ML19305C577
| ML19305C577 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 03/04/1980 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19305C579 | List: |
| References | |
| NUDOCS 8003310113 | |
| Download: ML19305C577 (13) | |
Text
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jo UNITED STATES g
NUCLEAR REGULATORY COMMISSION a
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WASHINGTON D.C.20555
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VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 56 License No. DPR-32 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated October 14, 1977, as supplemented, 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; 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 (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 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 satisified.
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, 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to the license amendment, and paragraph 3.B of Facility Operating License No. DPR-32 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 56
, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION A. Schwencer, Chief Operating Reactors Branch No.1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: March 4, 1980 l
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VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 55 License No. DPR-37 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated October 14, 1977, as supplemented, 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; 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 (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 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.
I I
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to the license amendment, and paragraph 3.B of Facility Operating License No. DPR-37 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 55, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications -
Date of Issuance: March 4,1980 s
ATTACHMENT TO LICENSE AMENDMENT NOS.56 AND 55 FACILITY OPERATING-LICENSE NOS. DPR-32 AND DPR-37
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DOCKET NOS. 50-280 AND 50-281 1
x 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 area of change.
Remove Insert-3.1 -1 3.1-1 3.1-2 3.1-2 3.1-23 3.1-23 3.1-24~
3.1-25 4.1-8 4.1-8 i
4.1-9a 4
6.6-16a 6.6-16a t
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TS 3.1-1 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR COOLANT SYSTEM Applicability Applies to the operating status of the Reactor Coolant System.
Objectives To specify those limiting conditions for operation of the Reactor Coolant System which must be met to ensure safe reactor operation.
These conditions relate to: operational components, heatup and cooldown, I
leakage, reactor coolant activity, oxygen and chloride concentrations, minimum temperature for criticality, and reactor coolant system overpres-sure mitigation.
A.
Operational Components Specifications 1.
Reactor Coulant Pumps a.
A reactor shall not be brought critical with less than two pumps, in non-isolated loops, in operation.
b.
If an unscheduled loss of one or more reactor coolant pumps occurs while operating below 10% rated power (P-7) and results in Icss than two pumps in service, the affcceed Amendment No. Sf>. IJnit 1 Amentiment No. 55 linit 7
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plant shall be shutdoen and the reactor made subcritical The shutdown by inserting all control banks into the core.
rods may remain withdrawn, A minimum of one pump in a non-isoalted loop, or one c.
residual heat removal pump and its associated flow path, shall be in operation during reactor coolant boron concen-tration reduction.
Reactor power shall not exceed 50% of rated power with only d.
two pumps in operation unless the overtemperature AT trip setpoints have been changed in accordance with Section 2.3, af ter which power shall not exceed 60% with the inactive loop stop valves open and 65% with the inactive loop stop valves closed.
L When all three pumps have been idle for > 15 minutes, the e.
first pump shall not be started unless: (1) a bubble exists in the pressurizer or (2) the secondary water temperature of each steam generator is less than 50*F above each of the RCS I
cold leg temperatures.
2.
Steam Generator A minimum of two steam generators in non-isolated loops shall be operable when the average reactor coolant temperature is greater than 350*F.
1 3.
Pressurizer Safety valves One valve shall be operable whenever the head is on the a.
reactor vensel, except during hydrontatic tests.
Ami ntimenL fio. 'a f,, fin i t t Amendment No. 55, Unit 2 1
6.
T.S. 3.1-23 References (1) FSAR 4.2 (2) FSAR 9.2 C.
Reactor Coolant System Overpressure Mitigation Specification The Reactor Coolant system overpressure mitigating system shall.be 1.
operable as described briow.
Whenever the reactor coolant average temperature is greater than a.
350'F, a bubble shall exist in the pressurizer with the necessary sprays and heaters operable.
b.
Uhenever the reactor coolant average temperature is j[ 3,50*F and the reactor vessel head is bolted:
(1) A maximum of one charging pump operable.
(2) Two charging pumps shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the motor circuit breakers have been removed from their power supply or the benchboard control switch is in the " PULL-TO-LOCK" position.
(3) Two operable Power Operated Relief Valves (PORV) with a lif t setting of j[ 435 psig, or (4) A bubble in the pressurizer with a maximum pressurizer narrow range level of 33%. After a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, two PORV's must also be operable, or (5) The Reactor Coolant system vented through one opened PORV, or an equivalent size opening.
2.
The requirements of Specification 3.1.G.1.b may be modified as follows:
One PORV may be inoperable for a period not to exceed 7 days. If a.
the inoperabic PORV, is not restored to operable status withiu 7 days, then depressurize the RCS and open one PORV within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
i Amenilnent No. 56, Unit 1 Amonilmen t No. r, r,, Ilnit 7
T.S. 3.1-24 b.
With both PORV's inoperabic, depressurize the RCS uithin 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> unless specification 3.1.C.l.b.(4) is in ef fect. When the RCS has been depressurized, open one PORV or establish the conditions Maintain the RCS depressurized until both PORV's listed below.
have been restored to operable status.
(1) A maximum Pressurizer narrow range level of 33%.
(2) The series RHR inlet valves opened and their respective breakers locked open or an alternate letdown path operable.
(3) Limit charging flow to less than 150 gpm.
(4) Safety Injection accumulator discharge valves clos.ed and their respective breakers locked open.
When the conditions noted in 3.1.G.2.b.(1) through 3.1.C.2.b. (4) c.
above are required to be established, their implementation shall be' verified at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3.
In the event that the Reactor Coolant System Overpressure !!1tigating System is used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Speci-i fication 6,6.4 within 30 days. The report shall describe the circum-stances initiating the transient, the effect of the !!itigating system or the administrative cc.trols on the transient and any corrective actions necessary to prevent recurrence.
Basis The operability of two PORV's or the RCS vented through an opened PORV ensures that the Reactor Vessel will be protected from pressure trans-icnts which could exceed the limits of Appendix G to 10 CFR Part 50 when the Reactor Coolant average temperature is < 350*F and the Reactor Vessel llead bolted. When the Reactor Coolant averapc temperature is > 350*F overpressure protection is provided' by a bevole in the pressurizer and/or pressurtzer safety valves. A single p0RV has adequate relievinr.
Amendment No. 56, Unit 1 Amendment No. 55, finit 7
T.S. 3.1-25 capability to protect the Reactor Vessel from overpressurization when the transient is limited to either (1) the start of an idle Reactor Coolant Pump with the secondary t;ater temperature of a steam generator < 50*F above the RCS cold leg temperature or (2) the start of a charging pump and its injection into a water solid RCS.
The limitation for a maximum of one charging pump allowed operable and the surveillance required to verify that two charging pumps to be inoperable below 350*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV, or equivalent.
A maximum Pressurizer narrow range level of 33% has been selected to provide sufficient time, approximately 10 minutes, for operator response in case of a malfunction resulting in maximum charging flow from one charging pump (600 gpm). Operator action would be initiated by at least two alarms that h
would occur between the normal operating level and the maximum allowable level (33%). When both PORV are inoperable and it is impossible to manually open at least one PORV, additional administrative controis shall be implemented to prevent a pressure transient that would exceed the limits of Appendix C to 10 CFR Part 50.
The requirements of this specification are only applicable when the Reactor Vessel head is bolted. When the Reactor Vessel head is unbolted, a RCS pressure of < 100 psig will lift the head, thereby creating a relieving capability equivalent to at least one PORV.,
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Amendment flo. 56, tinit 1 l
Amendment flo. 55, tinit 7
TABLE 4.1-1 (Continued)
Channel Description-Check Calibrate Test Remarka 25.
Turbine First Stage Pressure S
R H
26.-
. Emergency Plan Radiation Instruments *M R
M 27.
Environmental Radiation Monitors
- M N.A._
N.A.
TLD Dosimeters
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28.-
Logic Channel Testing N.A.
N.A.
M 29.
Terbinc Overspeed Protection N.A.
R R
Trip Channel (Electrical) 30.
Turbine Trip Set Point N.A.
R R
Stop valve closure or low EN fluid pressure 31.
Seismic Instrumentation M
SA N
32.
Reactor Trip Breaker N.A.
N.A.
M 33.
Reactor Coolant Pressure (Low)
N.A.
R N.A.
l S - Each Shif t M - Ponthly P - Prior to each startup if not done previous week l
D - Daily W - Weekly R - Each Refueling Shutdown gg mm NA - Not applicable BW - Every two weeks EE SA - Seniannually AP - After each startup if not done previous week
$$ Q - Every 90 effective full power days 55
=
- See Specification 4.1D P.
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TABLE 4.1.2A (CONTINUED)
FSAR SECTION Kl'.lT.RI:llCE DESCRIPTION TEST FREQUENCY.
16.
Reactor vessel Overpressure Functiona1 & Setpoint Prior to decreasing RCS temperature None below 3500F and monthly while the.
14Jtigating System (except backup RCS is <3500F and the Reactor Vessel air supply)
Head is bolted.
17.
Reactor Vessel overpressure, Setpoint Refuellug Nonc lif tigation System llackup Kir Supply e
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TS 6.6-163 With no fire suppression water system operable, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; c.
notify the Commission outlining the action taken and the plans and schedule for restoring the system to operable status.
d.
With redundant fire suppression water system component inoperable for more than 14 days, submit a Special Report to the Com-,sion within the next 10 days outlining the cause of inoperability and the plans for restoring the component to operable statua.
fire protection system inoperable for more than 14 e.
With the CO2 days, submit a Special Report to the Commission within the next 10 days outlining the cause of inoperability and the plans for restoring the system to operable status.
f.
With the Records Vault halon fire protection system inoperable f
for more than 14 days, submit a Special Report to the Commission l
within the next 10 days outlining the cause of inoperability and the plans for restoring the system to cperable status.
In the event that the Reactor Vessel Overpressure Mitigating g.
System is used to mitigate a RCS pressure transient, submit a Special Report to the Commission within 30 days.,The report shall describe the circumstances initiating the transient, the effect of the PORVs or the administrative controls on the transient and any corrective action necessary to prevent recurrence.
Amendment No. 56, linit 1 Amendment. No. %, fini t. 7
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