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Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data ML20211A9981999-07-12012 July 1999 Draft,Probabilistic Safety Assessment, Risk Info Matrix, Risk Ranking of Systems by Importance Measure ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20155D9961998-10-31031 October 1998 Rev 0 to GE-NE-B13-01980-24, Fracture Mechanics Evaluation on Observed Indication at N3A Steam Outlet Nozzle to Shell Weld at Cooper Nuclear Station ML20151Q1211998-08-14014 August 1998 Rev 0 to Control of Hazard Barriers ML20236P2971998-07-0707 July 1998 Rev 2 to NPPD CNS Strategy for Achieving Engineering Excellence ML20238E9201997-03-31031 March 1997 Geris 2000 Invessel Sys Alternate Method for Compliance to Reg Guide 1.150 ML20133F4561996-10-31031 October 1996 Engineering Self Assessment Follow-Up, for NPPD Cooper Nuclear Station ML20113B2831996-05-31031 May 1996 USI A-46 Seismic Evaluation Rept Vol I/V ML20113B2891996-05-0707 May 1996 Rev 1 to USNRC USI A-46 Resolution Ssel & Relay Evaluation Rept, Vol I of Viii ML20101K9321996-02-29029 February 1996 Determination of Loc for Cooper Feedwater Nozzle Fracture Mechanics Evaluation, for Feb 1996 ML20117N5601996-02-23023 February 1996 Engineering Self Assessment 960205- 23 ML20095G2001995-12-31031 December 1995 Fracture Mechanics Evaluation of UT Indications Found During 1995 Reexam of FW Nozzle to Shell Welds at Cns ML20095G0961995-12-18018 December 1995 Stresses from Applied Loadings for Level A,B,C & D Conditions Considered in Core Spray Line Fracture Mechanics Evaluation ML20094P2731995-11-30030 November 1995 Internal Core Spray Line Flaw Evaluation at Cns ML20078J4181994-11-0808 November 1994 Rev 0 to, Cooper Nuclear Station Restart Readiness Program ML20072U7391994-09-0101 September 1994 CNS Diagnostic Self Assessment Jul-Aug 1994 ML20113C1921994-07-31031 July 1994 Generation of Conservative Design & Median-Centered In-Structure Response Spectra for Cooper Nuclear Plant Control & Reactor Bldgs ML17352A8691993-05-31031 May 1993 Technical Rept, Assessment of Aging Degradation of Civil/ Structural Features at Selected Operating Nuclear Power Plants. ML20044G1701993-05-0404 May 1993 Seismic Occurrence of 930330, Engineering Evaluation ML20086S8811991-12-13013 December 1991 Fracture Mechanics Evaluation of UT Indications Found in Cooper Feedwater-Nozzle-to Shell Weld ML20086Q1961991-11-0101 November 1991 Emergency Preparedness Assessment Rept ML20065L6761990-11-30030 November 1990 Initial Simulator Certification Submittal ML20246C6101989-08-31031 August 1989 Dcrdr Suppl III to Summary Rept ML20151F9501988-06-27027 June 1988 Change 1,Rev 1 to SAIC-86/1797, Summary of Human Factors Activities Related to Cooper Nuclear Station Plant Mgt Info Sys & Spds ML20235U4311987-09-23023 September 1987 Turbine Bypass Valve Out-of-Svc Evaluation Summary. Related Info Encl ML20235U3931987-09-21021 September 1987 Turbine Bypass Valve Out-of-Svc Assessment. W/Records of 870921 Telcons ML20235H2741987-05-31031 May 1987 Mechanical Level Instrument Availability Analysis ML20209G8701987-01-31031 January 1987 Dcrdr Suppl II to Summary Rept ML20199C8431986-06-12012 June 1986 Rev C to Failure Mode & Effects Analyses ML20198H0201986-05-15015 May 1986 Nebraska Public Power District Response to Initial Requirements of IE Bulletin 85-003 ML20154N4741986-03-31031 March 1986 Safety Evaluation of Main Steam Line High Flow Setpoint for Cooper Nuclear Station ML20154L3281986-02-27027 February 1986 Dcrdr Suppl to Summary Rept ML20107D5231985-02-0404 February 1985 Detailed Control Room Design Review Summary Rept ML20099D4841985-02-0101 February 1985 Rev 2 to Detailed Descriptions of Displays for Cooper Nuclear Station Spds ML20087A9491984-03-0101 March 1984 Safety Parameter Display Sys Safety Analysis ML20083C6191983-11-23023 November 1983 Containment Purge & Vent Valve Operability Rept ML20073R2061983-04-30030 April 1983 Wetwell-Drywell Vacuum Breaker Valves:Long-Term Program Structural Evaluation ML20076E8611983-04-30030 April 1983 Nuclear Mgt Appraisal Rept for Nebraska Public Power District ML20064C0131982-12-31031 December 1982 10CFR50,App R Supplementary Info Rept to Vols I & II Submitted to NRC 820628 ML20084Q5361982-07-31031 July 1982 BWR Owners Group Position on NRC Reg Guide 1.97,Rev 2 ML20052C0741982-04-30030 April 1982 Plant Unique Analysis Rept:Mark I Containment Program. ML20052J0011982-02-12012 February 1982 to IF-300 Redundant Yoke,NUREG-0612 Evaluation. ML20038B7251981-10-31031 October 1981 Evaluation Rept,Oct,1981. ML20009A5351981-01-26026 January 1981 Meteorological Monitoring Sys Development Plan for Emergency Preparedness for Ne Public Power District,Cooper Nuclear Station & Brownsville,Ne, Preliminary Rept ML19295D8031980-10-31031 October 1980 Master List of Class IE Equipment. ML19262C3801980-01-21021 January 1980 Proposed Process Control Program ML20125B9931979-10-31031 October 1979 Voltage Drop Analysis Computations, Revision 1 ML19270G3301979-06-0606 June 1979 Offsite Dose Assessment Manual. ML19270G3331979-06-0606 June 1979 Radiological Environ Monitoring Manual. 1999-09-30
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20212K9781999-09-30030 September 1999 Safety Evaluation Accepting USI A-46 Implementation Program ML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data ML20217G7461999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Cooper Nuclear Station ML20217A1691999-09-22022 September 1999 Part 21 Rept Re Engine Sys,Inc Controllers,Manufactured Between Dec 1997 & May 1999,that May Have Questionable Soldering Workmanship.Caused by Inadequate Personnel Training.Sent Rept to All Nuclear Customers ML20212C5001999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Cooper Nuclear Station ML20211D6491999-08-25025 August 1999 Part 21 Rept Re Nonconformance within LCR-25 safety-related Lead Acid Battery Cells Manufactured by C&D.Analysis of Cells Completed.Analysis of Positive Grid Matl Shows Nonconforming Levels of Calcium within Positive Grid Alloy ML20210R0381999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Cooper Nuclear Station ML20210J2921999-07-29029 July 1999 Special Rept:On 990406,OG TS & Associated Charcoal Absorbers Were Removed from Svc.Caused by Scheduled Maint on Hpci. Evaluation of Offsite Effluent Release Dose Effects Was Performed to Ensure Plant Remained in Compliance ML20209H8281999-07-15015 July 1999 Safety Evaluation Accepting GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Cooper Nuclear Station ML20211A9981999-07-12012 July 1999 Draft,Probabilistic Safety Assessment, Risk Info Matrix, Risk Ranking of Systems by Importance Measure ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20209E1061999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Cns.With ML20196B3851999-06-17017 June 1999 Summary Rept of Facility Changes,Test & Experiments,Per 10CFR50.59 for Period 970901-990331.Summary of Commitment Changes Made During Same Time Period Also Encl ML20195K2851999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Cooper Nuclear Station.With ML20206P0481999-05-12012 May 1999 Safety Evaluation Concluding That NPP Established Acceptable Program to Verify Periodically design-basis Capability of safety-related MOVs at CNS & Adequately Addressed Actions Requested in GL 96-05 ML20206J0811999-05-0404 May 1999 Rev 14 to CNS QA Program for Operation ML20206P9751999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Cooper Nuclear Station ML20205Q0891999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Cooper Nuclear Station.With ML20204G8951999-03-15015 March 1999 CNS Inservice Insp Summary Rept Fall 1998 Refueling Outage (RFO-18) ML20207M9231999-03-12012 March 1999 Amended Part 21 Rept Re Cooper-Bessemer Ksv EDG Power Piston Failure.Total of 198 or More Pistons Have Been Measured at Seven Different Sites.All Potentially Defective Pistons Have Been Removed from Svc Based on Encl Results ML20204B3701999-03-11011 March 1999 SER Accepting Third 10-year Interval Inservice Insp Plan Requests for Relief for RI-17,Rev 1 and RI-25,Rev 0.Request for Relief RI-13,Rev 2 Involving Snubber Testing & Is Being Evaluated in Separate Report ML20204C9751999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Cooper Nuclear Station ML20199E6751999-01-14014 January 1999 Monthly Operating Rept for Dec 1998 for Cooper Nuclear Station ML20195B9191998-12-31031 December 1998 1998 NPPD Annual Rept. with ML20196J9641998-12-0707 December 1998 Safety Evaluation Accepting Licensee Third 10-yr Interval Inservice Insp Plan Request for Relief RI-27,rev 1 ML20198D2471998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Cooper Nuclear Station.With ML20196A2861998-11-23023 November 1998 SER Re Core Spray Piping Weld for Cooper Nuclear Station. Staff Concluded That Operation During Cycle 19 Acceptable with Indication re-examined During RFO 18 ML20196A5241998-11-23023 November 1998 Safety Evaluation Accepting Proposed Alternative to Use UT Techniques Qualified to Objectives of App Viil as Implemented by PDI Program in Performing RPV Shell Weld & Shell to Flange Weld Examinations ML20196A5061998-11-23023 November 1998 Safety Evaluation Re Flaw Indication Found in Main Steam Nozzle to Shell Weld NVE-BD-N3A at Cns.Plant Can Be Safely Operated for at Least One Fuel Cycle with Indication in as-is Condition ML20196C4241998-11-20020 November 1998 Rev 1 to Cooper Nuclear Station COLR Cycle 19 ML20195H1761998-11-17017 November 1998 SER Authorizing Proposed Alternative in Relief Requests RV-06,RV-07,RV-09,RV-11,RV-12 & RV-15 Pursuant to 10CFR50.55a(a)(3)(ii).RV-08 Granted Pursuant to 10CFR50.55a(f)(6)(i) & RV-13 Acceptable Under OM-10 ML20195F8601998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Cooper Nuclear Station.With ML20155D9961998-10-31031 October 1998 Rev 0 to GE-NE-B13-01980-24, Fracture Mechanics Evaluation on Observed Indication at N3A Steam Outlet Nozzle to Shell Weld at Cooper Nuclear Station ML20154Q5661998-10-0505 October 1998 Rev 0 to CNS COLR Cycle 19 ML20154L5381998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Cooper Nuclear Station.With ML20151Z6141998-09-16016 September 1998 SER Accepting Util Responses to NRC Bulletin 95-002 for Cooper Nuclear Station ML20154F7931998-08-31031 August 1998 Rev 0 to J11-03354-10, Supplemental Reload Licensing Rept for CNS Reload 18,Cycle 19 ML20153B1101998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Cooper Nuclear Station ML20237E7771998-08-20020 August 1998 Revised COLR Cycle 18 for Cooper Nuclear Station ML20151Q1211998-08-14014 August 1998 Rev 0 to Control of Hazard Barriers ML20237C0591998-07-31031 July 1998 Monthly Operating Rept for Jul 1998 for Cooper Nuclear Station ML20236R9131998-07-20020 July 1998 SER Accepting Rev 13 to Quality Assurance Program for Operation Policy Document for Plant ML20236P2971998-07-0707 July 1998 Rev 2 to NPPD CNS Strategy for Achieving Engineering Excellence ML20236R0931998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Cooper Nuclear Station ML20249A7701998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Cooper Nuclear Station ML20247G6131998-05-13013 May 1998 Part 21 Rept Re Defect Contained in Automatic Switch Co, Solenoid Valves,Purchased Under Purchase Order (Po) 970161. Caused by Presence of Brass Strands.Replaced Defective Valves ML20247G0951998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Cooper Nuclear Station ML20237B6861998-04-24024 April 1998 Vols I & II to CNS 1998 Biennial Emergency Exercise Scenario, Scheduled for 980609 ML20217A1531998-04-16016 April 1998 Closure to Interim Part 21 Rept Submitted to NRC on 970929. New Date Established for Completion of Level I & 2 Setpoint Project Committed to in .Final Approval of Setpoint Calculations Will Be Completed by 980531 ML20216G5331998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Cooper Nuclear Station 1999-09-30
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MD2-43-0386 March 1986 I DRF B21-00288 SAFETY EVALUATION OF MAIN STEAM LINE HIGH FLOW SETPOINT FOR COOPER NUCLEAR STATION Prepared by:
L. L. Chi, Senior Engineer Application Engineering Services Reviewed by:
J. R. Pobre, Principal Engineer Plant Piping Design Reviewed by: 8L401 R. R. Ghosh, Lead System Engineer Reactor Systems Design Reviewed by:
~ L! Rash, Principal Engineer icensing Services Ob Approved by:
M L. Sozii, Mdiddr Application Engineering Services 8603170330 860311 PDR ADOCK 05000298 p PDR GEN ER AL h, ELECTRIC NUCLEAR ENERGY BUSINESS OPERATIONS GENERAL ELECTRIC COMPANY e 175 CURTNER AVENUE
- SAN JOSE, CAUFCRNIA 95125
IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please read Carefully The only undertakings of General Electric Company respecting inform-ation in this document are contained in the contract between the customer and General Electric Company, as identified in the purchase order for this report and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than the customer or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, General Electric Company makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this docu-ment.
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l TABLE OF CONTENTS Eage,
- 1. Introduction 1
- 2. Basis for MSL High Flow Setpoint 2
- 3. Objectiive of Safety Evaluation 3
- 4. Safety Evaluation 4
- 5. Impact on USAR 5
- 6. References 6 e
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- 1. INTRODUCTION The purpose of the main steam line (MSL) high flow instrument is to detect large steam line breaks and isolate the reactor thereby limiting the potential for radioactive release outside the containment. The MSL high flow is one of the various methods to isolate the reactor in case of a MSL break outside the containment. The MSL high flow is primarily
- for detection of large breaks. For small breaks or steam leak, other methods of isolation are used. Other methods of isolation for Cooper Nuclear Station (CNS) include temperature sensors and radiation monitors in the main steam tunnel, low steam line pressure, and low reactor water level (Level 1). The effectiveness of the other isolation methods depends on the size of the break and, for area dependent monitors, on the location of the monitors.
- 2. BASIS FOR MSL HIGH FLOW SETPOINT The MSL flow rate is sensed from a differentis' pressure instrument in the flow velocity limiter (venturi) in each steam line. The maximum flow loss from a steam line break is limited by the venturi in each steam line. This maximum flow is approximately 200% of nuclear boiler rated (NBR) steam flow. The maximum setpoint for the MSL high flow trip is therefore less than 200% of NBR steam flow. The minimum setpoint
- must be above 100% of NBR steam flow to avoid spurious trips and allow for continuous operation. A typical value for a BWR/4 is 140% of NBR steam flow. This setpoint would allow on line testing of the mai'n steam isolation valves (MSIVs) since closing one valve would result in approximately 133% of NBR steam flow in the remaining three lines.
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- 3. OBJECTIVE OF SAFETY EVALUATION The Technical Specifications for CNS specifies that the setpoint for the differential pressure instrument should be less than that equivalent to 140% of NBR steam flow (Reference 1). However, the differential pressure setpoint was originally established as 118 psid for each of the four stes.a lines which corresponds to approximately 150% of NBR steam flow. This apparent anomaly is described in more detail in Section 5 of
- this report. Also, this data applies to an ideal steam line configurat-ion where a differential pressure of 48.86 psid is' calculated for the 100% NBR steam flow (Reference 2). Due to variation of the steam line configurations, the differential pressures for the four MSLs at CNS have a spread of 48 psid for MSL C to 56 psid for MSL A during normal full power operation. The differential pressure for MSL C is in agreement with the theoretical calculation. This means that the 118 psid setpoint would be considerably less than 150% of NBR steam flow for the other steam lines. The setpoint of 118 psid accounts for such variations of
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the affects of. steam line configurations and is considered as a reason-able setpoint for CNS. Nevertheless, this setpoint is in apparent violation with the plant Technical Specifications. Therefore, a safety evaluation was performed to determine the safety implication of allowing e i
the MSL high flow setpoint to remain at approximately 150% instead of 140% of NBR steam flow.
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- 4. SAFETY EVALUATION The variation of the MSL high flow setpoint from 150 to 140% of the NBR steam flow would not affect the analysis for a guillotine break of the MSL assumed in Section XIV-6 of the Updated Safety Analysis Report (USAR) for CNS (Reference 3). Thus, the safety implication of the setpoint of 150% of rated flow is limited to the difference in ability to detect a break between 150% to 140%. As noted in Reference 3, a setpoint of 140% of rated steam flow would detect steam line breaks greater than 0.3 ft 2. The dose release for such a break is only
~7 of that allowed by 10CFR100. A setpoint of 150% of NBR steam 2.7x10 2
flow would detect steam line breaks greater than 0.38 ft , gg,,, ,37 breaks below 0.3 ft 2are now being detected by other sensors, high temperature, etc, a high flow setpoint of 150% for CNS would increase the break detection requirement of the other sensors from 0.3 to 0.38 ft 2, a mere increase of 0.08 ft 2. The increase in break size detectable o by the high flow instruments does not necessarily affect the ability of the other sensors to detect and isolate the break. The response time for the other sensors would decrease becau.se of the larger break area.
The larger break area would result in larger break flow and depressurize the main steam lines faster. The temperature sensors would also respond faster. If feedt.ater and other high pressure makeup systems are not available, the low water level isolation setpoint would also be reached earlier with a larger break. A conservative evaluation shows that the difference in tote.1 dose release between the two break sizes is approximately 10%. This 10% increase in the extremely small dose calculated for the 140% setpoint will not significantly change the existing margin for the limits allowed by 10CFR100. Therefore, the ability of the plant to detect and isolate a MSL break outside the containment would not be affected by the equivalent setpoint of 150% of NBR steam flow (118 psid) for the MSL high flow. Therefore, the 118 psid setpoint does not present a safety concern and does not have any l safety implication.
- 5. IMPACT ON USAR The setpoint for the MSL high flow for CNS was set at 150% of NBR steam flow versus 140% because the design of each venturi was based on 105% of NBR steam flow. That is, the rated steam flow of the venturi was designed to 105% of NBR and the high flow setpoint became 140% of 105%
of NBR steam flow (1.4 X 1.05 X NBR) which resulted in the 150% setting.
The value of 105% of NBR steam flow has been used as a conservative basis for some safety analyses and for consistency with the rating of
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the turbine which is typically at 105% of NBR.
For the postulated event of MSL break outside the containment, the venturi will limit the break flow through the broken steam line. The maximum flow through the venturi is approximately 200% of rated steam flow. As described above, however, the designed flow is actually 200%
of 105% of NBR steam flow. This raises a concern to the basis of the "200% of rated steam flow" assumed in the safety analysis in Reference
- 3. For each of the four venturis, the 100% of NBR steam flow for CNS is 0
2.39x106 lb/hr or 664 lb/sec, 105% of NBR is 2.51x10 lb/hr or 697 6
lb/sec, 200% of NBR is 4.78x10 lb/hr or 1328 lb/sec, and 200% of 105%
6 of NBR is 5.02x10 lb/hr or 1393 lb/sec. The MSL break outside containment analysis in Reference 3 conservatively assumes that the f maxtema break flow used is 1500 lb/see which is more than 200% of either the NBR steam flow or the 105% of NBR steam flow. Therefore, the safety analysis for the MSL break outside the containment was performed with a conservative flow rate which in consistent with the actual size of the venturi. This demonstrates that sizing the venturi at 105% of NBR steam l flow is not an unreviewed safety question.
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- 6. REFERENCES
- 1. " Cooper Nuclear Station License and Technical Specifications",
Docket 50-298, DPR-46.
- 2. " Flow Element Components", General Electric Company, Design Specification Data Sheet, 21A1058AT, Rev. O.
- 3. " Nebraska Public Power District Coopar Nuclear Station Updated Safety Analysis Report", Section XIV-6, Docket 50-298.
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