ML14023A748
| ML14023A748 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/25/2014 |
| From: | James Kim Plant Licensing Branch 1 |
| To: | Heacock D Dominion Nuclear Connecticut |
| Kim J | |
| References | |
| TAC ME9733 | |
| Download: ML14023A748 (126) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 25, 2014 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
MILLSTONE POWER STATION, UNIT NO.3- ISSUANCE OF AMENDMENT RE: RISK-INFORMED JUSTIFICATION FOR THE RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCY REQUIREMENTS TO A LICENSEE CONTROLLED PROGRAM (ADOPTION OF TSTF-425, REVISION 3)
Dear Mr. Heacock:
The Commission has issued the enclosed Amendment No. 258 to Renewed Facility Operating License No. NPF-49 for the Millstone Power Station, Unit No. 3, in response to your application dated October 4, 2012, as supplemented by letters dated January 4, 2013, April 17, 2013, and October 30, 2013.
The amendment modifies the Technical Specifications (TSs) by relocating specific surveillance frequencies to a licensee controlled program with the adoption of Technical Specification Task Force (TSTF)-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control -
[Risk-Informed Technical Specification Task Force (RITSTF)] Initiative 5b." Additionally, the amendment adds a new program, the Surveillance Frequency Control Program, toTS Section 6, Administrative Controls.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely,
~~
James Kim, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-423
Enclosures:
- 1. Amendment No. 258 to NPF-49
- 2. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DOMINION NUCLEAR CONNECTICUT. INC.
DOCKET NO. 50-423 MILLSTONE POWER STATION, UNIT NO. 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 258 Renewed License No. NPF-49
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by the applicant dated October 4, 2012, as supplemented by letters dated January 4, 2013, April 17, 2013, and October 30, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-49 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 258, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the renewed license. DNC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3. This license amendment is effective as of the date of issuance, and shall be implemented within 90 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
~~~
Benjamin G. Beasley, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the License and Technical Specifications Date of Issuance: February 25, 2014
ATTACHMENT TO LICENSE AMENDMENT NO. 258 RENEWED FACILITY OPERATING LICENSE NO. NPF-49 DOCKET NO. 50-423 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Insert Page 4 Page 4 Replace the following pages of the Appendix A Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 1-8 1-8 3/4 1-1 3/4 1-1 3/4 1-2 3/4 1-2 3/4 1-3 3/4 1-3 3/4 1-8 3/4 1-8 3/4 1-8a 3/4 1-8a 3/4 1-21 3/4 1-21 3/4 1-23 3/4 1-23 3/4 1-25 3/4 1-25 3/4 1-26 3/4 1-26 3/4 1-27 3/4 1-27 3/4 2-2 3/4 2-2 3/4 2-8 3/4 2-8 3/4 2-10 3/4 2-10 3/4 2-20 3/4 2-20 3/4 2-21 3/4 2-21 3/4 2-26 3/4 2-26 3/4 2-27 3/4 2-27 3/4 3-1 3/4 3-1 3/4 3-10 3/4 3-10 3/4 3-11 3/4 3-11 3/4 3-12 3/4 3-12 3/4 3-13 3/4 3-13 3/4 3-16 3/4 3-16 3/4 3-36 3/4 3-36 3/4 3-37 3/4 3-37 3/4 3-38 3/4 3-38 3/4 3-39 3/4 3-39 3/4 3-40 3/4 3-40 3/4 3-41 3/4 3-41
Remove Insert 3/4 3-42 3/4 3-42 3/4 3-45 3/4 3-45 3/4 3-53 3/4 3-53 3/4 3-58 3/4 3-58 3/4 3-59a 3/4 3-59a 3/4 3-62 3/4 3-62 3/4 3-63 3/4 3-63 3/4 3-83 3/4 3-83 3/4 4-1 3/4 4-1 3/4 4-2a 3/4 4-2a 3/4 4-4 3/4 4-4 3/4 4-5a 3/4 4-5a 3/4 4-6a 3/4 4-6a 3/4 4-7 3/4 4-7 3/4 4-11 3/44-11 3/4 4-11 b 3/4 4-11 b 3/4 4-13 3/4 4-13 3/4 4-21a 3/4 4-21 a 3/4 4-23 3/4 4-23 3/4 4-23a 3/4 4-23a 3/4 4-29 3/4 4-29 3/4 4-39 3/4 4-39 3/4 5-1 3/4 5-1 3/4 5-2 3/4 5-2 3/4 5-4 3/4 5-4 3/4 5-5 3/4 5-5 3/4 5-9 3/4 5-9 3/4 5-10 3/4 5-10 3/4 6-1 3/4 6-1 3/4 6-6 3/4 6-6 3/4 6-7 3/4 6-7 3/4 6-9 3/4 6-9 3/4 6-11 3/4 6-11 3/4 6-12 3/4 6-12 3/4 6-13 3/4 6-13 3/4 6-15 3/4 6-15 3/4 6-18 3/4 6-18 3/4 6-19 3/4 6-19 3/4 6-20 3/4 6-20 3/4 6-22 3/4 6-22 3/4 7-5 3/4 7-5 3/4 7-5a 3/4 7-5a 3/4 7-6 3/4 7-6 3/4 7-8 3/4 7-8 3/4 7-9a 3/4 7-9a 3/4 7-11 3/4 7-11
Remove Insert 3/4 7-12 3/4 7-12 3/4 7-13 3/4 7-13 3/4 7-16 3/4 7-16 3/4 7-17 3/4 7-17 3/4 7-20 3/4 7-20 3/4 7-21 3/4 7-21 3/4 8-3a 3/4 8-3a 3/4 8-4 3/4 8-4 3/4 8-5 3/4 8-5 3/4 8-7 3/4 8-7 3/4 8-8 3/4 8-8 3/4 8-11 3/4 8-11 3/4 8-12 3/4 8-12 3/4 8-17 3/4 8-17 3/4 8-18a 3/4 8-18a 3/4 9-1 3/4 9-1 3/4 9-1a 3/4 9-1 a 3/4 9-2 3/4 9-1 a 3/4 9-4 3/4 9-2 3/4 9-8 3/4 9-4 3/4 9-9 3/4 9-8 3/4 9-11 3/4 9-9 3/4 9-12 3/4 9-11 3/4 10-2 3/4 9-12 3/4 10-4 3/4 10-2 3/4 10-5 3/4 10-4 6-17g 3/4 10-5 6-17g
(2) Technical Specifications The Technical Specifications contained in Appendix A, revised through Amendment No. 258 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated into the license. DNC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3) DNC shall not take any action that would cause Dominion Resources, Inc.
(DRI) or its parent companies to void, cancel, or diminish DNC's Commitment to have sufficient funds available to fund an extended plant shutdown as represented in the application for approval of the transfer of the licenses for MPS Unit No. 3.
(4) Immediately after the transfer of interests in MPS Unit No. 3 to DNC, the amount in the decommissioning trust fund for MPS Unit No. 3 must, with respect to the interest in MPS Unit No. 3, that DNC would then hold, be at a level no less than the formula amount under 10 CFR 50.75.
(5) The decommissioning trust agreement for MPS Unit No. 3 at the time the transfer of the unit to DNC is effected and thereafter is subject to the following:
(a) The decommissioning trust agreement must be in a form acceptable to the NRC.
(b) With respect to the decommissioning trust fund, investments in the securities or other obligations of Dominion Resources, Inc. or its affiliates or subsidiaries, successors, or assigns are prohibited.
Except for investments tied to market indexes or other non-nuclear-sector mutual funds, investments in any entity owning one or more nuclear power plants are prohibited.
(c) The decommissioning trust agreement for MPS Unit No. 3 must provide that no disbursements or payments from the trust, other than for ordinary administrative expenses, shall be made by the trustee until the trustee has first given the Director of the Office of Nuclear Reactor Regulation 30 days prior written notice of payment. The decommissioning trust agreement shall further contain a provision that no disbursements or payments from the trust shall be made if the trustee receives prior written notice of objection from the NRC.
(d) The decommissioning trust agreement must provide that the agreement can not be amended in any material respect without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.
Renewed License No. NPF-49 Amendment No. 258
TABLE 1.1 FREQUENCY NOTATION NOTATION FREQUENCY s At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
w At least once per 7 days.
M At least once per 31 days.
Q At least once per 92 days.
SA At least once per 184 days.
R At least once per 18 months.
S/U Prior to each reactor startup.
N.A. Not applicable.
p Completed prior to each release.
SFCP At the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3 1-8 Amendment No. 258
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 1 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 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 untrippab1e control rod(s);
- b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1 at the frequency specified in the Surveillance Frequency Control Program 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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 3.1 0.1.
MILLSTONE - UNIT 3 3/4 1-1 Amendment No. 6B, +H, ~. R-8-,
ti9 258
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within +/- 1% ~k at the frequency specified in the Surveillance Frequency Control Program. This comparison shall consider at least the following factors:
- 1) Reactor Coolant System boron concentration,
- 2) Control rod position,
- 3) Reactor Coolant System average temperature,
- 4) Fuel bumup based on gross thermal energy generation,
- 5) Xenon concentration, and
- 6) Samarium concentration.
The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel bumup of 60 EFPD after each fuel loading.
MILLSTONE - UNIT 3 3/4 1-2 Amendment No. W 258
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).*
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 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 the frequency specified in the Surveillance Frequency Control Program by consideration of the following factors:
- 1. Reactor Coolant System boron concentration,
- 2. Control rod position,
- 3. Reactor Coolant System average temperature,
- 4. Fuel bumup 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 the frequency specified in the Surveillance Frequency Control Program.
- Additional SHUTDOWN MARGIN requirements, if required, are given in Specification 3.3.5.
MILLSTONE - UNIT 3 3/4 1-3 Amendment No. W, H, -l-64, :t.-1-1, US, ti9 258
REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - COLD SHUTDOWN -LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to
- a. the limits specified in the CORE OPERATING LIMITS REPORT (COLR) for MODE 5 with RCS loops not filled* or
- b. the limits specified in the COLR for MODE 5 with RCS loops filled* with the chemical and volume control system (CVCS) aligned to preclude reactor coolant system boron concentration reduction.
APPLICABILITY: MODE 5 LOOPS NOT FILLED ACTION:
- a. With the SHUTDOWN MARGIN less than the above, 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.
- b. With the CVCS dilution flow paths not closed and secured in position in accordance with Specification 3 .1.1.2(b), immediately close and secure the paths or meet the limits specified in the COLR for MODE 5 with RCS loops not filled.
SURVEILLANCE REQUIREMENTS 4.1.1.2.1 The SHUTDOWN MARGIN shall be determined to be within the limits specified in the COLR:
- a. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the 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 the frequency specified in the Surveillance Frequency Control Program 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,
- Additional SHUTDOWN MARGIN requirements, if required, are given in Specification 3.3.5.
MILLSTONE- UNIT 3 3/4 1-8 Amendment No. W, 99, i--l-3-, +64, ~
258
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 5) Xenon concentration, and
- 6) Samarium concentration.
4.1.1.2.2 At the frequency specified in the Surveillance Frequency Control Program the following valves shall be verified closed and locked. The valves may be opened on an intermittent basis under administrative controls except as noted.
Valve Number Valve Function Valve Position
- 1. V304(Z-) Primary Grade Water Closed to eves
- 2. V120(Z-) Moderating Hx Outlet Closed
- 3. V147(Z-) BTRS Outlet Closed
- 4. V797(Z-) Failed Fuel Monitoring Closed Flushing
- 5. V100(Z-) Resin Sluice, CVCS Cation Closed Bed Demineralizer
- 6. V571(Z-) Resin Sluice, CVCS Cation Closed Bed Demineralizer
- 7. Vl11(Z-) Resin Sluice, CVCS Cation Closed Bed Demineralizer
- 8. Vl12(Z-) Resin Sluice, CVCS Cation Closed Bed Demineralizer
- 9. V98(Z-)N99(Z-) Resin Sluice, CVCS Mixed Closed Bed Demineralizer
- 10. V569(Z-)N570(Z-) Resin Sluice, CVCS Mixed Closed Bed Demineralizer
- 11. V107(Z-)/V109(Z-) Resin Sluice, CVCS Mixed Closed Bed Demineralizer
- 12. V108(Z-)N110(Z-) Resin Sluice, CVCS Mixed Closed Bed Demineralizer
- 13. V305(Z-)* Primary Grade Water Closed to Charging Pumps
- This valve may not be opened under administrative controls.
MILLSTONE - UNIT 3 3/4 1-8a Amendment No. 99-;-
258
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION ACTION: (Continued)
- c. A power distribution map is obtained from the movable incore detectors and F Q(Z) and F N are verified to be within their limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; and ~H
- d. THERMAL POWER level is reduced to less than or equal to 75%
of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER.
- c. With more than one rod trippable but inoperable due to causes other than addressed by ACTION a. above, POWER OPERATION may continue provided that:
- 1. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the remainder of the rods in the bank(s) with the inoperable rods are aligned to within +/- 12 steps of the inoperable rods while maintaining the rod sequence and insertion limits of Specification 3.1.3.6.
The THERMAL POWER level shall be restricted pursuant to Specification 3 .1.3 .6 during subsequent operation, and
- 2. The inoperable rods are restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- d. With more than one rod misaligned from its group step counter demand height by more than +/-12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifYing the individual rod positions at the frequency specified in the Surveillance Frequency Control Program except during time intervals when the rod position deviation monitor is inoperable, then verifY the group positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE- UNIT 3 3/4 1-21 Amendment No. §G, 6B, +9+, ~
258
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 1 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 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 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 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- b. With a maximum of one demand position indicator per bank inoperable:
- 1. 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 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
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 the frequency specified in the Surveillance Frequency Control Program 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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
4.1.3.2.2 Each ofthe 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 the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE- UNIT 3 3/4 1-23 Amendment No. W, W, ~. ~. ~
258
REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full-length (shutdown and control) rod drop time from the fully withdrawn position shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:
- a. Tavg greater than or equal to 500°F, and
- b. All reactor coolant pumps operating.
APPLICABILITY: MODES 1 and 2.
ACTION:
- a. With the drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
- b. With the rod drop times within limits but determined with three reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 65% of RATED THERMAL POWER with the reactor coolant stop valves in the nonoperating loop closed.
SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality:
- a. For all rods following each removal of the reactor vessel head, and
- b. Deleted
- c. At the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE - UNIT 3 3/4 1-25 Amendment No. W, ~' M6, ~'
258 ti-7
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).
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 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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 the frequency specified in the Surveillance Frequency Control Program.
- 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. W, ~
258
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* **.
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 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or
- b. Reduce THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3 .6 The position of each control bank shall be determined to be within the insertion limits at the frequency specified in the Surveillance Frequency Control Program except during time intervals when the rod insertion limit monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
- With Kerr greater than or equal to 1.
MILLSTONE - UNIT 3 3/4 1-27 Amendment No. §G, W, ~
Reissued by NRC Letter dated September 27, 2006 258
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.1.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by:
- a. Monitoring the indicated AFD for each OPERABLE excore channel at the frequency specified in the Surveillance Frequency Control Program when the AFD Monitor Alarm is OPERABLE:
- b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values ofthe indicated AFD shall be assumed to exist during the interval preceding each logging.
4.2.1.1.2 The indicated AFD shall be considered outside of its limits when two or more OPERABLE excore channels are indicating the AFD to be outside the limits.
4.2.1.1.3 When in base load operation, the target flux difference of each OPERABLE excore channel shall be determined by measurement at the frequency specified in the Surveillance Frequency Control Program. The provisions of Specification 4.0.4 are not applicable.
4.2.1.1.4 When in base load operation, the target flux difference shall be updated at the frequency specified in the Surveillance Frequency Control Program by either determining the target flux difference in conjunction with the surveillance requirements of Specification 4.2.1.1.3 or by linear interpolation between the most recently measured value and the calculated value at the end of cycle life. The provisions of Specification 4.0.4 are not applicable.
MILLSTONE- UNIT 3 3/4 2-2 Amendment No. ~' W 258
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
- c. Satisfying the following relationship:
M F~TP X K(Z)
F 0 (Z):S; PxW(Z) for P>0.5 M F~TP X K(Z)
F 0 (Z):::;; W(Z) x _ for P:::;; 0.5 05 where F 0 M(z) is the measured F o(Z) increased by the allowances for manufacturing tolerances and measurement uncertainty, FQ RTP is the F Q limit, K(Z) is the normalized F 0 (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. F0 RTP, 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 Fo(Z) was last determined,*** or (2) At the frequency specified in the Surveillance Frequency Control Program, whichever occurs first.
- e. With the maximum value of F~(Z)
K(Z) over the core height (Z) increasing since the previous determination ofF 0 M(z),
either ofthe following ACTIONS shall be taken:
(1) Increase F0 M(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. §.G, W, 99, ~' -19, 258 ti9
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
- b. Duri~ base load operation, if the THERMAL POWER is decreased below APL 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 ifFQ(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 F QM(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:
M FSTP X K(Z) NO FQ (Z):::; p X W(Z)BL for P > APL 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, Pis the relative THERMAL POWER, and W(Z)sL is the cycle-dependent function that accounts for limited power distribution transients encountered during base load operation.
FQRTP, K(Z), and W(Z)sL 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 APLNO for the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to mapping, and (2) At the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE- UNIT 3 3/4 2-10 Amendment No. §.9, W, 99, HG, ~
258
POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION: (Continued)
- b. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate that the RCS total flow rate is N
restored to within the limits specified above and in the COLR and F ~H is restored to within the above limit, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
- c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2. and/or b., above; subsequent N
POWER OPERATION may proceed provided that F ~Hand indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable operation prior to exceeding the following THERMAL POWER levels:
- 1. A nominal 50% of RATED THERMAL POWER,
- 2. A nominal 75% of RATED THERMAL POWER, and
- 3. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of attaining greater than or equal to 95% of RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.2.3.1.1 The provisions of Specification 4.0.4 are not applicable.
N 4.2.3.1.2 F ~H shall be determined to be within the acceptable range:
- a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and
- b. At the frequency specified in the Surveillance Frequency Control Program.
4.2.3.1.3 The RCS total flow rate shall be determined to be within the acceptable range by:
- a. Verifying by precision heat balance that the RCS total flow rate is
~ 363,200 gpm and greater than or equal to the limit specified in the COLR within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 90% of RATED THERMAL POWER after each fuel loading, and MILLSTONE - UNIT 3 3/4 2-20 Amendment No. @, 19, -1:{}(}, ~.
~
258
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
- b. Verifying that the RCS total flow rate is :2:: 363,200 gpm and greater than or equal to the limit specified in the COLR at the frequency specified in the Surveillance Frequency Control Program.
4.2.3.1.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at the frequency specified in the Surveillance Frequency Control Program.
4.2.3.1.5 DELETED.
4.2.3.1.6 DELETED.
MILLSTONE- UNIT 3 3/4 2-21 Amendment No. ti, @, ~'
258
POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued)
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; and
- 3. Identify and 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 QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or until verified at 95% or greater RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% ofRATED THERMAL POWER by:
- a. Calculating the ratio at the frequency specified in the Surveillance Frequency Control Program when the alarm is OPERABLE, and
- b. Calculating the ratio at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during steady-state operation when the alarm is inoperable.
4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm that the normalized symmetric power distribution, obtained from two sets of four symmetric thimble locations or full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE- UNIT 3 3/4 2-26 Amendment No. §'1-, W, 258
POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the limits specified in the CORE OPERATING LIMITS REPORT (COLR):
- a. Reactor Coolant System Tavg* and
- b. Pressurizer Pressure.
APPLICABILITY: MODE 1.
ACTION:
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.5 Each of the above DNB-related parameters shall be verified to be within the limits specified in the COLR at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE- UNIT 3 3/4 2-27 Amendment No. §=!-, W, U8, 258
3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks 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 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.
4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be verified to be within its limit at the frequency specified in the Surveillance Frequency Control Program. Neutron detectors and speed sensors are exempt from response time verification. Each verification shall include at least one train and one channel (to include input relays to both trains) per function.
MILLSTONE- UNIT 3 3/4 3-1 Amendment No. 4§., '1-9, 9+, -l-00, -l--&1, 258
TABLE 4.3-1
~ REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS r
r TRIP
\/l ANALOG ACTUATING MODES FOR 0
ztTJ CHANNEL DEVICE WHICH I
CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE
~ FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED
~ 1. Manual Reactor Trip N.A. N.A. N.A. SFCP(14) N.A. 1, 2, 3*, 4*, 5*
w
- 2. Power Range, Neutron Flux
SFCP(4, 6),
w
+:>.
w SFCP(4, 5)
_. I
- 4. Deleted
- 5. Intermediate Range SFCP SFCP(4, 5) SIU(l) N.A. N.A. 1***,2 N ~ 6. Source Range, Neutron Flux SFCP SFCP(4, 5) SIU(l), N.A. N.A. 2**,3*,4*,5*
U1 ~
co ::s SFCP(9)
Cl.
~ 7. Overtemperature ~ T SFCP SFCP SFCP N.A. N.A. 1, 2 g
z0
- 8. Overpower ~T SFCP SFCP SFCP N.A. N.A. 1, 2 jt 9. Pressurizer Pressure--Low SFCP SFCP SFCP(18) N.A. N.A. 1*****
J:~ 10. Pressurizer Pressure--High SFCP SFCP SFCP(18) N.A. N.A. 1, 2 J3 11. Pressurizer Water Level--High SFCP SFCP SFCP N.A. N.A. 1*****
~t j 12. Reactor Coolant Flow--Low SFCP SFCP SFCP N.A. N.A. 1
~~~$
TABLE 4.3-1 (Continued}
REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS
~
....... TRIP r ANALOG ACTUATING MODES FOR r
\/1 CHANNEL DEVICE WHICH 0 CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE
~ FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED 2 13.
I Steam Generator Water SFCP SFCP SFCP(18) N.A. N.A. 1, 2
...., Level--Low-Low w 14. Low Shaft Speed- N.A. SFCP(13) SFCP N.A. N.A. 1 Reactor Coolant Pumps
- 15. Turbine Trip
- a. Low Fluid Oil N.A. SFCP N.A. SIU(l, 10)**** N.A. 1 w Pressure
~
w b. Turbine Stop Valve N.A. SFCP N.A. SIU(1, 10)**** N.A. 1
- I Closure
- 16. Deleted
- 17. Reactor Trip System Interlocks
Neutron Flux, P-6
"'> b. Low Power Reactor N.A. SFCP(4) SFCP N.A. N.A. 1
~3(1) Trips Block, P-7 0.. c. Power Range Neutron N.A. SFCP(4) SFCP N.A. N.A. 1 3
(1)
Flux, P-8 a
z 0
~~ e. Power Range Neutron N.A. SFCP(4) SFCP N.A. 1, 2 N.A.
J~ Flux, P-10
~~ f. Turbine Impulse N.A. SFCP SFCP N.A. N.A. 1 Chamber Pressure,
~~ P-13
~
r TABLE 4.3-1 (Continued) r r./'1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS
~tTl TRIP I
ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH
~
........ CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE
- j FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED
- 18. Reactor Trip Breaker N.A. N.A. N.A. SFCP(7,11) N.A. 1, 2, 3*,
4*,5*
- 19. Automatic Trip and N.A. N.A. N.A. N.A. SFCP(7) 1, 2, 3*,
~ Interlock Logic 4*,5*
.+;:..
-'-f N
- 20. DELETED
- 21. Reactor Trip Bypass N.A. N.A. N.A. SFCP(7,15) N.A. 1, 2, 3*,
Breaker SFCP(16) 4*,5*
- 22. DELETED N>
U1 3 OJ (D
- s a
-z (D
- s 0
~t
~~
1E
~~
~t ~~
TABLE 4.3-1 (Continued)
TABLE NOTATIONS
- When the Reactor Trip System breakers are closed and the Control Rod Drive System is capable of rod withdrawal.
- Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
- Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
- Above the P-9 (Reactor Trip/Turbine Interlock) Setpoint.
- Above the P-7 (At Power) Setpoint (1) If not performed in previous 31 days.
(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Specification 4.0.4 are not applicable to entry into MODE 2 or 1.
(3) Single point comparison ofincore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(5) Detector plateau curves shall be obtained, and evaluated and compared to manufacturer's data. For the Source Range, Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(6) Incore- Excore Calibration, above 75% of RATED THERMAL POWER. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(7) Each train shall be tested at the frequency specified in the Surveillance Frequency Control Program.
(8) (Not used)
(9) Surveillance in MODES 3*, 4*, and 5* shall also include verification that permissives P-6 and P-1 0 are in their required state for existing plant conditions by observation of the permissive annunciator window.
MILLSTONE- UNIT 3 3/4 3-13 Amendment No. W, :m, W-9, tiG, 258
INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance ofthe ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.
4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME* of each ESFAS function shall be verified to be within the limit at the frequency specified in the Surveillance Frequency Control Program. Each verification shall include at least one train and one channel (to include input relays to both trains) per function.
- The provisions of Specification 4.0.4 are not applicable for response time verification of steam line isolation for entry into MODE 4 and MODE 3 and turbine driven auxiliary feedwater pump for entry into MODE 3.
MILLSTONE - UNIT 3 3/4 3-16 Amendment No. 4§., +9, %, +00, -l-8-7, 258
TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION
~
........ SURYEILLANCE REQUIREMENTS r
r TRIP C/1 0 ANALOG ACTUATING MODES ztr1 CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE
§2........ FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
.....J 1. Safety Injection (Reactor Trip, VJ Feedwater Isolation, Control Building Isolation (Manual Initiation Only), Start Diesel Generators, and Service Water)
VJ
+;:.. a. Manual Initiation N.A. N.A. N.A. SFCP N.A. N.A. N.A. 1, 2, 3, 4 VJ VJ I
Logic and Actuation Relays
U1 3 Low co (1)
(1) Low
- 1 0
z 2. Contamment Spray
- a. Manual Initiation N.A. N.A. N.A. SFCP N.A. N.A. N.A. 1, 2, 3, 4
~~
~ b. Automatic Actuation N.A. N.A. N.A. N.A. SFCP(l) SFCP(l) SFCP(4) 1, 2, 3, 4
~
Logic and Actuation J~ Relays
~~ c. Containment SFCP SFCP SFCP N.A. N.A. N.A. N.A. 1, 2, 3, 4
~~
Pressure-High-3
TABLE 4.3-2 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION
~
....... SURVEILLANCE REQUIREMENTS l'
l' TRIP C/J ANALOG ACTUATING MODES
~
m CHANNEL DEVICE MASTERSLAVE FOR WHICH CHANNEL CHANNEL OPERATIONALOPERATIONALACTUATIONRELAY RELAY SURVEILLANCE
~ FUNCTIONAL UNIT CHECK CALIBRATIONTEST TEST LOGIC TESTIEST TEST IS REQUIRED
,_., 3. Containment Isolation (J.)
- a. Phase "A" Isolation
- 1. Manual Initiation N.A. N.A. N.A. SFCP N.A. N.A. N.A. 1, 2, 3, 4
...__ Logic and Actuation
+:>.
(J.)
I Relays (J.)
-.)
- 3. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
- b. Phase "B" Isolation
- 1. Manual Initiation N.A. N.A. N.A. SFCP N.A. N.A. N.A. 1, 2, 3, 4
- 2. Automatic Actuation N.A. N.A. N.A. N.A. SFCP(1) SFCP(1)SFCP(4) 1, 2, 3, 4 Logic and Actuation N> s
()1 co (D Relays
~
s (D
g
~ c. DELETED
~~ 4. Steam Line Isolation
~;:g d. Manual Initiation J~ 1. Individual N.A. N.A. N.A. SFCP N.A. N.A. N.A. 1, 2, 3, 4
~~ ~~ 2. System N.A. N.A. N.A. SFCP N.A. N.A. N.A. 1, 2, 3, 4
~t ~~
TABLE 4.3-2 (Continued)
~
........ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION t'""" SURVEILLANCE REQIIIREMENTS t'"""
r/J 0 TRIP zt'rJ ANALOG CHANNEL ACTUATING DEVICE MASTERSLAVE MODES FOR WHICH CHANNELCHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE
~ FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
- 3 4. Steam Line Isolation (Continued)
VJ
- b. Automatic Actuation N.A. N.A. N.A. N.A. SFCP(1) SFCP(1) SFCP(4) 1, 2, 3, 4 Lofcic and Actuation Re ays
---.J::o.
VJ d. Steam Line Pressure- SFCP SFCP SFCP N.A. N.A. N.A. N.A. 1, 2, 3 VJ I
Low 00
- 5. Turbine Trip and Feedwater Isolation
- a. Automatic Actuation N.A. N.A. N.A. N.A. SFCP(1) SFCP(1)SFCP(4) 1, 2 N> Lo~ic and Actuation (J1 co s
(!)
Re ays
(!)
Level-High-High g
z c. Safety Injection N.A. N.A. N.A. SFCP N.A. N.A. N.A. 1, 2 0 Actuation Logic
~~ d. Tave Low Coincident N.A. SFCP SFCP N.A. N.A. N.A. N.A. 1, 2 with Reactor Trip (P-4)
J~
J~
~~
~~
TABLE 4.3-2 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION
-~
r' r'
SURVEILLANCE REQUIREMENTS TRIP r/)
ANALOG ACTUATING MODES
~tTl CHANNEL DEVICE MASTERSLAVE FOR WHICH CHANNELCHANNEL OPERATIONALOPERATIONALACTUATIONRELAY RELAY SURVEILLANCE
~ FUNCTIONAL UNIT
...., 6. Auxiliary Feedwater
(.;,.)
CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
- a. Manual Initiation N.A. N.A. N.A. SFCP N.A. N.A. N.A. 1, 2, 3
(.;,.)
- c. Steam Generator Water SFCP SFCP SFCP N.A. N.A. N.A. N.A. 1, 2, 3
.j:;>.
(.;,.) Level-Low-Low I
(.;,.)
\0 d. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
- e. Loss-of-Offsite Power See Item 8. below for all Loss ofPower Surveillance.
- f. Containment See Item 2. above for all CDA Surveillance Requirements.
Depressurization Actuation (CDA)
N ~ 7. Control Building Isolation Ul (11 CP [ a. Manual Actuation N.A. N.A. N.A. SFCP N.A. N.A. N.A.
- a (11
- b. Manual Safety Injection N.A. N.A. N.A. SFCP N.A. N.A. N.A. 1, 2, 3, 4 g
z Actuation 0
~~ Logic and Actuation J:~ Relays J:~ d. Containment Pressure-- SFCP SFCP SFCP N.A. N.A. N.A. N.A. 1, 2, 3
~~ High-1
~$~~
TABLE 4.3-2 (Continued)
~
........ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION r SURVEILLANCE REQUIREMENTS rC/).
0 TRIP z
tTl ANALOG ACTUATING MODES CHANNEL DEVICE MASTERSLAVEFOR WHICH c:: CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAYSURVEILLANCE
~ FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED w 7. Control Building Isolation (Continued)
- Ventilation Radiation w 8. Loss ofPower
+:>.
VJ a. 4 kV Bus Undervoltage N.A. SFCP N.A. SFCP(3) N.A. N.A. N.A. 1, 2, 3, 4 I
+:>.
0 (Loss ofVoltage)
- 9. Engineered Safety Features Actuation System N
U1 a>
(1)
Interlocks co ::s a. Pressurizer Pressure, N.A. SFCP SFCP N.A. N.A. N.A. N.A. 1, 2, 3 0..
a(1)
P-11
- s
0 P-12
~t c. Reactor Trip, P-4 N.A. N.A. N.A. SFCP N.A. N.A. N.A. 1, 2, 3 3~ 10. Emergency Generator N.A. N.A. N.A. N.A. SFCP(1,2) N.A. N.A. 1, 2, 3, 4
~c!3 Load Sequencer
~$ ~:\3 11. Cold Leg Injection SFCP SFCP SFCP N.A. N.A. N.A. N.A. 1, 2, 3
~~*
Permissive, P-19
TABLE 4.3-2 (Continued)
TABLE NOTATION
- 1. Each train shall be tested at the frequency specified in the Surveillance Frequency Control Program.
- 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 the frequency specified in the Surveillance Frequency Control Program.
- 3. At the frequency specified in the Surveillance Frequency Control Program, 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.
- 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.
- MODES 1, 2, 3, and 4.
During movement of recently irradiated fuel assemblies.
MILLSTONE - UNIT 3 3/4 3-41 Amendment No. #, 14, 19, i-00, -H9,
+9&, ~. ;H-9, n9, ~. U3-,
258
INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING FOR PLANT OPERATIONS LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels for plant operations 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 for plant operations exceeding the value shown in Table 3.3-6, adjust the Setpoint to within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or declare the channel inoperable.
- b. With one or more radiation monitoring channels for plant operations inoperable, take the ACTION shown in Table 3.3-6.
- c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.1 Each required radiation monitoring instrumentation channel for plant operations shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST for the MODES and at the frequencies shown in Table 4.3-3.
MILLSTONE - UNIT 3 3/4 3-42 Amendment No.~'
258
~
l' l'
\/).
......, TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT
~
trJ OPERATIONS SURVEILLANCE REQUIREMENTS ANALOG
~
......, CHANNEL MODES FOR WHICH V-l CHANNEL CHANNEL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED
- 1. Containment
- a. Deleted V-l
~
V-l
- b. RCS Leakage Detection
~
I
- 2) Deleted
- 2. Fuel Storage Pool Area Monitors
- N )>-
(.]1 3 TABLE NOTATIONS co (1)
- l s
(1)
- With fuel in the fuel storage pool area.
a z
0 3~
~$:
J~
~~
J~~
INSTRUMENTATION REMOTE SHUTDOWN INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The Remote Shutdown Instrumentation transfer switches, power, controls and monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
- a. With the number of OPERABLE remote shutdown monitoring channels less than the Minimum Channels OPERABLE as required by Table 3.3-9, restore the inoperable channel(s) to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- b. With one or more Remote Shutdown Instrumentation transfer switches, power, or control circuits inoperable, restore the inoperable switch(s)/circuit(s) to OPERABLE status within 7 days, or be in HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- c. Entry into an OPERATIONAL MODE is permitted while subject to these ACTION requirements.
SURVEILLANCE REQUIREMENTS 4.3.3.5.1 Each required remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6.
4.3.3.5.2 Each Remote Shutdown Instrumentation transfer switch, power and control circuit including the actuated components, shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE - UNIT 3 3/4 3-53 Amendment No. ~. '1-9-, GG, 258
~
L' L'
TABLE 4.3-6
[/).
....., REMOTE SHUTDOWN MONITORING INSTRUMENTATION 0 SURVEILLANCE REQUIREMENTS ztTl I CHANNEL CHANNEL
~.....,........ INSTRUMENT CHECK CALIBRATION
- 1. Reactor Trip Breaker Indication SFCP N.A.
w
- 4. Steam Generator Pressure SFCP SFCP w
---w
.j::. 5. Steam Generator Water Level SFCP SFCP I
VI 6. Auxiliary Feedwater Flow Rate SFCP SFCP 00
- 9. Reactor Coolant System Pressure SFCP SFCP (Wide Range)
(J1 a 12. Containment Pressure SFCP SFCP co ~
- l 0..
~ 14. Source Range Count Rate SFCP* SFCP
- l
~
0 z 15. Intermediate Range Amps SFCP SFCP
~~ 16. Boric Acid Tank Level SFCP SFCP J~
- When below P-6 (intermediate range neutron flux interlock setpoint).
~~
LIMITING CONDITION FOR OPERATION (Continued) action taken, the cause of the inoperability, and the plans and schedule for restoring the channel to OPERABLE status.
- f. With the number of OPERABLE channels for the reactor vessel water level monitor less than the minimum channels OPERABLE requirements of Table 3.3-10, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if repairs are feasible without shutting down or:
- 1. Initiate an alternate method of monitoring the reactor vessel inventory;
- 2. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the channel(s) to OPERABLE status; and
- 3. Restore the channel(s) to OPERABLE status at the next scheduled refueling.
- g. Entry into an OPERATIONAL MODE is permitted while subject to these ACTION requirements.
SURVEILLANCE REQUIREMENTS 4.3 .3 .6.1 Each required accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies shown in Table 4.3-7.
4.3.3.6.2 Deleted MILLSTONE - UNIT 3 3/4 3-59a Amendment No. 41, ~. %, -t-e-, ~
258
~
r r[/)
...., TABLE 4.3-7 0 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ztTJ I
CHANNEL
~ INSTRUMENT CHANNEL CHECK CALIBRATION
...., 1. Containment Pressure w
- 2. Reactor Coolant Outlet Temperature - T HOT (Wide Range) SFCP SFCP w
..j:::..
wI
- 3. Reactor Coolant Inlet Temperature - T COLD (Wide Range) SFCP SFCP 0\
N
- 4. Reactor Coolant Pressure - Wide Range SFCP SFCP
- 7. Steam Generator Water Level- Narrow Range SFCP SFCP
- 8. Steam Generator Water Level - Wide Range SFCP SFCP
(J'1 co a>-
(1) 11. Auxiliary Feedwater Flow Rate SFCP SFCP
- s a0..
(1)
- 12. Reactor Coolant System Subcooling Margin Monitor SFCP SFCP a 13. Containment Water Level (Wide Range) SFCP SFCP 0
z 14. Core Exit Thermocouples SFCP SFCP
~~ 15. DELETED J~
~~
~
r r
r:/1 TABLE 4.3-7 (Continued)
ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
~
tr1 CHANNEL CHANNEL
~ INSTRUMENT CHECK CALIBRATION
~ 16. Containment Area- High Range Radiation Monitor SFCP SFCP*
w
- 17. Reactor Vessel Water Level SFCP SFCP**
- 18. Deleted
~
wI 0\
w
- CHANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/h and a one point calibration check of the detector below 10 R/h with an installed or portable gamma source.
- Electronic calibration from the ICC cabinets only.
N U1 s
0'\ (!)
- l
~
(!)
g z
?
~~
J~
~~
~~
~
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 the frequency specified in the Surveillance Frequency Control Program that shall include verification that the Shutdown Margin Monitor is set per the CORE OPERATING LIMITS REPORT (COLR).
- b. At the frequency specified in the Surveillance Frequency Control Program VERIFY the minimum count rate (counts/sec) as defined within the COLR.
MILLSTONE - UNIT 3 3/4 3-83 Amendment No. -l-64, ~'
258
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 Four reactor coolant loops shall be OPERABLE and in operation.
APPLICABILITY: MODES 1 and 2.*
ACTION:
With less than the above required reactor coolant loops in operation, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified in operation and circulating reactor coolant at the frequency specified in the Surveillance Frequency Control Program.
- See Special Test Exceptions Specification 3.1 0.4.
MILLSTONE - UNIT 3 3/4 4-1 Amendment No.~'
258
REACTOR COOLANT SYSTEM HOT STANDBY SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability.
4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 17% at the frequency specified in the Surveillance Frequency Control Program.
4.4.1.2.3 The required reactor coolant loops shall be verified in operation and circulating reactor coolant at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE - UNIT 3 3/4 4-2a Amendment No. BB, 258
REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION (continued)
- b. With less than the above required reactor coolant loops in operation and the Control Rod Drive System is capable of rod withdrawal, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> open the Reactor Trip System breakers.
- c. With no loop in operation, suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.1.2 and immediately initiate corrective action to return the required loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required pump(s), if not in operation, shall be determined OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability.
4.4.1.3.2 The required steam generator(s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 17% at the frequency specified in the Surveillance Frequency Control Program.
4.4.1.3.3 The required loop(s) shall be verified in operation and circulating reactor coolant at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE - UNIT 3 3/4 4-4 Amendment No. -14§., +9+, ~'
258
REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION ACTION:
- a. With less than the required RHR loop(s) OPERABLE or with less than the required steam generator water level, immediately initiate corrective action to return the inoperable RHR loop to OPERABLE status or restore the required steam generator water level as soon as possible.
- b. With no RHR loop in operation, suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3 .1.1.1.2 and immediately initiate corrective action to return the required RHR loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at the frequency specified in the Surveillance Frequency Control Program.
4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at the frequency specified in the Surveillance Frequency Control Program.
4.4.1.4.1.3 The required pump, if not in operation, shall be determined OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power availability.
MILLSTONE- UNIT 3 3/4 4-Sa Amendment No. -t-S-1, +9+, ~'
258
REACTOR COOLANT SYSTEM COLD SHUTDOWN -LOOPS NOT FILLED SURVEILLANCE REQUIREMENTS 4.4.1.4.2.1 The required pump, if not in operation, shall be determined OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power availability.
4.4.1.4.2.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE- UNIT 3 3/4 4-6a Amendment No. ~' -l-9f.,
258
REACTOR COOLANT SYSTEM LOOP STOP VAL YES 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 1, 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- b. *( 1) With one or more RCS loop stop valves closed, maintain the valve(s) closed and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.1.5 Verify each RCS loop stop valve is open and the power removed from the valve operator at the frequency specified in the Surveillance Frequency Control Program.
- ( 1)All required ACTIONS of ACTION Statement 3.4.1.5.b shall be completed whenever this action is entered.
MILLSTONE- UNIT 3 3/4 4-7 Amendment No. ~' ~
258
REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.3.1 The pressurizer shall be OPERABLE with:
- a. at least two groups of pressurizer heaters, each having a capacity of at least 175 kW; and
- b. water level maintained at programmed level +/-6% of full scale (Figure 3.4-5).
APPLICABILITY: MODES 1 and 2.
ACTION:
- a. With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- b. With pressurizer water level outside the parameters described in Figure 3.4-5, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> restore programmed level to within+/- 6% of full scale, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- c. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the Reactor Trip System breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.3.1.1 The pressurizer water level shall be verified to be within programmed level+/- 6% of full scale at the frequency specified in the Surveillance Frequency Control Program.
4.4.3 .1.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by energizing the heaters and measuring circuit current at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE- UNIT 3 3/4 4-11 Amendment No. W(}, -l-eG, ~.
258
REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.3.2 The pressurizer shall be OPERABLE with:
- a. at least two groups of pressurizer heaters, each having a capacity of at least 175 kW; and
- b. water level less than or equal to 89% of full scale.
APPLICABILITY: MODE 3 ACTION:
- a. With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of being declared inoperable, or be in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- b. With the pressurizer otherwise inoperable, be in HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.3 .2.1 The pressurizer water level shall be determined to be less than or equal to 89% of full scale at the frequency specified in the Surveillance Frequency Control Program.
4.4.3 .2.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by energizing the heaters and measuring circuit current at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE- UNIT 3 3/4 4-llb Amendment No. -l-6B, ;t-l-B, 258
REACTOR COOLANT SYSTEM RELIEF VALVES SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE by:
- a. Performance of a CHANNEL CALIBRATION at the frequency specified in the Surveillance Frequency Control Program; and
- b. Operating the valve through one complete cycle of full travel during MODES 3 or 4 at the frequency specified in the Surveillance Frequency Control Program; and
- c. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV high pressurizer pressure actuation channels, but excluding valve operation, at the frequency specified in the Surveillance Frequency Control Program; and
- d. Verify the PORV high pressure automatic opening function is enabled at the frequency specified in the Surveillance Frequency Control Program.
4.4.4.2 Each block valve shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. in Specification 3.4.4.
MILLSTONE - UNIT 3 3/44-13 Amendment No. 8-8-, ~' -l4l-, ~'
258 U-G,
REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION ACTION: (Continued)
- 2. Appropriate grab samples of the containment atmosphere are obtained and analyzed for particulate radioactivity within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and at least once per 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> thereafter, and
- 3. A Reactor Coolant System water inventory balance is performed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and at least once per 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> thereafter.
Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:
- a. Containment Atmosphere Particulate Radioactivity Monitoring System-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 Monitoring System-performance of CHANNEL CALIBRATION at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE- UNIT 3 3/4 4-21a Amendment No. ;?;44.,
258
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System operational LEAKAGE shall be demonstrated to be within each of the above limits by:
- a. Deleted
- b. Deleted
- c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2250 +/- 20 psia at the frequency specified in the Surveillance Frequency Control Program with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4;
- - - - - - - - - - - - - - - - NOTES- - - - - - - - - - - - - - -
- 1. Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
- 2. Not applicable to primary to secondary LEAKAGE.
- d. Performance of a Reactor Coolant System water inventory balance at the frequency specified in the Surveillance Frequency Control Program;
Nom---------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
- e. Verification that primary to secondary LEAKAGE is:::;; 150 gallons per day through any one Steam Generator at the frequency specified in the Surveillance Frequency Control Program, and;
- f. Monitoring the Reactor Head Flange Leakoff System at the frequency specified in the Surveillance Frequency Control Program.
4.4.6.2.2(l)( 2)Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying LEAKAGE to be within its limit:
(1) The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.
2
( ) This surveillance is not required to be performed on Reactor Coolant System Pressure Isolation Valves located in the RHR flow path when in, or during the transition to or from, the shutdown cooling mode of operation.
MILLSTONE - UNIT 3 3/4 4-23 Amendment No. i:-GG, ~. -H4, ~
258 ~.;s&,
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS (Continued)
- a. At the frequency specified in the Surveillance Frequency Control Program,
- b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months,
- c. Deleted
- d. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to automatic or manual action or flow through the valve, and
- e. When tested pursuant to Specification 4.0.5.
MILLSTONE - UNIT 3 3/4 4-23a Amendment No. B-&
258
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.8.1 VerifY the specific activity of the reactor coolant less than or equal to 81.2 microCuries per gram DOSE EQUIVALENT XE-133 at the frequency specified in the Surveillance Frequency Control Program.*
4.4.8.2 VerifY the specific activity of the reactor coolant less than or equal to 1.0 microCuries per gram DOSE EQUIVALENT I-131 at the frequency specified in the Surveillance Frequency Control Program,* and between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change of greater than or equal to 15% RATED THERMAL POWER within a one hour period.
- Surveillance only required to be performed for MODE 1 operation, consistent with the provisions of Specification 4.0.1.
MILLSTONE - UNIT 3 3/4 4-29 Amendment No.~'
258
REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Demonstrate that each required PORV is OPERABLE by:
- a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at the frequency specified in the Surveillance Frequency Control Program thereafter when the PORV is required OPERABLE;
- b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at the frequency specified in the Surveillance Frequency Control Program; and
- c. Verifying the PORV block valve is open and the PORV Cold Overpressure Protection System (COPPS) is armed at the frequency specified in the Surveillance Frequency Control Program when the PORV is being used for overpressure protection.
4.4.9.3.2 Demonstrate that each required RHR suction relief valve is OPERABLE by:
- a. Verifying the isolation valves between the RCS and each required RHR suction relief valve are open at the frequency specified in the Surveillance Frequency Control Program; and
- b. Testing pursuant to Specification 4.0.5.
4.4.9.3.3 When complying with 3.4.9.3.4, verify that the RCS is vented through a vent pathway
~ 2.0 square inches at the frequency specified in the Surveillance Frequency Control Program for a passive vent path and at the frequency specified in the Surveillance Frequency Control Program for unlocked open vent valves.
4.4.9.3.4 Verify that no Safety Injection pumps are capable of injecting into the RCS at the frequency specified in the Surveillance Frequency Control Program.
4.4.9.3.5 Verify that a maximum of one centrifugal charging pump is capable of injecting into the RCS at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE - UNIT 3 3/4 4-39 Amendment No. ':/-9-, 8-G, +00, .ffi, 1:-5-1, 258 +9+,~,
3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:
- a. The isolation valve open and power removed,
- b. A contained borated water volume of between 6618 and 7030 gallons,
- c. A boron concentration of between 2600 and 2900 ppm, and
- d. A nitrogen cover-pressure of between 636 and 694 psia.
APPLICABILITY: MODES 1, 2, and 3*.
ACTION:
- a. With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.5.1 Each accumulator shall be demonstrated OPERABLE:
- a. At the frequency specified in the Surveillance Frequency Control Program by:
- 1) VerifYing that the contained borated water volume and nitrogen cover-pressure in the tanks are within their limits, and
- 2) VerifYing that each accumulator isolation valve is open.
- b. At the frequency specified in the Surveillance Frequency Control Program and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solution volume increase of greater than or equal to 1%
of tank volume by verifYing the boron concentration of the accumulator solution.
This surveillance is not required when the volume increase makeup source is the RWST.
- Pressurizer pressure above 1000 psig.
MILLSTONE - UNIT 3 3/4 5-1 Amendment No. H, -5+, W, -l-00, 258
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- c. At the frequency specified in the Surveillance Frequency Control Program when the RCS pressure is above 1000 psig by verifying that the associated circuit breakers are locked in a deenergized position or removed.
MILLSTONE - UNIT 3 3/4 5-2 Amendment No. -l-00, -ill-,
258
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
- a. At the frequency specified in the Surveillance Frequency Control Program by verifying that the following valves are in the indicated positions with power to the valve operators removed:
Valve Number Valve Function Valve Position 3SIH*MV8806 RWST Supply to SI Pumps OPEN 3SIH*MV8802A SI Pump A to Hot Leg Injection CLOSED 3SIH*MV8802B SI Pump B to Hot Leg Injection CLOSED 3SIH*MV8835 SI Cold Leg Master Isolation OPEN 3SIH*MV8813 SI Pump Master Miniflow OPEN Isolation 3SIL *MV8840 RHR to Hot Leg Injection CLOSED 3SIL *MV8809A RHR Pump A to Cold Leg OPEN Injection 3SIL *MV8809B RHR Pump B to Cold Leg OPEN Injection
- b. At the frequency specified in the Surveillance Frequency Control Program by:
- 1) Verifying that the ECCS piping, except for the operating centrifugal charging pump(s) and associated piping, the RSS pump, the RSS heat exchanger and associated piping, is full of water, and
- 2) Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:
- 1) For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and
- 2) At least once daily ofthe areas affected (during each day) within containment by containment entry and during the final entry when CONTAINMENT INTEGRITY is established.
- d. At the frequency specified in the Surveillance Frequency Control Program by:
- 1) Verifying automatic interlock action of the RHR System from the Reactor Coolant System by ensuring that with a simulated signal greater than or equal to 412.5 psia the interlocks prevent the valves from being opened.
MILLSTONE - UNIT 3 3/4 5-4 Amendment No. 6G, '1-9, -l-00, ~. -!41, 258 36, ~.
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 2) A visual inspection of the containment sump and verifYing that the subsystem suction inlets are not restricted by debris and that the sump components (strainers, etc.) show no evidence of structural distress or abnormal corrosion.
- e. At the frequency specified in the Surveillance Frequency Control Program by:
- 1) VerifYing that each automatic valve in the flow path actuates to its correct position on a Safety Injection actuation test signal, and
- 2) VerifYing that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:
a) Centrifugal charging pump, b) Safety Injection pump, and c) RHRpump.
- 3) VerifYing that the Residual Heat Removal pumps stop automatically upon receipt of a Low-Low RWST Level test signal.
- f. By verifYing that each of the following pump's developed head at the test flow point is greater than or equal to the required developed head when tested pursuant to Specification 4.0.5:
- 1) Centrifugal charging pump
- 2) Safety Injection pump
- 3) RHRpump
- 4) Containment recirculation pump
- g. By verifYing the correct position of each electrical and/or mechanical position stop for the following ECCS throttle valves:
- 1) Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of each valve stroking operation when the ECCS subsystems are required to be OPERABLE, and
- 2) At the frequency specified in the Surveillance Frequency Control Program.
ECCS Throttle Valves Valve Number Valve Number 3SIH*V6 3SIH*V25 3SIH*V7 3SIH*V27 MILLSTONE- UNIT 3 3/4 5-5 Amendment No. W, H4, +§-§., ;w6, 258 ~'~
EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:
- a. A contained borated water volume between 1,166,000 and 1,207,000 gallons,
- c. A minimum solution temperature of 40°F, and
- d. A maximum solution temperature of 50°F.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.5.4 The RWST shall be demonstrated OPERABLE:
- a. At the frequency specified in the Surveillance Frequency Control Program by:
- 1) Verifying the contained borated water volume in the tank, and
- 2) Verifying the boron concentration of the water.
- b. At the frequency specified in the Surveillance Frequency Control Program by verifying the RWST temperature.
MILLSTONE- UNIT 3 3/4 5-9 Amendment No. H, 6B, 258
EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 pH TRISODIUM PHOSPHATE STORAGE BASKETS LIMITING CONDITION FOR OPERATION 3.5.5 The trisodium phosphate (TSP) dodecahydrate Storage Baskets shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:
With the TSP Storage Baskets inoperable, restore the system TSP Storage Baskets to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.5.5 The TSP Storage Baskets shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying that a minimum total of 974 cubic feet ofTSP is contained in the TSP Storage Baskets.
MILLSTONE - UNIT 3 3/4 5-10 Amendment No. H-5-, ~'
258
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 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
- a. At the frequency specified in the Surveillance Frequency Control Program by verifying that all penetrations(!) not capable of being closed by OPERABLE containment automatic isolation valves,{l) 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; and
- b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3.
- c. Deleted (I)
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 during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.
(2)
In MODE 4, the requirement for an OPERABLE containment isolation valve system is satisfied by use of the containment isolation actuation pushbuttons.
(3)
Isolation devices in high radiation areas may be verified by use of administrative means.
MILLSTONE - UNIT 3 3/4 6-1 Amendment No. §9, -t-54-, -l-86, :t.-t-6 258
CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION Continued
- c. With the containment air lock inoperable, except as specified in ACTION a. or ACTION b. above, immediately initiate action to evaluate overall containment leakage rate per Specification 3.6.1.2 and verify an air lock door is closed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Restore the air lock to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:
- a. By verifying leakage results in accordance with the Containment Leakage Rate Testing Program. Containment air lock leakage test results shall be evaluated against the leakage limits of Technical Specification 3.6.1.2. (An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test).
- b. Deleted
- c. At the frequency specified in the Surveillance Frequency Control Program by verifying that only one door in each air lock can be opened at a time.
MILLSTONE- UNIT 3 3/4 6-6 Amendment No. §9, ~' +86, ~'
258
CONTAINMENT SYSTEMS CONTAINMENT PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment pressure shall be maintained greater than or equal to 10.6 psia and less than or equal to 14.0 psia.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With the containment pressure less than 10.6 psia or greater than 14.0 psia, restore the containment pressure to within the limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.1.4 The primary containment pressure shall be determined to be within the limits at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE - UNIT 3 3/4 6-7 Amendment No.~
258
CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall be maintained greater than or equal to 80°F and less than or equal to 120°F.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With the containment average air temperature less than 80°F or greater than 120°F, restore the average air temperature to within the limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arithmetical average of the temperatures at the following locations and shall be determined at the frequency specified in the Surveillance Frequency Control Program:
Location
- a. 94 ft elevation, E outside crane wall
- b. 86 ft elevation, NW outside crane wall
- c. 75ft elevation, W Steam Generator platform
- d. 75 ft elevation, E Steam Generator platform
- e. 45 ft elevation, Pressurizer cubicle, crane wall MILLSTONE - UNIT 3 3/4 6-9 Amendment No. 258
CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.7 Each containment purge supply and exhaust isolation valve shall be OPERABLE and each 42-inch containment shutdown purge supply and exhaust isolation valve shall be closed and locked closed.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
- a. With a 42-inch containment purge supply and/or exhaust isolation valve open or not locked closed, close and/or lock close that valve or isolate the penetration(s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.1. 7.1 The containment purge supply and exhaust isolation valves shall be verified to be locked closed and closed at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE - UNIT 3 3/4 6-11 Amendment No. ,
258
CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT QUENCH SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Quench Spray subsystems shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With one Containment Quench Spray subsystem inoperable, restore the inoperable system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.2.1 Each Containment Quench Spray subsystem shall be demonstrated OPERABLE:
- a. At the frequency specified in the Surveillance Frequency Control Program, by:
- 1) Verifying that each valve (manual, power operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position; and
- 2) Verifying the temperature of the borated water in the refueling water storage tank is between 40°F and 50°F.
- b. By verifying that each pump's developed head at the test flow point is greater than or equal to the required developed head when tested pursuant to Specification 4.0.5;
- c. At the frequency specified in the Surveillance Frequency Control Program, by:
- 1) Verifying that each automatic valve in the flow path actuates to its correct position on a CDA test signal, and
- 2) Verifying that each spray pump starts automatically on a CDA test signal.
- d. By verifying each spray nozzle is unobstructed following maintenance that could cause nozzle blockage.
MILLSTONE - UNIT 3 3/46-12 Amendment No. §., §.(}, -Hm, ~. B-§-,
258 +7-1' ;?;%, ;?ti
CONTAINMENT SYSTEMS RECIRCULATION SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 Two independent Recirculation Spray Systems shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With one Recirculation Spray System inoperable, restore the inoperable system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the inoperable Recirculation Spray System to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.2.2 Each Recirculation Spray System shall be demonstrated OPERABLE:
- a. At the frequency specified in the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;
- b. By verifying that each pump's developed head at the test flow point is greater than or equal to the required developed head when tested pursuant to Specification 4.0.5;
- c. At the frequency specified in the Surveillance Frequency Control Program by verifYing that on a CDA test signal, each recirculation spray pump starts automatically after receipt of an RWST Low-Low signal;
- d. At the frequency specified in the Surveillance Frequency Control Program, by verifYing that each automatic valve in the flow path actuates to its correct position on a CDA test signal; and
- e. By verifYing each spray nozzle is unobstructed following maintenance that could cause nozzle blockage.
MILLSTONE- UNIT 3 3/4 6-13 Amendment No. §G, 99, ~' ~'
258 +F/-, ~' ~' ~'
CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 The containment isolation valves shall be OPERABLE. (l) (2 )
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With one or more of the isolation valve(s) inoperable, maintain at least one isolation barrier OPERABLE in the affected penetration(s), and:
- a. Restore the inoperable valve(s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or
- b. Isolate the affected penetration(s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of deactivated automatic valve(s) secured in the isolation position(s), or
- c. Isolate the affected penetration(s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of closed manual valve(s) or blind flange(s); or
- d. Isolate the affected penetration that has only one containment isolation valve and a closed system withm 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> by use of at least one closed and deactivated automatic valve, closed manual valve, or blind flange; or
- e. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.3.1 DELETED 4.6.3.2 Each isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at the frequency specified in the Surveillance Frequency Control Program by:
- a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position,
- b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position, and
- c. Verifying that on a Containment High Radiation test signal, each purge supply and exhaust Isolation valve actuates to its isolation position.
4.6.3.3 The isolation time of each power-operated or automatic valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5.
(1) The provisions of this Specification are not applicable for main steam line isolation valves.
However, provisions of Specification 3. 7 .1.5 are applicable for main steam line isolation valves.
2
( ) Containment isolation valves may be opened on an intermittent basis under administrative controls.
MILLSTONE - UNIT 3 3/46-15 Amendment No. ::2-8-, ~' H-, %, ti, 258 :M, ~' ~' ~'
CONTAINMENT SYSTEMS 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM STEAM JET AIR EJECTOR LIMITING CONDITION FOR OPERATION 3.6.5.1 The inside and outside isolation valves in the steam jet air ejector suction line shall be closed.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With the inside or outside isolation valves in the steam jet air ejector suction line not closed, restore the valve to the closed position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.5.1.1 The steam jet air ejector suction line outside isolation valve shall be determined to be in the closed position by a visual inspection prior to increasing the Reactor Coolant System temperature above 200°F and at the frequency specified in the Surveillance Frequency Control Program.
4.6.5.1.2 The steam jet air ejector suction line inside isolation valve shall be determined to be locked in the closed position by a visual inspection prior to increasing the Reactor Coolant System temperature above 200°F.
MILLSTONE - UNIT 3 3/4 6-18 Amendment No. -t-00, 258
CONTAINMENT SYSTEMS 3/4.6.6 SECONDARY CONTAINMENT SUPPLEMENTARY LEAK COLLECTION AND RELEASE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.6.1 Two independent Supplementary Leak Collection and Release Systems shall be OPERABLE with each system comprised of:
- a. one OPERABLE filter and fan, and
- b. one OPERABLE Auxiliary Building Filter System as defined in Specification 3.7.9.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With one Supplementary Leak Collection and Release System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.6.1 Each Supplementary Leak Collection and Release System shall be demonstrated OPERABLE:
- a. At the frequency specified in the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying a system flow rate of 7600 cfm to 9800 cfm and that the system operates for at least 10 continuous hours with the heaters operating.
- b. At the frequency specified in the Surveillance Frequency Control Program and following painting, fire, or chemical release in any ventilation zone communicating with the system by:
- 1) Verifying that the system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978,
- and the system flow rate is 7600 cfm to 9800 cfm; MILLSTONE - UNIT 3 3/4 6-19 Amendment No. :2o-, £, s::t, +00, H:3-,
258 :2,%, :g:;.,
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 2.5% when tested in accordance with ASTM D3803-89 at a temperature of 30°C (86°F) and a relative humidity of 70%; and
- 3) Verifying a system flow rate of7600 cfm to 9800 cfm during system operation when tested in accordance with ANSI N51 0-1980.
- c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, by verifying, within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,
- shows the methyl iodide penetration less than or equal to 2.5% when tested in accordance with ASTM 03803-89 at a temperature of30°C (86°F) and a relative humidity of70%:
- d. At the frequency specified in the Surveillance Frequency Control Program by:
- 1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.25 inches Water Gauge while operating the system at a flow rate of 7600 cfm to 9800 cfm,
- 2) Verifying that the system starts on a Safety Injection test signal, and
- 3) Verifying that the heaters dissipate 50 +/-5 kW when tested in accordance with ANSI N510-1980.
- ANSI N510-1980 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52, Revision 2, March 1978.
MILLSTONE - UNIT 3 3/4 6-20 Amendment No. ;?;, £, ~. -l-00, ~.
258 ~. +84, ~.
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT LIMITING CONDITION FOR OPERATION 3.6.6.2 Secondary Containment shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With Secondary Containment inoperable, restore Secondary Containment to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENT 4.6.6.2.1 OPERABILITY of Secondary Containment shall be demonstrated at the frequency specified in the Surveillance Frequency Control Program by verifying that each door in each access opening is closed except when the access opening is being used for normal transit entry and exit.
4.6.6.2.2 At the frequency specified in the Surveillance Frequency Control Program, verify each Supplementary Leak Collection and Release System produces a negative pressure of greater than or equal to 0.4 inch water gauge in the Auxiliary Building at 24' -6" elevation within 120 seconds after a start signal.
MILLSTONE - UNIT 3 3/4 6-22 Amendment No. g:::j-, ~' -+/-R, ~'
258 ~.
PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION ACTION: (Continued)
Inoperable Equipment Required ACTION
- e. Three auxiliary feedwater e.
pumps in MODE 1, 2, or 3.
- - - - - - - NOTE --------
LCO 3.0.3 and all other LCO required ACTIONS requiring MODE changes are suspended until one AFW pump is restored to OPERABLE status.
Immediately initiate ACTION to restore one auxiliary feedwater pump to OPERABLE status.
SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:
- a. At the frequency specified in the Surveillance Frequency Control Program by:
- - - - - - - - - - - NOTE - - - - - - - - - - - - - - -
Auxiliary feedwater pumps may be considered OPERABLE during alignment and operation for steam generator level control, if they are capable of being manually realigned to the auxiliary feedwater mode of operation.
Verifying each auxiliary feedwater manual, power operated, and automatic valve in each water flow path and in each required steam supply flow path to the steam turbine driven auxiliary feedwater pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.
- b. At least once per 92 days on a STAGGERED TEST BASIS, tested pursuant to Specification 4.0.5, by:
- 1) Verifying that on recirculation flow each motor-driven pump develops a total head of greater than or equal to 3385 feet;
- 2) Verifying that on recirculation flow the steam turbine-driven pump develops a total head of greater than or equal to 3 780 feet when the secondary steam supply pressure is greater than 800 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
MILLSTONE- UNIT 3 3/4 7-5 Amendment No. 96, +00, H-1, -!-39, 258 206, 235,
PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
- c. At the frequency specified in the Surveillance Frequency Control Program by verifying that each auxiliary feed water pump starts as designed automatically upon receipt of an Auxiliary Feedwater Actuation test signal. For the steam turbine-driven auxiliary feedwater pump, the provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
- 4. 7 .1.2.2 An auxiliary feed water flow path to each steam generator shall be demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days prior to entering MODE 2 by verifying flow to each steam generator.
MILLSTONE - UNIT 3 3/4 7-5a Amendment No.~'
258
PLANT SYSTEMS DEMINERALIZED WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The demineralized water storage tank (DWST) shall be OPERABLE with a water volume of at least 334,000 gallons.
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
With the DWST inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:
- a. Restore the DWST to OPERABLE status or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, or
- b. Demonstrate the OPERABILITY of the condensate storage tank (CST) as a backup supply to the auxiliary feedwater pumps and restore the DWST to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The DWST shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying the water volume is within its limits when the tank is the supply source for the auxiliary feedwater pumps.
- 4. 7 .1.3 .2 The CST shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying that the combined volume of both the DWST and CST is at least 384,000 gallons of water whenever the CST and DWST are the supply source for the auxiliary feedwater pumps.
MILLSTONE - UNIT 3 3/4 7-6 Amendment No. W, 258
TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY
- 1. Gross Radioactivity At the frequency specified in the Surveillance Determination Frequency Control Program.
- 2. Isotopic Analysis for DOSE a) Once per 31 days, whenever the gross EQUIVALENT 1-131 radioactivity determination indicates Concentration concentrations greater than 10% of the allowable limit for radioiodines.
b) At the frequency specified in the Surveillance Frequency Control Program, whenever the gross radioactivity determination indicates concentrations less than or equal to 10% of the allowable limit for radioiodines.
MILLSTONE - UNIT 3 3/4 7-8 Amendment No. 258
PLANT SYSTEMS STEAM GENERA TOR ATMOSPHERIC RELIEF BYPASS LINES LIMITING CONDITION FOR OPERATION 3.7.1.6 Each steam generator atmospheric reliefbypass valve (SGARBV) line shall be OPERABLE, with the associated main steam atmospheric relief isolation (block) valve in the open position.
APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.
ACTIONS:
- a. With one required SGARBV line inoperable, restore required SGARBV line to OPERABLE status within 7 days or be in at least MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in MODE 4 without reliance upon steam generator for heat removal within the next 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />. LCO 3.0.4 is not applicable.
- b. With two or more required SGARBV lines inoperable, restore all but one required SGARBV line to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in MODE 4 without reliance upon steam generator for heat removal within the next 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.7.1.6.1 Verify one complete cycle of each SGARBV at the frequency specified in the Surveillance Frequency Control Program.
- 4. 7.1.6.2 Verify one complete cycle of each main steam atmospheric relief isolation (block) valve at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE- UNIT 3 3/4 7-9a Amendment No. -l-5+,
258
PLANT SYSTEMS 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent reactor plant component cooling water safety loops shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With only one reactor plant component cooling water safety loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.7.3 At least two reactor plant component cooling water safety loops shall be demonstrated OPERABLE:
- a. At the frequency specified in the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and
- b. At the frequency specified in the Surveillance Frequency Control Program by verifying that:
- 1) Each automatic valve actuates to its correct position on its associated Engineered Safety Feature actuation signal, and
- 2) Each Component Cooling Water System pump starts automatically on an SIS test signal.
MILLSTONE - UNIT 3 3/4 7-11 Amendment No. Hf., ~
258
PLANT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4 At least two independent service water loops shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With only one service water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.7.4 At least two service water loops shall be demonstrated OPERABLE:
- a. At the frequency specified in the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and
- b. At the frequency specified in the Surveillance Frequency Control Program by verifying that:
- 1) Each automatic valve servicing safety-related equipment actuates to its correct position on its associated Engineered Safety Feature actuation signal, and
- 2) Each Service Water System pump starts automatically on an SIS test signal.
MILLSTONE -UNIT 3 3/4 7-12 Amendment No. -H-1, ~.
258
PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION
- 3. 7.5 The ultimate heat sink (UHS) shall be OPERABLE with an average water temperature of less than or equal to 75°F.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
If the UHS temperature is above 75°F, monitor the UHS temperature once per hour for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
If the UHS temperature does not drop below 75°F during this period, place the plant in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. During this period, if the UHS temperature increases above 77°F, place the plant in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.7.5 The UHS shall be determined OPERABLE:
- a. At the frequency specified in the Surveillance Frequency Control Program by verifying the average water temperature to be within limits.
- b. At least once per 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> by verifying the average water temperature tq be within limits when the average water temperature exceeds 70°F.
MILLSTONE- UNIT 3 3/4 7-13 Amendment No. H-9, 258
PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION ACTION: (Continued)
- e. With both Control Room Emergency Air Filtration Systems inoperable, or with the OPERABLE Control Room Emergency Air Filtration System required to be in the emergency mode by ACTION d. not capable of being powered by an OPERABLE emergency power source, or with one or more Control Room Emergency Air Filtration System Trains inoperable due to an inoperable CRE boundary, immediately suspend the movement of recently irradiated fuel assemblies.
SURVEILLANCE REQUIREMENTS 4.7.7 Each Control Room Emergency Air Filtration System shall be demonstrated OPERABLE:
- a. At the frequency specified in the Surveillance Frequency Control Program by verifying that the control room air temperature is less than or equal to 95°F;
- b. At the frequency specified in the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying a system flow rate of 1,120 cfm +/- 20% and that the system operates for at least 10 continuous hours with the heaters operating;
- c. At the frequency specified in the Surveillance Frequency Control Program and following painting, fire, or chemical release in any ventilation zone communicating with the system by:
- 1) Verifying that the system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Position C.S.a, C.S.c, and C.S.d of Regulatory Guide 1.52, Revisions 2, March 1978,
- and the system flow rate is 1,120 cfm +/- 20%;
- 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,
- shows the methyl iodide penetration less than or equal to 2.5% when tested in accordance with ASTM D3803-89 at a temperature of 30°C (86°F), a relative humidity of 70%, and a face velocity of 54 ft/min; and
- 3) Verifying a system flow rate of 1,120 cfm +/- 20% during system operation when tested in accordance with ANSI N510-1980.
MILLSTONE - UNIT 3 3/47-16 Amendment No. ~, H3, -l-8-1-, -l-84, :2{B, 258 ~, ~, ;M;,
PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
- d. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,
- shows the methyl iodide penetration less than or equal to 2.5%
when tested in accordance with ASTM D3803-89 at a temperature of 30°C (86°F),
and a relative humidity of 70%, and a face velocity of 54 ft/min.
- e. At the frequency specified in the Surveillance Frequency Control Program by:
- 1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.75 inches Water Gauge while operating the system at a flow rate of 1,120 cfm +/- 20%;
- 2) Deleted
- 3) Verifying that the heaters dissipate 9.4 +/-1 kW when tested in accordance with ANSI N51 0-1980.
- f. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria ofless than 0.05% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 1120 cfm +/- 20%; and
- g. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N 510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 1120 cfm +/- 20%.
- h. By performance of CRE unfiltered air inleakage testing in accordance with the CRE Habitability Program at a frequency in accordance with the CRE Habitability Program.
- ANSI N510-1980 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52, Revision 2, March 1978.
MILLSTONE- UNIT 3 3/4 7-17 Amendment No.~. +B-, i-8+, ~. tiG, ~
258
PLANT SYSTEMS 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.9 Two independent Auxiliary Building Filter Systems shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With one Auxiliary Building Filter System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. In addition, comply with the ACTION requirements of Specification 3 .6.6.1.
SURVEILLANCE REQUIREMENTS 4.7.9 Each Auxiliary Building Filter System shall be demonstrated OPERABLE:
- a. At the frequency specified in the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying a system flow rate of30,000 cfm +/-10% and that the system operates for at least 10 continuous hours with the heaters operating;
- b. At the frequency specified in the Surveillance Frequency Control Program and following painting, fire, or chemical release in any ventilation zone communicating with the system by:
- 1) Verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978,
- and the system flow rate is 30,000 cfm +/-10%;
- 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,
- shows the methyl MILLSTONE - UNIT 3 3/4 7-20 Amendment No. ;!-, %1, H:3-, +84, ~, ~'
258 ~
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS iodide penetration less than or equal to 2.5% when tested in accordance with ASTM 03803-89 at a temperature of30°C (86°F), a relative humidity of 70%, and a face velocity of 52 ftlmin; and
- 3) Verifying a system flow rate of 30,000 cfm +/-1 0% during system operation when tested in accordance with ANSI N510-1980.
- c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,
- shows the methyl iodide penetration less than or equal to 2.5% when tested in accordance with ASTM D3803-89 at a temperature of30°C (86°F), a relative humidity of 70%, and a face velocity of 52 ft/min;
- d. At the frequency specified in the Surveillance Frequency Control Program by:
- 1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.8 inches Water Gauge while operating the system at a flow rate of30,000 cfm +/-10%,
- 2) Verifying that the system starts on a Safety Injection test signal, and
- 3) Verifying that the heaters dissipate 180 +/-18 kW when tested in accordance with ANSI N510-1980.
- e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N51 0-1980 for a DOP test aerosol while operating the system at a flow rate of 30,000 cfm +/- 10%;
and
- f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N 51 0-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of30,000 cfm +/-10%.
- ANSI N51 0-1980 shall be used in place of ANSI N51 0-1975 referenced in Regulatory Guide 1.52, Revision 2, March 1978.
MILLSTONE- UNIT 3 3/4 7-21 Amendment No.~' 'trl-, H:3-, -l-84, ;we 258
ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (continued)
Inoperable tqmpment Keqmrea AL llUN
- e. Two diesel generators e.2 Restore one ot tne moperaoJe 01ese1 generators to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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 1E Distribution System shall be:
- a. Determined OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying correct breaker alignments, indicated power availability, and
- b. Demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program 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 the frequency specified in the Surveillance Frequency Control Program 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. W, 64, ~. +94, ~.
258 ~.
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b) Simulated loss-of-offsite power by itself, or c) Simulated loss-of-offsite power in conjunction with an ESF Actuation test signal, or d) An ESF Actuation test signal by itself.
- 6) VerifYing the generator is synchronized and gradually loaded in accordance with the manufacturer's recommendations between 4800-5000 kW* and operates with a load between 4800-5000 kW* for at least 60 minutes, and
- 7) VerifYing the diesel generator is aligned to provide standby power to the associated emergency busses.
- b. At the frequency specified in the Surveillance Frequency Control Program by:
- 1) VerifYing that the diesel generator starts from standby conditions and attains generator voltage and frequency of 4160 +/- 420 volts and 60 +/-
0.8 Hz within 11 seconds after the start signal.
- 2) VerifYing the generator is synchronized to the associated emergency bus, loaded between 4800-5000 kW* in accordance with the manufacturer's recommendations, and operate with a load between 4800-5000 kW* for at least 60 minutes.
The diesel generator shall be started for this test using one of the signals in Surveillance Requirement 4.8.1.1.2.a.5. This test, if it is performed so it coincides with the testing required by Surveillance Requirement 4.8.1.1.2.a.5, may also serve to concurrently meet those requirements as well.
- c. At the frequency specified in the Surveillance Frequency Control Program and after each operation of the diesel where the period of operation was greater than or equal to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> by checking for and removing accumulated water from the day tank;
- d. At the frequency specified in the Surveillance Frequency Control Program by checking for and removing accumulated water from the fuel oil storage tanks;
- e. By sampling new fuel oil in accordance with ASTM-D4057 prior to addition to storage tanks and:
- 1) By verifYing in accordance with the tests specified in ASTM-D975-81 prior to add1tion to the storage tanks that the sample has:
a) An API Gravity of within 0.3 degrees at 60°F, or a specific gravity of within 0.0016 at 60/60°F, when compared to the supplier's certificate, or an absolute specific gravity at 60/60°F of greater than or equal to 0.83 but less than or equal to 0.89, or an API gravity of greater than or equal to 27 degrees but less than or equal to 39 degrees;
- The operating band is meant as guidance to avoid routine overloading of the diesel.
Momentary transients outside the load range shall not invalidate the test.
MILLSTONE- UNIT 3 3/4 8-4 Amendment No. 4, 64, ill-, -l-3-+, -l-94, 258
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b) A kinematic viscosity at 40°C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes (alternatively, Say bolt viscosity, SUS at 100°F of greater than or equal to 32.6, but less than or equal to 40.1 ), if gravity was not determined by comparison with the supplier's certification; c) A flash point equal to or greater than 125°F; and d) Water and sediment less than 0.05 percent by volume when tested in accordance with ASTM-Dl796-83.
- 2. By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM-0975-81 are met when tested in accordance with ASTM-D975-81 except that: (1) the cetane index shall be determined in accordance with ASTM-D976 (this test is an appropriate approximation for cetane number as stated in ASTM-D975-81 [Note E]),
and (2) the analysis for sulfur may be performed in accordance with ASTM-Dl552-79, ASTM-D2622-82 or ASTM-D4294-83.
- f. At the frequency specified in the Surveillance Frequency Control Program by obtaining a sample of fuel oil in accordance with ASTM-D2276-78, and verifying that total particulate contamination is less than 10 mg/liter when checked in accordance with ASTM-D2276-78, Method A;
- g. At the frequency specified in the Surveillance Frequency Control Program, during shutdown, by:
- 1) DELETED
- 2) Verifying the generator capability to reject a load of greater than or equal to 595 kW while maintaining voltage at 4160 +/- 420 volts and frequency at 60 +/-3Hz;
- 3) Verifying the generator capability to reject a load of 4986 kW without tripping. The generator voltage shall not exceed 5000 volts during and 4784 volts following the load rejection;
- 4) Simulating a loss-of-offsite power by itself, and:
a) Verifying deenergization of the emergency busses and load shedding from the emergency busses, and b) Verifying the diesel starts from standby conditions on the auto-start signal, energizes the emergency busses with permanently connected loads within 11 seconds, energizes the auto-connected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 +/- 420 volts and 60 +/- 0.8 Hz during this test.
MILLSTONE- UNIT 3 3/4 8-5 Amendment No. 4, W, 64, ':P.t, -l-00, m, m, -H-&, +94, 258
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 8) Verifying that the auto-connected loads to each diesel generator do not exceed the 2000-hour rating of 5335 kW;
- 9) Verifying the diesel generator's capability to:
a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Be restored to its standby status.
- 10) Verifying that with the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injection signal overrides the test mode by:
( 1) returning the diesel generator to standby operation, and (2) automatically energizing the emergency loads with offsite power;
- 11) DELETED
- 12) Verifying that the automatic load sequence timer is OPERABLE with the interval between each load block within +/- 10% of its design interval; and
- 13) DELETED
- h. At the frequency specified in the Surveillance Frequency Control Program by starting both diesel generators simultaneously from standby conditions, during shutdown, and verifying that both diesel generators achieve generator voltage and frequency at 4160 +/- 420 volts and 60 +/- 0.8 Hz in less than or equal to 11 seconds; and
- 1. At the frequency specified in the Surveillance Frequency Control Program by draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite solution.
MILLSTONE - UNIT 3 3/4 8-7 Amendment No. 64, 19, +00, ~' 11, 258 -l-94, ~'
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
J. At the frequency specified in the Surveillance Frequency Control Program by verifying the diesel generator operates for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. During the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of this test, the diesel generator shall be loaded between 5400-5500kW*
and during the remaining 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> of this test, the diesel generator shall be loaded between 4800-5000kW*. The generator voltage and frequency shall be 4160 +/-
420 volts and 60 +/- 0.8 Hz within 11 seconds after the start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test.** Within 5 minutes after completing this 24-hour test, perform Specification 4.8.1.1.2.a.5) excluding the requirement to start the diesel from standby conditions.***
- k. At the frequency specified in the Surveillance Frequency Control Program by verifying that the fuel transfer pump transfers fuel from each fuel storage tank to the day tank of each diesel via the installed cross-connection lines.
- 1. At the frequency specified in the Surveillance Frequency Control Program by verifying that the following diesel generator lockout features prevent diesel generator starting:
- 1) Engine overspeed,
- 2) Lube oil pressure low (2 of 3 logic),
- 3) Generator differential, and
- 4) Emergency stop.
- The operating band is meant as guidance to avoid routine overloading of the diesel.
Momentary transients outside the load range shall not invalidate the test.
- Diesel generator loadings may include gradual loading as recommended by the manufacturer.
- If Surveillance Requirement 4.8.1.1.2.a.5) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test. Instead, the diesel generator may be operated between 4800-5000 kW for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or until operating temperature has stabilized.
MILLSTONE- UNIT 3 3/4 8-8 Amendment No. W, 64, -l-+9, +94, 258
ELECTRICAL POWER SYSTEMS 3/4.8.2 D.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum, the following D.C. electrical sources shall be OPERABLE:
- a. 125-volt Battery Bank 301A-1, and an associated full capacity charger,
- b. 125-volt Battery Bank 301A-2, and an associated full capacity charger,
- c. 125-volt Battery Bank 301B-1 and an associated full capacity charger, and
- d. 125-volt Battery Bank 301B-2 and an associated full capacity charger.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
- a. With either Battery Bank 301A-1 or 301B-1, and/or one ofthe required full capacity chargers inoperable, restore the inoperable battery bank and/or full capacity charger to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- b. With either Battery Bank 301A-2 or 301B-2 inoperable, and/or one of the required full capacity chargers inoperable, restore the inoperable battery bank and/or full capacity charger to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.8.2.1 Each 125-volt battery bank and charger shall be demonstrated OPERABLE:
- a. At the frequency specified in the Surveillance Frequency Control Program by verifying that:
- 1) The parameters in Table 4.8-2a meet the Category A limits, and
- 2) The total battery terminal voltage is greater than or equal to 129 volts on float charge.
MILLSTONE - UNIT 3 3/4 8-11 Amendment No. 64, 258
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- b. At the frequency specified in the Surveillance Frequency Control Program and within 7 days after a battery discharge with battery terminal voltage below 11 0 volts, or battery overcharge with battery terminal voltage above 150 volts, by verifying that:
- 1) The parameters in Table 4.8-2a meet the Category B limits,
- 2) There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 1o-6 ohm, and
- 3) The average electrolyte temperature of six connected cells is above 60°F.
- c. At the frequency specified in the Surveillance Frequency Control Program by verifying that:
- 1) The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration,
- 2) The cell-to-cell and terminal connections are clean, tight, and coated with anticorrosion material,
- 3) The resistance of each cell-to-cell and terminal connection is less than or equal to 150 X 1o- 6 ohm, and
- 4) Each battery charger will supply at least the amperage indicated in Table 4.8-2b at greater than or equal to 132 volts for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- d. At the frequency specified in the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery service test;
- e. At the frequency specified in the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. This performance discharge test may be performed in lieu of the battery service test required by Specification 4.8.2.ld.; and
- f. At least once per 18 months, during shutdown, by giving performance discharge tests of battery capacity to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.
MILLSTONE - UNIT 3 3/4 8-12 Amendment No. 64, 19, -!-00, -l-49, 258
ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
- a. With one of the required trains of A. C. emergency busses not OPERABLE, restore the inoperable train to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- b. With one A. C. vital bus either not energized from its associated inverter, or with the inverter not connected to its associated D.C. bus: (1) reenergize the A.C. vital bus within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; and (2) reenergize the A.C.
vital bus from its associated inverter connected to its associated D.C. bus within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- c. With one D.C. bus not energized from its associated battery bank, reenergize the D.C. bus from its associated battery bank within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.8.3.1 The specified busses shall be determined OPERABLE in the specified manner at the frequency specified in the Surveillance Frequency Control Program by verifYing correct breaker alignment and indicated voltage on the busses.
MILLSTONE - UNIT 3 3/4 8-17 Amendment No. 64, ~'
258
ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION SHUTDOWN LIMITING CONDITION FOR OPERATION (Continued)
- 4) Two 125 volt DC Busses consisting of:
a) Bus #301B-1 energized from Battery Bank #301B-1, and b) Bus #301B-2 energized from Battery Bank #301B-2.
APPLICABILITY: MODES 5 and 6.
ACTION:
With any of the above required electrical busses not energized in the required manner, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity additions that could result in loss of required SDM or boron concentration, movement of recently irradiated fuel assemblies, crane operation with loads over the fuel storage pool, or operations with a potential for draining the reactor vessel, initiate corrective action to energize the required electrical busses in the specified manner as soon as possible.
SURVEILLANCE REQUIREMENTS 4.8.3.2 The specified busses shall be determined energized in the required manner at the frequency specified in the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated voltage on the busses.
MILLSTONE - UNIT 3 3/4 8-18a Amendment No. +46, ~' ~'
258
3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION
- 3. 9 .1.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained sufficient to ensure that the more restrictive of the following reactivity conditions is met; either:
- a. A Kerrof0.95 or less, or
- b. A boron concentration of greater than or equal to the limit specified in the CORE OPERATING LIMITS REPORT (COLR).
Additionally, the CVCS valves of Specification 4.1.1.2.2 shall be closed and secured in position.
APPLICABILITY: MODE 6.
- ACTION:
- a. With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS and positive reactivity additions and 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 its equivalent until Kerr is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to the limit specified in the COLR, whichever is the more restrictive.
- b. With any of the CVCS valves of Specification 4.1.1.2.2 not closed** and secured in position, immediately close and secure the valves.
SURVEILLANCE REQUIREMENTS 4.9.1.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:
- a. Removing or unbolting the reactor vessel head, and
- b. Withdrawal of any full-length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.
4.9.1.1.2 The boron concentration of the Reactor Coolant System and the refueling cavity shall be determined by chemical analysis at the frequency specified in the Surveillance Frequency Control Program.
4.9.1.1.3 The CVCS valves of Specification 4.1.1.2.2 shall be verified closed and locked at the frequency specified in the Surveillance Frequency Control Program.
- The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
- Except those opened under administrative control.
MILLSTONE - UNIT 3 3/4 9-1 Amendment No. 5G, @, 99, -l-1-3-, ~'
258 ~,~,
REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1.2 The soluble boron concentration of the Spent Fuel Pool shall be greater than or equal to 800 ppm.
APPLICABILITY:
Whenever fuel assemblies are in the spent fuel pool.
ACTION:
- a. With the boron concentration less than 800 ppm, initiate action to bring the boron concentration in the fuel pool to at least 800 ppm within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and
- b. With the boron concentration less than 800 ppm, suspend the movement of all fuel assemblies within the spent fuel pool and loads over the spent fuel racks.
SURVEILLANCE REQUIREMENTS 4.9.1.2 Verify that the boron concentration in the fuel pool is greater than or equal to 800 ppm at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE- UNIT 3 3/49-la Amendment No. H, §-8, -l-89, ~'
258
REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 Two Source Range Neutron Flux Monitors shall be OPERABLE with continuous visual indication in the control room, and one with audible indication in the containment and control room.
APPLICABILITY: MODE6.
ACTION:
- a. With one of the above required monitors inoperable immediately suspend all operations involving CORE ALTERATIONS and operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3. 9. 1.1.
- b. With both of the above required monitors inoperable determine the boron concentration of the Reactor Coolant System within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
SURVEILLANCE REQUIREMENTS 4.9.2 Each Source Range Neutron Flux Monitor shall be demonstrated OPERABLE by performance of:
- a. A CHANNEL CHECK and verification of audible counts at the frequency specified in the Surveillance Frequency Control Program,
- b. A CHANNEL CALIBRATION at the frequency specified in the Surveillance Frequency Control Program.*
- Neutron detectors are excluded from CHANNEL CALIBRATION.
MILLSTONE - UNIT 3 3/4 9-2 Amendment No. M-'7-, ~' ~'
258
REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:
- a. The equipment access hatch shall be either:
- 1. closed and held in place by a minimum of four bolts, or
- 2. open under administrative control
- and capable of being closed and held in place by a minimum of four bolts,
- b. A personnel access hatch shall be either:
- 1. closed by one personnel access hatch door, or
- 2. capable of being closed by an OPERABLE personnel access hatch door, under administrative control,* and
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
- 1. Closed by an isolation valve, blind flange, or manual valve, or
- 2. Be capable of being closed under administrative control.*
APPLICABILITY: During movement of fuel within the containment building.
ACTION:
With the requirements of the above specification not satisfied, immediately suspend all operations involving movement of fuel in the containment building.
SURVEILLANCE REQUIREMENTS 4.9.4.a Verify each required containment penetrations is in the required status at the frequency specified in the Surveillance Frequency Control Program.
4.9.4.b DELETED
- Administrative controls shall ensure that appropriate personnel are aware that the equipment access hatch penetration, personnel access hatch doors and/or other containment penetrations are open, and that a specific individual(s) is designated and available to close the equipment access hatch penetration, a personnel access hatch door and/or other containment penetrations within 30 minutes if a fuel handling accident occurs. Any obstructions (e.g. cables and hoses) that could prevent closure of the equipment access hatch penetration, a personnel access hatch door and/or other containment penetrations must be capable of being quickly removed.
MILLSTONE - UNIT 3 3/4 9-4 Amendment No. ~' ~'
258
REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation.*
APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet.
ACTION:
With no RHR loop OPERABLE or in operation, suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9 .1.1 and suspend loading irradiated fuel assemblies in the core and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.9.8.1 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 2800 gpm at the frequency specified in the Surveillance Frequency Control Program.
- The RHR loop may be removed from operation for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8-hour period, provided no operations are permitted that could cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.1.
MILLSTONE - UNIT 3 3/4 9-8 Amendment No. -t-e-1, ~.
258
REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent residual heat removal (RHR) loops shall be OPERABLE, and at least one RHR loop shall be in operation.*
APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is less than 23 feet.
ACTION:
- a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status, or to establish greater than or equal to 23 feet of water above the reactor vessel flange, as soon as possible.
- b. With no RHR loop in operation, suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.1 and immediately initiate corrective action to return the required RHR loop to operation. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.9.8.2 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 2800 gpm at the frequency specified in the Surveillance Frequency Control Program.
- The RHR loop may be removed from operation for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8-hour period, provided no operations are permitted that could cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9 .1.1.
MILLSTONE- UNIT 3 3/4 9-9 Amendment No. -l-0-1, ~.
258
REFUELING OPERATIONS 3/4.9.10 WATER LEVEL- REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.10 At least 23 feet of water shall be maintained over the top of the reactor vessel flange.
APPLICABILITY: During movement of fuel assemblies or control rods within the containment when either the fuel assemblies being moved or the fuel assemblies seated within the reactor vessel are irradiated while in MODE 6.
ACTION:
With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or control rods within the reactor vessel.
SURVEILLANCE REQUIREMENTS 4.9.10 The water level shall be determined to be at least its minimum required depth at the frequency specified in the Surveillance Frequency Control Program.
MILLSTONE - UNIT 3 3/4 9-11 Amendment No. ~'
258
REFUELING OPERATIONS 3/4.9.11 WATER LEVEL- STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool.
ACTION:
- a. With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at the frequency specified in the Surveillance Frequency Control Program when irradiated fuel assemblies are in the fuel storage pool.
MILLSTONE - UNIT 3 3/49-12 Amendment No. -s::J.,
258
SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.1 0.2.1 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1.1, and 3.2.4 may be suspended during the performance ofPHYSICS TESTS provided:
- a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and
- b. The limits of Specifications 3.2.2.1 and 3.2.3.1 are maintained and determined at the frequencies specified in Specification 4.1 0.2.1.2 below.
APPLICABILITY: MODE 1.
ACTION:
With any of the limits of Specification 3.2.2.1 or 3.2.3.1 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1.1, and 3.2.4 are suspended, either:
- a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2.1 and 3.2.3.1, or
- b. Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.10.2.1.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at the frequency specified in the Surveillance Frequency Control Program during PHYSICS TESTS.
4.1 0.2.1.2 The Surveillance Requirements of the below listed specifications shall be performed at the frequency specified in the Surveillance Frequency Control Program during PHYSICS TESTS:
- a. Specifications 4.2.2.1.2 and 4.2.2.1.3, and
- b. Specification 4.2.3.1.2.
MILLSTONE - UNIT 3 3/4 10-2 Amendment No.~.
258
SPECIAL TEST EXCEPTIONS 3/4.10.3 PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5, and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:
- a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,
- b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set at less than or equal to 25% of RATED THERMAL POWER, and
- c. The Reactor Coolant System lowest operating loop temperature (Tavg) is greater than or equal to 541 °F.
APPLICABILITY: MODE 2.
ACTION:
- a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the Reactor trip breakers.
- b. With a Reactor Coolant System operating loop temperature (Tavg) less than 541 °F, restore Tavg to within its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutes.
SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at the frequency specified in the Surveillance Frequency Control Program during PHYSICS TESTS.
4.1 0.3.2 Each Intermediate and Power Range channel shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to initiating PHYSICS TESTS.
4.10.3.3 The Reactor Coolant System temperature (Tavg) shall be determined to be greater than or equal to 541 °F at the frequency specified in the Surveillance Frequency Control Program during PHYSICS TESTS.
MILLSTONE- UNIT 3 3/4 10-4 Amendment No. 258
SPECIAL TEST EXCEPTIONS 3/4.10.4 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.1 0.4 The limitations of Specification 3.4.1.1 may be suspended during the performance of STARTUP and PHYSICS TESTS provided:
- a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and
- b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set less than or equal to 25% of RATED THERMAL POWER.
APPLICABILITY: During operation below the P-7 Interlock Setpoint.
ACTION:
With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the Reactor trip breakers.
SURVEILLANCE REQUIREMENTS 4.1 0.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoint at the frequency specified in the Surveillance Frequency Control Program during STARTUP and PHYSICS TESTS.
4.10.4.2 Each Intermediate and Power Range channel, and P-7 Interlock shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to initiating STARTUP and PHYSICS TESTS.
MILLSTONE - UNIT 3 3/4 10-5 Amendment No. 258
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
- c. The program shall, as allowed by 10 CFR 50.55a(b)(3)(v), meet Subsection ISTA, "General Requirements" and Subsection ISTD, "Preservice and lnservice Examination and Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear Power Plants" in lieu of Section XI ofthe ASME BPV Code lSI requirements for snubbers, or meet authorized alternatives pursuant to 10 CFR 50.55a(a)(3).
- d. The 120-month program updates shall be made in accordance with 10 CFR 50.55a (including 10 CFR 50.55a(b)(3)(v)) subject to the limitations and modifications listed therein.
J. Surveillance Frequency Control Program This program provides controls for surveillance frequencies. The program shall ensure that surveillance requirements specified in the technical specification are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
- a. The Surveillance Frequency Control Program shall contain a list of frequencies of those surveillance requirements for which the frequency is controlled by the program.
- b. Changes to the frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.
- c. The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the frequencies established in the Surveillance Frequency Control Program.
6.8.5 Written procedures shall be established, implemented and maintained covering Section I.E, Radiological Environmental Monitoring, ofthe REMODCM.
6.8.6 All procedures and procedure changes required for the Radiological Environmental Monitoring Program (REMP) of Specification 6.8.5 above shall be reviewed by an individual (other than the author) from the organization responsible for the REMP and approved by appropriate supervision.
Temporary changes may be made provided the intent of the original procedure is not altered and the change is documented and reviewed by an individual (other than the author) from the organization responsible for the REMP, within 14 days of implementation.
MILLSTONE- UNIT 3 6-17g Amendment No.~
258
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 258 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-49 DOMINION NUCLEAR CONNECTICUT. INC.
MILLSTONE POWER STATION, UNIT NO. 3 DOCKET NO. 50-423
1.0 INTRODUCTION
By letter dated October 4, 2012 (Agencywide Document Access and Management System (ADAMS) Accession No. ML12284A213), as supplemented by letters dated January 4, 2013 (ADAMS Accession No. ML13008A328), April17, 2013 (ADAMS Accession No. ML13113A344), and October 30, 2013 (ADAMS Accession No. ML13309A752), Dominion Nuclear Connecticut, Inc. (the licensee) proposed changes to the Technical Specifications (TSs) for Millstone Power Station, Unit No. 3 (MPS3). The supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on December 11, 2012 (77 FR 73687).
The requested change is the adoption the of NRC-approved Technical Specifications Task Force (TSTF) traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Controi-[Risk-lnformed Technical Specification Task Force (RITSTF)] Initiative 5b" (Reference 1). When implemented, TSTF-425 relocates most periodic frequencies of TS surveillances to a licensee controlled program, the Surveillance Frequency Control Program (SFCP), and provides requirements for the new program in the Administrative Controls section of the TS. All surveillance frequencies are proposed to be relocated except:
Frequencies that reference other approved programs for the specific interval (such as the In-Service Testing Program or the Primary Containment Leakage Rate Testing Program);
Frequencies that are purely event-driven (e.g., "each time the control rod is withdrawn to the 'full out' position");
Frequencies that are event-driven, but have a time component for performing the surveillance on a one-time basis once the event occurs (e.g., "within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after thermal power reaching ;:: 95% [rated thermal power RTP] "); and Frequencies that are related to specific conditions (e.g., battery degradation, age, and capacity) or conditions for the performance of a surveillance requirement (e.g., "drywell to suppression chamber differential pressure decrease").
The requested change includes a new program to the TS Section 6, Administrative Controls as Specification 6.8.4. The new program is called the SFCP and describes the requirements for the program to control changes to the relocated surveillance frequencies. The TS Bases for each of the affected surveillance requirements are revised to state that the frequency is set in accordance with the SFCP. Some surveillance requirements TS Bases do not contain a discussion of the frequency. In these cases, the TS Bases describing the current frequency were added to maintain consistency with the TS Bases for similar surveillances. These instances are noted in the markup along with the source of the text. The proposed changes to TS Section 6, Administrative Controls, to incorporate the SFCP include a specific reference to Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," Revision 1 (Reference 2) as the basis for making any changes to the surveillance frequencies once they are relocated out of the TS.
In a letter dated September 19, 2007 (Reference 3), the NRC staff approved NEI 04-10, Revision 1, as acceptable for referencing in licensing actions to the extent specified and under the limitations delineated in NEI 04-10, and the safety evaluation providing the basis for NRC acceptance of NEI 04-10.
2.0 REGULATORY EVALUATION
In the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" published in the Federal Register (58 FR 39132, July 22, 1993), the NRC addressed the use of Probabilistic Safety Analysis (PSA, currently referred to as Probabilistic Risk Assessment or PRA) in Standard Technical Specifications. In discussing the use of PSA in Nuclear Power Plant TSs, the Commission wrote in part:
The Commission believes that it would be inappropriate at this time to allow requirements which meet one or more of the first three criteria [of 10 CFR 50.36]
to be deleted from Technical Specifications based solely on PSA (Criterion 4).
However, if the results of PSA indicate that Technical Specifications can be relaxed or removed, a deterministic review will be performed ....
The Commission Policy in this regard is consistent with its Policy Statement on "Safety Goals for the Operation of Nuclear Power Plants," 51 FR 30028, published on August 21, 1986. The Policy Statement on Safety Goals states in part, "* *
- probabilistic results should also be reasonably balanced and supported through use of deterministic arguments. In this way, judgments can be made * *
- about the degree of confidence to be given these (probabilistic) estimates and assumptions. This is a key part of the process of determining the degree of regulatory conservatism that may be warranted for particular decisions. This defense-in-depth approach is expected to continue to ensure the protection of public health and safety." ...
The Commission will continue to use PSA, consistent with its policy on Safety Goals, as a tool in evaluating specific line item improvements to Technical Specifications, new requirements, and industry proposals for risk-based Technical Specification changes.
Approximately two years later the NRC provided additional detail concerning the use of PRA in the "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities; Final Policy
Statement" published in the Federal Register (60 FR 42622, August 16, 1995). The Commission, in discussing the deterministic and probabilistic approach to regulation, and the Commission's extension and enhancement of traditional regulation, wrote in part:
PRA addresses a broad spectrum of initiating events by assessing the event frequency. Mitigating system reliability is then assessed, including the potential for multiple and common cause failures. The treatment therefore goes beyond the single failure requirements in the deterministic approach. The probabilistic approach to regulation is, therefore, considered an extension and enhancement of traditional regulation by considering risk in a more coherent and complete manner.
The commission provided its new policy, stating:
Although PRA methods and information have thus far been used successfully in nuclear regulatory activities, there have been concerns that PRA methods are not consistently applied throughout the agency, that sufficient agency PRA/statistics expertise is not available, and that the Commission is not deriving full benefit from the large agency and industry investment in the developed risk assessment methods. Therefore, the Commission believes that an overall policy on the use of PRA in nuclear regulatory activities should be established so that the many potential applications of PRA can be implemented in a consistent and predictable manner that promotes regulatory stability and efficiency. This policy statement sets forth the Commission's intention to encourage the use of PRA and to expand the scope of PRA applications in all nuclear regulatory matters to the extent supported by the state-of-the-art in terms of methods and data.
Implementation of the policy statement will improve the regulatory process in three areas: Foremost, through safety decision making enhanced by the use of PRA insights; through more efficient use of agency resources; and through a reduction in unnecessary burdens on licensees.
Therefore, the Commission adopts the following policy statement regarding the expanded NRC use of PRA:
( 1) The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy.
(2) PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state-of-the-art, to reduce unnecessary conservatism associated with current regulatory requirements, regulatory guides, license commitments, and staff practices. Where appropriate, PRA should be used to support the proposal for additional regulatory requirements in accordance with 10 CFR 50.109 (Backfit Rule). Appropriate procedures for including PRA in the process for changing regulatory requirements should be developed and followed. It is, of course, understood that the intent of this policy is that existing rules and regulations shall be complied with unless these rules and regulations are revised.
(3) PRA evaluations in support of regulatory decisions should be as realistic as practicable and appropriate supporting data should be publicly available for review.
(4) The Commission's safety goals for nuclear power plants and subsidiary numerical objectives are to be used with appropriate consideration of uncertainties in making regulatory judgments on the need for proposing and backfitting new generic requirements on nuclear power plant licensees.
In Title 10 of the Code of Federal Regulations (1 0 CFR) 50.36, the NRC established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation:
(1) Safety limits, limiting safety system settings, and limiting control settings; (2) Limiting conditions for operation; (3) Surveillance requirements; (4) Design features; and (5)
Administrative controls.
As stated in 10 CFR 50.36(c)(3), "Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." These categories will remain in the TSs following the proposed amendment. The new TS SFCP provides the necessary administrative controls to require that surveillances relocated to the SFCP are conducted at a frequency to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Changes to surveillance frequencies in the SFCP will be made using the methodology contained in NEI 04-10, including qualitative considerations, results of risk analyses, sensitivity studies and any bounding analyses, and recommended monitoring of structures, systems, and components (SSCs), and documentation requirements. Furthermore, changes to frequencies are subject to regulatory review and oversight of the SFCP implementation through the rigorous NRC review of safety-related sse performance provided by the reactor oversight program.
Licensees are required by TSs to perform surveillance test, calibration, or inspection on specific safety-related system equipment (e.g., reactivity control, power distribution, electrical, and instrumentation) to verify system operability. Surveillance frequencies, currently identified in TSs, are based primarily upon deterministic methods such as engineering judgment, operating experience, and manufacturer's recommendations. The licensee's use of NRC-approved methodologies identified in NEI 04-10 provides a way to establish risk-informed surveillance frequencies that complement the deterministic approach and support the NRC's traditional defense-in-depth philosophy.
The licensee's SFCP ensures that surveillance requirements specified in the TS are performed at intervals sufficient to assure the above regulatory requirements are met. Existing regulatory requirements, such as 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," and 10 CFR 50 Appendix B (corrective action program),
require licensee monitoring of surveillance test failures and implementing corrective actions to address such failures. One of these actions may be to consider increasing the frequency at which a surveillance test is performed. In addition, the SFCP implementation guidance in NEI 04-10 requires monitoring the performance of SSCs for which surveillance frequencies are decreased to assure reduced testing does not adversely impact the SSCs. These requirements, and the monitoring required by NEI 04-10, ensure that surveillance frequencies
are sufficient to assure that the requirements of 10 CFR 50.36 are satisfied and that any performance deficiencies will be identified and appropriate corrective actions will be taken.
Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," (Reference 4),
describes a risk-informed approach, acceptable to the NRC, for assessing the nature and impact of proposed permanent licensing-basis changes by considering engineering issues and applying risk insights. This regulatory guide also provides risk acceptance guidelines for evaluating the results of such evaluations.
RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," (Reference 5), describes an acceptable risk-informed approach specifically for assessing proposed TS changes.
RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," (Reference 6), describes an acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision making for light water-reactors.
3.0 TECHNICAL EVALUATION
The licensee's proposed adoption of TSTF-425 for MPS3 provides for administrative relocation of applicable surveillance frequencies, and provides for the addition of the SFCP to TS Section 6, Administrative Controls. TSTF-425 also requires the application of NEI 04-10 for any changes to surveillance frequencies within the SFCP. The licensee's application for the changes proposed in TSTF-425 included documentation regarding the PRA technical adequacy consistent with the requirements of RG 1.200. In accordance with NEI 04-10, PRA methods are used, in combination with plant performance data and other considerations, to identify and justify modifications to the surveillance frequencies of equipment at nuclear power plants. This is in accordance with guidance provided in RG 1.174 and RG 1.177 in support of changes to surveillance test intervals.
3.1 RG 1.177 Five Key Safety Principles RG 1.177 identifies five key safety principles required for risk-informed changes toTS. Each of these principles is addressed by the industry methodology document, NEI 04-10.
3.1.1 The Proposed Change Meets Current Regulations The regulatory requirement of 10 CFR 50.36(c)(3) states that TSs will include surveillances which are "requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." NEI 04-10 provides guidance for relocating the surveillance frequencies from the TSs to a licensee-controlled program by providing an NRC-approved methodology for control of the surveillance frequencies. The surveillances themselves would remain in the TSs, as required by 10 CFR 50.36(c)(3).
This change is consistent with other NRC-approved TS changes in which surveillance frequencies have been relocated to licensee-controlled documents, such as surveillances performed in accordance with the In-service Testing Program or the Primary Containment Leakage Rate Testing Program. Thus, this proposed change satisfies the first key safety principle of RG 1.177 by complying with current regulations.
3.1.2 The Proposed Change Is Consistent With the Defense-in-Depth Philosophy The defense-in-depth philosophy, the second key safety principle of RG 1.177, is maintained if:
- A reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation.
- Over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided.
- System redundancy, independence, and diversity are preserved commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g.,
no risk outliers). Because the scope of the proposed methodology is limited to the revision of surveillance frequencies, the redundancy, independence, and diversity of plant systems are not impacted.
- Defenses against potential common cause failures are preserved, and the potential for the introduction of new common cause failure mechanisms is assessed.
- Independence of physical barriers is not degraded.
- Defenses against human errors are preserved.
- The intent of the General Design Criteria in 10 CFR Part 50, Appendix A, is maintained.
TSTF-425 requires the application of NEI 04-10 for any changes to surveillance frequencies within the SFCP. NEI 04-10 uses both the core damage frequency (CDF) and the large early release frequency (LERF) metrics to evaluate the impact of proposed changes to surveillance frequencies. The guidance of RG 1.174 and RG 1.177 for changes to CDF and LERF is achieved by evaluation using a comprehensive risk analysis, which assesses the impact of proposed changes including contributions from human errors and common cause failures.
Defense-in-depth is also included in the methodology explicitly as a qualitative consideration outside of the risk analysis, as is the potential impact on detection of component degradation that could lead to an increased likelihood of common cause failures. Both the quantitative risk analysis and the qualitative considerations assure that a reasonable balance of defense-in-depth is maintained to ensure protection of public health and safety, satisfying the second key safety principle of RG 1.177.
3.1.3 The Proposed Change Maintains Sufficient Safety Margins The engineering evaluation that will be conducted by the licensee under the SFCP when frequencies are revised will assess the impact of the proposed frequency change to assure that sufficient safety margins are maintained. The guidelines used for making that assessment will include ensuring that the proposed surveillance test frequency change is not in conflict with approved industry codes and standards or adversely affects any assumptions or inputs to the safety analysis, or, if such inputs are affected, justification is provided to ensure sufficient safety margin will continue to exist.
The design, operation, testing methods, and acceptance criteria for SSCs, specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the Updated Final Safety Analysis Report and TS Bases), since these are not affected by changes to the surveillance frequencies.
Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis.
Thus, safety margins are maintained by the proposed methodology, and the third key safety principle of RG 1.177 is satisfied.
3.1.4 When Proposed Changes Result in an Increase in Core Damage Frequency or Risk.
the Increases Should Be Small and Consistent With the Intent of the Commission's Safety Goal Policy Statement RG 1.177 provides a framework for evaluating the risk impact of proposed changes to surveillance frequencies. This requires the identification of the risk contribution from impacted surveillances, determination of the risk impact from the change to the proposed surveillance frequency, and performance of sensitivity and uncertainty evaluations. TSTF-425 requires application of NEI 04-10 in the SFCP. NEI 04-10 satisfies the intent of the RG 1.177 safety principle for evaluating the change in risk, and for assuring that such changes are small.
3.1.4.1 Quality of the PRA The quality of the MPS3 PRA must be compatible with the safety implications of the proposed TS change and the role the PRA plays in justifying the change. That is, the more the potential change in risk or the greater the uncertainty in that risk from the requested TS change, or both, the more rigor that must go into ensuring the quality of the PRA.
RG 1.200 is NRC's developed regulatory guidance for assessing the technical adequacy of a PRA. Revision 2 of this RG (Reference 7) endorses (with comments and qualifications) the use of the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS)
RA-Sa-2009, "Addenda to ASME RA-S-2008 Standard for Levei1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," (Reference 8),
NEI 00-02, "PRA Peer Review Process Guidelines," (Reference 9) and NEI 05-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard" (Reference 10).
Revision 1 of this RG (Reference 6) had endorsed the internal events PRA standard ASME RA-Sb-2005, "Addenda to ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," (Reference 11 ). For the internal events PRA, there are no
significant technical differences in the standard requirements, and therefore assessments using the previously endorsed internal events standard are acceptable.
The licensee has performed an assessment of the PRA models used to support the SFCP using the guidance of RG 1.200 to assure that the PRA models are capable of determining the change in risk due to changes to surveillance frequencies of SSCs, using plant-specific data and models. Capability category II of the standard is required by NEI 04-10 for the internal events PRA, and any identified deficiencies to those requirements are assessed further to determine any impacts to proposed decreases to surveillance frequencies, including by the use of sensitivity studies where appropriate.
A Westinghouse Owners Group (WOG) peer review of the MPS3 internal events PRA was performed in 1999. All findings and observations (F&Os) from this peer review have been addressed. In addition, the licensee performed a self-assessment of the MPS3 internal events PRA in 2007, using the American Society of Mechanical Engineers (ASME) PRA Standard, ASME RA-Sb-2005, and the guidance in RG 1.200, Revision 1. The internal events PRA model (M31 OA) addressed F&Os from the self-assessment, and included model upgrades. Science Applications International Corporation performed a focused scope peer review in June 2012, considering clarifications and qualifications in Regulatory Guide (RG) 1.200, Revision 2 for the model upgrades against the ASME/ANS RA-Sa-2009 PRA standard. The licensee addressed the "gaps" between the internal events PRA model and the PRA standard from the self-assessment and the focused scope peer review, and provided them in Table 3 of the license amendment request (LAR). The staff's evaluation of these F&Os is given below.
Gap #1 for Supporting Requirement IE-A8. The F&O is related to the lack of documentation of interviews with plant personnel to determine if potential initiating events have been overlooked.
The licensee notes that informal discussions between plant personnel and PRA engineers had been done. In response to an NRC request for additional information (RAI) dated January 4, 2013, the licensee noted that documentation of interviews was not retained; however, a recent interview between the PRA staff and Operations staff validated that potential initiating events have been identified. The licensee stated that this interview satisfies the requirements of SR IE-AS. The licensee has also developed a process to document discussions with plant personnel going forward. The staff concludes that, based on this information, the licensee has dispositioned this F&O for the application.
Gap #2 for Supporting Requirement SY-A4. The licensee stated that this F&O was a documentation issue only. In response to an RAI dated April17, 2013, the licensee noted several systems for which walkdowns or interviews had been performed to confirm that the system analysis correctly reflects the as-built, as operated plant. The staff concludes that, based on this information, the licensee has dispositioned this F&O for the application.
Gap #3 for Supporting Requirement HR-G5. This F&O is that existing human error analysis documentation for talk-throughs with Operations staff is outdated, and no new operator survey information is provided to support the basis for revised or new human factor events (HFEs). In response to an RAI dated April 17, 2013, the licensee performed operator walkthrough/talk-throughs for both significant and non-significant HFEs. The final analysis resulted in no impact to the human error probabilities (HEPs) within the PRA model. The staff concludes that, based on this information, the licensee has dispositioned this F&O for the application.
Gap #4 for Supporting Requirement HR-G7. This F&O identified several numerical inconsistencies in the dependency analysis supporting the HEPs. In response to an RAI dated January 4, 2013, the licensee showed through a sensitivity analysis that updating the numerical inconsistencies results in a negligible impact on the overall risk (CDF and LERF) and the individual HEP importance measures. The licensee stated that until these inconsistencies have been incorporated in the internal events PRA model, a sensitivity study will be performed by increasing the HEPs by a factor of 10. The staff concludes, based on this information, that this F&O can be addressed by the methodology of NEI 04-10.
Gap #5 for Supporting Requirement IFPP-82. This F&O is related to flood propagation paths associated with the condensate polishing facility. The F&O notes that potential water propagation is bounded by the amount of water generated during a circulating water pipe break.
The licensee will perform a sensitivity analysis for this F&O by doubling the circulating water initiating event frequency. The staff concludes, based on this information, that this F&O can be addressed by the methodology of NEI 04-10.
Gap #6 for IFSO-A4. This F&O identified a need to incorporate internal flooding frequencies associated with non-piping failures (e.g., expansion joints, bellows, overfill, and inadvertent sprinkler actuation). The LAR notes that those non-piping failures not included in the internal flooding analysis are bounded by already analyzed flow rates. The licensee will perform a sensitivity analysis for this F&O by doubling the internal flooding event frequencies. The staff concludes, based on this information, that this F&O can be addressed by the methodology of NEI 04-10.
Gap #7 for IE-C 12. This F&O is to perform a reasonableness check of the expansion joint rupture frequencies modeled in the PRA. This is similar to the F&O identified in Gap #6, and the staff concludes that this F&O can also be addressed by the methodology of NEI 04-10.
Gap #8 for IFSN-A3. This F&O is a documentation issue only between internal flooding PRA notebooks discussing internal flooding operator actions, and, therefore has no impact on the risk calculations for this application.
Gap #9 for IFSN-A8. This F&O is related to an internal flooding notebook not considering the potential for flood barrier unavailability and flood pathways through floor drain check valves. In response to an RAI dated April 17, 2013, the licensee reviewed the inter-area propagation paths and did not identify any new flooding events. In addition, the licensee stated that revision of documentation addresses the focused peer review's findings and no update to the PRA model is planned or necessary. The staff concludes that, based on this information, the licensee has dispositioned this F&O for the application.
Gap #1 0 for IFEV-A5. This F&O notes that the internal flooding initiating events are in units of "per calendar year," which is conservative in that it has not taken into account the capacity factor. The staff concludes, based on this information, that this F&O can be addressed by the methodology of NEI 04-10.
Gap #11 for IFEV-A7. The F&O identified a need to revise the flooding analysis to document the process used to identify human-induced flooding scenarios. In response to an RAI dated January 4, 2013, the licensee described the process. The process included an assumption that only equipment within the area that may be affected by spray or jet impingement damage are
assumed to fail, and there is no propagation to other areas and no damage due to submersion.
In response to a follow-up RAI dated April 17, 2013, to provide justification for this assumption, the licensee re-evaluated flooding zones to augment the modeling of human-induced flooding events during maintenance activities. The review identified new maintenance-induced flooding events which could potentially occur. When these new events were added to the PRA model, the licensee found the impact on CDF and LERF to be negligible. The staff concludes, based on this information, that this F&O can be addressed by the methodology of NEI 04-10.
The staff also performed an audit on January 29 and 30, 2013 of closed F&Os for the internal events PRA model, and found that they had been appropriately addressed.
Based on the licensee's assessment using the applicable PRA standard and RG 1.200, the level of PRA quality, combined with the proposed evaluation and disposition of gaps, is sufficient to support the evaluation of changes proposed to surveillance frequencies within the SFCP, and is consistent with regulatory position 2.3.1 of RG 1.177.
3.1.4.2 Scope of the PRA The licensee is required to evaluate each proposed change to a relocated surveillance frequency using the guidance contained in NEI 04-10 to determine its potential impact on risk, due to impacts from internal events, fires, seismic, other external events, and from shutdown conditions. In cases where a PRA of sufficient scope or quantitative risk models were unavailable, the licensee uses bounding analyses, or other conservative quantitative evaluations. A qualitative screening analysis may be used when the surveillance frequency impact on plant risk is shown to be negligible or zero.
MPS3 has an internal events PRA model which has received a peer review, self assessment, and a focused scope peer review as discussed previously.
MPS3 does not have a PRA model for internal fire events or for external events as described in a follow-up RAI dated April17, 2013. In accordance with NEI 04-10, Revision 1 the licensee will perform an initial qualitative screening analysis, and if the qualitative information is not sufficient a bounding analysis will be performed. The licensee provided additional information on their bounding analysis approach in response to a follow-up RAI dated October 30, 2013. The bounding analysis will be performed in accordance with NEI 04-10, Rev. 1, Step 1Ob, and it will be based on risk insights and analysis documented in the MPS3 Individual Plant Examination of External Events (IPEEE) report with consideration of the IPEEE accident sequences, as well as relevant operating experience and additional risk insights obtained since the IPEEE study, in the context of the current plant configuration and operation. Therefore, the NRC staff finds this approach to be consistent with NEI 04-10, Step 1Ob guidance in performing a bounding analysis.
For shutdown events, according to the response to an RAI dated January 4, 2013, the licensee will qualitatively assess changes in the surveillance frequencies using guidance from NEI 04-10, Rev. 1.
The licensee's evaluation methodology is sufficient to ensure that the scope of the risk contribution of each surveillance frequency change is properly identified for evaluation and, thus, is consistent with regulatory position 2.3.2 of RG 1.177.
3.1.4.3 PRA Modeling The licensee's methodology includes the determination of whether the SSCs affected by a proposed change to a surveillance frequency are modeled in the PRA. Where the SSC is directly or implicitly modeled, a quantitative evaluation of the risk impact may be carried out.
The methodology adjusts the failure probability of the impacted SSCs, including any impacted common cause failure modes, based on the proposed change to the surveillance frequency.
Where the sse is not modeled in the PRA, bounding analyses are performed to characterize the impact of the proposed change to the surveillance frequency. Potential impacts on the risk analyses due to screening criteria and truncation levels are addressed by the requirements for PRA technical adequacy consistent with guidance contained in RG 1.200, and by sensitivity studies identified in NEI 04-10.
The licensee's approach for the evaluations of the impact of selected testing strategy (i.e., staggered testing or sequential testing) is consistent with the guidance of NUREG/CR-6141 and NUREG/CR-5497, as discussed in NEI 04-10.
Thus, through the application of NEI 04-10, the MPS3 PRA modeling is sufficient to ensure an acceptable evaluation of risk for the proposed changes in surveillance frequency, and is consistent with regulatory position 2.3.3 of RG 1.177.
3.1.4.4 Assumptions for Time Related Failure Contributions The failure probabilities of SSCs modeled in PRAs may include a standby time-related contribution and a cyclic demand-related contribution. NEI 04-10 criteria adjust the time-related failure contribution of SSCs affected by the proposed change to surveillance frequency. This is consistent with RG 1.177 Section 2.3.3 which permits separation of the failure rate contributions into demand and standby for evaluation of surveillance requirements. If the available data do not support distinguishing between the time-related failures and demand failures, then the change to surveillance frequency is conservatively assumed to impact the total failure probability of the SSC, including both standby and demand contributions. The SSC failure rate (per unit time) is assumed to be unaffected by the change in test frequency, and will be confirmed by the required monitoring and feedback implemented after the change in surveillance frequency is implemented. The process requires consideration of qualitative sources of information with regards to potential impacts of test frequency on sse performance, including industry and plant-specific operating experience, vendor recommendations, industry standards, and code-specified test intervals. Thus the process is not reliant upon risk analyses as the sole basis for the proposed changes.
The potential benefits of reduced surveillance frequency, including reduced downtime, lesser potential for restoration errors, reduction of potential for test caused transients, and reduced test-caused wear of equipment, are identified qualitatively, but not quantitatively assessed.
Thus, through the application of NEI 04-10, the licensee has employed reasonable assumptions with regard to extensions of surveillance test intervals, and is consistent with regulatory position 2.3.4 of RG 1.177.
3.1.4.5 Sensitivity and Uncertainty Analyses NEI 04-10 requires sensitivity studies to assess the impact of uncertainties from key assumptions of the PRA, uncertainty in the failure probabilities of the affected SSCs, impact on the frequency of initiating events, and any identified deviations from capability category II of the PRA standard. Where the sensitivity analyses identify a potential impact on the proposed change, revised surveillance frequencies are considered, along with any qualitative considerations that may bear on the results of such sensitivity studies. Required monitoring and feedback of sse performance once the revised surveillance frequencies are implemented will also be performed. Thus, through the application of NEI 04-10, the licensee has appropriately considered the possible impact of PRA model uncertainty and sensitivity to key assumptions and model limitations, and is consistent with regulatory position 2.3.5 of RG 1.177.
3.1.4.6 Acceptance Guidelines The licensee will quantitatively evaluate the change in total risk (including internal and external events contributions) in terms of CDF and LERF for both the individual risk impact of a proposed change in surveillance frequency and the cumulative impact from all individual changes to surveillance frequencies using the guidance contained in NRC approved NEI 04-10 in accordance with the TS SFCP. Each individual change to surveillance frequency must show a risk impact below 1E-6 per year for change to CDF, and below 1E-7 per year for change to LERF. These are consistent with the acceptance criteria of RG 1.174 for very small changes in risk. Where the RG 1.174 acceptance criteria are not met, the process either considers revised surveillance frequencies which are consistent with RG 1.174 or the process terminates without permitting the proposed changes. Where quantitative results are unavailable for comparison with the acceptance guidelines, appropriate qualitative analyses are required to demonstrate that the associated risk impact of a proposed change to surveillance frequency is negligible.
Otherwise, bounding quantitative analyses are required which demonstrate the risk impact is at least one order of magnitude lower than the RG 1.174 acceptance guidelines for very small changes in risk. In addition to assessing each individual sse surveillance frequency change, the cumulative impact of all changes must result in a risk impact less than 1E-5 per year for change to CDF, and less than 1E-6 per year for change to LERF, and the total CDF and total LERF must be reasonably shown to be less than 1E-4 per year and 1E-5 per year, respectively.
These are consistent with the acceptance criteria of RG 1.174, as referenced by RG 1.177 for changes to surveillance frequencies. The staff further notes that the licensee includes a provision to exclude the contribution to cumulative risk from individual changes to surveillance frequencies associated with insignificant risk increases (less than 5E-8 CDF and 5E-9 LERF) once the baseline PRA models are updated to include the effects of the revised surveillance frequencies.
The quantitative acceptance guidance of RG 1.174 is supplemented by qualitative information to evaluate the proposed changes to surveillance frequencies, including industry and plant-specific operating experience, vendor recommendations, industry standards, the results of sensitivity studies, and sse performance data and test history.
The final acceptability of the proposed change is based on all of these considerations and not solely on the PRA results. Post implementation performance monitoring and feedback are also required to assure continued reliability of the SSCs. The licensee's application of NEI 04-10 provides acceptable methods for evaluating the risk increase associated with proposed changes to surveillance frequencies, consistent with Regulatory Position 2.4 of RG 1.177. Therefore, the
proposed methodology satisfies the fourth key safety principle of RG 1.177 by assuring that any increase in risk is small consistent with the intent of the Commission's Safety Goal Policy Statement.
3.1.5 The Impact of the Proposed Change Should Be Monitored Using Performance Measurement Strategies The licensee's adoption of TSTF-425 requires application of NEI 04-10 in the SFCP. NEI 04-10 requires performance monitoring of SSCs whose surveillance frequency has been revised as part of a feedback process to assure that the change in test frequency has not resulted in degradation of equipment performance and operational safety. The monitoring and feedback includes consideration of maintenance rule monitoring of equipment performance. In the event of degradation of SSC performance, the surveillance frequency will be reassessed in accordance with the methodology, in addition to any corrective actions which may apply as part of the maintenance rule requirements. The performance monitoring and feedback specified in NEI 04-10 is sufficient to reasonably assure acceptable sse performance and is consistent with regulatory position 3.2 of RG 1.177. Thus, the fifth key safety principle of RG 1.177 is satisfied.
3.2 Addition of Surveillance Frequency Control Program to Administrative Controls The licensee proposes to include the SFCP and specific requirements into TS Section 6, Administrative Controls, at Section 6.8.4, Surveillance Frequency Control Program, as follows:
This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure that the associated Limiting Conditions for Operation are met.
- a. The Surveillance Frequency Control Program shall contain a list of Frequencies of the Surveillance Requirements for which the Frequency is controlled by the program.
- b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.
- c. The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.
The proposed program is consistent with the model application of TSTF-425 previously approved by the NRC, and is therefore acceptable.
3.3 Summarv and Conclusions The staff has reviewed the licensee's proposed relocation of some surveillance frequencies to a licensee controlled document, and controlling changes to surveillance frequencies in accordance with a new program, the SFCP, identified in the administrative controls of the TS.
The SFCP and TS Section 6.8.4 reference NEI 04-10, which provides a risk-informed methodology using plant-specific risk insights and performance data to revise surveillance frequencies within the SFCP. This methodology supports relocating surveillance frequencies from TS to a licensee-controlled document, provided those frequencies are changed in accordance with NEI 04-10 which is specified in the Administrative Controls of the TS.
The proposed licensee adoption of TSTF-425 and risk-informed methodology of NEI 04-10 as referenced in the Administrative Controls of TS, satisfies the key principles of risk-informed decision making applied to changes to TS as delineated in RG 1.177 and RG 1.174, in that:
- The proposed change meets current regulations;
- The proposed change is consistent with defense-in-depth philosophy;
- The proposed change maintains sufficient safety margins;
- Increases in risk resulting from the proposed change are small and consistent with the intent of the Commission's Safety Goal Policy Statement; and
- The impact of the proposed change is monitored using performance measurement strategies.
10 CFR 50.36(c)(3) states "Technical specifications will include items in the following categories: Surveillance Requirements. Surveillance Requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." Based on the above evaluation, the NRC staff finds that, with the proposed relocation of surveillance frequencies to an owner-controlled document that is administratively controlled in accordance with the TS SFCP, Dominion Nuclear Connecticut, Inc.
continues to meet the regulatory requirement of 10 CFR 50.36, and specifically, of 10 CFR 50.36(c}(3} with respect to surveillance requirements.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Connecticut State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in amounts, and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding as published in the Federal Register on December 11, 2012 (77 FR 73687). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be
conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
- 1. TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Controi-RITSTF Initiative 5b," March 18, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML090850642).
- 2. NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," April 2007 (ADAMS Accession No. ML071360456).
- 3. Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," September 19, 2007 (ADAMS Accession No. ML072570267).
- 4. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, May 2011 (ADAMS Accession No. ML100910006).
- 5. Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decision making:
Technical Specifications," Revision 1, May 2011 (ADAMS Accession No. ML100910008).
- 6. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 1, January 2007 (ADAMS Accession No. ML070240001 ).
- 7. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, March 2009 (ADAMS Accession No. ML090410014).
- 8. ASME/ANS PRA Standard ASME/ANS RA-Sa-2009, Addenda to ASME RA-S-2008, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications."
- 9. NEI 00-02, Revision 1 "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," Revision 1, May 2006 (ADAMS Accession No. ML061510621 ).
- 10. NEI 05-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard," Revision 0, August 2006.
- 11. ASME PRA Standard ASME RA-Sb-2005, Addenda to ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications."
Principal Contributor: D. O'Neal Date: February 25, 2014
February 25, 2014 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
MILLSTONE POWER STATION, NO. UNIT 3 -ISSUANCE OF AMENDMENT RE: RISK-INFORMED JUSTIFICATION FOR THE RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCY REQUIREMENTS TO A LICENSEE CONTROLLED PROGRAM (ADOPTION OF TSTF-425, REVISION 3)
Dear Mr. Heacock:
The Commission has issued the enclosed Amendment No. 258 to Renewed Facility Operating License No. NPF-49 for the Millstone Power Station, Unit No. 3, in response to your application dated October 4, 2012, as supplemented by letters dated January 4, 2013, April 17, 2013, and October 30, 2013.
The amendment modifies the Technical Specifications (TSs) by relocating specific surveillance frequencies to a licensee controlled program with the adoption of Technical Specification Task Force (TSTF)-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control -
[Risk-Informed Technical Specification Task Force (RITSTF)] Initiative 5b." Additionally, the amendment adds a new program, the Surveillance Frequency Control Program, toTS Section 6, Administrative Controls.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely, Ira/
James Kim, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-423
Enclosures:
- 1. Amendment No. 258 to NPF-49
- 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
PUBLIC Lp11-1 R/F RidsRgn1 MaiiCenter RidsNrrDorllpl1-1 Resource RidsNrrLAKGoldstein RidsNrrDraApla Resource RidsNrrDoriDpr Resource RidsNrrPMMillstone RidsNrrDssStsb Resource R. McKinley, Rl RidsOgcMaiiCenter Resource RidsAcrsAcnw_MaiiCenter Resource A ccesston No.: ML14023A748 *S ee memo d ated J anuary 16 2014 OFFICE LPL1-1/PM LPL 1-1/LA APLA/BC STSB/BC u.w,t.;/NLU t LPL 1-1/BC wt commen s NAME JKim KGoldstein HHossein* REIIiott JWachutka BBeasley DATE 1/27/14 1/27/14 1/16/14 2/4/14 2/12/14 2/25/14 OFFICIAL ~t:v\JfUJ (.;01 'Y