ML081120121

Technical Specifications, Amendment No. 231

ML081120121
Person / Time
Site:	Vermont Yankee
Issue date:	04/17/2008
From:	NRC/NRR/ADRO/DORL/LPLI-1
To:
Kim J, NRR/DORL, 415-4125
References
TAC MD7054
Download: ML081120121 (4)

Text

E. Entergy Nuclear Operations, Inc., pursuant to the Act and 10 CFR Parts.30 and 70, to possess, but not to separate, such byproduct and special nuclear material as may be produced by operation of the facility.

3. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified! below:

A. Maximum Power Level Entergy Nuclear Operations, Inc. is authorized to operate the facility at reactor core power levels not to exceed 1912 megawatts thermal in accordance with the Technical Specifications (Appendix A) appended hereto.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 231, are hereby incorporated in the license. Entergy Nuclear Operations, Inc. shall operate the facility in accordance with the Technical Specifications.

C. Reports Entergy Nuclear Operations, Inc. shall make reports in accordance with the

• requirements of the Technical Specifications.

D. This paragraph deleted by Amendment No. 226.

E. Environmental Conditions Pursuant to the Initial Decision of the presiding Atomic Safety and Licensing Board issued February 27, 1973, the following conditions for the protection of the environment are incorporated herein:

Amendment No. 2-6, 209, 226, 22-9 ,-23-, 231

VYNPS 3.10 LIMITING CONDITIONS FOR 4.10 SURVEILLANCE REQUIREMENTS OPERATION

4. 480 V Uninterruptible Power Systems From and after the date that one Uninterruptible Power System or its associated Motor Control Center are made or found to be inoperable for any reason, the requirements of Specification 3.5.A.4 shall be satisfied.

5. RPS Power Protection From and after the date that one of the two redundant RPS power protection panels on an in-service RPS MG set or alternate power supply is made or found to be inoperable, the associated RPS MG set or alternate supply will be taken out of service until the panel is restored to operable status.

C. Diesel Fuel C. Diesel Fuel There shall be a minimum 7 1. The quantity of diesel day supply of usable diesel generator fuel shall be fuel in the diesel fuel'oil logged weekly and after storage tank. each operation of the unit.

2. Once a month a sample of diesel fuel shall be taken, checked for quality in accordance with the applicable ASTM Standards and logged.

Amendment No. -a4, -I-, -4, 2-8-0, 2-2*-4,231 218

VYNPS 3.12 LIMITING CONDITIONS FOR 4.12 SURVEILLANCE REQUIREMENTS OPERATION F. Fuel Movement F. Fuel Movement The reactor shall be shut Prior to any fuel handling or down for a minimum of 24 movement in the reactor core, hours prior to fuel movement the licensed operator shall within the reactor core. verify that the reactor has been shut down for a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

G. Crane Operability G. Crane Operability

1. The Reactor Building 1. a. Within one month crane shall be operable prior to spent fuel when the crane is used cask handling.

for handling of a spent operations, an fuel cask. inspection of crane cables, sheaves, hook, yoke, and cask lifting trunnions will be made in accordance with the applicable ANSI Standard. A crane rope shall be replaced if any of the replacement!

criteria are met. I

b. No-load mechanical and electrical tests will be conducted prior to lifting the empty cask from its transport vehicle to verify proper operation of crane controls, brakes and lifting speeds. A functional test of the crane brakes will be conducted each time an empty cask is lifted clear of its transport vehicle.

Amendment No. 2-9, ,3231 235

(

VYNPS BASES: 3.12 & 4.12 (Cont'd)

E. The intent of this specification is to permit'the unloading of a portion of the reactor core for such purposes as inservice inspection requirements, examination of the core support plate, control rod, control rod drive maintenance, etc. This specification provides assurance that inadvertent criticality does not occur during such operation.

This operation is performed with the mode switch in the "Refuel" position to provide the refueling interlocks normally available during refueling as explained in the Bases for Specification 3.12.A. In order to withdraw more than one control rod, it is necessary to bypass the refueling interlock on each withdrawn control rod which prevents more than one control rod from being withdrawn at a time. The requirement that the fuel assemblies in the cell controlled by the control rod be removed from the reactor core before the interlock can be bypassed ensures that withdrawal of another control rod does not result in inadvertent criticality. Each control rod essentially provides reactivity control for the fuel assemblies in the cell associated with that control rod. Thus, removal of an entire cell (fuel assemblies plus control rod) results in a lower reactivity potential of the core.

One method available for unloading or reloading the core is the spiral unload/reload. Spiral reloading and unloading encompass reloading or unloading a cell on the edge of a continuous fueled region (the cell can be reloaded or unloaded in any sequence.) The pattern begins (for reloading) and ends (for unloading) around a single SRM. The spiral reloading pattern is the reverse of the unloading pattern, with the exception that two diagonally adjacent bundles, which have previously accumulated exposure in-core, and placed next to each of the four SRMs before the actual spiral reloading begins. The spiral reload can be to either the original configuration or a different configuration.

Additionally, at least 50% of the fuel assemblies to be reloaded into the core shall have previously accumulated a minimum exposure of 1000 Mwd/T to ensure the presence of a minimum neutron flux as described in Bases Section 3.12.B.

F. The intent of this specification is to assure that the reactor core has been shut down for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following power operation and prior to fuel handling or movement. The safety analysis for the postulated refueling accident assumed that the reactor had been shut down for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for fission product decay prior to any fuel handling which could result in dropping of a fuel assembly.

G. The operability requirements of the reactor building crane ensures that the redundant features of the crane have been adequately inspected just prior to using it for handling of a spent fuel cask. The redundant hoist system ensures that a load will not be dropped for any postulated credible single component failures. Crane inspections and crane rope replacement criteria shall meet the requirements of ANSI Standard B30.2-1967. Details of the design of the redundant features of the crane and specific testing requirements for the crane are delineated in the Vermont Yankee document entitled "Reactor Building Crane Modification" (December 1975).

H. The Spent Fuel Pool Cooling System is designed to maintain the pool water temperature below 125'F during normal refueling operations. If the reactor core is 'completely discharged, the temperature of the pool water may increase to greater than 125 0 F, The RHR System supplemental fuel pool cooling may be used under these conditions to maintain the pool water temperature less than 150 0 F.

Amendment No. -24,--3-7, &9, -4, 141-,24, 2 31 239