ML111790031

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Request for License Amendment Regarding Measurement Uncertainty Recapture (Mur) Power Uprate, Attachment 2A, Markup of Proposed Operating License and Technical Specifications Pages
ML111790031
Person / Time
Site: Braidwood  Constellation icon.png
Issue date: 06/23/2011
From:
Exelon Generation Co, Exelon Nuclear
To:
Office of Nuclear Reactor Regulation
Shared Package
ML111790026 List:
References
Download: ML111790031 (8)
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Text

ATTACHMENT 2A Markup of Proposed Operating License and Technical Specifications Pages Braidwood Station, Units I and 2 Facility Operating License Nos. NPF-72 and NPF-77 REVISED OPERATING LICENSE AND TECHNICAL SPECIFICATIONS PAGES Operating License (Units 1 and 2), Page 3 TS Page 1.1-6 TS Page 2.0-1 TS Page 3.4.1-1 TS Page 3.4.1-2 TS Page 5.6-4

(3) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. The license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and Is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below' (1) Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of egawatts thermal (100 percent rated power) in accordance with the conditions specified herein and other items identified in Attachment 1 to this license. The items identified in Attachment I to this license shall be completed as specified.

Attachment 1 is hereby incorporated Into this license.

(2) Technical Specifications The Technical Specifitt bns contained in Appendix A as revised through Amendment No.-4637 and the Environmental Protection Plan contained in I Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) EmeraencvPlannin_a in the event that the NRC finds that the lack of progress In completion of the procedures in the Federal Emergency Management Agency's final rule, 44 CFR Part 350, is an indication that a major substantive problem exists in achieving or maintaining an adequate state of emergency preparedness, the provisions of 10 CFR Section 50,54(sx2) will apply.

Amendment 163

material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Exelon Generation Company, LLC pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use In amounts are required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and TO, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. The license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or Incorporated below:

(1) Maximum Power level The licensee is authorized tVperate the facility at reactor core power levels not In excess of4656. 6 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein and other items identified in Attachment 1 to this license. The items identified in Attachment 1 to this license shall be completed as specified, Attachment I is hereby incorporated into this license.

(2) Technical Specifications The Technical Specific ns contained in Appendix A as revised through Amendment No:-183, and the Environmental Protection Plan contained in Appendix B, both of which are attached to License No. NPF-72, dated July 2, 1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Emergency Planning In the event that the NRC finds that the lack of progress in completion of the procedures In the Federal Emergency Management Agency's final rule, 44 CFR Part 350, is an indication that a major substantive problem exists in achieving or maintaining an adequate state of emergency preparedness, the provisions of 10 CFR Section 50.54(s)(2) will apply.

Amendment No, 468 ,

Definitions 1.1 1.1 Definitions PRESSURE AND The PTLR is the unit specific document that TEMPERATURE LIMITS provides the reactor vessel pressure and REPORT (PTLR) temperature limits including heatup and cooldown rates, and the pressurizer Power Operated Relief Valve (PORV) lift settings for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.6.

QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper RATIO (QPTR) excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector 3645 calibrated output to the average of the lower excore detector calibrated outputs, whichever is RATED THERMAL POWER RTP shall be a total` r core heat transfer (RTP) rate to the reactor coolant o.6 MWt.

REACTOR TRIP The RTS RESPONSE TIME shall be that time interval SYSTEM (RTS) RESPONSE from when the monitored parameter exceeds its RTS TIME trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

RECENTLY IRRADIATED FUEL Fuel that has occupied part of a critical reactor core within the previous 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Note that all fuel that has been in a critical reactor core is referred to as irradiated fuel.

BRAIDWOOD -- UNITS 1 & 2 1.1 - 6 Amendment 442

SLs 2.0 2.0 SAFETY LIMITS (SLs) and >_ 1.19 for the ABB-NV DNB correlation for a thimble cell and a typical 2.1 SLs cell 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, /,Reactor Coolant System (RCS) highest loop av age teerat re, and pressurizer pressure shall not exceed the limits pecified in the COLR; and the following SLs shall not e exceeds 2.1.1.1 In MODE 1, the Departure from Nuclea Boiling Ratio (DNBR) shall be maintained ? 1 24 f the WRB-2 DNB correlation for a thimble r° and 1.25 for the WRB-2 DNB correlation for a typical cell.

2.1.1.2 In MODE 2, the DNBR shall be maintained ^ 1.17 for the WRB-2 DNB correlation, and L30 for tl e W 3 DNB y,

` errel ati-.

2.1.1.3 In MODES I and , e peak fuel centerline temperature shall be ained as follows:

< 5080°F decreasing by 58°F per 10,000 MWD/MTU burnup 1.13 for the ABB-NV DNB for Westinghouse fuel, correlation and > 1.18 for the WLOP DNB

b. < 5173°F decreasing by 65°F per 10,000 MWD/MTU burnup correlation for AREVA NP fuel (Unit 1 only), and
c. < 5189°F decreasing by 65°F per 10,000 MWD/MTU burnup for AREVA NP fuel containing Gadolinia (Unit 1 only).

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained 2735 prig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

BRAIDW00D - UNITS 1 & 2 2.0 - 1 Amendment

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCD 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature (Ta,), and RCS total flow rate shall be within the limits specified below:

a. Pressurizer pressure within the limit specified in the COLR;
b. RCS average temperature (T,,,) within the limit specified in the COLR; and
c. RCS total flow rate ? 4-900 gpm and within the limit specified in the COLR.

386,000

NOTE --------------------------- -

Pressurizer pressure limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute; or
b. THERMAL POWER step > 10% RTP.

APPLICABILITY: MODE ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more RCS DNB A.1 Restore RCS DNB 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> parameters not within parameter(s) to limits. within limit.

B. Required Action and B.1 Be in MODE 2. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.

BRAIDWOOD - UNITS 1 & 2 3.4.1 - 1 Amendment 4-1-3

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEI LANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is within the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> limit specified in the COLR.

SR 3.4.1.2 Verify RCS average temperature ( Tav^,) is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> within the limit specified in the L OLR.

SR 3.4.1.3 Verify RCS total flow rate is ^t 399,90 gpm 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and within the limit specified in the SR 3.4.1.4 ----Y-------------- NOTE -------------------- 386,000 Not required to be performed until 7 days after >_ 90% RTP.

Verify by precision heat balance that RCS 18 months total flow rate i s ? 389,j99 gpm and within th e li m it spec i fied in th COLR.

386,000 BRAIDW00D - UNITS 1 & 2 3.4.1 - 2 Amendment 4-1-3

Reporting Reirements qu 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT LR (continued)

5. ComEd letter from D. Saccomando to the Office of Nuclear Reactor Regulation dated December 21, 1994, transmitting an attachment that documents a plicable sections of WCAP-11992!11993 and ComEd application of the UET methodology addressed in "Additional Information Regarding Application for Amendment to Facility Operating Licenses-Reactivity Control Systems."

WCAP-16009-P-A, Revision 0, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," January 2005.

7. WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code," August 1985.

WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using NOTRUMP Code," August 1985.

WCAP-10216-P-A, Revision 1, "Relaxation of Constant Axial Offset Control - FQ Surveillance Technical Specification," February 1994.

10. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions,"

September 1986;

c. Tt}core operating limits shall be determined such that all plicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met; and The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
11. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," October 1999; BRA I DWOOD - UNITS 1 & 2 5.6 -- 4 Amendment 444
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