ML070780688
| ML070780688 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/14/2007 |
| From: | NRC/NRR/ADRO/DORL/LPLI-2 |
| To: | |
| nerses V, NRR//DORL, 415-1484 | |
| Shared Package | |
| ml063540109 | List: |
| References | |
| TAC MC2634 | |
| Download: ML070780688 (9) | |
Text
-4-(2) Technical Spjecifications The Technical Specifications contained in Appendix A, revised through Amendment No. 236 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 INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTIO PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE...................................................................
2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE ..................................
2-1 FIGURE 2. 1-1 DELETED............................................................................
2-2 FIGURE 2.1-2 DELETED............................................................................
2-3 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS..............
2-4 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS......2-5 BASES SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE ................................................................
B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE ...............................
B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOTNTS
..........
B 2-3 MILLSTONE
-UNIT 3 iii Amendment No. 2244, 236
2.0 SAFETY
LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS.REACTOR CORE 2. 1.1 The combination of THERMAL POWER, Reactor Coolant System highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the CORE OPERATING LIMITS REPORT; and the following Safety Limits shall not be exceeded: 2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to 1. 17 for the W'RB-lI/WrRB-2 DNB correlations, 2.1.1.2 The peak fuel centerline temperature shall be maintained less than 5080'F, decreasing by 5 8'F p~er 10,000 MWD/MTU of burnup.APPLICABILITY:
MODES 1 and 2.ACTION: Whenever the Reactor Core Safety Limit is violated, restore compliance and be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.REACTOR COOLANT -SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.APPLICABILITY:
MODES 1, 2,3,4, and 5.ACTlION MODES 1 and 2: Whenever the Reactor Coolant System pressure has exceeded 2750 psia be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.MODES 3, 4 and 5: Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.MILLSTONE
-UNIT 3 2-1 MILLTONE-UIT 32-IAmendment No. 443,-244,236 THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE
-UNIT 3 2-2 MILLTONE-UIT 32-2Amendment No. 60, 244~,2 36 POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION N 3.2.3.1 The indicated Reactor Coolant System (RCS) total flow rate and F AH shall be maintained as follows: a. RCS total flow rate : 371 ,920 gpin and geater than or equal to the limit specified in the CORE OPERATING LIMITS REPORT (COLR), and b. FN <FRTP [10+P b. FH AH [lO+FH(l.O-P)I Where: 1) P THERMAL POWER RATED THERMAL POWER'2) FN N 2 FAN = Measured values of FAH obtained by using the movable incore detectors to obtain a power distribution map. The measured value N of FAH should be used since Specification 3.2.3. lb. takes into consideration a measurement uncertainty of 4% for incore measurement, 3) FRTP =TeFN lmta AE HRA OE nteC~3 FAN = h AHliiatRTDTEMLPWRnthCOR N 4) PF AH = The power factor multiplier for FAH provided in the COLR, and 5) The measured value of RCS total flow rate shall be used since uncertainties of 2.4% for flow measurement have been included in Specification 3.2.3. 1Ia.APPLICABILITY:
MODE 1.ACTION: N With the RCS total flow rate or FAH outside the region of acceptable operation:
- a. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either: 1. Restore the RCS total flow rate to within the limits specified above and in N the COLR and FAN to within the above limit, or 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux -High Trip Setpoint 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 />.3/42-19 3/42-19 Amendment No. --2,64, 69,44-14, 244-, 236 MILLSTONE
-UNIT 3 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 re.Stored to I within the limits specified above and in the COLR and FAH 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 POWER OPERATION may proceed N provided that FAH and 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 RCS total flow rate and FAHl 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 least once per 31 Effective Full Power Days.4.2.3.1.3 The indicated RCS total flow rate shall be verified to be within the acceptable range at lestone er1 hur wenth os -ecnty baiedvaueofFN leat oce er 2 hurswhe th mst ecetlyobtine vaue f AH , obtained per Specification 4.2.3.1.2, is assumed to exist.4.2.3.1.4 The RCS total flow rate indicators shall be subjected to a CHAN4NEL CALIBRATION at least once per 18 months. The measurement instrumentation shall be calibrated within 7 days prior to the performance of the calorimetric flow measurement.
MILLSTONE
-UNIT 3 3/4 2-20 Amendment No. 6G, 49,4-t0% 236 ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 69.1.5 Deleted CORE OPERATING LIMITS REPORT 6.9.1.6. a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:
I1. Reactor Core Safety Limit for Specification
- 2. 1. 1.2.Overtemperature AT and Overpower AT setpoint parameters for Specification 2.2. 1.3.SHUTDOWN MARGIN for Specifications 3/4.1.1.1.1, 3/4.1.1.1.2, and 3/4.1.1.2.
- 41. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance 5.limit for Specification 3/4.1.1.3.
Shutdown Rod Insertion Limit for Specification 3/4.1.3.5.
Control Rod Insertion Limits for Specification 314.1.3.6.
- 7. AXIAL FLUX DIFFERENCE Limits, target band, and APLOD~ for Specification 3/4.2.1.1.
- 8. Heat Flux Hot Channel Factor, K(z), W(z), APLMO, and W(Z)BL for Specification 3/4.2.2.1.
- 9. RCS Total Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor, and Power Factor Multiplier for Specification 3/4.2.3.10. DNB Parameters for Specification 3/4.2.5.1.1. Shutdown Margin Monitor minimum count Tate for Specification 3/4.3.5.1-2. Boron Concentration for Specification 3/4.9. 1.1.MILLSTONE
-UNIT 3 6--19a Amendment No. -24,44~, 69, W. 4-99, 2-lg,224 29, 236 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)6.9.1 .6.b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in: I1. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," (W Proprietary). (Methodologyfor Specifications 2.1.1.1 -Departure from Nucleate Boiling Ratio, 2.1.1.2--Peak.
Fuel Centerline Temperature, 3.1.1.3 --Moderator Temperature Coefficient, 3.1.3 .5-Shutdown Bank Insertion Limit, 3.1.3.6--Control Bank Insertion Limits, 3.2. 1--AXIAL FLUX DIFFERENCE, 3.2.2--Heat Flux Hot Channel Factor, 3.2.3--Nuclear Enthalpy Rise Hot Channel Factor, 3.1.1.1.1, 3.1.1.1.2, 3.1.1.2 -- SHUTDOWN MARGIN, 3.9. 1. 1-- Boron Concentration.)
2). T. M. Anderson to K. Kniel (Chief of Core Performance Branch, NRC) January 3 1, 1 980--
Attachment:
Operation and Safety-Analysis Aspects of an Improved Load Follow Package.3. NUREG-800, Standard Review Plan, U.S. Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981 Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAO C), Revision 2, July 198 1.4. WCAP- 10216-P-A-R1A, "RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION," (~Po rita .(Methodology for Specifications 3.2.1-AILFU DIFFRENCE
[Relaxed Axial Offset Control] and 3.2.2--Heat Flux Hot Channel Factor [W(z) surveillance requirements for EQ Methodology].)
- 5. WCAP-9561-P-A, ADD.3, "BART A-1: A COMPUTER CODE FOR THE BEST ESTIMATE ANALYSIS OF REFLOOD TRANSIENTS--SPECIAL REPORT: THIMBLE MODELING W~ ECCS EVALUATION MODEL," (W Proprietary). (Methodology for Specification 3.2.2--Heat Flux Hot Channel Factor.)6. WCAP-10266-P-A, Addendum 1, "TIHE 1981 VERSION OF THE WESTINGHOUSE ECCS EVALUATION MODEL USING THE BASH CODE," ( W Proprietary). (Methodology for Specification 3.2.2--Heat Flux Hot Channel-actor.)7. WCAP-l 1946, "Safety Evaluation Suppotig a More Negative EOL Moderator Temperature Coefficient TUecrhnical Spec'ifica-tion for the Millstone Nuclear Power Station U-nit 3," (-W Proprietary).
- 8. WCAP- 10054-P-A, "WESTINGHOUSE SMALL BREAK EGGS EVALUATION MODEL 17 USING THE_ NOTRUMP CODE," &W Propreay.(thdlg for Specification 3.2.2 -Heat Flux Hot Channel F actor.) reay.(ehdlg
- 9. WCAP-10079-P-A, "NOTRUMP -A NODAL TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE," &~ Proprietary). (Methodology for Specification 3.2.2 -Heat Flux Hot Chiannel1 Factor.)10. WCAP- 126 10, "VANTAGE+
Fuel Assembly Report," (.W Proprietary).(Methodology for Specification 3.2.2 -Heat Flux Hot Channel Factor.)MILLSTONE
-UNIT 3 6-20 Amendment No. -24, P, 60,69,84-, 4-29,+7-1 9, 24-8, 229, 236 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)11. Letter from V. L. Rooney (USNRC) to J. F. Opeka, "Safety Evaluation for Topical Report, NUS CO-I 152, Addendum 4, 'Physics Methodology for PWR Reload Design,' TAC No. M91815," July 18, 1995.12. Letter from E. J. Mroczka to the USNRC, "Proposed Changes to Technical Specifications, Cycle 4 Reload Submittal
-Boron Dilution Analysis," B 13678, December 4, 1990.13. Letter from D. H. Jaffe (USNRC) to E. J. Mroczka, "Issuance of Amendment (TAC No. 77924)," March 11, 199 1.14. Letter from M. H. Brothers to the USNRC, "Proposed Revision to Technical Specification, SHUTDOWN MARGIN Requirements and Shutdown Margin Monitor OPERABILITY for MODES 3, 4, and 5 (PTSCR 3-16-97), B 16447, May 9, 1997.15. Letter from J. W. Anderson (tJSNRC) to M. L. Bowling (NNECO), "Issuance of Amendment
-Millstone Nuclear Power Station, Unit No. 3 (TAC No. M98699)," October 21, 1998.16. WCAP-830 1, "LOCTA-rV Program: Loss-of-Coolant Transient Analysis." 17. WCAP-10054-P-A, Addendum 2, "Addendum to the Westinghouse Small Break EGGS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model." 18. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtempe-rature AT Trip Functions," (Westinghouse Proprietary Class 2).(Methodology for Specifications
- 2. 1.1 and 2.2. 1.)MILLSTONE
-UNIT 3 6-20a MILLTONE-UNT 3 -20aAmendment No. ~8+l, -170, 2-14, 229, 23 6