ML20101B850
| ML20101B850 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 03/11/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20101B856 | List: |
| References | |
| NUDOCS 9603150095 | |
| Download: ML20101B850 (9) | |
Text
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t UNITED STATES i
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NUCLEAR REGULATORY COMMISSION le
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WASHINGTON, D.C. Sam aaa1 j
l UNION ELECTRIC COMPANY CALLAWAY PLANT. UNIT 1 i
DOCKET NO. 50-483 i
AMENDMENT TO FACILITY OPERATING LICENSE l
Amendment No. 108 License No. NPF-30 1.
The Nuclear Regulatory Commission (the Commission) has found that:
l-i A.
The application for amendment by the Union Electric Company (licensee) dated December 9,1994, as supplemented by letters dated September 13, 1995, and February 9, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as i
amended (the Act) and the Commission's regulations set forth in 10 4
CFR Chapter I; i
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
Th'e issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-30 is hereby amended to read as follows:
9603150095 960311 PDR ADOCK 05000483 P
1 (2) Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.108 and the Environmental Protection Plan l
contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license. UE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This amendment is effective as of its date of issuance to be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION bA 4
.d 4%
Kristine M. Thomas, Project Manager Project Directorate IV-2 i
Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 11,1996 i
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ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.108 TO FACILITY OPERATING LICENSE NO. NPF-30 DOCKET NO. 50-483 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE INSERT 3/4 3-13 3/4 3-13 3/4 7-4a 3/4 7-4a 3/4 7-5 3/4 7-5 B 3/4 7-2 B 3/4 7-2
INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.
APPLICABILITY: As shown in Table 3.3-3.
ACTION:
a.
With an ESFAS Instnmentation 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.
With an ESFAS Instruentation or Interlock Trip Setpoint less 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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2-1 was satisfied for the affected channel, or 2.
Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
Equation 2.2-1 Z + R + 5 s TA Where:
Z = The value from Column Z of Table 3.3-4 for the affected
- channel, i
R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the 'as measured" value (in percent span) of the 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.
c.
With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Tame 3.3-3.
SURVEILLANCE REOUIREMENTS 4.3.2.1 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 (the provisions of Specification 4.0.4 are not applicable for the steam turbine-driven auxiliary feedwater pump for entry into MODE 3.
The ENGINEERED SAFETY FEATURES RESPONSE TIME testing for the steam turbine-driven auxiliary feedwater pump is required to be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 2900 psig in all steam generators).
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 at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3.
CALLAWAY - UNIT 1 3/4 3-13 Amendment No. 404,108
TABLE 3.3-3 g
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION f-g MINIMUM 2
TOTAL NO.
CHANNELS CHANNELS APPLICABLE 7
FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION E
- 1. Safety injection (Reactor Trip, M
Phase "A" Isolation Feedwater Isolation, Component Cooling
~
Water Turbine Trip Auxiliary
[
Feedwater-Motor-Driven Pump, Emergency Diesel Generator i
Operation, Containment Cool-ing, and Essential Service WaterOperation) i a.
Manual Initiation 2
1 2
1,2,3,4 18 1
b.
Automatic Actuation L
I Logic and Actuation I
wy Relays (SSPS) 2 1
2 1,2,3,4 14 c.
Containment Pressure-High-1 3
2 2
1,2,3 33*
d.
Pressurizer i
i Pressure - Low 4
2 3
1, 2, 3f 33*
i e.
Steam Line Pressure-i Low 3/ steam line 2/ steam line 2/ steam ifne 1, 2, 3#
33*
l g
any steam g
line s
t
[
?
a.
Manual Initiation 2 pair 1 pair 2 pair 1, 2, 3, 4 18 g
operated simultaneously b.
Automatic Actuation 2
1 2
1,2,3,4 14 Logic and Actuation Relays (SSPS) c.
Containment Pressure-4 2
3 1,2,3 16 High-3
e PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION ACTION: Continued e.
With two auxiliary feedwater pumps inoperable, 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 HOT SHUTDOW within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
f.
With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.
SURVEILLANCE REOUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:
a.
At least once per 31 days by:
1)
Verifying that each non-automatic valve in the flow path that 1s not locked, sealed, or otherwise secured in position is in l
its correct position; and 2)
Verifying that each automatic valve, other than Al-HV-30, 31, 32, and 33, in the flow path is in the fully open position whenever the Auxiliary Feedwater System is placed in automatic control or when above 10% RATED THERMAL POWER.
b.
At least once per 92 days on a STAGGERED TEST BASIS by:
l i
1)
Verifying that each motor-driven pump develops a discharge pressure of greater than or equal to 1535 psig on recirculation flow when tested pursuant to Specification 4.0.5; and l
2)
Verifying that the steam turbine-driven pump develops a i
1 discharge pressure of greater than er equal to 1625 psig at a j
flow of greater than or equal to 120 gpm (required to be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining ;t900 psig in all steam generators - the provisions of Specification 4.0.4 are not applicable for entry into MODE 3).
CALLAWAY - UNIT 1 3/4 7-4a Amendment No. 66,108
O PLANT SYSTEMS O
SURVEILLANCE REQUIREMENTS (Continued)
At least once per 18 months during shutdown by:
1 c.
1)
Verifying that automatic valves AL-HV-30, 31, 32, and 33 in the ESW supply to the auxiliary feedwater pumps actuate to their full open position upon receipt of an Auxiliary Feedwater Pump Suction Pressure-Low test signal, 2)
Verifying that each motor-driven auxiliary feedwater pump i
starts as designed automatically upon receipt of an Auxiliary Feedwater Actuation test signal, and 3)
Verifying that each auxiliary feedwater motor-operated discharge valve limits the flow to each steam generator from the motor-driven pump to less than or equal to 320 gpe.
d.
At least once per 18 months by verifying that the steam turbine-driven auxiliary feedwater pump starts as designed automatically upon receipt of an Auxiliary Feedwater Actuation Test signal (required to be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 2900 psig in all steam generators - the provisions of Specification 4.0.4 are not applicable for entry into MODE 3).
4.7.1.2.2 An auxiliary feedwater flow path shall be demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days prior to entering MODE 2 by verifying normal flow to at least two steam generators from one auxiliary feedwater pump.
CALLAWAY - UNIT 1 3/4 7-5 Amendment No. &&,108
~
0 3/4.1 PLANT SYSIENS l
o BASES 3/4.7.1 TURBINE CYCLE i
3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that the Secondary Coolant System pressure will be limited to within 110% (1320 psia) of its design pressure of 1200 psia during the most severe anticipated system l
operational transient. The maximum relieving capacity is associated with a Turbine trip from 102% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).
The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code. (1971 Edition). The total relieving capacity for all valves on all of the steam lines is (18.28 x 105) Ibs/h which is 115% of the total secondary steam flow of 15.85 x 105 lbs/h at 102% RATED THERMAL POWER. A minimum of two OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in I
i Table 3.7-2.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable i
within the limitations of the ACTION requirements on the basis of the reduction in Secondary Coolant System steam flow and THERMAL POWER required by the reduced Reactor Trip Settings of the Power Range Neutron Flux channels. The Reactor Trip Setpoint reductions are derived on the following bases:
l For four loop operation:
SP = (X) - (Y)(V) x (109)
Where:
I SP = Reduced Reactor Trip Setpoint in percent of RATED THERMAL POWER, V = Maximum number of inoperable safety valves per steam line, 4
i CALLAWAY - UNIT 1 B 3/4 7-1
PLANT SYSTEMS e
BASES SAFETY VALVES (Continued) 109 - Power Range Neutron Flux-High Trip Setpoint for four loop operation, X=
Total relieving capacity of all safety valves per steam line in 1bs/ hour,and Y=
Maximum relieving capacity of any one safety valve in 1bs/ hour.
3 /4.7.1. 2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss-of-offsite power in conjunction with a feedwater line break to one steam generator.
Testing of each electric motor-driven auxiliary feedwater pump on a fixed orifice recirculation flow and ensuring a discharge pressure of greater than or equal to 1535 psig verifies the capability of each pump to deliver a total piamp flow of E75 gpm at a steam generator pressure of 1221 psig. Testing the steam turbine-driven auxiliary feedwater pump at greater than or equal to 120 gym and ensuring a discharge pressure of greater than or equal to 1625 psig verifies the capability of the pump to deliver a total pump flow of 1145 gpa at a steam generator pressure of 1221 psig. The capacity of any auxiliary feedwater pump is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System tesiperature to less than 350'F when the RHR System may be placed into operation.
3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STAND 8Y conditions for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> with steam discharge to the atmosphere concurrent with total loss-of-offsite power and then a cooldown to 350'F at 50*F per hour. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical chracteristics.
3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture. This dose also includes the effects of a coincident 1 gpm reactor to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.
4 CALLAWAY - UNIT 1 B 3/4 7-2 Amendment No. 108