ML20095E285
| ML20095E285 | |
| Person / Time | |
|---|---|
| Site: | Callaway  (NPF-30-A-105) |
| Issue date: | 12/07/1995 |
| From: | Thomas K NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20095E286 | List: |
| References | |
| GL-93-05, GL-93-5, NUDOCS 9512150008 | |
| Download: ML20095E285 (17) | |
Text
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1 UNITED STATES
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g 3-NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. enmas amng 4
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W ION ELECTRIC COMPANY CALLAWAY PLANT. UNIT 1 DOCKET NO. 50-483 J
l AMENDMENT TO FACILITY OPERATING LICENSE 1
i Amendment No.105 License No. NPF-30 i
1.
The Nuclear Regulatory Comission (the Comission) has found that:
l 3
I A.
The application for amendment filed by Union Electric Company (UE, the licensee) dated June 23, 1995, complies with the standards and i
requirements of the Atomic Energy Act of 1954, as amended (the Act),
j t
and the Comission's rules and regulations-set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the j
4 provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by I
4 this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted i
i in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to th'e comon defense and security or to the health and safety of the public; and i
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E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as-indicated in the attachment to this license amendment, and para-graph 2.C.(2) of Facility Operating License No. NPF-30 is heyeby amended to read as follows:
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9 g
t 9512150008 951207 PDR ADOCK 05000483 P
PDR i
. (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 105, and the Environmental Protection Plan contained in Appendix B both of which are attached hereto, are hereby
'j incorporated into the license. UE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is. effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION MQya@MWW Kristine M. Thomas, Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
December 7, 1995 A
O e
i ATTACHMENT TO LICENSE AMENDMENT NO.105 OPERATING LICENSE NO. NPF-30 DOCKET NO. 50-483
]
Revise Appendix A Technical Specifications by removing the pages identified '
below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain vertical lines indicating the area of change. Overleaf pages are provided to maintain document completeness.
REMOVE INSERT 3/4 1-1 3/4 1-1 3/4 1-3 3/4 1-3 3/4 1-4*
3/4 1-4*
3/4 1-14 3/4 1-14 3/4 1-15' 3/4 1-15 3/4 3-41 3/4 3-41 3/4 4-9 3/4 4-9 3/4 4-10*
3/4 4-10*
3/4 4-19*
3/4 4-19*
3/4 4-20 3/4 4-20 3/4 6-13 3/4 6-13 3/4 6-32 3/4 6-32 B 3/4 1-4 B 3/4 1-4 B 3/4 1-5 B 3/4 1-5 l
l
- Denotes overleaf page e
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL SHUTDOWN MARGIN - Ty > 200*F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% Ak/k.
APPLICABILITY: MODES 3 and 4.
ACTION:
)
With the SHUTDOWN MARGIN less than 1.3% Ak/k, within 15 minutes 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 I hour after detection of an inoperable (untrippable) rod (s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod (s) is inoperable. The above required SHUTDOWN MARGIN shall be verified i
acceptable with an increased allowance for the withdrawn worth of the untrippable rod (s):
1 b.
At least once 'per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors:
1)
Reactor Coolant System boron concentration, 2)
Control rod position, 3)
Reactor Coolant System average temperature, 4)
Fuel burnup based on gross thermal energy generation, 5)
Xenon concentration and 6)
Samarium concentration.
d CALLAWAY - UNIT 1 3/4 1-1 Amendment No. M3,105
REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T.,s 200*F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1% ok/k.
APPLICABILITY: MODE 5.
ACTION:
With the SHUTDOWN MARGIN less than 1% ok/k, within 15 minutes 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 RE0VIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1% ok/k at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors:
1)
Reactor Coolant System boron concentration, 2)
Control rod position, 3)
Reactor Coolant System average temperature, 4)
Fuel burnup based on gross thermal energy generation, 5)
Xenon concentration, and 6)
Samarium concentration.
1 1
i CALLAWAY - UNIT 1 3/4 1-3 Amendment No. 403,105
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REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR'0PERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be within the limits specified in the Core Operating Limits Report (COLR). The maximum upper limit shall be less positive than +5 pcm/*F for power levels up to 70% RATED THERMAL POWER and a linear ramp from that point to O pcm/'F at 1005 RATED THERMAL POWER for the all rods withdrawn, beginning of cycle life (BOL) condition.
APPLICABILITY:
Beginning of Cycle Life (BOL) Limit - MODES 1 and 2*#
End of Cycle Life (EOL) Limit - MODES 1, 2, and 3#
ACTION:
a.
With the MTC more positive than the BOL limit specified in the l
COLR, operation in MODES 1 and 2 may proceed provided:
1.
Control rod withdrawal limits are established and maintained sufficient to restore the MTC to within ti;9 BOL limits specified I
in the COLR within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT ST/'lDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. These withdrawal limits shall be 1.i addition to the insertion limits of Specification 3.1.3.6; I
2.
The control rods are maintained within tt.e 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 i
restoring the positive MTC to within its limit for the all rods withdrawn condition.
b.
With the MTC more negative than the EOL limit specified in the COLR,'
l be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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- With K,ff greater than or equal to 1.
i See Special Test Exception Specification 3.10.3.
CALLAWAY - UNIT 1 3/4 1-4 Amendment No. AA,58 i
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REf4CTIVITY CONTROL SYSTEMJ 3/4.I.3 MOVABLE CON 1ROL ASSEMBLIFJ GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full-length shutdown and control rods shall be OPERABLE and -
positioned within 112 steps (indicated position) of their group step counter demand position.
APPLICABILITY: MODES 1* and 2*.
FJ10ff:
The ACTION to be taken is based on the cause of rod inoperability as follows:
l CAUSE OF INOPERABILITY ACTION More Than One Rod One Rod 1.
One or more rods untrippable.
(a)
(a) l 2.
Misaligned by more than fl2 steps (indicated (c)
(b) position) from its group step counter demand height or from any other rod in its group.
ACTION a - 1.1 Determine that the SHUTDOWN MARGIN is greater than or equal i
to 1.3% Ak/k, with an increased allowance for the withdrawn worth of the untrippable rod (s), within I hour, or i
1.2 Initiate boration to restore the SHUTDOWN MARGIN to greater than or equal to 1.3% Ak/k, within I hour; and 1
2.
Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION b - Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION c - POWER OPERATION may continue provided that within I hour:
1.
The rod is restored to OPERABLE status within the above alignment requirements, or
'* See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
CALLAWAY - UNIT 1 3/4 1-14 Amendment No. 51,103,105 i.
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REACTIVITY CONTROL SYSTEMS i
j-LIMITING CONDITION FOR OPERATION i
ACTION (Continued) j 2.
The rod is. declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within il2 7
steps of the inoperable rod while maintaining the rod sequence j
and insertion limits of Specification 3.1.3.6.
The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or i
j 3.
The rod is declared inoperable and-the SH'JTDOWN MARGIN is greater than or equal to 1.3% Ak/k.
POWER OPERATION may then continue provided that:
}
a)
A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm i
that the previously analyzed results of these accidents remain valid for the duration of operation under these i
L conoitions; i
j b)
A power distribution map is obtained from the movable within their limits wil(h 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; andincore detectors and F Z) i i
c)
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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER.
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' SURVEILLANCE REOUIREMENTS 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 i
least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the rod position deviation monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
i-4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined to be OPERABLE (trippable) by movement of at least 10 steps in any l
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one direction at least once per 92 days.
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4.1.3.1.3 Prior to reactor criticality, the rod drop time of the individual full-length shutdown and control rods from the fully withdrawn position shall be demonstrated to be less than or equal to 2.7 seconds from the beginning of decay of stationary gripper coil voltage to dashpot entry with T,2 551*F and all reactor coolant pumps operating:
[
a.
For all rods'following each removal of the reactor vessel head, and b.
For specifically affected individual rods following any maintenance n
on or modification to the Control Rod Drive System which could
' affect the drop time of those specific rods.
i CALLAWAY - UNIT 1
-3/4 1-15 Amendment No. 51,50,103,105
TABLE 4.3-3
, RADIATION NONITORING INSTRUNENTATION FOR PLANT J
n OPERATIONS SURVEILLANCE REQUIREMENTS p,
g ANALOG g
CHANNEL
' ' CHANNEL' CHANNEL OPERATIONAL NODES FOR WHICH 8
FUNCTIONAL UNIT.
CHECK CALIBRATION TEST SURVEILLANCE IS REQUIRED
.C h'
l.
Containment
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a.
Gaseous Radioactivity-RCS Leakage Detection S
R Q-1, 2, 3, 4
-l (GT-RE-31 & 32) b.
Particulate Radioactivity -
S R
Q 1, 2, 3, 4 l
2 RCS Leakage Detection 2.
Fuel Building w1 a.
Fuel Building Exhaust -
l Gaseous Radioactivity-S R
Q w
1 i
High (GG-RE-27 & 28) b.
Criticality-High Radiation Level S
R Q
- 1) Spent Fuel Pool (50-RE-37 & 38)-
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- 2) New Fuel Pool S
R Q
(SD-RE-35 & 36)
. [
3.
Control Room
[
. Air Intake-Gaseous g
Radioactivity-High S-R Q
All l
(GK-RE-04 & 05) 2
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L 8
- With fuel in the respective fuel storage pool.
- With irradiated fuel in the fuel storage areas or fuel building.
l REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with at least two groups of backup pressurizer heaters each having a capacity of at least 150 kW and a water level of less than or equal to 92% (1657 cubic feet).
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
a.
With one group of backup pressurizer heaters inoperable, restore at least two groups of backup heaters to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, b.
With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the Reactor trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTD0L'N within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE RE_0VIREMENTS 4.4.3.1 The pressurizer water level shall be determined to be within its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by energizing the heaters and measuring circuit current at least once each refueling interval.
l CALLAWAY - UNIT 1 3/4 4-9 Amendment No. 105
1, REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES
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l tINITING CONDITION FOR OPERATION i
3.4.4 Aoth' power-operated relief valves (PORVs) and their associated block l
valves shall be OPERABLE.
APPLICABILITY: 800ES 1, 2, and 3.*
l ACTION:
. With one or both PORV(s) inoperable because of excessive seat leakage, l
a.
within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERA 8LE status or close i
the associated block valve (s)with Dower maintained in the block valve (s);
otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT l
SHi!TDOWN within the followino 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With one PORV inoperable due to causes other than excessive seat.
leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the P0RV to OPERA 8LE status, or i
close its associated block valve and remove power from the block valve; restore the PORV to OPERA 8LE status within the following l
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in NOT STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
l With both PORV(s) inoperable due to causes other than excessive seat c.
i leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore at least one PORV to OPERABLE status or clcse its ' associated block valve ~ and remove power from the block valve and be in NOT STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT SHUTDOWN within the fo110 win 9 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
d.
With one or both 61ock valves inoperable, within I hour restore the block valve (s) to OPERA 8LE status or place its associated PORV(s) in manual control.
Restore at least one block valve to OPERABLE status within the next hour if both valves are inoperable; restore any remaining inoperable block valve to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
The prowtsions of Specification 3.0.4 are not applicable.
a.
50RVEILLANCE REQUIREMENTS
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4.4.4.1 In5dition to the requirements of Specification 4.0.5, each ruMV shall be demonstrated OPERA 8LE at least once per 18 sonths by performance of a l
CilANNEL Call 8 RATION of the actuation instrumentation.
j 4.4.4.2 Each block valve shall be demonstrated OPERA 8LE at least once per 92 days ty operating the valve through one complete cycle of full travel unless the block valve is :1csed in'o'rder to meet the requirements of ACil0N b. or c. in Specification 3.4.4.
d J
- With all RCS cold leg temperatures above 368'F.
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CALLAWAY - UNIT 1 3/4 4-10 Amendment No.83 1
OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
1 a.
No PRESSURE BOUNDARY LEAXAGE, b.
I gpm UNIDENTIFIED LEAKAGE.
c.
1 gpm total reactor-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator, d.
10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, e.
8 gpm per RC pump CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 + 20 psig, and f.
The leakage from each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be limited to 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm, at a Reactor Coolant System pressure of 2235 1 20 psig.*
2
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APPLICABILITY: MODES 1. 2, 3, and 4.
4 ACTION:
a.
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from i
Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
c.
With any Reactor Coolant System Pressure Isolation Yalve leakage greater than the above limit, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or be in at least HOT STANDBY within the
~next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with an RCS pressure of less than C00 psig.
- Test pressures less than 2235 psig but greaten than 150 psig are allowed.
Observed ledkage shall be adjusted for the actual test pressure up to 2235 psig assuming the leakage to be directly proportional to ' pressure differential to the one-half power.
i CALLAWAY, UNIT 1 3/4 4-19 Amendment No. 66
I REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
a.
Monitoring the containment atmosphere gaseous or particulate radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; b.
Monitoring the containment normal sump inventory and discharge at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; c.
Measurement of the CONTROLLED LEAKAGE from the reacter coolant pump seals when the Reactor Coolant System pressure is 2235 i 20 psig at least once per 31 days. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4; d.
Performance of a Reactor Coolant System water inventory balance at r
least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; and e.
Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:
a.
At least once per 18 months; b.
Prior to entering MODE 2 whenever the unit has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been l
performed in the previous 9 months; c.
Prior to returning the valve to service following maintenance, repair or replacement work on the valve; and d.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to automatic or manual action or flow through the valve except for valves BBPV8702 A/B and EJHV8701 A/B.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.
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L CAO.AWAY.-UNIT 1 3/4 4-20 Amendment No. 105 i
n.
lw s'
CONTAINMENT SYSTEMS l
3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS i
CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Containment Spray System capable of taking suction from the RWST and transferring suction to the containment sump, i
APPLICABILITY: MODES 1, 2, 3, and 4.
)
ACTION:.
With one Containment Spray System incperable, restore the inoperable Containment Spray System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the inoperable Containment Spray System to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS c 4.6.2.1 Each Containment Spray System 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 that is not locked, sealed, or otherwise secured in position, is in its correct position; b.
By verifying, that on recirculation flow, each pump develops a discharge pressure of greater than or equal to 250 psig when tested pursuant to Specification 4.0.5; c.
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 Containment Pressure-High-3 (CSAS) test signal, and i
2)# Verifying that each spray pump starts automatically on a Containment Pressure-High-3 (CSAS) test signal.
d.
At least once per 10 years by performing an air or smoke flow test l
through each spray header and verifying each spray nozzle is unobstructed.
- The specified 18 month frequency may be waived for Cycle I provided the surveillance is performed prior to restart following the first refueling outage or June 1,1986, whichever occurs first. The provisions of Specification 4.0.2 are reset from performance of this surveillance.
CALLAWAY - UNIT 1-3/4 6-13 Amendment No. 8,105 -
i
_ _. _ _ ~
i CONTAINMENT SYSTEMS HYDROGEN CONTROL SYSTEMS 4
r LIMITING CONDITION FOR OPERATION 3.6.4.2 A Hydrogen Control System shall be OPERABLE with two independent Hydrogen Recombiner Systems, i
t,PPLICABILITY:- MODES 1 and 2 ACTION:
1 i
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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
a-1-
SVRVEILLANCE RE0VIREHENTS i
4.6.4.2 Each Hydrogen Recombiner System shall be demonstrated OPERABLE:
)
a.
At-least once each refueling interval by verifying, during a Hydrogen 1
Recombiner System functional test, that the heater air temperature 4
increases to greater than or equal to 1150*F within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />; and i
}
b.
At least once each refueling interval by:
l 1)
Performing a CHANNEL CALIBRATION of all Hydrogen Recombiner l
System instrumentation and control circuits, t
2)
Verifying through a visual examination that there is no evidence of abnormal conditions within the Hydrogen Recombiner System 1
enclosure (i.e., loose wiring or structural connections, j
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.
4 4
CALLAWAY - UNIT 1 3/4 6-32 Amendment No.105 i
~
REACTIVITY CONTROL SYSTEMS
-BASES CORE REACTIVITY (Continued)
Following evaluations of the core design and safety analysis, the cause of the reactivity anomaly may be resolved.
If the cause of the reactivity anomaly is a mismatch in core conditions at the time of RCS boron
)
concentration sampling, then a recalculation of the RCS boron concentration requirements may be performed to demonstrate that core reactivity is behaving j
as expected.
If an unexpected physical change in the condition of the core has occurred, it must be evaluated and corrected, if possible.
If the cause of the reactivity anomaly is in the calculation technique, then the calculational models must be revised to provide more accurate predictions.
If any of these results are demonstrated, and it is concluded that the reactor core is acceptable for continued operation, then the boron letdown curve may i
be renormalized and power operation may continue.
If operational restrictions or additional surveillance requirements are necessary to ensure the reactor l
l core is acceptable for continued operation, then they must be defined.
The required completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is adequate for preparing whatever operating restrictions or surveillances that may be required to allow continued reactor operation.
3/4.1.3 MOVABLE CONTROL ASSEMBLILS
{
The specifications of this section ensure that: (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is j
maintained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within 112 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 assurance that the Digital Rod Position Indicator is operating correctly over the full range of indication. Since the Digital Rod Position System does not indicate i
the actual shutdown rod position between 18 steps and 210 steps, only points in the indicated ranges are picked for verification of agreement with demanded position. Shutdown and control rods are positioned at 225 steps or higher for fully withdrawn.
For purposes of determining compliance with Specification 3.1.3.1, l
untrippability of any control or shutdown rod invokes ACTION statement 3.1.3.1.a.
The rod is considered trippable if the rod was demonstrated OPERABLE during the last performance of Surveillance Requirement 4.1.3.1.2 and met the rod drop time criteria during the last performance of Surveillance Requirement 4.1.3.1.3.
Exercising each individual rod every 92 days provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped.
The 92 day frequency takes into consideration other information available to the' operator in the control rone and SR 4.1.3.1.1, which is performed more frequently and adds to the dete.ination of OPERABILITY of-the rods.
Between required performances B 3/4 1-4 Amendment No. 29,44,51,05,103,105 CALLAhAY - UNIT 1 i
\\e.
l' REACTIVITY CONTROL SYSTEMS
. BASES MOVABLE CONTROL ASSEMBLIES (Continued) 1 i
of SR 4.1.3.1.2 (determination of rod OPERABILITY by movement), if a rod (s) is discovered to be immovable, but remains trippable and aligned, the rod (s) is considered to be OPERABLE. At any time, if a rod (s) is immovable, a determination of the trippability (OPERABILITY) of the rod must be made, and i
appropriate action taken.
The ACTION statements which permit limited variations from the basic
- requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER.
These restrictions provide assurance of fuel rod integrity during continued operation.
In i
addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.
.The power reduction and shutdown time limits given in ACTION statements i
3.1.3.2.a.2, 3.1.3.2.b.2, and 3.1.3.2.c.2, respectively, are initiated at the l
time of discovery that the compensatory actions required for POWER OPERATION can no longer be met.
.The maximum rod drop time restriction is consistent with the assumed rod l
drop time used in the safety analyses. Measurement with T greater than or equal to 551*F and with all reactor coolant pumps operating,, ensures that the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.
Control rod positions and OPERABILITY of the rod position indicators are i
required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is i
inoperable. These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.
1 I
h CALLAWAY. UNIT I B 3/4 1-5 Amendment No. 51,103,105
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