ML050670662
| ML050670662 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/25/2005 |
| From: | Grecheck E Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 05-009 | |
| Download: ML050670662 (194) | |
Text
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Dominon Rope Ferry Road Waterford, CT 06385 February 25, 2005 U.S. Nuclear Regulatory Commission Serial No. 05-009 Attention: Document Control Desk MPS Lic/MAE RO One White Flint North Docket Nos. 50-336 11555 Rockville Pike 50-423 Rockville, MD 20852-2738 License Nos. DPR-65 NPF-49 DOMINION NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION UNITS 2 AND 3 LICENSE AMENDMENT REQUESTS (LBDCR 04-MP2-015, LBDCR 04-MP3-014) ADMINISTRATIVE CHANGES IN TECHNICAL SPECIFICATIONS Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) hereby requests to amend Operating License DPR-65 for Millstone Power Station Unit 2 (MPS2) and Operating License NPF-49 for Millstone Power Station Unit 3 (MPS3) by incorporating certain administrative changes into the MPS2 and MPS3 technical specifications. The proposed amendments do not involve a significant impact on public health and safety and do not involve a significant hazards consideration pursuant to the provisions of 10 CFR 50.92 (see Significant Hazards Consideration in Attachments 1 and 4). The Site Operations Review Committee and the Management Safety Review Committee have reviewed and concurred with the determinations. DNC requests issuance of the amendments for MPS2 and MPS3 prior to December 31, 2005, with the amendment to be implemented within 90 days of issuance. In accordance with 10 CFR 50.91(b), copies of the license amendment requests are being provided to the State of Connecticut. If you have any questions or require additional information, please contact Mr. Paul R. Willoughby at (804) 273-3572. Very truly yours, Eugene S. Grecheck Vice President - Nuclear Support Services
Serial No. 05-009 Docket Nos. 50-336/50-423 Administrative Changes In Technical Specifications Page 2 of 3 Attachments: (6)
- 1. Evaluation of MPS2 Proposed License Amendment
- 2. MPS2 Marked-up Pages
- 3. MPS2 Re-typed Pages
- 4. Evaluation of MPS3 Proposed License Amendment
- 5. MPS3 Marked-up Pages
- 6. MPS3 Re-typed Pages Commitments made in this letter: None.
cc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 Mr. V. Nerses Senior Project Manager Millstone Unit 2 U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8C2 Rockville, MD 20852-2738 Mr. G. F. Wunder Project Manager Millstone Unit 3 U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8B1A Rockville, MD 20852-2738 Mr. S. M. Schneider NRC Senior Resident Inspector Millstone Power Station
Serial No. 05-009 Docket Nos. 50-336/50-423 Administrative Changes InTechnical Specifications Page 3 of 3 COMMONWEALTH OF VIRGINIA ) COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Eugene S. Grecheck, who is Vice President - Nuclear Support Services, of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief. Acknowledged before me this 2 day of Amae 2005. My Commission Expires:( 3/ Notary Noayublic lc I . - (SEAL) .-
Serial No. 05-009 Docket Nos. 50-336 ATTACHMENT 1 LICENSE AMENDMENT REQUEST (LBDCR 04-MP2-015) ADMINISTRATIVE CHANGES TO TECHNICAL SPECIFICATIONS EVALUATION OF PROPOSED LICENSE AMENDMENT MILLSTONE POWER STATION UNIT 2 DOMINION NUCLEAR CONNECTICUT, INC.
Serial No. 05-009 Docket Nos. 50-336 Administrative Changes Attachment 1 Page 1 of 7 EVALUATION OF PROPOSED LICENSE AMENDMENT
1.0 DESCRIPTION
2.0 PROPOSED CHANGE
3.0 BACKGROUND
3.1 Recent Improvements in Processing and Implementing Technical Specification Changes at the Millstone Power Station 3.2 Reason for Proposed Amendment
4.0 TECHNICAL ANALYSIS
5.0 REGULATORY ANALYSIS
No Significant Hazards Consideration
6.0 ENVIRONMENTAL CONSIDERATION
Serial No. 05-009 Docket Nos. 50-336 Administrative Changes Attachment 1 Page 2 of 7
1.0 DESCRIPTION
Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) hereby requests to amend Operating License DPR-65 for Millstone Power Station Unit 2 (MPS2) by incorporating certain administrative changes into Millstone Unit 2 technical specifications (TS). The proposed changes include:
- 1. Changes to index pages to correct inconsistencies in titles and page numbering, which were inadvertently introduced in previous amendments.
- 2. The use of a capitalized typeface for all the DEFINED TERMS specified in Section 1.0, "DEFINITIONS," of the MPS2 TS. This requirement is specified as item 1.1 in this section.
- 3. Deletion of references to certain technical specifications which no longer exist in MPS2 TS.
- 4. Clarification of a structure name.
2.0 PROPOSED CHANGE
S Changes To Index Pages The following changes will make the index page numbering consistent with TS page numbers.
- 1. Index Page IV: The title "Reactivity Balance" corresponding to page 3/4 1-2 will be replaced with the word "DELETED," and the word "DELETED" corresponding to page 3/4 1-3 will be replaced with the title "Reactivity Balance."
- 2. Index Page V: Capitalized typeface will be used in Specification 3/4.4.1 titles for the TS DEFINED TERMS in accordance with Section 1.1 of Millstone Unit 2 TS.
- 3. Index Page XI: The page number corresponding to TS 3/4.3.3 will be changed from B 3/4 3-2 to B 3/4 3-2a, and the page number corresponding to TS 3/4.3.4 will be changed from B 3/4 3-5 to B 3/4 3-6.
- 4. Index Page XII: The page number corresponding to TS 3/4.4.2 will be changed from B 3/4 4-1 to B 3/4 4-1d, the page number corresponding to TSs 3/4.5.2 and 3/4.5.3 will be changed from B 3/4 5-1 to B 3/4 5-2, page number corresponding to TS 3/4.5.4 will be changed from B 3/4 5-2 to B 3/4 5-2d, and the page number corresponding to TS 3/4.6.3 will be changed from B 3/4 6-3 to B 3/4 6-3a.
Serial No. 05-009 Docket Nos. 50-336 Administrative Changes Attachment 1 Page 3 of 7
- 5. Index Page XIII: The page number corresponding to TS 3/4.7.3 will be changed from B 3/4 7-3 to B 3/4 7-3a, the page number corresponding to TS 3/4.7.4 will be changed from B 3/4 7-4 to B 3/4 7-3c, and the page number corresponding to TSs 3/4.9.3, 3/4.9.4, 3/4.9.5 will be changed from B 3/4 9-1 to B 3/4 9-1 a.
- 6. Index Page XIV: The page number corresponding to TSs 3/4.9.11 and 3/4.9.12 will be changed from B 3/4 9-2a to B 3/4 9-2b.
- 7. Index Page XV: The page number corresponding to FUEL STORAGE, Drainage and Capacity will be changed from 5-5 to 5-5a.
- 8. Index Page XVII: The page number corresponding to Section 6.9.1, CORE OPERATING LIMITS REPORT will be changed from 6-18 to 6-18a. Also, a misspelling of the word DIESEL is corrected in the title of section 6.24.
Changes Involving Capitalization of The TS DEFINED TERMS The following changes will capitalize the typeface for all the DEFINED TERMS as specified in Section 1.0, "DEFINITIONS," of the MPS2 TS. The requirement to capitalize the typeface of these terms is specified in section 1.1 of the TS.
- 1. Page 1-8: Capitalize the terms "VENTING" and "SITE BOUNDARY."
- 2. Page 3/4 1-20: Capitalize the term "ACTION."
- 3. Page 3/4 1-31: Capitalize the term "OPERABLE."
- 4. Page 3/4 1-19: Capitalize the term "MODE."
- 5. Page 3/4 2-10: Capitalize the term "ACTIONS."
- 6. Page 3/4 3-1: Capitalize the term "MODES."
- 7. Page 3/4 3-4: Capitalize the term "OPERABLE."
- 8. Page 3/4 3-9: Capitalize the term "MODES."
- 9. Page 3/4 3-23: Capitalize the term "OPERABILITY."
- 10. Page 3/4 3-24: Capitalize the term "MODES."
- 11. Page 3/4 3-31: Capitalize the term "ACTIONS." Also, replace the word "INSTALLATION" in the title with the word "INSTRUMENTATION", which is the correct title for this section.
- 12. Page 3/4 3-32: Capitalize the terms "ACTION" and "OPERABLE."
- 13. Page 3/4 3-33: Capitalize the term "ACTION."
- 14. Page 3/4 4-7f: Capitalize the term "ACTION."
- 15. Page 3/4 4-16: Capitalize the term "DOSE EQUIVALENT."
- 16. Page 3/4 6-12: Capitalize the term "ACTION."
- 17. Page 3/4 7-9b: Capitalize the term "ACTION" and add titles at top of page.
- 18. Pages 3/4 8-1, 8-1 a, 8-2, and 8-2a: Capitalize the term "ACTION."
- 19. Page 5-1: Capitalize the term "SITE BOUNDARY."
- 20. Page 6-4: Capitalize the terms "MODES", "COLD SHUTDOWN" and "REFUELING," and replace the word "supervision" with "supervising."
- 21. Page 6-19: Capitalize the term "SHUTDOWN MARGIN."
- 22. Page 6-26: Capitalize the term "UNRESTRICTED AREAS."
Serial No. 05-009 Docket Nos. 50-336 Administrative Changes Attachment 1 Page 4 of 7
- 23. Page 6-27: Capitalize the terms "UNRESTRICTED AREAS" and "SITE BOUNDARY."
Changes Involving Deletion of References to Certain Technical Specifications, Which no Longer Exist in MPS2 TS The following changes will delete references to certain TSs which, have been deleted from the MPS2 TS.
- 1. Page 3/4 3-5: The wording "or 3.1.1.2, as applicable" will be deleted. TS 3.1.1.2, which originally addressed SHUTDOWN MARGIN for Tavg < 200Q F, was deleted by Amendment 280 and replaced with another TS 3.1.1.2, "Reactivity Balance." Therefore, the reference in ACTION 4 to the old TS 3.1.1.2 is not relevant and will be deleted.
- 2. Page 3/4 4-1: In the APPLICABILITY section of TS 3.4.4.1 the Asterisk and associated footnote will be deleted. The footnote references TS 3.10.4, which was deleted by Amendment 280. Therefore, the reference in the footnote to TS 3.10.4 is not relevant and will be deleted.
- 3. Page 3/4 4-7b: In section 4.4.5.1.5.c the wording "In lieu of any report required pursuant to Specification 6.6.1," will be deleted. Amendment 239 deleted section 6.6. Therefore, the reference in the section 4.4.5.1.5.c to Specification 6.6.1 is not relevant and will be deleted.
Change Involving Clarification of a Structure Name Page 3/4 6-1: In surveillance requirement SR 4.6.1.1.e, the word "containment" will be added before "structural integrity." This change clarifies that containment structure integrity is verified in accordance with the Containment Tendon Surveillance Program.
3.0 BACKGROUND
3.1 Recent Improvements in Processing and Implementing Technical Specification Changes at the Millstone Power Station. DNC has employed a new system relying on Adobe FrameMaker software for implementing and tracking changes to the TSs. As part of the conversion to the new system, the MPS2 TS were reviewed for inconsistencies and errors that are administrative in nature and that were introduced as part of the implementation of previous license amendments.
Serial No. 05-009 Docket Nos. 50-336 Administrative Changes Attachment 1 Page 5 of 7 3.2 Reason for Proposed Amendment The proposed amendment is being requested to remove inconsistencies and administrative errors that were inadvertently introduced during the implementation of the previous license amendments that modified the MPS2 TS.
4.0 TECHNICAL ANALYSIS
Safety Summary The proposed changes include changes to index pages to correct inconsistencies in titles and page numbering, use of a capitalized typeface for all the DEFINED TERMS specified in Section 1.0, "DEFINITIONS," of the MPS2 TS, deletion of references to certain TS that no longer exist in MPS2 TS, and a clarification of a structure name. The proposed changes will remove inconsistencies and administrative errors that were inadvertently introduced during the implementation of previous license amendments that modified the MPS 2 TS. These changes are administrative in nature and do not alter any of the requirements of the affected TS. These changes do not alter any of the assumptions used in the safety analyses, nor do they cause any safety system parameters to exceed their acceptance limit. Therefore, the proposed changes have no adverse effect on plant safety. Additionally, these changes can be made without adverse impact to plant operations or to the health and safety of the public.
5.0 REGULATORY ANALYSIS
No Significant Hazards Consideration DNC has evaluated whether or not a significant hazards consideration (SHC) is involved with the proposed changes by addressing the three standards set forth in 10 CFR 50.92(c) as discussed below. Criterion 1: Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes are administrative in nature and do not alter any of the requirements of the affected TS. The proposed changes do not modify any plant equipment and do not impact any failure modes that could lead to an accident. Additionally, the proposed changes have no effect on the consequence of any analyzed
Serial No. 05-009 Docket Nos. 50-336 Administrative Changes Attachment 1 Page 6 of 7 accident since the changes do not affect any equipment related to accident mitigation. Based on this discussion, the proposed amendment does not increase the probability or consequences of an accident previously evaluated. Criterion 2: Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes are administrative in nature. They do not modify any plant equipment and there is no impact on the capability of the existing equipment to perform their intended functions. No system setpoints are being modified and no changes are being made to the method in which plant operations are conducted. No new failure modes are introduced by the proposed changes. The proposed amendment does not introduce accident initiators or malfunctions that would cause a new or different kind of accident. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. Criterion 3: Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. These changes are administrative in nature, which do not alter any of the requirements of the affected TS. The proposed changes do not affect any of the assumptions used in the accident analysis, nor do they affect any operability requirements for equipment important to plant safety. Therefore, the proposed changes will not result in a significant reduction in the margin of safety as defined in the. bases for technical specifications covered in this license amendment request. In summary, DNC concludes that the proposed amendment does not represent a significant hazards consideration under the standards set forth in 10 CFR 50.92(c).
6.0 ENVIRONMENTAL CONSIDERATION
DNC has determined that the proposed amendment would not change requirements with respect to use of a facility component located within the restricted area, as defined by 10 CFR 20, nor would it change inspection or surveillance requirements. DNC has evaluated the proposed change and has determined that the change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released off site, or (iii) a significant
Serial No. 05-009 Docket Nos. 50-336 Administrative Changes Attachment 1 Page 7 of 7 increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51 .22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
Serial No. 05-009 Docket No. 50-336 ATTACHMENT 2 LICENSE AMENDMENT REQUEST (LBDCR 04-MP2-015) ADMINISTRATIVE CHANGES TO TECHNICAL SPECIFICATIONS MARKED-UP PAGES MILLSTONE POWER STATION UNIT 2 DOMINION NUCLEAR CONNECTICUT, INC.
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY.........................................................................................................3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 REACTIVITY CONTROL ............................................ 3/4 1-1 in (SDM) ............................................ 3/4 1-1 ce ............................................ 3/4 1-2
. ... ............................................. 3/4 1-3 Bor D.~ion .................................. 3/41-4 Moderator Temperature Coefficient (MTC) ............. ................... 3/4 1-5 Minimum Temperature for Criticality ................................ 3/4 1-7 3/4.1.2 BORATION SYSTEMS . 3/41-8 DELETED .34 1-8 DELETED ......... 3/4 1-9 DELETED ...... 3/4 1-11 DELETED ...... 3/41-13 DELETED ...... 3/4 1-14 DELETED ...... 3/4 1-15 DELETED ...... 3/4 1-16 DELETED ...... 3/4 1-18 3/4.1.3 MOVABLE CONTROL ASSEMBLIES .3/4 1-20 CEA Group Position .3/4 1-20 Position Indicator Channels .3/4 1-24 CEA Drop Time .3/4 1-26 Shutdown CEA Insertion Limit .3/4 1-27 Regulating CEA Insertion Limits .3/4 1-28 Control Rod Drive Mechanisms. 3/4 1-31 MILLSTONE - UNIT 2 IV Amendment No. i-&,
404,446, i53, 4&, M, ie 5
/
INDEX Septembcr 25,2003 t LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE ................................................. 3/4 2-1 3/4.2.2 Deleted 3/4.2.3 TOTAL INTEGRATED RADIAL PEAKING FACTOR - Fr .................. 3/4 2-9 3/4.2.4 AZIMUTHAL POWER TILT .................. ............................... 3/4 2-10 3/4.2.5 Deleted 3/4.2.6 DNB MARGIN ......................................... 3/4 2-13 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION ..................................... 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION.................................................................................. 3/4 3-9 3/4.3.3 MONITORING INSTRUMENTATION ............................................... 3/4 3-24 Radiation Monitoring ............................................... 3/4 3-24 Remote Shutdown Instrumentation ........................ ....................... 3/4 3-28 Accident Monitoring ............................................... 3/4 3-31 3/4.3.4 CONTAINMENT PURGE VALVE ISOLATION SIGNAL ....................... 3/4 3-36 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION ... 3/4 4-1 Startup and Power Operation .. . 3/4 4-1 d ioHt S tand 3/4
/.44 4-la l............a..............
oLP s Hot Shutdown ................. ............................................... 3/4 4-lb old Shutdown. Reactor Coolant System Loops Filled ..... 3/4 4-id Cold Shutdown Reactor Coolant System Loops Not Filled .....
. 3/4 4-If
\'ReactorC6olant Pumps -'Cold Shutdown .... 3/4 4-l h Amendment No. 35, 38,66,69, 99, 04, MILLSTONE - UNIT 2 V 139, 4-3, 4, 479, 237, 245,249,250,
BASES SECTION PAGE 3/4.0 APPLICABILITY ...................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL ...................................... B 3/4 1-1 3/4.1.2 Deleted 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................................... B 3/4 1-2 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE ............................................ B 3/4 2-1 3/4.2.2 Deleted 3/4.2.3 TOTAL INTEGRATED RADIAL PEAKING FACTOR - lb ............. B 3/4 2-1 3/4.2.4 AZIMUTHAL POWER TILT ............... ............................. B 3/4 2-1 3/4.2.5 Deleted 3/4.2.6 DNB MARGIN .......................................................................................... B 3/4 2-2 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION ........................................... B 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION ................ B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION ............................................ B 3/4 3 G 3/4.3.4 CONTAINMENT PURGE VALVE ISOLATION SIGNAL ................... B 3/4 3-Y 6 MILLSTONE - UNIT 2 Xi Amendment No. 3-, 494,4-04,4-39,4845, b4 H 25 )
INDEX BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION .......................... B 3/4 4-1 3/4.4.2 SAFETY VALVES ........................................ B 3/4ED 3/4.4.3 RELIEF VALVES ........................................ B 3/4 4-2 3/4.4.4 PRESSURIZER ........................................ B 3/4 4-2a 3/4.4.5 STEAM GENERATORS ........................................ B 3/4 4-2a 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE ....................................... B 3/4 4-3 3/4.4.7 DELETED ............................. B 3/4 4-4 ( 3/4.4.8 SPECIFIC ACTIVITY............................................................................... B 3/4 4-4 3/4.4.9 PRESSURE/TEMPERATURE LIMITS.................................................... B 3/4 4-5 3/4.4.10 DELETED ............................ B 3/4 4-7 3/4.4.11 DELETED ............................ B 3/4 4-8 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS .......... B 3/4 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS ......... . B 3/4.5.4 REFUELING WATER STORAGE TANK (RWST) ........... B 3/4.5.5 TRISODIUM PHOSPHATE (TSP) .......... B 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT ...................................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS .............................. B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES ...................................... B 3/4 6() 3/4.6.4 COMBUSTIBLE GAS CONTROL........................................................... B 3/4 6-4 3/4.6.5 SECONDARY CONTAINMENT ............ .......................... B 3/4 6-5 MILLSTONE - UNIT 2 XII Amendment No. 66, 69, 2, 104, 4-i, 4-;95, Z, 4, 2"
INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE .............................................. B 3/4 7-1 3/4.7.2 DELETED ............................................... B 3/4 7-3 a' 3/4.7.3 REACTOR BUILDING CLOSED COOLING WATER SYSTEM .......... B 3/4 7 3/4.7.4 SERVICE WATER SYSTEM ............... ............................... B 3/4 3/4.7.5 DELETED ............................................ B 3/4 7-4 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ............. B 3/4 7-4 3/4.7.7 DELETED ............................................ B 3/4 7-5 3/4.7.8 SNUBBERS ............................................ B 3/4 7-5 3/4.7.9 DELETED ............................................. B 3/4 7-6 3/4.7.10 DELETED ............................................. B 3/4 7-7 3/4.7.11 ULTIMATE HEAT SINK ............................................. B 3/4 7-7 3/4.8 ELECTRICAL POWER SYSTEMS .... B 3/4 8-1 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ............... .............................. B 3/4 9-1 3/4.9.2 INSTRUMENTATION ............................................. B 3/4 9-1 3/4.9.3 DECAY TIME ............................................. B 3/4 9 e 3/4.9.4 CONTAINMENT PENETRATIONS ................................ ............. B 3/4 9 1 3/4.9.5 DELETED ............................................. B 3/4 9& 3/4.9.6 DELETED ............................................. B 3/4 9-2 3/4.9.7 DELETED ............................................. B 3/4 9-2 3/4.9.8 SHUTDOWN COOLING AND COOLING RECIRCULATION ............B 3/4 9-2 MILLSTONE - UNIT 2 XIII Amendment No. -3, -W 69, 9,6, 404, 44-, 4-91,240, 2,
Septe adder
, 3 INDEX BASES SECTION PAGE 3/4.9.9 and 3/4.9.10 DELETED ..................................... B3/4 9-2 3/4.9.11 and 3/4.9.12 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL NVATER LEVEL........................................................ B 3/49-2/
3/4.9.13 DELETED .B 3/4 9-3 3/4.9.14 DELETED .B 3/4 9-3 3/4.9.15 STORAGE POOL AREA VENTILATION SYSTEM .B 3/4 9-3 3/4.9.16 SHIELDED CASK................................................................................... B 3/4 9-3b 3/4.9.17 SPENT FUEL POOL BORON CONCENTRATION .B 3/4 9-3b 3/4.9.18 SPENT FUEL POOL - STORAGE .B 3/4 9-4 1) 3/4.9.19 SPENT FUEL POOL - STORAGE PATTERN .B 3/4 9-4 3/4.9.20 SPENT FUEL POOL - CONSOLIDATION .B 3/4 9-4 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN .B 3/4 10-1 3/4.10.2 GROUP HEIGHT AND INSERTION LIMITS .B 3/4 10-1 3/4.10.3 DELETED .B 3/4 10-1 3/4.10.4 DELETED .B 3/4 10-1 3/4.10.5 DELETED .B 3/4 10-1 UJ 3/4.11 DELETED 3/4.11.1 DELETED ................. B3/4 11-1 3/4.11.2 DELETED ................. B 3/4 11-1 3/4.11.3 DELETED ................. B 3/4 11-1 MILLSTONE - UNIT 2 XIV Amendment No. 69, 4-4, 4-09, 417, a--,4-a,4a, a4-, 25G, 2-4,-9
INDEX ap4bef v42OQ2- . DESIGN FEATUTRES SECTION PAGE 5.1 SITE LOCATION . . . . . . . . . . . . . . . ... . . . . . . . . . 5-1 5.2 DELETED 5.3 REACTOR CORE Fuel Assemblies . . . . . . . . . . . . . . . ... . . . . . . . . 5-4 Control Element Assemblies ................... . 5-4 5.4 DELETED - 5.5 DELETED c 5.6 FUEL STORAGE Criticality . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 Drainage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 Capacity . .... s . 5.7 DELETED 5.8 DELETED 5.9 DELETED MILLSTONE - UNIT 2 XV Amendment No. Jp, , so 2
I~~l)EX) Juy5, qgo;- ADMSTRATiV CONTROLS SECTION PAGE 6.9 REPORTING BEOUIREMENTS 6.9.1 ROUTINE REPORTS .................................................................... 6-16 STARTUP REPORTS .................................................................... 6-16 ANNUAL REPORTS .................................................................... 6-17 ANNUAL RADIOLOGICAL REPORT .................................................................... 6-18 MONTHLY OPERATING REPORT ....................... ............................................. 6-18 CORE OPERATING LIMITS REPORT................................................................................... 6.9.2 SPECIAL REPORTS .................................................................... 6-19 6.10 DELETED 6.11 RADIATION PROTECTION PROGRAM . 6-20 6.12 HIGH RADIATION AREA .. 6-20 6.13 SYSTEMS INTEGRITY.................................................................................................................6-23 6.14 IODINE MONITORING ..................... 6-23 6.15 RADIOLOGICAL EFFLUENT MONITORING ANDOFFSITE DOSE CALCULATION MANUAL (REMODCM).. 6-24 6.16 RADIOACTIVE WASTE TREATMENT . 6-24 6.17 SECONDARY WATER CHEMISTRY . 6-25 6.18 DELETED 6.19 CONTAINMENT LEAKAGE RATE TESTING PROGRAM . 6-26 6.20 RADIACTIVE EFFULENT CONTROLS PROGRAM . 6-26 6.21 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM . 6-28 6.22 REACTOR COOLANT PUMP FLYWHEEL INSPECTION PROGRAM . 6-28 6.23 TECHNICAL SPECIFICATION (IaS) BASES CONTROL PROGRAM . 6-28 6.24 DIEPEL FUEL OIL TEST PROGRAM ................................ 6-29 Amendment No. 9, 6,63,66, 4i3., MILLSTONE - UNIT 2 XVII 4-04, 4-14, 44&, 4^3,4,6,4-69,239,
'^ _ -. _ -
.3/(l q41Q DEFINITIONS VENTING 1.35 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during fig. Vent, used in system names, does not imply a VENTING process.
MEMBER(S) OF THE PUBLIC 1.36 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant. The term "REAL MEMBER OF THE PUBLIC" means an individual who is exposed to existing dose pathways at one particular location. SITE BOUNDARY 1.37 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee. UNRESTRICTED AREA 1.38 An UNRESTRICTED AREA shall be any area at or beyond he - bud to which access is not controlled by the licensee for purposes of protection f individuals from exposure to radiation and radioactive materials or any area within the sit used for residential quarters or industrial, commercial institutional and/or recreational purposes. STORAGE PATTERN 1.39 The Region B spent fuel racks contain a cell blocking device in every 4th rack location for A administrative control. This 4th location will be referred to as the blocked location. A STORAGE PATTERN refers to a blocked location and all adjacent and diagonal cell locations surrounding the blocked location within the respective region. MILLSTONE - UNIT 2 1-8 Amendment No. 1-04, 44, 5 8,
Septembcr 25, 200e REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES CEA POSITION LIMITING CONDITION FOR OPERATION 3.1.3.1 All CEAs shall be OPERABLE with each CEA of a given group positioned within 10 steps (indicated position) of all other CEAs in its group, and the CEA Motion Inhibit and the CEA Deviation Circuit shall be OPERABLE. APPLICABILITY: MODES 1i0) and 2). ACTION: INOPERABLE EQUIPMENT REQUIRED ACTION A. One or more CEAs trippable A. I Reduce THERMAL POWER to < 70% of the and misaligned from all other maximum allowable THERMAL POWER within 1 hour CEAs in its group by > 10 steps and restore CEA(s) misalignment within 2 hours or and < 20 steps. othervise be in MODE 3 within the next 6 hours. OR One CEA trippable and misaligned from all other CEAs in its group by 2 20 steps. B. CEA Motion Inhibit inoperable. B.l Verify the indicated position of each CEA to be within 10 steps of all other CEAs in its group within I hour and every 4 hours thereafter, and restore CEA Motion Inhibit to OPERABLE status within 6 hours or otherwise be in MODE 3 within the next 6 hours. B.2(2) Place and maintain the CEA drive system mode switch in either the "off' or "manual" position, and withdraw all CEAs in group 7 to 2 172 steps within 6 hours or otherwise be in MODE 3 within the next 6 hours. (I) See Special Test l i 3.10.2 (2) Performance of B.2 is allowed only when not in conflict with either Required Action A.l orC.I. I MILLSTONE - UNIT 2 3/4 1-20 Amendment No. M
Pprl 21, 198, - REACTIVITY CONTROL SYSTEMS ICONTROL ROD DRIVE MECHANISMS LIM!- N'G CON14DITION FOR OPERATION I i.1.3.7 The control rod drive mechanisms shall be de-energized. APPLICABILITY: MODES 3*, 4, 5 and 6, whenever the RCS boron concentra-tion is less than refueling concentration of Specification
3.9.1. ACTION
With any of the control rod drive mechanisms energized, restore the mechanisms to their de-energized state within 2 hours or immediately open the reactor trip circuit breakers. SURVEILLANCE REQUIREMENTS 4.1.3.7 The control rod drive mechanisms shall be verified to be .de-energized at least once per 24 hours. /
- The control rod drive mechanisms may be energized for MODE 3 as long as 4 reactor coolant pumps are OPERATING, the reactor coolant system temperature is greater than 5000F, the pressurizer pressure is greater than 2000 psia and the high power trip is
. L e MILLSTONE - UNIT 2 3/4 1-31 Amendment No. ++'}
Sver2520'3 POWER DISTRIBUTION LIMITS TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR - FTr LIMITING CONDITION FOR OPERATION 3.2.3 The calculated value of FTr shall be within the 100% power limit specified in the CORE OPERATING LIMITS REPORT. The FTr value.shall include the effect of AZIMUTHAL POWER TILT. APPLICABILITY: MODE 1 with THERMAL POWER >20% RTP*. ACTION: With FTr exceeding the 100% power limit within 6 hours either:
- a. Reduce THERMAL POWER to bring the combination of THERMAL POWER and FTr to within the power dependent limit specified in the CORE OPERATING LIMITS REPORT and withdraw the CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6; or
- b. Be in at least HOT STANDBY.
SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 FTr shall be determined to be within the 100% power limit at the following intervals:
- a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading,
- b. At least once per 31 days of accumulated operation in A , and
- c. Within four hours if the AZIMUTHAL POWER TILT (Tq) is > 0.020.
4.2.3.3 FTr shall be determined by using the incore detectors to obtain a power distribution map with all CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump Combination.
- See Special Test Exception 3.10.2 MILLSTONE - UNIT 2 3/4 2-9 Amendment No. 38, 52, 49,90,99, 443,439,448, 4..,4.64,230, -...
scptz 1 ~L,~ 20j POWER DISTRIBUTION LIMITS AZIMUTHAL POWER TILT - TQ LIMITING CONDITION FOR OPERATION 3.2.4 The AZIMUTHAL POWER TILT (Tq) shall be < 0.02. APPLICABILITY: MODE I with THERMAL POWER > 50% of RATED THERMAL POWER0l). ACTION:
- a. With the indicated Tq > 0.02 but < 0.10, either restore Tq to < 0.02 within 2 hours or verify the TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR (FTr) is within the limit of Specification 3.2.3 within 2 hours and once per 8 hours thereafter. Or otherwise, reduce THERMAL POWER to < 50% of RATED THERMAL POWER within the next 4 hours.
- b. With the indicated Tq > 0.10, perform the following actions: (2)
- 1. Verify the TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR (FTr) is within the limit of Specification 3.2.3 within 2 hours; and
- 2. Reduce THERMAL POWER to *50% of RATED THERMAL POWER within 2 hours; and
- 3. Restore Tq < 0.02 prior to increasing THERMAL POWER. Correct the cause of the out of limit condition prior to increasing THERMAL POWER.
Subsequent power operation above 50% of RATED THERMAL POWER may proceed provided that the measured Tq is verified < 0.02 at least once per hour for 12 hours, or until verified at 95% of RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.2.4.1 Verify T is within limit at least once every 12 hours. The provisions of Specification 4.0.4 are not applicable for entering into MODE I with THERMAL POWER > 50% of RATED THERMAL POWER from MODE 1. (1) See Special Test Exception (2) All subsequent Required must be completed if power reduction commences prior to restoring Tq < 0.10. MILLSTONE - UNIT 2 3/4 2-10 Amendment No. 38, A2, 90, ,55, IM
June 10, 1996 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor protective instrumentation channels and bypasses of Table 3.3-1 shall be OPERABLE. APPLICABILITY: As shown in Table 3.3-1. ACTION: As shown in Table 3.3- 1. SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor protective instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the mody& and at the frequencies shown in Table 4.3-1. HODE-5 4.3.1.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation. 4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Neutron detectors are exempt (f from response time testing. Each test shall include at least one channel per function such that all UJ channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-I. MILLSTONE - UNIT 2 3/4 3-1 Amendment No. X, +9&
September 25, 20023 n TABLE 3.3-1 (Continued) TABLE NOTATION
- With the protective system trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.
(a) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 5% of RATED THERMAL POWER. (b) Trip may be manually bypassed when steam generator pressure is < 800 psia and all CEAs are fully inserted; bypass shall be automatically removed when steam generator pressure is 2 800 psia. (c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 15% of RATED THERMAL POWER. _F (d) Trip does not need to be opceif all the control rod drive mechanisms are de-energized or if the RCS boron concentration is greater than or equal to the refueling concentration of Specification 3.9. 1. (e) DELETED (f) AT Power input to trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 2 5% of RATED THERMAL POWER. ACTION STATEMENTS ACTION I - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 4 hours and/or open the protective system trip breakers. ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may continue provided the following conditions are satisfied:
- a. The inoperable channel is placed in either the bypassed or tripped condition within 1 hour. The inoperable channel shall either be restored to OPERABLE status, or placed in the tripped condition, within 48 hours.
- b. Within 1 hour, all functional units receiving an input from the inoperable channel are also declared inoperable, and the appropriate actions are taken for the affected functional units.
- c. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 48 hours, provided one of the inoperable channels is placed in the tripped condition.
MILLSTONE - UNIT 2 3/4 3-4 Amendment No. 9, 38,72, 44-6, 4-39, 225, 2Q a,- )a-
sept~he 25 20 TABLE 3.3-1 (Continued) ACTION STATEMENTS ACTION 3 - NOT USED ACTION 4 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, immediately verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 F 3.,-1 appleele, and at least once per 4 hours thereafter. ACTION 5 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours. ACTION 6 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours. MILLSTONE - UNIT 2 3/4 3-5 Amendment No. 225, ,-
Scptembcr 25, 2003 en INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The engineered safety feature actuation system instrumentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.34. APPLICABILITY: As shown in Table 3.3-3. ACTION:
- a. With an engineered safety feature actuation system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, either adjust the trip setpoint to be consistent with the value specified in the Trip Setpoint column of Table 3.3-4 within 2 hours or declare the channel inoperable and take the ACTION shown in Table 3.3-3.
- b. With an engineered safety feature actuation system instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.
SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each engineered safety feature acutation system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during thead4eand at the frequencies shown in Table 4.3-2. 0D3 4.3.2.1.2 The logic for the bypasses shall be demo ated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels aoc1 by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation. MILLSTONE - UNIT 2 3/4 3-9 Amendment No. 498, a2
ScpltzeMel.T 203 IN STRUM FNTATI ON ENGiNEERED SAFETY FEATURE ACTUATION SYSTEM SENSOR CABINET POWER SUPPLY DRAWERS LIMITING CONDITION FOR OPERATION 3.3.2.2 The engineered safety feature actuation system Sensor Cabinets (RC02A I, RC02B2, RC02C3 & RC02D4) Power Supply Drawers shall be OPERABLE and energized from the normal power source with the backup power source available. The normal and backup power sources for each sensor cabinet is detailed in Table 3.3-5a: CABINET NORMAL POWER BACKUP POWER RC02AI VA-10 VA-40 RC02B2 VA-20 VA-30 RC02C3 VA-30 VA-20 RC02D4 VA-40 VA-10 Table 3.3-5a APPLICABILITY: MODES 1, 2, 3 and 4 ACTION: With any of the Sensor Cabinet Power Supply Drawers inoperable, or either the normal or backup power source not available as delineated in Table 3.3-5a, restore the inoperable Sensor Cabinet Power Supply Drawer to OPERABLE status within 48 hours or be in COLD SHUTDOWN within the next 36 hours. SURVEILLANCE REQUIREMENTS 4.3.2.2.1 The engineered safety feature actuation system Sensor Cabinet Power Supply Drawers shall be determined OPERABLE once per shift by visual inspection of the power supply drawer indicating lamps. p II 4.3.2.2.2 Verify the of the Sensor Cabinet Power Supply auctioneering circuit at least one per 18 months. MILLSTONE - UNIT 2 3/4 3-23 Amendment No. 479,H2
Septernber 90 9004 t INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm/trip setpoints within the specified limits. APPLICABILITY: As shown in Table 3.3-6. ACTION:
- a. With a radiation monitoring channel alarmltrip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 2 hours or declare the channel inoperable.
- b. With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.1.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the io4s and at the frequencies shown in Table 4.3-3. j3 4.3.3.1.2 DELETED 4.3.3.1.3 Verify the response time of the control room isolation channel at least once per 18 months. MILLSTONE - UNIT 2 3/4 3-24 Amendment No. ,24-5, , 4
V v4-- L- W.25, 2003N Septcmbcr AIDMEN&IT NG ACCIDENT MONITORING LIMITING CONDITION FOR OPERATION 3.3.3.8 The accident monitoring instrumentation channels shown in Table 3.3-11 shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. A per Table 3.3-11 .
SURVEILLANCE REQUIREMENTS 4.3.3.8 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7. MILLSTONE - UNIT 2 3/4 3-31 Amendment No. 66, 4-5A, 82
TABLE 3.3-11 CA ACCIDENT MONITORING INSTRUMENTATION r Minimum Total No. Channels t.J Instrument of Channels . 0
- 1. Pressurizer Water Level 2 I I1 z 2. Auxiliary Feedwater Flow Rate
- 2/S.Q
- 3. RCS Subcooled/Superheat Monitor 2 2 H I/valve 3
- 4. PORV Position Indicator Acoustic Monitor
- 5. PORV Block Valve Position I/valve 3 Indicator z0 M
- 6. Safety Valve Position Indicator I/valve 3 Acoustic Monitor ID 7. Containment Pressure (Wide Range) 2 I 4 t3
- 8. Containment Water Level (Narrow Range) 7t##
I tj 9. Containment Water Level (Wide Range) 4
- 10. Core Exit Thermocouples quadrant 2 CETs in any 5 n
of 2 core quadrants ID
- 11. Main Steam Line Radiation Monitor 3 3 6
- 12. Reactor Vessel Coolant Level l 2* 1* 8
- A channel is eight (8) sensors in a probe.1llA channel is 4e if four (4) or more sensors, two (2) or more in the upper four and two (2) or more in the lower four, arepepbe.
#f# Refer to ACTION statement in Technical Specification 3.4.6. 1.
Szpct.br 2, 00 TABLE 3.3-11 (Continued) ACTION STATEMENTS ACTION I - With the number of OPERABLE channels less than the MINIMUM CHANNELS OPERABLE requirements of Table 3.3-1 1, either restore the inoperable channel(s) to OPERABLE status within 30 days or be in HOT STANDBY within the next 12 hours. ACTION 2 - With the number of channels OPERABLE less than the MINIMUM CHANNELS OPERABLE, determine the subcooling margin once per 12 hours. ACTION 3 - With any individual valve position indicator inoperable, obtain quench tan temperature, level and pressure information, and monitor discCTo temperature once per shift to determine valve position. This is not require if the PORV block valve is closed with power removed in accordance with Specification 3.4.3.a or 3.4.3.b. ACTION 4 - a. With the number of OPERABLE accident monitoring instrumentation channels less than the total number of channels shown in Table 3.3-1 1, restore the inoperable channel(s) to OPERABLE status within 7 days, or submit a special report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction, the plans for restoring the channel(s) to OPERABLE status, and any alternate methods in affect for estimating the applicable parameter during the interim.
- b. With the number of OPERABLE accident monitoring instrumentation channels less than the MINIMUM CHANNELS OPERABLE requirements of Table 3.3-1 1, restore the inoperable channel(s) to OPERABLE status within 48 hours, or submit a special report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction, the plans for restoring the channel(s) to OPERABLE status, and any alternate methods in affect for estimating the applicable parameter during the interim.
MILLSTONE - UNIT 2 3/4 3-33 Amendment No. 420, 2.82-
Scptember 14, 000 REACTOR COOLANT SYSTEM COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.4.1 Two reactor coolant loops shall be OPERABLE and in operation. APPLICABILITY: MODES and I m
&l-ACTION:
With the requirements of the above specification not met, be in at least HOT STANDBY within 6 hours. SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation at least - once per 12 hours. (9 3S Spet To Exemption 1 .10.4. - MILLSTONE - UNIT 2 3/4 4-1 Amendment No. 0, 69, 20, 2449)
Apl-M9-REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.1.5 Reports
- a. Following each inservice inspection of steam generator tubes, the number of tubes plugged and sleeved in each steam generator shall be reported to the Commission within 15 days.
- b. The complete results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the period in which this inspection was completed. This report shall include:
- 1. Number and extent of tubes inspected.
- 2. Location and percent of wall-thickness penetration for each indication of an imperfection.
- 3. Identification of tubes plugged or sleeved.
- c. Results of steam generator tube inspections which fall into Category C-3 shall be reported pursuant to.J 0 CFR 50.72. In lieu of any report Iequired pouuat to-Speei4iGain-6.6, 4'4 Special Report pursuant to Specification 6.9.2 shall be submitted prior to resumption of plant operation and shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.
MILLSTONE - UNIT 2 3/4 4-7b .Amendment No. 22, 37, R, 9,414
Aps9r198~- TABLE 4.4-6 STEAM GENERATOR T UBE INSPECTTON =TT= r" 1ST SAMPLE INSPECTION _ SAMPLE INSPECT -3RD SANR SAYiPE¢TION Sample Size Result Akrequired Result Aequired Res quired l A minimum of C-1 None N/AN/A N/4 S tubes per C-2 Repair defective C-l None / NA_ S. G tubes and inspect C-2 Repair defective tubes aI No additional 25 tubes in inspect additional 45 esCRepaird ciX Z\) this S.G.* i hsSG - eardfc~v tubes* C-3 Perform for C-3 result of first sample C-3 Perform4 for q-3 N/A N/A result of first sampfe C-3 Inspection all tubes All other Non N/A N/A 3 in this S.G, repair S.Gs are C- I defective tubes and Some S.Gs Perform ae6 for C-2 N/A N/A inspect 25 tubes in C-2 but no result of second sample each other SG* additional Prompt notification S.G are C-3 t4 _4 to NRC pursuant to Additional Inspect all tubes in each N/A N/A 3 10 CFR 50.72 S.G is C-3 S.G and repair defective _ _) 0 tubes.* Prompt notification to NRC pursuant to 10 CFR 50.72
-El av_
S = 3N% Where N is the number of steam generators in the unit, and n is the number of steam generators inspected during an inspection n
- Repair of defective tubes shall be limited to plugging with the exception of those. tubes which may be sleeved. Tubes with defective sleeves shall be plugged.
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,,,a .,a . ,, ., , *.+/-. , a, .. . .L, Jl ... a a 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 1.0 p Ci/gram Dese Ettih'e t4 1-131 I MILLSTONE - UNIT 2 3/4 4-16
3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained. APPLICABILITY: MODES 1, 2,3 and 4. ACTION: Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
- a. At least once per 31 days by verifying that all penetrations(') not capable of being closed by OPERABLE containment automatic isolation valves(2) and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, (3)except for valves that are open under administrative control as permitted by Specification 3.6.3.1.
- b. At least once per 31 days by verifying the equipment hatch is closed and sealed.
- c. By verifying the containment air lock is in compliance with the requirements of Specification 3.6.1.3.
- d. After each closing of a penetration subject to type B testing (except the containment air lock), if opened following a Type A or B test, by leak rate testing in accordance.yith the Containment Leakage Rate Testing Program.
- e. By verifyingbtructural integrity in accordance with the Containment Tendon Surveillance Program.
(1) Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed, or otherwise secured in the closed position. These penetrations shall be verified closed prior to entering MODE 4 from MODE 5, if not performed within the previous 92 days. (2) In MODE 4, the requirement for an OPERABLE containment automatic isolation valve system is satisfied by use of the containment isolation trip pushbuttons (3) Isolation devices in high radiation areas may be verified by use of administrative means. MILLSTONE - UNIT 2 3/4 6-1 Amendment No. , 95, 03, 2410, 21-5, A)l
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CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY AND COOLING SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.2.1 Two containment spray trains and two containment cooling trains, with each cooling train consisting of two containment air recirculation and cooling units, shall be OPERABLE. APPLICABILITY: MODES 1, 2 and 3*. ACTION: (t Inoperable Equipment
<EER Required A4ien A w
- a. Onecontainment a.1 Restore the inoperable containment spray train to spray train OPERABLE status within 72 hours or be in HOT STANDBY within the next 6 hours and reduce pressurizer pressure to less than 1750 psia within the following 6 hours.
- b. One containment b.1 Restore the inoperable containment cooling train to cooling train OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours.
- c. 1One containment c. 1 Restore the inoperable containment spray train or the Ispray train inoperable containment cooling train to OPERABLE status AND . within 48 hours or be in HOT SHUTDOWN within the next One containment 12 hours.
cooling train
- d. Two d.I Restore at least one inoperable containment cooling train to containment OPERABLE status within 48 hours or be in HOT cooling trains SHUTDOWN within the next 12 hours.
- e. All other e. I Enter LCO 3.0.3 immediately.
combinations SURVEILLANCE REQUIREMENTS 4.6.2.1.1 Each containment spray train shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying each containment spray manual, power operated, and automatic valve in the spray train flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
- The Containment Spray System is not required to be OPERABLE in MODE 3 if pressurizer pressure is < 1750 psia.
MILLSTONE - UNIT 2 3/4 6-12 Amendment No. 145, 22, 26, 28
'LPN :-rt 4- PLRNTT TE SYsS
=r _
LIMITING CONDITION FOR OPERATION (Continued)
- a. With two or more of the feedwater isolation components inoperable in the same flow path, either -
- 1. Restore the inoperable component(s) to OPERABLE status within 8 hours until e ' applies, or
- 2. Isolate the affected flow path within 8 hours, and verify that the inoperable feedwater isolation components are closed or isolated/secured once per 7 days, or
- 3. Be in HOT SHUTDOWN within the next 12 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.6 Each feedwater isolation valve/feedwater pump trip circuitry shall be demonstrated OPERABLE at least once per 18 months by:
- a. Verifying that on 'A' main steam isolation test signal, each isolation valve actuates to its isolation position, and
- b. Verifying that on 'B' main steam isolation test signal, each isolation valve actuates to its isolation position, and
- c. Verifying that on 'A' main steam isolation test signal, each feedwater pump trip circuit actuates, and
- d. Verifying that on 'B' main steam isolation test signal, each feedwater pump trip circuit actuates.
MILLSTONE - UNIT 2 3/4 7-9b Amendment No. 4-I
Jmmry-4-2O2~ 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. Two physically independent circuits between the offsite transmission network and the onsite Class IE distribution system, and
- b. Two separate and independent diesel generators each with a separate fuel oil supply tank containing a minimum of 12,000 gallons of fuel.
APPLICABILITY: MODES 1, 2,3 and 4. ACTION: Inoperable Equipment Required Aeti
- a. One offsite a.l Perform Surveillance Requirement 4.8.1.1.1 for circuit remaining offsite circuit within I hour prior to or after entering this condition, and at least once per 8 hours thereafter.
AND a.2 Restore the inoperable offsite circuit to OPERABLE status within 72 hours or be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. MILLSTONE - UNIT 2 3/4 8-1 Amendment No. 4-, 477, 492, 24-, 2 I5, 6-'
jafty ~OO~ ELECTRICAL POWER SYSTEMS ACTION (Continued) Inoperable Equipment Required Atenie
- b. One diesel b. I Perform Surveillance Requirement .8.1.1 forthe generator offsite circuit within 1 hour prior t a tor er entering this condition, and at least once per 8 ours ereafter.
AND b.2 Demonstrate OPERABL diesel genera or is not inoperable due to corn on cause failure ithin 24 hours or perform Surveilla e Requirement 4. .1.1.2.a.2 for the OPERABLE di el generator withi 24 hours. AND b.3 Verify the stea -driven auxiliary feed ater pump OPERABLE ( ODES 1,2, and 3 onl. If this.(@9 condition is no satisfied within 2 hour , be in X least HOT STAND within the next 6 ho s and HOT SHUTDOWN ithin the following 6 urs. AND b.4 (Applicable on if the 14 day allowe outage time specified in Statement b.5 is to be used.) Verify the required Millstone Unit No. 3 dies 1generator(s) is/ are OPERABLE and the Millstone Un t No. 3 SBO diesel generator is available within I our prior to or after entering this condition, and at le t once per 24 hours thereafter. Restore any inoperab required Millstone Unit No. 3 diesel generator OPERABLE status and/or Millstone Unit No. 3 SB diesel generator to available status within 72 hours or b in HOT STANDBY within the next 6 hours and OLD SHUTDOWN within the following 30 h urs. AND b.5 Restore the inoperable diesel generator to PERABLE MILLSTONE - UNIT 2 status within 72 hours (within 14 days if Statement b.4 is met) or be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. 3/4 8-la Amendment No.4,4-57,7, 92,234, 2-5+,Q-
'I
ELECTRICAL POWER SYSTEMS _1_11
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AA He _ a - I ACTION (Continued) Inoperable Equipment Required l IC. One offsite C.I Perform Surveillance Requirement circuit remaining offsite circuit within I per 8 hours thereafter. AND AND One diesel generator c.2 Demonstrate OPERABLI iesel genera r is not inoperable due to comrnn cause failure ithin 8 hours or perform Surveillan Requirement 4. .1.1.2.a.2 for the OPERABLE di el generator within 8 hours. AND c.3 Verify the ste -driven auxiliary feedw er pump is OPERABL (MODES 1, 2, and 3 only). If this condition i .not satisfied within 2 hours, e in at least HOT STA BY within the next 6 hour and HOT SHUTD within the following 6 ho rs. AND c.4 Resto one inoperable A.C. source to ERABLE I statu within 12 hours or be in HOT STANDBY within the xt 6 hours and COLD SHUTDOkN within the foll ing 30 hours. AND C.5 Res ore remaining inoperable A.C. sour e to I OP RABLE status following the time r quirements of Statements a or b above based o the initial loss of the remaining inoperable A.C. source
- d. Two offsite d.1 Restore one of the inoperable offsite sour es to circuits OPERABLE status within 24 hours or be n HOT STANDBY within the next 6 hours.
AND d.2 Following restoration of one offsite source r store remaining inoperable offsite source to OPE LE status following the time requirements o Statement a above based on the initial loss of the remaining inoperable offsite source. MILLSTONE - UNIT 2 3/4 8-2 Amendment No. 1-67, -7, A34, my64-
ELECTRICAL POWER SYSTEMS ACTION (Continued) 'iCTON Inoperable Equipment Required Adz f
- e. Two diesel e.I Perform Surveillance Requirement 4.8.1.1 1 for the generators offsite circuits within I hour and at least o ce per 8 hours thereafter.
AND e.2 Restore one of the inoperable diesel generator to OPERABLE status within 2 hours or be in HO STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. AND e.3 Following restoration of one diesel generator res re remaining inoperable diesel generator to OPERABLE status following the time requirements of AeeR Statement b above based on the initial loss of the remaining inoperable diesel generator. SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Verify correct breaker alignment and indicated power available for each required offsite circuit at least once per 24 hours. MILLSTONE - UNIT 2 314 8-2a Amendment No. 431-, 231, I)
September 17, 2002 5.0 DESIGN FEATURES 1- ZVO ND qty 5.1 SITE LOCATION _ I The Unit 2 Containment Building iscated on the site at Millstone Point in Waterford, Connecticut. The nearest ske-bet~ndmy on land is 2034 feet northeast of the containment building wall (1627 feet northeast of the elevated stack), which is the minimum distance to the boundary of the exclusion area as described in 10 CFR 100.3. No part of the site that is closer than these distances shall be sold or leased except to Dominion Nuclear Connecticut, Inc. or its corporate affiliates for use in conjunction with normal utility operations. 5.2 DELETED MILLSTONE - UNIT 2 5-1 Amendment No. 26, 270
Septembcr 17, 2002 TABLE 6.2-1(3) MINIMUM SHIFT-CREW COMPOSITION(2) APPLICABLE MODES LICENSE CATEGORY 1, 2,3 & 4 5&6 Senior Reactor Operator 2 IM Reactor Operator 2 l Non-Licensed Operator 2 1 Shift Technical Advisor 1(4 ) None Required (1) Does not include the licens Senior eactor or Senior Reactor Operator Limited to Fuel Handling individual CORE ALTERATIONS after the initial fuel loading. (2) The above shift crew composition and the radiation protection technician of Section 6.2.2 may be less than the minimum requirements for a period of time not to exceed 2 hours in 6J 9 order to accommodate unexpected absence provided expeditious actions are taken to fill the required position. (3) Requirements for minimum number of licensed operators on shift during operation in other than eeld-44down or are .1i OCFR50.54(m). (4) The Shift Technical Advi or position can be filled by either of the two Senior Reactor Operators (a dual-role i ividual), if he meets the requirements of Specification 6.3.1 .b. 1. MILLSTONE - UNIT 2 6-4 Amendment No. 66, , 92, 136, 63, 49,276
SeptcniL 2,, 2003 e ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (CONT.)
- 8. XN-NF-621 (P)(A), "Exxon Nuclear DNB Correlation for PWR Fuel Designs," Exxon Nuclear Company.
- 9. XN-NF-82-06(P)(A), and Supplements 2,4 and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup," Exxon Nuclear Company.
- 10. ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWd/MTU," Advanced Nuclear Fuels Corporation.
- 11. XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," Exxon Nuclear Company.
- 12. ANF-89-151(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation.
- 13. EMF-1961 (P)(A), "Statistical Setpoint/Transient Methodology for Combustion Engineering Type Reactors," Siemens Power Corporation.
- 14. EMF-2130(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," Framatome ANP.
- 15. EMF-92-153(P)(A) and Supplement 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation.
- c. The core operating limits shall be determined so that all applicable limits (e.g.,
fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as sh y , and transient and accident analysis limits) of the safety analysis are met.
- d. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator, Region I, and one copy to the NRC Resident Inspector within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:
- a. Deleted MILLSTONE - UNIT 2 6-19 Amendment No. 148,4-3, 28,250 26O, R
May4a ,2003L-e ADMINISTRATIVE CONTROLS 6.19 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of the primary containment as required by I OCFR50.54(o) and I OCFR5O, Appendix J, Option B as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995. The peak calculated primary Containment internal pressure for the design basis loss of coolant accident is Pa. The maximum allowable primary containment leakage rate, La, at Pa, is 0.5% of primary containment air weight per day. Leakage rate acceptance criteria are:
- a. Primary containment overall leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and < 0.75 La for Type A tests;
- b. Air lock testing acceptance criteria are:
- 1. Overall air lock leakage rate is
- 0.05 La when tested at > Pa.
- 2. For each door, pressure decay is *0.1 psig when pressurized to 2 25 psi for at least 15 minutes.
The provisions of SR 4.0.2 do not apply for test frequencies specified in the Primary Containment Leakage Rate Testing Program. The provisions of SR 4.0.3 are applicable to the Primary Containment Leakage Rate Testing Program. 6.20 RADIOACTIVE EFFLUENT CONTROLS PROGRAM This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the REMODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
- a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance CM;
- b. Limitations on the concentratof radioactive material released in liquid effluents to e 4 a onforming to ten times the concentration values in (j Appendix B, Table 2, Column 2 to 10CFR20.1001-20.2402; Ill MILLSTONE - UNIT 2 6-26 Amendment No. 20,250, 2%7
MaI,15W0 ^ ADMINISTRATIVE CONTROLS 6.20 RADIOACTIVE EFFLUENT CONTROLS PROGRAM
- c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the REMODCM;
- d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to conforming to 10 CFR 50, Appendix I;
-ion of cumuative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the REMODCM at least every 31 days.
Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the REMODCM at least every 31 days; ,J
- f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;
- g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the .i budry shall be in accordance with the following:
- 1. For noble gases: a dose rate
- 500 mrem/yr to the whole ody and a dose rate < 3000 mrem/yr to the skin, and
- 2. For iodine-131, iodine 133, tritium, and all radionucli sin particulate form with half-lives greater than 8 days: a dose rate 1500 mrem/yr to any organ; iggUr L o stP fie-
- h. Limitations on the annu quarterly aid ses resulting from noble ses released in gaseous uents from each unit o areas beyond the s conforming toI FR 50, Appendix I;
- i. Limitation n the annual and quart y doses to a member of the public from iodine-1i, iodine-133, tritium, all radionuclides in particulate form with half lives >tdays in gaseous effl ts released from each unit to areas beyond the-4te betd~iy, conforming to1CER 50, Appendix I; and
- j. Limitations on the an al dose or dose commitment to any member of the public, beyond the site bou;;y due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.
The provisions of Specification 4.0.2 and Specification 4.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency. MILLSTONE - UNIT 2 6-27 Amendment No. 250,4 ;&
)
Serial No. 05-009 Docket No. 50-336 ATTACHMENT 3 LICENSE AMENDMENT REQUEST (LBDCR 04-MP2-015) ADMINISTRATIVE CHANGES TO TECHNICAL SPECIFICATIONS RE-TYPED PAGES MILLSTONE POWER STATION UNIT 2 DOMINION NUCLEAR CONNECTICUT, INC.
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY ............... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 REACTIVITY CONTROL .3/4 1-1 Shutdown Margin (SDM) . 3/4 1-1 DELETED .3/4 1-2 Reactivity Balance . 3/4 1-3 Boron Dilution . 3/4 1-4 Moderator Temperature Coefficient (MTC) . 3/41-5 Minimum Temperature for Criticality . 3/4 1-7 3/4.1.2 BORATION SYSTEMS . 3/4 1-8 DELETED .3/4 1-8 DELETED .3/4 1-9 DELETED .3/4 1-11 DELETED .3/4 1-13 DELETED .3/4 1-14 DELETED .3/4 1-15 DELETED .3/4 1-16 DELETED .34 1-18 3/4.1.3 MOVABLE CONTROL ASSEMBLIES .3/4 1-20 CEA Group Position .3/4 1-20 Position Indicator Channels .3/4 1-24 CEA Drop Time .3/4 1-26 Shutdown CEA Insertion Limit .3/4 1-27 Regulating CEA Insertion Limits .3/4 1-28 Control Rod Drive Mechanisms .3/4 1-31 MILLSTONE - UNIT 2 IV Amendment No. 3, 404, 416, 4-53, 4-85, M, 8,
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE ............................................... 3/4 2-1 3/4.2.2 Deleted 3/4.2.3 TOTAL INTEGRATED RADIAL PEAKING FACTOR - . ................. 3/4 2-9 3/4.2.4 AZIMUTHAL POWER TILT ............ ............................... 3/4 2-10 3/4.2.5 Deleted 3/4.2.6 DNB MARGIN .................................... 3/4 2-13 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION .................................... 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION ............................................... 3/4 3-9 3/4.3.3 MONITORING INSTRUMENTATION .............................................. 3/4 3-24 Radiation Monitoring .............................................. 3/4 3-24 Remote Shutdown Instrumentation .............................................. 3/4 3-28 Accident Monitoring ............................................... 3/4 3-31 3/4.3.4 CONTAINMENT PURGE VALVE ISOLATION SIGNAL ...................... 3/4 3-36 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION ............................. 3/4 4-1 Startup and Power Operation .............................................. 3/4 4-1 HOT STANDBY ............................................... 3/4 4-la HOT SHUTDOWN .............................................. 3/4 4-lb COLD SHUTDOWN - Reactor Coolant System Loops Filled ................... 3/4 4-1 d COLD SHUTDOWN - Reactor Coolant System Loops Not Filled ............. 3/4 4-If Reactor Coolant Pumps - COLD SHUTDOWN .......................................... 3/4 4-1h MILLSTONE - UNIT 2 V Amendment No. -3, 3, 66,69,99, 40449,+83,, 49+,2-3,2445,249,
-2$50,282,
INDEX BASES SECTION PAGE 3/4.0 APPLICABILITY ...................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL ...................................... B 3/4 1-1 3/4.1.2 Deleted 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................................... B 3/4 1-2 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE .B 3/4 2-1 3/4.2.2 Deleted 3/4.2.3 TOTAL INTEGRATED RADIAL PEAKING FACTOR - FT. B 3/4 2-1 3/4.2.4 AZIMUTHAL POWER TILT .B 3/4 2-1 3/4.2.5 Deleted 3/4.2.6 DNB MARGIN .B 3/4 2-2 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION ................................................... B 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION ................B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION ................................................. B 3/4 3-2a 3/4.3.4 CONTAINMENT PURGE VALVE ISOLATION SIGNAL .................... B 3/4 3-6 MILLSTONE - UNIT 2 xi Amendment No. -3, 49, 404, -9, 185, 245,28;,
INDEX BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION ......................... B 3/4 4-1 3/4.4.2 SAFETY VALVES ....................................... B 3/4 4-Id 3/4.4.3 RELIEF VALVES ........................................ B 3/4 4-2 3/4.4.4 PRESSURIZER ........................................ B 3/4 4-2a 3/4.4.5 STEAM GENERATORS......................................................................... B 3/4 4-2a 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE ....................................... B 3/4 4-3 3/4.4.7 DELETED ....................................... B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY............................................................................... B 3/4 4-4 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ......................... .............. B 3/4 4-5 3/4.4.10 DELETED ........................................ B 3/4 4-7 3/4.4.11 DELETED ....................................... B 3/4 4-8 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS ..... B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS........................................................................ B 3/4 5-2 3/4.5.4 REFUELING WATER STORAGE TANK (RWST) ..... B 3/4 5-2d 3/4.5.5 TRISODIUM PHOSPHATE (TSP)........................................................... B 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT ...................................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS ............................. B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES ..................................... B 3/4 6-3a 3/4.6.4 COMBUSTIBLE GAS CONTROL........................................................... B 3/4 6-4 3/4.6.5 SECONDARY CONTAINMENT............................................................. B 3/4 6-5 MILLSTONE - UNIT 2 XII Amendment No. 66, 69, :, 404, 453, 48, 24-, 264,266,
INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ............................................... B 3/4 7-1 3/4.7.2 DELETED ............................................... B 3/4 7-3 3/4.7.3 REACTOR BUILDING CLOSED COOLING WATER SYSTEM ....... B 3/4 7-3a 3/4.7.4 SERVICE WATER SYSTEM ............................................... B 3/4 7-3c 3/4.7.5 DELETED ............................................... B 3/4 7-4 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ............B 3/4 7-4 3/4.7.7 DELETED ............................................... B 3/4 7-5 3/4.7.8 SNUBBERS ............................................... B 3/4 7-5 3/4.7.9 DELETED ............................................... B 3/4 7-6 3/4.7.10 DELETED ............................................... B 3/4 7-7 3/4.7.11 ULTIMATE HEAT SINK .............. ................................. B 3/4 7-7 3/4.8 ELECTRICAL POWER SYSTEMS .... B 3/4 8-1 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION .............................................. B 3/4 9-1 3/4.9.2 INSTRUMENTATION .............................................. B 3/4 9-1 3/4.9.3 DECAY TIME ............................................. B 3/4 9-la 3/4.9.4 CONTAINMENT PENETRATIONS ............................................. B 3/4 9-la 3/4.9.5 DELETED ............................................. B 3/4 9-Ia 3/4.9.6 DELETED .............................................. B 3/4 9-2 3/4.9.7 DELETED .............................................. B 3/4 9-2 3/4.9.8 SHUTDOWN COOLING AND COOLING RECIRCULATION ............ B 3/4 9-2 MILLSTONE - UNIT 2 XIII Amendment No. , W, 69,96, 4-04, 44;, 449,240,242,
INDEX BASES SECTION PAGE 3/4.9.9 and 3/4.9.10 DELETED ................................................ B 3/4 9-2 3/4.9.11 and 3/4.9.12 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL WATER LEVEL....................................................... B 3/4 9-2b 3/4.9.13 DELETED .B 3/4 9-3 3/4.9.14 DELETED .B 3/4 9-3 3/4.9.15 STORAGE POOL AREA VENTILATION SYSTEM .B 3/4 9-3 3/4.9.16 SHIELDED CASK................................................................................... B 3/4 9-3b 3/4.9.17 SPENT FUEL POOL BORON CONCENTRATION .B 3/4 9-3b 3/4.9.18 SPENT FUEL POOL - STORAGE .B 3/4 9-4 3/4.9.19 SPENT FUEL POOL - STORAGE PATTERN .B 3/4 9-4 3/4.9.20 SPENT FUEL POOL - CONSOLIDATION .B 3/4 9-4 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN ................ B 3/4 10-1 3/4.10.2 GROUP HEIGHT AND INSERTION LIMITS .B 3/4 10-1 3/4.10.3 DELETED .B 3/4 10-1 3/4.10.4 DELETED .B 3/4 10-1 3/4.10.5 DELETED. B 3/4 10-1 3/4.11 DELETED 3/4.11.1 DELETED ................. B 3/4 11-1 3/4.11.2 DELETED ................. B 3/4 11-1 3/4.11.3 DELETED .................. B 3/4 I -I MILLSTONE - UNIT 2 XIV Amendment No. 69,4-04, 4-09, 4-17, 154, 45,48-, 4-, &, 74,N
INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE LOCATION ........................ 5-1 5.2 DELETED 5.3 REACTOR CORE Fuel Assemblies ........................ 5-4 Control Element Assemblies ........................ 5-4 5.4 DELETED 5.5 DELETED 5.6 FUEL STORAGE Criticality ........................ 5-5 Drainage ........................ 5-5a Capacity.......................................................................................................................... 5-5a 5.7 DELETED 5.8 DELETED 5.9 DELETED MILLSTONE - UNIT 2 XV Amendment No. 404, 409, 20,
INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS ........................................................ 6-16 STARTUP REPORTS ........................................................ 6-16 ANNUAL REPORTS ........................................................ 6-17 ANNUAL RADIOLOGICAL REPORT ........................................................ 6-18 MONTHLY OPERATING REPORT ........................................................ 6-18 CORE OPERATING LIMITS REPORT ........................................................ 6-18a 6.9.2 SPECIAL REPORTS ........................................................ 6-19 6.10 DELETED 6.11 RADIATION PROTECTION PROGRAM ........................................................ 6-20 6.12 HIGH RADIATION AREA ........................................................ 6-20 6.13 SYSTEMS INTEGRITY ........................................................ 6-23 6.14 IODINE MONITORING ........................................................ 6-23 6.15 RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REMODCM) ........................................................ 6-24 6.16 RADIOACTIVE WASTE TREATMENT ........................................................ 6-24 6.17 SECONDARY WATER CHEMISTRY ........................................................ 6-25 6.18 DELETED 6.19 CONTAINMENT LEAKAGE RATE TESTING PROGRAM ....................................... 6-26 6.20 RADIOACTIVE EFFULENT CONTROLS PROGRAM ............................................... 6-26 6.21 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ........................ 6-28 6.22 REACTOR COOLANT PUMP FLYWHEEL INSPECTION PROGRAM ................... 6-28 6.23 TECHNICAL SPECIFICATION (TS) BASES CONTROL PROGRAM ...................... 6-28 6.24 DIESEL FUEL OIL TEST PROGRAM ....................... ................................. 6-29 MILLSTONE - UNIT 2 XVII Amendment No. 49, 6,63,66,4,03, 404,444,448,443,4-63,469,239,2-50, 26G,264, 20, 276,27,
DEFINITIONS VENTiNG 1.35 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process. MEMBER(S' OF THE PUBLIC 1.36 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant. The term "REAL MEMBER OF THE PUBLIC" means an individual who is exposed to existing dose pathways at one particular location. SITE BOUNDARY 1.37 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee. UNRESTRICTED AREA 1.38 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the SITE BOUNDARY used for residential quarters or industrial, commercial institutional and/or recreational purposes. STORAGE PATTERN 1.39 The Region B spent fuel racks contain a cell blocking device in every 4th rack location for administrative control. This 4th location will be referred to as the blocked location. A STORAGE PATTERN refers to a blocked location and all adjacent and diagonal cell locations surrounding the blocked location within the respective region. MILLSTONE - UNIT 2 1-8 Amendment No. 404, 1447, 418, 4-2,
REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES CEA POSITION LIMITING CONDITION FOR OPERATION 3.1.3.1 All CEAs shall be OPERABLE with each CEA of a given group positioned within 10 steps (indicated position) of all other CEAs in its group, and the CEA Motion Inhibit and the CEA Deviation Circuit shall be OPERABLE. APPLICABILITY: MODES j(1)and2(1). ACTION: INOPERABLE EQUIPMENT REQUIRED ACTION A. One or more CEAs trippable A.1 Reduce THERMAL POWER to < 70% of the and misaligned from all other maximum allowable THERMAL POWER within I hour CEAs in its group by > 10 steps and restore CEA(s) misalignment within 2 hours or and < 20 steps. otherwise be in MODE 3 within the next 6 hours. OR One CEA trippable and misaligned from all other CEAs in its group by 2 20 steps. B. CEA Motion Inhibit inoperable. B.1 Verify the indicated position of each CEA to be within 10 steps of all other CEAs in its group within 1 hour and every 4 hours thereafter, and restore CEA Motion Inhibit to OPERABLE status within 6 hours or otherwise be in MODE 3 within the next 6 hours.
.OR B.2(2) Place and maintain the CEA drive system mode switch in either the "off' or "manual" position, and withdraw all CEAs in group 7 to 2 172 steps within 6 hours or otherwise be in MODE 3 within the next 6 hours.
(1) See Special Test Exception 3.10.2 (2) Performance of ACTION B.2 is allowed only when not in conflict with either Required Action I A.l orC.I. MILLSTONE - UNIT 2 3/4 1-20 Amendment No. 32, 284,
REACTIVITY CONTROL SYSTEMS CONTROL ROD DRIVE MECHANISMS LIMITING CONDITION FOR OPERATION 3.1.3.7 The control rod drive mechanisms shall be de-energized. APPLICABILITY: MODES 3*, 4, 5 and 6, whenever the RCS boron concentration is less than refueling concentration of Specification 3.9.1. ACTION: With any of the control rod drive mechanisms energized, restore the mechanisms to their de-energized state within 2 hours or immediately open the reactor trip circuit breakers. SURVEILLANCE REQUIREMENTS 4.1.3.7 The control rod drive mechanisms shall be verified to be de-energized at least once per 24 hours.
- The control rod drive mechanisms may be energized for MODE 3 as long as 4 reactor coolant pumps are OPERATING, the reactor coolant system temperature is greater than 5000 F, the pressurizer pressure is greater than 2000 psia and the high power trip is OPERABLE. I MILLSTONE - UNIT 2 314 1-31 Amendment No. 44,
POWER DISTRIBUTION LIMITS TOTAL UNRODDED INTEGRATF1D RADIAL PEAKING FACTOR - FTr LIMITING CONDITION FOR OPERATION 3.2.3 The calculated value of FTr shall be within the 100% power limit specified in the CORE OPERATING LIMITS REPORT. The FTr value shall include the effect of AZIMUTHAL POWER TILT. APPLICABILITY: MODE I with THERMAL POWER >20% RTP*. ACTION: With FTr exceeding the 100% power limit within 6 hours either:
- a. Reduce THERMAL POWER to bring the combination of THERMAL POWER and FTr to within the power dependent limit specified in the CORE OPERATING LIMITS REPORT and withdraw the CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6; or
- b. Be in at least HOT STANDBY.
SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 FTr shall be determined to be within the 100% power limit at the following intervals:
- a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading,
- b. At least once per 31 days of accumulated operation in MODE 1, and I
- c. Within four hours if the AZIMUTHAL POWER TILT (Tq) is > 0.020.
4.2.3.3 FTr shall be determined by using the incore detectors to obtain a power distribution map with all CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump Combination.
- See Special Test Exception 3.10.2.
MILLSTONE - UNIT 2 3/4 2-9 Amendment No. -3&, 62, 9, 90,99, A4,49,44, 45, -4864, -M, -M,
POWER DISTRIBUTION LIMITS AZIMUTHAL POWER TILT - To LIMITING CONDITION FOR OPERATION 3.2.4 The AZIMUTHAL POWER TILT (Tq) shall be < 0.02. APPLICABILITY: MODE I with THERMAL POWER> 50% of RATED THERMAL POWERMl). ACTION:
- a. With the indicated Tq > 0.02 but < 0. 10, either restore Tq to < 0.02 within 2 hours or verify the TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR (FTr) is within the limit of Specification 3.2.3 within 2 hours and once per 8 hours thereafter. Or otherwise, reduce THERMAL POWER to
- 50% of RATED THERMAL POWER within the next 4 hours.
- b. With the indicated Tq > 0.10, perform the following actions: (2)
- 1. Verify the TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR (FTr) is within the limit of Specification 3.2.3 within 2 hours; and
- 2. Reduce THERMAL POWER to *50% of RATED THERMAL POWER within 2 hours; and
- 3. Restore Tq
- 0.02 prior to increasing THERMAL POWER. Correct the cause of the out of limit condition prior to increasing THERMAL POWER Subsequent power operation above 50% of RATED THERMAL POWER may proceed provided that the measured Tq is verified < 0.02 at least once per hour for 12 hours, or until verified at 95% of RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.2.4.1 Verify T is within limit at least once every 12 hours. The provisions of Specification 4.0.4 are not applicable for entering into MODE I with THERMAL POWER > 50% of RATED THERMAL POWER from MODE 1. (l) See Special Test Exception 3.10.2. (2) All subsequent Required ACTIONS must be completed if power reduction commences prior to restoring Tq < 0.10. MILLSTONE - UNIT 2 3/4 2-10 Amendment No. 38, A, 90, 439,4-55, 2,
3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor protective instrumentation channels and bypasses of Table 3.3-1 shall be OPERABLE. APPLICABILITY: As shown in Table 3.3-1. ACTION: As shown in Table 3.3-1. SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor protective instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-1. 4.3.1.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation. 4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Neutron detectors are exempt from response time testing. Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1. M ILLST ONE -UN}IT 2 3/4 3-1 Amendment No. X, 495,
TABLE 3.3-1 (Continued) TABLE NOTATION
- With the protective system trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.
(a) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 2 5% of RATED THERMAL POWER. (b) Trip may be manually bypassed when steam generator pressure is < 800 psia and all CEAs are fully inserted; bypass shall be automatically removed when steam generator pressure is > 800 psia. (c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 2 15% of RATED THERMAL POWER. (d) Trip does not need to be OPERABLE if all the control rod drive mechanisms are de-energized or if the RCS boron concentration is greater than or equal to the refueling concentration of
*Specification 3.9.1.
(e) DELETED (f) AT Power input to trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 2 5% of RATED THERMAL POWER. ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 4 hours and/or open the protective system trip breakers. ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may continue provided the following conditions are satisfied:
- a. The inoperable channel is placed in either the bypassed or tripped condition within I hour. The inoperable channel shall either be restored to OPERABLE status, or placed in the tripped condition, within 48 hours.
- b. Within 1 hour, all functional units receiving an input from the inoperable channel are also declared inoperable, and the appropriate actions are taken for the affected functional units.
- c. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 48 hours, provided one of the inoperable channels is placed in the tripped condition.
MILLSTONE - UNIT 2 3/4 3-4 Amendment No. 9, A8, 2, 1416, 49,
,,224, 2
TABLE 3.3-1 (Continued) ACTION STATEMENTS ACTION 3 -NOT USED ACTION 4 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, immediately verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1, and at least once I per 4 hours thereafter. ACTION 5 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours. ACTION 6 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours. MILLSTONE - UNIT 2 3/4 3-5 Amendment No. 2i2, 2,
INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The engineered safety feature actuation system instrumentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4. APPLICABILITY: As shown in Table 3.3-3. ACTION:
- a. With an engineered safety feature actuation system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.34, either adjust the trip setpoint to be consistent with the value specified in the Trip Setpoint column of Table 3.3-4 within 2 hours or declare the channel inoperable and take the ACTION shown in Table 3.3-3.
- b. With an engineered safety feature actuation system instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.
SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each engineered safety feature acutation system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-2. 4.3.2.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation. MILLSTONE - UNIT 2 3/4 3-9 Amendment No. 498, 282,
INSTRUMENTATION ENGINEERED SAFETY FEATURE ACTUATION SYSTEM SENSOR CABINET POWER SUPPLY DRAWERS LIMITING CONDITION FOR OPERATION 3.3.2.2 The engineered safety feature actuation system Sensor Cabinets (RC02A 1, RC02B2, RC02C3 & RC02D4) Power Supply Drawers shall be OPERABLE and energized from the normal power source with the backup power source available. The normal and backup power sources for each sensor cabinet is detailed in Table 3.3-5a: CABINET NORMAL POWER BACKUP POWER RC02AI VA-10 VA-40 RC02B2 VA-20 VA-30 RC02C3 VA-30 VA-20 RC02D4 VA-40 VA-10 Table 3.3-5a APPLICABILITY: MODES 1, 2,3 and 4 ACTION: With any of the Sensor Cabinet Power Supply Drawers inoperable, or either the normal or backup power source not available as delineated in Table 3.3-5a, restore the inoperable Sensor Cabinet Power Supply Drawer to OPERABLE status within 48 hours or be in COLD SHUTDOWN within the next 36 hours. SURVEILLANCE REQUIREMENTS 4.3.2.2.1 The engineered safety feature actuation system Sensor Cabinet Power Supply Drawers shall be determined OPERABLE once per shift by visual inspection of the power supply drawer indicating lamps. 4.3.2.2.2 Verify the OPERABILITY of the Sensor Cabinet Power Supply auctioneering circuit at least once per 18 months. I MILLSTONE - UNIT 2 3/4 3-23 Amendment No. 49, I,
INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm/trip setpoints within the specified limits. APPLICABILITY: As shown in Table 3.3-6. ACTION:
- a. With a radiation monitoring channel alarm/trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 2 hours or declare the channel inoperable.
- b. With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.1.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-3. 4.3.3.1.2 DELETED 4.3.3.1.3 Verify the response time of the control room isolation channel at least once per 18 months. MILLSTONE - UNIT 2 3/4 3-24 Amendment No. 4-I-, 245, I, 284,
INSTRUMENTATION I ACCIDENT MONITORING LIMITING CONDITION FOR OPERATION 3.3.3.8 The accident monitoring instrumentation channels shown in Table 3.3-11 shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. ACTIONS per Table 3.3-1 1. I SURVEILLANCE REQUIREMENTS 4.3.3.8 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.
MILLSTONE - UNIT 2 3/4 3-3 1 Amendment No. 66, 141-, 2,
TABLE 3.3-11 r4 ACCIDENT MONITORING INSTRUMENTATION C,, Minimum 3 Total No. Channels Instrument of Channels OPERABLE ACTION I
- 1. Pressurizer Water Level 2 I 1
- 2. Auxiliary Feedwater Flow Rate 2/S.G. I/S.G. 1 0
- 3. RCS Subcooled/Superheat Monitor 2 2
- 4. PORV Position Indicator I/valve UIvalve 3 Acoustic Monitor
- 5. PORV Block Valve Position 1/valve 1/valve 3 Indicator Z
0 6. Safety Valve Position Indicator 1/valve 1/valve 3 Acoustic Monitor
- 7. Containment Pressure (Wide Range) 2 4
- 8. Containment Water Level (Narrow Range) 7##
- 9. Containment Water Level (Wide Range) 2 4
- 10. Core Exit Thermocouples 4 CETs/core quadrant 2 CETs in any 5 0T of 2 core quadrants
- 11. Main Steam Line Radiation Monitor 3 3 6
- 12. Reactor Vessel Coolant Level 2* 1* 8
- A channel is eight (8) sensors in a probe. A channel is OPERABLE if four (4) or more sensors, two (2) or more in the upper four and two (2) or more in the lower four, are OPERABLE. I
## Refer to ACTION statement in Technical Specification 3.4.6. 1.
TABLE 3.3-11 (Continued) ACTION STATEMENTS ACTION I - With the number of OPERABLE channels less than the MINIMUM CHANNELS OPERABLE requirements of Table 3.3-1 1, either restore the inoperable channel(s) to OPERABLE status within 30 days or be in HOT STANDBY within the next 12 hours. ACTION 2 - With the number of channels OPERABLE less than the MINIMUM CHANNELS OPERABLE, determine the subcooling margin once per 12 hours. ACTION 3 - With any individual valve position indicator inoperable, obtain quench tank temperature, level and pressure information, and monitor discharge pipe temperature once per shift to determine valve position. This ACTION is not I required if the PORV block valve is closed with power removed in accordance with Specification 3.4.3.a or 3.4.3.b. ACTION 4 - a. With the number of OPERABLE accident monitoring instrumentation channels less than the total number of channels shown in Table 3.3-1 1, restore the inoperable channel(s) to OPERABLE status within 7 days, or submit a special report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction, the plans for restoring the channel(s) to OPERABLE status, and any alternate methods in affect for estimating the applicable parameter during the interim.
- b. With the number of OPERABLE accident monitoring instrumentation channels less than the MINIMUM CHANNELS OPERABLE requirements of Table 3.3-1 1, restore the inoperable channel(s) to OPERABLE status within 48 hours, or submit a special report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction, the plans for restoring the channel(s) to OPERABLE status, and any alternate methods in affect for estimating the applicable parameter during the interim.
MILLSTONE - UNIT 2 3/4 3-33 Amendment No. 420, I,
REACTOR COOLANT SYSTEM COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.4.1 Two reactor coolant loops shall be OPERABLE and in operation. APPLICABILITY: MODES I and 2. ACTION: With the requirements of the above specification not met, be in at least HOT STANDBY within 6 hours. SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation at least once per 12 hours. MILLSTONE - UNIT 2 3/4 4-1 Amendment No. Se, 69, 20, 249,
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5. 1.5 Reports
- a. Following each inservice inspection of steam generator tubes, the number of tubes plugged and sleeved in each steam generator shall be reported to the Commission within 15 days.
- b. The complete results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the period in which this inspection was completed. This report shall include:
- 1. Number and extent of tubes inspected.
- 2. Location and percent of wall-thickness penetration for each indication of an imperfection.
- 3. Identification of tubes plugged or sleeved.
- c. Results of steam generator tube inspections which fall into Category C-3 shall be reported pursuant to 10 CFR 50.72. A Special Report pursuant to Specification 6.9.2 shall be submitted prior to resumption of plant operation and shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.
MILLSTONE - UNIT 2 3/4 4-7b Amendment No. -3, 52, 89, 4i,
TABLE 4.4-6 STEAM GENERATOR TUBE INSPECTION 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION H Sample Size Result ACTION Required Result ACTION Required Result ACTION Required 0 z A minimum of C-I None N/A N/A N/A N/A S tubes per C-2 Repair defective C-1 None N/A N/A tjl S.G. tubes and inspect C-2 Repair defective tubes and C-l None 2: t" additional 25 tubes in inspect additional 45 tubes in C-2 Repair defective this S.G.* this S.G.* tubes* W C-3 Perform ACTION for C-3 result of first sample C-3 Perform ACTION for C-3 N/A N/A result of first sample C-3 Inspection all tubes All other None N/A N/A in this S.G., repair S.G.s are C-CD defective tubes and I Q inspect 25 tubes in Some S.G.s Perform ACTION for C-2 N/A N/A each other S.G.* C-2 but no result of second sample Prompt notification additional 3 S.G. are C-3 to NRC pursuant to 03 10 CFR 50.72 Additional Inspect all tubes in each S.G. N/A N/A C-D S.G. is C-3 and repair defective tubes.* z+ Prompt notification to NRC 0 pursuant to 10 CFR 50.72 B S = 3 N% Where N is the number of steam generators in the unit, and n is the number of steam generators inspected during an inspection n
- Repair of defective tubes shall be limited to plugging with the exception of those tubes which may be sleeved. Tubes with defective sleeves shall be plugged.
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co ab "20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT 1-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 1.0 p Ci/gram DOSE EQUIVALENT I-13 1 I MILLSTONE - UNIT 2 3/4 4-16 Amendment No.
3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained. APPLICABILITY: MODES I,2,3 and 4. ACTION: Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
- a. At least once per 31 days by verifying that all penetrations(') not capable of being closed by OPERABLE containment automatic isolation valves(2) and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions,(3) except for valves that are open under administrative control as permitted by Specification 3.6.3.1.
- b. At least once per 31 days by verifying the equipment hatch is closed and sealed.
- c. By verifying the containment air lock is in compliance with the requirements of Specification 3.6.1.3.
- d. After each closing of a penetration subject to type B testing (except the containment air lock), if opened following a Type A or B test, by leak rate testing in accordance with the Containment Leakage Rate Testing Program.
- e. By verifying Containment structural integrity in accordance with the Containment Tendon Surveillance Program.
(1) Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed, or otherwise secured in the closed position. These penetrations shall be verified closed prior to entering MODE 4 from MODE 5, if not performed within the previous 92 days. (2) In MODE 4, the requirement for an OPERABLE containment automatic isolation valve system is satisfied by use of the containment isolation trip pushbuttons (3) Isolation devices in high radiation areas may be verified by use of administrative means. MILLSTONE - UNIT 2 3/4 6-1 Amendment No. 2A, 95, 203,240, 24-, pg.,
CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY AND COOLING SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.2.1 Two containment spray trains and two containment cooling trains, with each cooling train consisting of two containment air recirculation and cooling units, shall be OPERABLE. APPLICABILITY: MODES 1, 2 and 3*. ACTION: I Inoperable Equipment Required ACTION I
- a. Onecontainment a.1 Restore the inoperable containment spray train to spray train OPERABLE status within 72 hours or be in HOT STANDBY within the next 6 hours and reduce pressurizer pressure to less than 1750 psia within the following 6 hours.
- b. One containment b.1 Restore the inoperable containment cooling train tol cooling train OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours.
- c. One containment c.1 Restore the inoperable containment spray train or the spray train inoperable containment cooling train to OPERABLE status AND within 48 hours or be in HOT SHUTDOWN within the next One containment 12 hours.
cooling train
- d. Two d.l Restore at least one inoperable containment cooling train to containment OPERABLE status within 48 hours or be in HOT cooling trains SHUTDOWN within the next 12 hours.
- e. All other e.1 Enter LCO 3.0.3 immediately.
combinations SURVEILLANCE REQUIREMENTS 4.6.2.1.1 Each containment spray train shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying each containment spray manual, power operated, and automatic valve in the spray train flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
- The Containment Spray System is not required to be OPERABLE in MODE 3 if pressurizer pressure is < 1750 psia.
MILLSTONE - UNIT 2 314 6-12 Amendment No. 24-S, 2S, 26, 28,
PLANT SYSTEMS MAIN FEEDWATER ISOLATION COMPONENTS (MFICs) LIMITING CONDITION FOR OPERATION (Continued)
- a. With two or more of the feedwater isolation components inoperable in the same flow path, either:
- 1. Restore the inoperable component(s) to OPERABLE status within 8 hours until ACTION 'a' applies, or
- 2. Isolate the affected flow path within 8 hours, and verify that the inoperable feedwater isolation components are closed or isolated/secured once per 7 days, or
- 3. Be in HOT SHUTDOWN within the next 12 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.6 Each feedwater isolation valve/feedwater pump trip circuitry shall be demonstrated OPERABLE at least once per 18 months by:
- a. Verifying that on 'A' main steam isolation test signal, each isolation valve actuates to its isolation position, and
- b. Verifying that on 'B' main steam isolation test signal, each isolation valve actuates to its isolation position, and
- c. Verifying that on 'A' main steam isolation test signal, each feedwater pump trip circuit actuates, and
- d. Verifying that on 'B' main steam isolation test signal, each feedwater pump trip circuit actuates.
MILLSTONE - UNIT 2 3/4 7-9b Amendment No. M8,
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. Two physically independent circuits between the offsite transmission network and the onsite Class I E distribution system, and
- b. Two separate and independent diesel generators each with a separate fuel oil supply tank containing a minimum of 12,000 gallons of fuel.
APPLICABILITY: MODES 1,2, 3 and 4. ACTION: Inoperable Equipment Required ACTION I
- a. One offsite a.l Perform Surveillance Requirement 4.8.1.1.1 for circuit remaining offsite circuit within I hour prior to or after entering this condition, and at least once per 8 hours thereafter.
AND a.2 Restore the inoperable offsite circuit to OPERABLE status within 72 hours or be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. MILLSTONE - UNIT 2 -3/4 8-1 Amendment No. 4, 17?, 492,23-,
, 264,
ELECTRICAL POWER SYSTEMS ACTION (Continued) Inoperable Equipment Required ACTION
- b. One diesel b.1 Perform Surveillance Requirement 4.8.1.1.1 for the generator offsite circuit within 1 hour prior to or after entering this condition, and at least once per 8 hours thereafter.
AND b.2 Demonstrate OPERABLE diesel generator is not inoperable due to common cause failure within 24 hours or perform Surveillance Requirement 4.8.1.1 .2.a.2 for the OPERABLE diesel generator within 24 hours. AND b.3 Verify the steam-driven auxiliary feedwater pump is OPERABLE (MODES 1, 2, and 3 only). If this condition is not satisfied within 2 hours, be in at least HOT STANDBY within the next 6 hours and HOT SHUTDOWN within the following 6 hours. AND b.4 (Applicable only if the 14 day allowed outage time specified in ACTION Statement b.5 is to be used.) Verify the required Millstone Unit No. 3 diesel generator(s) is/are OPERABLE and the Millstone Unit No. 3 SBO diesel generator is available within I hour prior to or after entering this condition, and at least once per 24 hours thereafter. Restore any inoperable required Millstone Unit No. 3 diesel generator to OPERABLE status and/or Millstone Unit No. 3 SBO diesel generator to available status within 72 hours or be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. AND b.5 Restore the inoperable diesel generator to OPERABLE status within 72 hours (within 14 days if ACTION Statement b.4 is met) or be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. MILLSTONE - UNIT 2 3/4 8-1 a Amendment No. 4-, 4AW, 4-92,234-, 25,4264,
ELECTRICAL POWER SYSTEMS ACTION (Continued) Inoperable Equipment Required ACTION
- c. One offsite lC. One offsite c.l Perform Surveillance Requirement 4.8.1.1.1 for circuit remaining offsite circuit within I hour and at least once per 8 hours thereafter.
AND AND One diesel generator c.2 Demonstrate OPERABLE diesel generator is not inoperable due to common cause failure within 8 hours or perform Surveillance Requirement 4.8.1.1 .2.a.2 for the OPERABLE diesel generator within 8 hours. AND c.3 Verify the steam-driven auxiliary feedwater pump is OPERABLE (MODES 1, 2, and 3 only). If this condition is not satisfied within 2 hours, be in at least HOT STANDBY within the next 6 hours and HOT SHUTDOWN within the following 6 hours. AND c.4 Restore one inoperable A.C. source to OPERABLE status within 12 hours or be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. AND c.5 Restore remaining inoperable A.C. source to OPERABLE status following the time requirements of ACTION Statements a or b above based on the initial loss of the remaining inoperable A.C. source.
- d. Two offsite d. I Restore one of the inoperable offsite sources to circuits OPERABLE status within 24 hours or be in HOT STANDBY within the next 6 hours.
AND d.2 Following restoration of one offsite source restore remaining inoperable offsite source to OPERABLE status following the time requirements of ACTION Statement a above based on the initial loss of the remaining inoperable offsite source. MILLSTONE - UNIT 2 3/4 8-2 Amendment No. 467, 1-77, 2-4-, 261,
ELECTRICAL POWER SYSTEMS ACTION (Continued) Inoperable Equipment Required ACTION
- e. Two diesel e.1 Perform Surveillance Requirement 4.8.1.1.1 for the generators offsite circuits within 1 hour and at least once per 8 hours thereafter.
AND e.2 Restore one of the inoperable diesel generators to OPERABLE status within 2 hours or be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. AND e.3 Following restoration of one diesel generator restore remaining inoperable diesel generator to OPERABLE status following the time requirements of ACTION Statement b above based on the initial loss of the remaining inoperable diesel generator. SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Verify correct breaker alignment and indicated power available for each required offsite circuit at least once per 24 hours. MILLSTONE - UNIT 2 3/4 8-2a Amendment No. 1-31, 23-, 277,
5.0 DESIGN FEATURES 5.1 SITE LOCATION The Unit 2 Containment Building is located on the site at Millstone Point in Waterford, Connecticut. The nearest SITE BOUNDARY on land is 2034 feet northeast of the containment I building wall (1627 feet northeast of the elevated stack), which is the minimum distance to the boundary of the exclusion area as described in 10 CFR 100.3. No part of the site that is closer than these distances shall be sold or leased except to Dominion Nuclear Connecticut, Inc. or its corporate affiliates for use in conjunction with normal utility operations. 5.2 DELETED MILLSTONE - UNIT 2 5 1 Amendment No. 216,470,
TABLE 6.2-1(3) MINIMUM SHIFT-CREW COMPOSITION(2) APPLICABLE MODES LICENSE CATEGORY 1, 2,3 & 4 5&6 Senior Reactor Operator 2 1() Reactor Operator 2 1 Non-Licensed Operator 2 1 Shift Technical Advisor 1(4) None Required (1) Does not include the licensed Senior Reactor or Senior Reactor Operator Limited to Fuel Handling individual supervising CORE ALTERATIONS after the initial fuel loading. I (2) The above shift crew composition and the radiation protection technician of Section 6.2.2 may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence provided expeditious actions are taken to fill the required position. (3) Requirements for minimum number of licensed operators on shift during operation in MODES other than COLD SHUTDOWN or REFUELING are contained in I IOCFR50.54(m). (4) The Shift Technical Advisor position can be filled by either of the two Senior Reactor Operators (a dual-role individual), if he meets the requirements of Specification 6.3.1.b.l. MILLSTONE - UNIT 2 6-4 Amendment No. 66, ,9, -, 1,63, 49+,,24,
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (CONT.)
- 8. XN-NF-621(P)(A), "Exxon Nuclear DNB Correlation for PWR Fuel Designs," Exxon Nuclear Company.
- 9. XN-NF-82-06(P)(A), and Supplements 2,4 and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup," Exxon Nuclear Company.
- 10. ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWdIMTU," Advanced Nuclear Fuels Corporation.
- 11. XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," Exxon Nuclear Company.
- 12. ANF-89-1 51 (P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation.
- 13. EMF-1961 (P)(A), "Statistical Setpoint/Transient Methodology for Combustion Engineering Type Reactors," Siemens Power Corporation.
- 14. EMF-2130(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," Framatome ANP.
- 15. EMF-92-153(P)(A) and Supplement 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation.
- c. The core operating limits shall be determined so that all applicable limits (e.g.,
fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met.
- d. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator, Region 1, and one copy to the NRC Resident Inspector within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:
- a. Deleted MILLSTONE - UNIT 2 6-19 Amendment No. 448, 463, 228, 250 20, 284,,
ADMINISTRATIVE CONTROLS 6.19 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of the primary containment as required by I OCFR50.54(o) and I OCFR50, Appendix J, Option B as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995. The peak calculated primary Containment internal pressure for the design basis loss of coolant accident is Pal The maximum allowable primary containment leakage rate, La, at Pa' is 0.5% of primary containment air weight per day. Leakage rate acceptance criteria are:
- a. Primary containment overall leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and < 0.75 La for Type A tests;
- b. Air lock testing acceptance criteria are:
- 1. Overall air lock leakage rate is < 0.05 La when tested at 2 Pa.
- 2. For each door, pressure decay is
- 0.1 psig when pressurized to 2 25 psi for at least 15 minutes.
The provisions of SR 4.0.2 do not apply for test frequencies specified in the Primary Containment Leakage Rate Testing Program. The provisions of SR 4.0.3 are applicable to the Primary Containment Leakage Rate Testing Program. 6.20 RADIOACTIVE EFFLUENT CONTROLS PROGRAM This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the REMODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
- a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the REMODCM;
- b. Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS, conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to IOCFR 20.1001-20.2402; MILLSTONE - UNIT 2 6-26 Amendment No. X0, MS,-276,
ADMINISTRATIVE CONTROLS 6.20 RADIOACTIVE EFFLUENT CONTROLS PROGRAM
- c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the REMODCM;
- d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS, conforming to 10 CFR 50, Appendix I;
- e. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the REMODCM at least every 31 days.
Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the REMODCM at least every 31 days;
- f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;
- g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY shall be in accordance with the following:
I. For noble gases: a dose rate < 500 mrem/yr to the whole body and a dose rate
- 3000 mrem/yr to the skin, and
- 2. For iodine-131, iodine 133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: a dose rate < 1500 mrem/yr to any organ;
- h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY, conforming to 10 CFR 50, Appendix I;
- i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY, conforming to 10 CFR 50, Appendix l; and
- j. Limitations on the annual dose or dose commitment to any member of the public, beyond the SITE BOUNDARY, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.
The provisions of Specification 4.0.2 and Specification 4.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency. MILLSTONE - UNIT 2 6-27 Amendment No. 2i5, 2I6,
Serial No. 05-009 Docket No. 50-423 ATTACHMENT 4 LICENSE AMENDMENT REQUEST (LBDCR 04-MP3-014) ADMINISTRATIVE CHANGES TO TECHNICAL SPECIFICATIONS EVALUATION OF PROPOSED LICENSE AMENDMENT MILLSTONE POWER STATION UNIT 3 DOMINION NUCLEAR CONNECTICUT, INC.
Serial No. 05-009 Docket No. 50-423 Administrative Changes Attachment 4 Page 1 of 7 EVALUATION OF PROPOSED LICENSE AMENDMENT
1.0 DESCRIPTION
2.0 PROPOSED CHANGE
S
3.0 BACKGROUND
3.1 Recent Improvements in Processing and Implementing Technical Specification Changes at the Millstone Power Station 3.2 Reason for Proposed Amendment
4.0 TECHNICAL ANALYSIS
5.0 REGULATORY ANALYSIS
No Significant Hazards Consideration
6.0 ENVIRONMENTAL CONSIDERATION
Serial No. 05-009 Docket No. 50-423 Administrative Changes Attachment 4 Page 2 of 7
1.0 DESCRIPTION
Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) hereby requests to amend Operating License NPF-49 for Millstone Power Station Unit 3 (MPS3) by incorporating certain administrative changes into MPS3 technical specifications (TS). The proposed changes include:
- 1. Changes to index pages *to correct inconsistencies in titles, which were inadvertently introduced in previous amendments.
- 2. The use of a capitalized typeface for all the DEFINED TERMS specified in Section 1.0, "DEFINITIONS," of the MPS3 technical specifications. This requirement is specified as item 1.1 in this section.
- 3. Correction to the list of amendments affecting page 3/4 1-23 (list at bottom of the page).
- 4. Deletion of ACTION 27 on page 3/4 3-44 of Table 3.3-6.
- 5. Deletion of references to certain technical specifications that no longer exist in MPS3 TS.
- 6. Corrections to station and owner titles and other typographical errors in MPS3 Operating License Appendix B, "Environmental Protection Plan."
- 7. Miscellaneous typographical errors in the technical specifications.
2.0 PROPOSED CHANGE
S Changes To Index Pages The following changes will correct inconsistencies in titles.
- 1. Index Page vii: The titles for Tables 3.4.2 and 4.4.2 will be replaced with the word "DELETED."
- 2. Index Page viii: The titles for Tables 4.4.5 and for "Pressurizer" will be replaced with the word "DELETED."
Changes Involving Capitalization of The TS DEFINED TERMS The following changes will capitalize the typeface for all the DEFINED TERMS as specified in Section 1.0, "DEFINITIONS," of the MPS3 TS. The requirement to capitalize the typeface of these terms is specified in section 1.1 of the TS.
Serial No. 05-009 Docket No. 50-423 Administrative Changes Attachment 4 Page 3 of 7
- 1. Index Page iv: Capitalize the terms "SHUTDOWN MARGIN" and "COLD SHUTDOWN."
- 2. Index Page vii: Capitalize the terms "HOT STANDBY," "HOT SHUTDOWN" and "COLD SHUTDOWN."
- 3. Index Page xix: Capitalize the term "CORE OPERATING LIMITS REPORT."
- 4. Pages 3/4 1-1, 3/4 1-3, 3/4 1-26, 3/4 1-27 and 3/4 2-5: Capitalize the term "CORE OPERATING LIMITS REPORT."
- 5. Pages 3/4 2-6 and 3/4 2-8: Capitalize the term "ACTIONS."
- 6. Page 3/4 2-10: Capitalize the term "THERMAL POWER."
- 7. Page 3/4 2-11: Capitalize the term "ACTIONS."
- 8. Page 3/4 3-7: Capitalize the term "OPERABLE."
- 9. Page 3/4 3-41: Capitalize the term "OPERABLE" and "CHANNEL CALIBRATION."
- 10. Pages 3/4 3-59, 3/4 61 and 3/4 82: Capitalize the term "OPERABLE."
- 11. Page 3/4 3-83: Capitalize the term "CORE OPERATING LIMITS REPORT."
- 12. Page 3/4 4-7: Capitalize the terms "ACTIONS" and "ACTION."
- 13. Page 3/4 4-8: Capitalize the term "MODE."
- 14. Page 3/4 4-12: Capitalize the term "OPERABLE."
- 15. Page 3/4 4-20: Capitalize the term "ACTION."
- 16. Page 3/4 4-21: The term "inoperable" should be in lower case.
- 17. Pages 3/4 8-1, 3/4 8-2, 3/4 8-3 and 3/4 8-3a: Capitalize the term "ACTION."
- 18. Page 5-1: Capitalize the term "SITE BOUNDARY."
- 19. Page 6-1 9a: Capitalize the term "SHUTDOWN MARGIN."
- 20. Page 6-20: Capitalize the terms "AXIAL FLUX DIFFERENCE" and "SHUTDOWN MARGIN."
- 21. Page 6-20a: Capitalize the term "MODES."
- 22. Page 6-21: Capitalize the term "SHUTDOWN MARGIN."
- 23. Page 6-25: Capitalize the terms "UNRESTRICTED AREAS" and "SITE BOUNDARY."
- 24. Page 6-26: Capitalize the term "SITE BOUNDARY."
Correction to The List of Amendments Affecting page 3/4 1-23 Review of Amendment 206, issued by NRC letter dated, July 24, 2002 shows that page 3/4 1-23 was not one of the pages affected by this amendment. Therefore, Amendment 206 is deleted from the list of amendments affecting page 3/4 1-23 (list at bottom of the page). Deletion of ACTION 27 on Page 3/4 3-44 of Table 3.3-6 The ACTION numbers in Section 3/4.3, Instrumentation, follow a single sequence. ACTIONS 1 through 13 are in Table 3.3-1, ACTIONS 14 through 26 are in Table 3.3-3 and ACTIONS 27 through 29 are in Table 3.3-6. Amendment 220, which was issued on
Serial No. 05-009 Docket No. 50-423 Administrative Changes Attachment 4 Page 4 of 7 September 14, 2004, added a new ACTION 27 for loss of power, 4KV Undervoltage in Table 3.3-3. Current TS contains also ACTION 27 in Table 3.3-6, which is labeled as "Not used." Therefore ACTION 27 in Table 3.3-6 (which is not in use) is deleted to avoid confusion. Changes Involving Deletion of References to Certain Technical Specifications that no Longer Exist In Millstone Unit 3 TS The following changes will delete references to certain TSs which have been deleted from the Millstone Unit 3 TS. Page 6-20: The wording "and 3/4.2.1.2" in Item 6 and "and 3/4.2.2.2" in item 7 of Specification 6.9.1.6.a will be deleted. Amendment 217, issued by NRC letter dated December 10, 2003, deleted both TS 3/4.2.1.2 and 3/4.2.2.2, which originally addressed three loop operation. Corrections to Station and Owner Titles And Other Typographical Errors in Appendix B of the Operating License, "Environmental Protection Plan." The word "Nuclear" is deleted from the station name to become "Millstone Power Station" and the owner name is changed from "Northeast Nuclear Energy Company" to Dominion Nuclear Connecticut, Inc." These changes are made to reflect the current station and owner names. Other typographical errors are also corrected as noted on the marked-up pages. Miscellaneous Typographical Errors In The Technical Specifications
- 1. Page 3/4 1-3, Valve ID is changed from "3CHS-V305" to "3CHS*V305."
- 2. Page 3/4 4-21, the word "followng" is changed to "following."
3.0 BACKGROUND
3.1 Recent Improvements in Processing and Implementing Technical Specification Changes at the Millstone Power Station. DNC has employed a new system relying on Adobe FrameMaker software for implementing and tracking changes to the TSs. As part of the conversion to the new system, the MPS2 TS were reviewed for inconsistencies and errors that are administrative in nature and that were introduced as part of the implementation of previous license amendments.
Serial No. 05-009 Docket No. 50-423 Administrative Changes Attachment 4 Page 5 of 7 2.1 Reason for Proposed Amendment The proposed amendment is being requested to remove inconsistencies and administrative errors that were inadvertently introduced during the implementation of the previous license amendments that modified the MPS3 TS.
4.0 TECHNICAL ANALYSIS
Safety Summary The proposed changes include changes to index pages to correct inconsistencies in titles, use of a capitalized typeface for all the DEFINED TERMS specified in Section 1.0, "DEFINITIONS" of the MPS3 TS, deletion of references to certain TS that no longer exist in MPS3 TS, deletion of a duplicate ACTION number, and a correction to station and owner names in Appendix B to the Operating License. The proposed changes will remove inconsistencies and administrative errors that were inadvertently introduced during the implementation of previous license amendments that modified the Millstone Unit 3 TS. These changes are administrative in nature and do not alter any of the requirements of the affected TS. These changes do not alter any of the assumptions used in the safety analyses, nor do they cause any safety system parameters to exceed their acceptance limit. Therefore, the proposed changes have no adverse effect on plant safety. Additionally, these changes can be made without adverse impact to plant operations or to the health and safety of the public.
5.0 REGULATORY ANALYSIS
No Significant Hazards Consideration DNC has evaluated whether or not a significant hazards consideration is involved with the proposed changes by addressing the three standards set forth in 10 CFR 50.92(c) as discussed below. Criterion 1: Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes are administrative in nature and do not alter any of the requirements of the affected TS. The proposed changes do not modify any plant equipment and do not impact any failure modes that could lead to an accident.
Serial No. 05-009 Docket No. 50-423 Administrative Changes Attachment 4 Page 6 of 7 Additionally, the proposed changes have no effect on the consequence of any analyzed accident since the changes do not affect any equipment related to accident mitigation. Based on this discussion, the proposed amendment does not increase the probability or consequences of an accident previously evaluated. Criterion 2: Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes are administrative in nature. They do not modify any plant equipment and there is no impact on the capability of the existing equipment to perform their intended functions. No system setpoints are being modified and no changes are being made to the method in which plant operations are conducted. No new failure modes are introduced by the proposed changes. The proposed amendment does not introduce accident initiators or malfunctions that would cause a new or different kind of accident. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. Criterion 3: Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. These changes are administrative in nature and do not alter any of the requirements of the affected TS. The proposed changes do not affect any of the assumptions used in the accident analysis, nor do they affect any operability requirements for equipment important to plant safety. Therefore, the proposed changes will not result in a significant reduction in the margin of safety as defined in the bases for technical specifications covered in this license amendment request. In summary, DNC concludes that the proposed amendment does not represent a significant hazards consideration under the standards set forth in 10 CFR 50.92(c).
6.0 ENVIRONMENTAL CONSIDERATION
DNC has determined that the proposed amendment would not change requirements with respect to use of a facility component located within the restricted area, as defined by 10 CFR 20, nor would it change inspection or surveillance requirements. DNC has evaluated the proposed change and has determined that the change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant
Serial No. 05-009 Docket No. 50-423 Administrative Changes Attachment 4 Page 7 of 7 increase in the amounts of any effluent that may be released off site, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
Serial No. 05-009 Docket No. 50-423 ATTACHMENT 5 LICENSE AMENDMENT REQUEST (LBDCR 04-MP3-014) ADMINISTRATIVE CHANGES TO TECHNICAL SPECIFICATIONS MARKED-UP PAGES MILLSTONE POWER STATION UNIT 3 DOMINION NUCLEAR CONNECTICUT, INC.
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PA-CB 3/4.0 APPLICABILITY ................................................... 3/40-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR Shudow ~ a 1 ~s......................................................3/4 1-1
- MODE , AND 5 LOOPS FILLED ........................ 3/4 1-3 FIGURE 3.1-1 DELETED.......... ..... ......................................................................... 3/4 1-4 FIGURE 3.1-2 DELETED.... 3/4 1-5 FIGURE 3.1-3 DELET .......... .................................................................................. 3/41-6 FIGURE 3.1-4 DEL ED....................................... 3/41-7 Loops Not Filled .3/4 1-8 FIGURE 3.1-5 DELETED .................................... 3/4 1-9 Moderator Temperature Coefficient ........... ........................ 3/4 1-10 Minimum Temperature for Criticality ................................... 3/4 1-12 3/4.1.2 BORATION SYSTEMS DELETED ........ 3/4 1-13 DELETED ....... 3/4 1-14 DELETED ......... 314 1-15 DELETED ....... 3/4 1-16 DELETED ..... 3/41-17 DELETED ..... 3/41-18 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height ....... 3/4 1-20 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD ...................... 3/4 1-22 Position Indication Systems - Operating ........................ ................ 3/4 1-23 MILLSTONE - UNIT 3 iv Amendment No. , 60,99, 49, 241, 2+)
Decembor 10,2>003s
- INDIZX' LlMfING CONDITIONS FOR OPERATION ANSURVEILLA E REQUIREMENTS SECO-N PAGE TABLE 3.3-13 DELETED TABLE-A.3-9 D 3/4.3.4 DELETED 3/4.3.5 SHUTDOWN MARGIN MONITOR ............................ 3/43,-82 3/4.4 REA'CTO COOLANT SYSTEM
'3/4.4.1 REACTOR COOLANT LOOPS AND COObLANTCIRCULATION
- Startup anidP6We ptio ..................................... -.1
. o SS.t b... = '. ... ...... 3/442 Hot Sbit.o...;:. :.. . 3;4 4-3 ColdL'Shutd.oops. *--
.3/4 4-5 ColdShutd oso . ........................ ...... 3/44-6 Loop Stop Valves .3/444-7 Isolated Loop Startup .. . . 3/4 4-8 3/4.4.2 SAFETY VALVES ........ 3/4.4.2 SAY VALVES
~~~~~~~~.............----.---;--------
---------------. ......-;.-.3/.4..4-93.-.I/4 4r9 DELETED ; : ;.... 3/4 4-10 3/4.4.3 PRESSURIZER Startup and Power Operation . . . .3/4 4-11 FIGURE 3.4-5 PRESSURIZER LEVEL CONTROL ................................. . 3/4 4- la Hot Standby ..... . . . 3/4 4-1 lb 3/4.4.4 RELIEF VALVES . ........................
. 3/44;12 3/4.4.5 STEAM GENERATORS ................................ ;.;.;.;:: ..... ; 3/4 4-14 TABLE 4.4-1 M tMUJMNUMBER OF. STEAM GENERATORS TO-BE INSPECTED DURING INSERVICE INSPECTION ;. . .... 3/4 4-19 TABLE 4.4-2 STEAM GENERATOR TIBE INSPECTION.,.: ;. :, 3/4 4-20 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage*Deteotion Systems
.............. 3/4 4-21 Operational Leakage; ..............
; ... .. .... :. .... 3/4 4-22 TABLE 3.4-1 REACTOR COOLANT S URE ISOLATION VALVES' 3/44-24 3/4.4.7 DELETED.............................................. 3/4 4-25
.TABLE43.4--B.4 sw'i1eiiENnsTY-rmT L ................................ 314 4-26 ; PE f1 Lf[qf- ...............,.;;.;.;...3. 4-7 3/4.4.8 SPECIFIC ACTIVITY .......................... . . . ....3/44-28 MILLSTONE - UNIT 3 ' vii Amendment No. 460, 468, 44-3, 497i _2,+9
Decombcr 10, 2003 e INDEX LIMIMNG CONDIONS FOR OPEgRATIONAND SURVEILLANCE REQUIREMENTS sEcTONN FIGURE 3.4-1 DOSE EQUIVALENT 1-131 REACTOR COOLANT SPEGIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >1 uCigram DOSE EQUIVALENT 1-131 .,;...... , , 3144-30
- TABLE4.4-4 REACTOR COOLANT SPECIFIC ACTIVIIY SAMPIE AND ANALYSIS PROGRAM ............................ ; ; ;;:'
.2.3144-31 314.4.9 PRESSURE/TMPERAtURE OlvfITS. ............................. 3/4 4-33 FIGURE 3:4-2 REACTOR COOLANT SYSTEM.HEATUP LiIATI6NS.-
APPLICABLE UP TO 10 EFPY .... 3 3..................
/4 4-34 FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS -
APPLICABLE UP TO 10 EFPY...................... ; 3/4 4-35 TA53L 4.4-5 - ^ ^t fr9IAL EUR :ILL{NGEPRO GRAM
.; : D ..................... ,..........;;
3/4 4-36 Pr~............... ---.---. 3/4 4-37 Overpressure Protection Systems ....................... 3/4 4-38 FIGURE 3.4-4a High Setpoint PORV Curve For the Cold Overpressure Protection System .3/4 4-40 FIGURE 3.4-4b Low Setpoint PORV Curve For the Cold Overpressure Protection System .3/4 4-41 3/. 4 0 DELET D .................................. ............................................ 3/ 4-4 3/4.4.10 DELETED.3/4 4-42 .3/4.4.11. DELETED .3/4 4-43 3/4.5 EMERGENCY CORE COOLING SYSTEMS -3/4.5.1 ACCUMULATORS ............. ....... .. 3/4:5-1
*,.... . 4 ..............-. .. * '. . . EQU., AL...
3/4.5.2 ECCS SUBSYSTEMS ""Tivg GREATER IAN"OR EQUAL TO 350°FT0350 ....................... 0F ;................;
. . ;........................... 3/4 3.55-3 3/4:5.3 ECCS SUBSYSTEMS - Tag LESS THAN 350°F .: .. . 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK ................................... 3/4 5-9 3/4.5.5. pH TRISODIUM PHOSPHATE STORAGE BASKETS .......................... 3/4 5-10 3/4.6 CONTAINMET SYSTEMS
- 3/4.6.1. PRIMARY CONTAINMENT
,Containment Integrity ................ ;:;.; ;.;.: 3/4 6-1
*ContainmentLeakage . ;...;.: ... 3/4 6-2 Containment Air Locks .3/4 6-5 Containment Pressure .3/4 6-7 MILLSTONE - UNIT .3 viii Amendment No. .59, 87,69+ , 204,
, * ,4 * -
May45 INDEX ADMINlISTRATIVE CONTROLS SECTION PAGE 6.8 PROCEDURES AND PROGRAMS ........... . . . . . . . . . 6-14 6.9. REPORTING REOUIREMENTS . . 6-17 6.9.1 ROUTINE REPORTS .. . . . . . . . . . . . . . . . . . . . . . . . 6-17 Startup Report . . . . . . . . . . . . . . . . . . . . . . . . 6-17 Annual Reports . . . . . . . . . . . . . . . . . . . . . . . ; 6-18 Occupational Radiation Exposure Report . . . . . . . . . . . 6-18 9 Annual Radiological Environmental Operating Report . 6-19 Annual Radioactive Effluent Release Report . . . . . . . . . . 6-19 Monthly Operating Reports . . . . . . . . . . . . . . . . . . .6-19 C-r O ng Lti ts-Lii eRport ................ 6-19a 6.9.2 SPECIAL R PORTS . . . . . . . . . 6-21 6.10 DELETEDC 6.11 RADIATION PROTECTION PROGRAM . .. . .. .. .. . . . . . . . 6-21 . 6.12 HIGH RADIATION AREA . . . . . . . . . . . . . . . . . . . .... ..... 6-21i 6.13 RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REMODCM1 . . . . . . . . . . . . . . . . . 6-24 . 6.14 RADIOACTIVE WASTE TREATMENT . . . . . . . . . . . . . . . . . . . 6-24 6.15 RADIOACTIVE EFFLUENT CONTROLS PROGRAM . . . . . . . . . . . . . . . 6-25 6.16 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM . . . . . . . . . . . 6-26 6.17 REACTOR COOLANT PUMP FLYWHEEL INSPECTION PROGRAM . . . . . ... . 6-26 6.18 TECHNICAL SPECIFICATIONS (TS) BASES CONTROL PROGRAM ....... . 6-26 6.19 COMPONENT CYCLIC-OR TRANSIENT LIMIT . . . . . . . . . . . . . . . 6-27 . MILLSTONE - UNIT 3 XiX Amendment No. Mn, Rn, P, W 0914
*7py, FM Z4+5-lM,
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1,1.1 BORATION CONTROL SHUTDOWN MARGIN - MODES ] AND 2 LIMITING CONDITION FOR OPERATION 3.1.1.1.1 The SHUTDOWN MARGIN shall be within the limits specified in the e4/ Gpo~igLinit neOx(COLR). APPLICABILITY: MODES 1 and ACTION: With the SHUTDOWN MARGIN not within the limits specified in the COLR, immediately initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be within the limits specified in ( the COLR: 1)
- a. Within 1 hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s);
- b. When in MODE I or MODE 2 with Kff greater than or equal to I at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6;
- c. When in MODE 2 with Keff less than 1, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;
- d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specification 4.1.1.1.2, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and
- See Special Test Exceptions Specification 3.10.1.
MILLSTONE - UNIT 3 3/4 1-1 Amendment No. 69, 4-3, 2-7, 24-8
Marefr9T2004- e 3/41 REACTWITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - MODES 3.4 AND 5 LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.1.1.1.2 The SHUTDOWN MARGIN shall be within the limits specifiedinthe Cere _ -d -e OL a e _1 APPLICABILITY: MODES 3,4 and 5 ACTION: With the SHUTDOWN MARGIN less than the required value, immediately initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.2.1 The SHUTDOWN MARGIN shall be determined to be within the limits specified in the COLR:
- a. Within I hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s); and
- b. At least once per 24 hours by consideration of the following factors:
- 1. Reactor Coolant System boron concentration,
- 2. Control rod position,
- 3. Reactor Coolant System average temperature,
- 4. Fuel burnup based on gross thermal energy generation,
- 5. Xenon concentration, and
- 6. Samarium concentration.
4.1.1.1.2.2 Valve 3CHS-V305 shall be verified closed and locked at least once per 31 days.
- Additional SHUTDOWN MARGIN requirements, if required, are given in Specification 3.3.5.
MILLSTONE - UNIT 3 3/4 1-3 Amendment No. 60, 44A*6, 4-4, A 218,
Dz~cbzr 10, 200 REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 The Digital Rod Position Indication System and the Demand Position Indication System shall be OPERABLE and capable of determining the control rod positions within 12 steps. APPLICABILITY: MODES I and 2. ACTION:
- a. With a maximum of one digital rod position indicator per bank inoperable:
- 1. Determine the position of the nonindicating rod(s) indirectly by the movable incore detectors'at least once per 8 hours and immediately after any motion of the nonindicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours.
- b. With a maximum of one demand position indicator per bank inoperable:
I. Verify that all digital rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.2.1 Each digital rod position indicator shall be determined to be OPERABLE by verifying that the Demand Position Indication System and the Digital Rod Position Indication System agree within 12 steps at least once per 12 hours except during time intervals when the rod position deviation monitor is inoperable, then compare the Demand Position Indication System and the Digital Rod Position Indication System at least once per 4 hours. 4.1.3.2.2 Each of the above required digital rod position indicator(s) shall be determined to be OPERABLE by verifying that the digital rod position indicators agree with the demand position indicators within 12 steps when exercised over the full-range of rod travel at least once per 24 months. 9 MILLSTONE - UNIT 3 3/4 1-23 Amendment No. MO, 7, P,@2 R-1-7
March 11, 19j REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMITS I IMTTTN(7 CaNnuTTTAm FOR OPFRATTMN 3.1.3.5 All shutdown rods shall be limited in physical insertion as specified in the c (COLR). APPLICABILITY: MODES 1* and 2* ** ACTION: With a maximum of one shutdown rod inserted beyond the insertion limits specified in the COLR except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour either:
- a. Restore the rod to within the limit specified in the COLR, or
- b. Declare the rod to be inoperable and apply Specification 3.1.3.1.
,11RVFTI I ANrF RFQIITRFMFNTC 4.1.3.5 Each shutdown rod shall be determined to be within the insertion limits specified in the COLR:
- a. Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality, and
- b. At least once per 12 hours thereafter.
- See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
** With Keff greater than or equal to 1.
.MILLSTONE - UNIT 3 3/4 1-26 Amendment No. G>2
REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as specified in the core operating limits report (COLR). APPLICABILITY: MODES 1* and 2* ** LMT & ACTION: With the control banks inserted beyond the insertion limits specified in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1.2:
- a. Restore the control banks to within the limits within 2 hours, or
- b. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank posi-tion using the insertion limits specified in the COLR, or
- c. Be in at least HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals when the rod insertion limit monitor is inoperable, then verify the individual rod positions at least once per 4 hours.
*See Special Test Exceptions Specifications 3.10.2 and 3.10.3. **With Keff greater than or equal to 1.
MILLSTONE - UNIT 3 3/4 27 Amendment No. go 4G>Z 0007
D-eembe 10-,2003-Q-POW9R DISTRIBUTION LIMITS 314.2.2 HEAT FLUX HOT CHANNEL FAVOR - LIMITING CONDITION FORXOPERATION , .. 3.2.2.1 FQ(Z) shallle limited h y the foll6wg relationships:' F RTP FQ(Z) <F K(Z) for' P>i 0.5 F FQ-(Z) . K(Z) for'P *0.5 FQRTP_ the FQ;iimithit RATE-DTHERMAL POWER(RTP) provided in the cor. operatingPRERe). (coW CPE A LrMTS R Where: ER ' aapn dEL RATED'. THERMAI :PQOWER K(Z) = the normalized FQ(Z) as a function of core height specified in the COLR APPLICABILITY: MODE 1. ACTION: With FQ(Z) exceeding its limit:
- a. For RAOC operation with Specification 4.2.2.1 .2.b not being satisfied or for base load operation with Specification 4.2.2.1.4.b not being satisfied:
(1) Reduce THERMAL POWER at least 1% for each 1% FQ(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 ho uis`; POWER OPERATION may proceed for up to a total~of 72 hours; su1sequent POWER OPERATION may proceed provided the Overpower AT Trip setpoints have been reduced at least 1% for each 1% FQ(Z) exceeds the limit, and MILLSTONE - UNIT 3 314 2-5 Amendment No. 'M, 60,99, 420, A,
-21
lofan POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (C ontinued) (2) Identify and correct the cause of the out-of-limit condition priorith increasing THERMAL POWER above the reduced limit required by iteim (1) above; THERMAL POWER may then be increased provided FQ(Z) is demonstrated through incore inaipping to be within its limits.
- b. For RAOC operation with Specification 4.2.2.1.2.c not being satisfied, one of the following aStions shall be taken:
(1) Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specified in the CORE OPERATING LIMiTS REPORT by at least 1% AFD for each percent FQ(Z) exceeds its limits. Within S hours, reset the AFD alarm setpoints to these modified limits, or (2): Verify, that the requirement§ of Specification 4;2.21.3 for base load operation are satisfied 'and enter base load operation... Where it is necessary to-calculate the percent that FQ(Z) exceeds the limits for item (1) above, it shall be. cailated as the maximum percent over the core height (Z), consistent with Sfion4.24.1.2.f, that FQ(Z) exceeds its limit by the following expression: EMt~ FQM(Z) x W(Z) 1 P. x ( Ji x 100 for P > 0.5
.(QZ. x W(Z)
X** 10O for P*0.5'
- x. K(Z)
- c. For base load operation with Specification 4.2.2.1.4.c not being satisfied, one of the following aeon shall be taken (1) Place the core in an equil i ondition where the limit in 4.2.2.1.4.c is satisfied, and remeasure FQM(Z), or MILLSTONE --UNIT 3. 3/4 2-6 Amendment No. .42, A,
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
- c. Satisfying the following relationship:
FR7P x K(Z) Fa"(Z) * .for P > 0.5 P x W(Z) F,'w(Z) < Fg. ()for P :5 0.5 W(Z) x 0.5 where Fa(Z) is the measured Fu(Z) increased by the allowances for manufacturing tolerances and measurement uncertainty, Fw is the FQ limit, K(Z) is the normalized F0 (Z) as a function of core height, P is the relative THERMAL POWER, and W(Z) is the cycle-dependent function that accounts for power distribution transients encountered during normal operation. FATP, K(Z), and W(Z) are specified in the CORE OPERATING LIMITS REPORT as per Specification 6.9.1.6.
- d. Measuring F'0(Z) according to the following schedule:
(1) Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which FQ(Z) was last determined,*** or (2) At least once per 31 Effective Full Power Days, whichever occurs first.
- e. With the maximum value of Fa (Z)
K(Z) over the core height (Z) increasing since the previous determination of FM4(Z), either of the followingt h be taken: (1) Increase F M(Z) by an appropr actor specified in the COLR and verify that this value satisfies the relationship in Specification 4.2.2.1.2.c, or
*** During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and power distribution map outlined.
MILLSTONE - UNIT 3 3/4 2-8 Amendment No. 0, 0, Ai J0 4t
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
- b. During base load operation, if the THERMAL POWER is decreased below APLND then the conditions of 4.2.2.1.3.a shall be satisfied before reentering base load operation.
4.2.2.1.4 During base load operation F,(Z) shall be evaluated to determine if FQ(Z) is within its limit by:
'a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER above APLND.
- b. Evaluate the computed heat flux hot channel factor by performing both of the .foll1owi ng:
(1) Determine the computed heat flux hot channel factor, F.M(Z), by increasing the measured FQM(Z) component of the power distribution map by 3% to account for manufacturing tolerances and further increase the value by 5% to.account for measurement uncertainties, and (2) Verify that FQ'am(Z) satisfies the requirements of Specification 3.2.2.1 for all core plane regions, i.e., 0 - 100% inclusive.
- c. Satisfying the following relationship:
Fm(Z) < F) x KZ) for P > APLND Px W(Z)BL where: FM(Z) is the measured F0,(Z) increased by the allowances for manufacturing tolerances and measurement uncertainty, FRTl is the F. limit, K(Z).is the normalized F,(Z) as a function of core height, P is the relative THERMAL POWER, and W(Z)8 is the cycle-dependent function that accounts for limited power distribution transients encountered during base load operation. FnTPI K(Z), and W(Z)L are specified in the COLR as per Specification 6.9.1.6.
- d. Measuring Fm(Z) in conjunction with target flux difference determi-nation according to the following schedule:
(1) Prior to entering base load operation after satisfying Sec-tion 4.2.2.1.3 unless a full-core flux map has been taken in the previous 31 EFPD with the relative thermal-power having been maintained above APLND for the 24 hours to mapping, and 1eeML eowJeg (2) At least once per 31 Effective Full Power Days. MILLSTONE - UNIT 3 3/4 2-10 Amendment No. p, IF 4,
POweR DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
- e. With the maximum value of FQM(Z)
K(Z) over the core height (Z) increasing since the previous determination of FQM(Z), either of the following ac" shall be taken:
~CI
- 1) Increase FQM(Z) by appropriate factor specified in the COLR and verify that this value satisfies the relationship in Specification 4.2.2.1.4.c, or
- 2) FQM(Z) shall be measured at least once per 7 Effective Full Power Days until 2 successive maps indicate that the maximum value of F M(Z)
K(Z) over the core height (Z) is not increasing.
- f. The limits specified in 4.2.2.1.4.c and 4.2:2.1.4.e are not applicable in the following core plane regions:
- 1) Lower core region 0% to 15%, inclusive.
- 2) Upper core region 85% to 100%, inclusive.
4.2.2.1.5 When FQ(Z) is measured for reasons other than meeting the requirements of Specifications 4.2:2.1.2 or 4.2.2.1.4, an overall measured FQ(Z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty. ..MILL.STONE - UNIT 3 3/4 2-11 Amendment No. -O, 60, 99, 20, F . 7"
.REAL ~~SIRTirE~iNUflMEi MINIMUM TOTAL NO; . .; CHANNELS CHANNELS APPLICABLE td2 FUNCTIONALUNI OF "TO TRIRP ORABLE MODES 3 1.
2.. Manual Reactor Trip Power Range, Neutron Flux 2 2 . 1 1 2 2 1,2 3*, 4*,.5* ' 1 11,
. 2
- a. High Setpoint 4 - . .2 3 1,2 : 2
- b. Low Setpoint 4 2- 3 2
- 3. Power Range, Neutron Flux High Positive Rate 4 2: 3 1,2 *2.
- 4. Deleted W *2 Intermediate Range, Neutron Flux 2
- 1. 3 W 5.
t!j 6. Source Range, Neuton Flux
- a. S5 2 i, .. 2 2## 4
- b. Shutdown 2 1, :2 3*, 4*, 5* 11 :
- 7. Overtemperature AT 4 2 3 1,2 6 (F\
- 8. Overpower.AT 4
- 2:' 3 1,2 6
- 9. Pressurizer Pressure--Low 4 2 3 1** 6 (1) 0 10. Pressurizer Pressure--High 4 2. 3 1,2 6 (1) 0i 0Z
- 11. Pressurizer Water Level-High 3 2. 2 1** 6 p
.p S 11
O3t24I94-TABLE 3.3-1 (Continued) ACTION STATEMENTS (Continued) ACTION 9 - (Not used) ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours; however-, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE. ACTION 11 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or open the Reactor Trip System breakers within the next hour. ACTION 12 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition Within 6 hours, and
- b. When the Minimum Channels OPERABLE requirement is met, the inoperable channel may be bypassed for up to 4 hours for surveillance testing of the Turbine Control Valves.
ACTION 13 - With one of the diverse trip features (undervoltage or shunt trip attachments) inoperable, restore it to OPERABLE status within 48 hours or declare the breaker inoperable and apply ACTION 10. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to
- OPERABLE status.
ACTION 13A - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable Channel to OPERABLE status within 6 hours or be in at least HOT STANDBY within the next 6.hours; however, one channel may be bypassed for up to 4 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is oeR L VILLSTONE - WNIT 3
- 3/4 3 Amendment No. 70, {t, Cosa
Sctmbcr 25, 2003 9-INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The engineered safety feature actuation system instrumentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.34. APPLICABILITY: As shown in Table 3.3-3. ACTION:
- a. With an engineered safety feature actuation system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, either adjust the trip setpoint to be consistent with the value specified in the Trip Setpoint column of Table 3.3-4 within 2 hours or declare the channel inoperable and take the ACTION shown in Table 3.3-3.
- b. With an engineered safety feature actuation system instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.
SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each engineered safety feature acutation system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the mde-and at the frequencies shown in Table 4.3-2. ez-3 4.3.2.1.2 The logic for the bypasses shall be demo ated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels ax by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation. MILLSTONE - UNIT 2 3/4 3-9 Amendment No. 498, 28
July' TABLE 4.3-2 (Continuedj TABLE NOTATION
- 1. Each train shall be tested at least every 62 days on a STAGGERED TEST BAS 0o Peru31Lv
- 2. This surveillance may be performed continuously by the emergency generator a sequencer auto test system as long as the EGLS auto test system is demonstrated by the performance of an ACTUATION LOGIC TEST at least once per 92 days.
- 3. On a monthly basis, a loss of voltage condition will be initiated at each undervoltage monitoring relay to verify individual relay operation. Setpoint verification and actuation of the associated logic and alarm relays will be performed as part of the arhan-k caibmfien required once per 18 months. (
- 4. For Engineered Safety Features Actuation System functional units with only Potter &
Brumfield MDR series relays used in a clean, environmentally controlled cabinet, as discussed in Westinghouse Owners Group Report WCAP- 13900, the surveillance interval for slave relay testing is R.
- MODES1,2,3,4,5and6.
During fuel movement within containment or the spent fuel pool. MILLSTONE - UNIT 3 3/4 3-41 Girt4 by 1 1atvdlJlylOF Amendment No. 45, -4, U79, I-Os,4-29, 4*98, 20324-9,
September 14, 2004 TABLE 3.3-3 (Continued) ACTION STATEMENTS (Continued) ACTION 27 - a. With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- 1. The inoperable channel is placed in the tripped condition within 6 hours, and
- 2. the Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours for surveillance testing of other channels per Specification 4.3.2.1.
- b. With the number of OPERABLE channels one less than the Minimum Channels required OPERABLE:
- 1. Place one channel in bypass and other channel in trip condition within one hour and restore one channel to OPERABLE status in 48 hours, OR
- 2. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
MILLSTONE - UNIT 3 3/4 3-25a Amendment No. 220
June 27, 19=98 --Q TABLE 3.3- (Continued) TABLE NOTATIONS
- With fuel in the fuel storage pool areas.
AseaTe of. ACTION 28 - Un not-used--9__ ACTION STATEMENTS With less than the Minimum Channels OPERABLE requirement, fuel ciY movement may continue for up to 30 days provided an appropriate portable continuous monitor with the same Alarm Setpolnt Is provided in the fuel storage pool area. Restore the inoperable monitors to OPERABLE status within 30 days or suspend all operations involving fuel movement in the fuel storage pool areas. ACTION 29 - With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1. MILSTONE - UNIT 3 3/4 3-44 Amendment No. ff#.v t-0or
INSTRUMENTATIDN ACCIDENT MONITORING' INSTRENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE. APPLLCABILITY: MODES 1, 2, and 3. ACTION:
-a. With the number of OPERABLE accident monitoring instrumentation channels except the containment area high range radiation monitor, the containment hydrogen monitor, and reactor vessel water level, less than the Total Number of Channels shown in Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
- b. With the number of OPERABLE accident monitoring instrumentation channels except the containment area-high range radiation monitor, the containment hydrogen monitor, and reactor vessel water level less than the Minimum Channels OPERABLE requirements of
'Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
- c. With the number of OPERABLE channels for the containment area-high range radiation monitor less than required by either the total or the Minimum Channels OPERABLE requirements, initiate an alternate method of monitoring the appropriate parameter(s), within 72 hours, and either restore the inoperable channel(s) to OPERABLE status within 7 days or prepare and submit a Special Report to the Commis-sion, pursuant to Specification 6.9.2, within 14 days that provides actions taken, cause of the inoperability, and the plans and sched-ule for restoring the channels 0 status.
- d. With the number of OPE clne s for the containment hydrogen monitors less than e total number of channels shown in Table 3.3-10; resto the inoperable channel to OPERABLE status within 30 days or e in at least HOT STANDBY within the next 6 hours and in at least 1 SHUTDOWN within the following 6 hours. With the number of channels for the containment hydrogen monitors less than the minimum channels OPERABLE requirement of Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours: and in at least HOT SHUTDOWN within the following 6 hours.
- e. With the number of OPERABLE channels for the reactor vessel water level monitor less than the Total number of Channels shown in Table 3.3-10, either restore the inoperable channel to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the MILLSrtDNE - UNlIT 3 3/4 3-S9 Amendment No. X7, 77;6)
LI S-. C)C
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INSTRUMENTATION 3/4.3.5 SHUTDOWN MARGIN MONITOR LIMITING CONDITION FOR OPERATION 3.3.5 Two channels of Shutdown Margin Monitors shall be OPERABLE
- a. With a minimum count rate as designated in the CORE OPERATING LIMITS REPORT (COLR), or OEgFL
- b. If the minimum count rate in Specification 3.3. cannot be met, then the Shutdown Margin Monitors may be made with a lower minimum count rate, as specified in the COLR, by borating the Reactor Coolant System above the requirements of Specification 3.1.1.1.2 or 3.1.1.2. The additional boration shall be:
- 1. A minimum of 15 0 ppm above the SHUTDOWN MARGIN requirements specified in the COLR for MODE 3, or 6I
- 2. A minimum of 350 ppm above the SHUTDOWN MARGIN requirements specified in the COLR for MODE 4, MODE S with RCS loops filled, and MODE 5 with RCS loops not filled.
APPLICABILITY: MODES 3 *,4, and 5. ACTION:
- a. With one Shutdown Margin Monitor inoperable, restore the inoperable channel to OPERABLE status within 48 hours.
- b. With both Shutdown Margin Monitors inoperable or one Shutdown Margin Monitor inoperable for greater than 48 hours, immediately suspend all operations involving positive reactivity changes via dilution and rod withdrawal.
Verify the valves listed in Specification 4.1.1.2.2 are closed and secured in position within the next 4 hours and at least once per 14 days thereafter.** Verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1.2 or 3.1.1.2, as applicable, within 1 hour and atleast once per 12 hours thereafter.
- The shutdown margin monitors may be blocked during reactor startup in accordance with approved plant procedures.
- The valves may be opened on an intermittent basis under administrative controls as noted in Surveillance 4.1.1.2.2.
MILLSTONE - UNIT 3 3/4 3-82 Amendment No. 464, 241, 4-
INSTRUMENTATION 3/4.3.5 SHUTDOWN MARGIN MONITOR (continued) SURVEILLANCE REQUIREMENTS 4.3.5 a. Each of the above required shutdown margin monitoring instruments shall be demonstrated OPERABLE by an ANALOG CHANNEL OPERATIONAL TEST at least once per 92 days that shall include I verification that the Shutdown Margin Monitor is set per the dabsr- eF Po R 3 oeT
- b. A-e per 4 hours VERIFY the minimum count rate (counts/sec) as defined within the COLR.
MILLSTONE - UNIT 3 3/4 3-83 Amendment No. 164-- 0495
REACTOR COOLANT SYSTEM LOOP STOP VALVES .. e-LIMTING CONDITION FOR OPERATION . 1. . 3.4.1.5 Each RCS loop stop valve shall be ojipn and the power removed from the valve operator. APPLICABILITY: MODES 1, 2, 3 and 4.. ACTION:
- a. Withpower.available to ont or moteio op stop valve operators, remove power from the loop stop valve operators within 30 minutes br be in HOT STANDBY jig tle next 6 hours and COLD SHUTDOWN within the following'30 hours.
0() With one ornmore RCS loop stop valves closed, maintain the valve(s) closed and be in HOT STANDBY within 6 hours and COLD SHUTDOWN within the next 30 hours..- SURVEILLANCE REQUIWENTS 4.4.1.5 Verify each RCS loop stop valve is open and the power removed foim 'the val"v' eirdtor at least once per 3 Idays. I-(l) All requiredaction o toS tatement 3.4.1.5.b shall be completed whenever this action is entered. RoTIN U A'1&rro isJ MILLSTONE - UNIT 3 3/4 4-7 Amendment No I>:*
Jaafyt27-'- REACTOR COOLANT SYSTEM ISOLATED LOOP STARTUP LIMITING CONDITION FOR OPERATION . 3.4.1.6 A reactor coolant loop shall remain isolated with power removed from the associated RCS loop stop valve operators until:
- a. The temperature at the cold leg of the isolated loop is within 20'F of the highest cold leg temperature of the operating loops, and
- b. The boron concentration of the isolated loop is greater than or equal to the boron concQatratiotrequ redby. Specifications 3.1.1;1.2 or33.1.1.2 for* 5 or Specificiation 3.9.1.1 for 6.
APPLI' 'CADI TY: MODES 5 and 6. AM lQIN: . ,
- a. With the requirements of the above specification satisfied, do n bpen the isolated loop stop valves. / n p t SU RVI HILLANCE REQUIREMENTS /,, . .
4.4.1.6.1 The isolated loop cold leg tern ture shall be determined to be wit in 20'F of the highest cold leg temperature of the o'ting loops within 30 minutes prior to pening the cold leg stop valve. 4.4.1.6.2 The isolated loop bar, concentration shall be determined to be grea r than or equal to the boron concentration r by Specifications 3.1.1.1.2 or 3.1.1.2 for 5 or Specification 3.9.1.1 for within 2 hours prior to opening the hot or cold leg stop valve. MILLSTONE - UNIT 3 3/4 4-8 Amendment No. 42, 1, 60, 4-54;02.
REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4. Both power-operat6d relief valves (PORVs) and their associated block valves shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one or both PORV(s) inoperable because of excessive seat.
leakage, within I hour. either restore the PORV(s) to OPERABLE status or close the associated block valve(s) with power maintained to the block valve(s); otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. With one PORV inoperable due to causes other than excessive'seat leakage, within I hour either restore the PORV.to OPERABLE status or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours or be in HOT STANDBY within the next.6 hours and in HOT SHUTDOWN within the following 6 hours.
- c. With both PORVs inoperable due to causes other than excessive seat leakage, within 1 hour either restore at least one PORV to OPERABLE status or close its associated block valve and remove power from the block valve and be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- d. With one or both block valve(s) inoperable, within I hour restore the block valve(s) to OPERABLE status, or place its associated PORV(s) control switch to "CLOSE." Restore at least one block valve to OPERABLE status within the next hour if both block valves are inoperable; restore any remaining inoperable block valve to status within 72 hours; otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- e. Entry into an OPERATIONAL MODE is permitted while subject to these ACTION requirements.
A
'~Le-MILLSTONE - UNIT 3 3/4 4-12 Amendment No. P7, 99, 45
'-4 TABLE 4.4-2 I-r- --4 STEAM GENERATOR TUBE INSPECTION A CT 0 C) 1ST SAMPLE INSPECTION 2DSAMPLE INSPE , 3RD SAMPLE INSPECTION '-I --4 Sample Size Result Acta Required Result Ao~;Required Result A44n Required CAJ A minimum of C-1 None N. A. N. A. N. A. N. A. S Tubes per S. G. C-2 Plug defective tubes C-1 None N. A. N. A. and inspect additional Plug defective tubes C-1 None 2S tubes in this S. G. C-2 and inspect additional C-2 Plug defective tubes 4S tubes in this S. G. P rorV to o Perfor-m14bo for C-3 C-3 result of first _____ _____sample CA) Perforr for_ C-3 C-3 result of first N. A. N. A.
- ~sample/
C-3 Inspect all tubes in All other this S. G., plug de. S. G.s are None N. A. N. A. fective tubes and C-1 / inspect 2S 2S.tubes inspecthe tuesiin p f Jrfo Some S. G.s Perform ae~tefn for N. A. N. A. each other S. G. C-2 but no C-2 result of second additional sample Notificatior to NRC S. G. are pursuant to §50.72 C-3 (b)(2) of 10 CFR Additional Inspect all tubes in Part 50 S. G. is C-3 each S. G. and plug defective tubes. Notification to NRC N. A. N. A. pursuant to §50.72 (b)(2) of 10 CF R
.__ _Part 50 S = 3 N % Where N is the number of steam generators in the unit, and n is the number of steam generators inspected n during an inspection
REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS Jul 2,4O-LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shal.l be OPERABLE:
- a. Either the Containment Atmosphere Gaseous or Particulate Radioactivity Monitoring System, and
- b. The Containment Drain Sump Level or Pumped Capacity Monitoring System APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
- a. With both the Containment Atmosphere Gaseous and Particulate Radioactivity Monitors rINPE-RA , operation may continue for up to 30 days provided th ontainment Drain Sump Level or Pumped Capacity Monitoring System 's OPERABLE and gaseous grab samples of the containment atmos ere are obtained at least once per 12 hours and analyzed for gros noble gas activity within the subsequent 2 hours; otherwise, be in/at least HOT STANDBY within the next 6 hours and in COLD S in the following 30 hours.
- b. With the on ai nt Drain Sump Level or Pumped Capacity Monitoring System , operation may continue for up to 30 days provided either the Containment Atmosphere Gase6us or Particulate Radioactivity Monitoring System is OPERABLE; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the followng 30 hours.
SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:
- a. Containment Atmosphere Gaseous and Particulate Radioactivity Monitoring Systems-performance of CHANNEL CHECK, CHANNEL CALIBRATION, and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, and
- b. Containment Drain Sump Level and Pumped Capacity Monitoring System-performance of CHANNEL CALIBRATION at least once per 24 months.
MILLSTONE - UNIT 3 3/4 4-21 Amendment No. 77, 7y, 700, 7/,
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES A'Iget 26, 2002 OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. Two physically independent circuits between the offsite transmission network and the onsite Class AE Distribution System, and
- b. Two separate and independent diesel generators, each with:
- 1) A separate day tank containing a minimum volume of 278 gallons of fuel,
- 2) A separate Fuel Storage System containing a minimum volume of 32,760 gallons of fuel,
- 3) A separate fuel transfer pump,
- 4) Lubricating oil storage containing a minimum total volume of 280 gallons of lubricating oil, and
- 5) Capability to transfer lubricating oil from storage to the diesel generator unit.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: 2 Inoperable Equipment Required Aet4ef
- a. One offsite a.1 Perform Surveillance Requirement circuit 4.8.1.1.1.a for remaining offsite circuit within 1 hour prior to or after entering this condition, and at least once per 8 hours thereafter.
AND a.2 Restore the inoperable offsite circuit to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.
- b. One diesel b.1 Perform Surveillance Requirement generator 4.8.1.1.1.a for the offsite circuits within 1 hour prior to or after entering this condition, and at least once per 8 hours thereafter.
AND b.2 Demonstrate OPERABLE diesel generator is not inoperable due to common cause failure within 24 hours or perform Surveillance Requirement 4.8.1.1.2.a.5 for the OPERABLE diesel generator within 24 hours. AND MILLSTONE - UNIT 3 3/4 8-1 Amendment No. XA74,/fZ,,2a4-e
ELECTRICAL POWER SYSTEMS A"Q"' ~ 202 LIMITING CONDITION FOR OPERATION ACTION (continued) 170 Inoperable Equipment IReuired AIto,
- b. One diesel b.3 Verify all required system , sub stems, generator trains, components, and d vices hat depend on the remaining OPERABLE diesel generator as a source of emergency ower a e OPERABLE, and the steam-d iven uxiliary feedwater pump is OPERABI (MOD S 1, 2, and 3 only). If these condi ions a e not satisfied within 2 hours be i at least HOT STANDBY within the n xt 6 ours and in COLD SHUTDOWN within th foll wing 30 hours.
AND b.4 (Applicable only if th 14 y allowed -- outage time specified n Action Statement b.5 is to be used). V rify the required Millstone Unit No. 2 iesel generator(s) is/are OPERABLE and t e Millstone Unit No. 3 SBO diesel gene ator is available within 1 hour prior o or after entering this condition, and t least once per 24 hours thereafter. estore any inoperable required Millstone nit No. 2 diesel generator to OPERABLE status and/or Millstone Unit No. 3 SBO diesel generator to available states within 72 hours or be in at least HOT ANDBY within the next 6 hours and COLD S UTDOWN within the following 30 ho rs. AND b.5 Restore the i operable diesel generator to OPERABLE sta s within 72 hours (within 14 days if Aetion Statement b.4 is met) or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.
- c. One offsite c.l Perform Surveillance Requirement circuit 4.8.1.1.1.a for remaining offsite circuit within 1 hour and at least once per 8 hours thereafter.
AND AND One diesel generator c.2 Demonstrate OPERABLE diesel generator is not-inoperable due to common cause failure within 8 hours or perform Surveillance Requirement 4.8.1.1.2.a.5 for the OPERABLE diesel generator within 8 hours. AND I MILLSTONE - UNIT 3 3/4 8-2 Amendment No. X1
ELECTRICAL POWER SYSTEMS Auagutl 26, 2002 - LIMITING CONDITION FOR OPERATION ACTION (continued) ( Inoperable Equipment Required
- c. One offsite c.3 Verify all required syste s, subsystems, circuit trains, components, and devi es that depend on the remaining OPERABL dit sel generator AND as a source of emergency powE.r are OPERABLE, and the steam- rivE auxiliary One diesel feedwater pump is OPERA LE ( DES 1, 2, and generator 3 only). If these cond tion! are not satisfied within 2 hour , be in at least HOT STANDBY within the ext hours and in COLD SHUTDOWN within th fol owing 30 hours.
AND c.4 Restore one inoperable .C. ource to OPERABLE status within 1 ho irs-or be in at least HOT STANDBY withi *the next 6 hours and COLD SHUTDOWN withi the following 30 hours. AND c.5 Restore remaining ino erable A.C. source to OPERABLE status foll ing th time requirements of State ents a. or b. above based on the initial I ss of the remaining inoperable A.C. so rce.
- d. Two offsite d.1 Restore one of the inoperabl offsite circuits sources to OPERABLE status w hin 24 hours or be in at least HOT STANDB within the next 6 hours.
AND d.2 Following restoration of one offsite source, restore remaining in perable offsite source to OPERABLE s tus following the time requirements of Act en Statement
- a. above based on the initial loss of the remaining inoperable offsite source.
- e. Two diesel e.1 Perform Surveillance Requirement generators 4.8.1.1.1.a for the offsite circuits within 1 hour and at least once per 8 hours thereafter.
AND MILLSTONE - UNIT 3 3/4 8-3 Amendment No. 2-1-9
ELECTRICAL POWER SYSTEMS August 2O64,VJ LIMITING CONDITION FOR OPERATION ACTION (continued) qgp Inoperable Equipment Required AcG4+ef\
- e. Two diesel e.2 Restore one of the inoperable dies 1 generators -generators to OPERABLE status withi 2 hours or be in at least HOT STANDBY within the next 6ihours and COLD SHUTDOWN withi the following 30 hours.
AND e.3 Following restoration of one diesel generator, restore remaining inoperabl diesel generator to OPERABLE status following the time requirements of A w.e; Statement b. above based on the initial loss of the..remaining inoperable diesel generator. SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above requited independent circuits between the offsite transmission network and the Onsite Class IE Distribution System shall be:
- a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and
- b. Demonstrated OPERABLE at least once per 18 months during shutdown by transferring (manually and automatically) unit power supply from the normal circuit to the alternate circuit.
4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:*
- a. At least once per 31 days on a STAGGERED TEST BASIS by:
- 1) Verifying the fuel level in the day tank,
- 2) Verifying the fuel level in the fuel storage tank,
- 3) Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day tank,
- 4) Verifying the lubricating oil inventory in storage,
- 5) Verifying the diesel starts from standby conditions and achieves generator voltage and frequency at 4160 + 420 volts and 60 + 0.8 Hz. The diesel generator shall be started for this test by using one of the following signals:
a) Manual, or
- All planned starts for the purpose of these surveillances may be preceded by an engine prelube period.
MILLSTONE - UNIT 3 3/4 8-3a Amendment No. 7p, PA, 717, IAd, 40--
5.0 DESIGN FEATURES - tsar6 DVN AeC 5.1 SITE LOCATION The Unit 3 Containment Building is located on e site at Millstone Point in Waterford, Connecticut. -The nearest b on land is 1719 feet northeast of the containment building wall (1627 feet northeast of the elevated stack), which is the minimum distance to the boundary of the exclusion area as described in 10 CFR 100.3. No part of the site that is closer than these distances shall be sold or leased except to Dominion Nuclear Connecticut, Inc. or its corporate affiliates for use in conjunction with Ix normal utility operations. 5.2 DELETED MILLSTONE - UNIT 3 5-1 Amendment No. .- 2.
ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 6.9.1.5 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator Region I, and one copy to the NRC Resident Inspector, no later than the 15th of each month following the calendar month covered by the report. CORE OPERATING LIMITS REPORT 6.9.1.6 a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:
- 1. Ove A C ower AT setpoint parameters for Specification 2.2.1,
- 2. huLix1 for Specifications 3/4.1.1.1.1, 3/4.1.1.1.2, and 3/4.1.1.2,
- 3. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3.
MILLSTONE - UNIT 3 6-19a Amendment No. 24, 3 69, 6, i-&&, I
Mff*i9, 2004-4 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)
- 4. Shutdown Rod Insertion Limit for Specification 3/4.1.3.5,
- 5. Control Rod Insertion Limits for Specification 3/4.1.3.6,
- 6. Axialsrlux erenc¢mi gband et op
- 7. Heat Flux Hot Channel Factor, K(z), W(z), APLYD, and for Specificationot9 3/4.2.2.1 and V44 ) \
- 8. Nuclear Enthalpy Rise Hot Channel Factor, Power Factor Mult Ii for Specification 314.2.3.\\
- 9. DNB Parameters for Specification 3/4.2.5.
- 10. Shutdown Margin Monitor minimum count rate for Specification /4. .5.
- 11. Boron Concentration for Specification 3/4.9.1.1.
6.9.1.6.b The analytical methods used to determine the core operating limits s 11 e those previously reviewed and approved by the NRC in:
- 1. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVAL]A ION METHODOLOGY," (W[ Proprietary). (Methodology for Specifications 3.1.1.3--Moderator Temperature Coefficient, 3.1.3.5-- tdown Bank Insertion Limit, 3.1.3.6--Control Bank Insertion Limits, 3.2.1- Pigforenee, 3.2.2--Heat Flux Hot Channel Factor, 3.2.3--Nuclear Enthalpy Rise Hot hannel Factor, 3.1.1.1.1, 3.1.1.1.2, 3.1.1.2 -- 3.9.1. -- Boro Concentration.)
- 2. T. M. Anderson to K. Kniel (Chief of Core Performance Branch, NRC), January 31, 1980-
Attachment:
Operation and Safety-Analysis Aspects of an Impro ed Load Follow Package.
- 3. NUREG-800, Standard Review Plan, U.S. Nuclear Regulatory Commis ion, Section 4.3, Nuclear Design, July 1981 Branch Technical Position CPB .3-1, Westinghouse Constant Axial Offset Control (CAOC), Revision 2, July 981.
- 4. WCAP-10216-P-A-RIA, "RELAXATION OF CONSTANT AXIAL OF SET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION," l\
(W Proprietary). (Methodology for Specifications 3.2.1--Ati444ue Di. efee [Relaxed Axial Offset Control] and 3.2.2--Heat Flux Hot Channel Factor [W(z) surveillance requirements for FQ Methodology].) s 5. WCAP-9561-P-A,ADD.3, "BARTA-I:ACOMPUTERCODEFORTHEBEST ESTIMATE ANALYSIS OF REFLOOD TRANSIENTS--SPECIAL REPORT: THIMBLE MODELING W ECCS EVALUATION MODEL," (Yl Proprietary). (Methodology for Specification 3.2.2--Heat Flux Hot Channel Factor.)
- 6. WCAP-10266-P-A, Addendum 1, "THE 1981 VERSION OF THE WESTINGHOUSE ECCS EVALUATION MODEL USING THE BASH CODE,"
(W Proprietary). (Methodology for Specification 3.2.2--Heat Flux Hot Channel FactoT.) MILLSTONE - UNIT 3 6-20 Amendment No. 24, 3;, 60, 69, 84, 420,440, 8-) J
MW647f24_- ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)
- 7. WCAP- 11946, "Safety Evaluation Supporting a More Negative EOL Moderator Temperature Coefficient Technical Specification for the Millstone Nuclear Power Station Unit 3," QW- Proprietary).
- 8. WCAP-10054-P-A, "WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL.17 USING THE NOTRUMP CODE," (EW Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
- 9. WCAP-10079-P-A, "NOTRUMP - A NODAL TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE," (NY Proprietary). (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
- 10. WCAP-12610, "VANTAGE+ Fuel Assembly Report," (Y Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
- 11. Letter from V. L. Rooney (USNRC) to J. F. Opeka, "Safety Evaluation for Topical Report, NUSCO-152, Addendum 4, 'Physics Methodology for PWR Reload Design,' TAC No. M91815," July 18, 1995.
- 12. Letter from E. J. Mroczka to the USNRC, "Proposed Changes to Technical Specifications, Cycle 4 Reload Submittal - Boron Dilution Analysis," B13678, December 4, 1990.
- 13. Letter from D. H. Jaffe (USNRC) to E. J. Mroczka, "Issuance of Amendment (TAC No. 77924),' 11, 1 9 9 g.
Q at toopnsnb~
- 14. Letter from M. H. B to the USNRC, "Proposed Revision to Technical Specification, 9wd Requirements and Shutdown Margin Monitor ZA °for and 5 (PTISCR 3-16-97), B16447, May 9, 1997.
Letter from J. erson (USNRC) to M. L. Bowling (NNECO), "Issuance of Amnendment - Millstone Nuclear Power Station, Unit No. 3 (TAC No. M98699)," October21, 1998.
- 16. WCAP-8301, "LOCTA-IV Program: Loss-of-Coolant Transient Analysis."
- 17. WCAP-10054-P-A, Addendum 2, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model."
- 18. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," (Westinghouse Proprietary Class 2).
(Methodology for Specification 2.2.1.) MILLSTONE - UNIT 3 6-20a Amendment No. 4l-,4-70, 2
May4&2-e'~ ADMINISTRATIVE CONTROLS 6.9.1.6.c The core operating limits shall be determined so that all applica-ble limits (e.g. fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as sh' t4na4 4*, and transient and accident analysis limits) of the safety analysis are methi TD or v3 MA 6.9.1.6.d The CORE OPERATING LIMITS REPORT, including yTid-cle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control -Desk with copies to the Regional Adminis-trator and Resident Inspector.
.SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. -Nuclear Regulatory
- Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator Region I and one copy to the NRC Resident Inspector, within the time period specified for each report.
6.10 Deleted. 6.11 RADIATION PROTECTIOH PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be -approved, maintained, and adhered to for all operations involving personnel radiation exposure. 6.12 HIGH RADIATION AREA ) As provided in paragraph 20.1601(c:) of 10 CFR Part 20, the following controls shall be applied to bigh radiation areas in place of the controls required by paragraph 20.el601a) and (b) of 10-CFR Part 20: 97 6.12.1 High Radiation Areas .wlth. Dose Rates Not Exceedl ng 1.0 rem/hour at 30 Ce'ntimeters f rom -the-Radiation Source or -from -any surface Penetrated -hv the RadjtDo. a; Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.
- 6. Access to, and activities in, each such area shall. bo controlled by
'jsteans of -aRadiation-Werk Permit tRWP) .or equivalent that includes
*specificatlon of radi tion dose rites in the idinediate-work-area(s) and other Appropriate radiation protecti6n equipnient and measures.
- c.' IndiViduals qUalified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requiremeht for an RWP or equivalent while performing their
-assigned duties provided that they are otherwise following plant radiation protection protedures for entry to, exit from, and work in such areas..
- d. Each individual or group entering such an area shall possess:
- 1. A radiation monitoring device that continuously displays radiation.dose rates in the area, or MILLSTONE'- UNIT 3 6-21 Amendment No. 7A, pp. Pt M. -
091t 709. 272. '777. '-'13 C
ADMINTITRAT1VF DiJTRn1 S 6.15 RADIOACTIVE EFFLUENT CONTROLS PROGRAM This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses'to' members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the REMODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
- a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests :and setpoint determination in accordance with the methodology in the REODCM;
- b. Limitations on the concentrations of radioactive material released in
.liquid effluents to , conforming to ten times the concentration values.1 Appendix B. Table 2, Column 2 to 10 CFR 20.1001-20.2402; VT FzI=EPR
- c. Monitoring, sampling, and analysis of radioa ve liquid and gaseous effluents in accordance with 30 CFR 20.1302 an with the methodology and parameters in the REHODCH;
- d. Limitations on the annual and quarterl doses or dose commitment to a member of the public from rad oactlv materials in liquid effluents released from each unit to , conforming to 10 CFR 50, Appendix I;I
- e. *Determjnation of cumulative dose Contributions from radioactive effluents
-for the current calendar quarter and current calendar year in accordance with the methodoloy and parameters-in the REHODCH at least every 31 days.
Determination of projected dose contributions from radioactive effluents in accordance with the methodology. in the RENODCH at least every 31 days;
- f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems. are used to reduce -releases of radioactivity when the projected
.doses in a period.of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix 1;
- g. Limitations on the dose rate resulting from radioactive-material released in gaseous effluents-.from the site to areas at or beyond the shall be in accordance with the following:.
- 1. For noble gases: a dose rate - 500 mrem/yr to the whole body and
- dose rate < 3000 mrenVyr to the skin, and
- 2. For. iodine-131, iodine-13 , and all adionuclides in particulate form wit f-lives greater than days: a dose rate <
*500jmrem/yr to a gan; h.. Limitations on annual and quarterly ai s resulting from noble. gases released i gaseous effluents from o ch.unit to areas. beyond the siut euminy, conforming to 10 CFR ,Appendix I;
- 4. Litnitations on the annual and tery doses to a member of the public from iodine-131 iodine-133, tr and all radionuclides in particulate 1um,
. form with half lives > 8 da s n gaseous effluents released from each unit to areas beyond the sy , conforminj to 10 CFR 50, Appendix'l; and 6-25 IIIL STON
- UIT Hnn Too 36-25Amendment 0915
ADMINISTRATIVECONTROLS -X__ S. Limitations an the annual coe or dose commitment to any member of the e public beyond the , due to.releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190. The provisions of Specification 4.0.2 and Specification 4.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency. 6.16 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM A program shall. be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provided (1) representative measurements of radioactivity in the highest potential exposure pathwas, and (2)verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the REMODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the fol lowing:,
- a. Monitoring sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the REMODCM.
- b. A Land Use Census to-ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and
- c. Participation in' a Interlaboratory Comparison Program .to. ensure that ihdependent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.
6.17 REACTOR COOLANT PUMP FLYWHEEL INSPECTION PROGRAM This program shall provide for the inspection of each reactor coolant pump flywheel by either qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle of one-half the outer radius or a surface examination (magnetic particle testing and/or penetrant testing) of exposed surfaces defined by the volume of the disassembled flywheels at least once every lD years. 6.18 TECHNICAL SPEOIFICATIONS (TS1 BASES CONTROL 'PROGRAM This program provides a means for processing changes to the Bases of these Technical Specifications:
- a. Changes to the Bases Vf the TS shall be made under appropriate
.administrative controls and reviews.
- b. Licensees may make changes to Bases without prior NRC approval provided the changes. do not require either of the following:
- 1. A change in the TS incorporated in the license or
- 2. A change to updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
MILLSTONE - UNIt 3 6-26 Amendment No. , Ej, FEZ, S--
APPENDIX B TO FACILITY OPERATING LICENSE NO. NPF-49 MILLSTONE NUCUEAZ-OWER STATION, UNIT 3 toiLo'4 A1/LJCLFe,'IP. CVAIC--C NC) DOCKET NO. 50423 ENVIRONMENTAL PROTECTION PLAN (NONRADIOLOGICAL)
MILLSTONE POfwEIZOWER STATION, UNIT NO. 3 ENVIRONMENTAL PROTECTION PLAN (NONRADIOLOGICAL) TABLE OF CONTENTS Section Page 1.0 Objectives of the Environmental Protection Plan . . . . . 1-1 2.0 Environmental Protection Issues .................... 2-1 3.0 Consistency Requirements ....................... . . 3-1 3.1 Plant Design and Operation ..................... . . . 3-2 3.2 Reporting Related to the NPDES Permit and State Certifications ..... 3-3 3.3 Changes Required for Compliance with Other Environmental Regulations ... 3-3 4.0 onditions . . .. ................................................. 4-1 4.1 r portant Environmental Events . . . . . 4-1 4.2 Environmental Monitoring . . . .. 4-1 5.0 Administrative Procedures ............................ .. . . 5-1 5.1 Review.....................................................................................................................................5-1 5.2 Records Retention........................................................................................................................5-1 5.3 Changes in Environmental Protection Plan . . . . . 5-2. 5.4 Plant Reporting Requirements.............................................................................. . ................. 5-2 Millstone Unit 3
RoY h 4- -{ Pn4, January31, 1986 1.0 Objectives of the Environmental Protection Plan The Environmental Protection Plan (EPP) is to provide for protection of non-radiological environmental values during operation of the nuclear facility. The principal objectives of the EPP are as follows: (1) Verify that the facility is operated in an environmentally acceptable manner, as established by the Final Environmental Statement - Operating Licensing Stage (FES-OL) and other NRC environmental impact assessments. (2) Coordinate NRC requirements and maintain consistency with other Federal, State and local requirements for environmental protection. (3) Keep NRC informed of the environmental effects of facility construction and of actions taken to control those effects. Environmental concerns identified in the FES-OL which relate to water quality matters are regulated by way of the licensee's NPDES permit. Millstone Unit 3 1-1
Janmary3l9 'n86 2.0 Environmental Protection Issues . In the FES-OL dated December, 1984, the staff considered the environmental impacts associated with the operation of Millstone Cuele Power Station, Unit No. 3. No environmental issues were identified which required study or license conditions to resolve environmental concerns and to assure adequate protection of the environment. Millstone Unit 3 2-1
written evaluation of such activity and obtain prior NRC approval. When such activity involves a change in the EPP, such activity and change to the EPP may be implemented only in accordance with an appropriate license amendment as set forth in Section 5.3 of this EPP. A proposed change, test or experiment shall be deemed to involve an unreviewed environmental questio it concerns: (1) a matter which may result in a significant increase in any adverse environmental impact previously evaluated in the FES-OL, environmental impact appraisals, or in any decisions of the Atomic Safety and Licensing Board; or ( 2 ) a significant change in effluents or power level; or (3) a matter, not previously reviewed and evaluated in the documents specified in (1) of this Subsection, which may have a significant adverse environmental impact. The licensee shall maintain records of changes in facility design or operation and of tests and experiments carried out pursuant to this Subsection. These records shall include written evaluations which provide bases for the determination that the change, test, or experiment does not'involve an unreviewed environmental question or constitute a decrease in the effectiveness of this EPP to meet the objectives specified in Section 1.0. The licensee shall include as part of the Annual Environmental Operating Reporit Subsection 5.4.1) brief descriptions, analyses, interpretations, and evaluations of such changes tests and experiments. I; Millstone Unit 3 3-2
JouaqJI19a69-5.0 Administrative Procedures 5.1 Review The licensee shall provide for review of compliance with the EPP. The reviews shall be conducted independently of the individual or groups responsible for performing the specific activity. A description of the organization structure utilized to achieve the independent review function and results of the review activities shall be maintained $made available for inspection. 52 Records Retention Records and logs relative to the environmental aspects of station operation shall be made and retained in a manner convenient for review and inspection. These records and logs shall be made available to NRC on request. Records of modifications to station structures, systems and components determined to potentially affect the continued protection of the environment shall be retained for the life of the station. All other records, data and logs relating to this EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies. Millstone Unit 3 5-1
Januay31,' 0 86 5.3 Changes in Environmental Protection Plan Requests for changes in the EPP shall include an assessment of the environmental impact of the proposed change and a supporting justification. Implementation of such changes in the EPP shall not commence prior to NRC approval of the proposed changes in the form of a license amendment incorporating the appropriate revision to the EPP. 5.4 Plant Reporting Requirements 5.4.1 Routine Reports An Annual Environmental Operating Report mplementation of this EPP for the previous year shall be submitted to the NRC on or before May I of each year. The initial report shall be submitted on or before May I of the year following start of commercial operation of the plant. The report shall include summaries and analyses of the results of the environmental protection activities required by Subsection 4.2 of this EPP for the report period, including a comparison with related preoperational studies, operational controls (as appropriate) and previous nonradiological environmental monitoring reports, and an assessment of the observed impacts of the plant operation on the environment. If harmful effects or evidence of trends toward Millstone Unit 3 5-2
January 31, 1986 irreversible damage to the environment are observed, the licensee shall provide a detailed analysis of the data and a proposed course of mitigating action. The Annual Environmental Operating Report shall also include: (1) A list of EPP noncompliancefd the corrective actions taken to remedy them. (2 ) A list of all changes in station design-or operation, tests, and experiments made in accordance with Subsection 3.1 which involved a potentially significant unreviewed environmental question. (3) A list of nonroutine reports submitted in accordance with Subsection 5.4.2. In the event that some results are not available by the report due date, the report shall be submitted noting and explaining the missing results. The minsing results shall be submitted as soon as possible in a supplementary report. 5.4.2 Nonroutine Reports A written report shall be submitted to the NRC within 30 days of occurrence of a nonroutine event. The report shall (a) describe, analyze, and evaluate the event, including extent and magnitude of the impact, and plant operating Millstone Unit 3 .5-3
January 31, 1986 characteristics, (b) describe the probable cause of the event, (c) indicate the action taken to correct the reported event, (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (e) indicate the agencies notified and their preliminary responses Events reportable under this subsection which also require reports to other Federal, State or local agencies shall be reported in accordance with those reporting requirements in lieu of the requirements of this Subsection. The NRC shall be provided with a copy of such report at the same time it is submitted to the other agency. Millstone Unit 34 54
Serial No. 05-009 Docket No. 50-423 ATTACHMENT 6 LICENSE AMENDMENT REQUEST (LBDCR 04-MP3-014) ADMINISTRATIVE CHANGES TO TECHNICAL SPECIFICATIONS RE-TYPED PAGES MILLSTONE POWER STATION UNIT 3 DOMINION NUCLEAR CONNECTICUT, INC.
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY ......................................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - MODES I AND 2 .............................................. 3/4 1-1 SHUTDOWN MARGIN - MODES 3,4, AND 5 LOOPS FILLED ............. 3/4 1-3 FIGURE 3.1-1 DELETED .......................................................... 3/4 1-4 FIGURE 3.1-2 DELETED .......................................................... 3/4 1-5 FIGURE 3.1-3 DELETED .......................................................... 3/4 1-6 FIGURE 3.1-4 DELETED ......................................................... 3/4 1-7 SHUTDOWN MARGIN - COLD SHUTDOWN - Loops Not Filled .................................... 3/4 1-8 FIGURE 3.1-5 DELETED .................................... 3/4 1-9 Moderator Temperature Coefficient .................................... 3/4 1-10 Minimum Temperature for Criticality .................................... 3/4 1-12 3/4.1.2 BORATION SYSTEMS DELETED ....... 3/4 1-13 DELETED ....... 3/4 1-14 DELETED ....... 3/4 1-15 DELETED ....... 3/4 1-16 DELETED .............................................................. 3/41-17 DELETED ..... 3/4 1-18 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height ....... 3/4 1-20 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD ...................... 3/4 1-22 Position Indication Systems - Operating ......................................... 3/4 1-23 MILLSTONE - UNIT 3 iv Amendment No. 50, 60,99, 49, 2, 24-,
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-13 DELETED TABLE 4.3-9 DELETED 3/4.3.4 DELETED 3/4.3.5 SHUTDOWN MARGIN MONITOR ................................. 3/4 3-82 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation . 3/4 4-1 HOT STANDBY .3/4 4-2 HOT SHUTDOWN .3/4 4-3 COLD SHUTDOWN - Loops Filled .3/4 4-5 COLD SHUTDOWN - Loops Not Filled .3/4 4-6 Loop Stop Valves .3/4 4-7 Isolated Loop Startup .3/4 4-8 3/4.4.2 SAFETY VALVES .3/4 4-9 DELETED .3/4 4-10 3/4.4.3 PRESSURIZER Startup and Power Operation ............................................................. 3/4 4-11 FIGURE 3.4-5 PRESSURIZER LEVEL CONTROL .................................................... 3/4 4-1 la Hot Standby ............................................................. 3/4 4-1 lb 3/4.4.4 RELIEF VALVES ............................................................. 3/4 4-12 3/4.4.5 STEAM GENERATORS ............................................................. 3/4 4-14 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION ............................... 3/4 4-19 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION .......................................... 3/4 4-20 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems ............................................................. 3/4 4-21 Operational Leakage ............................................................. 3/4 4-22 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES ..3/4 4-24 3/4.4.7 DELETED ............................................................. 3/4 4-25 TABLE 3.4-2 DELETED ............................................................. 3/4 4-26 TABLE 4.4-3 DELETED ............................................................. 3/4 4-27 3/4.4.8 SPECIFIC ACTIVITY ............................................................. 3/4 4-28 MILLSTONE - UNIT 3 vii Amendment No. 46i, 4-64, 1, 49-3, 49-7,244,247,
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE FIGURE 3.4-1 DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >lI Ci/gram DOSE EQUIVALENT I-131 ....................................... 3/4 4-30 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM ....................................... 3/4 4-31 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ....................................... 3/4 4-33 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 10 EFPY....................................... 3/4 4-34 FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS - APPLICABLE UP TO 10 EFPY ....................................... 3/4 4-35 TABLE 4.4-5 DELETED .... 3/4 4-36 DELETED .... 3/4 4-37 Overpressure Protection Systems .... 3/4 4-38 FIGURE 3.4-4a High Setpoint PORV Curve For the Cold Overpressure Protection System .... 3/4 4-40 FIGUIRE 3.4-4b Low Setpoint PORV Curve For the Cold Overpressure Protection System .... 3/4 4-41 3/4.4.10 DELETED .... 3/4 4-42 3/4.4.11 DELETED .... 3/4 4-43 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ............ 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 3500 F ....................................... 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 3500 F .................................... 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK ....................................... 3/4 5-9 3/4.5.5 pH TRISODIUM PHOSPHATE STORAGE BASKETS ........................ 3/4 5-10 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity ... 3/4 6-1 Containment Leakage ... 3/4 6-2 Containment Air Locks ... 3/4 6-5 Containment Pressure ... 3/4 6-7 MILLSTONE - UNIT 3 viii Amendment No. -9, A, 99, 44I, 204, 24-,
INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.8 PROCEDURES AND PROGRAMS ........................... 6-14 6.9 REPORTING REOUIREMENTS ........................... 6-17 6.9.1 ROUTINE REPORTS .. 6-17 Startup Report .6-17 Annual Reports .6-18 Occupational Radiological Exposure Report .6-18 Annual Radiological Environmental Operating Report .6-19 Annual Radioactive Effluent Release Report .6-19 Monthly Operating Reports .6-19 CORE OPERATING LIMITS REPORT .6-19a I 6.9.2 SPECIAL REPORTS ........................ 6-21 6.10 DELETED 6.11 RADIATION PROTECTION PROGRAM . 6-21 6.12 HIGH RADIATION AREA . 6-21 6.13 RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REMODCM) .. 6-24 6.14 RADIOACTIVE WASTE TREATMENT . 6-24 6.15 RADIOACTIVE EFFLUENT CONTROLS PROGRAM . 6-25 6.16 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM . 6-26 6.17 REACTOR COOLANT PUMP FLYWHEEL INSPECTION PROGRAM . 6-26 6.18 TECHNICAL SPECIFICATIONS (TS) BASES CONTROL PROGRAM . 6-26 6.19 COMPONENT CYCLIC OR TRANSIENT LIMIT . 6-27 MILLSTONE - UNIT 3 xix Amendment No. 56, 69,86, 4A, 4-8, 204, 242,245,
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1.1 BORATION CONTROL SHUTDOWN MARGIN - MODES 1 AND 2 LIMITING CONDITION FOR OPERATION 3.1.1.1.1 The SHUTDOWN MARGIN shall be within the limits specified in the CORE OPERATING LIMITS REPORT (COLR). APPLICABILITY: MODES I and 2*. ACTION: With the SHUTDOWN MARGIN not within the limits specified in the COLR, immediately initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be within the limits specified in the COLR:
- a. Within I hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s);
- b. When in MODE I or MODE 2 with Keff greater than or equal to I at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6;
- c. When in MODE 2 with Keff less than 1, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;
- d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specification 4.1.1.1.2, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and
- See Special Test Exceptions Specification 3.10.1.
MILLSTONE - UNIT 3 3/4 1-] Amendment No. 60, 4-, 24I, 18,
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - MODES 3.4 AND 5 LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.1.1.1.2 The SHUTDOWN MARGIN shall be within the limits specified in the CORE OPERATING LIMITS REPORT (COLR).* I APPLICABILITY: MODES 3,4 and 5 ACTION: With the SHUTDOWN MARGIN less than the required value, immediately initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.2.1 The SHUTDOWN MARGIN shall be determined to be within the limits specified in the COLR:
- a. Within I hour after detection of an inoperable control rod(s) and at least once per 12 hours thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s); and
- b. At least once per 24 hours by consideration of the following factors:
- 1. Reactor Coolant System boron concentration,
- 2. Control rod position,
- 3. Reactor Coolant System average temperature,
- 4. Fuel burnup based on gross thermal energy generation,
- 5. Xenon concentration, and
- 6. Samarium concentration.
4.1.1.1.2.2 Valve 3CHS*V305 shall be verified closed and locked at least once per 31 days. I
- Additional SHUTDOWN MARGIN requirements, if required, are given in Specification 3.3.5.
MILLSTONE - UNIT 3 3/4 1-3 Amendment No. 60, 443,464,24-7, 248,
REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 The Digital Rod Position Indication System and the Demand Position Indication System shall be OPERABLE and capable of determining the control rod positions within +/-12 steps. APPLICABILITY: MODES I and 2. ACTION:
- a. With a maximum of one digital rod position indicator per bank inoperable:
- 1. Determine the position of the nonindicating rod(s) indirectly by the movable incore detectors at least once per 8 hours and immediately after any motion of the nonindicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours.
- b. With a maximum of one demand position indicator per bank inoperable:
I. Verify that all digital rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.2.1 Each digital rod position indicator shall be determined to be OPERABLE by verifying that the Demand Position Indication System and the Digital Rod Position Indication System agree within 12 steps at least once per 12 hours except during time intervals when the rod position deviation monitor is inoperable, then compare the Demand Position Indication System and the Digital Rod Position Indication System at least once per 4 hours. 4.1.3.2.2 Each of the above required digital rod position indicator(s) shall be determined to be OPERABLE by verifying that the digital rod position indicators agree with the demand position indicators within 12 steps when exercised over the full-range of rod travel at least once per 24 months. MILLSTONE - UNIT 3 314 1-23 Amendment No. 500, 2-7, I
REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT (COLR). I APPLICABILITY: MODES I* and 2***. ACTION: With a maximum of one shutdown rod inserted beyond the insertion limits specified in the COLR except for surveillance testing pursuant to Specification 4.1.3.1.2, within I hour either:
- a. Restore the rod to within the limit specified in the COLR, or
- b. Declare the rod to be inoperable and apply Specification 3.1.3.1.
SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be within the insertion limits specified in the COLR:
- a. Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality, and
- b. At least once per 12 hours thereafter.
- See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
- With Keff greater than or equal to I.
MILLSTONE - UNIT 3 3/4 1-26 Amendment No. 60,
REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT (COLR). I APPLICABILITY: MODES l
- and 2***.
ACTION: With the control banks inserted beyond the insertion limits specified in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1.2:
- a. Restore the control banks to within the limits within 2 hours, or
- b. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the insertion limits specified in the COLR, or
- c. Be in at least HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals when the rod insertion limit monitor is inoperable, then verify the individual rod positions at least once per 4 hours.
- See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
- With Keff greater than or equal to 1.
MILLSTONE - UNIT 3 3/4 1-27 Amendment No. M, 64,
POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F_) LIMITING CONDITION FOR OPERATION 3.2.2.1 FQ(Z) shall be limited by the following relationships: RTP FQ(Z)
- Q P
. K(Z) for P> 0.5 RTP FQ(Z)
- 0 K(Z) for P<0.5 FQRTP = the FQ limit at RATED THERMAL POWER (RTP) provided in the CORE OPERATING LIMITS REPORT (COLR). I Whee p= THERMAL POWER an RATED THERMAL POWER' K(Z) = the normalized FQ(Z) as a function of core height specified in the COLR.
APPLICABILITY: MODE l. ACTION: With FQ(Z) exceeding its limit:
- a. For RAOC operation with Specification 4.2.2.1.2.b not being satisfied or for base load operation with Specification 4.2.2.1.4.b not being satisfied:
(I) Reduce THERMAL POWER at least 1% for each 1% FQ(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower AT Trip setpoints have been reduced at least 1% for each 1% FQ(Z) exceeds the limit, and MILLSTONE - UNIT 3 3/4 2-5 Amendment No. M, 60, 99, 4-20, 170, 247,
POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued) (2) Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by item (I) above; THERMAL POWER may then be increased provided FQ(Z) is demonstrated through incore mapping to be within its limits.
- b. For RAOC operation with Specification 4.2.2.1 .2.c not being satisfied, one of the following ACTIONS shall be taken:
(1) Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specified in the CORE OPERATING LIMITS REPORT by at least 1% AFD for each percent FQ(Z) exceeds its limits. Within 8 hours, reset the AFD alarm setpoints to these modified limits, or (2) Verify that the requirements of Specification 4.2.2.1.3 for base load operation are satisfied and enter base load operation. Where it is necessary to calculate the percent that FQ(Z) exceeds the limits for item (I) above, it shall be calculated as the maximum percent over the core height (Z), consistent with Specification 4.2.2.1 .2.f, that FQ(Z) exceeds its limit by the following expression: FFQ (Z) XW(Z) Q FRTP I x100 forP>0.5 I x K(Z) FQ (Z) x W(Z) QFTP -I x 100 for P < 0.5 05 x K(Z)
- c. For base load operation with Specification 4.2.2.1.4.c not being satisfied, one of the following ACTIONS shall be taken:
(1) Place the core in an equilibrium condition where the limit in 4.2.2.1.4.c is satisfied, and remeasure FQM(Z), or MILLSTONE - UNIT 3 3/4 2-6 Amendment No. 99,A, 470,
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
- c. Satisfying the following relationship:
F P x K(Z) F M Z)<QfoP> . (Z)* PxW(Z) foPO. FRTP xKZ F M (Z) FQ ()for P < 0.5 FQ(Z W(Z) xO.5 frP*. where FQM(Z) is the measured FQ(Z) increased by the allowances for manufacturing tolerances and measurement uncertainty, FQRTP is the FQ limit, K(Z) is the normalized FQ(Z) as a function of core height, P is the relative THERMAL POWER, and W(Z) is the cycle-dependent function that accounts for power distribution transients encountered during normal operation. FQRTP, K(Z), and W(Z) are specified in the CORE OPERATING LIMITS REPORT as per Specification 6.9.1.6.
- d. Measuring FQM(Z) according to the following schedule:
(1) Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which FQ(Z) was last determined,*** or (2) At least once per 31 Effective Full Power Days, whichever occurs first.
- e. With the maximum value of M
FQ (Z) K(Z) over the core height (Z) increasing since the previous determination of FQM(Z), either of the following ACTIONS shall be taken: (1) Increase FQM(Z) by an appropriate factor specified in the COLR and verify that this value satisfies the relationship in Specification 4.2.2.1.2.c, or
- During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and power distribution map outlined.
MILLSTONE - UNIT 3 3/4 2-8 Amendment No. ", 60, 99, 4G-, 4,
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
- b. During base load operation, if the THERMAL POWER is decreased below APLND then the conditions of 4.2.2.1.3.a shall be satisfied before reentering base load operation.
4.2.2.1.4 During base load operation FQ(Z) shall be evaluated to determine if FQ(Z) is within its limit by:
- a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER above APLND.
- b. Evaluate the computed heat flux hot channel factor by performing both of the following:
(1) Determine the computed heat flux hot channel factor, FQM(Z), by increasing the measured FQM(Z) component of the power distribution map by 3% to account for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties, and (2) Verify that FQM(Z) satisfies the requirements of Specification 3.2.2.1 for all core plane regions, i.e., 0 - 100% inclusive.
- c. Satisfying the following relationship:
FRTP xKZ M <K(Z)for F P > APLND FQ (Z)* P xW(Z)B3L where: FQM(Z) is the measured FQ(Z) increased by the allowances for manufacturing tolerances and measurement uncertainty, FQRTP is the FQ limit, K(Z) is the normalized FQ(Z) as a function of core height, P is the relative THERMAL POWER, and W(Z)BL is the cycle-dependent function that accounts for limited power distribution transients encountered during base load operation. FQRTP, K(Z), and W(Z)BL are specified in the COLR as per Specification 6.9.1.6.
- d. Measuring FQM(Z) in conjunction with target flux difference determination according to the following schedule:
(1) Prior to entering base load operation after satisfying Section 4.2.2.1.3 unless a full core flux map has been taken in the previous 31 EFPD with the relative THERMAL POWER having been maintained above APLND for the 24 hours prior to mapping, and (2) At least once per 31 Effective Full Power Days. MILLSTONE - UNIT 3 3/4 2-10 Amendment No. -O, 60, 99, 470,
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
- e. With the maximum value of FK(Z)
K(Z) over the core height (Z) increasing since the previous determination of FQM(Z), either of the following ACTIONS shall be taken: I
- 1) Increase FQM(Z) by appropriate factor specified in the COLR and verify that this value satisfies the relationship in Specification 4.2.2.1.4.c, or
- 2) FQM(Z) shall be measured at least once per 7 Effective Full Power Days until 2 successive maps indicate that the maximum value of FQM(Z)
K(Z) over the core height (Z) is not increasing.
- f. The limits specified in 4.2.2.1.4.c and 4.2.2.1.4.e are not applicable in the following core plane regions:
- 1) Lower core region 0°%o to 15%, inclusive.
- 2) Upper core region 85% to 100%, inclusive.
4.2.2.1.5 When FQ(Z) is measured for reasons other than meeting the requirements of Specifications 4.2.2.1.2 or 4.2.2.1.4, an overall measured FQ(Z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty. MILLSTONE - UNIT 3 3/4 2-11 Amendment No. M, 60, 99, 4, 40,
TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM H: TOTAL NO. CHANNELS CHANNELS APPLICABLE 0- FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION z I. Manual Reactor Trip 2 1 2 1,2 I Hn tj 2 1 2 3*, 4*, 5* 11
- 2. Power Range, Neutron Flux
- a. High Setpoint 4 2 3 1,2 2
- b. Low Setpoint 4 2 3 1###, 2 2 W 3. Power Range, Neutron Flux It High Positive Rate 4 2 3 1,2 2
- 4. Deleted
- 5. Intermediate Range, Neutron Flux 2 I 2 I###, 2 3
- 6. Source Range, Neutron Flux
- a. STARTUP 2 I 2 2#11 4 I
- b. Shutdown 2 1 2 3*, 4*, 5* 11 3
01
- 7. Overtemperature AT 4 2 3 1,2 6 0*
- 8. Overpower AT 4 2 3 1,2 6 3
ah
- 9. Pressurizer Pressure--Low 4 2 3 I** 6 (1) co CDh
- 10. Pressurizer Pressure--High 4 2 3 1,2 6 (1)
- 11. Pressurizer Water Level--High 3 2 2 I** 6 P
P
TABLE 3.3-1 (Continued) ACTION STATEMENTS (Continued) ACTION 9- (Not used) ACTION 10- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE. ACTION 11- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or open the Reactor Trip System breakers within the next hour. ACTION 12- With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 6 hours, and
- b. When the Minimum Channels OPERABLE requirement is met, the inoperable channel may be bypassed for up to 4 hours for surveillance testing of the Turbine Control Valves.
ACTION 13- With one of the diverse trip features (undervoltage or shunt trip attachments) inoperable, restore it to OPERABLE status within 48 hours or declare the breaker inoperable and apply ACTION 10. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status. ACTION 13A- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable Channel to OPERABLE status within 6 hours or be in at least HOT STANDBY within the next 6 hours; however, one channel may be bypassed for up to 4 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE. I MILLSTONE - UNIT 3 3/4 3-7 Amendment No. i@, 8,
TABLE 4.3-2 (Continued) TABLE NOTATION I. Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
- 2. This surveillance may be performed continuously by the emergency generator load sequencer auto test system as long as the EGLS auto test system is demonstrated OPERABLE by the performance of an ACTUATION LOGIC TEST at least once per 92 I days.
- 3. On a monthly basis, a loss of voltage condition will be initiated at each undervoltage monitoring relay to verify individual relay operation. Setpoint verification and actuation of the associated logic and alarm relays will be performed as part of the CHANNEL CALIBRATION required once per 18 months. I
- 4. For Engineered Safety Features Actuation System functional units with only Potter &
Brumfield MDR series relays used in a clean, environmentally controlled cabinet, as discussed in Westinghouse Owners Group Report WCAP- 13900, the surveillance interval f6r slave relay testing is R.
- MODES 1, 2, 3, 4, 5 and 6.
During fuel movement within containment or the spent fuel pool. MILLSTONE - UNIT 3 3/4 3-41 Amendment No. 45,44,9, 400, 129, 4-98,20,24-9,
TABLE 3.3-6 (Continued) TABLE NOTATIONS
- With fuel in the fuel storage pool areas.
ACTION STATEMENTS I ACTION 28 - With less than the Minimum Channels OPERABLE requirement, fuel movement may continue for up to 30 days provided an appropriate portable continuous monitor with the same Alarm Setpoint is provided in the fuel storage pool area. Restore the inoperable monitors to OPERABLE status within 30 days or suspend all operations involving fuel movement in the fuel storage pool areas. ACTION 29 - With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1. MILLSTONE - UNIT 3 3/4 3-44 Amendment No. 46, 4-29,
INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With the number of OPERABLE accident monitoring instrumentation channels except the containment area high range radiation monitor, the containment hydrogen monitor, and reactor vessel water level, less than the Total Number of Channels shown in Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
- b. With the number of OPERABLE accident monitoring instrumentation channels except the containment area-high range radiation monitor, the containment hydrogen monitor, and reactor vessel water level less than the Minimum Channels OPERABLE requirements of Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
- c. With the number of OPERABLE channels for the containment area-high range radiation monitor less than required by either the total or the Minimum Channels OPERABLE requirements, initiate an alternate method of monitoring the appropriate parameter(s), within 72 hours, and either restore the inoperable channel(s) to OPERABLE status within 7 days or prepare and submit a Special Report to the Commission, pursuant to Specification 6.9.2, within 14 days that provides actions taken, cause of the inoperability, and the plans and schedule for restoring the channels to OPERABLE status.
- d. With the number of OPERABLE channels for the containment hydrogen monitors less than the total number of channels shown in Table 3.3-10, restore the inoperable channel to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours. With the number of OPERABLE channels for the containment hydrogen monitors less than the minimum channels OPERABLE requirement of Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
- e. With the number of OPERABLE channels for the reactor vessel water level monitor less than the Total number of Channels shown in Table 3.3-10, either restore the inoperable channel to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the MILLSTONE - UNIT 3 3/4 3-59 Amendment No. 4, 5-7, X76,
C-, TABLE 3.3-10 (Continued) 0 ACCIDENT MONITORING INSTRUMENTATION z4 M~ TOTAL MINIMUM NO. OF CHANNELS E INSTRUMENT CHANNELS OPERABLE
- 16. Containment Area - High Range Radiation Monitor 2 I
- 17. Reactor Vessel Water Level 2* 1*
- 18. Containment Hydrogen Monitor 2 1
- 19. Neutron Flux 2 I w
- A channel consists of eight sensors in a probe. A channel is OPERABLE if four or more sensors, half or more in the upper head region and half or more in the upper plenum region, are OPERABLE. I E3 0.
=3 zo
INSTRUMENTATION 3/4.3.5 SHUTDOWN MARGIN MONITOR LIMITING CONDITION FOR OPERATION 3.3.5 Two channels of Shutdown Margin Monitors shall be OPERABLE
- a. With a minimum count rate as designated in the CORE OPERATING LIMITS REPORT (COLR), or
- b. If the minimum count rate in Specification 3.3.5.a cannot be met, then the Shutdown Margin Monitors may be made OPERABLE with a lower minimum count rate, as specified in the COLR, by borating the Reactor Coolant System above the requirements of Specification 3.1.1.1.2 or 3.1.1.2. The additional boration shall be:
I. A minimum of 150 ppm above the SHUTDOWN MARGIN requirements specified in the COLR for MODE 3, or
- 2. A minimum of 350 ppm above the SHUTDOWN MARGIN requirements specified in the COLR for MODE 4, MODE 5 with RCS loops filled, and MODE 5 with RCS loops not filled.
APPLICABILITY: MODES 3*, 4, and 5. ACTION:
- a. With one Shutdown Margin Monitor inoperable, restore the inoperable channel to OPERABLE status within 48 hours.
- b. With both Shutdown Margin Monitors inoperable or one Shutdown Margin Monitor inoperable for greater than 48 hours, immediately suspend all operations involving positive reactivity changes via dilution and rod withdrawal.
Verify the valves listed in Specification 4.1.1.2.2 are closed and secured in position within the next 4 hours and at least once per 14 days thereafter.** Verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1.2 or 3.1.1.2, as applicable, within 1 hour and at least once per 12 hours thereafter.
- The shutdown margin monitors may be blocked during reactor startup in accordance with approved plant procedures.
- The valves may be opened on an intermittent basis under administrative controls as noted in Surveillance 4.1.1.2.2.
MILLSTONE - UNIT 3 3/4 3-82 Amendment No. 464, 4-7, 2-8,
INSTRUMENTATION 3/4.3.5 SHUTDOWN MARGIN MONITOR (continued) SURVEILLANCE REQUIREMENTS 4.3.5 a. Each of the above required shutdown margin monitoring instruments shall be demonstrated OPERABLE by an ANALOG CHANNEL OPERATIONAL TEST at least once per 92 days that shall include verification that the Shutdown Margin Monitor is set per the CORE OPERATING LIMITS REPORT (COLR). I
- b. At least once per 24 hours VERIFY the minimum count rate (counts/sec) as defined within the COLR.
MILLSTONE - UNIT 3 3/4 3-83 Amendment No. 464,
REACTOR COOLANT SYSTEM LOOP STOP VALVES LIMITING CONDITION FOR OPERATION 3.4.1.5 Each RCS loop stop valve shall be open and the power removed from the valve operator. APPLICABILITY: MODES I,2,3 and 4. ACTION:
- a. With power available to one or more loop stop valve operators, remove power from the loop stop valve operators within 30 minutes or be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.
b.(') With one or more RCS loop stop valves closed, maintain the valve(s) closed and be in HOT STANDBY within 6 hours and COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.4.1.5 Verify each RCS loop stop valve is open and the power removed from the valve operator at least once per 31 days. (I) All required ACTIONS of ACTION Statement 3.4.1.5.b shall be completed whenever this I action is entered. MILLSTONE - UNIT 3 3/4 4-7 Amendment No. 2417,
REACTOR COOLANT SYSTEM ISOLATED LOOP STARTUP LIMITING CONDITION FOR OPERATION 3.4.1.6 A reactor coolant loop shall remain isolated with power removed from the associated RCS loop stop valve operators until:
- a. The temperature at the cold leg of the isolated loop is within 20'F of the highest cold leg temperature of the operating loops, and
- b. The boron concentration of the isolated loop is greater than or equal to the boron concentration required by Specifications 3.1.1.1.2 or 3.1.1.2 for MODE 5 or Specification 3.9.1.1 for MODE 6.
APPLICABILITY: MODES 5 and 6. ACTION:
- a. With the requirements of the above specification not satisfied, do not open the isolated loop stop valves.
SURVEILLANCE REQUIREMENTS 4.4.1.6.1 The isolated loop cold leg temperature shall be determined to be within 20'F of the highest cold leg temperature of the operating loops within 30 minutes prior to opening the cold leg stop valve. 4.4.1.6.2 The isolated loop boron concentration shall be determined to be greater than or equal to the boron concentration required by Specifications 3.1.1.1.2 or 3.1.1.2 for MODE 5 or Specification 3.9.1.1 for MODE 6 within 2 hours prior to opening the hot or cold leg stop valve. MILLSTONE - UNIT 3 3/4 4-8 Amendment No. 42, 7, 60, 4-5, 2 ,
REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4. Both power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one or both PORV(s) inoperable because of excessive seat leakage, within 1 hour either restore the PORV(s) to OPERABLE status or close the associated block valve(s) with power maintained to the block valve(s); otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. With one PORV inoperable due to causes other than excessive seat leakage, within I hour either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours or be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- c. With both PORVs inoperable due to causes other than excessive seat leakage, within 1 hour either restore at least one PORV to OPERABLE status or close its associated block valve and remove power from the block valve and be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- d. With one or both block valve(s) inoperable, within I hour restore the block valve(s) to OPERABLE status, or place its associated PORV(s) control switch to "CLOSE." Restore at least one block valve to OPERABLE status within the next hour if both block valves are inoperable; restore any remaining inoperable block valve to OPERABLE status within 72 hours; otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- e. Entry into an OPERATIONAL MODE is permitted while subject to these ACTION requirements.
MILLSTONE - UNIT 3 3/4 4-12 Amendment No. -, SS, -141-,
TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION r I ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result ACTION Required Result ACTION Required Result ACTION Required A minimum of CT- None N.A. N.A. N.A. S Tubes per C-l None N.A.N.A. S.Q C-2 Plug defective tubes C2 Plug defective tubes C None w and inspect additional and inspect additional C-2 Plug defective tubes 2S tubes in this S.G. 4S tubes in this S.GQ Perform ACTION for C-3 result of first sample Perform ACTION for N.A. N.A. C-3 C-3 result of first sample C-3 Inspect all tubes in All other None N.A
- 0. this S.G, plug de- S.G.s are fective tubes and C-1 inspect 2S tubes in Some S.s Perform ACTION for N.A. N.A.
each other S.G C-2 but no C-2 result of second z additional sample U3 Notification to NRC S.G are pursuant to §50.72 C-3 (b)(2) of 10 CFR Additional Inspect all tubes in each N.A. N.A 24 Part 50 S.G. is C-3 S.G and plug defective tubes. Notification to NRC pursuant to
§50.72 (b)(2) of 10 CFR Part 50 S= 3N /Where N is the number of steam generators in the unit, and n is the number of steam generators inspected during an inspection
REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:
- a. Either the Containment Atmosphere Gaseous or Particulate Radioactivity Monitoring System, and
- b. The Containment Drain Sump Level or Pumped Capacity Monitoring System APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
- a. With both the Containment Atmosphere Gaseous and Particulate Radioactivity Monitors inoperable, operation may continue for up to 30 days provided the Containment Drain Sump Level or Pumped Capacity Monitoring System is OPERABLE and gaseous grab samples of the containment atmosphere are obtained at least once per 12 hours and analyzed for gross noble gas activity within the subsequent 2 hours; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With the Containment Drain Sump Level or Pumped Capacity Monitoring System inoperable, operation may continue for up to 30 days provided either the Containment Atmosphere Gaseous or Particulate Radioactivity Monitoring System is OPERABLE; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:
- a. Containment Atmosphere Gaseous and Particulate Radioactivity Monitoring Systems-performance of CHANNEL CHECK, CHANNEL CALIBRATION, and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, and
- b. Containment Drain Sump Level and Pumped Capacity Monitoring System-performance of CHANNEL CALIBRATION at least once per 24 months.
MILLSTONE - UNIT 3 3/4 4-21 Amendment No. 4-, -9, 400, 33, I06,
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. Two physically independent circuits between the offsite transmission network and the onsite Class I E Distribution System, and
- b. Two separate and independent diesel generators, each with:
- 1) A separate day tank containing a minimum volume of 278 gallons of fuel,
- 2) A separate Fuel Storage System containing a minimum volume of 32,760 gallons of fuel,
- 3) A separate fuel transfer pump,
- 4) Lubricating oil storage containing a minimum total volume of 280 gallons of lubricating oil, and
- 5) Capability to transfer lubricating oil from storage to the diesel generator unit.
APPLICABILITY: MODES 1, 2,3, and 4. ACTION: bquipment Required ACIIUN Inoperable Equipment Required ACTION I
- a. One offsite circuit a.1 Perform Surveillance Requirement 4.8.1.1 .1 .a for remaining offsite circuit within 1 hour prior to or after entering this condition, and at least once per 8 hours thereafter.
AND a.2 Restore the inoperable offsite circuit to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.
- b. One diesel generator b.l Perform Surveillance Requirement 4.8.1.l.1.a for the offsite circuits within 1 hour prior to or after entering this condition, and at least once per 8 hours thereafter.
AND b.2 Demonstrate OPERABLE diesel generator is not inoperable due to common cause failure within 24 hours or perform Surveillance Requirement 4.8.1.1 .2.a.5 for the OPERABLE diesel generator within 24 hours. AND MILLSTONE - UNIT 3 3/4 8-1 Amendment No. 64, 97, 42,20,
ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (continued) Inoperable Equipment I Required ACTION
- b. One diesel generator b.3 Verify all required systems, subsystems, trains, components, and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are OPERABLE, and the steam-driven auxiliary feedwater pump is OPERABLE (MODES 1, 2, and 3 only). If these conditions are not satisfied within 2 hours, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
AND b.4 (Applicable only if the 14 day allowed outage time specified in ACTION Statement b.5 is to be used). Verify the required Millstone Unit No. 2 diesel generator(s) is/are OPERABLE and the Millstone Unit No. 3 SBO diesel generator is available within 1 hour prior to or after entering this condition, and at least once per 24 hours thereafter. Restore any inoperable required Millstone Unit No. 2 diesel generator to OPERABLE status and/or Millstone Unit No. 3 SBO diesel generator to available status within 72 hours or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. AND b.5 Restore the inoperable diesel generator to OPERABLE status within 72 hours (within 14 days if ACTION Statement b.4 is met) or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.
- c. One offsite circuit c.1 Perform Surveillance Requirement 4.8.1.1. .Lafor remaining offsite circuit within 1 hour and at least AND once per 8 hours thereafter.
AND One diesel generator c.2 Demonstrate OPERABLE diesel generator is not inoperable due to common cause failure within 8 hours or perform Surveillance Requirement 4.8.1.1 .2.a.5 for the OPERABLE diesel generator within 8 hours. AND MILLSTONE - UNIT 3 3/4 8-2 Amendment No. 240,
ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (continued) Inoperable Equipment Required ACTION C. One offsite circuit c.3 Verify all required systems, subsystems, trains, AND components, and devices that depend on the remaining OPERABLE diesel generator as a source One diesel generator of emergency power are OPERABLE, and the steam-driven auxiliary feedwater pump is OPERABLE (MODES 1, 2, and 3 only). If these conditions are not satisfied within 2 hours, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. AND c.4 Restore one inoperable A.C. source to OPERABLE status within 12 hours or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. AND c.5 Restore remaining inoperable A.C. source to OPERABLE status following the time requirements of ACTION Statements a. or b. above based on the initial loss of the remaining inoperable A.C. source.
- d. Two offsite circuits d.1 Restore one of the inoperable offsite sources to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours.
AND d.2 Following restoration of one offsite source, restore remaining inoperable offsite source to OPERABLE status following the time requirements of ACTION Statement a. above based on the initial loss of the remaining inoperable offsite source.
- e. Two diesel generators e.1 Perform Surveillance Requirement 4.8.1.1. .La for the offsite circuits within I hour and at least once per 8 hours thereafter.
AND MILLSTONE - UNIT 3 3/4 8-3 Amendment No. 2410,
ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (continued) Inoperable Equipment Required ACTION
- e. Two diesel generators e.2 Restore one of the inoperable diesel generators to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.
AND e.3 Following restoration of one diesel generator, restore remaining inoperable diesel generator to OPERABLE status following the time requirements of ACTION Statement b. above based on the initial loss of the remaining inoperable diesel generator. SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the Onsite Class IE Distribution System shall be:
- a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and
- b. Demonstrated OPERABLE at least once per 18 months during shutdown by transferring (manually and automatically) unit power supply from the normal circuit to the alternate circuit.
4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:*
- a. At least once per 31 days on a STAGGERED TEST BASIS by:
- 1) Verifying the fuel level in the day tank,
- 2) Verifying the fuel level in the fuel storage tank,
- 3) Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day tank,
- 4) Verifying the lubricating oil inventory in storage,
- 5) Verifying the diesel starts from standby conditions and achieves generator voltage and frequency at 4160 d 420 volts and 60 4 0.8 Hz. The diesel generator shall be started for this test by using one of the following signals:
a) Manual, or
- All planned starts for the purpose ofthese surveillances may be preceded by an engine prelube period.
MILLSTONE - UNIT 3 3/4 8-3a Amendment No. 4,64, 4142, 194, 210,
5.0 DESIGN FEATURES 5.1 SITE LOCATION The Unit 3 Containment Building is located on the site at Millstone Point in Waterford, Connecticut. The nearest SITE BOUNDARY on land is 1719 feet northeast of the containment I building wall (1627 feet northeast of the elevated stack), which is the minimum distance to the boundary of the exclusion area as described in 10 CFR 100.3. No part of the site that is closer than these distances shall be sold or leased except to Dominion Nuclear Connecticut, Inc. or its corporate affiliates for use in conjunction with normal utility operations. 5.2 DELETED MILLSTONE - UNIT 3 5-1 Amendment 24-2,
ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 6.9.1.5 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator Region I, and one copy to the NRC Resident Inspector, no later than the 15th of each month following the calendar month covered by the report. CORE OPERATING LIMITS REPORT 6.9.1.6 a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:
- 1. Overtemperature AT and Overpower AT setpoint parameters for Specification 2.2.1,
- 2. SHUTDOWN MARGIN for Specifications 3/4.1.1.1.1, 3/4.1.1.1.2, and 3/4.1.1.2, I
- 3. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3.
MILLSTONE - UNIT 3 6-19a Amendment No. 24, 34, 69, 6, 48, 28,
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)
- 4. Shutdown Rod Insertion Limit for Specification 3/4.1.3.5,
- 5. Control Rod Insertion Limits for Specification 3/4.1.3.6,
- 6. AXIAL FLUX DIFFERENCE Limits, target band, and APLND for Specification 3/4.2.1.1,
- 7. Heat Flux Hot Channel Factor, K(z), W(z), APLND, and W(Z)BL for Specification 3/4.2.2.1.
- 8. Nuclear Enthalpy Rise Hot Channel Factor, Power Factor Multiplier for Specification 3/4.2.3.
- 9. DNB Parameters for Specification 3/4.2.5.
- 10. Shutdown Margin Monitor minimum count rate for Specification 3/4.3.5.
- 11. Boron Concentration for Specification 3/4.9.1.1.
6.9.1.6.b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in: I. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," OMProprietary). (Methodology for Specifications 3.1.1.3--Moderator Temperature Coefficient, 3.1.3.5--Shutdown Bank Insertion Limit, 3.1.3.6--Control Bank Insertion Limits, 3.2.1--AXIAL FLUX DIFFERENCE, 3.2.2-Heat Flux Hot Channel Factor, 3.2.3--Nuclear Enthalpy Rise Hot Channel Factor,3.1.1.1.1, 3.1.1.1.2,3.1.1.2 -- SHUTDOWN MARGIN,3.9.1.1 -- Boron Concentration.)
- 2. T. M. Anderson to K. Kniel (Chief of Core Performance Branch, NRC), January 31, 1980--
Attachment:
Operation and Safety-Analysis Aspects of an Improved Load Follow Package.
- 3. NUREG-800, Standard Review Plan, U.S. Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981 Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial 0 set Control (CAOC), Revision 2, July 1981.
- 4. WCAP-10216-P-A-RIA, "RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION,"
(W Prorietg). (Methodology for Specifications 3.2.1--AXIAL FLUX DIFFERENC [Relaxed Axial Offset Control] and 3.2.2--Heat Flux Hot Channel Factor [W(z) surveillance requirements for FQ Methodology].)
- 5. WCAP-9561-P-A, ADD. 3, "BARTA-1: A COMPUTER CODE FOR THE BEST ESTIMATE ANALYSIS OF REFLOOD TRANSIENTS--SPECIAL REPORT:
THIMBLE MODELING W ECCS EVALUATION MODEL," (I Proprietary). (Methodology for Specification 3.2.2--Heat Flux Hot Channel Factor.)
- 6. WCAP-10266-P-A, Addendum 1, "THE 1981 VERSION OF THE WESTINGHOUSE ECCS EVALUATION MODEL USING THE BASH CODE,"
fWProprietary). (Methodology for Specification 3.2.2--Heat Flux Hot Channel Pa~ctor.) MILLSTONE - UNIT 3 6-20 Amendment No. 24,3-7, 60,69, 9+, 20, 470, 214-,
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)
- 7. WCAP-1 1946, "Safety Evaluation Supporting a More Negative EOL Moderator Temperature Coefficient Technical Specification for the Millstone Nuclear Power Station Unit 3," (! Proprietary).
- 8. WCAP-10054-P-A, "WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL. 17 USING THE NOTRUMP CODE," (W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
- 9. WCAP-10079-P-A, "NOTRUMP - A NODAL TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE," (W Proprietary). (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
- 10. WCAP-12610, "VANTAGE+ Fuel Assembly Report," M Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
- 11. Letter from V. L. Rooney (USNRC) to J. F. Opeka, "Safety Evaluation for Topical Report, NUSCO-I 52, Addendum 4, 'Physics Methodology for PWR Reload Design,' TAC No. M91815," July 18, 1995.
- 12. Letter from E. J. Mroczka to the USNRC, "Proposed Changes to Technical Specifications, Cycle 4 Reload Submittal - Boron Dilution Analysis," B 13678, December 4, 1990.
- 13. Letter from D. H. Jaffe (USNRC) to E. J. Mroczka, "Issuance of Amendment (TAC No. 77924)," March 11, 1991.
- 14. Letter from M. H. Brothers to the USNRC, "Proposed Revision to Technical Specification, SHUTDOWN MARGIN Requirements and Shutdown Margin Monitor OPERABILITY for MODES 3,4, and 5 (PTSCR 3-16-97), B 16447, May 9, 1997.
- 15. Letter from J. W. Anderson (USNRC) to M. L. Bowling (NNECO), "Issuance of Amendment - Millstone Nuclear Power Station, Unit No. 3 (TAC No. M98699),"
October 21, 1998.
- 16. WCAP-8301, "LOCTA-IV Program: Loss-of-Coolant Transient Analysis."
- 17. WCAP-10054-P-A, Addendum 2, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model."
- 18. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," (Westinghouse Proprietary Class 2).
(Methodology for Specification 2.2.1.) MILLSTONE - UNIT 3 6-20a Amendment No. 8+, 4170, M,
ADMINISTRATIVE CONTROLS 6.9.1.6.c The core operating limits shall be determined so that all applicable limits (e.g. fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met. 6.9.1.6.d The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector. SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator Region I, and one copy to the NRC Resident Inspector, within the time period specified for each report. 6.10 Deleted. 6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CER Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure. 6.12 HIGH RADIATION AREA As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601 (a) and (b) of 10 CFR Part 20: 6.12.1 High Radiation Areas with Dose Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation
- a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.
- b. Access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent; that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.
- c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.
- d. Each individual or group entering such an area shall possess:
- 1. A radiation monitoring device that continuously displays radiation dose rates in the area, or MILLSTONE - UNIT 3 6-21 Amendment No. 24, 40, , 69, 04, 4173, 242, 24S,
ADMINISTRATIVE CONTROLS 6.15 RADIOACTIVE EFFLUENT CONTROLS PROGRAM This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the REMODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
- a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the REMODCM;
- b. Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS, conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402;
- c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the REMODCM;
- d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS, conforming to 10 CFR 50, Appendix I;
- e. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the REMODCM at least every 31 days.
Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the REMODCM at least every 31 days;
- f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;
- g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY shall be in accordance with the following:
- 1. For noble gases: a dose rate
- 500 mremlyr to the whole body and a dose rate < 3000 mrem/yr to the skin, and
- 2. For iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: a dose rate
- 1500 mrem/yr to any organ;
- h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY, conforming to 10 CFR 50, Appendix I;
- i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY, conforming to 10 CFR 50, Appendix I; and MILLSTONE - UNIT 3 6-25 Amendment No. -- 9, 24-S,
ADMINISTRATIVE CONTROLS
- j. Limitations on the annual dose or dose commitment to any member of the public, beyond the SITE BOUNDARY, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.
The provisions of Specification 4.0.2 and Specification 4.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency. 6.16 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the REMODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:
- a. Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the REMODCM.
- b. A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and
- c. Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.
6.17 REACTOR COOLANT PUMP FLYWHEEL INSPECTION PROGRAM This program shall provide for the inspection of each reactor coolant pump flywheel by either qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle of one-half the outer radius or a surface examination (magnetic particle testing and/or penetrant testing) of exposed surfaces defined by the volume of the disassembled flywheels at least once every 10 years. 6.18 TECHNICAL SPECIFICATIONS (TS) BASES CONTROL PROGRAM This program provides a means for processing changes to the Bases of these Technical Specifications:
- a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
- b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
- 1. A change in the TS incorporated in the license or
- 2. A change to updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
MILLSTONE - UNIT 3 6-26 Amendment No. 4-8, 204, 214-2, 2-,
APPENDIX B TO FACILITY OPERATING LICENSE NO. NPF-49 MILLSTONE POWER STATION, UNIT 3 DOMINION NUCLEAR CONNECTICUT, INC. DOCKET NO. 50-423 ENVIRONMENTAL PROTECTION PLAN (NONRADIOLOGICAL) Amendment No.
MILLSTONE POWER STATION, UNIT NO. 3 ENVIRONMENTAL PROTECTION PLAN (NONRADIOLOGICAL) TABLE OF CONTENTS Section Page 1.0 Objectives of the Environmental Protection Plan .................................................... 1-1 2.0 Environmental Protection Issues .................................................... 2-1 3.0 Consistency Requirements .................................................... 3-1 3.1 Plant Design and Operation .................................................... 3-2 3.2 Reporting Related to the NPDES Permit and State Certifications ............... ............ 3-3 3.3 Changes Required for Compliance with Other Environmental Regulations ............ 3-3 4.0 Environmental Conditions .................................................... 4-1 4.1 Unusual or Important Environmental Events .................................................... 4-1 4.2 Environmental Monitoring .................................................... 4-1 5.0 Administrative Procedures .................................................... 5-1 5.1 Review .................................................... 5-1 5.2 Records Retention .................................................... 5-1 5.3 Changes in Environmental Protection Plan ....................................... ............. 5-2 5.4 Plant Reporting Requirements .................................................... 5-2 MILLSTONE - UNIT 3 Amendment No.
2.0 Environmental Protection Issues In the FES-OL dated December, 1984, the staff considered the environmental impacts associated with the operation of Millstone Power Station, Unit No. 3. No environmental issues were I identified which required study or license conditions to resolve environmental concerns and to assure adequate protection of the environment. MILLSTONE - UNIT 3 2-1 Amendment No.
written evaluation of such activity and obtain priorNRC approval. When such activity involves a change in the EPP, such activity and change to the EPP may be implemented only in accordance with an appropriate license amendment as set forth in Section 5.3 of this EPP. A proposed change, test or experiment shall be deemed to involve an unreviewed environmental question if it concerns: (I) a matter which may result in a significant increase in any adverse environmental impact previously evaluated in the FES-OL, environmental impact appraisals, or in any decisions of the Atomic Safety and Licensing Board; or (2) a significant change in effluents or power level; or (3) a matter, not previously reviewed and evaluated in the documents specified in (1) of this Subsection, which may have a significant adverse environmental impact. The licensee shall maintain records of changes in facility design or operation and of tests and experiments carried out pursuant to this Subsection. These records shall include written evaluations which provide bases for the determination that the change, test, or experiment does not involve an unreviewed environmental question or constitute a decrease in the effectiveness of this EPP to meet the objectives specified in Section 1.0. The licensee shall include as part of the Annual Environmental Operating Report (per Subsection 5.4.1) brief descriptions, analyses, interpretations, and evaluations of such changes, tests and experiments. MILLSTONE - UNIT 3 3-2
5.0 Administrative Procedures 5.1 Review The licensee shall provide for review of compliance with the EPP. The reviews shall be conducted independently of the individual or groups responsible for performing the specific activity. A description of the organization structure utilized to achieve the independent review function and results of the review activities shall be maintained and made available for inspection. 5.2 Records Retention Records and logs relative to the environmental aspects of station operation shall be made and retained in a manner convenient for review and inspection. These records and logs shall be made available to NRC on request. Records of modifications to station structures, systems and components determined to potentially affect the continued protection of the environment shall be retained for the life of the station. All other records, data and logs relating to this EPP shall be retained for five years or, where applicable, in accordance with the requirements of oiher agencies. MILLSTONE - UNIT 3 5-1
5.3 Changes in Environmental Protection Plan Requests for changes in the EPP shall include an assessment of the environmental impact of the proposed change and a supporting justification. Implementation of such changes in the EPP shall not commence prior to NRC approval of the proposed changes in the form of a license amendment incorporating the appropriate revision to the EPP. 5.4 Plant Reporting Requirements 5.4.1 Routine Reports An Annual Environmental Operating Report describing implementation of this EPP for the previous year shall be submitted to the NRC on or before May I of each year. The initial report shall be submitted on or before May I of the year following start of commercial operation of the plant. The report shall include summaries and analyses of the results of the environmental protection activities required by Subsection 4.2 of this EPP for the report period, including a comparison with related preoperational studies, operational controls (as appropriate) and previous nonradiological environmental monitoring reports, and an assessment of the observed impacts of the plant operation on the environment. If harmful effects or evidence of trends toward MILLSTONE - UNIT 3 5-2
irreversible damage to the environment are observed, the licensee shall provide a detailed analysis of the data and a proposed course of mitigating action. The Annual Environmental Operating Report shall also include: (1) A list of EPP noncompliances and the corrective actions taken to remedy them. (2) A list of all changes in station design or operation, tests, and experiments made in accordance with Subsection 3.1 which involved a potentially significant unreviewed environmental question. (3) A list of nonroutine reports submitted in accordance with Subsection 5.4.2. In the event that some results are not available by the report due date, the report shall be submitted noting and explaining the missing results. The missing results shall be submitted as soon as possible in a supplementary report. 5.4.2 Nonroutine Reports A written report shall be submitted to the NRC within 30 days of occurrence of a nonroutine event. The report shall (a) describe, analyze, and evaluate the event, including extent and magnitude of the impact, and plant operating MILLSTONE - UNIT 3 5-3
characteristics, (b) describe the probable cause of the event, (c) indicate the action taken to correct the reported event, (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (e) indicate the agencies notified and their preliminary responses. I Events reportable under this subsection which also require reports to other Federal, State or local agencies shall be reported in accordance with those reporting requirements in lieu of the requirements of this Subsection. The NRC shall be provided with a copy of such report at the same time it is submitted to the other agency. MILLSTONE - UNIT 3 5-4}}