ML20076K732
| ML20076K732 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 10/26/1994 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20076K734 | List: |
| References | |
| NUDOCS 9411020037 | |
| Download: ML20076K732 (7) | |
Text
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Jt UNITED STATES E
NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 20566 4001
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VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 i
VERMONT YANKEE NUCLEAR POWER STATION j
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.141 License-No. DPR-28 1.
The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The application for amendment filed by the Vermont Yankee Nuclear Power Corporation (the licensee) dated December 6, 1993, 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
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in compliance with the Commission's regulations; i
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.
2.
Accordingly, the license is amended by changes to the Technical Specifica-
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tions as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-28 is hereby amended to read as follows:
9411020037 941026 1
PDR ADOCK 0500 1
P
. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.141, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
Th s license amendment is effective as of its date of issuance and i
3.
shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION L
V Walter R. Butler, Director Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
October 26, 1994
5 l
ATTACHMENT TO LICENSE AMENDMENT NO.141 i
FACILITY OPERATING LICENSE NO. OPR-28 DOCKET NO. 50-271 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.
Remove Insert 121 121 122 122 143 143 144 144
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VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION 2.
Operability testing of safety-related pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Ccde and applicable Addenda as required by 10 CFR 50, Section 50.55a(g),
except where specific l
written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g) (6) (i).
F.
Jet Pumps F.
Jet Pumps 1.
Whenever the reactor is 1.
Whenever there is in the startup/ hot recirculation flow with standby or run modes, the reactor in the all jet pumps shall be startup/ hot standby or intact and all operating run modes, jet pump jet pumps shall be integrity and operable.
If it is operability shall be determined that a jet checked daily by pump is inoperable, an verifying that the orderly shutdown shall following two conditions be initiated and the do not occur reactor shall be in a simultaneously:
cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
a.
The recirculation pump flow differs by more than 10%
from the established speed-flow characteristics.
b.
The indicated total core flow is more than 10% greater than the core flow value derived from established power-core flow relationships.
2.
Flow indication from 2.
In the event that the each of the twenty jet jet pump (s) fail the pumps shall be verified tests in Specifications prior to initiation of 4.6.F.1.a and 4.6.F.1.b, reactor startup from a determine their cold shutdown condition.
operability by verifying that each individual jet pump AP% deviation from average loop AP does not vary from its normal established deviation by I
more than 10%.
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Amendment No.
99, 94, 141 121
VYllPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION 3.
The indicated core flow 3.
The surveillance is the sum of the flow requirements of 4.6.F.1 indication from each of and 4.6.F.2 do not apply the twenty jet pumps.
to the idle loop and If flow indication associated jet pumps failure occurs for two when in single loop or more jet pumps, operation.
immediate corrective action shall be taken.
4.
The baseline data If flow indication for required to evaluate the all but one jet pump conditions in cannot be obtained Specifications 4.6.F.1 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> an and 4.6.F.2 shall be orderly shutdown shall acquired each operating be initiated and the cycle.
Baseline data reactor shall be in a for evaluating 4.6.F.2 cold shutdown condition while in single loop within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
operation shall be updated as soon as G.
Sincle Loop Operation practical after enteririg single loop operation.
1.
The reactor may be started and operated or G.
Sinole Loop operation operation may continue with a single 1.
With one recirculation recirculation loop pump not in operation, provided that:
core flow between 34%
and 45% of rated, and a.
The designated core thermal power adjustments for greater than the limit APRM flux scram and specified in rod block trip Figure 3.6.4 (Region 2),
settings (Specifi-establish baseline APRM cations 2.1.A.l.a and LPRMW neutron flux and 2.1.B.1, noise levels prior to Table 3.1.1 and entering this region, Table 3.2.5), rod provided that baseline block monitor trip values have not been setting established since the (Table 3.2.5), MCpR last core refueling.
fuel cladding Baseline values shall be integrity safety established with one limit (Specifi-recirculation pump not ention 1.1.A), and in operation and core MCPR operating thermal power less than limits and MAPLHGR or equal to the limit limits, provided in specified in the Core Operating Figure 3.6.4.
Limits Report, are initiated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
During the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, either these adjustments must be completed or the reactor brought to Hot Shutdown.
(1)
Detector Levels A and C of one LPRM string per core octant plus detector Levels A and C of one LPRM string in the c.cnter.;f the core shall be monitored.
Amendment No. M, 94, 4M, 14l 122 a
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VYNPS i
l l
BASES:
3.6 and 4.6 (Cont'd) throughout plant life. The inservice inspection and testing programs are performed in accordance with 10CFR50, Section 50.55a(g) except where specific relief has been granted by the NRC.
These inspection and testing programs provide further assurance that gross defects are i
not occurring and ensure that safety-related components remain operable.
l The type of inspection planned for each component depends on location, accessibility, and type of expected defect. Direct visual examination is proposed wherever possible since it is sensitive, fast, and reliable.
Magnetic particle and liquid penetrant inspections are planned where practical, and where added sensitivity is required. Ultrasonic testing and radiography shall be used where defects can occur on concealed surfaces.
Generic Letter 88-01 established the NRC position for in-service inspection of BWR austenitic stainless steel piping susceptible to Intergranular Stress Corrosion Cracking (IGSCC).
The ir-service inspection and testing programs presented at this time are based on a thorough evaluation of present technology aad state-of-the-art inspection and testing techniques.
F.
Jet Pumps Failure of a jet pump nozzle assembly hold down mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break.
Therefore, if a failure occurred, repairs must be made.
l The following factors form the basis for the surveillance requirements:
A break in a jet pump decreases the flow resistance characteristic of the external piping loop causing the recirculation pump to operate at a higher flow condition when compared to previous operation.
The change in flow rate of the failed jet pump produces a change in the indicated flow rate of that pump relative to the other pumps in that loop. Comparison of the data with a normal relationship or pattern provides the indication necessary to detect a failed jet pump.
The jet pump flow deviation pattern derived from the diffuser to lower plenum differential pressure readings will be used to further evaluate jet pump operability in the event that the jet I
l pumps fail the tests in Specifications 4.6.F.1.a and b.
Amendment No +G, 146, +49'l4l 143
e VYNPS BASES:
3.6 and 4.6 (Cont'd)
Agreement of indicated core flow with established power-core flow relationships provides the most assurance that recirculation flow is not bypassing the core through inactive or broken jet pumps. This bypass flow is reverse with respect to normal jet pump flow.
The indicated total core flow is a summation of the flow indications for the twenty individual jet pumps. The total core flow measuring instrumentation sums reverse jet pump flow as though it were forward flow (except in the case of single loop operation when reverse flow is subtracted from the total jet pump flow). Thus, the indicated flow is higher than actual core flow by at least twice the normal flow through any backflowing pump.
Reactivity inventory is known to a high degree of confidence so that even if a jet pump failure occurred during a shutdown period, subsequent power ascension would promptly demonstrate abnormal control rod withdrawal for any power-flow operating map point.
A nozzle-riser system failure could also generate the coincident failure of a jet pump body; however, the converse is not true.
The lack of any substantial stress in the jet pump body makes failure impossible without an initial nozzle-riser system failure.
G.
Single Loop Oparation Continuous operation with one recirculation loop was justified in
" Vermont Yankee Nuclear Power Station Single Loop Operation",
NEDo-30060, February 1983, with the adjustments specified in Technical Specification 3.6.G.I.a.
APRM and/or LPRM oscillations in excess of those specified in Section 3.6.G.l.b could be an indication that a condition of thermal hydraulic /neutronic instability exists and that appropriate remedial action should be taken.
By restricting core flow to greater than or equal to 34% of rated, which corresponds to the core flow at the 80%
rod line with 2 recirculation pumps running at minimum speed, the region of the power / flow map where these oscillations are most likely to occur is avoided (Region 1 of Figure 3.6.4).
These specifications are based upon the guidance of GE SIL #380, Revision 1, dated February 10, 1984.
During single loop operation, the idle recirculation loop is isolated by electrically disarming the recirculation pump motor generator set drive motor, until ready to resume two loop operation. This is done to prevent a cold water injection transient caused by an inadvertent pump startup.
Under single loop operation, the flow control is placed in the manual mode to avoid control oscillations which may occur in the recirculatien flow control system under these conditions.
H.
Recirculation System The largest recirculation break area assumed in the ECCS evaluation was 4.14 square feet.
Amendment No. 44, 90, 44, 141 144
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