ML20216C611
| ML20216C611 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 04/02/1998 |
| From: | Westreich B NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20216C617 | List: |
| References | |
| NUDOCS 9804140514 | |
| Download: ML20216C611 (9) | |
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4 UNITED STATES g
g NUCLEAR REGULATORY COMMISSION
't WASHINGTON, D.C. 20555 0001
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j UNION ELECTRIC COMPANY CALLAWAY PLANT UNIT 1
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l DOCKET NO. 50-483 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.124 License No. NPF-30 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Callaway Plant Unit 1 (the facility) Facility Operating License No. NPF-30 filed by the Union Electric Company (the Company),
dated October 17,1997, as supplemented by letters dated March 3,1998, and March 17,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; I
B.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance: (i) 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.
The issuance of this license 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 Specifications 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:
9804140514 980402 PDR ADOCK 05000483 P
1 Y:
2 (2) Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through
'i Amendment No.124 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The i,ansee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance to be implemented within 30 days from the date ofissuance.
FOR THE NUCLEAR REGULATORY COMMISSION Y@(b
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Barry C. Westreich, Project Manager Project Directorate IV-2 Division of Reactor Projects lil/IV Office of Nuclear Reactor Regulation -
Attachment:
Changes to the Technical Specifications Date of issuance: April 2, 1998
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~ ATTACHMENTTO LICENSE AMENDMENT NQ.124 FACILITY OPERATING LICENSE NO. NPF-30 DOCKET NO. 50-483 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised _ pages are identified by Amendment number and contain marginallines indicating the areas of change. The corresponding overleaf pages are also provided to q
maintain document completeness.
REMOVE INSERT 3/4 4-30 3/4 4-30 3/4 4-31 3/4 4-31 3/4 4-36 3/4 4-36 B 3/4 4-7 8 3/4 4-7 8 3/4 4-8 B 3/4 4-8
' B 3/4 4-15 B 3/4 4-15 B 3/4 4-16 B 3/4 4-16 B 3/4 5-2 B 3/4 5-2 l
1
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REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE t!MITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit Ibes shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
a.
A maximum heatup of 100*F in any 1-hour period.
b.
A maximum cooldown of 100'F in an'y 1-hour period, and c.
A maximum temperature change of less than or egal to 10*F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
APPLICABILITY: At all times.
ACTION:
i j
Wit h any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perfore an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least 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 reduce the RCS T,,, and pressure to less than 200*F and 500 psig, respectively, 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 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 einutes during system heatup, cocidown, and inservice leak and hydrostatic testing operations.
4.4.9.1.2 The reactor vessel material irradiation survefilance specti 2ns shall be removed and examined, to detersins changes in anterial proper'.ies, as reoutred by 10 CFR Part 50 Appendix H.
The results of these examina-l tions shall be used to update Figures 3.4-2, 3.4-3, and 3.4-4.
CALLAWAY - UNIT 1 3/4 4-29 Amendment No. 76 S
MATERIAL Mt0PERTY BASIS s
UMITING MATERIAL LOWER SHELL PLATE N3 LIMmNG ART VALUES AT 20 EFPY:
1MT,100.47 3MT,84.2 Y 2500 l
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^ 2250 ll me I
em
" 2000 - -
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1750 unices,rist W
OP: RATION I
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4eerpring:
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OPERATION 6
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1000 y'
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c o o i. s o w n ll s
750 - -
i natas 85
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4 rh,
u ll 500_
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0 0
50 100 150 200 '250 300 350 400 450 500 Indicated Temperature (Deg.F)
FIGURE 3.4-3 Callaway Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown i
Rates of 0. 20, 40, 60 and 100'F/hr) Applicable for the First 20 EFPY (With Margins for Instrumentation Errors) Includes Vessel flange requirements of 170*F and 561 psig per 10 CFR 50. Appendix G CALLAWAY - UNIT 1 3/4 4-31 Amendment No. 36-74.124
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:
a.
Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the P0RV is required OPERABLE; b.
Performance of a CHANNEL CALI8 RATION on the PORV actuation channel at least once per 18 months; and c.
Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORV is being used for overpressure protection.
4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection as follows:
a.
For RHR suction relief valire 8708B:
By verifying at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that RHR RCS suction isolation valves (RRSIV) EJ-HV-8701B and 8B-PV-87028 are open.
b.
For RHR suction relief valve 8708A:
By verifying at least once.per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that RRSIV EJ-HV-8701A and BB-PV-8702A are open.
c.
Testing pursuant to Specification 4.0.5.
4.4.9.3.3 With the RCS vented, verify the vent pathway at least i
once per 31 days when the pathway is provided by a valve (s) that t
is locked, sealed, or otherwise secured in the open position; otherwise, verify the vent pathway every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
CALLAWAY - UNIT 1 3/4 4-35 AmeAtleert No. ft,83
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)
- 2. These limit lines shall be calculated periodically using methods provided below.
- 3. System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure vessel Code. Section,XI.
l Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature. RT effective full power years (EFPY)'of service life.
die.attheendof20 20 EFPY service life period is chosen such that the limiting RT at the 1/4T location in the core 1
region 1s greater than the RT of the-limYting unirradiated material.
The selection of such a limiting k assures that all components in the Reactor
~ Coolant System will be operated c,onservativeiy in accordance with applicable Code requirements.
The reactor vessel materials have been tested to determine their initial RT,: tion and resultant fast neutron (E greater than 1 MeV) irradiation can the results of these tests are shown in Table B 3/4.4-1. Reactor opera cause an increase in the RT Therefore, an adjusted reference temperature.
7 based upon the fluence and c,opper content and nickel content of the material l
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in question, can be predicted using Figure B 3/4.4-1 and the largest value of l
computed by either Regulatory Guide 1.99. Revision 2. " Effects of ART *IualElementsonPredictedRadiationDamagetoReactorVesselMaterials."
Reslc i
or the Westinghouse Copper Trend Curves shown in Figure B 3/4.4-2. The heatup l
and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT,the pressure and temperature sensing
_l at the end of 20 EFPY as well as adjustments for possible errors in instruments.
Capsules are removed in accordance with the requirements of ASTM l
E185-73 and 10 CFR Part 50. Appendix H.
The lead factor represents the i
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CALLAWAY - UNIT 1 B 3/4 4-7 Amendment No. 35.75.103.124
l i
BASES HEATUP (Continued)
The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.
Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.
The OPERABILITY of two PORVs, two RHR suction relief valves, one RHR suction relief valve and one PORV. or an RCS vent opening of at least 2 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 368*F.
Either PORV or either RHR suction relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either:
(1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50 F above the RCS cold leg temperatures, or (2) the start of a centrifugal charging pump and/or the normal charging pump and its injection j
into a water-solid RCS.
4 In addition to opening RCS vents to meet the requirement of Specification 3.4.9.3c.
it is acceptable to remove a 3ressurizer Code safety valve, open a PORV block valve and remove power from tie valve operator in conjunction with
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disassembly of a PORV and removal of its internals, or otherwise open the RCS.
COLD OVERPRESSURE The Maximum Allowed PORV Setpoint for the Cold Overpressure Mitigation System (COMS) is derived by analysis which models the performance of the COMS assuming various mass input and heat input transients. Operation with a PORV setpoint less than or equal to the maximum setpoint ensures that Appendix G criteria will not be violated with consideration for 1) a maximum pressure overshoot beyond the PORV setpoint which can occur as a result of time delays in signal processing and valve opening: 2) a 50*F heat transport effect made CALLAWAY - UNIT 1 B 3/4 4-15 Amendment No. 42,83.103.124 i
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l EMERGENCY CORE COOLING SYSTEMS BASES l
ECCS SUBSYSTEMS (Continued)
The limitation for one centrifugal charging pump and the normal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging i
pumps except the required OPERABLE charging Jump to be inoperable in MODES 4 and 5 and in MODE 6 with the reactor vessel lead on. 3rovides assurance that a mass addition pressure transient can be relieved by tie operation of a single PORV or RHR suction relief valve.
In addition, the requirement to verify all Safety Injection pumps to be inoperable in MODE 4. in MODE 5 with the water level above the top of the reactor vessel flange, and in MODE 6 with the reactor vessel head on and with the water level above the top of the reactor vessel flange. 3rovides assurance that the mass addition can be relieved by a single PORV or lHR suction relief valve.
I With the water level not above the top of the reactor vessel flange and I
with the vessel head on. Safety Injection pumps may be available to mitigate j
the effects of a loss of decay heat removal during partially drained conditions.
The Surveillance Requirements, which are provided to ensure the OPERABILITY of each component, ensure that, at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained.
The safety analyses make assumptions with respect to:
(1) both the maximum and minimum total system resistance. (2) both the maximum and minimum branch l
injection line resistance, and (3) the maximum and minimum ranges of potential pump performance.
These resistances and ranges of pump performance are used to calculate the maximum and minimum ECCS flows assumed in the safety analyses.
j The centrifugal charging pump minimum flow Surveillance Requirement provides the absolute minimum injected flow assumed in the safety analyses.
The maxims total system resistance defines the range of minimum flows l
(including the minimum flow Surveillance Requirement), with respect to pump head, that is assumed in the safety analyses.
The chargingpumptejalsystemresistance((Pf,c,U0,,refore,thecentrifugal
) must not be greater than o
i 1.004E-02 ft/gpm. @ ere P is pump discharge pressure in feet. P,c,is RCS o
pressure in feet, and 0, is the total pump flow rate in gpm.
The safety injection pump minimum flow Surveillance Requirement 3rovides the absolute minimum injected flow assumed in the safety analyses.
T1e maximum total system resistance defines the range of minimum flows (including the minimum flow Surveillance Requirement), with respect to pump head, that is assumed in the safety analyses Therefore,thesafetyinjectionpumptotgl where P is pump disc $ar,c )/0o,.
system resistance ((P -P
) must not be greater than 0.423E-02 ft/gpm.
l ge pressure in feet. P,c,is RCS pressure in feet, and l
o l
0, is the total pump flow rate in gpm.
I CALLAWAY - UNIT 1 B 3/4 5-2 Amendment No. '2. ".68.124