ML19347F709
| ML19347F709 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 05/12/1981 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19347F710 | List: |
| References | |
| NUDOCS 8105260092 | |
| Download: ML19347F709 (16) | |
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UNITED STATES y y e q~ i NUCLEAR REGULATCRY COMMISSION
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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. 67 License No. DPR-32 1.
The Nuclear Regulatory Commissico (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated November 14, 1980, 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; l
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.
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.
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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. OpR-32 is hereby amended to read as follows:
(B) Tec'hnical Specifications The Technica1' Specifications contained in Appendices A and B, as revised through Amendment No. 67, 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.
FQR,THE NUCL AR REGULATORY COMMISSION J fi
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kl St;even A. Varga, Chis f Operating Reactors nch No. 1 Division of Licensi i
Attachnent:
Changes to the Technical Specifications
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Date of Issuance:
MAY 121981 e
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UNrrED STATES
{ 'Q,y(J',9, NUCLEAR REGULATORY COMMISSION 5.*,Mh C
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VIRGINIA ELECTRIC AND POWER COMPAN_Y_
DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY CPERATING LICENSE Amendment No. 67 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 November 14, 1980, complies with as amended (the Act)quirements of the Atomic Energy Act of 1954, the standards and re 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 provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of tk public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment v:ill not be inim1 cal 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.
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Accordingly, the. license is amended by changes to the Technical Specifications as indicated in the attachment to this license 4
amendment, and paragraph 3.B of Facility Operating License No. PPR-37 is hereby amended to read as follows:
(B) Tech'nical Specifications _
The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 67, 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 fb_
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Reven A. Varga,, h reff Operating Reacters B anch No. 1 Division of Licens n
Attachment:
Changes to the Technical Specifications Dat'e of Issuance:
MAY l 21981
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ATTACHMENT TO LICENSE AMENDMENTS AMEllDMENT NO. 67 TC FACILITY OPERATING LICENSE f!O. DPR-32 AMENDMENT NO. 67 TO FACILITY OPERATING LICENSE NO. DPR-37 I
DOCKET NOS. 50-280 AND 50-281 Revise Appendix A as follows:
Remove Pages Insert Pages 3.1 -1 3.1-i 3.1 -2 3.1-2 3.1 -2a 3.5-2 3.5-2 3.5-3 3.10-1 3.10-1 3.10-2 3.10-2 3.10-3 3.10-3 3.10-4 3.10-4 3.10-5 3.10-5 3.10-6 3.10-6 3.10-7 D
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s TS 3.1-1 3.0 DMTING 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, leakage, rea'etor coolant activity, oxygen and chloride concentrations, minimum temperature for criticality, and reactor coolant system overpres-sure mitigation.
A.
Operational Components Specifications 1.
Reactor Coolant Pumps 4
a.
A reactor shall not be brought critical with less than two pumps, in non-isolated loops, in operation.
1 Amendment No. 67 & 67
1 TS 3.1-2 b.
If an unscheduled loss of one or more reactor coolant pumps occurs while operating below 10% rated power (P-7) and 1
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- re3alts in less than two pumps in service, the affected i
plant shall be shutdown and the reactor made suberitical by inserting all control banks into the core. The shutdown 3-rods may remain withdrawn.
When the average reactor coolant loop temperature is greater c.
than 350*F, the following conditions shall be met:
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1.
At least two reactor coolant loops shall be operable.
2.
At least one reactor coolant loop shall be in operation.
d.
When the average reactor coolant loop temperature is less than or equal to 350*F, the following conditions shall be met:
1.
A minimum of two non-isolated loops, consisting of any combination of reactor coolant loops or residual heat removal loops, shall be operable, except as specified in Specification 3.10.A.6.
2.
At least one reactor coolant loop or one residual heat removal l'oop shall be in cperation, except as specified in Specification 3.10.A.6.
Amendment No. 57 & 57
l TS 3,1-2a Reactor power shall not exceed 50% of rated power with only e.
two pumps in operation unless the overtemperature AT trip I
setpoints have been changed in accordance with Section 2.3, after which power shall not exceed 60% with the inactive loop stop valves open and 65% with the inactive loop stop valves closed.
f.
When all three pumps have been idle for > 15 minutes, the fir'st 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 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.
3.
Pressurizer Safety Valves One valve shall be operable whenever the head is on the a.
reactor vessel, except during hydrostatic tests.
Amendment ib. 67 & $7
TS 3.5-2 1.
One residual heat removal pump may be out of service, provided immediate attention is directed to making repairs.
2.
One residual heat remova'l heat exchanger may be out of service, provided immediate attention is directed to making repairs.
Basis
- The Residual Heat Removal System is required to bring the Reactor Coolant System from conditions of approximately 350*F and pressures between 400 and 450 psig to cold shutdown conditions. Heat removal at greater temperatures is by the Steam and Power Conversion System. The Residual Heat Removal System is provided with two pumps and two heat exchangers.
If one of the two pumps and/or one of the two heat exchangers is not operative, safe operation of the unit is not affected;-however, the time for cooldown to cold shutdown conditions is extended.
The NRC requires that the series motorized valves in the line connecting I
the RHRS and RCS be provided with pressure interlocks to prevent them from l
opening when the reactor coolant system is at pressure.
References FSAR Section 9.3 - Residual Heat Removal System.
o knencment No. 67 & 67 w-e
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TS 3.10-1 l
3.10 REFUE;ING t
Applicability.
Applies to operatinf limitations during refueling operations.
Ojective To assure that no accident could occur during refueling operations that would affect public health and ' safety.
Specification A.
During refueling operations the following conditions are satisfied:
1.
The equipment door and at least one door in the personnel air lock shall be properly closed. For those systems which provide a direct path from containment atmosphere to the outside atmosphere, all automatic containment isolation valves in the unit shall be operable or at least one valve shall be closed in each line penetrating the containment.
2.
The Containment Vent and Purge System and the area and airborne radiation monitors which initiate isolation of this system, shall be tested and verified to be operable immediately prior to refueling cperations.
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TS 3.10-2 3.
At least one source range neutron detector shall be in service at all times when the reactor vessel head is unbolted. Whenever I
core, geometry or coolant chemistry is being changed, subcritical neutron flux shall be continuously monitored by at least two source range neutron detectors, each with continuous visual indi-cation in the Main Control Room and one with audible indication within the containment. During core fuel loading phases, there shall be a minimum neutron count rate detectable on two operating source range neutron detectors with the exception of initial core loading, at which time a minimum neutron count rate need be established only when there are eight (8) or more fuel assemblies loaded into the reactor vessel.
4.
Manipulator crane area radiation levels and airborne activity levels within the containment and airborne activity levelr in the ventilation exhaust duct shall be continuously monitored during refueling. A manipulator crane high radiation alarm or high airborne activity level alarm within the containment will automatically stop the purge venti-lation fans and automatically close the containment purge isolation valves.
5.
Fuel pit bridge area radiation levels and ventilation vent exhaust airborne activity levels shall be continuously monitored during refueling. The fuel build.ing exhaust will be continuously bypassed through the iodine filter bank during refueling procedures, prior to discharge through the ventilation vent.
Amendment No. 6' 2 67 a
TS 3.10-3 l
6.
At least one residual heat removal pump and heat exchanger shall be operable to circulate reactor coolant. The residual heat removal I
loop.may be removed from operation for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period during the performance of core alterations or reactor vessel surveil-lance. inspections.
7.
Two residual heat removal pumps and heat exchangers shall be operable to circulate reactor coolant when the water level above the top.of the reactor pressure vessel flange is less than 23 feet.
8.
At least 23 feet of water shall be maintained over the top of the reactor pressure vessel flange dur.ng movement of fuel assemblies.
9.
When the reactor vessel head is unbolted, a minimum boron concen-tration of 2,000 3pm shall be maintained in any filled portion of the Reactor Coolant System and shall be checked by sampling at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
10.
Direct communication between the Main Control Room and the refueling cavity manipulator crane shall be available whenever changes in core geometry are taking place.
11.
No movement of irradiated fuel in the reactor core shall be accomplished l
until the reactor has been suberitical for a period of at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
l Amendment No. 67 & 67 l
a TS 3.10-4 12.
A spent fuel cask or heavy loads exceeding '110 percent of the weight of a fuel assembly (not including fuel handling tool) shall not be I
moved over spent inel, and only one spent fuel assembly will be handled at one time over the reactor or the spent fuel pit.
13.
A spent fuel cask shall not be moved into the Fuel Building until such time as the NRC has reviewed and approved the spent fuel cask drop evaluation.
B.
If any one of the specified limiting conditions for refueling is not met, refueling of the reactor shall cease, work shall be initiated to correct the conditions so that the specified limit is set, and no operations which increase the reactivity of the core shall be made.
C.
After initial fuel loading and after each core refueling. operation and prior to reactor operation at greater than 75% of rated power, the movable incore detector system shall be utilized to verify proper power distribution.
Basis Detailed instructions, the above specified precautions and the design of the
. fuel handling equipment, which incorporates built-in interlocks and safety features, provide assurance that an accident, which would result in a hazard to public health and safety, will not occur during refueling operations.
When no change is being made in core geometry, one neutron detector is Amendment No. 67 & 67
TS 3.10-5 sufficient to monitor the core and permits maintenance of the out-of-function instrumentation. Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. Containment high radiation levels and high airborne activity. levels automatically stop and isolate the Containment Purge System. The fuel building ventilation exhaust is diverted through charcoal filters whenever refueling an in progress. At least one flow path is required for cooling and mixing the coolant contained in the reactor vessel so as to maintain a uniform boron concentration an[to remove residual heat.
The shutdown margin established by Specification A-9' maintains the' core suberitical, even with all of the control rod assemblies withdrawn from the core.
During refueling, the reactor, refueling water cavity is filled with approximately 220,000 gal of water borated to at least 2,000 ppm boron. The boron concentra-tion of this water is sufficient to maintain the reactor suberitical by approxi-mately 10% A k/k in the cold shutdown condition with all control rod assemblies inserted and also to maintain the core suberitical by approximately 1% with no control rod assemblies inserted into the reactor. Periodic checks of refueling water baron concentration assure the proper shutdown margin. Specification h-10 allows the Control Room Operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movecent.
In addition to the above safeguards, interlocks are used during refueling to assure safe handling of the fuel assemblies. An excess weight interlock is provided on the lifting hoist to prevent cove =ent of more than one fuel tsse=bly at a ti=e.
The spent fuel transfer echaris: ca: accccodate c ly
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i one fuel asse=bly at a time.
smer.d:ent Ib.
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TS 3.10-6 Upon each completion of core loading and installation of the reactor vessel head, specific mechanical and electrical tests vill be performed prior to I
initial criticql.ity.
The fuel handling accident has been analyzed based on the activity that could be released from fuel rod gaps of 204 rods of the highest power assembly with 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay period following power operation at 2550 MWt for 23,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
h The requirements detailed in Specification 3.10 provide assurance that refueling unit conditions conform to the operating conditions assumed in the a'ccident analysis.
Detailed procedures and checks insure that fuel assemblies are loaded in the
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proper locations in the core. As an additional check, the moveable incore detector system will be used to verify proper power distribution. This system is capable of revealing any assembly enrichment error or loading error which l
could cause power shapes to be peaked in excess of design value.
- Fuel rod gap activity from 204 rods of the highest power 15x15 assembly is greater than fuel rod gap activity from 264 rods of the highest power 17x17 demonstration assembly.
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Amendment No. 67 & 67 l
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TS 3.10-7 References FSAR Section 5.,2 Containment Isolation 1
FSAR Section 6.3 Consequence Limiting Safeguards FSAR Section 9.12 Fuel Handling System FSAR Section 11.3 Radiation Protection FSAR Section 13.3 Table 13.3-1 FSAR Section 14.4.1 Fuel Handling Accidents FSAR Supplement: Volume I: Question 3.2 I
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Amendment No. 67 & 67 i
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