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Category:Letter type:NL
MONTHYEARNL-21-034, Notification of Expected Date of Transfer of Ownership of Nuclear Units to Holtec Indian Point 2, LLC and Holtec Indian Point 3, LLC; and Notification of Receipt of All Required Regulatory Approvals2021-05-26026 May 2021 Notification of Expected Date of Transfer of Ownership of Nuclear Units to Holtec Indian Point 2, LLC and Holtec Indian Point 3, LLC; and Notification of Receipt of All Required Regulatory Approvals NL-21-039, Response to Request for Additional Information - License Amendment Request to Revise the Indian Point Nuclear Generating Unit No. 3 Licensing Basis to Incorporate the Installation and Use of a New Auxiliary2021-05-20020 May 2021 Response to Request for Additional Information - License Amendment Request to Revise the Indian Point Nuclear Generating Unit No. 3 Licensing Basis to Incorporate the Installation and Use of a New Auxiliary NL-21-033, Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel2021-05-11011 May 2021 Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel NL-21-032, Termination of Emergency Response Data System Feed to the U.S. Nuclear Regulatory Commission at Indian Point Energy Center2021-05-11011 May 2021 Termination of Emergency Response Data System Feed to the U.S. Nuclear Regulatory Commission at Indian Point Energy Center NL-21-005, Cancellation of Commitments Related to Beyond-Design-Basis External Events Seismic and Flooding Actions2021-05-11011 May 2021 Cancellation of Commitments Related to Beyond-Design-Basis External Events Seismic and Flooding Actions NL-21-030, Submittal of 2020 Annual Radiological Environmental Operating Report2021-05-0606 May 2021 Submittal of 2020 Annual Radiological Environmental Operating Report NL-21-027, Registration of Spent Fuel Cask Use2021-04-20020 April 2021 Registration of Spent Fuel Cask Use NL-21-021, Registration of Spent Fuel Cask Use2021-04-19019 April 2021 Registration of Spent Fuel Cask Use NL-21-017, Pre-Notice of Disbursement from Decommissioning Trusts2021-04-0808 April 2021 Pre-Notice of Disbursement from Decommissioning Trusts NL-21-010, Submittal of 2020 Annual Fitness for Duty Program Performance Data Report and Fatigue Management Program Data Report2021-02-17017 February 2021 Submittal of 2020 Annual Fitness for Duty Program Performance Data Report and Fatigue Management Program Data Report NL-21-006, Relief Request IP3-ISI-RR-16, Proposed Alternative to American Society of Mechanical Engineers Code Case N-513-4 Inspection Requirement2021-02-10010 February 2021 Relief Request IP3-ISI-RR-16, Proposed Alternative to American Society of Mechanical Engineers Code Case N-513-4 Inspection Requirement NL-21-014, Response to 2nd Round Request for Additional Information - License Amendment Request to Revise the Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device2021-01-26026 January 2021 Response to 2nd Round Request for Additional Information - License Amendment Request to Revise the Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device NL-20-082, Notice of Planned Transfer of Decommissioning Funds2020-12-14014 December 2020 Notice of Planned Transfer of Decommissioning Funds NL-20-081, Pre-Notice of Disbursement from Decommissioning Trusts2020-12-0909 December 2020 Pre-Notice of Disbursement from Decommissioning Trusts NL-20-080, Report in Accordance with 10 CFR 71.95(a) for Failure to Comply with Certificate of Compliance No. 71-93212020-11-19019 November 2020 Report in Accordance with 10 CFR 71.95(a) for Failure to Comply with Certificate of Compliance No. 71-9321 NL-20-079, (IP2 and IP3) - Request for a One-Time Exemption from 10 CFR 73, Appendix B, Section VI, Subsection C.3.(I)(1) Regarding Annual Force-on-Force (FOF) Exercises, Due to Covid 19 Pandemic2020-11-12012 November 2020 (IP2 and IP3) - Request for a One-Time Exemption from 10 CFR 73, Appendix B, Section VI, Subsection C.3.(I)(1) Regarding Annual Force-on-Force (FOF) Exercises, Due to Covid 19 Pandemic NL-20-077, Submittal of Quality Assurance Program Manual Revision 22020-11-0909 November 2020 Submittal of Quality Assurance Program Manual Revision 2 NL-20-078, Response to Requests for Additional Information - License Amendment Request to Revise the Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device2020-11-0909 November 2020 Response to Requests for Additional Information - License Amendment Request to Revise the Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device NL-20-076, Revision of Commitment Related to Nuclear Reactor Safeguards Interim Compensatory Measure - Section B.5.b Issue Regarding Spent Fuel Dispersal2020-11-0202 November 2020 Revision of Commitment Related to Nuclear Reactor Safeguards Interim Compensatory Measure - Section B.5.b Issue Regarding Spent Fuel Dispersal NL-20-069, One-time Scheduler Exemption Request from 10 CFR 50, Appendix E Biennial Emergency Preparedness Exercise Requirements Due to COVID-19 Public Health Emergency2020-10-0808 October 2020 One-time Scheduler Exemption Request from 10 CFR 50, Appendix E Biennial Emergency Preparedness Exercise Requirements Due to COVID-19 Public Health Emergency NL-20-070, Response to Requests for Additional Information, License Amendment Request to Revise the Indian Point Nuclear Generating Unit No. 3 Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device2020-10-0202 October 2020 Response to Requests for Additional Information, License Amendment Request to Revise the Indian Point Nuclear Generating Unit No. 3 Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device NL-20-067, Redacted Version of Response to U.S. Nuclear Regulatory Commission Region I Letter Regarding Algonquin Incremental Market Project Pipeline2020-09-16016 September 2020 Redacted Version of Response to U.S. Nuclear Regulatory Commission Region I Letter Regarding Algonquin Incremental Market Project Pipeline NL-20-064, 10 CFR 50.59(d)(2) Summary Report of Changes, Tests and Experiments2020-09-0101 September 2020 10 CFR 50.59(d)(2) Summary Report of Changes, Tests and Experiments NL-20-060, Status of Remaining Actions for Generic Letter 2004-022020-08-11011 August 2020 Status of Remaining Actions for Generic Letter 2004-02 NL-20-057, Cancellation of Commitment Related to Large Break LOCA Reanalysis2020-07-30030 July 2020 Cancellation of Commitment Related to Large Break LOCA Reanalysis NL-20-0851, 30-Day 10 CFR 21 Notification - Continuously Energized Eaton D26 Relays Could Fail to Deenergize Because of an Organic C3 Insulating Material2020-07-22022 July 2020 30-Day 10 CFR 21 Notification - Continuously Energized Eaton D26 Relays Could Fail to Deenergize Because of an Organic C3 Insulating Material NL-20-051, Submittal of Quality Assurance Program Manual, Revision 1 for the Indian Point Energy Center2020-07-0707 July 2020 Submittal of Quality Assurance Program Manual, Revision 1 for the Indian Point Energy Center NL-20-052, Unsatisfactory 10 CFR 26 Fitness-For-Duty Blind Performance Testing Results2020-07-0707 July 2020 Unsatisfactory 10 CFR 26 Fitness-For-Duty Blind Performance Testing Results NL-20-012, Application to Revise Provisional Operating License and Technical Specifications2020-06-30030 June 2020 Application to Revise Provisional Operating License and Technical Specifications NL-20-050, Response to U.S. Nuclear Regulatory Commission Region I Letter Regarding Algonquin Incremental Market Project Pipeline2020-06-24024 June 2020 Response to U.S. Nuclear Regulatory Commission Region I Letter Regarding Algonquin Incremental Market Project Pipeline NL-20-041, Registration of Unit 3 Spent Fuel Cask Use2020-05-13013 May 2020 Registration of Unit 3 Spent Fuel Cask Use NL-20-042, Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel2020-05-12012 May 2020 Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel NL-20-033, Technical Specifications Proposed Change - Permanently Defueled Technical Specifications2020-04-28028 April 2020 Technical Specifications Proposed Change - Permanently Defueled Technical Specifications NL-20-038, Response to U.S. Nuclear Regulatory Commission Region I Letter Regarding Algonquin Incremental Market Project Pipeline2020-04-23023 April 2020 Response to U.S. Nuclear Regulatory Commission Region I Letter Regarding Algonquin Incremental Market Project Pipeline NL-20-035, Response to Request for Additional Information - Temporary Exemption Request from 10 CFR Appendix R, Section Iii.H Due to COVID-19 Pandemic2020-04-16016 April 2020 Response to Request for Additional Information - Temporary Exemption Request from 10 CFR Appendix R, Section Iii.H Due to COVID-19 Pandemic NL-20-034, Temporary Exemption Request from 10 CFR Appendix R, Section Iii.H Due to COVID-19 Pandemic2020-04-13013 April 2020 Temporary Exemption Request from 10 CFR Appendix R, Section Iii.H Due to COVID-19 Pandemic NL-20-021, Proposed License Amendment to Revise the Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device2020-03-24024 March 2020 Proposed License Amendment to Revise the Licensing Basis to Incorporate the Installation and Use of a New Auxiliary Lifting Device NL-20-020, Submittal of 2019 Annual Fitness for Duty Performance Data Report Update2020-02-26026 February 2020 Submittal of 2019 Annual Fitness for Duty Performance Data Report Update NL-20-015, Submittal of 2019 Annual Fitness for Duty Program Performance Data Report and Fatigue Management Program Data Report2020-02-10010 February 2020 Submittal of 2019 Annual Fitness for Duty Program Performance Data Report and Fatigue Management Program Data Report NL-20-008, Transmittal of Presentation Slides for Partially Closed Pre-Submittal Meeting to Discuss a Planned License Amendment Request to Replace the Fuel Handling Building Crane2020-01-0606 January 2020 Transmittal of Presentation Slides for Partially Closed Pre-Submittal Meeting to Discuss a Planned License Amendment Request to Replace the Fuel Handling Building Crane NL-19-094, 2018 Annual 10 CFR 50.46 Emergency Core Cooling System Evaluation Changes Report2019-12-16016 December 2019 2018 Annual 10 CFR 50.46 Emergency Core Cooling System Evaluation Changes Report NL-19-084, Application for Order Consenting to Transfers of Control of Licenses and Approving Conforming License Amendments2019-11-21021 November 2019 Application for Order Consenting to Transfers of Control of Licenses and Approving Conforming License Amendments NL-19-093, Proposed Technical Specifications (TS) Changes - Indian Point Nuclear Generating Unit 3 TS SR 3.7.7.2 and TS 3.7.6, Required Action A.12019-11-21021 November 2019 Proposed Technical Specifications (TS) Changes - Indian Point Nuclear Generating Unit 3 TS SR 3.7.7.2 and TS 3.7.6, Required Action A.1 NL-19-092, Request for Rescission of Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)2019-11-20020 November 2019 Request for Rescission of Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049) NL-19-043, Request for Partial Exemption from Record Retention Requirements in 10 CFR 50.122019-10-22022 October 2019 Request for Partial Exemption from Record Retention Requirements in 10 CFR 50.12 NL-19-073, Request for Relaxation of Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)2019-10-22022 October 2019 Request for Relaxation of Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049) NL-19-078, Supplement to Technical Specifications Proposed Change - Permanently Defueled Technical Specifications2019-10-22022 October 2019 Supplement to Technical Specifications Proposed Change - Permanently Defueled Technical Specifications NL-19-091, Independent Spent Fuel Storage Installation (Isfsi), Registration of Spent Fuel Cask Use2019-10-17017 October 2019 Independent Spent Fuel Storage Installation (Isfsi), Registration of Spent Fuel Cask Use NL-19-090, Registration of Unit 2 Spent Fuel Cask Use2019-10-0909 October 2019 Registration of Unit 2 Spent Fuel Cask Use NL-19-079, 50.59(d)(2) Summary Report of Changes, Tests and Experiments2019-09-26026 September 2019 50.59(d)(2) Summary Report of Changes, Tests and Experiments 2021-05-06
[Table view] Category:Fuel Cycle Reload Report
MONTHYEARML19106A0522019-04-0404 April 2019 Submittal of Core Operating Limits Report for Cycle 21 NL-18-020, Core Operating Limits Report for Cycle 242018-04-10010 April 2018 Core Operating Limits Report for Cycle 24 NL-17-043, Core Operating Limits Report for Cycle 202017-04-13013 April 2017 Core Operating Limits Report for Cycle 20 NL-17-023, Revision to the Core Operating Limits Report2017-02-22022 February 2017 Revision to the Core Operating Limits Report NL-16-070, Core Operating Limits Report2016-06-15015 June 2016 Core Operating Limits Report NL-16-017, Mid-Cycle Revision to the Core Operating Limits Report for Cycle 192016-02-0404 February 2016 Mid-Cycle Revision to the Core Operating Limits Report for Cycle 19 NL-14-057, Submittal of Core Operating Limits Report Cycle 222014-04-0909 April 2014 Submittal of Core Operating Limits Report Cycle 22 NL-13-0303, Submittal of Revised Core Operating Limits Report2013-04-18018 April 2013 Submittal of Revised Core Operating Limits Report NL-12-022, Unit. 2, Revised Core Operating Limits Report2012-04-0303 April 2012 Unit. 2, Revised Core Operating Limits Report NL-11-049, Revised Core Operating Limits Report2011-05-0404 May 2011 Revised Core Operating Limits Report NL-10-070, Revised Core Operating Limits Report, Mid-Cycle Revision2010-07-0707 July 2010 Revised Core Operating Limits Report, Mid-Cycle Revision NL-10-034, Revised Core Operating Limits Report, Cycle 202010-05-0303 May 2010 Revised Core Operating Limits Report, Cycle 20 NL-09-047, Revised Core Operating Limits Report, Cycle 162009-05-14014 May 2009 Revised Core Operating Limits Report, Cycle 16 NL-08-114, Revised Core Operating Limits Report for Indian Point Unit 22008-07-11011 July 2008 Revised Core Operating Limits Report for Indian Point Unit 2 NL-06-085, 2006 Summary Reports for In-Service Inspection and Repairs or Replacements2006-08-14014 August 2006 2006 Summary Reports for In-Service Inspection and Repairs or Replacements NL-06-069, Revised Core Operating Limits Report, Cycle 182006-06-22022 June 2006 Revised Core Operating Limits Report, Cycle 18 NL-05-069, Cycle 14 Core Operating Limit Report (Colr), Revision 142005-05-11011 May 2005 Cycle 14 Core Operating Limit Report (Colr), Revision 14 NL-05-022, Cycle 17 Core Operating Limit Report (COLR)2005-02-16016 February 2005 Cycle 17 Core Operating Limit Report (COLR) NL-03-080, Core Operating Limits Report for Cycle 132003-05-0808 May 2003 Core Operating Limits Report for Cycle 13 NL-02-143, Cycle 16 Core Operating Limit Report (COLR)2002-11-15015 November 2002 Cycle 16 Core Operating Limit Report (COLR) 2019-04-04
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Text
Enteray Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB Entergy P.O. Box 249 Buchanan, N.Y. 10511-0249 Tel (914) 254-6710 Robert Walpole Licensing Manger NL-13-030 April 18, 2013 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Revised Core Operating Limits Report Indian Point Nuclear Generating Unit No. 3 Docket No. 50-286 License No. DPR-64
Dear Sir or Madam:
Enclosures 1 and 2 to this letter provide Entergy Nuclear Operations Core Operating Limits Report (COLR) for Indian Point 3 Cycle 18. This report is submitted in accordance with Technical Specification 5.6.5.d.
There are no new commitments contained in this letter. If you have any questions or require additional information, please contact me at 914-254-6710.
Sincerely, f bi ccýk RW/mb cc: next page bl
NL-13-030 Docket No. 50-286 Page 2 of 2
Enclosure:
- 1. 3-GRAPH-RPC-16, Rev. 6, Core Operating Limits Report cc: Mr. William Dean, Regional Administrator, NRC Region 1 Mr. Douglas Pickett, Senior Project Manager, NRC NRR DORL IPEC NRC Resident Inspector's Office Mr. Francis J. Murray, President and CEO, NYSERDA (w/o enclosure)
Ms. Bridget Frymire, New York State Department of Public Service
ENCLOSURE 1 TO NL-13-030 3-GRAPH-RPC-16, Rev. 6, Core Operating Limits Report ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT UNIT 3 NUCLEAR POWER PLANT DOCKET No. 50-286
Procedure Use Is: Control Copy:
Nuclear Northeast l] Continuous Effective Date: _____ __
0 Reference 0 Information Page 1 of 12 3-GRAPH-'RPC-1 6-1 5"Revision: 6 CORE'OPIIERATING LIMITS REPORT l%pprovedBy:
3- 2 te, 13 D'a Q16/Designe'e Procedure Sponsor, ICal II Procedure Owner PARTIAL REVISION
No: 3-GRAPH-RPC-16 Rev: 6 CORE OPERATING LIMITS REPORT Page 2 of 12 REVISION
SUMMARY
(Page 1 of 1) 1.0 REASON FOR REVISION 1.1 Incorporate Cycle 18, changes. The only change from the cycle 17 COLR is an update to the applicable cycle number. (EC-30830) 2.0
SUMMARY
OF CHANGES 2.1 Changed reference from Cycle 17 to Cycle 18 in NOTE prior to TS 2.1.1.
(EC-30830) [Editorial 4.6.13]
TABLE OF CONTENTS SECTION PAGE TS 2.1.1 Reactor C ore S Ls ............................................................................................... 4 TS 3.1.1 Shutdown Margin (SDM) ..................................................................................... 4 TS 3.1.3 Moderator Temperature Coefficient (MTC) .......................................................... 4 TS 3.1.5 Shutdown Bank Insertion Limits .......................................................................... 5 TS 3.1.6 Control Bank Insertion Limits .............................................................................. 5 TS 3.2.1 Heat Flux Hot Channel Factor (FQ(Z)) ................................................................ 5 TS 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor FAH..............................5 TS 3.2.3 Axial Flux Difference (AFD) (Constant Axial Offset Control (CAOC) Methodology). 6 TS 3.3.1 R PS Instrum entation ............................................................................................ 6 TS 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)
L im its ....................................................................................................................................... 6 TS 3.9.1 Refueling Boron Concentration ............................................................................ 6 ATTACHMENTS Attachment 1 OVERTEMPERATURE AT ALLOWABLE VALUE ....................................... 7 Attachment 2 OVERPOWER AT ALLOWABLE VALUE .................................................... 8 FIGURES Figure 1 Reactor Core Safety Limit - Four Loops in Operation ........................................ 9 Figure 2 Rod Bank Insertion Limits .................................................................................. 10 Figure 3 Hot Channel Factor Normalized Operating Envelope ......................................... 11 Figure 4 Axial Flux Difference Envelope Limits ................................................................. 12
No: 3-GRAPH-RPC-16 Rev: 6 CORE OPERATING LIMITS REPORT Page 4 of 12 NOTE The data presented in this report applies to Cycle 18 ONLY and may NOT be used for other cycles of operation. Also, it applies only to operation at a maximum power level of 3188.4 MWt. Any technical change to this document may require a Safety Evaluation to be performed in accordance with 10 CFR 50.59.
TS 2.1.1 Reactor Core SLs In MODE 1 and 2, the combination of thermal power level, pressurizer pressure, and Reactor Vessel inlet temperature SHALL not exceed the limits shown in Figure 1. The safety limit is exceeded if the point defined by the combination of Reactor Vessel inlet temperature and power level is at any time above the appropriate pressure line.
TS 3.1.1 Shutdown Margin (SDM)
The shutdown margin SHALL be greater than or equal to 1.3% Ak/k.
TS 3.1.3 Moderator Temperature Coefficient (MTC)
The MTC upper limit SHALL be < 0.0 Ak/k/OF at hot zero power.
The MTC lower limit SHALL be less negative than or equal to:
-38.0 pcm/°F @ 300 ppm
-44.5 pcm/°F @ 60 ppm
-47.0 pcm/°F @ 0 ppm
TS 3.1.5 Shutdown Bank Insertion Limits The Shutdown Banks SHALL be fully withdrawn when the reactor is in MODE 1 and MODE 2. Shutdown Banks with a group step counter demand position > 225 steps are considered fully withdrawn because the bank demand position is above the top of the active fuel.
TS 3.1.6 Control Bank Insertion Limits The Control Bank Insertion Limits for MODE 1 and MODE 2 with keff 1.0 are as indicated in Figure 2. Control Bank Insertion Limits apply to the step counter demand position.
Each control bank shall be considered fully withdrawn at > 225 steps.
TS 3.2.1 Heat Flux Hot Channel Factor (F(_QLZ NOTE
" P is the fraction of Rated Thermal Power (RTP) at which the core is operating.
" K(Z) is the fraction given in Figure 3 and Z is the core height location of Fa.
IF P > .5, Fa(Z) < (2.30 / P) x K(Z)
IF P < .5, FQ(Z) < (4.60) x K(Z)
TS 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor FAH NOTE P is the fraction of Rated Thermal Power (RTP) at which the core is operating.
FNH< 1.65 { 1 +0.3(1 -P)}
No: 3-GRAPH-RPC-16 Rev: 6 CORE OPERATING LIMITS REPORT Page 6 of 12 TS 3.2.3 Axial Flux Difference (AFD) (Constant Axial Offset Control (CAOC)
Methodologv)
The Indicated limit is the Target Band; i.e., the Target +/- 5%
The AFD shall be maintained within the ACCEPTABLE OPERATION portion of Figure 4, as required by TS 3.2.3.
TS 3.3.1 RPS Instrumentation
- 1. Overtemperature AT Allowable Value as referenced in Technical Specifications Table 3.3.1-1, Function 5, Note 1 Refer to Attachment 1
- 2. Overpower AT Allowable Value as referenced in Technical Specifications Table 3.3.1-1, Function 6, Note 2 Refer to Attachment 2 TS 3.4.1 RCS Pressure, Temperature, and Flow Degarture from Nucleate Boiling (DNB) Limits The following DNB related parameters are applicable in MODE 1:
- a. Reactor Coolant System loop Tavg <- 576.70 F for full-power Tavg = 572.0°F
- b. Pressurizer Pressure > 2204 psig
- c. Reactor Coolant System Total Flow Rate Ž 364,700 gpm TS 3.9.1 Refueling Boron Concentration When required by Technical Specification 3.9.1, the minimum boron concentration in the RCS, Refuel Canal, and Reactor Cavity SHALL be the more restrictive of either _>2050 ppm or that which is sufficient to provide a shutdown margin > 5% Ak/k.
Attachment 1 (Page 1 of 1)
OVERTEMPERATURE AT ALLOWABLE VALUE The Overtemperature AT Function Allowable Value SHALL NOT exceed the Technical Specification Table 3.3.1-1, Note 1 value.
The following provides the computed value:
AT < ATo [K1 - K2 [(1 + T1s)/(1 + t 2 s)] (T - T') + K3 (P - P') - f1(AI)]
Where: AT is measured RCS AT, OF (measured by hot leg and cold leg RTDs).
AT, is the loop specific indicated AT at RTP, OF.
s is the Laplace transform operator, sec 1 .
T is the measured RCS average temperature, OF.
T' is the loop specific indicated Tavg at RTP, °F < 572.00 F.
P is the measured pressurizer pressure, psig.
P' is the nominal RCS operating pressure, __2235 psig.
K1 *1.26 K2 >_0.022/OF K3 > 0.00070/psi rl > 25.0 sec T2 -- 3.0 sec fd(Al) = 4.00[ -15.75- (qt - qb)] when qt - qb < - 15.75% RTP 0% of RTP when -15.75% RTP < qt - qb < 6.9% RTP
+3.33[(qt - qb) - 6.9]1 when qt - qb > 6.9% RTP Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.
Attachment 2 (Page 1 of 1)
OVERPOWER AT ALLOWABLE VALUE The Overpower AT Function Allowable Value SHALL NOT exceed the Technical Specification Table 3.3.1-1, Note 2 value.
The following provides the computed value:
AT < AT, [K4 - K5 [(R3s)/(1 + ms)](T) - K6(T - T") - f2 (AI)]
Where: AT is measured RCS AT, OF (measured by hot leg and cold leg RTDs).
ATo is the loop specific indicated AT at RTP, OF.
s is the Laplace transform operator, sec1.
T is the measured RCS average temperature, OF.
T" is the loop specific indicated Tavg at RTP, °F < 572.0°F.
K4 < 1.10 K5 > 0.0175/°F for increasing T K6 > 0.0015/°F when T > T" 0/°F for decreasing T 0/°F when T < T" T3 _Ž10 sec f2(AI) = 0
Figure 1 Reactor Core Safety Limit - Four Loops in Operation (Page 1 of 1) 247 r I: U AICC*EP13AI3LI" 640 ---- -_0, ;0 sa 640 6.
LL I...**' ".
0) 4-01 0 10 20 30 40 50 60 70 W8 90 '08 1 to Reactor Power (Percent of 3216 MWt)
[Conservative relative to 3188.4 MWt; use as-is for operation at 3188.4 MWt]
No: 3-GRAPH-RPC-16 Rev: 6 CORE OPERATING LIMITS REPORT Page 10 of 12 Figure 2 Rod Bank Insertion Limits (Page 1 of 1)
(Four Loop Operation) 104 Step Overlap 230 220 210 200 190 180 170 Cz 160
-C 150 U) 140 a) 130 C')
120 110 0
100 90 0~ 80 Cr 70 60 50 40 30 20 10 0
0 10 20 30 40 50 60 70 80 90 100 Power (Percent of 3188.4 MWt)
No: 3-GRAPH-RPC-16 Rev: 6 CORE OPERATING LIMITS REPORT Page 11 of 12 Figure 3 Hot Channel Factor Normalized Operating Envelope (For S. G. Tube Plugging up to 10%)
(Page 1 of 1) 1.2 N
1 0
0 0.8 U-(U 0.
FQ = 2.30
- 0. 0.6 FQ k(z) Elev (ft) 0 2.30 1.0 0.0 N
2.30 1.0 6.0 0.4 -2.30 1.0 12.0 0
Z 0.2 0
0 2 4 6 8 10 12 Core Height (ft)
No: 3-GRAPH-RPC-16 Rev: 6 CORE OPERATING LIMITS REPORT Page 12 of 12 Figure 4 Axial Flux Difference Envelope Limits (Page 1 of 1) 1 95 90 85 80 C6 75 0~
70 0
CL 65 60 55 50
-40 -30 -20 -10 0 10 20 30 40 Axial Flux Difference