ML20247M935
| ML20247M935 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 05/14/1998 |
| From: | Thomas C NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20247M939 | List: |
| References | |
| NUDOCS 9805260420 | |
| Download: ML20247M935 (8) | |
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UNITED STATES l
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NUCLEAR REGULATORY COMMISSION g
WASHINGTON, D.C. 30666-0001
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i VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.161 License No. DPR-28 l
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment filed by the Vermont Yankee Nuclear Power Corporation (the licensee) dated March 20,1998, 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: (l) 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 si of the Commission's regulations and all applicable requirements have been satisfied.
i 9805260420 980514 PDR ADOCK 05000271
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. 2. Accordingly, the license is hereby amended to authorize revision of the Final Safety Analysis Report (FSAR) as set fe,rth in the application for amendment by Vermont Yankee I
Nuclear Power Corporation dated March 20,1998. The licensee shall submit the revised description authorized by this amendment with the next update of the FSAR in acco' #r.e with 10 CFR 50.71(e).
- 3. This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Cac.o M Cecil O. Thomas, Director Project Directorate 1-3 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical i
Specification Date of issuance: May 14, 1998 I
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ATTACHMENT TO LICENSE AMENDMENT NO.181 FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 Replace the following pages of Appendix A Technical Specification Bases with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove JDaart 165 165 165s 165a 165b 188 188 188a l
t VYNPS BASES:
3.7 (Cont'd)
The vacuum relief system from the pressure suppression chamber to Reactor Building consists of two 100% vacuum relief breakers (2 parallel sets of 2 valves in series).
Operation of either system will maintain the pressure differential less than 2 psig; the external design pressure is 2 psig. With one vacuum breaker out of service there is no immediate threat to accident mitigation or primary containment and, therefore, reactor operation can be continued for 7 days while repairs are being made.
The espacity of the ten (10) drywell vacuum relief valves is sized to limit the pressure differential between the suppression chamber and drywell during post-accident drywell cooling operations to the design limit of 2 psig. They are sized on the basis of the Bodega Bay pressure suppression tests. The ASME Boiler and Pressure Vessel Code,Section III, Subsection B, for this vessel allows eight (8) operable valves, therefore, with two (2) valves secured, containment integrity is not impaired.
Each drywell-suppression chamber vacuum breaker is fitted with a-redundant pair of limit switches to provide fail-safe signals to panel mounted indicators in the Reactor Building and alarms in the Control Room when the disks are open more than 0.050" at all points along the seal surface of the disk. These switches are capable of transmitting the disk closed to open signal with 0.01" movement of the switch plunger.
Continued reactor operation with failed components is justified because of the redundance of components and circuits and, most importantly, the accessibility of the valve lever arm and position reference external to the valve. The fail safe feature of the alarm circuits assures operator attention if a line l
fault occurs.
The requirement to inert the containment is based on the l
recommendation of the Advisory Committee on Reactor Safeguards.
This recommendation, in turn, is based on the assumption that several percent of the zirconium in the core will undergo a reaction with steam during the less-of-coolant accident. This reaction would release sufficient hydrogen to result in a flammable concentration in the primary containment building.
The oxygen concentration is therefore kept below 4% to minimize the possibility of hydrogen combustion.
General Electric has estimated that less than 0.1% of the zirconium would react with steam following a less-of-coolant due to operation of emergency core cooling equipment.
This quantity of zirconium would not liberate enough hydrogen to form a combustible mixture.
The use of the 18" purge and vent flow path isolation valves AC-7A i
(16-19-7A), AC-7B (16-19-7B), AC-8 (16-19-8), AC-10 (16-19-10) has been restricted to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year. Normal plant operations (other than inerting and de-inerting) will have AC-8 and AC-10 closed and nitrogen will be supplU 3 to the drywell via the 1" nitrogen makeup supply. The differential pressure maintained between the drywell and torus will allow the nitrogen to " bubble over" into the suppression l
chamber. A normally open AC-6B (3") allows for venting. A normally j
closed AC-6A (3") is periodically opened for performance of Amendment No. 49, ":::: Ch;ng;, 110, 444 161 165 l
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J VYNPS BASES:
3.7 (Cont'd) surveillance such as monthly torus to drywell vacuum breaker tests.
Procedurally, when AC-6A is open, AC-6 and AC-7 are closed to prevent overpressurization of the SBGT system or the reactor building ductwork, should a LOCA occur.
For this and similar analyses performed, a spurious opening of AC-6 or AC-7 (one of the closed containment' isolation valves) is not assumed as a failure h
simultaneous with a postulated LOCA.
Analyses demonstrate that for normal plant operation system alignments, including surveillance such as those described above, that SBGT integrity would be maintained if a LOCA was postulated. Therefore, during normal plant operations, the 90 hour0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> clock does not apply. Accordingly, opening of the 18 inch atmospheric control isolation valves AC-7A, AC-7B, AC-8 and AC-10 will be limited to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per calendar year (except for performance of the subject valve stroke time surveillance - in which case the appropriate corresponding valves are closed to protect 3
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equipment should a LOCA occur). This restriction will apply whenever primary containment integrity is required. The 90 hour0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> clock will apply anytime purge and vent evolutions can not assure the integrity of the SDGT trains or related equipment.
B. and C. Standby Gas Treatment System and Secondary Containment System The secondary containment is designed to minimize any ground level
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release of radioactive materials which might result from a serious accident. The Reactor Building provides secondary containment during reactor operation, when the drywell is sealed and in service; the Reactor Building provides primary containment when the reactor is shutdown and the drywell is open, as during refueling. Because the secondary containment is an integral part of the complete containment system, secondary containment is required at all times that primary containment is required except, however, for initial fuel loading and low power physics testing.
In the Cold Shutdown condition or the Refuel Mode,the probability and consequence of the LOCA are reduced due to the pressure and temperature limitations in these conditions.
Therefore, maintaining Secondary Containment Integrity is not required in the Cold Shutdown condition or the Refuel Mode, except for other situations for which significant releases of radioactive material can be postulated, such an during operations with a potential for draining the reactor vessel, during alteration of the Reactor Core, or during movement of irradiated fuel assemblies or the fuel cask in the secondary containment.
With the reactor in the Run Mode, the Startup Mode, or the Hot Shutdown condition, if Secondary containment Integrity is not maintained, Secondary Containment Integrity must be restored within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during the Run Mode, the Startup Mode, and the Hot Shutdown condition.
This time period also ensures that the probability of an accident (requiring Secondary Containment Integrity) occurring during periods where Secondary Co7tainment Integrity is not maintained, is mininal.
If Secondary Containment Amendment Nc. l i 7, 161 165a
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VYNPS-B&Sfd 3 '. 7 (Cont'd)
Integrity cannot be restored within the required time period, the plant must be brought to a mode or condition in which the LCO does not apply.
Movement of irradiated fuel assemblies or the fuel cask in the secondary containment, alteration of the Reactor Core, and operations with the potential for draining the reactor vessel can be postulated
.to cause fission product release to the secondary containment.
In.
such cases, the secondary containment is the only barrier to' release of fission products to the environment. Alteration of the Reactor Core and movement of irradiated fuel assemblies and the fuel cask must be immediately suspended if secondary Containment Integrity is not maintained.
Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend operations with the potential.for draining the reactor vessel to minimize the probability of a vessel draindown and subsequent potential.for fission product release. Actions must continue until operations with the potential for draining the reactor vessel are-suspended.
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i Amsndment No. 161 165b l
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2 VYNPS BASES:
1.8 (Cont'd)
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Ventilation Exhaust Treatment The requirement that the AOG Building and Radwaste Building HEPA filters be.used when specified provides rensonable assurance that the release of radioactive materials in gaseous effluents will be kept "as low as is reasonab.y achievable".
This specification implements the requirements of 10CFR Part 50.36a and the design objective of Appendix I to.10CFR Part 50.
The requirements governing the use of the appropriate portions of the gaseous radwaste filter systems were specified by the'NRC in NUREG-0473 Revision 2 (July 1979) as a suitable fraction of the guide set forth in Sections II.B and II.C of Appendix I, 10CFR Part 50, for gaseous effluents.
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Exolosive Gas Mixture The hydrogen monitors are used to detect possible hydrogen buildups which could result in a possible hydrogen explosion. Automatic isolation of the off-gas flow would prevent the hydrogen explosion and possible damage to the augmented off-gas system. Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materiels will be controlled.
K.
Steam Jet Air Eiector (SJAE) l Restricting the gross radioactivity release rate of gases from the main condenser SJAE provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10CFR Part 50.
L.
Primary Containment (MARK I)
This specification provides reasonable assurance that releases from containment purging / venting operations will be filtered through the Standby Gas Treatment System (SBGT) so that the annual dose limits of 10CFR Part 20 for Members of the Public in areas at and beyond the Site Boundary will not be exceeded. The dose objectives of Specification 3.8.G restrict purge / venting operations when the Standby Gas Treatment System is not in use and gives reasonable assurance that all releases from the plant will be kept "as low as is reasonably achievable". The specification requires the use of SBGT only when Iodine-131, Iodine-133 or radionuclides in particulate form with half-lives greater than 8 days in containment exceeds the levels in Table 1, Column 3, to Appendix B of 10CFR 20.1001-20.2401 since
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the filter system is not considered effective in reducing noble gas radioactivity from gas streams.
The use of the 18" purge and vent flow path isolation valves AC-7A (16-19-7A), AC-7B (16-19-7B), AC-8 (16-19-8), AC-10 (16-19-10) has l
been restricted to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year. Normal plant operations (other y
than inerting and de-inerting) will have AC-8 and AC-10 closed and 1
nitrogen will be supplied to the drywell via the 1" nitrogen makeup supply. The differential pressure maintained between the drywell i
j Amendment No. 03, 151 161 188 l
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VYNPS BASES:
3.8 (Cont'd) and torus will allow the nitrogen to abubble over" into the suppression chamber. A normally open AC-6B (3") allows for venting.
A normally closed AC-6A (3") is periodically opened for performance of curveillances such as monthly torus to drywell vacuum breaker testa.
Procedurally, when AC-6A is open, AC-6 and AC-7 are closed to prevent overpressurization of the SBGT system or the reactor building ductwork, should a LOCA occur.
For this and similar analyses performv'., a spurious opening of AC-6 or AC-7 (one of the closed containment isolation valves) is not assumed as a failure simultaneous with a postulated LOCA.
Analyses demonstrate that for normal plant operation system alignments, including surveillance such as those described above, that SBGT integrity would be maintained if a LOCA was postulated.
Therefore, during normal plant operations, the 90 hour0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> clock does'not apply. Accordingly, opening of the 18 inch atmospheric control isolation valves AC-7A, AC-7B, AC-8 and AC-10 will be limited to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per calendar yrAr (except for performance of tha subject valve stroke time surveillance - in which case the appropriate corresponding valves are closed to protect equipment should a LOCA occur).
This restriction will apply whenever primary containment integrity is required.
The 90 hour0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> clock will apply anytime purge and vent evolutions can not assure the integrity of the SBGT trains or related equipment.
I M.
Total Dose (4 0CFR190)
This specification is provided to meet the dose limitations of 40CFR Part 190 to Members of the Public in areas at and beyond the Site Boundary.
The specification requires the preparation and submittal of a Specific Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I.
For sites containing up to 4 reactors, it is highly unlikely that the resultant doce to a Member of the Public will exceed the dose limits of 40CFR Part 190 if the individual reactors i
remain within the reporting requirement level. The Special Report will describe a course of action that should result in the limitation Amendment No.161 188a 1
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