Letter Sequence Other |
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Results
Other: AECM-83-0146, Forwards Response to Five of Six Items Contained in Re Util Work Addressing Containment Issues.Evaluation of Condensation Oscillation Loads Produced by Discharges from RHR HX Relief Valve Will Be Completed by Late Apr, AECM-84-0443, Forwards Approach to Chugging,Assessment of RHR Steam Discharge Condensation Oscillation in Mark III Containments, Per 830323 Commitment.Ae Evaluation Will Be Completed by mid-Dec 1984, AECM-85-0018, Forwards Evaluation of RHR Relief Valve Condensation Oscillation Loads on Structures & Piping.Info Demonstrates That All Loads Which Could Be Produced by Actuation of RHR HX Relief Valves Acceptable, AECM-85-0046, Provides Basis That Weir Wall Overflow Following Postulated Inadvertent Upper Containment Pool Dump Does Not Involve Significant Safety Problem.Evaluation Plans & Justification Provided, AECM-85-0233, Forwards Weir Wall Overflow Probability Analysis. Because of Very Low Probability of Occurrence & Leak of Safety Significance for Worst Case Postulated Overflow (Ref AECM-85/0046),corrective Actions Not Warranted, AECM-85-0376, Submits Info Re Wier Wall Overflow,Per 851021 Request.Topics Addressed Include Overflow Effect Upon Uninsulated Drywell Piping & Evaluation of Recirculation Piping Relative to Predicted Water Level.Ge Rept Re Containment Issue Encl, AECM-86-0143, Clarifies Planned Resolution of Possible Loads from Chugging Caused by Steam & Gas Venting Into Suppression Pool Through RHR Relief Valve Discharge Line.Rhr HX Vent Line Header Will Have Dedicated Discharge Line to Pool, ML20097F052, ML20137Q230, ML20204J145
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MONTHYEARAECM-83-0146, Forwards Response to Five of Six Items Contained in Re Util Work Addressing Containment Issues.Evaluation of Condensation Oscillation Loads Produced by Discharges from RHR HX Relief Valve Will Be Completed by Late Apr1983-03-23023 March 1983 Forwards Response to Five of Six Items Contained in Re Util Work Addressing Containment Issues.Evaluation of Condensation Oscillation Loads Produced by Discharges from RHR HX Relief Valve Will Be Completed by Late Apr Project stage: Other ML20137Q2301983-07-31031 July 1983 Containment Issue Isolation Transient Project stage: Other ML20097F0521984-03-31031 March 1984 Approach to Chugging,Assessment of RHR Steam Discharge Condensation Oscillation in Mark III Containments Project stage: Other AECM-84-0443, Forwards Approach to Chugging,Assessment of RHR Steam Discharge Condensation Oscillation in Mark III Containments, Per 830323 Commitment.Ae Evaluation Will Be Completed by mid-Dec 19841984-09-0707 September 1984 Forwards Approach to Chugging,Assessment of RHR Steam Discharge Condensation Oscillation in Mark III Containments, Per 830323 Commitment.Ae Evaluation Will Be Completed by mid-Dec 1984 Project stage: Other AECM-85-0018, Forwards Evaluation of RHR Relief Valve Condensation Oscillation Loads on Structures & Piping.Info Demonstrates That All Loads Which Could Be Produced by Actuation of RHR HX Relief Valves Acceptable1985-01-24024 January 1985 Forwards Evaluation of RHR Relief Valve Condensation Oscillation Loads on Structures & Piping.Info Demonstrates That All Loads Which Could Be Produced by Actuation of RHR HX Relief Valves Acceptable Project stage: Other AECM-85-0046, Provides Basis That Weir Wall Overflow Following Postulated Inadvertent Upper Containment Pool Dump Does Not Involve Significant Safety Problem.Evaluation Plans & Justification Provided1985-02-25025 February 1985 Provides Basis That Weir Wall Overflow Following Postulated Inadvertent Upper Containment Pool Dump Does Not Involve Significant Safety Problem.Evaluation Plans & Justification Provided Project stage: Other ML20126J2671985-06-11011 June 1985 Forwards Request for Addl Info Re Steam Condensing Mode of RHR Sys.Info Requested by 850715 Project stage: RAI AECM-85-0287, Responds to NRC 850611 Request for Addl Info Re Steam Condensing Mode.Listed Methods for Predicting Chugging Will Be Evaluated.Evaluation Scheduled for Completion by 8512201985-09-13013 September 1985 Responds to NRC 850611 Request for Addl Info Re Steam Condensing Mode.Listed Methods for Predicting Chugging Will Be Evaluated.Evaluation Scheduled for Completion by 851220 Project stage: Request AECM-85-0233, Forwards Weir Wall Overflow Probability Analysis. Because of Very Low Probability of Occurrence & Leak of Safety Significance for Worst Case Postulated Overflow (Ref AECM-85/0046),corrective Actions Not Warranted1985-10-0404 October 1985 Forwards Weir Wall Overflow Probability Analysis. Because of Very Low Probability of Occurrence & Leak of Safety Significance for Worst Case Postulated Overflow (Ref AECM-85/0046),corrective Actions Not Warranted Project stage: Other AECM-85-0376, Submits Info Re Wier Wall Overflow,Per 851021 Request.Topics Addressed Include Overflow Effect Upon Uninsulated Drywell Piping & Evaluation of Recirculation Piping Relative to Predicted Water Level.Ge Rept Re Containment Issue Encl1985-11-27027 November 1985 Submits Info Re Wier Wall Overflow,Per 851021 Request.Topics Addressed Include Overflow Effect Upon Uninsulated Drywell Piping & Evaluation of Recirculation Piping Relative to Predicted Water Level.Ge Rept Re Containment Issue Encl Project stage: Other AECM-86-0012, Discusses Status of Util Activities in Response to NRC 850611 Request for Addl Info Re Review of Possible Condensation Oscillation Loads from Chugging. Dedicated Lines from RHR HX Noncondensible Vents Preferred Resolution1986-01-28028 January 1986 Discusses Status of Util Activities in Response to NRC 850611 Request for Addl Info Re Review of Possible Condensation Oscillation Loads from Chugging. Dedicated Lines from RHR HX Noncondensible Vents Preferred Resolution Project stage: Request AECM-86-0143, Clarifies Planned Resolution of Possible Loads from Chugging Caused by Steam & Gas Venting Into Suppression Pool Through RHR Relief Valve Discharge Line.Rhr HX Vent Line Header Will Have Dedicated Discharge Line to Pool1986-05-16016 May 1986 Clarifies Planned Resolution of Possible Loads from Chugging Caused by Steam & Gas Venting Into Suppression Pool Through RHR Relief Valve Discharge Line.Rhr HX Vent Line Header Will Have Dedicated Discharge Line to Pool Project stage: Other AECM-86-0175, Corrected Response to 860326 Request for Addl Info Re Listed Humphrey Concerns,Including Addl Submerged Structure Loads & Containment Open Area,Correcting Omission of Page 2 from Attachment 31986-08-14014 August 1986 Corrected Response to 860326 Request for Addl Info Re Listed Humphrey Concerns,Including Addl Submerged Structure Loads & Containment Open Area,Correcting Omission of Page 2 from Attachment 3 Project stage: Request ML20204J1451987-03-23023 March 1987 SER Supporting Licensee Responses to Humphrey Concerns on Safety of Mark III Containment Design Project stage: Other ML20204J1161987-03-23023 March 1987 Forwards SER Re Licensee Responses to Humphrey Concerns on Mark III Containment Design.Humphrey Concerns Satisfactorily Resolved.Licensee Will Continue to Prohibit Use of RHR Steam Condensing Mode Until Design Mods Implemented Project stage: Approval 1985-11-27
[Table View] |
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Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML20206J1201999-04-30030 April 1999 Redacted ME-98-001-00, Pressure Locking & Thermal Binding Test Program on Two Gate Valves with Limitorque Actuators ML20206T7991999-01-31031 January 1999 Iodine Revolatizitation in Grand Gulf Loca ML20155F1961998-09-0101 September 1998 Engineering Rept for Evaluation of BWR CR Drive Mounting Flange Cap Screw ML20206J1271998-04-30030 April 1998 Pressure Locking Thrust Evaluation Methodology for Flexible Wedge Gate Valves ML20138A2041997-03-31031 March 1997 Pre-SALP Rept for SALP Period 960225-970906 ML20199G7191997-01-28028 January 1997 Rev 0 to Grand Gulf Nuclear Station Engineering Rept GGNS-97-0002 for GL 96-06 Evaluation of Drywall & Containment Penetrations ML20141A1321996-07-31031 July 1996 Prediction of Onset of Fission Gas Release from Fuel in Generic Bwr ML20115A5851996-06-20020 June 1996 GGNS Engineering Rept for Evaluation of Ampacity Deratings for Thermo-Lag Fire Barrier Encl Cables in Fire Areas/Zones OC214,OC302,OC308 & 1A539 ML20108D0861996-04-30030 April 1996 Surveillance Specimen Program Evaluation for Grand Gulf Nuclear Station ML20115A5831996-04-18018 April 1996 Engineering Rept for Evaluation of Ampacity Deratings for Thermo-Lag Fire Barrier Encl Cables in Fire Areas/Zones OC202,OC402,OC702 & 1A316 ML20100J8621996-02-22022 February 1996 Engineering Rept Safety/Relief Valves Safety Function Lift Setpoint Tolerance Relaxation Summary Rept ML20094M5241995-11-20020 November 1995 Self Assessment SALP Period 940227-960224 ML20078S1351995-02-0808 February 1995 SWS Operational Performance Insp (Swsopi) for Ggns ML20072C9891994-04-30030 April 1994 Survey of Melcor Assessment & Selected Applications, Third Draft ML20063H7791993-12-31031 December 1993 1993 Self Assessment ML20046A2601993-06-18018 June 1993 Safety Sys Functional Assessment, Low Pressure Core Spray & RCIC Sys. ML18010B0841993-05-0505 May 1993 NRC Licensing Submittal Review of Licensing Conditions Imposed by NUREG-1216. ML18010A9521992-11-30030 November 1992 NRC Licensing Submittal Review of Licensing Conditions Imposed by NUREG-1216. ML20116K8001992-11-13013 November 1992 Quarterly Status Rept for Period Ending 920930, 'Degraded Core Accident Hydrogen Control Program.' ML20095K0951992-04-30030 April 1992 Resolution of High Groundwater Levels at Grand Gulf Nuclear Station ML20086K4741991-10-31031 October 1991 Cycle 6 Plant Transient Analysis ML20086K4991991-10-31031 October 1991 Cycle 6 Reload Analysis ML20086K5111991-10-31031 October 1991 LOCA Analysis for Single Loop Operation ML20086K5191991-10-31031 October 1991 Nonproprietary Siemens Nuclear Power Corp Grand Gulf Unit 1 Reload XN-1.3,Cycle 4 Mechanical Design Rept Suppl 1 ML20029C3461991-03-31031 March 1991 Simulator Certification Rept for Grand Gulf Nuclear Station. ML20073K8451991-03-31031 March 1991 Entergy Nuclear Performance Rept,Mar 1991 ML20043B6871990-03-31031 March 1990 Degraded Core Accident Hydrogen Control Program, Quarterly Status Rept for Jan-Mar 1990 ML19353B1281989-12-0101 December 1989 Number 1 Low Pressure Turbine Rotor Ultrasonic Insp Summary. ML19332D1691989-09-30030 September 1989 Quarterly Status Rept for Jul-Sept 1989 Degraded Core Accident Hydrogen Control Program. AECM-89-0165, Degraded Core Accident Hydrogen Control Program, Quarterly Status Rept for Quarter Ending 8906301989-06-30030 June 1989 Degraded Core Accident Hydrogen Control Program, Quarterly Status Rept for Quarter Ending 890630 ML20235X2621989-02-28028 February 1989 Criticality Safety Analysis of Grand Gulf Nuclear Station Unit 1 Spent Fuel Storage Racks W/Gaps in Neutron Absorbing Panels ML20235N3271989-02-22022 February 1989 Rept on Evaluation of Torsion Shear Stress Components in Design of Pipe Supports ML20247B1901988-11-30030 November 1988 Revised Flow Dependent Thermal Limits AECM-88-0021, Degraded Core Accident Hydrogen Control Program, Quarterly Rept Covering Period 871031-12311987-12-31031 December 1987 Degraded Core Accident Hydrogen Control Program, Quarterly Rept Covering Period 871031-1231 AECM-87-0163, Degraded Core Accident Hydrogen Control Program, Quarterly Status Rept for Apr-June 19871987-08-27027 August 1987 Degraded Core Accident Hydrogen Control Program, Quarterly Status Rept for Apr-June 1987 ML20065N4861987-04-30030 April 1987 Flux Wire Dosimeter Evaluation for Grand Gulf Nuclear Power Station,Unit 1 ML20214U1711987-03-31031 March 1987 Technical Rept 86.2GG, Verification of Individual Plant Evaluation for Grand Gulf. W/Jr Siegel 870313 Release Ltr AECM-87-0102, Degraded Core Accident Hydrogen Control Program, Quarterly Status Rept for Quarter Ending on 8703311987-03-31031 March 1987 Degraded Core Accident Hydrogen Control Program, Quarterly Status Rept for Quarter Ending on 870331 AECM-87-0015, Degraded Core Accident Hydrogen Control Program, Quarterly Status Rept for Quarter Ending 8612311986-12-31031 December 1986 Degraded Core Accident Hydrogen Control Program, Quarterly Status Rept for Quarter Ending 861231 ML20213G2221986-11-10010 November 1986 Degraded Core Accident Hydrogen Control Program, Quarterly Status Rept for Quarter Ending 860930 ML20151A8791986-09-30030 September 1986 Rev 1 to SPDS Safety Analysis ML20215L2311986-09-30030 September 1986 Rev 0 to Grand Gulf Nuclear Station Safety Relief Valve Fatigue Evaluation ML20203N4171986-04-30030 April 1986 Rev 2 to Tdi Owners Group App Ii:Generic Maint Matrix & Justifications ML20199L1091986-04-0303 April 1986 Station Blackout Evaluation Rept,Grand Gulf Nuclear Station ML20207G6991986-03-31031 March 1986 Feedwater Heaters Out-of-Svc Analysis,Grand Gulf Units 1 & 2 ML20155B0911986-02-28028 February 1986 Single Loop Operating Analysis ML20205J7381986-02-18018 February 1986 Degraded Core Accident Hydrogen Control Program, Quarterly Status Rept for Quarter Ending 851231 ML20141N8871986-02-13013 February 1986 Metallurgical Evaluation of Recirculation Piping Ctr Cross Surface Cracking Following Induction Heat Stress Improvement,Grand Gulf Nuclear Station - Unit 1, Final Rept L-09-100, Rev 0 to MPL-09-100, Grand Gulf In-Plant Safety Relief Valve Test Fatigue Evaluation Final Rept1986-01-31031 January 1986 Rev 0 to MPL-09-100, Grand Gulf In-Plant Safety Relief Valve Test Fatigue Evaluation Final Rept ML20137E4471985-11-19019 November 1985 Metallurgical Evaluation of Recirculation Piping Ctr Cross Surface Cracking Following Induction Heat Stress Improvement,Grand Gulf Nuclear Station - Unit 1, Interim Rept 1999-04-30
[Table view] Category:REPORTS-TOPICAL (BY MANUFACTURERS-VENDORS ETC)
MONTHYEARML20077G5671991-04-30030 April 1991 Vols 1 & 2 to Grand Gulf 1 ANF-1.4 Design Rept Mechanical, Thermal-Hydraulic & Neutronic Design for Advanced Nuclear Fuels 9X9-5 Fuel Assemblies ML20043E8041990-06-0101 June 1990 Nonproprietary Criticality Safety Analysis for Grand Gulf Fuel Storage Racks W/ANF-1.4 Fuel Assemblies. ML20235U8191987-08-31031 August 1987 Cycle 3 Plant Transient Analysis ML20099L4791984-11-30030 November 1984 Revised Vols 1-4, Tdi Diesel Generator Design Review & Quality Revalidation Rept ML20137Q2301983-07-31031 July 1983 Containment Issue Isolation Transient ML20072D2941983-05-11011 May 1983 Nonproprietary Version of Grand Gulf Drywell Break Sensitivity Summary ML20040D6511981-12-31031 December 1981 Nonproprietary Version of OPS-37A36, Verification of CLASIX-3 Computer Program. ML20049J5471979-12-31031 December 1979 SAP4G07 User Manual Static & Dynamic Analysis of Mechanical & Piping Components by Finite Element Method. 1991-04-30
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217F9921999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Grand Gulf Nuclear Station,Unit 1.With ML20212F5641999-09-23023 September 1999 SER Concluding That All of ampacity-related Concerns Have Been Resolved & Licensee Provided Adequate Technical Basis to Assure That All of Thermo-Lag Fire Barrier Encl Cables Operating within Acceptable Ampacity Limits ML20211Q3171999-09-0909 September 1999 Safety Evaluation Accepting BWROG Rept, Prediction of Onset of Fission Gas Release from Fuel in Generic BWR, Dtd July 1996 ML20216E4881999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Grand Gulf Nuclear Station.With ML20211A6921999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Grand Gulf Nuclear Station,Unit 1.With ML20209J1961999-07-12012 July 1999 Special Rept 99-001:on 990528,smoke Detectors Failed to Alarm During Performance of Routine Channel Functional Testing.Applicable TRM Interim Compensatory Measure of Establishing Roving Hourly Fire Patrol Was Implemented ML20196K4981999-07-0101 July 1999 Safety Evaluation Authorizing PRR-E12-01,PRR-E21-01, PRR-P75-01,PRR-P81-01,VRR-B21-01,VRR-B21-02,VRR-E38-01 & VRR-E51-01.Concludes That Alternatives Proposed by EOI Acceptable ML20209G0691999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Grand Gulf Nuclear Station,Unit 1.With ML20196A1161999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Grand Gulf Nuclear Station.With ML20206Q4831999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Grand Gulf Nuclear Station Unit 1.With ML20206J1201999-04-30030 April 1999 Redacted ME-98-001-00, Pressure Locking & Thermal Binding Test Program on Two Gate Valves with Limitorque Actuators 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 ML20205P8771999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Grand Gulf Nuclear Station,Unit 1.With 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 ML20207K5141999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Grand Gulf Nuclear Station,Unit 1.With ML20206T7991999-01-31031 January 1999 Iodine Revolatizitation in Grand Gulf Loca ML20207A8301998-12-31031 December 1998 1998 Annual Operating Rept for Ggns,Unit 1 ML20206R9501998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Grand Gulf Nuclear Station,Unit 1.With ML20206D7721998-12-31031 December 1998 South Mississippi Electric Power Association 1998 Annual Rept ML20198E2481998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Grand Gulf Nuclear Station,Unit 1.With ML20195F4121998-11-13013 November 1998 Rev 16 to GGNS-TOP-1A, Operational QA Manual (Oqam) ML20195C4841998-11-0606 November 1998 SER Accepting QA Program Change to Consolidate Four Existing QA Programs for Arkansas Nuclear One,Grand Gulf Nuclear Station,River Bend Station & Waterford 3 Steam Electric Station Into Single QA Program ML20195C2791998-11-0505 November 1998 BWR Feedwater Nozzle Inservice Insp Summary Rept for GGNS, NUREG-0619-00006 ML20195F4801998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Grand Gulf Nuclear Station,Unit 1.With ML20155C1351998-10-26026 October 1998 Rev B to Entergy QA Program Manual ML20154K2391998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Grand Gulf Nuclear Station Unit 1.With ML20155F1961998-09-0101 September 1998 Engineering Rept for Evaluation of BWR CR Drive Mounting Flange Cap Screw ML20153B2161998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Grand Gulf Nuclear Station,Unit 1.With ML20237B6661998-07-31031 July 1998 Monthly Operating Rept for Jul 1998 for Grand Gulf Nuclear Station,Unit 1 ML20236R0231998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Grand Gulf Nuclear Station,Unit 1 ML20155J0811998-05-31031 May 1998 10CFR50.59 SE for Period Jan 1997 - May 1998 ML20249B1251998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Grand Gulf Nuclear Station,Unit 1 ML20248B6261998-05-11011 May 1998 Rev 6 to Grand Gulf Nuclear Station COLR Safety-Related ML20217Q6701998-05-0606 May 1998 SER Approving Proposed Postponement of Beginning Augmented Exam Requirements of 10CFR50.55a(g)(6)(ii)(A)(2) at Grand Gulf for Circumferential Shell Welds for Two Operating Cycles ML20206J1271998-04-30030 April 1998 Pressure Locking Thrust Evaluation Methodology for Flexible Wedge Gate Valves ML20247F3591998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Grand Gulf Nuclear Plant,Unit 1 ML20217M8951998-04-30030 April 1998 QA Program Manual ML20217P8281998-04-0707 April 1998 Safety Evaluation Accepting Relief Authorization for Alternative to Requirements of ASME Section Xi,Subarticle IWA-5250 Bolting Exam for Plants,Per 10CFR50.55a(a)(3)(i) ML20217P0381998-04-0606 April 1998 Safety Evaluation Supporting Amend 135 to License NPF-29 ML20217A0291998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Grand Gulf Nuclear Sation,Unit 1 ML20216J4211998-03-18018 March 1998 SER Approving Alternative to Insp of Reactor Pressure Vessel Circumferential Welds for Grand Gulf Nuclear Station ML20216J2021998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Grand Gulf Nuclear Station,Unit 1 ML20203A2891998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Grand Gulf Nuclear Station ML20247B4111997-12-31031 December 1997 1997 Annual Financial Rept for South Mississippi Electric Power Association ML20203H9741997-12-31031 December 1997 1997 Annual Operating Rept, for Ggns,Unit 1 ML20198P1121997-12-31031 December 1997 Monthly Operating Rept for Dec 1997 for Grand Gulf Nuclear Station,Unit 1 ML20203B5581997-12-0404 December 1997 Special Rept 97-003:on 971111,valid Failure of Div 2 EDG Occurred,Due to Jacket Water Leak.Failure Reported,Per Plant Technical Requirements Manual Section 7.7.2.2 ML20203K4031997-11-30030 November 1997 Monthly Operating Rept for Nov 1997 for Grand Gulf Nuclear Station,Unit 1 ML20199H3711997-11-19019 November 1997 SER Accepting Approving Request Relief from Requirements of Section XI, Rule for Inservice Insp of NPP Components, of ASME for Current or New 10-year Inservice Insp Interval IAW 50.55(a)(3)(i) of 10CFR50 ML20199F3431997-11-18018 November 1997 SER Accepting Rev 15 of Operational Quality Assurance Manual for Grand Gulf Nuclear Station,Unit 1 1999-09-09
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Text
. . __ _ _ --
. GENERAL ELECTRIC CO.
NUCLEAR ENERGY DIVISION SAN JOSE, CA 9512S DESIGN MEMO #123-8324 DRF #124-98 CONTAINMENT ISSUE ISOLATION TRANSIENT Prepared by: i V N* 1 II[I1 H. M. Srivastava, Principal ' Engineer Plant Piping Design l
i l
Reviewed by: N*k- maw H. L. Hwang, Principa10 Engineer i
Plant Dynamic Methods & Applications l-i r
7_/2.l&)
Approved by: .C ,
. C. Atwell, Manager Plant Piping Design l
l
' JULY 1983 0512050190 PDR ADOCK h PDR h 16 l
P
_ . _ - - _ _ . - - _ _ _ , - _ _ --_ _ _ _1.__._._._.__--_____.__- _ . _ _ _ _ _ _ _ _ _ . . _ _
CONTENTS .
f*81
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . . . . . I 2.0 PURPOSE . . . . . . . . . . . . . . . . . . . . . . . 1 3.0 CLASSIFICATION OF EVENT . . . . . . . . . . . . . . . 1 4.0 DRYWELL PLOODING TRANSIENT STRESS EVALUATION . . . . . 1 4.1 RECIRCULATION PUMP . . . . . . . . . . . . . . . 1 4.2 RECIRCULATION PIPING . . . . . . . . . . . . . . 1 S.O RESULTS . . . . . . . . . . . . . . . . . . . . . . . 2 5.1 RECIRCULATION PUMP . . . . . . . . . . . . . . . 2
.; S.2 RECIRCULATION PIPING . . . . . . . . . . . . . . 2
6.0 CONCLUSION
S AND RECO M NDATIONS . . . . . . . . . . . 3
7.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . 4 APPENDIX A - LION 401 - COMPITTER PROGRAM . . . . . . . S e
e e a , , , - - - -,_q , , - - ,. . , , , ,, _~ .-.-n-, . - - - - , - .--- _.-~ , - . - , , ,-,, , - - .,nn,, , , -- - - , , , -. , - - - - - , , - - - , - - - + - - - - - . . . - - - - ,
.i, 1,0 INTRODUCTION.
- Various Scenarios lead to conditions where suppression pool water may overflow / backflow over the weir, splashing onto, or partial immersion of, recirculation pumps and piping become possible. 1hezzal shock, and resulting fatigue, will add to the system imposed (Service Levels A and B) fatigue life for the recirculation piping system and.may consume'all remaining useful life. Occurrence of any other concurrent, or subsequent, loading could result in a recirculation break LOCA.
2.0 PURPOSE.
The purpose of this report is to show that if this incredible event, were to occur it would not damage the piping system to an extent whereby loading could cause a recirculation break LOCA.
3.0 CLASSIFICATION OF EVENT.
Drywell flooding is"a rare event and it is postulated to occur only once or twice during a 40 year plant life. This event can be classified as Service Level C or Service Level D event (Reference 7.1) and no
- fatigue analysis is required. .
4.0 DRYWELL FLOODING TRANSIENT STRESS EVALUATION.
Although this event is classified as Service Levels C or D event, simplified fatigue analyses were made (Reference 7.2) for the pump casing and a typical BWR-6 recirculation piping system subjected to this postulated transient. The assumptions for this evaluation were as follows.
4.1 RECIRCULATION PUMP The present capability of the recirculation pumps as documented by the Byron Jackson stress report, states the pump casing can withstand a 0
maximum temperature gradient of 350 F without any fatigue evaluation.
However, a fatigue evaluation was made with a pump casing temperature gradient of 4500F.
4.2 RECIRCULATION PIPING A fatigue evaluation was made for this event end the stresses were added to the piping stresses due to system loading and thermal transients.
The thermal gradient was evaluated for this event using the LION 401 computer program. The surface temperature of the pipe was assumed to be 700 immediately after water reaches the pipe surface. The boundary temperature and heat transfer coefficient were conservatively assumed as follows: .
4 I
i
l i
. l i
Temperature " 528 - h" . ;
'F BTU HR-FT *F
\
70 .
0 : Time (Sec) M H 0 15 Time (Sec)
TEMPERATURE PR0cILE HEAT TRANSFER COEFFICIENT Theheattransfegto,waterincreasesrapidlyin15secondsto 1000 STU/hr F Ft and partially destroy the insulation. This causes o
high thermal gradient in the pipe. ,
5.0 RESULTS.
The results of the evaluation (Section 4.0) were:
5.1 RECIRCULATION PUMP 5.1.1 The calculated allowable cycles for this transient were 150 cycles which gives a fatigue usage factor of 0.007 (1/150).
5.1.2 Distortion There is a chance the 4500F temperature difference would cause local yielding such that a dimensional check of the critical parts would be required. The recirculation pump motor cannot tolerate flooding without subsequent cleaning, oil change and drying its' winding.
These operations and check have to be done after this event.
5.1.3 Fracture Toughness ,
The pump casing is cast sustenitic stainless steel, so brittle fracture is not a concern.
The pump cover case bolts are ferritic steel. The mean temperature of the cover is: 1/2 (450) + 100 = 3500F. This temperature is above NDTT (Nil Ductility Transition Temperature) of ferritic steel.
5.2 RECIRCULATION PIPING .
5.2.1 The calculated maximum temperature gradients for this transient
- described in Section 4.2 were:
AT g = 354*F AT = 88'F 2
T'g,
= 46'F (Due to thick pump casing and thin pipe) 2
l c. .
The stresses due to the above thermal gradients were added to the
- - stresses due to the system loads and thermal gradients. The allowable cycles with these stresses were 900, which gives a fatigue usage factor of 0.001 (1/900).
5.2.2 Distortion .
The temperature gra' d ient may distort the pipe at the pipe to pump casing weld location which will not affect the function of the recirculation piping.
5.2.3 Fracture Toughness For the austenitic stainless steel recirculation piping, fracture is not a concern for this e.ent.
6.0 CONCLUSION
S AND REC 00NENDATIONS.
The above transient is similar to events of " Improper Start of a Cold Loop", except the temperature shock AT is 3980F instead of 4500F. So the transient is not totally new for the recirculation piping system design. This event is not a safety concern based on the fatigue 4 evaluation and the following reasons.
i 1) The stresses produced by the event are in a category (secondary 4 peak) that do not require evaluation except for Service Levels A 4 8 conditions. These peak stresses produced by the thermal shock are important only for fatigue and fatigue usage which, for a few rare events, is not required by the Code or by NRC rules.
- 2) If it were necessary to consider the fatigue usage due to this thermal shock, calculations show; based on worst case conditions, that significant fatigue usage would not result unless there were more than one hundred such cycles.
- 3) Under a worst case condition the potential damage to the piping could be slight distortion at the weld joints. The worst case condition is defined as the insulation being removed and a 4500 temperature difference between the outside and inside of the recirculation pipe. In the event that suppression pool water immersed part of the recirculation piping, we would recommend the insulation of the piping be removed and the weld joints connecting the recirculation piping to the recirculation pump be visually examined for deformation at the next shutdown.
Additionally, a dimensional and alignment check of the pump is recommended. The pump motor must be reconditioned by decontamination and drying the insulation, an electrical check, and,an oil change.
This assumes the motor was flooded.
t 3
4 ..
7.0 REFERENCES
7.1 ASE Boiler and Pressure Vessel Code,Section III Division I - 1980 Edition upto and including Winter 1982 Addenda.
7.2 Design Record File 8124-98, Recirculation flooding.
8 0 e e
.4.
APPENDIX A LION 401 PROGRAM LION 401 is a digital computer program which is used to solve the steady state or transient temperature distribution in any three-dimensional configuration.
The heat source may be externally conducted or internally generated.
In addition to the solving of heat conduction in structural eler.tnts, LION 401 may also be used in such cases as forced convection, free conver. ion, or radiation where the output will yield temperatures and heat fluxes for points representing the surface of the structure.
The program solves the transient heat conduction equations for a three-dimansional field using a first forward difference method.
Input to the program consists of structural geometry, physical properties, boundary conditions, internal heat generation rates and coolant flow properties and rates.
1
=
1 l
5-
. .- . . -. . - . . . . . . .