAto - 3 .. a W / Q:{D.Tg$q. Fe..;FY@?.?.@ ' OPERAB .a;:>... t..'.)m:T,;-
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'<through~atileast a 2 square nch vent within the next 8 hours. .. :.: _ c. c hl$ccy,7,v..+ q .$$9,WM:h?'90wd bd4h@.' re.lial. OMoe.S ; - ' ~ h'-,MM.M . M ; M P.With both PORVsAinoPerab .?! Q V.@('yf.;1[.N.T.Wieast a Z; square' inch ve.re,:. eep t a"., ~7~ jk.TER rdId OPddeS P nt within 8 hou.- ','. % ". 'f, 'n c : -c.b In:the event either the PORVsAor the RCS vent (s) are used to mitigate 7e ^~...*. " an RCS pressure L.-ansienti, a 5pecial Report shall be prepared and ~ d't i % i : subattted to-the Commission pursuant to Specification 6.9.2 within . -ht. ". ' J.'y
- 30. days. The report. shall describe the circumstances initiating the i-?.
- f..~ transient,. the effect of the PORVs r RCS vent (s) on the transient, and, any corrective action nece.ssary to prevent recurrence.
The prov.isions of Specification 3.0.4-are not applicab ) 4_ l@. wcI5n (21 -~ d l l s a f 1 e O ~ CALLAWAY - UNIT 1 '3/4 4-34
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v e ? - d.> MW.Y:-RT~$Vertfying tha PORY isolation ^ valve is open at least once per-72 hours W. - c.6.M: c:.; h .f w M.W-.. z.en the PORV: is b.eing.used for overpressure protection. rWy.&.:e.. m.,.,..,. p... :,.,... > m?.... >. .,;.,a.:. e.,..-.... .E T.M ; M'1,. ,').e.,.,.c s.iM7.W. M:&... 9 9..E %"i: c .y .. y_...x... n.. .@.... _. p,:g,.,...ym'14' 419.I% The RCS Vent (s) shalT be verified to be open at least once pe 2 'P..... ,.g . -? M..,y~.f,12'. hours." when.the vent (s).is being used for overpressura protection. . ~.... . ~... - - n..
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$. + 'Q:.. }~.,. 'Except wnen the vent pathway is~ providad with a valve which is locked, sealed, u.e - or otherwise secured in-the open. position, then verify these valves. open at S. , least onca per_31 days. w 72 e. f ..,-d w ,,,..ei..w..V' 4 4 .. ',.W.Og.<,y%,v e 7. 2--
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.y l .~. -... ... v. . ~ 4. O'PatAEE. A.' d,, M 0 Mn O m3; E ~ Amw mea cre %wer WAm.. b cyMer r'pcer.l ak l. east once pa 31 dqs W 3 A7ota $m16w.B are cpel ch knd om-pr / 12 houx5 i ges ike_ 'EIS. OuddM OSF cell cuogressee ? k Te@ puxsur.m 4 ": p S ct M M - CALLAWAY - UNIT 1. 3/4-4-35 g
CONTAINMENT SYSTEMS ' CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 13.6.1.6 LThe structural integrity of the containment vessel shall-be maintained at'a level consistent with the acceptance criteria in Specification 4.6.1.6. -APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: 4 a. With the structural integrity not conforming to.the requirements of Specification 4.6.1.6.1, be.in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. b. With the structural integrity of the containment not conforming at a-level consistent with the acceptance criteria of Specification 4.6.1.6.2,' restore structural integrity or complete an engineering evaluation that assures structural integrity. prior-to increasing reactor coolant temperature above 200*f. ( SURVEILLANCE REQUIREMENTS
- 4.6.1.6.1' Containment Vessel Tendons, End Anchorages and Adjacent Concrete Surfaces.
The containment vessel tendons' structural integrity shall.be cemon-strated at the end of 1, 3, and 5 years following the initial containment vessel structural integrity test and at'5 year intervals thereafter. The tendons' structural integrity shall be demonstrated by: a. Determining that a random but representative sample of at least 11 l-tendons (4 inverted U and 7-hoop) each have an observed lift-off ~ L force within predicted limits'for each. For each subsequent inspec-l . tion one tendon-from'each group may be kept unchanged to develop a history and to correlate the observed data. If the observed lift-off force of any one tendon i.n the original sample population lies between the predicted lower limit and 90% of the predicted lower . limit, two tendons,.one on each side of this tendon should be checked for their lift-off forces. If both of these adjacent' tendons are L
- found to be within.their predicted limits, all three tendons should
.be restored to the required-level of integrity. This single-deficiency [ may be considered unique and acceptable. Unless there is abnormal i degradation of the containment vessel during the first three inspec-tions, the. sample population for subsequent inspections shall-include e . CALLAWAY-- UNIT:1' 3/4.6-8 i
i CONTAINMENT' SYSTEMS SURVEILLANCE RFQUIREMENTS (Continued) at least 6 tendons (3 inverted U and 3 hoop). If more than one tendon has an observed lift-off force between the predicted lower limit and 90% of the predicted lower limit, or with one tendon below 90% of the predicted lower limit, it shall be considered as evidence of possible abnormal degradation for the purposes of Specification 4.6.1.6.lg.; b. Performing tendon detensioning, inspections, and material tests on a previously stressed tendon from each group (inverted U and hoop). -A randomly selected tendon from each group shall be completely detensioned in order to identify broken or damaged wires and deter-mining that over the entire length of the removed wire that: 1) The tendon wires are free of unacceptable (pitting of 1/64 inch or deeper and minimum of 1/32 inch in diameter) corrosion, cracks, and damage. The presence of unacceptable corrosioa, cracks, or other damage shall be considered evidence of p:::ib b abnormal degradation of the containment structure for the purposes of Specification 4.6.1.6.lg.; 2) There are no changes in the presence or physical appearance of the sheathing filler grease. Abnormal changes in the presence or physical appearance of the sheathing filler grease shall be . considered evidence of p:::ibh abnormal degradation of the con-tainment structure for the purposes of Specification 4.6.1.6.lg. ; and 3) A minimum tensile strength of 240,000 psi (guaranteed ultimate strength of the tendon material) exists for at least three wire samples (one from each end and one at mid-length) cut from each removed wire. Failure of any one of the wire samples to meet the minimum tensile strength test shall be considered as evidence of pc::ib h abnormal degradation of the containment vessel structure for the purposes of Specification 4.6.1.6.lg. Performing tendon retensioning of those tendons detensioned for c. inspection to their observed lift-off force with a tolerance limit of +6%. During retensioning of these tendons, the changes in load and elongation should'be measured simultaneously at a minimum of three approximately equally spaced levels of force between zero and the seating force. If the elongation corresponding to a specific load differs by more than 5% from that recorded during installation, an investigation should be made to ensure that the difference is not related to wire failures or slip of wires in anchorages; d. Assuring the observed lift-off stresses adjusted to account for elastic losses exceed the average minimum design value given below: -Inverted U 139 ksi Hoop: Cylind - 147 ksi Dome 134 ksi CALLAWAY - UNIT 1 2/4 6 ') L
-e CONTAINMENT SYSTEMS dirul l . SURVEILLANCE REQUIREMENTS (Continued) Verifying the OPERABILITY of the sheathing filler grease by assuring: . e. '1) If the installed quantity of grease exceeds that withdrawn by
- S% or more, an investigation shall be conducted to assure that excessive leakage has not occurrad in the tendon duct system,
'2) Minimum grease coverage exists for.the different parts of the anchorage system, and 3) The chemical properties of the filler material are within the tolerance limits as specified by the manufacturer. Failure to satisfy Specification 4.6.1.6.le. 2) or 3) above for OPERABILITY of the sheathing filler grease shall be considered as evidence of ?:::fil: abnormal degradation of the containment structure for~the' purposes of-Specification 4.6.1.6.lg. f. Determining through inspection that no apparent degradation has occurred in the visual appearance of the end anchorage or the cor. crete surfaces adjacent to the end anchorages. If apparent degradation has occurred in.the visual appearance of the end anchorage or the' concr'ete surfaces adjacent to the end. anchorages, it shall be con- -sidered as evidence of pe::iti: abnormal degradation of the contain-ment structure.for the purposes of Specification 4.6.1.6.lg.; and g. If evidence of p;;;ibic-abnormal degradation of the containment structure is detected during the performance and/or evaluation of the results of the above tests, the following actions shall be completed: 1)- Reported to the NRC within 10 days, 2) Perform an engineering evaluation demonstrating the continued ability of the containment structure to perform its design function. If continued containment integrity cannot be assured by engineering analysis within 90 days, ACTION a. required by Specification-3.6.1.6 shall be taken, and 3) Provide a determination of the cause of the apparent degradation and performan.e of any corrective actions necessary to ensure continued containment integrity. 4.6.1.6.2 Containment Vessel Surfaces. The structural integrity of the exposed accessible interior and exterior surfaces of the containment. vessel, ~ including the liner plate, shall be' determined during the shutdown for each Type'A containmentileakage rate. test (reference Specification 4.6.1.2) by a visual-inspection of these surfaces. This inspection shall be performed prior to the Type ' A ' containment 11eakage rate test to verify no apparent changes in appearance or other abnormal degradation. 4 1 CALLAWAY - UNIT.F 3/a'6-10 n ' ' -___._,_,.,.u___._.
l f CONTAINMENT SYSTEMS 3/4.6.3' CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 The containment isolation valves specified in Table 3.6-1 shall be OPERABLE with isolation times as shown in Table 3.6-1. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one or more of the containment isolation valve (s) specified in Table 3.6-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetratica that is open cnd: a. Restore the inoperable valve (s) to OPERABLE status within 4 hours, or b. Isolate each affected penetration within 4 hours by use of at least one deactivated automatic valve secured in the isolation position and the provisions of Specification 3.0.4 are not applicable, or c. Isolate each affected penetration within 4 hours by use of at least one closed manual valve or blind flange and the provisions of Specification 3.0.4 are not applicable, or d. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.6.3.1 The containment isolation valves specified in Table 3.6-1 shall be demonstrated.0PERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test, and verification of . isolation time. P 4 'CALLAWAY - UNIT 1 3/4 6-16
M ELECTRICAL POWER SiSTEMS A.C. SOURCES SHUTDOWN - LIMITING CONDITION FOR 0PERATION 3.8.1.2 LAs a minimum, the following A.C. electrical power sources shall be ~ ' OPERABLE: a. One circuit between the offsite transmission network and the Onsite Class 1E Distribution System, and b. One diesel generator with: 1) A day tank containing a minimum volume of 390 gallons of fuel, 2). ;A fuel storage system containing a minimum volume of 85,300 gallons of fuel, and 3) A fuel transfer pump. APPLICABILITY: MODES 5 and 6. ACTION: With less than the above minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, ~ positive reactivity-changes,- movement of irradiated fuel, or crane operation with. loads over the spent fuel-pool, ind uithi-S hour:, deprc;:uri:: and vent -.th: R:::ttr C 1:nt Sy; tem thr ugh at 1:::t a 2 squera inch v ht. In addition, when in t00E 5 with the reactor coolant loops not filled, or in MODE 6 with the water level less than.23 feet above the reactor' vessel flange, immediately initiate corrective' action to restore the required sources to OPERABLE status as soon as possible. l t-SURVEILLANCE REOUIREMENTS 4.8.1.2' The above required A'.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the requirements of Specifications 4.8.1.1.1, 4.8.1.1.2 (except for Specification 4.8.1.1.2a. 5)), and 4.8.1.1.3. CALLAVAY - UNIT 1-3/4 8-8 --er s wmmwe
ELECTRICAL POWER SYSTEMS f' D.C. SOURCES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, the following 0.C. electrical sources shall be OPERABLE: a. 125-Volt Battery Bank NK11 and NK13, and its associated full capacity charger NK21 and NK23, or b. 125-Volt Battery Bank NK12 and NK14, and its associated full capacity chargers NK22 and NK24. ~ APPLICABILITY: MODES 5 and 6. ACTION: a. With the required battery bank inoperable, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes or movement of irradiated fuel; initiate corrective action to restore the required battery bank to OPERABLE status as soon as possible, cnd within S hcurs, deprc;;urice cnd vcat th Rccctor Cocicnt Systc through ct lecst c 2 ;qucrc inch vent. b. With the required full-capacity charger inoperable, demonstrate the OPERABILITY of its associated battery bank by performing Specification 4.8.2.la.1) within 1 hour, and at least once per 8 hours thereafter. If any Category A limit in Table 4.8-2 is not met, declare the battery inoperable. SURVEILLANCE REQUIREMENTS 4.8.2.2 The above required 125-volt battery banks and associated chargers shall be demonstrated OPERABLE in accordance with Specification 4.8.2.1. 4 CALLAWAY - UNIT 1 3/4 8-12
i.' .. a. y :. v.+.. 1 ga hh h hh hh 'i 'D W:.. J ~~. - P r REACTOR COOLANT SYSTEM .) -/. S.Q.:c-in c;; s ~ . g'. ' M 7/ ~ ^ Y. ,.. :..g 7, j .,.n . 71.y,re.. 7, 4 g : Y; M BASES' M ", *).31.,. . g' i ~ '? ~' i.1 A; ::- y. - uy.,,..f., y.. j.. g J HEATUP (Continued) 'f ' e,,, '".,, g t..N MW2 .m ,, :.f, ':IXZ ,.:Q .p%. c;&.. s. s ys. W A.s.:n. :;'ie. VOW.H.;Y ' V% m 4 QN. p : &:1,. .n + y. p e v w,,-:.;.:.-. m ..o . d?pg't? >cTheJ'use.'of.the< composite curve is necessary to set conservative heatup - f ' 3hm%cour.tationsibecause'it.it possible for conditions to exist such that over t m M Q c. 3Timi '.. :S se of the. heatup ramp the controlling condition switches from the inside ' ..hk.#It.ta the outside and the pressure limit must at, all times be based.on analysis NE.31%of the most' critical criterionJ ' Mhl U.Q %@d-E:h%;..d t.e. '?h 6; t ~' s '^*h~' .3@g/;MipO.kFiria1192ths 'compositisi ~c'u;rves for the,heatuo r '3 7,y in J fr .y $ rata-data are adjusted for possible eirort in the pressure and temperature i 3;. . rig.y;jQsensing l nstruments by: the values l indicated on the respective curves. - c., T,,{lj L A+nfC,*,fi$;$WO.P 4C.::}}.D%.WO.'k -?*T.CW ' :'. W/lM.WWAlthough'the prescrizer operates :itr, temperature. ranges above those 3f@b yhich.there is' reasort for concern of.nonductile failure, operating limits " Md. are provided to assure compatibility cf operation with the fatigue analysis ~ ' $@@4.Mrlperformed in accordance with the ASME' Code rpirements. f Mij@;(ff"CFMF? ~ 2' WN[ ' ' foo EH7. Wrt8W rtid.(- Ofdues , y?. '? The OPERABILITY ~ of two PORVshor an RCS vent opening of at least 2. square Mg inches ensures that the RCS will.be protectea from pressure transients which S py @NMcould exceed the limits. of Appendi.$G to:10 CFR Part 50 when one or more, g[L ed C 3.'.r;4 =JVg{g,qpf ar the RCS ccTdslegs are less thanfor equal te-3 .Ef ther PORV&nas - re i t-gsthe transient' is. limited to eithed.?(1) the start of an idle RCP with the .S~ TMW6M/Usscondary water temperature of ths# steam generator less than or equal to 50*F [T%above the RCS ' cold leg temperaturiiis,'or (2) the start of a centrifugal charging -M -. pump and its injectiori into. a water-solid. RCS. y l -9 3/4.4.10 STRUCTURAL INTEGRITY y~ "*, s.%. : ~ - . The inservice inspectiod and testing programs for ASME Code Class 1, 2,. ' and 1 compo.nents ensure that the structural integrity and operational readiness -of these components wilt be-maintained at an acceptable level throughout the life of the plant. These programs are ist accordance with Section XI of-the i ASME Boiler and Pressure Vessel Code and applicable Addenda. as required by la CFR 50.55a(g), except where specific written relief has been granted by the-Commission pursuant to 10 CFR 50.55a(g)(6)(i). Components of the Reactor Coolant System were designed to provide access j-to permit inservice inspections in accordance with Section XI of the ASME ' Soiler and Pressure-Vessel Code,1974: Edition and Addenda. through Summer 1975. P y T 9 CALLAWAY - UNIT L 83/44-15
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l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE J 3/4.0 APPLICABILITY............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - T > 200 F........................... 3/4 1-1 avg 3/4 1-3 Shutdown Margin - Tavg $ 200 F........................... Moderator Temperature Coefficient........................ 3/4 1-4 Minimum Temperature for Criticality...................... 3/4 1-6 3/4.1.2 - BORATION SYSTEMS Flow Path - Shutdown..................................... 3/4 1-7 Flow Paths - Operating................................... 3/4 1-8 Charging Pump - Shutdown................................. 3/4 1-9 Charging Pumps - Operatin............................... 3/4 1-10 M 5 44 Borate d Water Sourc e., - Hele :tm.......................... 3/41-1g Borated Water source - S.- 8-' 9 Borated Water Sources - 0"edes i,tt 3cr: ting........................ 3/4 1-123 M MOVABLE 'ONTROL ASSEMBLIES 3/4.1.3 C Group Height............................................. 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R0D............................................... 3/4 1-16 Position Indication Systems - Operating.................. 3/4 1-17 Position Indication System - Shutdown.................... 3/4 1-18 Rod Drop Time............................................ 3/4 1-19 4 Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Rod Insertion Limits............................. 3/4 1-21 FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP 0PERATION............................... 3/4 1-22 FIGURE 3.1-2 (BLANK)............................................. 3/4 1-23 IV CALLAWAY - UNIT 1
p
== WIIIls h INDEX e LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE EMERGENCY CORE COOLING SYSTEMS (Continued) -3/4.5.5 00RON INJECTION SYSTEii B e re n I n j e c t i o n Ta n k.......................... '........... 3/4 0 10 5 10 3/4. 5.4- > REFUELING WATER STORAGE TANK............................. 3/4 5-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT . Containment Integrity.................................... 3/4 6 4 Containment Leakage...................................... 3/4 6-2 Containment Air Locks.................. 3/4 6-4 Internal Pressure........................................ 3/4 6-6 Air Temperature.......................................... 3/4 6-7 Containment Vessel Structural Integrity.................. 3/4 6-8 Containment Ventilation System........................... 3/4 6-11 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System................................. 3/4 6-13 Spray Additive System.................................... 3/4 6-14 Containment Cooling System.......~........................ 3/4 6-15 3/4.6.3 CONTAINMENT ISO LATION VALVES............................. 3/4 6-16 TABLE 3.6-1 CONTAINMENT ISOLATION VALVES.......................... 3/4 6-18 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers....................................... 3/4 6-30 Hydrogen Control Systems................................. 3/4 6-31 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE .SafetyValves.......................[.................... 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP OPERATION................... 3/4 7-2 3/4 7-2 TABLE 3.7-2 (BLANK)......... CALLAWAY - UNIT 1 IX s
DUT INDEX BASES SECTICN' PAGE REACTOR COOLANT SYSTEM (Continued) 3/4.4.5 STEAM GENERATORS.......................................... B 3/4 4-3 3/4.4.6 REACTOR COO LANT SYSTEM LEAKAGE............................ B 3/4 4-4 3/4.4.7 CHEMISTRY................................................. B 3/4 4-5 3/4.4.8 -SPECIFIC ACTIVITY......................................... B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................... B 3/4 4-6 TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS.......................... B 3/4 4-10 FIGURE B'3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) As A FUNCTION OF FULL POWER SERVICE LIFE........................ B 3/4 4-12 FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER ON SHIFT OF RTNDT FOR REACTOR VESSEL STEELS EXPOSED TO IRRADIATION AT 550 F........................... B 3/4 4-13 3/4. 4.10 STRUCTURAL INTEGRITY..................................... B.3/4 4-15 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.............................................. B 3/4 5-1 and 3/4.s.4 3/4.5.2 e+wk 3/4. 5. 3*ECCS SUBSYSTEMS............................... B 3/4 5-1 3/4.5.4 00RCN INJECTION SY5 TEM..'.................................. 0 /4 5 2 l 1/4.5.5 REFUELING WATER STORAGE TANK.............................. B 3/4 5-2 P/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT....................................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS...................... B 3/4 6-3 3/4.6.3 CONTAINMENT ISO LATION VALVES.............................. B 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L................................... B 3/4 6-4 J CALLAWAY - UNIT 1 XV m
-REACTIVITY-CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS FLOW PATH - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, and of the following boron injection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergency power source: a. A flow path from the Boric Acid Storage System via a boric acid transfer pump and a centrifugal charging pump to the Reactor Coolant System if the Boric Acid Storage System in Specification 3.1.2.5a, or 3,f.'2.64. (c,5 etsklO Moce)is OPERABLE; or b. The flow path from the refueling water storage tank via a centrifugal charging pump to the Reactor Coolant System if the refueling water storage tank in Specification 3.1.2.5b.4 is OPERABLE. W 3.I 3.lo 4 (,as dic})A Ny Mb0 APPLICABILITY: MODES 4, 5, and 6. ACTION: Witit none of the above flow paths OPERABLE or capable of being powered from an OPERABLE emergency puwer source,. suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REQUIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE at least once per 31 days by verifying that each valve (manual, power operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. CALL /,WAY - UNIT 1 3/4 1-7 o
DD[/:T REACTIVITY CONTROL SYSTEMS p 31 a< E Mooe.s 5 & (, BORATED WATER SOURCE - -El'UTDS"' LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE: a. A Boric Acid Storage System with: 2368 1) A minimum contained borated water volume of 4913 gallons, 2) Between 7000 and 7700 ppm of boron, and 3) A minimum solution temperature of 65*F. b. The refueling water storage tank (RWST) with: 55 9I6 1) A minimum contained borated water volume of-53,,500 gallons, 2) A minimum boron concentration of 2000 ppm, and 3) A minimum solution temperature of 37 F. APPLICABILITY: MODES 5 and 6. . ACTION: -With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REOUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE: a. At least once per 7 days by: 1) Verifying the boron concentration of the water, 2) Verifying the contained borated water volume, and 3) Verifying the Boric Acid Storage System solution temperature when it is the source of borated water. b. At least once per 24 hours by verifying the RWST temperature when it is the source of borated water and the outside' air terrorature is less than 37*F. CALLAWAY - UNIT 1 3/4 1-11
REACTIVITY CONTROL SYSTEMS [ /t Hoot 4 U f f BORATED WATER SOURCES - 0^ RAT!!'O ' LIMITING C0flDITION FOR OPERATION one of 3.1.2.6 As a minimum,^the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2: I a. A Boric Acid Storage System with: n,658 1) A minimum contained borated water volume of -1G;-1+2 gallons, 2) Between 7000 and 7700 ppm of boron, and 3) A minimum solution temperature of 65*F. b. The refueling water storage tank (RWST) with: 1) A minimum contained borated water "olume of 394,000 gallons, 2) Between 2000 and 2100 ppm of boron, 3) A minimum solution temperature of 37*F, and 4) A caximum solution temperature of 100 F. APPLICABILITY: MODEl 1, 2, 2, :nd 4. ACTION: no borafed techer source, OPERABLE 3 a. With th; Scri: ' i d--s k % g Sy:te: 'n per:b!: and being used 2: := cf th; abre; requir:d b r^t;d unter%:o:urs or, be in.at 1;;st "0T urces restore the storage system to OPERABLE status within fa ST/fDSY withir the n:AS hour; and berated to. SHUTDOWh' t'ARCIt!
- quiv:?:nt t at 1 ::t 1% 'k/k at 200 F; meter -the Scric Acid Stor:g; Syst;.T. to OPERACLE statu:
ithin th; ncxt 7 days er bc in COLD SHUTOOWN within the next se hours. 24 b. h'ith thc Rh'ST incperab?;, rc; tor; the tank.tc OPERACLE statu within and 4n 1 hcur er be '. m.at 102:t ",0T ST'"_;D0Y-within the acxt S hcur . 3 3. _.. : _n u.__...,,.- -, n - o n m. ~,,.o -.u.... 3 wu .~ m SuRVEnt.ACC6 RGGUtDEA.WCT5 4.l.7.G TN. abmic Vegtored berstd wal.cr source Shlt k dhowsb+-ed CeERAst.e b3 h perGormwe, M e.ac.k of h reiuwemtats oS Spectfstahon 4. t. L. 5. CALLAWAY - UNIT 1 3/4 1-12
REACTIVITY CONTROL SYSTEMS g) i ucces 1,7 43 5# BORATED WATER SOURCES - SPERAT-INS LIMITING CONDITION FOR OPERATION 7 3.1.2.K As a minimum, the following borated water source (s) thall be OPERABLE as_ required by Specification 3.1.2.2: a. A Boric Acid Storage System with: 1) A minimum contained borated water volume of 10,142 gallons, n,Gsa 2) Between 7000 and 7700 ppm of boron, and 3) A minimum solution temperature of 65'F. b. The refueling water storage tank (RWST) with: 1) A minimum contained borated water volume of 394,000 gallons, 2) Between 2000 and 2100 ppm of boron, 3) A minimum solution temperature of 37*F, and 4) A maximum solution temperature of 100'F. APPLICABILITY: MODES 1, 2, 3, :. '- ACTION: a. With the Boric Acid Storage System inoperable and being used as one of the above required borated water sources, restore the storage system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent to at least 1% ak/k at 200 F; restore the Boric Acid Storage System tc OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. b. With the RWST incperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the fo11cwing 30 hours. i t e e CALLAWAY - UNIT 1 3/4 1-1
REACTIVITY CONTROL SYSTEMS 4 SURVEILLANCE REQUIREMENTS 'I 4.1.2.5 Each borated water source shall be demonstrated OPERABLE: a. At least once per 7 days by: 1) Verifying the baron concentration in the water, 2) Verifying the contained borated water volume of the water source, and 3) Verifying the Boric Acid Stcrage System solution temperature when it is the source of torated water. b. At least once per. 24 hours by verifying the RWST temperature when the outside air temperature is either less than 37 F or greater than 100*F. e eee b 9 d<_ { e 0 CALLAWAY - Uti!T 1 3/4 1-1/ m... m.
REACTIVITY CONTROL SYSTEMS h7 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Ell'u'j GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1' All full-length shutdown and control rods shall be OPERABLE and positioned within i 12 steps (indiceted position) of their group step counter demand position. APPLICABILITY: MODES la and 2*. ACTION: a. With one or more full-length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour and be in HOT STANDBY within 6 hours, b'. With more than one full-length rod inoperable or misaligned from the group step counter demand position by more than i 12 steps (indicated position), be in HOT STANDBY within 6 hours. c. . With one full-length rod trippable but inop'erable due to causes other than addressed by ACTION a., dove, or misaligned from its group step counter demand height by more than 12 steps (indicated i-po=ition), POWER OPERATION may continue provided that within 1 hour:
- 1. -.The rod is restored to OPERABLE status within the above alignment requirements, or 2.
The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within 't'12 steps of the inoperable rud while maintaining the rod sequence and insertion limits of Figures 3.1-1 and 3.1-2. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3. The rod is declared inoperable and the SHUTDOWN MAPGIN 4 requirement of Specification 3.1.1.1 is satisfied. POWER i OPERATION'may then continue provided that: l a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions; b) ' The SriUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at-least once per 22 hours; ' Soc 4cdoes
- See-Special Test Exceptiens^3.10.2 and 3.10.3.
5 3/4 1-1/ CALLAWAY - UNIT 1 f4
m _ _. DRUT REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION 'hCTION(Continued) i c) A power distribution map is obtained from the movable .incore detectors and F (Z) and F are veri ed to be q H within their limits within 72 hours; and d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and ~ ~ within the following 4 hours the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER. p I. t -SURVEILLANCE REQUIREMENTS I e 4.1.3.1.1.The position of each full-length rod shall be determined to be within the group demand. limit by verifying the individual rod positions at least once per 12 hours except during time intervals when the rod position deviation monitor'is inoperable, then verify the group positions at least once L per 4 hours. 4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any onc direction at least once per 31 days. t PI - t 1-;f CALLAWAY - TWIT 1 ~ 3 /.t =. . =. -
DRIFT ~ -TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION ~ IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment loss o'f Reactor Coolant from Small Ruptured Pipes or from Cracks in Large Pipes Which Actuates-the Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal at Full Power Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident) Major Secondary Coolant System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection) e e b e 9 6 7 CALLAWAY - UNIT 1 3/4 1-1/
QQ1(( REACTIVITY CONTROL SYSTEMS UIUll B -POSITION'INDIdATIONSYSTEMS-0PERATING LIMITING CONDITION FOR-OPERATION 13.1.3.2 The Digital Rod Position Indication System and the Demand Position Indication System shall be OPERABLE and capable of determining the control
- rod positions within i 12 steps.
APPLICABILITY: MODES 1 and 2. ACTION: a. With a maximum of one digital rod position indicator per bank inoperable either: 1. Determine the position of the nonindicating roc (s) indirectly 'by the movable incore detectors at least once per 8 hours and immediately after any motion of the nonindicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours. b. With a maximum of'one demand position indicator per bank inoperable either: 1. ' Verify that all digital rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours, or 2. Reduce THERMAL POWER to less than'50% of RATED THERMAL POWER within 8 hours. SURVEILLANCE REOUIREMENTS 4.1.3.2 Each digital rod position indicator shall be determined to be OPERABLE by verifying that the Demand Position Indication System and the Digital Rod Position. Indication System agree within 12 steps at least once per 12 hours except during' time intervals when the rod position deviation monitor is inoperable, _ then compare 'the Demand Position Indication System and the Digital Rod Position Indication System'at least once per 4 hours. 8 ~ 2/4 1-1/ .CALLAWAY - UNIT 1
m REACTIVITY CONTROL SYSTEMS wilru' POSITION INDICIATION SYSTEM-SHUTDOWN LIMITING CONDITION FOR OoERATION 3.1.3.3 One digital rod position indicator (excluding demand position indica-tion) shall be OPERABLE and capable of determining the control rod position within i 12 steps for each shutdown or control rod not fully inserted. APPLICABILITY: MODES _3*#, 4*# and 5*#. ACTION: With less than the above required position' indicator (s) OPERABLE, immediately open the Reactor Trip System breakers. SURVEILL NCE REQUIREMENTS 4.1.3.3 Each of the above required digita'l rod position indicator (s) shall be determi ?d to be OPERABLE by verifying that the digital rod position indicator agrees with the demand position indicator within 12 steps when exercised over the full-range of rod travel at least once per 18 months. "With the Reactor Trip System breakers in the closed position.
- See Special Test Exceptiony3.10.5.
Specihcaben 9 CALLAWAY - UNIT 1 3/43-1/
~ REACTIVITY CONTROL SYSTEMS R0D OROP TIME -LIMITING CONDITION FOR OPERATION I 3.-l. 3. 4 The individual full-length shutdown and control rod drop time from the fully withdrawn position shall be less than or equal to 2.2 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with: T,yg greater than or equal to 551 F, and a. b. All Reactor Coolant pump.e operating. APPLICABILITY: MODES 1 and 2. ACTION: a. With the rod drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit 4 prior to' proceeding to MODE 1 or 2. b. With the rod drop times within limits but determined with three reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 66% of RATED THERMAL POWER. i SURVEILLANCE REOUIREMENTS 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality: a.. For all rods following each removal of the reactor vessel head, b. For specifically affected individual rods following any maintenance on'or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and .l c. At least once per 18 months. ~ 10 e fCALLAWAY.- UNIT 1 3/4 1->6
REACTIVITY CONTROL SYSTEMS [ SHUTOOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn. APPLICABILITY: MODES 1* and 2*#. ACTION: With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour either: a. Fully withdraw the rod, or l b. Declare the rod to be inoperable and apply Specification 3.1.3.1. l t SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn: a. Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality, and b. At least once per 12 hours thereafter. Spetdaabons
- See Special Test Exceptions ^3.10.2 and 3.10.3.
- With K,ff greater than or equal to 1.
I CALLAWAY - UNIT 1 3/4 1-27
REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figures 3.1-1 and 3.1-2. APPLICABILITY: MODES 1* and 2*#. ACTION: With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2: a. Restore the control banks to within the limits within 2 hours, or b. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the above figures, or c. Be in at least HOT STANDBY within 6 hours. SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12-hours except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours. Specihcattom
- See Special Test Exceptions ^3.10.2 and 3.10.3.
- With K,ff, greater than or equal to 1.
2 CALLAWAY - UNIT 1 3/4 1-22
CRAFT (FULLY WITHDRAWN)! 228 ~/(29.3%', 228E /(79.3 /o!E25lh% ~~ ~ 220 f y- -/ ~ ' ~ ./. ./ j- /: DANK B:- I 180-c r- ~ (100 % 161). l 1M (0% 161)- / /; N. if . / =T 3 140 / g- [ BANK C E / g 5- [ P / 13 100 I e / / 5 / / ![ BANK Dr-" a G. y .f, a. r/ r......... / (0 % 46) / g 20' 0' - - - - - ~ ~ ~ ~ 'Oi 20l 40; 60 80 100 OWER Pement) (FULLY INSERTED)l FIGURE 3.1-1 ROD BANK IriSERTI0tl LIMITS VERSUS THERMAL POWER - FOUR LOOP OPERATION 3 CALLAWAY - 'JNIT 1 3/4 1-22'
LI ' ~., :, Figure 3.1-2 left blank pending NRC approval of three loop operation G e CALLAWAY - UNIT 1 3/4 1-2
g - 2 P J d ATTACIDIENT 4 Specifications Not Resolved i 4 6 s i e + . <f
f D / FT' POWER DISTRIBUTION LIMITS y ]pf l LIMITING CONDITION FOR OPERATION ACTION (Continued) b. Within 24 hours of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that the combination of R and RCS total flow rate are restored to within the above limits, or reduce THERMAL POWER to less than 5% of RATED THERPAL POWER within the next 2 hours, and Identify and correct the cause of the out-of-limit condition prior c. to increasing THERMAL DOWER above the reduced THERMAL POWER limit required by ACTION a.2. and/or b., above; subsequent POWER OPERATION may proceed provided that the combination of R and indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable operation shown on Figure 3.2-3 prior to exceeding the following THERMAL POWER levels: 1. A nominal 50% of RATED THERMAL POWER, 2. A nominal 75% of RATED THERMAL POWER, and 3. Within 24 hours of attaining greater than or eoual to 95% of RATED THERMAL POWER. SURVEILLANCE REOUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 The combination of indicated RCS total flow rate and R shall be determined to be within the region of acceptable operation of Figure 3.2-3: a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and b. At least once per 31 Effective Full Power Days. 4.2.3.3 The indicated RCS total flow rate shall be verified to be within the region cf acceptable operation of Figure 3.2-3 at least once per 12 hours when the most recently obtained value of R, obtained per Specification 4.2.3.2, is assumed to exist. 4.2.3.4 The RCS loop flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. 4.2.3.5 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months, taith i dam prior to performin3 h eru.tuun hesb batece, measuremewt, der icmperebne.h, mst.ruvuentabon used be deleyrm \\ccatootcc pressure,bc calibrated. , and 4c4bater ued.un AP m h L.alorimd.ric. Seeb cakut abow skatt 4.L 3 4 The %eg* der Moduris skall *. inspecked ad deaWed d necessary al, leasl TLD EAY - Y t i 3/4 2-10
TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE 1. Triaxial Peak Recording Accelerographs a. Radwaste Base Slab i 1.0 g 1 b. Control Room i 1.0 g I c. ESW Pump Facility i 1.0 g 1 d. Ctmt Structure i 2.0 g i e. Auxiliary Bldg. SI Pump Suctions i 1.0 g 1 f. SGB Piping i 2.0 g 1 g. SGB Support i 1.0 g 1 2. Triaxial Time History and Response Spectrum Recording System, Monitoring the Following Accelerometers (Active) a. Ctmt. Base Slab i 1.0 g 1 b. Ctmt. Oper. Floor i 1.0 g 1 c. Reactor Support i 1.0 g 1 d. Aux., Bldg. Base Slab i 1.0 g i e. Aux. Bldg. Control Room Air Filters i 1.0 g 1 f]' Cf f. Free Field i 0.5 g 1 3. Triaxial Response-Spectrum Recorder (Passive) a. Ctmt. Base Slab i 1.0 g 1 4. Triaxial Seismic Switches ACCELERATION LEVEL / DIREGiON 0.12 g-.@ss.D @ o os 1 0.20 ;.i3s.1%s.83s a. OBE Ctmt. Base Slab 20s 1 b. SSE Ctat. Base Slab c. OBE Ctmt. Oper. F1. 0.13 g.los.103 '3S 1 d. SSE Ctmt. Oper. F1. 8-24-g. 14s.1 %. 28 s 1 1 e. System Trigger 0.01 g.013 013 013 CALLAWAY - UNIT 1 3/4 3-44 ... =.
~ TABLE 3.3-16 ,O_ ACCIDENT MONITORING INSTRUMENTATION R TOTAL MINIMUM NO. OF CHANNELS INSTRUMENT CHANNELS OPERABLE E 1. Containment Pressure y a. Normal Range 2 1 b. Extended Ranga 2 1 g 2. Reactor Coolant Outlet Temperature - THOT (Wide Range) 2 1 i 3. Reactor Coolant Inlet Temperature - TCOLD.(Wide Range) 2 1 4. Reactor Coolant Pressure - Wide Range 2 1 l 5. Pressurizer Water Level 2 1 2/ steam generator 1/ steam generator 6. Steam Line Pressure l R 7. Steam Generator Water Level - Narrow Range 1/ steam generator 1/ steam generator b 8. Steam Generator' Water Level - Wide Range 1/ steam generator 1/ steam generator h 9. Refueling Water Storage Tank Water Level ~. 2 1 10. Containment Hydrogen Concentration Level 2 1 11. Auxiliary Feedwater Flow Rate 1/ steam generator 1/ steam generator i 12. Reactor Coolant System Subcooling Margin Monit 2 1 13. PORV Position Indicator
- 1/ Valve 1/ Valve 14.
PORY Block Valve Position Indicator ** 1/ Valve 1/ Valve j 15. Safety Valve Position Indicator 1/ Valve 1/ Valve l
- 16.
Containment Water Level 2 1 i 17. Containment Radiation Level (High Range) 2 1 18. Thermocouple / Core Cooling Detection System. 4/ core quadrant 2/ core quadrant i i
TABLE 3.3-10 (Centinued) ACCIDENT MONITORING INSTRUMENTATION r-E TOTAL MIhIMUM . 4 NO. OF CHANNELS INSTRUMENT CHANNELS OPERABLE E 10. "eeeter Ceelent "ediaties :.e w 1/leep 1/le:p 20. Unit Vent - High Range Noble Gas Monitor 1 1 . w Reactor Vessel Water Leve h 1 2 21. 22. Steam Relief - Noble Gas Monit 4 4 23. Source Range'- Neutron F1 2 1 - 24. Auxiliary Feedwater Pump Tur Exhaust - Noble Gas Monito 2 1 Y s { TABLE NOTATIONS ~
- These instruments need not be required OPERABLE until prior to STARiUT following the first ]
refuelina outage.
- Not applicable if the associated block valve is in the closed position.
.l
- Not applicable if the block valve is verified in the closed position and power is removed.
4.. e s -1 6
J. I .g TABLE'4.'3-7 ~ h . ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS g l g CHANNEL CHANNEL 1 INSTRUPENT CHECK CALIBRATION Nh 1. Containment Pressure M R h 2. Reactor Coolant Outlet Temperature - THOT (Wide Range) M R l l 3. Reactor Coolant Inlet. Temperature - TCOLD (Wide Range) M R l 4. Reactor Coolant Pressure - Wide Range M R 5. Pressurizer Water Level M R 6. Steam Line Pressure M R i 7. Steam Generator Water Level - Narrow Range M R 8. Steam Generator Water Level - Wide Range M R 9. Refueling Water Storage Tank Water Level ' M R y 10. Containment Hydrogen Concentration Level M R m 11. Auxiliary Feedwater Flow Rate M R m 12. Reactor Coolant System Subcooling Margin Monit M R
- 13. -PORY Position Indicator
- M N.A.
14. PORY Block Valve Position Indicator ** M N.A. 15. Safety Valve Position Indicator M N.A. 16. Containment Water level M R 17. Containment Radiation Level (High Range) M R***
- 18. ^ Thermocouple / Core Cooling Detection System M
R s E 9 l
c ~ TABLE 4.'-7 (Continued) 3 SE' ACCIDENT MONITORING INSTRUMENTATIOR SURVEILLANCE' REQUIREMENTS g N CHANNEL CHANNEL . i ~ INSTRUMENT CHECK CALIBRATION E ' 10.
- t;r0;; hats;dkthn1.;;;[
R
- 20.. Unit Vent - High Range Noble Gas Monitor M
R ^ s Reactor Vessel Water Leve h M R 2L 22. Steam Relief - Noble Gas Monito M R SourceRange-NeutronFluh M R ^ 23. 24. Auxiliary Feedwater Pump Turb*ne Exhaust - Noble Gas Monit # M R = T, TABLE NOTATIONS u -s 0%
- These instruments need not be required OPERABLE until prior to STARTUP following the ir g ueling_nutage-
- Not applicable if-the associated block valve is in the closed position.
- Not applicable if the block valve is = verified in the closed position and power is removed.
- CHANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector,.for range decades above 10R/h and a one point calibration (. heck of the detector below 10R/h with an installed or. portable gamma source.
- Not applicable if the block valve is = verified in the closed position and power is removed.
\\ ,e s e 1 g 9 9
7 ;.-._.,_-._,,., ,;,.g., TABLE 3.3-13 (Continued) n> E RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION 6 4 MINIMUM CHANNELS t INSTRUMENT OPERABLE APPLICABILITY ACTION Eq 4. Radwaste Building Vent System e a. Noble Gas Activity Monitor 1 38, 40 Providing Alarm and Automation Termination of Release (GT-RE-10) i GH 43 b. Iodine Sampler 1 ~ Particulate Sampler 1 43 c. 45 d. Flow Rate N.A. w s 39 } -e. Sampler Flow Rata Monitor 1 - 4 r l
- c:p D
9 H 4 4 i i w l l t
TABLE 4.3-9 (Continued) g h-RADI0ACTIVEGASEOUSEFFLUENTMONITORINGINSTRUMENTATIONSURVEILLANCEREQUIREMNITS ANALOG CHANNEL MODES FOR WHICH E CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST IS REQUIRED 4. Radwaste Building Vent System a. Noble Gas Activity Monitor - D, P M, P R(3) Q(1) Providing Alarm and Automatic Termination of Release (M-RE-10) CH b. Iodine Sampler W N.A. N.A. N.A. c. Particulate Sampler W N.A. N.A. N.A. y d. Flow Rate F.A. N.A. R(7) N.A. y e. Sampler Flow Rate Monitor D N.A. R Q I O I I
- t3 3::=
m l: i l l l l l ! 1 1. l t
EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 BORON INJECTION SYSTEM j BORON INJECTION TANK MITING CONDITION FOR OPERATION x 3.5.5 he boron injection tank sh'all be OPERABLE with: a. minimum contained borated water volume of 90 gallons, and b. Ab on concentration of between 2000 and 2 0 ppm. APPLICA8ILITY: M c5 1, 2, and 3. ACTION: With the boron injection ank inoperable, r tor the tank to OPERA 8LE status within 1 hour dr be in H0 TANDBY and bor ed to a SHUTDOWN MARGIN equivalent to 1% ok/k at 200 F within t e next 6 hou s; restore the tank to OPERABLE status within the next 7 days r be in f T SHUTDOWN within the next 12 hours. SilRVEILLANCE REQUIREMENTS 4.5.5 The boron infection ank shall be demo trated OPERABLE by: a. Verifying the ontained borated water lume at least once per 7 days, and b. Verifying he baron concentration of the wa r in the tank at least once per days. CALLAWAY - UNIT 1 3/4 5-10
EMERGENCY CORE COOLING SYSTEMS - i F 5 i 3/4.5.5 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 5 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with: a. A minimum contained borated water volume of 394,000 gallons, b. A boron concentration of between 2000 and 2100 ppm of boron, A minimum solution temperature of 37 F, and c. d. A maximum solution temperature of 100 F. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the RWST inoperable, rest:re the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLA?q (lj REMENTS 4
- 4. 5.6 The RWST shall be demonstrated OPERABLE:
a. At least once per 7 days by: 1) Verifying the contained borated water volume in the tank, and 2) Verifying the boron concentration of the water. b. At least once per 24 hours by verifying the RWST temperature when the outside air temperature is either less than 37 F or greater than 100*F. 10 CALLAWAY - UNIT 1 3/4 5-J{ me e. e a
- sian *
- ed>,essee e>ege a-@*
- G N.4# gD +N
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b. If any periodic Type A test fails to meet either 0.75 L, or 0.75 L, g the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet either 0.75 L, or 0.75 L, a Type A test shall be performed at g least every 18 months until two consecutive Type A tests meet either 0.75 L, or 0.75 L at which time the above test schedule may be t resumed; c. The accuracy of each Type A test shall be verified by a supplemental test which: 1) Confirms the accuracy of the test by verifying that the contain-ment leakage rate calculated in accordance with ANSI N45.4-1972, i Appendix C, is within 25% of the containment leakage rate measured prior to the introduction of the superimposed leak, 2) Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test, and 3) Requires that the rate at which gas is injected into the contain-ment or bled from the containment during the supplemental test ) is between 0.75 L, and 1.25 L,. ,,/ ~ Type B and C tests shall be conducted with gas at a pressure not d. less than P,, 48 psig, at intervals no greater than 24 months except for tests involving: 1) Air locks, and 2) Purge supply and exhaust isolation valves with resilient material seals. e. Air locks shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.3; f. Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE by the requirements of gpecifications4.6.1.7(2)and4.6.1.7.4,asapplicab,le;and ,,The provisions of Spec fication 4.0.2 are not applicable. i g. 3, =.... e e8 ^ 3/4 6-3
- CALLAWAY - UNIT 1
CONTAINMENT SYSTEMS SURVEILLANCE R QUIREMENTS 4.6.1.3 'Each containment air lock shall be demonstrated OPERABLE: ~sg a. Within 72 hours following each closing, except when the air lock is 4 being used for multiple entries, then at least once per 72 hours, by { verifying that the seal leakage is less than 0.005 L, as determined ) by precision flow measurements when measured for at least 30 seconds with the volume between the seals at a constant pressure of 10 psig; ) j w b. By conducting overall air lock leakage tests at not less than P,, 48 psig, and verifying the overall air lock leakage rate is within its limit: 1) At least once per 6 months,# and 2) Prior to establishing CONTAINMENT INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability.* c. At least once per 6 months by verifying that only one door in each air lock can be opened at a time. -4, o,oos' # * ~;du &' y fgs}logh 4 T-2em
- The provis. ions of Specification 4.0 2 are not applicable.-
- This represents'an exemption to Appendix J of 10 CFR Part 50.
CALLAWAY-bNIT1 3/4 6-5 a
=- E g}I m PLANT SYSTEl15 M LI 3/4.7.11 FIRE BARRIER PENETRATIONS LIMITING CONDITION FOR OPERATION 3.7.11 All fire barrier penetrations (walls, floor / ceiling:, cable tray enclosures, and other fire barriers) separating safety related fire areas or separating portions of redundant systems important to safe shutdown within a fire area and all sealing devices in fire rated assembly penetrations (fire doors; fire windows', fire dampers',"c,able, piping, and ventilation duct penetration seals)shall be OPERABLE. (fg APPLICABILITY: At all times. ACTION: a. With one or more of the above required fire barrier penetrations inoperable, within 1 hour establish a continuous fire watch on at least one side of the affected penetration, or verify the OPERABILITY of fire detectors on at least one side of the inoperable fire barrier and establish an hourly fire watch patrol. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. i SURVEILLANCE REQUIREMENTS 4.7.11.1 At least once per 18 months the above required fire rated assemblies and penetration sealing devices shall be verified OPERABLE by performing a visual inspection of: a. The exposed surfaces of each fire rated assembly, b. Each fire window / fire damper and associated hardware, and c. At least 10% of each type (electrical and mechanical) of sealed pene-tration. If-apparent changes in appearance or abnormal degradations are found, a visual inspection of an additional 10% of each type of sealed penetration shall be made. This inspection process shall continue until a 10% sample with no apparent changes in appearance or abnormal degradation is found. Samples shall be selected such that t each penetration seal will be inspected every'15 years. 4.7.11.2 Each of the above required fire doors shall be verified OPERABLE by inspecting the automatic hold open, release and closing mechanism and latches at least once per 6 months, and by verifying: a. The OPERABILITY of the Fire Door Supervision System for each electri-cally.' supervised fire door by performing a TRIP ACTUATING DEVICE OPERATIONAL. TEST at least once per 31 days, b. That each locked closed fire door is closed at least once per 7 days, .. _ c. That doors with automatic hold-open and release mechanisms are free of obstructions at least once per 24 hours and performing a functional test at least once per 18 months, and d. That each, unlocked fire door without electrical supervision is closed at least once per 24 hours. CALLAWAY - UNIT 1 3/4 7-26
1 "' ? REACTIVITY CONTROL SYSTEMS L'i l& BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) . The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting MTC value -4.1 x 10 4 ak/k/ F. The MTC value of -3.2 x 10 4 ak/k/F represents a conservative value (with correc-tions for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these correcticns to the limiting MTC value of -4.1 x 10 4 ok/k/ F. The Surveillance Requirements for measurement of the MTC at the beginning and naar the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551*F. This limitation is required to ensure: (1) the moderator temperature coefficient is within its analyzed tamperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel ~is above its minimum RT temperature. NDT 3/4.1.2 BORATION SYSTEMS The Boration Systems ensure that negative reactivity control is available during each MODE of facility operation. The components required to perform this function inclu e: (1) borated water sources, (2) centrifugal charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators. 350 With the RCS average temperature above e902F, a minimum of two boron injection flow paths are required to ensure single functional capability in boration capability of c:g renders one of the flow paths inoperable.fThe the event an assumed fgi13.. e flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% ak/k after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires l'IA5B M,-147 gallons of 7000 ppm borated water from the boric acid storage tanks or 83, % 9 7 2,09 5-gallons of 2000 ppm borated water from the RWST. "W k CALLAWAY - UNIT 1 8 3/4 1-2 opW1Ae4. n SS mt.usary to e$ute, that a,4 w mas ad4.h m peer.urc W.s w a c w w r,w v N operabon o% a Sw}lo Ponv or HHH myi reheV valva.
REACTIVITY CONTROL SYSTEMS b5ES 80 RATION SYSTEMS (Continued) With the RCS temperature below 200 F, one Boration System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable. The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable in' MODES 4, 5, and 6 provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORVer an RHR suction rehef valve.. The boron capability required below 200 F is sufficient to provide a SHUTDOWN MARGIN of 1% Ak/k after xenon decay and cooldown from 200 F to 140 F. This condition requires either gallons of 7000 ppm borated water from the boric acid storage tanks or 12,117 gallons of 200L ppm borated water from the R\\lST. N,oM Z%8 The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics. The limits on contained water volume dnd baron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within Containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The OPERABILITY of one Boration System during REFUELING ensures that this system is available for reactivity control while in MODE 6. 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is main-tained, and (3) the potential effects of rod misalignment on associated acci-dent analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control red alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within 12 steps at 24, 48, 120 and 228 steps withdrawn for the Control Banks and 18, 210 and 228 steps withdrawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication. Since the Digital Rod Position System does not indicate the actual shutdown rod position between 18 steps and 210 steps, only points in the indi-cated ranges are picked for verification of agreement with demanded position. CALLAWAY - UNIT 1 8 3/4 1-3
-.- =___. -. POWER DISTRIBUTION LIMITS n BASES ~ HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE j HOT CHANNEL FACTOR (Continued) The Radial Peaking Factor, Fg (Z), is measured periodically to provide assurance that the Hot Channel Factor, F (Z), remains within its limit. The j F,y limit for RATED THERMAL POWER (FRTP)qas. provided in the R x Factor Limit Report per Specification 6.9.1.9 was determined from expected power control maneuvers over the full range of burnup conditions in the core. l l .WhenRCSflowrateandFharemeasured,noadditionalallowancesare necessary prior to comparison with the limits of Figure 3.2-3. Measurement 4 errorsof2%forRCStotalflowrateand4%forFhhavebeenallowedforin determination of the design DN8R value. i. The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate indicators. Potential fouling of the feedwater venturi which might not be i i detected could bias the result from the precision heat balance in a non-conservative canner. Therefore,
- enalty ef 0.1' fea undetected fee!*aa ef x ;qrM g/.the feedwater, venturi is ' eled
- d
- eige-- ?.2-3.. <.... m ~........ -.......... a w... a......a m --- 4 +..PuToc.m4 Any fee!'n; 9f:5 ef;;ht is l
rekehg hdth:5dibhvehicha!N5t3hhEENEbE35hNStAE'.'Ifdetact5d,5ction'shall be taken before performing subsequent precision heat balance measurements, 1.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measuremet or the venturi shall be cleaned to eliminate j the fouling. The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the acceptable region of operation shown on Figure 3.2-3. l 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis, ~ ~~ i Radial power distribution measurements are made during STARTUP testing and j periodically during power operation. t The limit of 1.02, at which corrective action is required, provides DN6 and linear heat generation rate protecticn with x y plane power tilts. A . limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt. The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing q the maximum allowed power by 35 for e ch percent of tilt in excess of 1. CALLAWAY - UNIT 1 8 3/4 2-5 e.... -.............. .. -~- -.. -. - -. - - - - __-,__,____,.,m.____-._ .-,-___-,--,___,._.,-._~_,__.m-m.
EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued) The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and Safety Injection pumps except the required OPERABLE charging pump to be inoperable in MODES 4 and 5 and in MODE 6 with the reactor vessel head on provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. The Surveillance Requirements provided to ensure OPERABILITY of each component ensure, that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in t% ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses. The Surveillance Requirements for leakage testing e of ECCS check valves ensure that a failure of one valve will not cause an intersystem LOCA. The Surveillance Requirement to vent the ECCS pump casings and accessible, i.e.., can be reached without personnel hazard or high radiation dose, discharge piping ensures against inoperable pump.s caused by gas binding or water hammer in ECCS piping. 3/AS.4 BORON INJECTION SYSTEM Th of the Boron Injection System as part of the ECJ,Le that sufficient e-tractivity is injected into the c poreounteract any positive increase in rea'litiv4 caused by RCS sys am tooldown. RCS cooldown can be caused by inadvertent depr(essuM2 ion oss-of-coolant accident, or a steam line rupture. The limits on i n tank minimum contained volume Qoron concentration p tr' pn e that the assumptions used in the Steam Line b7ea analysis are met -TrFe contained water volume limit includes an allowance for water de because of tank discharge line location or other physical characteristics. 4 3/4.5.5 REFUELING WATER STORAGE TANK The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the core, and (2) the reactor will remain subcritical in the cold condition folicwing mixing of the RWST cnd the RCS water CALLAWAY - UNIT 1 8 3/4 S-2 ... ~. --}}