ML20248E099
| ML20248E099 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 08/02/1989 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20248E101 | List: |
| References | |
| NUDOCS 8908110238 | |
| Download: ML20248E099 (12) | |
Text
_ _ _ _ _ _ _
e?'""4 UNITED STATES i ' ' l-i NUCLEAR REGULATORY COMMISSION h-
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WASHINGTON, D. C. 20555 T
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VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SUFRY POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 131 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 August 5,1988, as supplemented January 25,19J9, co.nplies with the standards and requirements of the Atcr.iic Energy Act of 1954, as amended (the Act) and the Commissien's rules and regulations set forth in 10 CFR Chapter I; B.
The f acility 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 authorize 3 by this amendment can be concucted without endangering the healti.
and safety of the public, and (ii) that such activities will be conducted in conpliance 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 acendment 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 TL:hnical Specifications as indicated in the attachrent to this lictnse amendment, and paragraph 3.3 of Facility Operating License No. DPR-32 is hereby amended to read as follows:
I 890G110238 890802 I
PDR ADOCK 05000280
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2 (B). Technical Specifications The Technical Specifications contained in Appendix A, as revised. hrough Amendment No. 131, 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 1
/
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.. rbert H. Berkow, Director Project Directorate 11-2 Division of Reactor Projects-1/II Office of Nuclear Reactor Reculation
Attachment:
Changes.to the Technical Specifications Date of Issuance: August 2. 1989
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UNITED STATES i
'i NUCLEAR REGULATORY COMMISSION
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i WASHINGTON. D. C. 20555 sg.....f.
VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 131 License No. DPR-37 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated August 5,1988, as supplemented January 25, 1989, 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 Comission; 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon j
defense and security or to the bealth and safety of the public; and l
E.
The issuance of this amentent is in accordance with 10 CFR Part l
51 of the Comission's regulations and all applicable requirements I
have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amentent, and paragraph 3.B of Facility Operating License ho. DPR-37 is hereby amended to read as follows:
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(B) Technica1' Specifications I
The Technical Specifications contained in Appendix A, as revised through Amendnent ho.131, 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
' 4}/) / g T. w L Herbert N. Berkow, Director Project Directorate 11-2 Division of Reactor Projects-1/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: August 2, 1989 l
1 ATTACHMENT TO LICENSE AMENDMEN1 l
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AMENDMENT NO.
131 FACILITY CPERATING LICENSE NO. DPR-32 AMENDMENT tiO.
131 FACILITY CPERATING LICENSE NO. OPR-37 DOCKET f405. 50-280 AND 50-281 l
l Revise Appendix A as follows-1 1
Remove Pages Insert Pages TS 3.12-8 TS 3.12-8 TS 3.12-10 TS 3.12-10 TS 3.12-11 TS 3.12-11 TS 3.12-13 TS 3.12-13 TS 3.12-13a TS 4.1-6 TS 4.1-6 TS 4.1-7 TS 4.1-7
- g TS 3.12-8 L
AT and Overtemperature aT trip settings shall be reduced by the equivalent of 2% power for every 1% quadrant to average power tilt.
C.
Inocerable Control Rods 1.
A control rod assembly shall be considered inoperable if the assembly cannot be moved by the drive mechanism or the assembly remains misaligned from its group step demand position by more than 24 steps during the " Thermal Soak" period, as defined in Section 3.12.E.1.b, or 12 steps otherwise during power operation.
No tolerance limit is required in the shutdown modes, but a rod shall be considered inoperable if the rod position indicators do not verify rod movement upon demand.
Additionally, a full length control rod shall be considered inoperable if its rod drop time is greater than 2.4 seconds to dashpot entry.
2.
No more than one inoperabie control rod assembly shall be permitted when the reactor is critical.
3.
If more than one control rod assembly in a given bank is out of service because :/ a single failure external to the indiriouel rod drive mechanism, (i.e.
programming circuitry), the provisions of Specifications 3.12.C.1 and 3.12.C.2 shall not apply and the reactor may remain l
critical for a period not to exceed two hours provided innediate attention is directed toward making the i
necessary repairs.
In the event the affected asseelies l
cannot be returned to service within this specified 1
period, the reactor will be brought to hot shutdown co.',d i ti on s.
4 The provisions of Specifications 3.12.C.1 and 3.12.C.2 shall not apply during physics tests in which the assemblies are intentionally misaligned.
5.
Power operation may continue with one rod inoperable provided that within one hour either:
a.
the rod is no longer inoperable as defined in Speci-I fication 3.12.C.1, or Amendment Nos.131 and 131
4 TS 3.12-10 5.
If power has been reduced in accordance aith Specification 3.12.C.5.b, power may be increased above 75% power provided that:
a)
'an analysis has been performed to determine the hot channel factors and the resulting. allowable power level based.on the limits ~ of Specification 3.12.B;1, and a) an evaluation of the effect of operating at the increased-power level on the accident analyses of Table 3.12-1 has been completed.
D.
Core Quadrant Power Balance:
1.
If the reactor is operating above 75% of rated power with one excore nuclear channel out of service, the core quaarant power balance shall be determined:
a.
Once per day, and b.
After a change _ in power level greater than 10% or more than 30 inches of control rod motion.
2.
The core quadrant power balance shall be determined by one of the following methods:
.a.
Movable detectors (at least two per quadrant) b.
Core exit thermocouple (at least four per quadrant)
E.
Rod position Indicator Cher
',1 1.
Rod Position Indication shall be provided as follows:
a.
Above 50% power, the rod potition indication system shall be operable and capable of determining the control rod positions to within !!2 steps of their respective group step demand counter indications, b.
From movement of control banks to achieve criticality up to 50% power, the rod position indication system shall be operable and capable of determining the control rod positions to within !24 steps of their respective gr7up step demand counter indications for a maximum of one hour out of twenty-four, and to within 212 steps otherwise.
During the one-hour
" Thermal Soak" period, the step demand counters shall be operable and capable of determining the group demand positions to within 22 steps.
Amendment Nos. 131 and 131
TS 3.12-11 c.
In hot, intermediate and cold shutdown conditions, the step demand counters shall be operable and capable of determining the group demand positions to within 12 steps.
The rod position indicators shall be available to verify rod movement upon demand.
-2.
If a rod position indicator channel is out of service, then:
a.
For operation above 50% of rated power, the position of the RCC shall be checked indirectly using the movable incore detectors at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and immediately after any motion of the non-indicating rod exceeding 14 steps, or b.
Reduce Power to less than 50% of rated power within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
During operations below 50% of rated power, no special monit', ring is required.
3.
If more than one rod position (RPI) indicator channel per group or two RPI channels per bank are inoperable during control bank motion to achieve criticality or power operations, then the requirements of Specification 3.0.1 will be followed.
Basis The reactivity control concept assumed for operation is that reactivity changes accompanying changes in reactor power are compensated by control rod assembly motion.
Reactivity changes associated with xenon, samarium, fuel depletion, and large changes in reactor coolant temperature (operating temperature to cold shutdown) are compensated for by changes in the soluble boron concentration. During power operation, the shutdown l
groups are fully withdrawn and control of power is by the control groups.
l A reactor trip occurring during power operation will place the reactor L
into the hot shutdown condition.
The control rod assembly insertion limits provide for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod assembly remains fully withdrawn, with sufficient margins to meet the assumptions used in the accident analysis.
In addition, they provide a limit Amendment Nos. 131 and 131
TS 3.12-13 in service, the effects of malpositioned control rod assemblies are observable from nuclear and process information displayed in the Main Control Room and by core thermocouple and in-core movable detectors.
Below 50% ' power, no special monitoring is required for malpositioned control rod assemblies with inoperable rod position indicators because, even with an unnoticed complete assembly misalignment (full length control rod assembly 12 feet out of alignment with its bank), operation at 50% steady state power does not result in exceeding core limits.
The." Thermal St ik" allowance below 50% power, during which the rod position indication system tolerance requirem'ent is
- relaxed, provides time for the system to reach thermal equilibrium.
A total of one hour in twenty-four is available for this allowance, which may be a continuous hour or may consist of discrete, shorter intervals.
For such a short period of time, a misaligned rod does not pose an unacceptable risk.
At these conditions, the rod position indicators should still be used to verify rod movement but not their exact location.
The tolerance is tightened after one hour t> ensure that the thermal overshoot does not concea!
.6 n.tual rod misalignment.
The reliance upon the step. demand counters at hot ano cold shutdown conditions shifts the monitoring of rod position from the rod position indication system to the more reliable demand counters when RCS temperature is changing greatly but the core remains saxritical.
The step demand counters also provide the therma' soak period.
precise group demand positions durin3 l
The specified control rod assembly drop time is consistent with safety analyses that have been performed.
I Amendment Nos. 131 and 131
4 r TS 3.12-13a l An inoperable control rod assembly imposes. additional demand!.
on the operators.
The permissible number of inoperable control rod assemblies is limited to one in crder to limit the' magnitude of-the operating burden, but such a failure would not prevent dropping of -the operable control rod assemblies upon reactor trip.
' Two criteria have been chosen as a design basis for fuel
- performance related to fission gas release, pellet temperature, and cladding mechanical properties.
First, the peak value of fuel centerline temperature must not exceed 4700'F.
- Second, the minimum DNBR in the core must not be less than the applicable design limit in normal operation or in'short term transients.
)
I Amendment Nos. 131 and 131
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