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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML15261A4681998-09-0404 September 1998 Safety Evaluation Supporting Amends 232,232 & 231 to Licenses DPR-38,DPR-47 & DPR-55,respectively ML20198J7651997-10-15015 October 1997 Safety Evaluation Accepting 10-yr Interval Insp Program Plan Alternatives for Listed Plants Units ML20148H2501997-06-0505 June 1997 Safety Evaluation Accepting Proposed Restructuring of Util Through Acquisition Of,& Merger W/Panenergy Corp ML20134N7121997-02-20020 February 1997 Safety Evaluation Accepting Relief Request 96-04 for Plant ML20236L5971994-12-29029 December 1994 SER in Response to 940314 TIA 94-012 Requesting NRR Staff to Determine Specific Mod to Keowee Emergency Power Supply Logic Must Be Reviewed by Staff Prior to Implementation of Mod ML20236Y0221987-07-0808 July 1987 Safety Evaluation Clarifying Determination of Acceptability of Test Duration for Performance of Integrated Leak Rate Test at Plant ML20214S8731986-11-26026 November 1986 Safety Evaluation Re LER 269/86-006 Concerning Moderator Temp Coefficient.Concurs W/Licensee Assessment,Actions Taken or to Be Taken.Four Impacted FSAR Events All Remained within Respective Safety Criteria ML20236X6941986-08-0505 August 1986 Supplemental Safety Evaluation Supporting Seismic Qualification of Auxiliary Feedwater Sys (Multi-Plant Action C-14).SALP Input & 861029 Fr Re Proposed Rules Section Encl ML20236X5371985-10-22022 October 1985 SER Re Seismic Qualification of Plant Auxiliary Feedwater Sys.Nrr Request for Addl Info Encl ML20134N7761985-05-15015 May 1985 Safety Evaluation Accepting Util 831104 Response to Generic Ltr 83-28 Item 1.1 Re post-trip Review Program & Procedures ML20236X5771982-12-29029 December 1982 SER Addressing Concerns of Safe Shutdown of Plant Following Fire,Internal Flooding of Turbine Bldg & Security Incidents ML20126M5011981-05-22022 May 1981 Safety Evaluation for Unit 2 Concluding That Conformance W/Requirements & Completion of Corrective Actions by 820630 Will Ensure Compliance w/800523 Memorandum & Order.Info Re Deficiencies Noted Requested within 90 Days ML20214J2911979-12-20020 December 1979 Safety Evaluation Re Preliminary Design for safety-grade Anticipatory Reactor Trips on Loss of Main Feedwater &/Or Turbine Trip 1998-09-04
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20206P1501999-01-0505 January 1999 LER 98-S03-00:on 981207,security Officer Discovered Uncontrolled Safeguards Info Drawing.Caused by Failure to Follow Established Procedures & Policies.Drawing Was Controlled by Site Security.With ML20216F9931998-12-31031 December 1998 Piedmont Municipal Power Agency 1998 Annual Rept ML20198E6381998-12-17017 December 1998 LER 98-S02-00:on 981130,security Access Was Revoked Due to Falsification of Criminal Record.Individual Was Escorted from Protected Area & Unescorted Access Was Restricted. with ML20153G4601998-09-30030 September 1998 USI A-46 Seismic Evaluation Rept, Vols 1-2 ML17354B0971998-09-0909 September 1998 Part 21 Rept Re Possible Machining Defect in Certain One Inch Stainless Steel Swagelok Front Ferrules,Part Number SS-1613-1.Caused by Tubing Slipping Out of Fitting at Three Times Working Pressure of Tubing.Notified Affected Utils ML15261A4681998-09-0404 September 1998 Safety Evaluation Supporting Amends 232,232 & 231 to Licenses DPR-38,DPR-47 & DPR-55,respectively ML20248F7441998-05-31031 May 1998 Reactor Vessel Working Group,Response to RAI Regarding Reactor Pressure Vessel Integrity ML20247L9041997-12-31031 December 1997 1997 Annual Rept for Duke Energy Corporation & Saluda River Electric Cooperative,Inc,Financial Statements as of Dec 1997 & 1996 Together W/Auditors Rept ML20198J7651997-10-15015 October 1997 Safety Evaluation Accepting 10-yr Interval Insp Program Plan Alternatives for Listed Plants Units ML20148S3141997-06-30030 June 1997 Ro:On 970422,Oconee Unit 2 Was Shut Down Due to Leak in Rcs. Leak Was Caused by Crack in Pipe to safe-end Weld Connection at RCS Nozzle for HPI Sys A1 Injection Line.Unit 1 Was Shut Down to Inspect Hpis Injection Lines & Implement Ldst Mods ML20148H2501997-06-0505 June 1997 Safety Evaluation Accepting Proposed Restructuring of Util Through Acquisition Of,& Merger W/Panenergy Corp ML20210E3591997-03-27027 March 1997 Part 21 Rept Re Sorrento Electronics Inc Has Determined Operation & Maint Manual May Not Adequately Define Requirements for Performing Periodic Surveillance of SR Applications.Caused by Hardware Failures.Revised RM-23A ML20134N7121997-02-20020 February 1997 Safety Evaluation Accepting Relief Request 96-04 for Plant ML20138L2151997-01-31031 January 1997 Monthly Operating Repts for Jan 1997 for Oconee Nuclear Station,Units 1,2 & 3 ML20138L2281996-12-31031 December 1996 Revised Monthly Operating Repts for Dec 1996 for Oconee Nuclear Station,Units 1,2 & 3 ML20133C1231996-12-23023 December 1996 Informs Commission of Staff Review of Request for License Amends from DPC to Perform Emergency Power Engineered Safeguards Functional Test on Three Oconee Nuclear Units ML20115F2471996-07-0303 July 1996 Part 21 Rept Re Piping (Small Portion of Unmelted Matl Drawn Lengthwise Into Bar During Drawing Process) Defect That Existed in Bar as Received from Mill.Addl Insp Procedure for Raw Matl Instituted ML20107M8931995-10-31031 October 1995 Nonproprietary DPC Fuel Reconstitution Analysis Methodology ML17353A4341995-10-31031 October 1995 Rev 1 to BAW-2245, Initial Rt of Linde 80 Welds Based on Fracture Toughness in Transition Range. ML17264A1181995-07-31031 July 1995 Response to Part (1) of GL 92-01,Rev 1,Suppl 1. ML20086M0851995-06-29029 June 1995 DPC TR QA Program ML20077R3631994-12-31031 December 1994 Monthly Operating Repts for Dec 1994 for Bfnpp ML20236L5971994-12-29029 December 1994 SER in Response to 940314 TIA 94-012 Requesting NRR Staff to Determine Specific Mod to Keowee Emergency Power Supply Logic Must Be Reviewed by Staff Prior to Implementation of Mod ML20064L2001994-01-31031 January 1994 Final Rept EPRI TR-103591, Burnup Verification Measurements on Spent-Fuel Assemblies at Oconee Nuclear Station ML20062K7481993-12-0101 December 1993 ISI Rept for Unit 2 McGuire 1993 Refueling Outage 8 ML20056E5171993-08-31031 August 1993 Technical Review Rept, Tardy Licensee Actions ML20046C1291993-08-0202 August 1993 LER 93-007-00:on 930701,determined That Unit 1 Ssf Rc Makeup Sys Inoperable in Past Due to Design Deficiency.Operations Procedures Revised to Reflect Newly Calculated Operating Limits for Rc Makeup Pump,Rcps & RCS.W/930802 Ltr ML20056G0131993-07-27027 July 1993 Rev 0 to ISI Rept Unit 2 Oconee 1993 Refueling Outage 13 ML20044G5311993-05-26026 May 1993 Suppl to 921207 Part 21 Rept Re Declutch Sys Anomaly in Certain Types of Valve Actuators Supplied by Limitorque Corp.Limitorque Designed New Declutch Lever Which Will Be Available in First Quarter 1993 ML20126J5961992-12-31031 December 1992 Part 21 Rept Re Potential Loss of RHR Cooling During Nozzle Dam Removal.Nozzle Dams May Create Trapped Air Column Behind Cold Leg Nozzle Dam.Mod to Nozzle Dams Currently Underway. Ltrs to Affected Utils Encl ML20117A5981992-11-23023 November 1992 Special Rept:On 921119,ability of Control Battery Racks to Withstand Seismic Event Could Not Be Confirmed & Batteries Declared Inoperable.Batteries Expected to Be Restored in TS Required Time ML20097G0421992-05-31031 May 1992 Analysis of Capsule OCIII-D Duke Power Company Oconee Nuclear Station Unit-3 ML20077D0671991-11-15015 November 1991 Nonproprietary Version of Rev 0 to Boric Acid Corrosion of Oconee Unit 1 Upper Tubesheet ML20067A5241990-12-31031 December 1990 Final Submittal in Response to NRC Bulletin 88-011, 'Pressurizer Surge Line Thermal Stratification.' ML20042F3541990-04-30030 April 1990 Special Rept Re Failure to Prevent Performance Degradation of Reactor Bldg Cooling Units.Caused by Mgt Deficiency & Inadequate Program.Cooling Unit Declared Inoperable & Removed from Svc for Cleaning & Placed Back in Operation ML17348A1621990-03-27027 March 1990 Part 21 Rept Re Matls W/Programmatic Defects Supplied by Dubose Steel,Inc.Customers,Purchase Order,Items & Affected Heat Numbers Listed ML19332D5391989-10-31031 October 1989 Core Thermal-Hydraulic Methodology Using VIPRE-01. ML20042F2321989-08-31031 August 1989 Nonproprietary DCHF-1 Correlation for Predicting Critical Heat Flux in Mixing Vane Grid Fuel Assemblies. ML20205F3211988-10-10010 October 1988 Part 21 Rept Re Potential Deviation from Tech Spec Concerning Ry Indicators Due to Operating Temp Effect on Analog Meter Movement.Initially Reported on 881006.Customers Verbally Notified on 881006-07 ML20154K2091988-09-0909 September 1988 Rev 0 to Response to NRC Bulletin 88-005,Nonconforming Matls Supplied by Piping Supplies,Inc at Folsom,Nj & West Jersey Mfg Co.... Proprietary Procedure 1404.1, Leeb Hardness Testing (Equotip).... Encl.Procedure Withheld ML20245D9541988-09-0606 September 1988 Part 21 Rept Re Condition Involving Inconel 600 Matl Used to Fabricate Steam Generator Tube Plugs & Found to Possess Microstructure Susceptible to Stress Corrosion Cracking ML20245B6061988-08-31031 August 1988 Inadequate NPSH in HPSI Sys in Pwrs, Engineering Evaluation Rept ML20239A6991987-11-30030 November 1987 Addendum 1 to Rev 2 to Integrated Reactor Vessel Matl Surveillance Program (Addendum) ML20236T0791987-11-25025 November 1987 Advises LER 269/87-09,re Degradation of More than One Functional Unit of Emergency Power Switching Logic for Units 2 & 3,in Preparation & Will Be Submitted by 871215. Incident Originally Discussed in Special Rept ML20236Q9491987-10-31031 October 1987 Monthly Operating Repts for Oct 1987 ML20235W9611987-09-30030 September 1987 Monthly Operating Repts for Sept 1987 ML20234B1861987-08-31031 August 1987 Monthly Operating Repts for Aug 1987 ML20237K4761987-07-31031 July 1987 Monthly Operating Repts for Jul 1987 ML20236Y0221987-07-0808 July 1987 Safety Evaluation Clarifying Determination of Acceptability of Test Duration for Performance of Integrated Leak Rate Test at Plant ML20235S6311987-06-30030 June 1987 Monthly Operating Repts for June 1987 1999-01-05
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/ o UNITED STATES E" ~1 NUCLEAR REGULATORY COMMISSION 5 :
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SAFETY-EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION MODERATOR TEMPERATURE COEFFICIENT, LER 269/86-006 0CONEE NUCLEAR STATION, UNIT NO. 1 DUKE POWER COMPANY DOCKET NO. 50-269
1.0 INTRODUCTION
By letter dated May 12, 1986, Duke Power Company (DPC) submitted a Licensee Event Report (LER) No. 269/86-006 (Ref. 1). This LER reports the results of a measurement which indicated that the end of cycleJ(EOC) mocerator temperature coefficient (MTC) was more negative than assumed in the safety analysis for Oconee, Unit 1, Cycle 9. The staff has reviewed the LER, information submitted by DPC by letter dated April 23, 1986 (Ref. 2),
pertinent secticns of the FSAR, and information received in a telephone conference call,on June 20, 1986.
MTCisthechEn#ge in reactivity that results from a change in the temperature of the water in the core. It is measured in units of Ap/AT, or a change in reactivity per degree Fahrenheit change in the average moderator temperature. For power excursion transients, the more negative MTC is desirable. That is not the case for overcooling accidents. As the assumed MTC becomes more negative, added positive reactivity feedback occurs from primary system overcooling. A more negative MTC raises the potential that the core could achieve recriticality during the transient.
- 2. 0 BACKGROUND On April 10, 1986 Duke Power Company (DPC), the licensee for Oconee Nuclear Power Station, Unit 1, concluded that the hot full power (HFP)
, moderator4 temperature coefficient (MTC) at the end of cycle (E0C) 9 was
-3.36x10 ap/*F. This MTC was based on the analysis of measured data taken on February 10, 1986 with the reactor at 98% of full rated power and at a soluble boron concentration of 46 parts per 4million (ppm). Since the MTC was more negative than the value of -3.0x10 Ap/*F assumed in the Cycle 9 safety analyses, it resulted in a condition outside of the design basis of the plant and, therefore, was reportable under the -
provisions of 10 CFR 50.72 and 10 CFR 50.73. The LER 269/86-006 was sent by DPC for Oconee, Unit 1 pursuant to th'ese regulations.
No corrective actions were necessary for Cycle 9 because the end of Cycle 9 occurred on February 13, 1986 shortly after the measurements were taken.
Noaccidentsortransients.gccurredduringthetimethattheHFPMTCwas more negative than -3.0x10 op/*F. Therefore, the public health and safety 8612030407 861126 PDR ADOCK 05000269 S PDR
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s 3 s were not affected. DPC reported that no corrective actions were required for Oconee, mUnits 2 and 3 since these units were operating at cycle burnups and solu'sle boron concentrations whict would not produde MTCs more negative
~4 thar: -3.0x10 ap/ F. However, an MTC measurement was performed on Marcn 11, '
1986 on Oconee, Unit 2 with the reactor at 274 effective full power days (EFPD) into Cycle 8. This measurement yielded a value for the MTC of
~4 '
-2."88x10 W *F.
3.0 LVALUATION
' This measurement, along with the previous me,asurement for Oconee Unit 1, confirmec the non-cNi's'ervatism in DPC's Metermiriation of MTCs for these extended :ycle cores. OPC attributes this deficiency tc.a normalization of its MTC calculations to B&W's MTC calculations3 and of a lack ofa data
- o bencamark calculations in the applicable burnup range of the reactor
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cycle. OPC reported in the- LER that it would remove this calculational bias in its MTC calculations ~. DPC states that additional measurements will be taken on an Oconee unit during a shutdown, due to whatever cause, which occurs after 300 EFPDs and leads to a xenon free core.
DPC reviewed the FSAR to determine the' transients and accidents that would be Adversely impacted by a more negative HFP 'Ei)C MTC. Youreventswere 3
ide'ntified as requiring further evaluation (Ref. 2). These events are (1) the cold water accident (FSAR Section:15.5), (2) the control rod misalignment accidents (FSAR Section 15.7), (3) the rod ejection accident (FSAR Section 15.12), and (4) the steam line break accident (FSAR Section 15.13). OPC used the RETRAN 02-M00003 computer s g code to calculate the system responses to the cold water accident and the' control rod
~4 alsalignment accident. The analyses used an MTC of -3.5x10 W *F.
The results:showed that the consequences of a cold water v, .
accidentwere more severe than the consequences with an MTC of -3.0x10 4 ap/*F. The
~
consequences for the control rod misaligntient accidents (a control rod drop accident was analyzed) were shown to be insensitive to the'MTC.
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I Although RETRAN 02-M00003 has been accepted by the staff, each utility using the code is. required to submit its own plant specific application justifying the use of the code. This has not been done by DPC. However, the use of the code by DPC to determine the system responses for the two events described above is acceptable for the purpose of evaluating the safety significance of having been outside of the design basis of the Oconee, Unit 1 safety analysis near EOC 9 with the more negative MTC.
DPC reevaluated the results for the rod ejection acciden't analysis presented in the FSAR. These results show that a more negative MTC would nbt increase the consequences of ~a control rod ejection accident.
In its review ofge information presented in the FSAR, DPC states that the HFP MTC is not controlling regarding the shutdown margin because its core design methodology requires at least a 1% Ak/k shutdown margin at hot zero power (HZP). In fact, DPC states that, in practice, the actual E0C shutdown margin exceeds the required 1% Ak/k margin. For moderator cool-down below 532*F, DPC will use the HZP MTC as being the applicable MTC rather than the HFP MTC that evidently was used in the FSAR analysis.
In a telephone conference c:ll with the staff on June 20, 1986, DPC stated :
+
that the MTC used in deriving the normal operating power-imbalance envelope was based on an MTC inherent w the.nodalocode calculations. The MTC obtained with the nodal code (ard not normalized to the B&W MTCs) apparently yields acceptable values of the MTC when compared to the measured data. Therefore, since the nodal code MTCs were used in ,
deriving the power-imbalance envelope, the more negative measured EOC MTC did not affect these limits.
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4.0 CONCLUSION
S Based on its review, the staff concurs with the DPC assessment, actions taken, or to be taken concerning the more negative MTC than assumed in the safety analyses for.EOC 9 for Oconee, Unit 1. In particular, the assessment showed that the four FSAR events that were impacted all remained within their respective safety criteria. In addition, OPC will design reload Oconee cores such'that the HFP HTC will be no more negative than -3.5x10 -4 ap/*F and the HZP MTC will be no more negative
-4 than -3.0x10 ap/ F. Finally, OPC will perform HTC measurements as opportunities be.qdp'e available in'the Oconee, Units.
5.0 REFERENCES
- 1. LER.269/86-006 Oconee Unit 1 Cycle 9 End of Cycle Hot Full Power Lower Moderator Temperature Coefficient, May 12, 1986.
- 2. Letter from H. B. Tucker (Duke Power Company) to H. R. Denton, (NRC),
April 23, 1986.
Dated: Nov' ember 26, 1986 ,
Principal Contributor:
D. Fieno 9
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