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Category:TECHNICAL SPECIFICATIONS
MONTHYEARML20211M6651999-09-0101 September 1999 Errata Page 4-45,reflecting Proposed Changes Requested in ML20211D1551999-08-20020 August 1999 Proposed Tech Specs Pages,Revising Degraded Voltage Relay as-left Setpoint Tolerances ML20210J1261999-07-29029 July 1999 Proposed Tech Specs Revising ESF Sys Leakage Limits in post-accident Recirculation Surveillance TSs ML20195E6201999-06-0404 June 1999 Proposed Tech Specs,Modifying Conditions Which Allow Reduction in Number of Means for Maintaining Decay Heat Removal Capability During Shutdown Conditions ML20206R1171999-05-13013 May 1999 Proposed Tech Specs Section 3.1.1,incorporating Administrative Updating & Changing Bases Statement ML20205H0781999-04-0101 April 1999 Proposed Tech Specs Adding LCO Action Statements,Making SRs More Consistent with NUREG-1430,correcting Conflicts or Inconsistencies Caused by Earlier TS Revs & Revising SFP Sampling ML20203B0511999-02-0202 February 1999 Proposed Tech Specs Expanding Scope of Systems & Test Requirements for post-accident RB Sump Recirculation ESF Systems & Increasing Max Allowable Leakage of TS 4.5.4 for Applicable Portions of ESF Systems Outside of Containment ML20196H5361998-12-0303 December 1998 Proposed Tech Specs Reflecting Decrease in RCS Flow Resulting from Revised Analysis to Allow Operation of Plant with 20% Average Level of SG Tubes Plugged Per SG ML20196G4861998-12-0303 December 1998 Non-proprietary Proposed Tech Specs,Consenting to Transfer & Authorization for Amergen to Possess,Use & Operate TMI-1 Under Essentially Same Conditions & Authorizations Included in Existing License ML20196F8661998-11-25025 November 1998 Proposed Tech Specs Revised Pages for TS Change 277 Changing Surveillances Specs for OTSG ISI for TMI Cycle 13 RFO Exams Which Would Be Applicable for One Cycle of Operation Only. with Certificate of Svc ML20154Q6271998-10-19019 October 1998 Proposed Tech Specs Adding Operability & SRs for Remote Shutdown Sys Similar to Requirements in NUREG-1430, Std Tech Specs - B&W Plants, Section 3.3.18 ML20154P8661998-10-19019 October 1998 Proposed Tech Specs,Providing Allowable RCS Specific Activity Limit Based on OTSG Insp Results Performed Each Refueling Outage ML20249B2421998-06-11011 June 1998 Proposed Tech Specs Re Alternate High Radiation Area Control ML20217J8201998-03-25025 March 1998 Proposed Tech Specs Page 6-1,reflecting Change in Trade Name of Owners & Operator of TMI-1 & Correcting Typo ML20217E5311998-03-23023 March 1998 Proposed Tech Specs Pages for Section 3.1.2 to Incorporate New Pressure Limits for Reactor Vessel IAW 10CFR50,App G for Period of Applicability Through 17.7 EFPY ML20217G0201997-10-0303 October 1997 Proposed Tech Specs Re Revised Pages 4-80 & 4-81 Previously Submitted ML20211F3781997-09-24024 September 1997 Proposed Tech Specs Revising Steam Line Break Accident Dose Consequence ML20211C3421997-09-19019 September 1997 Proposed Tech Specs Pages 3.8-3.9b to TS Section 3.1.4 Providing More Restrictive Limit of 0.35 Uci/Gram Dose Equivalent I-131 & Clarifying UFSAR Analysis ML20211C2431997-09-19019 September 1997 Proposed Tech Specs Re Decay Heat Removal Sys Leakage ML20210K0011997-08-14014 August 1997 Proposed Tech Specs Revising TMI-1 UFSAR Section 14.1.2.9 Environ Dose Consequences for TMI-1 Steam Line Break Analysis ML20141J4051997-08-12012 August 1997 Proposed Tech Specs Revising Surveillance Specification for Once Through Steam Generator Inservice Insp for TMI-1 Cycle 12 Refueling (12R) Exams Applicable to TMI-1 Cycle 12 Operation ML20198E7941997-07-30030 July 1997 Proposed Tech Specs Incorporating Addl Sys Leakage Limits & Leak Test Requirements for Systems Outside Containment Which Were Not Previously Contained in TS 4.5.4 Nor Considered in TMI-1 UFSAR DBA Analysis Dose Calculations for 2568 Mwt ML20151K2071997-07-25025 July 1997 Revised TS Page 6-19 Replacing Corresponding Page Contained in 970508 Transmittal of TS Change Request 264 ML20141E1491997-05-0808 May 1997 Proposed Tech Specs,Consisting of Change Request 264, Incorporating Addl NRC-approved Analytical Methods Used to Determine TMI-1 Core Operating Limits ML20140E2471997-04-21021 April 1997 Proposed Tech Specs 3.3.1.2,changing Required Borated Water in Each Core Flood Tank to 940 ft,4.5.2.1.b,lowering Surveillance Acceptance Criteria for ECCS HPI Flow to 431 Gpm & 3.3.1.1.f Re Operability of Decay Heat Valves ML20134M1211997-02-0707 February 1997 Proposed Tech Specs,Incorporating Certain Improvements from Revised STS for B&W plants,NUREG-1430 ML20133D2821996-12-24024 December 1996 Proposed Tech Specs 3.15.3 Re Auxiliary & Fuel Handling Bldg Air Treatment Sys ML20132F3301996-12-16016 December 1996 Proposed Tech Specs,Reflecting Change in Legal Name of Operator of Plant from Gpu Corp to Gpu Inc & Reflecting Plant License & TS Registered Trade Name of Gpu Energy ML20135D0991996-12-0303 December 1996 Proposed Tech Specs Incorporating Certain Requirements from Revised B&W Std TS,NUREG-1430 ML20135C5321996-12-0202 December 1996 Proposed Tech Specs Re Relocation of Audit Frequency Requirements ML20128H4121996-10-0303 October 1996 Errata to Proposed Ts,Adding Revised Table of Contents & Making Minor Editorial Corrections ML20117H0451996-08-29029 August 1996 Proposed Tech Specs,Consisting of Change Request 257, Incorporating Certain Improvements from STS for B&W Plants (NUREG-1430) ML20113D1971996-06-28028 June 1996 Proposed Tech Specs,Consisting of Change Request 259, Allowing Implementation of Recently Approved Option B to 10CFR50,App J ML20112C8791996-05-24024 May 1996 Proposed Tech Specs Re Pages for App A.Certificate of Svc Encl ML20101R1571996-04-10010 April 1996 Proposed Tech Specs,Revising Addl Group of Surveillances in Which Justification Has Been Completed ML20100J6931996-02-22022 February 1996 Proposed Tech Specs,Consisting of Change Request 254, Revising Proposed TS Page 4-46 on Paragraph 4.6.2 That Provides Addl Testing Requirements in Case Battery Cell Parameters Not Met ML20095J2311995-12-21021 December 1995 Proposed Tech Specs,Raising Low Voltage Action Level to 105 Volts DC ML20092K5471995-09-20020 September 1995 Revised Tech Spec Pages 4-31,4-32 & 4-33,incorporating Change in TS Section 4.4.1.5 ML20091L2911995-08-23023 August 1995 Proposed Tech Specs Page 6-11a,incorporating Ref to 10CFR20.1302 ML20087F4921995-08-11011 August 1995 Proposed TS Section 3.2 Re Makeup,Purification & Chemical Addition Sys Requirements ML20085N1601995-06-22022 June 1995 Proposed Tech Specs Revising Replacement Pages in Package & Remove Outdated Pages,In Response to NRC Request for Addl Info ML20084L4351995-06-0101 June 1995 Proposed TS 5.3.1.1,describing Use of Advanced Clad Assemblies ML20084N3501995-06-0101 June 1995 Proposed Tech Specs,Deleting RETS & Relocating TSs Per Guidance in GL 89-01 & NUREG-1430 ML20084B3361995-05-24024 May 1995 Proposed Tech Specs Re Change in Surveillance Test Requirements for source-range Nuclear Instrumentation ML20083N7841995-05-17017 May 1995 Proposed Tech Specs,Consisting of Change Requests 252, Removing Chemical Addition Sys Requirements from TS to COLR ML20079A7811994-12-23023 December 1994 Proposed Tech Specs Page 3-32a ML20069A6781994-05-20020 May 1994 Proposed Tech Specs,Supporting Cycle 10 Control Rod Trip Insertion Time Testing ML20065M6831994-04-19019 April 1994 Proposed Tech Specs,Reflecting Deletion of Audit Program Frequency Requirements ML20065K0451994-04-11011 April 1994 Proposed Tech Specs Reflecting Relocation of Detailed Insp Criteria,Methods & Frequencies of Containment Tendon Surveillance Program to FSAR & Providing Direct Ref to Existing Tendon Surveillance Program ML20073C7731994-03-22022 March 1994 Proposed Tech Specs Re Control Rod Trip Insertion Time Testing 1999-09-01
[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20211M6651999-09-0101 September 1999 Errata Page 4-45,reflecting Proposed Changes Requested in ML20211D1551999-08-20020 August 1999 Proposed Tech Specs Pages,Revising Degraded Voltage Relay as-left Setpoint Tolerances ML20210S7691999-08-12012 August 1999 Rev 10 to 6610-PLN-4200.02, TMI Emergency Dose Calculation Manual ML20210J1261999-07-29029 July 1999 Proposed Tech Specs Revising ESF Sys Leakage Limits in post-accident Recirculation Surveillance TSs ML20195E6201999-06-0404 June 1999 Proposed Tech Specs,Modifying Conditions Which Allow Reduction in Number of Means for Maintaining Decay Heat Removal Capability During Shutdown Conditions ML20206R1171999-05-13013 May 1999 Proposed Tech Specs Section 3.1.1,incorporating Administrative Updating & Changing Bases Statement ML20206R6531999-05-13013 May 1999 Rev 39 to TMI Modified Amended Physical Security Plan ML20205H0781999-04-0101 April 1999 Proposed Tech Specs Adding LCO Action Statements,Making SRs More Consistent with NUREG-1430,correcting Conflicts or Inconsistencies Caused by Earlier TS Revs & Revising SFP Sampling ML20204B5291999-03-12012 March 1999 Rev 9 to 6610-PLN-4200.02, TMI Emergency Dose Calculation Manual (Edcm) ML20203B0511999-02-0202 February 1999 Proposed Tech Specs Expanding Scope of Systems & Test Requirements for post-accident RB Sump Recirculation ESF Systems & Increasing Max Allowable Leakage of TS 4.5.4 for Applicable Portions of ESF Systems Outside of Containment ML20196G4861998-12-0303 December 1998 Non-proprietary Proposed Tech Specs,Consenting to Transfer & Authorization for Amergen to Possess,Use & Operate TMI-1 Under Essentially Same Conditions & Authorizations Included in Existing License ML20196H5361998-12-0303 December 1998 Proposed Tech Specs Reflecting Decrease in RCS Flow Resulting from Revised Analysis to Allow Operation of Plant with 20% Average Level of SG Tubes Plugged Per SG ML20196F8661998-11-25025 November 1998 Proposed Tech Specs Revised Pages for TS Change 277 Changing Surveillances Specs for OTSG ISI for TMI Cycle 13 RFO Exams Which Would Be Applicable for One Cycle of Operation Only. with Certificate of Svc ML20154P8661998-10-19019 October 1998 Proposed Tech Specs,Providing Allowable RCS Specific Activity Limit Based on OTSG Insp Results Performed Each Refueling Outage ML20154Q6271998-10-19019 October 1998 Proposed Tech Specs Adding Operability & SRs for Remote Shutdown Sys Similar to Requirements in NUREG-1430, Std Tech Specs - B&W Plants, Section 3.3.18 ML20154D5491998-10-0101 October 1998 Cancellation Notification of Temporary Change Notice 1-98-0066 to Procedure 6610-PLN-4200.02 ML20206C0911998-09-0101 September 1998 Rev 17 to Odcm ML20249B2421998-06-11011 June 1998 Proposed Tech Specs Re Alternate High Radiation Area Control ML20216E9751998-04-13013 April 1998 Emergency Dose Assessment Users Manual, for Insertion Into Rev 7 of Edcm ML20216E9491998-04-0909 April 1998 Rev 7,Temporary Change Notice 1-98-003 to 6610-PLN-4200.02, Edcm, Changing Pages 2 & 57 & Adding New Emergency Dose Assessment Users Manual ML20217J8201998-03-25025 March 1998 Proposed Tech Specs Page 6-1,reflecting Change in Trade Name of Owners & Operator of TMI-1 & Correcting Typo ML20217E5311998-03-23023 March 1998 Proposed Tech Specs Pages for Section 3.1.2 to Incorporate New Pressure Limits for Reactor Vessel IAW 10CFR50,App G for Period of Applicability Through 17.7 EFPY ML20202B2061998-01-30030 January 1998 Rev 7,Temporary Change Notice 1-98-0013 to 6610-PLN-4200.02, Edcm ML20198T4721997-12-31031 December 1997 TMI-1 Cycle 12 Startup Rept ML20217G0201997-10-0303 October 1997 Proposed Tech Specs Re Revised Pages 4-80 & 4-81 Previously Submitted ML20211F3781997-09-24024 September 1997 Proposed Tech Specs Revising Steam Line Break Accident Dose Consequence ML20211C2431997-09-19019 September 1997 Proposed Tech Specs Re Decay Heat Removal Sys Leakage ML20211C3421997-09-19019 September 1997 Proposed Tech Specs Pages 3.8-3.9b to TS Section 3.1.4 Providing More Restrictive Limit of 0.35 Uci/Gram Dose Equivalent I-131 & Clarifying UFSAR Analysis ML20210K0011997-08-14014 August 1997 Proposed Tech Specs Revising TMI-1 UFSAR Section 14.1.2.9 Environ Dose Consequences for TMI-1 Steam Line Break Analysis ML20141J4051997-08-12012 August 1997 Proposed Tech Specs Revising Surveillance Specification for Once Through Steam Generator Inservice Insp for TMI-1 Cycle 12 Refueling (12R) Exams Applicable to TMI-1 Cycle 12 Operation ML20198E7941997-07-30030 July 1997 Proposed Tech Specs Incorporating Addl Sys Leakage Limits & Leak Test Requirements for Systems Outside Containment Which Were Not Previously Contained in TS 4.5.4 Nor Considered in TMI-1 UFSAR DBA Analysis Dose Calculations for 2568 Mwt ML20151K2071997-07-25025 July 1997 Revised TS Page 6-19 Replacing Corresponding Page Contained in 970508 Transmittal of TS Change Request 264 ML20217M7251997-06-22022 June 1997 Rev 16 to Procedure 6610-PLN-4200.01, Odcm ML20141E1491997-05-0808 May 1997 Proposed Tech Specs,Consisting of Change Request 264, Incorporating Addl NRC-approved Analytical Methods Used to Determine TMI-1 Core Operating Limits ML20140E2471997-04-21021 April 1997 Proposed Tech Specs 3.3.1.2,changing Required Borated Water in Each Core Flood Tank to 940 ft,4.5.2.1.b,lowering Surveillance Acceptance Criteria for ECCS HPI Flow to 431 Gpm & 3.3.1.1.f Re Operability of Decay Heat Valves ML20134M1211997-02-0707 February 1997 Proposed Tech Specs,Incorporating Certain Improvements from Revised STS for B&W plants,NUREG-1430 ML20133D2821996-12-24024 December 1996 Proposed Tech Specs 3.15.3 Re Auxiliary & Fuel Handling Bldg Air Treatment Sys ML20132F3301996-12-16016 December 1996 Proposed Tech Specs,Reflecting Change in Legal Name of Operator of Plant from Gpu Corp to Gpu Inc & Reflecting Plant License & TS Registered Trade Name of Gpu Energy ML20135D0991996-12-0303 December 1996 Proposed Tech Specs Incorporating Certain Requirements from Revised B&W Std TS,NUREG-1430 ML20135C5321996-12-0202 December 1996 Proposed Tech Specs Re Relocation of Audit Frequency Requirements ML20128H4121996-10-0303 October 1996 Errata to Proposed Ts,Adding Revised Table of Contents & Making Minor Editorial Corrections ML20117H0451996-08-29029 August 1996 Proposed Tech Specs,Consisting of Change Request 257, Incorporating Certain Improvements from STS for B&W Plants (NUREG-1430) ML20113D1971996-06-28028 June 1996 Proposed Tech Specs,Consisting of Change Request 259, Allowing Implementation of Recently Approved Option B to 10CFR50,App J ML20112C8791996-05-24024 May 1996 Proposed Tech Specs Re Pages for App A.Certificate of Svc Encl ML20138B4641996-05-0606 May 1996 Rev 14 to Procedure 6610-PLN-4200.01, Odcm ML20101R1571996-04-10010 April 1996 Proposed Tech Specs,Revising Addl Group of Surveillances in Which Justification Has Been Completed ML20100J6931996-02-22022 February 1996 Proposed Tech Specs,Consisting of Change Request 254, Revising Proposed TS Page 4-46 on Paragraph 4.6.2 That Provides Addl Testing Requirements in Case Battery Cell Parameters Not Met ML20096F0521995-12-31031 December 1995 TMI-1 Cycle 11,Startup Rept ML20095J2311995-12-21021 December 1995 Proposed Tech Specs,Raising Low Voltage Action Level to 105 Volts DC ML20092M1941995-09-21021 September 1995 TMI-1 Pump & Valve IST Program 1999-09-01
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!T"RCPCLITAIT EDISON COMPANY JERSEY CENTRAL PCWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC CCMPANY THREE MILE ISLAND NUCLEAR STATION UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Stecification Chance Request No. 56 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. EPR-50 for Three Mile Island Nuclear Station Unit 1. As a part of this request ,
proposed replacement pages for Appendix A are also included.
METROPOLITAN EDISON CCMPANY By Vice President Sworn and subscribed to me this M day of , 1977.
e Notary Pu kic
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$910 290 7 1480 254
4 U'iITED STATES OF AMERICA IIUCLEAR REGULATORY CCIGIISSIO:I IN THE GTTER OF DOCKET NO. 50-289 LICE:!SE NO. DPR-50 METROPOLITA'i EDISON COMPANY This is to certify that a copy of Technical Specification Change Request No. 56 to Appendix A of the Operating License for "'hree Mile Island Nuclear Station Unit 1, has, on the date given 'celow, been filed with the U. S. Nuclear Regulatory Conmission and been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennsylvania by deposit in the United States mail, addresced as follows: .
Mr. '4eldon B. Arehart Mr. Harr/' B. Reese, Jr.
Board of Supervisors of Board of County Conmissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Court House Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY By / b iice President Dated: May 2h,1977 1480 255
ihr<e Mile Island nuclear Station Unit 1 (TMI-1)
Operating License Uo. DPR-50 Docket Ko. 50-289, ,
Technical Specification Change Recuest No. 56 The licensee requests that the attached changed pages replace pages 3-33, 34, 3ha, 35 of Appendix A of the existing Technical Specification.
Reason for Proposed Chsnge The reason for this change is to provide uniformity in responding to abnormal CRD System conditions by specifying time limits for completing corrective action. This change vill also eliminate the need for a licensee event report when prompt corrective action is taken.
Safety Analysis Justifying Change The time limits stated in this proposed change are conservative with respect to B&W Standard Technical Specification and they provide assurance that the operating limits of the fuel vill not be exceeded. This proposed change does not involve an unreviewed safety questien in that (i) the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased, (ii) the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created; and (iii) the margin of safety defined in the basis for any technical specification is not reduced.
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1480 256
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3 5.2 co:: TROL 'sCD 0?ct? AliD Po'ER DISTRIBUTIoII LI: CTS Anoli cability This specification applies to power distribution and operation of control rods during power operation.
Objective To assure an acceptable core power distribution during power operation, to set a limit on potential reactivity insertion fron a hypothetical control rod
- ejection, and to assure core suberiticality after a reactor trip.
Specification 3 5.2.1 The available shutdown margin shall not be less than one percent Ak/k with the highest worth control rod fully withdrawn.
3 5 2.2 Operatie- vith inoperable rods:
- a. Operation with more than one inoperable rod as defined in Specification h.7.1 and h.7.2.3 in the safety or regulating rod banks shall not be permitted. If more than one rod becomes inoperable, the reactor shall be placed in hot shutdown within four hours. Operation may continue i~f the system is restored to no more than one inoperable rod and operability of the remaining rods is demonstrated within 2h hours by a 2" exercise.
- b. If a control rod in the regulating and/or safety rod banks is declared inoperable in the withdrawn position as defined in Specification Paragraph 4.7 1.1 and h.7 1.3, an evaluation shall be initiated immediately to verify the existence of one percent Ak/k hot shutdown margin. Boration may be initiated to increase the available rod worth either to compensate for the vorth of the inoperable rod or until the regulating Tahki~are fully withdrawn, whichever occurs first. Simultaneously, a program of exercising the remaining regulating and safety rods shall be initiated to verify operability.
- c. If within one hour of determination of an inoperable rod as defined in Specification h.7.1, it is not determined that a one percent Ak/k hot shutdown cargin exists cisbining the vorth of the inoperable rod with each of the othe rods, the reactor shall be brought to the hot standby condition until this cargin is established.
1480 ?57 3-23
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- d. Following the determination of an inoperable rod as defined in Specification h.7.1, all rods shall be exercised within 2h hourc and exercised weekly until the rod problem is solved.
- e. If a control rod in the regulating or cafety rod groups is declared inoperable per k.7 1.2, pover chall be reduced to 607, of the thermal power allowable for the reactor coolant pump combination within one hour.
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1480 258 3-33a
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- f. If s control rod in the regulating or axial power shaping groups in deelsred inoperable per Specification h.T.l.2.,
operation may continue provided the rods in the group are positioned such that the rod that was declared inoperable is maintained within allowable group average position limits of Specification 4.7.1.2.
- g. If the inoperable rod in Paragraph "e" above is in groups 5, o, 7, or 8, the other rods in the group may be trimmed to the same positicn. Normal operation of 100 percent of the thermal power allowable for the reactor coolant pump combination may then continue provided tnat the rod that was declared inoperable is maintained witPla allowable group average position limits in 3.5 2 5 3.5 2.3 The worth of single inserted control rods during criticality are limited by the restrictions of Specification 3.13.5 and the control Rod Position Limits defined in Specification 3.5.2 5 3 5 2.h Quadrant tilt:
- a. Except for physics tests the quadrant tilt shall not exceed
+2.665 as datermined using the full incore detector system.
- b. When the fu11 incore detector system is not available and except for physics tests quadrant tilt shall not exceed +1.hT5
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as determined using the minimum incore detector system.
- c. When neither incore detector system above is available and except for physics tests quadrant tilt shall not exceed +0.81%
as determined using the power range channels displayed on the console for each quadrant (out of core detector system).
- d. Except for physics tests if quadrant tilt exceeds the tilt limit power shall be reduced immediately to below the power level cutoff (see Figures 3 5-2A, 3 5-2B and 3.5-2C)T "MdFe-over, the power level cutoff value shall be reduced 2 percent for each 1 percent tilt in excess of the tilt limit. For less than four pump operation, thermal power shall be reduced 2 percent of the thermal power allovable for the reactor coolant pump combination for each 1 percent tilt in excess of the tilt limit.
- e. Within a period of h hours, the quadrant power tilt shall be reduced to less than the tilt limit except for physics tests, or the following adjustments in setpoints and limits shall te made:
- 1. The protection c/ stem reactor pover/ imbalance envelope trip setpoints shs11 te reduced 2 percent in pcVer for ea= rercen: t11:.
3-3' 1480 259
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- 2. 7te control rod group withdrawal limits (Figures 3.5-2A, 3.5-2B, 3.5-2C, 3.5-23, 3 5-2E, 3 5-2F, 3.5-2K, 3.5-2L, and 3.5-2M) shall be reduced 2 percent in power for each 1 percent tilt in excesc of the tilt limit.
- 3. The operational imbalan:e limits (Figure 3.5-2G, 3.5-2H and 3.5-2I) shall be reduced 2 percent in power for each 1 percent tilt in excess of the tilt limit.
- f. Except for physics or diagnostic testing, i' quadrant tilt is in excess of +25 725 determined using the full incore detector system (FIT), or +2k.09% determined using the minimum incore '
detector system (MIT) if the FIT is not available, or +21 39%
determined using the out of core detector system (0"') when neither the FIT nor MIT are.available, the reactor will be placed in the hot shutdown condition. Diagnosite testing during power operation with c. quadrant tilt is permitted provided that the thermal power allowable is restricted as stated in 3.5.2.h.d above.
- g. Quadrant tilt shall be monitored on a minimum frequency of once every two hours during power operation above 15 percent of rated power.
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O ~G j 3 5.2.5 Control Rod Fositions: g
[fiY'Ci)i/d
- a. Operating rod group overlap shall not exceed 25% +5% between two sequential groups except for physics tests. If overlap exceeds 25% +5%, the reactor shall be placed in hot shutdown within four hours unless the overlap condition is corrected.
- b. Position limits are specified for regulating and axial power shaping centrol rods. Except for physics tests or exercising control rods, the regulating control rod insertion /vithdrawal limits are specified on Figures 3 5-2A, 3.5-2B, and 3 5-2C for four pump operation and Figures 3 5-2D, 3.5-2E, and 3.5-2F for three or two pump operation. Also excepting physics tests or exercising control rods, the axial power shuping control rod insertion /vithdrawal limits are specified on Figures 3.5-2K, 3 5-2L, and 3 5-2M. If any of these control rod position limits are exceeded, corrective measures shall be taken immedi-ately to achieve an acceptable control rod position. Acceptable control rod positions shall be attained within four hours.
- c. Except for physics tests, power shall not be increased above the power level cutoff (See Figures 3 5-2A, 3 5-2B and 3.5-2C) unless the xenon reactivity is within 10% of the equilibrium value for operation at rated power and asymptotically approach-ing stability.
- d. Core imbalance shall be monitored on a minimum frequency of once every two hours during power operation above 40% of rated power. Except for physics tests, corrective measures (reduction
- of imbalance by APSR movements and/or reduction in reactor power) shell be taken to maintain operation within the envelope define'l by Figures 3 5-2G, 3.5-2H and 3 5-2I. If the imbalance is not within the envelope defined by Figures 3 5-2G, 3.5-2H and 3 5-21 corrective measures shall be taken to achieve an acceptable imbalance. If an acceptable imbalance is not achieved within four hours, reactor power shall be reduced until imbalance limits are met. _ _ _ . - _
- e. Safety rod limits are given in 3.1.3 5 3 5 2.6 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent.
3.5 2 7 A power map shall be taken at periodic intervals of 10 to 30 full power days using the incore instrumentation detection system, to verify that the power distribution is within the limits shown in Figure 3. 5-2I.
ssses 1480 261 The power-imbalance envelope defined in Figures 3 5-2G, 3 5-2H, and 3.5-2I is buJad on IOCA ana]yses which have defined the taximum linear heat rate (see sicure 3.5-2J ) such that the taximum clad terperature vill not exceed the
.inal Acceptance Criteria (2200F). Operation outside of the power imbalance envelope alone does not constitute a situstion that would cause the Final Acceptance Criteria to be exceedeJ should s LOCA recur. The power impalance envelope represent; the boundary of opers'. en l'.mited by the Final Acceptance Criteria only 12 the con:rcl rods are at ;he withdrawal /incertion limits as defined by 2-1