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| SAFETY EVALUATION BY THE 3FFICE OF NUCLEAR REACTOR REGULATION RELATING TO TECHNICAL SPECIFICATION CHANGES AND UNIT 2/ CYCLE 2 RELOAD DUKE POWER COMPANY CATAWBA UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 INTRODUCTION By letter oated November 13, 1987 (Ref. 1), Duke Power Company (the 'icensee) requested Changes to the Technical Specifications for Catawba Nuclear Station Units 1 and 2, to reflect the Unit 2 refueling and the addition of the Boron Dilution Mitigation System for Unit 2. In addition deletion of Surveillance 4.3.3.12.1(b) which requires an ANALOG CHANNEL OPERATIONAL CHECK prior to startup is requested for both units. A second letter dated December 11, 1987 (Ref 2) provided a discussion of the Justification and No Significant Hazards Considerations. Additional infonnation and justification were provided in letters dated January 15 and 20, 1988 (Refs. 10 and 11). | | SAFETY EVALUATION BY THE 3FFICE OF NUCLEAR REACTOR REGULATION RELATING TO TECHNICAL SPECIFICATION CHANGES AND UNIT 2/ CYCLE 2 RELOAD DUKE POWER COMPANY CATAWBA UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 INTRODUCTION By letter oated November 13, 1987 (Ref. 1), Duke Power Company (the 'icensee) requested Changes to the Technical Specifications for Catawba Nuclear Station Units 1 and 2, to reflect the Unit 2 refueling and the addition of the Boron Dilution Mitigation System for Unit 2. In addition deletion of Surveillance 4.3.3.12.1(b) which requires an ANALOG CHANNEL OPERATIONAL CHECK prior to startup is requested for both units. A second {{letter dated|date=December 11, 1987|text=letter dated December 11, 1987}} (Ref 2) provided a discussion of the Justification and No Significant Hazards Considerations. Additional infonnation and justification were provided in letters dated January 15 and 20, 1988 (Refs. 10 and 11). |
| EVALUATION A. Unit 2 Cycle 2 Reload | | EVALUATION A. Unit 2 Cycle 2 Reload |
| : 1. General Design The Catawba Unit 2, Cycle 2. reactor core contains 193 Optimized Fuel Assemblies. During the Cycle 1/2 refueling 64 Region 1 fuel assemblies will be replaced with 64 Region 4 fuel assemblies. The Region 4 fuel is very similar to that used in Regions 1, 2, and 3. | | : 1. General Design The Catawba Unit 2, Cycle 2. reactor core contains 193 Optimized Fuel Assemblies. During the Cycle 1/2 refueling 64 Region 1 fuel assemblies will be replaced with 64 Region 4 fuel assemblies. The Region 4 fuel is very similar to that used in Regions 1, 2, and 3. |
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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20211A9791999-08-20020 August 1999 Safety Evaluation Granting Licensee Request for Approval of Proposed Relief from Volumetric Exam Requirements of ASME B&PV Code,Section Xi,For Plant,Unit 2 ML20209E4361999-07-0909 July 1999 SER Agreeing with Licensee General Interpretation of TS LCO 3.0.6,but Finds No Technical Basis or Guidance That Snubbers Could Be Treated as Exception to General Interpretation ML20196K6631999-07-0707 July 1999 Safety Evaluation Supporting Licensee 990520 Position Re Inoperable Snubbers ML20206P5201999-05-14014 May 1999 Safety Evaluation Accepting GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves ML20205S5551999-04-21021 April 1999 Safety Evaluation Accepting Response to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design Basis Accident Conditions ML20205N3651999-04-12012 April 1999 Safety Evaluation Accepting IPE of External Events Submittal ML20196J8351998-12-0808 December 1998 Safety Evaluation Granting Relief Request Re Relief Valves in Diesel Generator Fuel Oil Sys ML20196C0251998-11-27027 November 1998 SER Accepting Clarification on Calibration Tolerances on Trip Setpoints for Catawba Nuclear Station ML20196A6881998-11-25025 November 1998 Safety Evaluation Granting Relief Request 98-02 Re Limited Exam for Three Welds ML20247M0861998-05-21021 May 1998 SER Accepting 1997 Rev to Catawba UFSAR Submitted on 970925. Rev Added Analysis for Postulated Accident Involving Dropping of Sf Pool Weir Gate Onto Sf Assemblies ML20216E1771998-04-13013 April 1998 Safety Evaluation Accepting Relief Request 98-01 for Catawba Nuclear Station Units 1 & 2 from Requirements of ASME Boiler & PV Code for Second 10-year Interval Program for Inservice Testing of Pumps & Valves ML20217M4211998-04-0303 April 1998 Safety Evaluation Approving Request for Relief 97-04, non-code Repair Valves.Relief Granted Retroactively to Unit 1 & Expired Dec 1997.Relief for Unit 2 Will Expire at End of Cycle 9 Outage or Next Scheduled Outage Exceeding 30 Days ML20199A5521998-01-22022 January 1998 Safety Evaluation Re Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power Operated Gate Valves, for Cns,Units 1 & 2 ML20199A6421998-01-22022 January 1998 Safety Evaluation Accepting Proposed EALs Changes for Plant, Units 1 & 2.Concludes That Changes Consistent W/Guidance in NUMARC/NESP-007,w/variations as Identified & Accepted in Review & Meets Requirements of 10CFR50.47(b)(4) ML20198P9641998-01-15015 January 1998 SE Granting Relief Request 97-03 for Second 10-yr Interval Inservice Insp Program ML20198R9951997-10-30030 October 1997 Safety Evaluation Authorizing Request for Approval of Alternative to Exam Requirement of Reactor Vessel Shell Weld,Per 10CFR50.55a(g)(6)(ii)(A)(5) ML20198J7651997-10-15015 October 1997 Safety Evaluation Accepting 10-yr Interval Insp Program Plan Alternatives for Listed Plants Units ML20211F8801997-09-22022 September 1997 Safety Evaluation Supporting Second ten-year Interval Inservice Inspection Program Plan & Associated Requests for Relief for Catawba Nuclear Station Unit 1 ML20149K8281997-07-29029 July 1997 SER Granting Request for Exemption from Requirements of 10CFR70.24 for Units 1 & 2 ML20141E1121997-06-16016 June 1997 Safety Evaluation Approving Licensee Position That UFSAR Table 6-77 Be Revised to Identify SA-1 & SA-4 Instead of SA-3 & SA-6 as Containment Isolation Valves ML20148H2501997-06-0505 June 1997 Safety Evaluation Accepting Proposed Restructuring of Util Through Acquisition Of,& Merger W/Panenergy Corp ML20141G6701997-05-20020 May 1997 Safety Evaluation Accepting Proposed Alternative Use of TS Requirement for Code Class Snubbers ML20135B3051997-02-27027 February 1997 Safety Evaluation Granting Second 10 Yr ISI Program Plan & Associated Requests for Relief ML20134P2421997-02-20020 February 1997 Safety Evaluation Accepting TR BAW-10199P for Ref in Plants Licensing Documentation & Use in Licensing Applications ML20134L4081996-11-19019 November 1996 SER Accepting Performance of Plant Standby Nuclear Svc Water Pond ML20134G5551996-11-0707 November 1996 Safety Evaluation Accepting Proposed Application of BWU-Z CHF Correlation for Plants Mark-BW 17x17 Type Fuel ML20129E4851996-10-0101 October 1996 Safety Evaluation Recommending That Relief Request 96-02 Be Granted,Per 10CFR50.55a(g)(6)(i),per Request ML20056D9611993-07-30030 July 1993 SER Accepting Licensee 930325-0429 Submittals of Technical Info to Support Continued Operation of Facility for Remainder of Fuel Cycle 7 ML20055H9151990-07-27027 July 1990 Safety Evaluation Accepting Actions Taken to Resolve NRC Bulletin 88-002, Rapidly Propagating Fatigue Cracks in Steam Generators Tubes ML20248C0731989-08-0303 August 1989 Sser Accepting 880601,0909 & 890602 Changes to ATWS Mitigation Sys Actuation Circuitry for Plants ML20246L4221989-05-12012 May 1989 Safety Evaluation Supporting Revs 15 & 6 to Pump & Valve Inservice Testing Program & Relief Requests ML20150C2651988-06-28028 June 1988 Safety Evaluation Supporting Licensee Assessment to Fracture Toughness Requirements for Protection Against PTS Events,Per 10CFR50.61 ML20150C2611988-06-28028 June 1988 Safety Evaluation Supporting Licensee Assessment to Fracture Toughness Requirements for Protection Against PTS Events,Per 10CFR50.61 ML20154H2171988-05-18018 May 1988 Safety Evaluation Accepting Util 880414 Submittal Re Reload Startup Physics Test Program ML20148H1001988-01-22022 January 1988 Safety Evaluation Supporting Util 871113 Proposed Tech Spec Changes Reflecting Unit 2 Cycle 2 Refueling & Addition of Boron Dilution Mitigation Sys for Unit 2 ML20236M9631987-11-0606 November 1987 Safety Evaluation Accepting Util Proposed ATWS Mitigating Sys Actuation Circuitry for Facilities,Per 10CFR50.62(c)(1) & Pending Final Resolution of Tech Spec Issue ML20236D4451987-09-30030 September 1987 SER Re Licensee 870814 & 21 Responses Re Single Failure Potential in Nuclear Svc Water Sys.Nuclear Svc Water Sys Meets Requirements of GDC 5 & 44 Re Sharing & Provisions for Suitable Redundancy in Cooling Water Sys ML20236F5881987-07-29029 July 1987 Safety Evaluation Supporting Util 831104,841102,1231 & 851203 Responses to Items 3.1.1,3.1.2,3.2.1,3.2.2 & 4.5.1 of Generic Ltr 83-28 ML20215D2701987-06-11011 June 1987 SER Accepting Rev 13 to Unit 1 & Rev 4 to Unit 2 Pump & Valve Inservice Testing,Per Generic Ltr 83-28,Item 2.1 (Part 2) ML20214T4061987-06-0202 June 1987 Safety Evaluation Supporting Util 860424 Submittal Re Turbine Sys Maint Program for Early Detection of Cracking in Low Pressure Turbine Wheels ML20214M3361987-05-22022 May 1987 Safety Evaluation Supporting Util Rept Entitled, Rod Swap Methodology Rept for Startup Physics Testing ML20209B1511987-01-28028 January 1987 SER Supporting Util Responses to Generic Ltr 83-28,Item 4.5.2 Re Reactor Trip Sys on-line Testing ML20207Q5851987-01-15015 January 1987 SER Re Generic Ltr 83-28,Item 2.1 (Part 2) Concerning Vendor Interface Programs for Reactor Trip Sys Components.Util Response Acceptable.Item Closed ML20207N3011987-01-0808 January 1987 Safety Evaluation on Util 830309 & 851025 Requests for Relief from 10CFR50.55a Requirements Re Pump & Valve Inservice Testing Program.Relief Granted W/Listed Exceptions ML20207P6341986-08-31031 August 1986 Safety Evaluation Accepting Util 851120 & 860324 Proposals to Eliminate Arbitrary Intermediate Pipe Breaks in Select List of High Energy Piping Sys ML20199L3941986-07-0202 July 1986 SER Providing Final Conclusions & Recommendations from Evaluation of Tdi Owners Group Program to Validate & Update Quality of Tdi Diesel Generators ML20211E4251986-06-10010 June 1986 Draft SER Re Util 850624 Response to Generic Ltr 83-28,Items 4.1,4.2.1 & 4.2.2 Re Preventive Maint Program for Reactor Trip Breakers/Maint & Trending.Position on Item 4.1 Acceptable.Position on Items 4.2.1 & 4.2.2 Unacceptable ML20207T4241985-12-30030 December 1985 SER Re SPDS Based on Documentation & 850513-15 Audit.Spds Does Not Fully Meet Requirements of Suppl 1 to NUREG-0737. Five Listed Variables Should Be Added to Spds.Interim Implementation Acceptable ML20138M9011985-12-17017 December 1985 Draft SER on Util 850329 Response to 841003 Request for Addl Info Re Hydrogen Control Measures for Plant.Addl Info & Analyses Required Re Effect of Upper Compartment Burns on Air Return Fan ML20128F9921985-06-21021 June 1985 SER Based on Util 831104 Response to Generic Ltr 83-28,Item 1.1 Re post-trip Review Program Description & Procedures. Program & Procedures Acceptable 1999-08-20
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217H0201999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Catawba Nuclear Station,Units 1 & 2 ML20216E5401999-09-0707 September 1999 Special Rept:On 990826,discovered That Meteorological Sys Upper Wind Speed Cup Set Broken,Causing Upper Wind Channel to Be Inoperable.Cup Set Replaced & Channel Restored to Operable Status on 990826 ML20212B4711999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Catawba Nuclear Station,Units 1 & 2 ML20217H0321999-08-31031 August 1999 Revised Monthly Operating Rept for Aug 1999 for Catawba Nuclear Station,Units 1 & 2 ML20211B1281999-08-31031 August 1999 Dynamic Rod Worth Measurement Using Casmo/Simulate ML20211A9791999-08-20020 August 1999 Safety Evaluation Granting Licensee Request for Approval of Proposed Relief from Volumetric Exam Requirements of ASME B&PV Code,Section Xi,For Plant,Unit 2 ML20211F3441999-08-17017 August 1999 Updated non-proprietary Page 2-4 of TR DPC-NE-2009 ML20211C1291999-08-17017 August 1999 ISI Rept Unit 1 Catawba 1999 RFO 11 ML20210R1051999-08-0606 August 1999 Special Rept:On 990628,cathodic Protection Sys Was Declared Inoperable After Sys Did Not Pass Acceptance Criteria of Bimonthly Surveillance.Work Request 98085802 Was Initiated & Connections on Well Anode Were Cleaned or Replaced ML20210S2891999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Catawba Nuclear Station,Units 1 & 2 ML20212B4871999-07-31031 July 1999 Revised Monthly Operating Rept for July 1999 for Catawba Nuclear Station,Units 1 & 2 ML20209E4361999-07-0909 July 1999 SER Agreeing with Licensee General Interpretation of TS LCO 3.0.6,but Finds No Technical Basis or Guidance That Snubbers Could Be Treated as Exception to General Interpretation ML20196K6631999-07-0707 July 1999 Safety Evaluation Supporting Licensee 990520 Position Re Inoperable Snubbers ML20210S2951999-06-30030 June 1999 Revised Monthly Operating Rept for June 1999 for Catawba Nuclear Station,Units 1 & 2 ML20209H4501999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Catawba Nuclear Station,Units 1 & 2 ML20209H4561999-05-31031 May 1999 Revised Monthly Operating Rept for May 1999 for Catawba Nuclear Station,Units 1 & 2 ML20206T4771999-05-31031 May 1999 Rev 3 to UFSAR Chapter 15 Sys Transient Analysis Methodology ML20196L1881999-05-31031 May 1999 Non-proprietary Rev 1 to DPC-NE-3004, Mass & Energy Release & Containment Response Methodology ML20196A0001999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Catawba Nuclear Station,Units 1 & 2 ML20206P5201999-05-14014 May 1999 Safety Evaluation Accepting GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves ML20206N8391999-05-0404 May 1999 Rev 16 to CNEI-0400-24, Catawba Unit 1 Cycle 12 Colr ML20206R1811999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Catawba Nuclear Station,Units 1 & 2 ML20196A0041999-04-30030 April 1999 Revised Monthly Operating Repts for Apr 1999 for Catawba Nuclear Station,Units 1 & 2 ML20206N8261999-04-22022 April 1999 Rev 15 to CNEI-0400-24, Catawba Unit 1 Cycle 12 Colr. Page 145 of 270 of Incoming Submittal Not Included ML20205S5551999-04-21021 April 1999 Safety Evaluation Accepting Response to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design Basis Accident Conditions ML20205N3651999-04-12012 April 1999 Safety Evaluation Accepting IPE of External Events Submittal ML18016A9011999-04-12012 April 1999 Part 21 Rept Re Defect in Component of DSRV-16-4,Enterprise DG Sys.Caused by Potential Problem with Connecting Rod Assemblies Built Since 1986,that Have Been Converted to Use Prestressed Fasteners.Affected Rods Should Be Inspected ML20206R1931999-03-31031 March 1999 Revised Monthly Operating Repts for Apr 1999 for Catawba Nuclear Station,Units 1 & 2 ML20205P9521999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Catawba Nuclear Station,Units 1 & 2 ML20205P9561999-02-28028 February 1999 Revised Monthly Operating Repts for Feb 1999 for Catawba Nuclear Station,Units 1 & 2 ML20204C9111999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Catawba Nuclear Station,Units 1 & 2 ML20203A2581999-02-0505 February 1999 Safety Evaluation of TR DPC-NE-3002-A,Rev 2, UFSAR Chapter 15 Sys Transient Analysis Methodology. Rept Acceptable. Staff Requests Duke Energy Corp to Publish Accepted Version of TR within 3 Months of Receipt of SE ML20204C9161999-01-31031 January 1999 Revised Monthly Operating Repts for Jan 1999 for Catawba Nuclear Station,Units 1 & 2 ML20199K8711999-01-13013 January 1999 Inservice Insp Rept for Unit 2 Catawba 1998 Refueling Outage 9 ML20199E3071998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Catawba Nuclear Station,Units 1 & 2 ML20216F9931998-12-31031 December 1998 Piedmont Municipal Power Agency 1998 Annual Rept ML20205E9441998-12-31031 December 1998 1998 10CFR50.59 Rept for Catawba Nuclear Station,Units 1 & 2, Containing Brief Description of Changes,Tests & Experiments,Including Summary of Ses.With ML20206P2081998-12-31031 December 1998 Special Rept:On 981218,inoperability of Meteorological Monitoring Instrumentation Channels,Was Observed.Caused by Data Logger Overloading Circuit.Replaced & Repaired Temp Signal Processor ML20203A4101998-12-22022 December 1998 Rev 16 to CNEI-0400-25, Catawba Unit 2 Cycle 10 Colr ML20203A4041998-12-22022 December 1998 Rev 14 to CNEI-0400-24, Catawba Unit 1 Cycle 11 Colr ML20198B1341998-12-14014 December 1998 Revised Special Rept:On 980505,discovered That Certain Fire Barriers Appeared to Be Degraded.Caused by Removal of Firestop Damming Boards.Hourly Fire Watches Established in Affected Areas ML20196J8351998-12-0808 December 1998 Safety Evaluation Granting Relief Request Re Relief Valves in Diesel Generator Fuel Oil Sys ML20199E3221998-11-30030 November 1998 Revised MOR for Nov 1998 for Catawba Nuclear Station,Units 1 & 2 Re Personnel Exposure ML20198E3151998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Catawba Nuclear Station,Units 1 & 2 ML20196C0251998-11-27027 November 1998 SER Accepting Clarification on Calibration Tolerances on Trip Setpoints for Catawba Nuclear Station ML20196A6881998-11-25025 November 1998 Safety Evaluation Granting Relief Request 98-02 Re Limited Exam for Three Welds ML20196D4041998-11-19019 November 1998 Rev 1 to Special Rept:On 980618,determined That Method Used to Calibrate Wind Speed Instrumentation Loops of Meteorological Monitoring Instrumentation Sys Does Not Meet TS Definition for Channel Calibration.Procedure Revised ML20195E5521998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Catawba Nuclear Station,Units 1 & 2 ML20198E3261998-10-31031 October 1998 Revised Monthly Operating Repts for Oct 1998 for Catawba Nuclear Station,Units 1 & 2 ML20154M7661998-10-12012 October 1998 LER 98-S01-00:on 980913,terminated Vendor Employee Entered Protected Area.Caused by Computer Interface Malfunction. Security Retained Vendor Employee Badge to Prevent Further Access & Computer Malfunction Was Repaired.With 1999-09-07
[Table view] |
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SAFETY EVALUATION BY THE 3FFICE OF NUCLEAR REACTOR REGULATION RELATING TO TECHNICAL SPECIFICATION CHANGES AND UNIT 2/ CYCLE 2 RELOAD DUKE POWER COMPANY CATAWBA UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 INTRODUCTION By letter oated November 13, 1987 (Ref. 1), Duke Power Company (the 'icensee) requested Changes to the Technical Specifications for Catawba Nuclear Station Units 1 and 2, to reflect the Unit 2 refueling and the addition of the Boron Dilution Mitigation System for Unit 2. In addition deletion of Surveillance 4.3.3.12.1(b) which requires an ANALOG CHANNEL OPERATIONAL CHECK prior to startup is requested for both units. A second letter dated December 11, 1987 (Ref 2) provided a discussion of the Justification and No Significant Hazards Considerations. Additional infonnation and justification were provided in letters dated January 15 and 20, 1988 (Refs. 10 and 11).
EVALUATION A. Unit 2 Cycle 2 Reload
- 1. General Design The Catawba Unit 2, Cycle 2. reactor core contains 193 Optimized Fuel Assemblies. During the Cycle 1/2 refueling 64 Region 1 fuel assemblies will be replaced with 64 Region 4 fuel assemblies. The Region 4 fuel is very similar to that used in Regions 1, 2, and 3.
Region 4 fuel assemblies have a smaller rod plenum spring than those used in Regior;s 1, 2, and 3. This new spring design is being generally incorporated by Westinghouse and the justification was submitted in Reference 3. The Region 4 fuel has been designed according to the fuel performance model in WCAP 8785 (Ref. 4). Tne fuel is designed and operated so that clad flattenin as provided by the Westinghouse model in WCAP Ref 5).8377 For(g allwill not occ fuel regions, the fuel rod internal pressure design basis, which is discussed and shown acceptable in WCAP-8964 (Ref. 6) is satisfied, 1 The licensee provided a 2eload Safety Evaluation (RSE) for Catawba 2, l Cycle 2, as an attachment to Reference 1. The RSE pre:ents a l Cycle-specific evaluation for Cycle 2 which demonstrates that the l core reload will not adversely affect the safety of the plant. This l evaluation was performed utilizing the approved reload design l t
methods of WCAP-9273-U-A (Ref. 7).
l 8801270128 880122 PDR ADOCK 05000413-P PDR
- 2. Nuclear Design The Cycle 2 Core loading is designed to meet an [Fn(Z)xP] ECCS limit of less than or equal to 2.32xK(Z). Adherence to Yhe Fn limit is obtained by using the F TS surveillance described in WCAP-10217-A (Ref. 8). F surveillabceispartoftheRelaxedAxialOffsetControl (RA0C) and rkplaces the previous F surveillance by~ comparing a measuredFnlimit.ThisprovidesaMoreconvenientformofassuring plant operNtion below then F limit while retaining the intent of using a measured parameter Yo verify operation below TS limits. The above discussion is consistent with Reference 8 which was approved.
Thus, the staff finds that the TS change to Fg surveillance is acceptable.
RAOC will be employed in Cycle 2 to enhance operational flexibility during non steady state operation. RA0C makes use of available margin by expanding the allowable di band, particularly at reduced power. ROAC is described in Reference 8 and was approved by the staff. Thus, it is acceptable for use in Catawba Unit 2.
During operation at or near steady state equilibrium conditions core peaking factors are significantly reduced due to the limited amount of xenon skewing possible under these operating conditions. The licensee proposes to use Base load TS to recognize this reduction in core peaking factors. The proposed Base load TS are identical to those that the staff has previously approved for McGuire Units 1 and 2, and Catawba Unit 1 and are therefore acceptable.
The RSE provides a table of Cycle 2 kinetics characteristics which are compared with the current limits based on previously approved accident analyses. The RSE also provides a table showing the results of the calculated Cycle 2 control rod worths and requirements ct the most limiting condition during the cycle (end-of-life). These results include a standard 10% allowance for calculational uncertainty, From this information, the staff concludes that sufficient control rod worth will be available to provide the required shutdown margin for Cycle 2 operation. Control rod insertion limits were increased for less than 100% power for Cycle 2. Since the required shutdown margin is maintained, the TS change proposed to reflect the increased insertion is acceptable.
- 3. Thermal and Hydraulic Design The thermal hydraulic methodology, DNBR correlation and core DNB limits used for Cycle 2 are consistent with the current licensing basis described in the FSAR and approved by the staff.
The power distributions produced by the cycle-specific RA0C analysis were analyzed for normal operation and Condition II events, limits on the 4 allowable operating flux difference as a function of power level from these considerations were found to be less restrictive than those l
resulting from LOCA Fn considerations. The Condition 11 analyses generate DNB core limYts and resultant Over-Temperature Delta-T setpoints. These generated a change to the F(41) function in the TS. The change is acceptable because it results from cycle-specific calculations using approved methods (Refs 7 and 8). Therefore, the staff concludes that the Cycle 2 thermal-hydraulic analysis is acceptable.
- 4. Accident Analysis The effects of the reload on the design basis and postulated accidents analyzed in the FSAR were examined. in all cases it was found that the effects were accommodated within the conservatism of the initial assumptions used in the previous applicable safety analysis, the safety evaluation performed in support of the RTD Bypass Elimination licensing submittal and the safety evaluation performed in support of the UHI Elimination licensing submittal.
(Refs.1, 2 and 111 A core reload can affect accident analysis input parameters through control rod worths, core peaking factors and core kinetic characteristics. The Cycle 2 parameters in each of these areas were examined and found to be within the bounds of the current limits.
- 5. Technical Specification Changes The Technical Specification changes for the Unit 2 Cycle 2 Reload I are:
- 1. RA0C and Axial Flux Difference Limits i
- 2. Fg Surveillance
- 3. Base load Technical Specifications ;
- 4. Rod Insertion Limits
- 5. OT AT f3 (61)
Acceptability of items 1 - 4 was discussed in Section 2 Nuclear Design. Acceptability of item 5 was discussed in Section 3, thermal and hydraulic design. The proposed changes are for Unit 2 only but the actuel change pages involve both Units. The revisions to the bases are also acceptable.
B. Boron Dilution Mitigatien Syst_em 1.0 Introduction The Boron Oilution Mitigation System (BDMS) which is being installed in Unit 2 is the same as the BDMS which was installed on Unit 1. The BDMS was described in letters dated June 6, 1986 and September 9, 1986 (Refs.
12 and 13).
- 2. Technical Specification Changes The changes for Unit'2 which deal with the Boron Oilution Mitigation System (BDMS) are to Specifications 4.1.1.1.3, 4.1.1.1.4, 4.1.1.2.2, Table 3. 3-1; item 6.b. Table 3.3-1; Action 5, Table 4.3-1; Note (9), 3/4.3.3.12, and 3/4.9.2. Changes to Specifications 4.3.3.12.1(b), 3.9.2.1 and 4.9.1.3 apply to both Units. Each change is discussed below.
The changes that apply to Unit 2 only are identical to the changes made to the Unit 1 Technical Specifications when the Boron Dilution Mitigation System was installed on that Unit. They were reviewed and approved at that time (Ref 9). It was requested that these changes not apply to Unit 2 until after the BDMS system has been calibrated, tested and declared operable. The licensee had stated (Ref.10) that all the Technical Specifications applicable to the boron dilution accident which are to be deleted will be adminstratively maintained in this interim period. We find this acceptable.
Specification 4.3.3.12.1(8)
This specification will be deleted since it requires surveillance prior to Mode 2 but the Specification itself is not applicable in Modes 1 and 2. We find this change acceptable.
Specification 3.9.2.1 This change is an editorial change which deletes a phrase "and control room" which appeared twice in the sentence. Thus it is acceptable.
Specification 3.9.1.3 This specification verifies that potential boron dilution flow paths are isolated when the unit is in Mode 6. The deletion of TS 3.9.1.3 is acceptable since the BDMS provides for automatic isolation of potential boron dilution flow paths.
CONCLUSION We have concluded, based on the-considerations discussed above, that: (1) there is reasonable assurance that the health and safet of the public will not be endangered by operation ir, the proposed manr.er 2 and (y) such activities will be conducted in compliance with the Comission's regulations and the issuance :
of these amendments will not be inimical to the common defense and security or the health and safety of the public.
References
- 1) letter from Hal B. Tucker (Duke Power Companyl to NRC, November 13, 1987
- 2) letter from Hal B. Tucker (Duke Power Company) to NRC, December 11, 1987
- 3) letter from E. P. Rahe Jr. (Westinghouse) to L. E. Phillps (NRC), April 12, 1984, NS-EPR-2893
Subject:
Fuel Handling load Curtain (6g vs 49)
- 4) Miller, J. V. (Ed.) "Improved Analytical Model used in Westinghouse Fuel Rod Design Computations, "WCAP-8785, October, 1976
- 5) George, R. A., (et al.), "Revised Clad Flattening Model, WCAP-8377, July, 1977
- 6) Risher, D. H., (et al.), "Safety Analyses for the Reused Fuel Rod Internal Pressure Design Basis," WCAP-8964, June, 1977
- 7) Bordelon, F. M., (et all, "Westinghouse Reload Safety Evaluation Methodology", WCAP-9273-A, July, 1985
- 8) Muller, R. W. , (et al.), "Relaxation of Constant Axial Offset Control-FnSurveillance Technical Specification, "WCAP-10217-A, June, 1983
- 9) Mercrandum to Kahtan Jabbour, NRC, from Charles E. Rossi, NRC, October 1, 1986, (TAC 61740)
- 10) Letter frw Hal B. Tucker (Duke Power Company) to NRC, January 20, 1988
- 11) letter from Hal B. Tucker (Duke Power Company) to NRC, January 15, 1988
- 12) Letter from Hal 8. Tucker (Duke Power Company) to Harold Denton (NRC),
June 6, 1986
- 13) Letter from Hal B. Tucker (Duke Power Company) to Harold Denton (NRC),
September 9, 1986 l
.. .. \
ENCLOSURE 2 Functional Areas
- 1. Management Involvement in Assuring Quality.
Rating: Category 3
- 2. Approach to Resolution of Technical Issues from a Safety Standpoint.
Rating: Category 3
- 3. Responsiveness to NRC Initiatives Rating: Category 3
- 4. Enforcement History N/A
- 5. Operational and Construction Events N/A
- 6. Staffing (including Management)
N/A
- 7. Training and Qualification Effectiveness N/A i
i
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