ML20069F554
| ML20069F554 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 06/01/1994 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20069F557 | List: |
| References | |
| GL-88-01, GL-88-1, NUDOCS 9406080353 | |
| Download: ML20069F554 (7) | |
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E UNITED STATES 57 NUCLEAR REGULATORY COMMISSION
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VERMONT YANKEE NVCLEAR POWER CORPORATION DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.139 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 July 14, 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; 8.
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.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 3.8 of Facility Operating License No. DPR-28 is hereby amended to read as follows:
9406000353 940601 PDR ADOCK 05000271 p
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Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.139, 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 its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/ 'ctI Walter R. Butler, Director Project Directorate I-3 Division of Reactor Projects - I/II' Office of Nuclear Reactor Regulation
Attachment:
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Changes to the Technical Specifications Date of Issuance: June 1, 1994
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ATTACHMENT TO LICENSE AMENDMENT N0.139 FACILITY OPERATING LICENSE N0. DPR-28 DOCKET N0. 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 108 108 108a 108a 122 122 123 123 1
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VYNPS VYNPS 3.6 LIMITING CONDITIONS FOR OPERATION 4.6 SURVEILLANCE REQUIREMENTS l
C.
Coolant Leakage C.
Coolant Leakage l
l la. Any time irradiated fuel is in the reactor vessel 1.
Reactor coolant system leakage, for the purpose of i
and reactor coolant temperature is above 212oF, satisfying Specification 3.6.C.1, shall be checked reactor coolant leakage into the primary and logged once per shift, not to exceed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
containment from unidentified sources shall not exceed 5 gpm.
In addition, the total reactor coolant leakage into the primary containment shall not exceed 25 gpm.
- b. While in the run mode, reactor coolant leakage into the primary containment from unidentified sources shall not increase by more than 2 gpm within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
2.
Both the sump and air sampling systems shall be operable during power operation.
From and after the date that one of these systems is made or found inoperable for any reason, reactor operation is permissible only during succeeding seven days.
3.
If these conditions cannot be met, initiate an orderly shutdown and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D.
Safety and Relief Valves D.
Safety and Relief Valves 1.
During reactor power operating conditions and 1.
Operability testing of Safety and Relief Valves whenever the reactor coolant pressure is greater shall be in accordance with Specification 4.6.E.
than 120 psig and temperature greater than 3500F, The lif t point of the safety and relief valves shall both safety valves shall be operable. The relief be set as specified in Specification 2.2.B.
valves shall be operable, except that if one relief valve is inoperable, reactor power shall be immediately reduced to and maintained at or below 95% of rated power.
2.
If Specification 3.6.D.1 is not met, initiate an orderly shutdown and the reactor coolant pressure shall be below 120 psig and 3500F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
18 Amendment No, M8, 139 C40V e
VYNPS VYNPS 3.6 LIMITING CONDITIONS FOR OPERATION 4.6 SURVEILLANCE REQUIRDILTS E.
Structural Integrity and Onerability Testing E.
Structural Integrit.y and Operability Testing The structural integrity and the operability of the 1.
Inservice inspection of safety-related components safety-related systems and components shall be shall be perfonned in accordance with Section XI maintained at the level required by the original of the ASME Boiler and Pressure Vessel Code and acceptance standards throughout the lif e of the plant.
applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
Inservice inspection of piping, identified in NRC Generic Letter 88-01, shall be performed in accordance with the staff positions on schedule, methods, and personnel and sample expansion included in the Generic Letter.
2.
Operability testing of safety-related pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and pressure Vesnel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g),
except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
t I
i Amendment 99, 139 108a (40\\1
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VYNPS 3.6 & 4.6 (Continued) greater than the limit specified for unidentified leakage; the probability is small that imperfections or cracks associated with such leakage would grow rapidly.
Leakage less than the limit specified can be detected within a few hours utilizing the available leakage detection systems.
If the limit is exceeded and the origin cannot be determined in a reasonably short time the plant should be shut down to allow further investigation and corrective action.
The 2 gpm increase limit in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period for unidentified leaks was established as an additional requirement to the 5 gpm limit by Generic Letter 88-01,
- NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping.*
The removal capacity f rom the drywell floor drain sump and the equipment drain sump is 50 gpm each. Removal of 50 gpm from either of these sumps can be accomplished with considerable margin.
D.
Safety and Relief Valves Parametric evaluations have shown that only three of the four relief valves are required to provide a pressure margin greater than the recommended 25 psi below the safety valve actuation settings as well as a MCPR > 1.06 for the limiting overpressure transient below 98% power.
Consequently, 95% power has been selected as a limiting power level for three valve operation.
For the purpose of this limiting condition a relief valve that is unable to actuate within tolerance of its set pressure is considered to be as inoperable as a mechanically malfunctioning valve.
Experience in safety valve operation shows that a testing of 50% of the safety valves per refueling outage is adequate to detect failures or deterioration. The tolerance value is specified in Section III of the ASME Boiler and Pressure Vessel Code as 1 1% of design pressure. An analysis has been performed which shows that with all safety valves set 1% higher the reactor coolant pressure safety limit of 1375 psig is not exceeded.
E.
Structural Integrity and Operability Testing A pre-service inspection of the components listed in Table 4.2-4 of the FSAR will be conducted after site erection to assure f reedom f rom defects greater than code allowance; in addition, this serves as a reference base for further inspections. Prior to operation, the reactor primary system will be f ree of gross defects.
In addition, the facility has been designed such that gross defects should not occur 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 not occurring and ensure that safety-related components remain operable.
Amendment No. us.139 122 C400
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VYNPS 3.6 & 4 (CO?rr' D)
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 def ects can occur on concealed surf aces.
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 in-service inspection and testing programs presented at this time are based on a thorough evaluation of present technology and state-of-the-art inspection and testing techniques.
F.
Jet Pumns 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.
The detection technique is as follows.
With the two recirculation pumps balanced in speed to within +51, the flow rates in both recirculation loops will be verified by main control Room monitoring instruments.
If the two flow rate values do not differ by more than 101, riser and nozzle assembly integrity has been verified. If they do dif f er by 10% or more the core flow rate measured by the jet pump dif fuser dif f erential pressure system must be checked against the core flow rate derived from the measured value of loop flow to core flow correlation.
If the difference between measured and derived core flow rate is lot or more (with the measured value higher) diffuser measurements will be taken to define the location within the vessel of f ailed jet pump nozzle (or riser) and the plant shut down for repairs.
If the potential blowdown flow area is increased, the system resistance to the recirculation pump is also reduced; hence, the af f ected drive pump will
- rtin out " to a substantially higher flow rate (approximately 115% to 120% for a single nozzle f ailure).
If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed.
Any imbalance
- between drive loop flow rates would be indicated by the plant process instrumentation. In addition, the af f ected jet pump would provide a leakage path past the core thus reducing the core flow rate.
The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (31 to 61) in the total core flow measure. This decrease, together with the loop flow increase, would result in a leak of correlation between measured and derived core flow rate.
Amendment No. 45, -14 8, 139 123 (40\\7 e
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