ML20155F529
| ML20155F529 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  (NPF-03-A-122, NPF-3-A-122) |
| Issue date: | 09/30/1988 |
| From: | De Agazio A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20155F532 | List: |
| References | |
| NUDOCS 8810130287 | |
| Download: ML20155F529 (11) | |
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UNITED STATES 8
NUCLEAR REGULATORY COMMISSION o
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TOLECO EDISON COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 122 License No. NPF-3 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by the Toledo Edison Company and The Cleveland Electric Illuminating Company (the licensees) dated l
July 15, 1988, complies with the tandards and requirements of the Atomic Energy Act of 1954, as ame ded (t'e Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I 8.
The facility will operate in conformity with ths application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendaent can be conducted without endangering the health and safety of the public, and (ii) th6t such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requiremnts have been satisfied.
2.
Facility Operating License No. NPF-3 is hereby amended by changes to the Technical Specifications as indicated in the attachment to this license amendment; oy deleting paragraph 2.C.(3)(t); and by revising paragraph 2.C.(2)toreadasfollows:
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. (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.122, are hereby incorporated in the license. The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION John N. Hannon, Director Project Directorate III-3 Division of Reactor Projects - III, IV, V. & Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: September 30, 1988
ATTACHMENT TO LICENSE AMEN 0 MENT NO. 122 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.
Remove Insert 3/4 7-4 3/4 7-4 3/4 7-5 3/4 7-5 B3/4 7-la B3/4 7-la B3/4 7-2 B3/4 7-2 r
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Table 4.7-1 1
M I
b MAIN STEAM LINE SAFETY VALVE LIFT SETTINGS
- 4 i
l NIMBER PER STEAM GENERATOR LIFT SETTING (i 1%)*
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2 1050 psig i
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7 1100 psig I
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- The lift setting pressure shall correspond to ambient conditions of the vaive j
at nominal operating temperature and pressure.
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9 PLANT SYSTEMS l
AUXILIARY FEEDVATER SYSTEM I
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LIMITING CONDITION FOR OPERATION t
3.7.1.2 Two independent steam generator auxiliary feedvater pumps and associated flow paths shall be OPERABLE.
APPLICABILITY:
MODES 1, 2 and 3.
i I
ACTION:
r Vith one Auxiliary Feedvater System inoperable, restore the a.
inoperable 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 i
HOT SHUTDOVN vithin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.7.1.2 Each Auxiliary Feedvater System shall be demonstrated OPERABLE:
At least once per 31 days on a STAGGERED TEST BASIS by a.
i l
1.
Verifying that each steam turbine driven pump develops a i
differential pressure of > 1070 psid on recirculation flov vhen the secondary steam supply pressure is greater than 800 psia, as i
measured on PI SP 12B for pump 1-1 and PI SP 12A for pump 1-2.
The provisions of Specification 4.0.4 are not applicable for i
entry into Mode 3.
t 2.
Verifying that each valve (pover operated or automatic) in the i
flow path is in its correct position.
3.
Verifying that all manual valves in the auxiliary feedvater pump i
suction and discharge liner that affect the system's capacity to
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deliver vater to the steam generator are locked in their proper position.
l 4.
Verifying that valves CV 196, CV 197, FV 32, TV 91 end FV 106 t
i are closed.
b.
At least once per 18 months by:
1.
Verifying that each automatic valve in the flow path actuates to its correct position on an auxiliary feedvater actuation test signal.
2.
Verifying that each pump starts automatically upon receipt of an auxiliary feedvater actuation test signal.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
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DAVIS-BESSE, UNIT 1 3/4 7-4 Amendment No. 74, 122
l PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 3.
Verifying that there is a flow path between each auxiliary feedvater pump and each steam generator by pumping water from the Condensate Storage Tank to the steam generator.
The flow path to the steam generator shall be verified by l
either steam generator level charge or Auxiliary Feedvater Safety Grade Flov Indication.
Verification of the Auxiliary Feedvater System's flow capscity is not required.
c.
The Auxiliary Feed Pump Turbine Steam Generator Level Control System shall be demonstrated OPERABLE by performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATION at least once per 18 months.
d.
The Auxiliary Feed Pump Suction Pressure Interlocks, and Auxiliary Feed Pump Turbine Inlet Steam Pressure Interlocks shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATION at least once per 18 months, After any modification or repair to the Auxiliary Feedvater System e.
that could affect the system's capability to deliver water to the steam generator, the affected flov path shall be demonstrated available as follows:
1.
If the modification or repair is downstream of the test flow line, the auxiliary feed pump shall pump veter from the Condensate Storage Tank to the steam generatori and the flow path availability vill be verified by steam generator level change or Auxiliary Feedvater Safety Grade Flov Indication.
2.
If the modification or repair is upstream of the test flov line, the auxiliary feed pump shall pump water through the Auxiliary Feedvater System to the test flov line; and the flow path availability vill be verified by flov indication in the test flow line.
(see note belov)
This Surveillance Testing shall be performed prior to entering MODE 3 if the modification is made in MODES 4, 5 or 6.
Verification of the Auxiliary Feedvater System's flow capacity is not required.
Note:
When conducting tests of the Auxiliary Feedvater System in MODES 1, 2, and 3 vhich require local manual realignment of valves that make the system inoperable, a dedicated individual shall be stationed at the valves (in communication with the control room) able to restore the valves to normal system OPERABLE status.
DAVIS-BESSE, UNIT 1 3/4 7-5 Amendment No. 4Y,44,44, 122
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PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) f.
Following each ext.'nded cold shutdown (> 30 days in MODE 5), by:
1.
Verifying 1 st there is a flow path between each auxiliary feedvater pump and each steam generator by pumping Condensate Storage Tank vater to the steam generator.
The flow path to the steam generator shall be verified by either steam generator level chante or Auxiliary Feedvater Safety Grade Flov Indication.
l Verification of the Auxiliary Feedvater System's flow capacity is not required.
l i
l i
DAVIS-BESSE, UNIT 1 3/4 7-5a Amendment No. 96
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3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure vill be limited to within 110% its design pressure of 1050 psig during the most severe anticipated system operational transient.
The maximum relieving capacity is associated with a turbine trip from 100%
RATED THERHAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).
The safety valve set pressures and relieving capacities are in accordance with l
Section III of ASME Boiler and Pressure Vessel Code, 1971 Edition. The Code requires the following:
1.
At least two pressure-relief valves are required to provide relieving capacity for steam systems.
2.
The capacity of the smallest pressure-relief valve shall not be less than 50 percent of that of the largest pressure-relief device.
3.
The set pressure of one of the pressure-relief devices shall not be greater than the maximum allovable working pressure of the system at design temperature.
4.
Total rated relieving capacity of the pressure-relief devices shall prevent a rise of more than 10 percent above system design pressure at design temperature under any pressure transients anticipated to arise.
l These requirements are, respectively, met as follows:
i 1.
Nine safety valves are installed per steam generator.
2.
The relief capacity of two of the nine safety valves per steam generator is 583,574 lbs/hr each, and the capacity of the remaining seven is 845,759 lbs/hr each.
3.
A minimum of two OPERABLE safety valves per steam generator, with a combined total relief capactty of at least 1,167,148 lbs/hr, one with a setpoint not greater than 1050 psig (+/-1%), and one with a setpoint not greater than 1100 psig (+/-1%).
4.
The total relieving capacity of all safety valves on both main steam lines is 14,175,000 lbs/hr which is 120 percent of the total secondary system flow of 11,7'0,000 lbs/hr at 100 percent of rated thermal power.
A maximum safety valve setpoint pressure of 1100 psig (+/-1%)
assures main steam system pressure remains belov 110 percent, or 1155 psig.
DAVIS-BESSE, UNIT 1 B 3/4 7-1 Amendment No. 48, 117
l 3/4.7 PLANT SYSTEMS BASES 3/4.7.1.1 SAFETY VALVES (Continued)
STARTUP and/or POVER OPERATION is allovable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flov and THERMAL POVER required by the reduced reactor trip settings of the High Flux channels.
The reactor trip setpoint reductions are derived on the follvoing bases:
SP = (X) - (Y)(V) x y where:
SP = reduced Trip Setpoint in percent of RATED THERMAL POVER (Not to Exceed V)
V = maximum number of inoperable safety valves per steam generator V = High Flux Trip Setpoint for four pump operation as specified in Table 2.2-1 X = Total relieving capacity of all safety valves per steam generator in lbs/ hour, 7,087,500 lbs/ hour Y = Maximum relieving capacity of any one safety valve in lbs/ hour, 845,759 lbs/ hour Z = Required relieving capacity per steam generator in lbs/hr, 6,585,600 lbs/hr.
3/4.7.1.2 AUXILIARY FEEDVATEP SYSTEMS j
Tho OPERABILITY of the Auxiliary Feedvater Systems ensures that the Reactor Coolant System can be cooled dovn to less than 280*F from normal operating conditions in the event of a total loss of offsite power.
Each steam driven auxiliary feedvater pump is capable of delivering a total feedvater flow of 600 gpm at a pressure of 1050 psig to the entrance of the steam generatote.
This capacity is sufficient to ensure that adequate feedvater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 280'F vhere the Decay Heat Removal System may be placed in operation.
Next page is B 3/4 7-2 DAVIS-BESSE, UNIT 1 B 3/4 7-la Amendment No. YY7.122
PLANT SYSTEMS BASES 3/4.7.1.2 AUXILIARY FEEDVATER SYSTEMS (Continued)
Following any modifications or repairs to the Auxiliary Feedvater System piping from the Condensate Storage Tank through auxiliary feed pumps to the steam generators that could affect the system's capability to deliver vater to the steam generators, following extended cold shutdovn, a flow path verification test shall be performed. This test may be conduwted in H0 DES 4, 5 or 6 using auxiliary steam to drive the auxiliary feed pumps turbine to demonstrate that the flow path exists from the Condensate Storage Tank to the steam genera'.scs via auxiliary feed pumps.
Verification of the turbine plant cooling vater valves (CV 196 and CV 197),
the startup feedvater pump suction valves (FV 32 and FV 91), and the startup feedvater pump discharge valve (FV 106) in the closed position is required to address the concerns associated with potential pipe failures in the auxiliary feedvater pump rooms, that could occur during operation of the startup feedvater pump.
3/4.7.1.3 CONDENSATE STORAGE FACILITIES The OPERABILITY of the Condensate Storage Tank vith the minimum vater volume ensures that sufficient vater is available to maintain the RCS at HOT STANDBY conditions for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> with steam discharge to atmosphere and to cooldovn the Reactor Coolant System to less than 280'F in the event of a total loss of offsite power or of the main feedvater system.
The contained vater volume limit includes an allowance for vater not usable because of tank discharge line location or other physical characteristics.
3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant offsite radiation dose vill be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose includes the effects of a coincident 1.0 GPM primary 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.
3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator vill blovdown in the event of a steam line rupture.
This restriction is required to 1) minimize the DAVIS-BESSE, UNIT 1 B 3/4 7-2 Amendment No. 58, 96, Rev. 11/18/87
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