View source for BVY 10-017, Technical Specifications Proposed Change No. 290 Administrative Changes

{{Adams
| number = ML101100462
| issue date = 04/13/2010
| title = Technical Specifications Proposed Change No. 290 Administrative Changes
| author name = Colomb M
| author affiliation = Entergy Nuclear Operations, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000271
| license number = DPR-028
| contact person = 
| case reference number = BVY 10-017
| document type = Letter type:BVY, Technical Specification, Bases Change
| page count = 17
}}

=Text=
{{#Wiki_filter:Entergy Nuclear Operations, Inc.
Vermont Yankee P.O. Box 0250 Entergy                                                    320 Governor Hunt Rd Vernon, VT 05354 Tel 802 257 7711 Michael J. Colomb Site Vice President BVY 10-017 April 13, 2010 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555

==SUBJECT:==
Technical Specifications Proposed Change No. 290 Administrative Changes Vermont Yankee Nuclear Power Station Docket No. 50-271 License No. DPR-28

==Dear Sir or Madam:==

In accordance with 10CFR50.90, Entergy Nuclear Operations, Inc. (ENO) is proposing an amendment to Operating License DPR-28 for the Vermont Yankee Nuclear Power Station (VY) by incorporating the attached proposed changes into the VY Technical Specifications (TS). The proposed changes are administrative in nature and do not materially change any technical requirements.
ENO has reviewed the proposed amendment in accordance with 10CFR50.92 and concludes it does not involve a significant hazards consideration. In accordance with 10CFR50.91, a copy of this application, with attachments, is being provided to the State of Vermont, Department of Public Service. to this letter provides a description and evaluation of the proposed changes. contains a markup of the current TS pages. Attachment 3 contains the retyped TS pages.
ENO requests review and approval of the proposed license amendment by May 1, 2011 with a 60 day implementation period from the date of the amendment approval.
There are no new regulatory commitments made in this letter.
If you have any questions on this transmittal, please contact Mr. James DeVincentis at 802-451-3150.
4oeo/

BVY 10-017 / Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct.
Executed on April 13, 2010.
Sincerely, MJC/PLC Attachments
: 1. Description and Evaluation of the Proposed Changes
: 2. Markup of the Current Technical Specifications Pages
: 3. Retyped Technical Specifications Pages cc:     Mr. Samuel J. Collins Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-14.15 Mr. James S. Kim, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 08C2A Washington, DC 20555 USNRC Resident Inspector Entergy Nuclear Vermont Yankee, LLC 320 Governor Hunt Rd Vernon, Vermont 05354 Mr. David O'Brien, Commissioner VT Department of Public Service 112 State Street - Drawer 20 Montpelier, Vermont 05620-2601

BVY 10-017 Docket No. 50-271 Attachment 1 Vermont Yankee Nuclear Power Station Proposed Change No. 290 Description and Evaluation of Proposed Changes

BVY 10-017 / Attachment 1 / Page 1 of 4
: 1.      

==SUMMARY==
DESCRIPTION The proposed amendment would revise the Vermont Yankee Nuclear Power Station (VY)
Technical Specifications (TS) by:
: 1. Updating the Table of Contents (TOC) as a result of changes made by License Amendments 230 and 239.
: 2. Updating the Applicability and Objective portions of TS 4.12 as result of changes made by License Amendment 239.
: 3. Revise TS 3.7.A.8 to revise wording that requires an "orderly shutdown be initiated immediately" if certain primary containment specifications "cannot" be met. The revised wording would still require that the reactor be in a cold shutdown condition within 24 hours if the applicable specifications are not met.
The proposed changes are considered administrative in nature and do not materially change any technical requirement. No corresponding TS Bases pages were found to be affected.
: 2.      DETAILED DESCRIPTION The following changes are proposed to the VY TS:
: 1. Revise the TOC to reflect that TS 3/4.6.1 Shock Suppressors and TS 3/4.12.G Crane Operability have been deleted. TS 3/4.6.1 was relocated to the VY Technical Requirements Manual (TRM), as authorized by License Amendment 230 (Reference (a)). TS 3/4.12.G was also relocated to the VY TRM, as authorized by License Amendment 239 (Reference (b)).
: 2. Revise the Applicability and Objective portions of TS 4.12 to remove reference to the testing and operability of the reactor building crane. License Amendment 239 (Reference (b)) authorized the relocation of the reactor building crane specifications, TS 3/4.12.G, to the VY TRM. Currently, the Applicability and Objective portions of TS 4.12 state:
Applicability:
Applies to the periodic testing of those interlocks and instruments used during refueling and to the testing of the reactor building crane.
Obiective:
To verify the operability of the instrumentation and interlocks used in refueling and the operability of the reactor building crane.
The proposed Applicability and Objective portions of TS 4.12 would be revised to state:
Applicability:
Applies to the periodic testing of those interlocks and instruments used during refueling.

BVY 10-017 / Attachment 1 / Page 2 of 4 Objective:
To verify the operability of the instrumentation and interlocks used in refueling.
: 3. Revise TS 3.7.A.8 to revise the wording that requires an immediate shutdown of the reactor be initiated if primary containment specifications TS 3.7.A.1 through 3.7.A.6 cannot be met.
Currently, TS 3.7.A.8 states:
If Specification 3.7.A.1 through 3.7.A.6 cannot be met, an orderly shutdown shall be initiated immediately and the reactor shall be in a cold shutdown condition within 24 hours.
Proposed TS 3.7.A.8 would be revised to state:
If Specification 3.7.A.1 through 3.7.A.6 are not met, the reactor shall be in a cold shutdown condition within 24 hours.
: 3.      TECHNICAL EVALUATION The proposed change to revise the TOC to reflect that TS 3/4.6.1 and TS 3/4.12.G are no longer contained in the TS is an administrative change as the relocation of TS 3/4.6.1 and TS 3/4.12.G to the VY TRM were previously approved by the NRC License Amendments 230 and 239 (References (a) and (b)), respectively. Similarly, the revision of the Applicability and Objective portions of TS 4.12.G to remove reference to the testing and operability of the reactor building crane is an administrative change as the relocation of TS 3/4.12.G to the VY TRM was approved by License Amendment 239 (Reference (b)). This was an oversight in the previous'amendments.'
VY identified these issues and entered them into the Corrective Action Process.
The proposed change to revise TS 3.7.A.8 would reword the requirement that an orderly shutdown be initiated immediately in the event that primary containment specifications 3.7.A.1 through 3.7.A.6 cannot be met.
The requirement to "immediately" initiate a reactor shutdown is considered unnecessary as it could lead to undesired manipulations of plant equipment in order to meet the intent of the word "immediately" in the current TS. Removing the requirement to immediately initiate a reactor shutdown would allow for an assessment of plant status and necessary planning of the required reactor shutdown to the cold shutdown condition within 24 hours. This is consistent with the approach used in the Standard Technical Specifications, NUREG-1433, (Reference (c)) for defining conditions and required actions for an inoperable primary containment.
Also, replacing the word "cannot" with "are not" is consistent with the intent of 10CFR50.36(c)(2) which uses the term "is not" for defining actions to be taken when an LCO "is not" met ("are not" is the plural of "is not"). This change makes the TS consistent with the intent of the regulation.
These are considered administrative changes.
: 4.       EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION The proposed amendment would revise Vermont Yankee Nuclear Power Station (VY) Technical Specifications (TS) by updating the Table of Contents and Applicability and Objective portions of TS 4.12 to reflect previous changes made by License Amendments 230 and 239. The proposed amendment would also revise TS 3.7.A.8 to reword the requirement to immediately initiate a

BVY 10-017 / Attachment 1 / Page 3 of 4 reactor shutdown      if certain primary containment specifications cannot be met. The revised wording would still require   that the reactor be in a cold shutdown condition within 24 hours if the applicable specifications are    not met. These changes are considered to be administrative in ,nature and do not materially change     any technical requirements.
Pursuant to 10CFR50.92, Entergy Nuclear Operations, Inc. (ENO) has reviewed the proposed change and concludes that the changes do not involve a significant hazards consideration since the proposed changes satisfy the criteria in 10CFR50.92(c). These criteria require that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The discussion below addresses each of these criteria and demonstrates that the proposed amendment does not constitute a significant hazard.
The proposed changes do not involve a significant hazards consideration because:
: 1.      The operation of Vermont Yankee Nuclear Power Station (VY) in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes are administrative in nature and do not involve any physical changes to the plant. The changes do not revise the methods of plant operation which could increase the probability or consequences of accidents. No new modes of operation are introduced.by the proposed changes such that a previously evaluated accident is more likely to occur or more adverse consequences would result.
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2.      The operation of VY in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes are administrative in nature and do not affect the operation of any systems or equipment, nor do they involve any potential initiating events that would create any new or different kind of accident. There are no changes to the design assumptions, conditions, configuration of the facility, or manner in which the plant is operated and maintained. The changes do not affect assumptions contained in plant safety analyses or the physical design and/or modes of plant operation.
Consequently, no new failure mode is introduced.
Therefore, the proposed changes do not create the' possibility of a new or different kind of accident from any previously evaluated.
: 3.      The operation of VY in accordance with the proposed amendment will not involve a siqnificant reduction in a margin of safety.
There are no changes being made to the Technical Specification (TS) safety limits or safety system settings. The operating limits and functional capabilities of systems, structures and components are unchanged as a result of these administrative changes. These changes do not affect any equipment involved in

BVY 10-017 / Attachment 1 / Page 4 of 4 potential initiating events or plant response to accidents. There is no change to the basis for any TS related to the establishment, or maintenance of, a nuclear safety margin. The proposed changes do not impact any safety limits, safety settings or safety margins.
Therefore, operation of VY in accordance with the proposed amendment will not involve a significant reduction in the margin to safety.
Based on the above, ENO concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10CFR50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.
: 5.       ENVIRONMENTAL CONSIDERATIONS This amendment request meets the eligibility criteria for categorical exclusion from environmental review set forth in 10CFR51.22(c)(9) as follows:
(i)      The amendment involves no significant hazards determination.
As described in this evaluation, the proposed change involves no significant hazards consideration.
(ii)     There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.
The proposed amendment does not involve any physical alterations to the plant configuration that could lead to a change in the type or amount of effluent release offsite.
(iii)    There is no significant increase in individual or cumulative occupational radiation exposure.
The proposed amendment does not involve a significant increase in individual or cumulative occupational radiation exposure.
Based on the above, ENO concludes that the proposed change meets the eligibility criteria for categorical exclusion as set forth in 10CFR51.22(c)(9). Pursuant to 10CFR51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
: 6.       REFERENCES a) Letter, USNRC to Entergy, "Vermont Yankee Nuclear Power Station - Issuance of Amendment RE: Adoption of Technical Specification Task Force (TSTF) Change TSTF-372, The Addition of Limiting Condition for Operation (LCO) 3.0.8 on the Inoperability of Snubbers (TAC No. MD 1664)," NVY 07-028, dated March 26, 2007 b) Letter, USNRC to Entergy, "Vermont Yankee Nuclear Power Station- Issuance of Amendment RE: Relocation of Reactor Building Crane Technical Specification (TAC No.
MD9725)," NVY 09-077, dated July 13, 2009 c) NUREG-1433 "Standard Technical Specifications General Electric Plants, BWR/4,"
Revision 3

BVY 10-017 Docket 50-271 Attachment 2 Vermont Yankee Nuclear Power Station Proposed Change No. 290 Markup of the Current Technical Specifications Pages

VYNPS TABLE OF CONTENTS (Continued)
LIMITING CONDITIONS OF OPERATION                                                        Page No.      SURVEILLANCE 3.4   REACTOR STANDBY LIQUID CONTROL SYSTEM .........                                             92   ...      4.4 A. Normal Operation ..........................                                             92   ...        A B. Operation with Inoperable Components ......                                             93   ...         B C. Standby Liquid Control System Tank-Borated Solution .................................                                              93   ...         C BASES                                                                                       97 3.5   CORE AND CONTAINMENT COOLING SYSTEMS ............                                           99   ...      4.5 A. Core Spray and Low Pressure Coolant Injection .................................                                             99   ...         A B. Containment Spray Cooling Capability ......                                            102   ...         B C. Residual Heat Removal (RHR) Service Water System ..............................                                            103   ...         C D. Station Service Water and Alternate Cooling Tower Systems ........................                                         104   ...         D E. High Pressure Coolant Injection (HPCI)
System .....................................                                     .... 105    ...         E F. Automatic Depressurization System .........                                            106   ...         F G. Reactor Core Isolation Cooling System (RCIC) . ...................................                                          107   ...         G H. Minimum Core and Containment Cooling System Availability                      ...........................                   108   ...         H I. Maintenance of Filled Discharge Pipe .......                                           109   ...         I BASES                                                                                      110 3.6   REACTOR COOLANT SYSTEM ............................                                        115   ...       4.6 A. Pressure and Temperature Limitations ......                                            115   ...         A B. Coolant Chemistry .............................                                        116   ...         B C. Coolant Leakage ............................                                           118   ...         C D. Safety and Relief Valves .....................                                         120   ...         D E. Structural Integrity and Operability Testing         .................................                                      120   ...         E F. Jet Pumps .................................                                            121   ...         F G. Single Loop Operation ........................                                         122 H. Recirculation System .........................                                         126
* I. eh*ls  .upprss..                            ..........................128                             I J. hermal Hydraulic Stability ..................                                        134   ...         J BASES                                                                                      138 3.7   STATION CONTAINMENT SYSTEMS ......................                                         146   ...       4.7 A. Primary Containment .......................                                           146   ...         A B. Standby Gas Treatment System.' ..............                                          152    ...        B C. Secondary Containment System ........                                     ......       155   ...         C D. Primary Containment Isolation Valves .....                                            158   ...         D E. Reactor Building Automatic Ventilation System Isolation Valves (RBAVSIVs) ........                                           158a    ...        E BASES                                                                                      163 Amendment No. 4-,   4&4,        4-, 95, 4-94, 2--,                           2                                  -ii-

VYNPS
                                                        -TABLE OF CONTENTS (Continued)
LIMITING CONDITIONS OF OPERATION                                            Page No. SURVEILLANCE 3.12  REFUELING AND SPENT FUEL HANDLING .............                                229       4.12 A. Refueling Interlocks ......................                                229         A B. Core Monitoring ...........................                                230         B C. Fuel Storage Pool Water Level .............                                231         C D. Control Rod and Control Rod Drive Maintenance ...............................                                232         D E. Extended Core Maintenance .................                                233         E F. Fuel Movement .............................                                235         F I       G.                          biltr . ... .. .. ... .. .. ..                     235         G H. (Spent Fuel Pool Water Temperature .........                                236         H BASES                                                                          237 3.13  Deleted .......................................                                240 5.0   DESIGN FEATURES ...............................                                253 6.0   ADMINISTRATIVE CONTROLS .......................                                255 6.1   RESPONSIBILITY ................................                                 255 6.2   ORGANIZATION ..................................                                 255 6.3   ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED.................................                            ...... 257 6.4   PROCEDURES ....................................                                 257 6.5   HIGH RADIATION AREA ...........................                                 257 6.6   REPORTING REQUIREMENTS ........................                                 258 6.7   PROGRAMS AND MANUALS ..........................                                 262 Amendment No. 3-3, 4-3,            6-,       84, 4&,     4-&&, a-94,,--v                           .- iv-

VYNPS 3.7 LIMITING CONDITIONS            FOR                  4.7  SURVEILLANCE REQUIREMENTS OPERATION ii. 24 Hours   prior  to reducing thermal power to less than 15% rated thermal power prior to the next shutdown.
: 8. If    Specification        3.7.A.1 through 3.7.A.6 e                         are AO met, a-: oheFl-y
                   &;d      the reactor     shall  be in a cold shutdown condition within 24 hours.
: 9. If Specification 3.7.A.7 cannot be met, and the
  )                primary containment oxygen concentration cannot be restored to less than 4% oxygen by volume within the subsequent 24 hour period, reactor thermal power shall be less than 49                 15% rated thermal power within the next 8 hours.
: 10. Drywell/Suppression Chamber d/p                                        10. Drywell/Suppression Chamber d/p.
: a.         Differential pressure. between                           a;   The! differential the drywell and                                  pressure between suppression chamber                              the drywell and shall be maintained                              suppression chamber
                              >1.7 psid while in                               shall be recorded the RUN MODE during                              once per shift.
the time period:.
2 i.From 24 hours after thermal, power is greater                         b. The operability of than 15% rated                                the low thermal power                                 differential following                                     pressure alarm startup, to                                   shall be verified once per week.
ii. 24 hours prior    to reducing thermal power to less than 15% rated thermal power prior to the next shutdown, iii.      Except   as<.
specified in 3.7.A.10.b.
Amendment  No. 4,4i i                                                                    L51 L2

VYNPS 3.12 LIMITING CONDITIONS FOR                 4.12 SURVEILLANCE REQUIREMENTS OPERATION 3.12 REFUELING AND SPENT FUEL HANDLING       4.12 REFUELING AND SPENT FUEL HANDLING Applicability:                               Applicability:
Applies to fuel handling, core               Applies to the periodic testing reactivity limitations, and spent            of those interlocks and fuel handling.                               instruments used during refueling.
I Objective:                                   Objective:
To assure core reactivity is                 To verify the operability of within capability of the control             instrumentation and interlocks rods, to prevent criticality                 used in refueling-during refueling, and to assure safe handling of spent fuel I
casks.
Specification:                               Specification:
A. Refueling Interlocks                     A. Refueling Interlocks The reactor mode switch shall                 Prior to any fuel handling, be locked in the "Refuel"                     with the Head off the reactor position during core                          vessel, the following alterations and;                              required refueling interlock inputs shall be functionally
: 1. The refueling interlocks                  tested once every 7 days:
shall be operable during                                                 )
(                in-vessel fuel movement                   a. All-rods-in;               (
for the equipment utilized                b. Refuel platform position;   )
  )
in moving fuel.
If one or more of the c.
d.
Refuel platform fuel grapple, fuel loaded7 Refuel platform frame K
required refueling                           mounted hoist, fuel interlocks are inoperable;                   loaded;
: e. Refuel platform monorail
(                Immediately suspend fuel                     mounted hoist, fuel movement with equipment                      loaded.
associated with the inoperable interlock(s),
                   -or-Immediately insert a control rod withdrawal block and verify all
(                control rods are fully inserted.
: 2. The refueling interlocks shall be operable except as specified in Specification 3.12.D and 3.12.E.
Amendment No.   -14, 24, '22                                                   229

BVY 10-017 Docket 50-271 Attachment 3 Vermont Yankee Nuclear Power Station Proposed Change No. 290 Retyped Technical Specifications Pages

VYNPS
                                                       *TABLE OF CONTENTS (Continued)
LIMITING CONDITIONS OF OPERATION                                     Page No. SURVEILLANCE 3.4   REACTOR STANDBY LIQUID CONTROL SYSTEM .........                          92   ...      4.4 A. Normal Operation ..........................                          92             A B. Operation with Inoperable Components ......                          93             B C. Standby Liquid Control System Tank-Borated Solution .................................                           93             C BASES                                                                    97 3.5   CORE AND CONTAINMENT COOLING SYSTEMS ..........                          99            4.5 A. Core Spray and Low Pressure Coolant Injection . ................................                        99             A B. Containment Spray Cooling Capability ......                         102             B C. Residual Heat Removal (RHR) Service Water System ...............................                        103             C D. Station Service Water and Alternate Cooling Tower Systems ......................                        104             D E. High Pressure Coolant Injection (HPCI)
System ....................................                         105             E F. Automatic Depressurization System ..........                        106              F G. Reactor Core Isolation Cooling System (RCIC) ....................................                        107             G H. Minimum Core and Containment Cooling System Availability .......................                         108             H I. Maintenance of Filled Discharge Pipe ......                         109              I BASES                                                                   110 3.6   REACTOR COOLANT SYSTEM ........................                         115            4.6 A. Pressure and Temperature Limitations ......                        115             A B. Coolant Chemistry .........................                         116             B C. Coolant Leakage ...........................                         118             C D. Safety and Relief Valves ..................                         120              D E. Structural Integrity and Operability Testing          .................................                  120             E F. Jet Pumps .................................                        121              F G. Single Loop Operation .....................                        122 H. Recirculation System ......................                         126 I. Deleted ...................................                        128              I J. Thermal Hydraulic Stability ...............                        134              J BASES                                                                   138 3.7   STATION CONTAINMENT                   SYSTEMS ...................       146   ...      4.7 A. Primary Containment .......................                        146             A B. Standby Gas Treatment System ..............                        152              B C. Secondary Containment System ..............                        155              C D. Primary Containment Isolation Valves ......                        158              D E. Reactor Building Automatic Ventilation System Isolation Valves (RBAVSIVs) ........                       158a              E BASES                                                                  163 Amendment No. 3,     -,        -3,     4-5,       4-9-3,   -,*,    -4                       -ii-

VYNPS TABLE OF CONTENTS (Continued)
LIMITING CONDITIONS OF OPERATION                                        Pa ge No. SURVEILLANCE 3.12 REFUELING AND SPENT FUEL HANDLING .............                             229       4.12 A. Refueling Interlocks ......................                            229         A B. Core Monitoring ...........................                            230         B C. Fuel Storage Pool Water Level .............                           231         C D. Control Rod and Control Rod Drive Maintenance ...............................                            232         D E. Extended Core Maintenance .................                            233         E F. Fuel Movement .............................                           235         F G. Deleted ...................................                           235         G H. Spent Fuel Pool Water Temperature .........                           236         H BASES                                                                       237 3.13  Deleted .......................................                            240 5.0   DESIGN FEATURES ...............................                            253 6.0  ADMINISTRATIVE               CONTROLS .......................               255 6.1   RESPONSIBILITY ................................                             255 6.2   ORGANIZATION ..................................                             255 6.3  ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED ......................................                             257 6.4   PROCEDURES ....................................                             257 6.5   HIGH RADIATION AREA ...........................                             257 6.6   REPORTING REQUIREMENTS ........................                             258 6.7   PROGRAMS AND MANUALS ..........................                             262 Amendment No.     -34, 4-3 , 6&-, 8-,                 45,  -i*,    4-94, 2-4                  iv

VYNPS 3.7 LIMITING CONDITIONS FOR                       4.7 SURVEILLANCE REQUIREMENTS OPERATION ii. 24 Hours prior to reducing thermal power to less than 15% rated thermal power prior to the next shutdown.
: 8. If Specification 3.7.A.1 through 3.7.A.6 are not met, the reactor shall be in a cold shutdown condition within 24 hours.
: 9. If Specification 3.7.A.7 cannot be met, and the primary containment oxygen concentration cannot be restored to less than 4% oxygen by volume within the subsequent 24 hour period, reactor thermal power shall be less than 15% rated thermal power within the next 8 hours.
: 10. Drywell/Suppression Chamber d/p
: a. Differential                           10. Drywell/Suppression pressure between                            Chamber d/p the drywell and suppression chamber                         a. The differential shall be maintained                              pressure between
                   >1.7 psid while in                               the drywell and The RUN MODE during                              suppression chamber the time period:                                 shall be recorded once per shift.
i.From 24 hours after thermal power is greater than 15% rated                         b. The operability of thermal power                               the low following                                  differential startup, to                                pressure alarm shall be verified ii.      24 hours prior                         once per week.
to reducing thermal power to less than 15%
rated thermal power prior to the next shutdown, iii. Except as specified in 3.7.A.10.b.
Amendment No.      2-3                                                          151

VYNPS 3.12 LIMITING CONDITIONS FOR                  4.12 SURVEILLANCE REQUIREMENTS OPERATION 3.12 REFUELING AND SPENT FUEL HANDLING        4.12 REFUELING AND SPENT FUEL HANDLING Applicability:                                Applicability:
Applies to fuel handling, core                Applies to the periodic testing reactivity limitations, and spent             of those interlocks and fuel handling.                                instruments used during refueling.
Objective:                                    Objective:
To assure core reactivity is                  To verify the operability of within capability of the control              instrumentation and interlocks rods, to prevent criticality                  used in refueling.
during refueling, and to assure safe handling of spent fuel casks.
Specification:                                Specification:
A. Refueling Interlocks                     A. Refuelina Interlocks The reactor mode switch shall                 Prior to any fuel handling, be locked in the "Refuel"                     with the Head off the reactor position during core                          vessel, the following alterations and;                              required refueling interlock inputs shall be functionally
: 1. The refueling interlocks                   tested once every 7 days:
shall be operable during in-vessel fuel movement                   a. All-rods-in; for the equipment utilized                b. Refuel platform position; in moving fuel.                           c. Refuel platform fuel grapple, fuel loaded; If one or more of the                     d. Refuel platform frame required refueling                           mounted hoist, fuel interlocks are inoperable;                    loaded;
: e. Refuel platform monorail Immediately suspend fuel                     mounted hoist, fuel movement with equipment                        loaded.
associated with the inoperable interlock(s),
                -or-Immediately insert a control rod withdrawal block and verify all control rods are fully inserted.
: 2. The refueling interlocks shall be operable except as specified in Specification 3.12.D and 3.12.E.
Amendment No. 4.-, -24, 242                                                      229}}