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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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UNITED STATFJi OF AMERICA NUCLEAR REGULATORY Cola 4ISSION IN THE MATTER OF DOCKET NO. 50-289 LICENSE NO. DPR-50 METROPOLITAN EI'ISON CCITANY This is to certify that a copy of Technical Specification Change Request No. 36, Amendment 1 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with the U. S. Nuclear Regulatory Commiscion and been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Penncylvania by deposit in the United States mail, addressed ac follows:
Mr. '4eldon B. Arehart Mr. Harry B. Reese, Jr.
Board of Supervicora of Board of County Commissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Court House Middletowri, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY
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METROPOLITAN EDISON COMPANY ! o-- t JERSEY CENTRAL POWER & LIGHT COMPANY h, 19; c./ '
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PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request Nc. 36 Amendment 1 Thir Technical Specification Change Request is submitted in support of request ',o change Appendix A to Operating License No. DPR-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 COMPANY 0
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Vice President W
Sworn and subscribed to me this day of , 1977 V /
NotaryPut[c 45 1469
Metropolitan Edison Company (Met-Ed)
Three Mile Island Nuclear Station Unit 1 (TMI-1)
Docket No. 50-289 Operating License No. DPR-50 Technical Scecification Chance Request No. 36, Amendment 1 The Licensee requests that the attached pages replace the corresponding existing Technical Specifications pages. This amendment implements the recommendation made in our letter of October 19,1976. (GQL 1407)
Reasons for Proposed Change It has been discovered that:
The fuel densification penalty was not properly incorporated into technical specifications prepared for cycle 2. Proper incorporation of this penalty would affect DNB based pressure-temperature limit curves such that they would be more restrictive.
A new rod bow penalty has been imposed as a result of Westinghouse experiments revealing a heretofore not considered phenomenon.
Babcock & Wilcox calculations confirmed that elimination of the internal vent valve bypass. flow penalty, as authorized by Nuclear Regulatory Commission letter of March 10, 1976, would compensate for this error and the imposed rod bov penalty. A more detailed discussion of =argins available for off-setting rod bow effects is presented in GQL lh07 Thus , eliminaticn of the internal vent valve flow penalty will allow continued use of present pressure temperature limits (Figure 2.3-1)
As a prerequisite for eliminating the vent valve flow penalty, the Commission required in its letter of March 10, 1976, "... testing to be conducted each refueling outage to confirm that no vent valve is stuck in an open position and that each va:.ve continues to exhibit co=plete freedom of movement." This surveillance requirement was performed during the last refueling outage, and is planned for the upcoming refueling outage. This proposed change incorporates this surveillance requirement into technical specifications.
Note: The proposed technical specification b.16 included in Technical Specification Change Request No. 13 (still under review) is no lenger needed, due to equipment modification. Therefore, Technical Specification Change Pequest No.13 has been retracted.
Safety Analysis Justfyinc P'coposed Chance Elimination of the vent valve flow penalty has been authorized by the Conmission.
Revised densification analysis indicates that the correct penalties are 5.935 DNBR (versus 1.885 in the Reload Report) and 3.h7% power peeking relative to DNBR (versus 1.065 quoted in the Relcad Repert).
}kb9
The variable low pressure trip setpoint for cycle 2 operation is based on the four pump open vent valve pressure-temperature limit curve presented in figure 2.1.3 of the present Technical Specifications (Curve 1).
Curves 2 and 3 represent the corresponding limits for 3 and 2 pump operation, respectively. Each curve is based on the assumption that the reactor is operating at the maximum achievable power level for that pump operating condition. In the original cycle 2 submittal (and in the cycle one technical specifications), Curve 1 incorporated the open vent valve penalty, while curves 2 and 3 did not. That is, the four pump limit curve was based upon operation with one vent valve open while the three and two pump limit curves assumed all vent valves remained shut. In revising Figure 2.1-3 to incorporate the corrected DUBR densification penalty, the basis for the four pump limit j curve was changed to eliminate the vent valve penalty. The combined effect l vas to move curve 1 to the right. However, in order to compensate for rod bov l effects, curve 1 was left unchanged for this submittal. The revised curves 2 and 3 have incorporated the new red bow and the increased densification penalty, therefore, the curves moved slightly to the left.
! The flux / flow trip setpoint for cycle 2 (1.08) is based on the one pump coastdown analysis. When the revised densification penalty is incorporated and the vent valve penalty is eliminated, the thermal-hydraulic limiting flux / flow setpoint is greater than 1.12 (this limit must be at least 1.11 to justify the tech spec setpoint of 1.08). It can also be shown that a thermal-hydraulic limit of 1.11 on the flux / flow setpoint can be justified by taking credit for 1/2 of the vent valve penalty.
The error found in the TMI-1, Cycle 2 DNBR densification penalty calculations resulted from the use of inconsistent heat flum (flux shape) and enthalpy rise in evaluating the DNER densification penalty. This error only affects the PT envelope and flux / flow ratio.
Based upon the above, it is determined that this change does not constitute a threat to the health and safety of the public, nor does it involve an unreviewed safety question.
I .47 1469 i
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TABLE OF CONTEiTS
( Section Page 4.6 DIERGENCY POWER SYSTE4 PERIODIC TESTS h-46 h.7 REACTOR CONTROL ROD SYSTE4 TESTS h- 48 4.7.1 CONTROL ROD DRIVE SYSTDI FUNCTIONAL TESTS h-48 h.7.2 CONTROL ROD PROGRM4 VERIFICATION h-50 h.8 MAIN STEAM ISOLATION VALVES h-51 h.9 DIERGENCY FEEDWATER PU'4PS PERIODIC TESTING k-52 h.9 1 TEST h-52 h.9.2 ACCEPTANCE CRITERIA h-52 h.10 REACTIVITY ANOMALIES h-53 h.ll SITE ENVIRONMENTAL RADIOACTIVITY SURVEI k-5h h.12 CONTROL ROOM FILTERING SYSTDi h-55 4.12.1 OPERATING TESTS h-55 h.12.2 FILTER TESTS h-55 h.13 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE h-56 h.lh REACTOR BUILDING PURGE EXHAUST SYSTDi h-57 h.15 MAIN STEAM SYSTDI INSERVICE INSPECTION k-58 h.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE h-59 5 DESIGN FEATURES 5-1 51 SITE 5-1 5.2 CONTAINMENT 5-2 5.2.1 REACTOR BUILDING 5-2
! ( 5.2.2 REACTOR BUILDING ISOLATION SYSTEM 5-3 l 5.3 REACTOR 5-h 5.3.1 REACTOR CORE 5-k 5.3.2 REACTOR COOLANT SYSTD4 5-4 5.h NEW AND SPENT FUEL S0TRAGE FACILITIES 5-6 5.h.1 NEW FUEL STORAGE 5-6 5.h.2 SPEiT FUEL STORAGE 5-6 5.5 AIR INTAKE TUNNEL FIRE PROTECTION SYSTE4S 5-8 1
6 ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 6-2 6.2.1 0FFSITE 6-2 6.2.2 FACILITY STAFF 6-2 6.3 STATION STAFF QUALIFICATIONS 6-3 6.4 TRAINING 6-3 6.5 REVIEW & AUDIT 6-3 6.5.1 PLAliT OPERATIONS REVIEW COMMITTEE (PORC) 6-3 6.5.2.A MET-ED CORPORATE TECHIIICAL SUPPORT STAFF 6-5 6.5 2.B GENERAL OFFICE REVIEW BOARD (GCRB) 6-7 6.6 REPORTABLE OCCURRENCE ACTION 6-10 6.7 OCCURRENCES INVOLVING A SAFETY LIMIT VIOLATION 6-10a 6.8 PROCEDURES 6-11
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The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level
( trip set point produced by the power to flow ratio provides overpower D!G protection fer all modes of pump operation. For every flow rate there is a maximum pernissible power level, and for every power level there is a minimum permissible lov flow rate. Typical power level and lov flow rate combinations for the pump situations of Table 2.3-1 are as follows:
- 1. Trip would occur when four reactor coolant pumps are operating if power is 108 percent and reactor flow rate is 100 percent, or flow rate is 92.6 percent and power level is 100 percent.
- 2. Trip would occur when three reactor coolant pumps are operating if power is 80.7 percent and reactor flow rate is Th.7 percent or flow rate is 69 2 percent and power level is 75 percent.
- 3. Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 52.9 percent and reactor flow rate is h9.2 percent or flow rate is h5.h percent and the power level is E9 percent.
The flux / flow ratios account for the maximum calibration and instrumentation errors and the maximum variation frcm the average value of the RC flow si6nal in such a manner that the reactor protective system receives a conservative indication of the RC flow.
No penalty in reactor coolant flow through the core was taken for an cpen i core vent valve because of the core vent valve survef.llance program during f
\ each refueling outage.
For safety analysis calculations the maximum calibration and instrumentation errors for the power level were used.
The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either pcVer peaking kW/ft limits or DIGR limits. The reactor power imbalance (pover in the top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of Figure 2.3-2 are produced. The pover-to-flow ratio reduces the power level trip and associated reactor power / reactor power-imbalance boundaries by 1.08 i percent for a one percent flow reduction.
- b. Pump monitors The redundant pump monitors prevent the minitram core DIGR frcs
- decreasing belov 1.3 by tripping the reactor due to the loss of
- reactor coolant pu=p(s). The pu=p monitors also restrict the I power level for the number of pumps in operation. .
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' c. Reactor coolant system pressure l
l During a startup accident fro = low power or a slow rod withdrawal i frcm high power, the system high pressure trip set point is reached before the nuclear overpower trip set point. The trip setting I limit shown in Figure 2.3-1 for high reactor coolant system pressure (2355 psig) has been established to maintain the system l
pressure below the safety limit (2750 psig) for any design transient.
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h.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE
( Aeolicability Applies to Reactor Internals Vent Valves.
Objective To verify that no reactor internals vent valve is stuck in t. open position and that each valve continues to exhibit freedom of move =ent.
Specification h.16.1 At intervals not exceeding the refueling interval, each reactor internals vent valve vill be tested to verify
, that no valve is stuck in the open position and that each j valve continues to exhibit freedom of movement.
Bases i t i Verifying vent valve freedom of movement insures that coolant flow does not bypass the core through reactor internals vent valves during operation and therefore insures the conservatism of Core Protection Safety limits as delineated in figures 2.1-1 and 2.1-3, and the flux / flow trip setpoint.
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l 1600 620 640 660 560 580 600 Reactor Outlet Temperature. *F REACTOR COOLANT FLOW POWER PUMPS OPERATING (TYPE OF LIMIT)
CURVE (L6S/HR) 139.8 x 106 (100",)* 112% Four Pumps (DNBR Limit) 1 66.7% inree Pumps (DN8R Limit) 2 104.5 x 106 (74.7",)
59.1% One Pump in Eacn loop (Quality Limit) 3 68.8 x 106 (49.2%)
- 106.5% of Cycle 1 Design Flow TMl-1, UNIT I, CYCLE 2 CORE PROTECTI0tl SAFETY Figure 2.1 3 1469 352