Reactor Protection & Control Process Instrumentation Replacement Project at DC Cook Nuclear Plant Units 1 & 2 TS Compliance Assessment.| ML17329A717 |
| Person / Time |
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| Site: |
Cook  |
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| Issue date: |
12/15/1992 |
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| From: |
AMERICAN ELECTRIC POWER SERVICE CORP. |
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| To: |
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| Shared Package |
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| ML17329A705 |
List: |
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| References |
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| 2985-SKF-01, 2985-SKF-01-R00, 2985-SKF-1, 2985-SKF-1-R, NUDOCS 9212180082 |
| Download: ML17329A717 (6) |
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Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML17335A5451999-09-28028 September 1999
Rev 1 to Containment Sump Level Design Condition & Failure Effects Analysis for Potential Draindown Scenarios.
ML17326A1481999-09-17017 September 1999
Independent Review of Control Rod Insertion Following Cold Leg Lbloca,Dc Cook,Units 1 & 2.
ML17335A5461999-08-0202 August 1999
Rev 0 to Evaluation of Cook Recirculation Sump Level for Reduced Pump Flow Rates.
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Unit 1 Steam Generator Layup Chemistry Out of Specification During 970924 Through 971231.
ML17325B4511998-12-15015 December 1998
Rev 1 to ER-98-009, Preliminary Waste Characterization of DC Cook SG Lower Assemblies.
ML17335A2741998-09-0101 September 1998
Rev 0 to ER-98-009, Preliminary Waste Characterization of DC Cook SG Lower Assemblies.
ML17335A1921998-07-10010 July 1998
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Containment Sump Operability Determination.
ML17333B0921997-10-0909 October 1997
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ML17333B0041997-08-14014 August 1997
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ML17333A9561997-07-21021 July 1997
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Generators U1R97 2 Volt Interim Plugging Criteria Rept.
ML17333A5641996-08-31031 August 1996
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DC Cook Nuclear Plant USI A-46 Safe Shutdown Equipment List Rept.
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ML17333A3321995-07-31031 July 1995
DC Cook Nuclear Plant Boric Acid Concentration Reduction.
ML17333A9051995-07-25025 July 1995
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ML17333A9061995-06-0505 June 1995
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ML17332A5851995-02-28028 February 1995
DC Cook Nuclear Plant Units 1 & 2 Addendum to Seismic PRA Notebook.
ML17332A7121994-12-0505 December 1994
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Rev 0 to HI-941183, Spent Nuclear Fuel Pool Thermal- Hydraulic Analysis Rept for DC Cook Nuclear Plant.
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ML17331B1331993-12-31031 December 1993
DC Cook Units 1 & 2 Main Steam Safety Valve Lift Setpoint Tolerance Relaxation.
ML17331B1411993-12-0909 December 1993
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ML17331B1961993-09-30030 September 1993
Control Rod Misalignment Analysis.
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DC Cook Nuclear Plant Hydrogen Control Evaluation Summary Rept.
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Engineering Analysis of Grounding Issues Reactor Protection Process Control Group Replacement Project DC Cook Nuclear Plants Units 1 & 2.
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Reactor Protection & Control Process Instrumentation Replacement Project at DC Cook Nuclear Plant Units 1 & 2 Reactor Protection Sys Functional Diversity Assessment.
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Reactor Protection & Control Process Instrumentation Replacement Project at DC Cook Nuclear Plants Units 1 & 2 Qualification Compliance.
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1999-09-28
[Table view]
Category:TEXT-SAFETY REPORT
MONTHYEARML17335A5641999-10-18018 October 1999
LER 99-024-00:on 990708,literal TS Requirements Were Not Met by Accumlator Valve Surveillance.Caused by Misjudgement Made in Conversion from Initial DC Cook TS to W Std Ts.Submitted License Amend Request.With 991018 Ltr
ML17335A5531999-10-0707 October 1999
LER 99-023-00:on 990907,inadequate TS Surveillance Testing of ESW Pump ESF Response Time Noted.Caused by Inadequate Understanding of Plant Design Basis.Surveillance Tests Will Be Revised & Implemented
ML17335A5631999-09-30030 September 1999
Monthly Operating Rept for Sept 1999 for DC Cook Nuclear Plant,Unit 1.With 991012 Ltr
ML17335A5621999-09-30030 September 1999
Monthly Operating Rept for Sept 1999 for DC Cook Nuclear Plant,Unit 2.With 991012 Ltr
ML17335A5481999-09-30030 September 1999
Non-proprietary DC Cook Nuclear Plant Units 1 & 2 Mods to Containment Sys W SE (Secl 99-076,Rev 3).
ML17335A5451999-09-28028 September 1999
Rev 1 to Containment Sump Level Design Condition & Failure Effects Analysis for Potential Draindown Scenarios.
ML17326A1291999-09-17017 September 1999
LER 99-022-00:on 990609,electrical Bus Degraded Voltage Setpoints Too Low for Safety Related Loads,Was Discovered. Caused by Lack of Understanding of Design of Plant.No Immediate Corrective Actions Necessary
ML17326A1481999-09-17017 September 1999
Independent Review of Control Rod Insertion Following Cold Leg Lbloca,Dc Cook,Units 1 & 2.
ML17326A1211999-08-31031 August 1999
Monthly Operating Rept for Aug 1999 for Cook Nuclear Plant, Unit 2.With 990915 Ltr
ML17326A1201999-08-31031 August 1999
Monthly Operating Rept for Aug 1999 for Cook Nuclear Plant, Unit 1.With 990915 Ltr
ML17326A1121999-08-27027 August 1999
LER 99-021-00:on 990728,determined That GL 96-01 Test Requirements Were Not Met in Surveillance Tests.Caused by Failure to Understand Full Extent of GL Requirements. Surveillance Procedures Will Be Revised or Developed
ML17326A1011999-08-26026 August 1999
LER 99-020-00:on 990727,EDGs Were Declared Inoperable.Caused by Inadequate Protection of Air Intake,Exhaust & Room Ventilation Structures from Tornado Missile Hazards. Implemented Compensatory Measures in Form of ACs
ML17326A0911999-08-16016 August 1999
LER 99-019-00:on 990716,noted Victoreen Containment Hrrms Not Environmentally Qualified to Withstand post-LOCA Conditions.Caused by Inadequate Design Control.Reviewing Options to Support Hrrms Operability in Modes 1-4
ML17326A0771999-08-0404 August 1999
LER 98-029-01:on 980422,noted That Fuel Handling Area Ventilation Sys Was Inoperable.Caused by Original Design Deficiency.Radiological Analysis for Spent Fuel Handling Accidents in Auxiliary Bldg Will Be Redone by 990830
ML17335A5461999-08-0202 August 1999
Rev 0 to Evaluation of Cook Recirculation Sump Level for Reduced Pump Flow Rates.
ML17326A0871999-07-31031 July 1999
Monthly Operating Rept for July 1999 for DC Cook Nuclear Plant,Unit 1.With 990812 Ltr
ML17326A0861999-07-31031 July 1999
Monthly Operating Rept for July 1999 for DC Cook Nuclear Plant,Units 2.With 990812 Ltr
ML17326A0741999-07-29029 July 1999
LER 99-018-00:on 990629,determined That Valve Yokes May Yield Under Combined Stress of Seismic Event & Static,Valve Closed,Stem Thrust.Caused by Inadequate Design of Associated Movs.Operability Determinations Were Performed for Valves
ML17326A0661999-07-26026 July 1999
LER 99-017-00:on 990625,noted That Improperly Installed Fuel Oil Return Relief Valve Rendered EDG Inoperable.Caused by Personnel Error.Fuel Oil Return Valve Was Replaced with Valve in Correct Orientation.With 990722 Ltr
ML17326A0651999-07-22022 July 1999
LER 98-014-03:on 980310,noted That Response to high-high Containment Pressure Procedure Was Not Consistent with Analysis of Record.Caused by Inadequate Interface with W. FRZ-1 Will Be Revised to Be Consistent with New Analysis
ML17326A0491999-07-13013 July 1999
LER 99-016-00:on 990615,TS Requirements for Source Range Neutron Flux Monitors Not Met.Caused by Failure to Understand Design Basis of Plant.Procedures Revised.With 990713 Ltr
ML17326A0331999-07-0101 July 1999
LER 99-004-01:on 971030,failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed Was Noted.Caused by Ineffective Mgt of Tss.Chemistry Personnel Have Been Instructed on Requirement to Follow TS as Written
ML17326A0511999-06-30030 June 1999
Monthly Operating Rept for June 1999 for DC Cook Nuclear Plant,Unit 2.With 990709 Ltr
ML17326A0501999-06-30030 June 1999
Monthly Operating Rept for June 1999 for DC Cook Nuclear Plant,Unit 1.With 990709 Ltr
ML17326A0151999-06-18018 June 1999
LER 99-014-00:on 990521,determined That Boron Injection Tank Manway Bolts Were Not Included in ISI Program,Creating Missed Exam for Previous ISI Interval.Caused by Programmatic Weakness.Isi Program & Associated ISI Database Modified
ML17325B6421999-06-0101 June 1999
LER 99-013-00:on 990327,safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Have Led to ECCS Pump Failure.Caused by Inadequate Original Design Application of Si.Throttle Valves Will Be Developed
ML17325B6311999-06-0101 June 1999
LER 99-S03-00:on 990430,vital Area Barrier Degradation Was Noted.Caused by Inadequate Insp & Maint of Vital Area Barrier.Repairs & Mods Were Made to Barriers to Eliminate Degraded & Nonconforming Conditions
ML17326A0061999-05-31031 May 1999
Monthly Operating Rept for May 1999 for Dcp.With 990609 Ltr
ML17326A0071999-05-31031 May 1999
Monthly Operating Rept for May 1999 for DC Cook Nuclear Plant,Unit 2.With 990609 Ltr
ML17325B6351999-05-28028 May 1999
LER 99-S02-00:on 990428,vulnerability in Safeguard Sys That Could Allow Unauthorized Access to Protected Area Was Noted. Caused by Inadequate Original Plant Design.Mods Were Made to Wall Opening to Eliminate Nonconforming Conditions
ML17265A8231999-05-24024 May 1999
LER 98-037-01:on 990422,determined That Ice Condenser Bypass Leakage Exceeds Design Basis Limit.Caused by Pressure Seal Required by Revised W Design Not Incorporated Into Aep Design.Numerous Matl Condition Walkdowns & Assessments Made
ML17325B6001999-05-20020 May 1999
LER 99-012-00:on 990420,concluded That Auxiliary Bldg ESF Ventilation Sys Not Capable of Maintaining ESF Room Temps post-accident.Caused by Inadequate Control of Sys Design Inputs.Comprehensive Action Plan Being Developed
ML17325B5861999-05-10010 May 1999
LER 99-002-00:on 990415,discovered That TS 4.0.5 Requirements Were Not Met Due to Improperly Performed Test. Caused by Incorrect Interpretation of ASME Code.App J Testing Will Be Completed & Procedures Will Be Revised
ML17325B5811999-05-0404 May 1999
LER 99-011-00:on 990407,air Sys for EDG Will Not Support Long Operability.Caused by Original Design Error.Temporary Mod to Supply Makeup Air Capability in Modes 5 & 6 Was Prepared
ML17325B5771999-05-0303 May 1999
LER 99-010-00:on 990401,RCS Leak Detection Sys Sensitivity Not in Accordance with Design Requirements Occurred.Caused by Inadequate Original Design of Containment Sump Level. Evaluation Will Be Performed to Clearly Define Design
ML17335A5301999-04-30030 April 1999
Monthly Operating Rept for Apr 1999 for DC Cook Nuclear Plant,Unit 1.With 990508 Ltr
ML17335A5291999-04-30030 April 1999
Monthly Operating Rept for Apr 1999 for DC Cook Nuclear Plant,Unit 2.With 990508 Ltr
ML17325B5581999-04-16016 April 1999
LER 99-006-00:on 990115,personnel Identified Discrepancy Between TS 3.9.7 Impact Energy Limit & Procedure 12 Ohp 4030.STP.046.Caused by Lack of Design Basis Control.Placed Procedure 12 Ohp 4030.STP.046 on Administrative Hold
ML17325B5471999-04-12012 April 1999
LER 99-009-00:on 990304,as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit Occurred.Cause Investigation for Condition Has Not Been Completed.Update to LER Will Be Submitted,Upon Completion of Investigation
ML17325B5321999-04-0707 April 1999
LER 99-S01-00:on 990308,discovered That Lock for Vital Gate Leading to Plant 4KV Switchgear Area Was Nonconforming & Vulnerable to Unauthorized Access.Caused by Inadequate Gate Design & Inadequate Procedures.Mods Are Being Made to Gate
ML17325B5161999-04-0101 April 1999
LER 99-007-00:on 981020,calculations Showed That Divider Barrier Between Upper & Lower Containment Vols Were Overstressed.Engineers Are Currently Working on Analyses of Loads & Stress on Enclosures
ML17325B5491999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for DC Cook Nuclear Plant Unit 2.With 990408 Ltr
ML17325B5441999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for DC Cook Nuclear Plant,Unit 1.With 990408 Ltr
ML17325B5221999-03-29029 March 1999
LER 99-001-00:on 960610,degraded Component Cooling Water Flow to Containment Main Steam Line Penetrations,Identified on 990226.Caused by Inadequate Understanding of Design Basis.Additional Investigations Ongoing
ML17325B4801999-03-18018 March 1999
LER 99-004-00:on 971030,failure to Perform TS Surveillance Analyses of Rc Chemistry with Fuel Removed Was Noted.Cause of Event Is Under Investigation.Corrected Written Job Order Activities Used to Control SD Chemistry Sampling
ML17325B4741999-03-18018 March 1999
LER 99-005-00:on 940512,determined That Rt Breaker Manual Actuations During Rod Drop Testing Were Not Previously Reported.Caused by Lack of Training.Addl Corrective Actions,Including Preventative Actions May Be Developed
ML17325B5671999-03-0202 March 1999
Summary of Unit 1 Steam Generator Layup Chemistry from 980101 to 990218.
ML17325B4631999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for DC Cook Nuclear Power Station,Unit 2.With 990308 Ltr
ML17325B4621999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for DC Cook Nuclear Plant,Unit 1.With 990308 Ltr
ML17325B4571999-02-24024 February 1999
LER 99-003-00:on 990107,CR Pressurization Sys Surveillance Test Did Not Test Sys in Normal Operating Condition.Caused by Failure to Recognize Door 12DR-AUX415 as Part of CR Pressure Boundary.Performed Walkdown of Other Doors
1999-09-30
[Table view] |
Text
REACTOR PROTECTION AND CONTROL PROCESS INSTRUMENTATION REPLACEMENT PROJECT AT DONALD C. COOK NUCLEAR PLANT UNITS 1 AND 2 TECHNICAL SPECIFICATION COMPLIANCE ASSESSMENT REPORT NO. 2985-SKF-01, REV 0 Prepared by: Date ~ /<<8 IHX
/
Concurred by: Date L /
Approved by: Date 2< i < a.
9212i80082 921216 ppp AOOCK 050003is PDR ii l
p
~Sub'ect:
This Report is to summarize the impact of the Cook Nuclear Plant Reactor Protection System Upgrade on the existing Technical Specifications.
References:
Cook Nuclear Plant Technical Specifications (both Unit 1 5 2).
Calculation 12-RPC-01 Loop Uncertainty/Setpoint Calculation For The Reactor Coolant Flow.
Calculation 12-RPC-04 Loop Uncertainty/Setpoint Calculation For Pressurizer Pressure.
Hurst Engineering Report 2985-HEI-01; Response Time Evaluation.
Instrument Society of America (ISA) draft recommended practice ISA-DRP67.04, Part II, Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation (Draft 10) .
Westinghouse Menu Driven Setpoint Calculation Program (WCAP-12741).
Issue:
The Cook Nuclear Plant, Reactor Protection System is being upgraded to Foxboro SPEC 200/SPEC 200 MICRO equipment. Several of the existing Technical Specifications contain definitions, testing, configuration, surveillance, and operability requirements which are potentially effected by the upgrade of this instrumentation. This evaluation was performed to document the impact of the Reactor Protection Upgrade on these Technical Specifications.
~Scc e:
The Technical Specifications were reviewed in detail to identify all areas of potential impact. The following areas were identified by this review:
- 1. Channel Functional Test Definition
- 2. Indicated DNB and TAVG operating parameters
- 3. Time Response Testing
- 4. Surveillance Intervals
- 5. Channel Operability Requirements
- 6. Permissives and Interlocks
- 7. Remote Shutdown Instrumentation
- 8. Appendix R Requirements
- 9. Post Accident Monitoring
- 10. Reactor Trip and Engineered Safety Feature Trip Setpoints
- 11. Reactor Trip and Engineered Safety Feature Allowable Values This report provides a summary of the impact of the Reactor Protection and Report g2985-SKF-01, Rev. 0 Page 1 of 4
Control System upgrade on each of these Technical Specification topics.
~ ~
Discussion:
Channel Functional Test Definition Based on information from other Plants performing similar projects, the definitions section of the Technical Specifications was reviewed. In particular the definition, of Channel Functional Test was scrutinized.
The current definition recognizes analog and bistable channels.
Addition of a digital channel functional test definition, as done in some of the similar upgrades, would imply that there is some acceptance of equipment performance based on the input or monitoring of digital signals. This implication would not be correct. Even though the new instrumentation uses card level digital processors for signal processing, all communication between modules is by analog signal. All test points used for evaluating performance are located in the analog portions of the instrument loops. All inputs and outputs used as acceptance criteria for channel calibration, channel check, and channel functional tests are analog. The existing definition for Channel Functional Test, as well as the other definitions, are both appropriate and correct.
- 2. Indicated DNB and TAVG operating parameters (T. S 3.2. 5)
Technical Specification 3.2.5 provides values for Reactor Coolant System Average Temperature, Pressurizer Pressure and Reactor Coolant System Total Flow Rate, The values provided include an instrument error allowance for the indication portion of these instrument loops. Based on review of the Unit 2 Cycle 9 Technical Specification submittal (AEP:NRC:1071E) these allowances are 2.7 degrees for Average Temperature, 63 psi for Pressurizer Pressure, and 3.5% Reactor Coolant Flow. Calculation 12-RPC-01 rev 0 shows the new Reactor Coolant flow indication uncertainty to be 3.1%.
Calculation 2-RPC-04 rev 0 shows the Pressurizer Pressure indication uncertainty to be 49 psi. Both of these values are moderately conservative and acceptable. Uncertainties associated with Tavg Indication will be revised as a Unit 2 Cycle 10 specific parameter to reflect current assumptions relative to temperature streaming and calculation methods.
- 3. Time Response Testing Time Response Testing is discussed in Report Number 2985-HEI-01.
Manufacturers Specifications indicate that the new equipment will meet the current Technical Specification response time requirements.
Instrument performance will be verified both during acceptance testing and during post installation tests.
Report 82985-SKF-01, Rev. 0 Page 2 of 4
4, Surveillance Intervals The drift specifications of the new equipment are better than the installed equipment. This improvement may be sufficient to allow extension of some of the Surveillance intervals using the previously approved Westinghouse Owners Group Methodology; however, application for extended surveillance intervals will be deferred until an extensive amount of plant specific data has been collected and analyzed.
- 5. Channel Operability Requirements
- 6. Permissives and Interlocks Design of the new equipment maintains the same functions and logic as the base design. There are no changes in either the number of channels or the channel power supply orientation of the Functional Units listed in the current Technical Specifications. The new equipment does not incorporate any channel bypass features. Existing Technical Specification Tables 3.3-1 and 3.3-3 accurately reflect the configuration of the new equipment. Similarly the various permissives and interlocks noted in the notes throughout Technical Specification Section 3.4.3 are not effected.
- 7. Remote Shutdown Instrumentation
- 8. Appendix R Requirements
- 9. Post Accident Monitoring Current Technical Specification requirements for Remote Shutdown Instrumentation, Appendix R Remote Shutdown Instrumentation, and Post-Accident Instrumentation were reviewed. The new instrumentation does not add or delete any plant parameters. No indicators are added, deleted, or moved. No Surveillance Interval changes are required.
The Reactor Protection and Control System Upgrade will have no impact on compliance with these Technical Specification requirements.
- 10. Reactor Trip and Engineered Safety Feature Trip Setpoints All Trip Setpoints noted in tables 2.2-1, Reactor Trip System Instrumentation Trip Setpoints and table 3,3-4, Engineered Safety Feature Actuation System Instrumentation Trip Setpoints were evaluated. Functional Units associated with Neutron Flux were not recalculated because the Nuclear Instrumentation is not effected by this project. The remaining Functional Units for Unit 2 have been recalculated.
Trip setpoint calculations were based both on Instrument Society of America (ISA) draft recommended practice ISA-dRP67.04, Part II, Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation (Draft 10) and the Westinghouse Menu Driven Setpoint Calculation Program, WCAP-12741 (STEPIT) . The ISA methodology was used to provide a completely new set of calculations Report ¹2985-SICF-01, Rev. 0 Page 3 of 4
based on manufacturers published data, qualification test reports, Plant specific procedures and Plant specific practices. These calculations address the entire instrument loop (process connection through bistable output) and include inaccuracies associated with cable insulation losses. The Westinghouse methodology was used as a direct comparison between the current H-Line equipment and the SPEC 200/SPEC 200 MICRO equipment. The STEPIT databases were first benchmarked against the existing setpoint studies performed by Westinghouse. The rack portions of the databases were then modified to reflect the new equipment. Accuracies used in the STEPIT databases reflect analog signal accuracies, this assured both the applicability and fidelity of the Westinghouse methodology. The new values from both calculations were then compared with the existing values contained in the Technical Specifications. In all cases the calculated Trip Setpoints were conservative to the existing Technical Specification values.
Reactor Trip and Engineered Safety Feature Allowable Values All Allowable Values noted in tables 2.2-1, Reactor Trip System Instrumentation Trip Setpoints and table 3.3-4, Engineered Safety Feature Actuation System Instrumentation Trip Setpoints were evaluated. Functional Units associated with Neutron Flux were not recalculated because the Nuclear Instrumentation System is not effected by this project. Allowable Values associated with the remaining Functional Units using the same methodology described above for Trip Setpoints.
The existing Technical Specification Allowable Values are conservative approximations of current calculated allowances. Of the 16 Allowable values examined, 6 are based on the original license practice of a 1%
or 10 psi difference between the trip and allowable values. The remaining values are conservative approximations of the values calculated by STEPIT. The conservatism built into the existing values is sufficient to bound the calculation variance caused by the improved drift characteristics of the SPEC 200/SPEC 200 MICRO equipment.
==
Conclusion:==
The replacement of the existing Foxboro H-line Reactor Protection System Instrumentation with Foxboro SPEC 200/SPEC 200 M ICRO Instrumentation does not generate the need for a Technical Specification Amendment.
held(SKF$ 1590. WP Report g2985-SKF-Ol, Rev. 0 Page 4 of 4
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