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Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML20069F0001994-01-24024 January 1994 Vols 1-4 to Shoreham Decommissioning Project Termination Survey Final Rept ML20045C8881993-06-0808 June 1993 Vols 1 & 2 to Refueling Jib Crane 1T31-CRN-008A Incident Root Cause Analysis. W/One Oversize Encl ML20128P6451993-02-28028 February 1993 Snps Decommissioning Project Termination Survey Final Rept for Steam Turbine Sys (N31) ML20128P7431993-02-19019 February 1993 Rev 3 to 93X027, Nuclear QA Surveillance Rept ML20099H5781992-07-31031 July 1992 Rev 4 to Shoreham Defueled Sar ML20101K5791992-06-25025 June 1992 Long Island Power Authority Shoreham Decommissioning Project,Shoreham Nuclear Power Station,Technical Rept on Water Processing & Water Mgt Activities for Reactor Pressure Vessel & Wet Cutting Station ML20082M5081991-08-26026 August 1991 Rev 3 to Shoreham Defueled Sar ML20005F2511990-01-0505 January 1990 Shoreham Nuclear Power Station Defueled Sar. ML19332G1991989-09-18018 September 1989 Rev 0 to Radiological Safety Analysis for Spent Fuel Storage & Handling. ML20245E0181989-06-19019 June 1989 QC 1989 Staffing Rept ML20153G8941988-08-31031 August 1988 Rev 1 to Shoreham Nuclear Power Station Prompt Notification Sys Rept ML20196K6311988-06-29029 June 1988 QC Div 1988 Staffing Rept ML20148A4851988-02-29029 February 1988 Shoreham Nuclear Power Station PRA W/Supplemental Containment Sys ML20214U2021987-01-31031 January 1987 Technical Rept 86.2SH Verification of IPE for Shoreham. W/ 870313 Release Memo ML20210M6001986-12-31031 December 1986 Emergency Planning Federal Involvement in Preparedness Exercise at Shoreham Nuclear Plant. Related Correspondence ML20214T6021986-12-31031 December 1986 Rev 6 to Plant Design Assessment Rept for Safety/Relief Valves & LOCA Loads, Vols 1 & 2.Proprietary Suppl Withheld (Ref 10CFR2.790) ML20210K2961986-12-31031 December 1986 Nuclear Regulation,Unique Features of Shoreham Nuclear Plant Emergency Planning. Related Correspondence ML20214U3641986-10-31031 October 1986 Shoreham Startup & Low Power Testing Operations,Special Rept:Lilco QA Audit on Training & Qualifications ML20237H6381986-07-31031 July 1986 Compliance W/10CFR50,App I ML20211P6851986-06-30030 June 1986 Implications of Chernobyl-4 Accident for Nuclear Emergency Planning for State of Ny ML20203N4171986-04-30030 April 1986 Rev 2 to Tdi Owners Group App Ii:Generic Maint Matrix & Justifications SNRC-1207, Vols 1 & 2 of Colt Emergency Diesel Generator Info to Be Incorporated in Fsar. W/11 Oversize Drawings1985-11-30030 November 1985 Vols 1 & 2 of Colt Emergency Diesel Generator Info to Be Incorporated in Fsar. W/11 Oversize Drawings ML20128N9651985-05-31031 May 1985 New York Power Pool,1985 Summer Operating Reserve, Projection & Analysis ML20111C0961984-11-20020 November 1984 Rev 1 to Long Island Lighting Co,Shoreham Nuclear Power Station,Prompt Notification Sys Design Rept ML20091Q8851984-06-30030 June 1984 Colt Diesel Generator Summary for Shoreham Nuclear Power Station - Unit 1 ML20092N6631984-06-12012 June 1984 Seismic Survivability Study for MP-45 Diesel Generators ML20091M5451984-05-31031 May 1984 Design Review of Connecting Rods for Tdi DSRV-4 Series Diesel Generators, Final Rept Prepared for Tdi Diesel Generator Owners Group ML20091M5511984-05-23023 May 1984 Investigation of Types AF & Ae Piston Skirts, Final Rept Prepared for Tdi Diesel Generator Owners Group ML20091M5201984-05-22022 May 1984 Draft Final Rept, Evaluation of Emergency Diesel Generator Crankshafts at Shoreham & Grand Gulf, Prepared for Tdi Diesel Generator Owners Group ML20084E5421984-04-30030 April 1984 Emergency Diesel Generator Engine & Auxiliary Module Wiring & Termination Qualification to IEEE-383-1974 ML20087L3511984-03-31031 March 1984 Emergency Diesel Generator Air Start Valve Capscrew Dimension & Stress Analysis, Prepared for Transamerica Delaval,Inc (Tdi) Diesel Generator Owners Group ML20081C4721984-03-12012 March 1984 Design Review of Connecting Rod Bearing Shells for Transamerica Delaval Enterprise Engines ML20087L2851984-02-27027 February 1984 Control Bldg Category I Equipment Balance-of-Plant Qualification Level, Monthly Status Rept ML20086S2221984-02-27027 February 1984 Investigation of Types AF & Ae Piston Skirts ML20080U3971984-02-10010 February 1984 Rept on Special Lifting Devices ML20083F6511983-12-15015 December 1983 Analysis of Replacement Connecting Rod Bearings Emergency Diesel Generators,Fatigue Life Prediction,Shoreham Nuclear Power Station ML20083C6681983-12-0808 December 1983 Metallurgical Analysis of Cracked Piston Skirts from Emergency Diesel Generators,Shoreham Nuclear Power Station ML20081C1571983-10-20020 October 1983 Diesel Generator Status Rept ML20112J2501983-08-31031 August 1983 Critique of Hudson Inst/Lilco Defense of Shoreham Economics. Related Info Encl ML20082D8831983-08-30030 August 1983 Suppression Pool Local-to-Bulk Temp Difference,Shoreham Nuclear Power Station - Unit 1 ML20112J2981983-08-30030 August 1983 Lilco/Hudson Inst Rept on Shoreham:Analysis of Errors Re Property Taxes & Employment. Related Info Encl ML20081L6891983-08-29029 August 1983 Excerpt from Draft FSAR Section 5.2.8, Inservice Insp Program, & Section 5.2.8.1 Provisions for Access to Rcpb ML20085D8921983-07-26026 July 1983 Books 1-3 of Independent Design Review for Shoreham Nuclear Power Station, Final Technical Rept ML20072J9521983-06-30030 June 1983 Independent Design Review for Shoreham Nuclear Power Station, Executive Summary of Final Rept ML20058N6071983-06-30030 June 1983 Environ Qualification Rept for Class 1E Equipment for Shoreham Nuclear Power Station Unit 1 Lilco ML20072K0141983-06-27027 June 1983 Rev 1 to Independent Design Review,Shoreham Nuclear Power Plant ML20076M0861983-06-27027 June 1983 Rev 5 to Environ Qualification Rept for Class IE Equipment for Shoreham Nuclear Power Station Unit 1 ML20072F2401983-06-22022 June 1983 Element-By-Element Review of Lilco Transition Module of Shoreham Nuclear Power Station Offsite Radiological Emergency Response Plan ML20079R7351983-05-31031 May 1983 Shoreham Common Sensors Failures Evaluation Rept ML20073L5911983-04-30030 April 1983 Cable Separation Analysis Rept 1994-01-24
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20140G4481997-05-0101 May 1997 Part 21 Rept Re Potential Defect in Component of Dsrv & Dsr Enterprise Standby Diesel Generator Sys.Recommends That Springs Be Inspected on Periodic Basis,Such as During Refueling Outages ML20135D8011996-11-26026 November 1996 Part 21 Rept Re Two Safety Related Valves Supplied by Velan Valve Corp Were Not in Compliance W/Originally Supplied QA Documentation.Returned Valves to Velan in May 1996 & on 961120 Velan Advised That Valves Had Been Misplaced ML20080G4691995-01-26026 January 1995 Record of Telcon W/Nrc & Licensees 950126 to Clarify Position Re Dispositioning of Exempt Sources Listed in Section 6.3.3 of Shoreham Termination Survey Final Rept Dtd Oct 1994 ML20069F0001994-01-24024 January 1994 Vols 1-4 to Shoreham Decommissioning Project Termination Survey Final Rept ML20058K3841993-12-0909 December 1993 Part 21 Rept Re Potential Defect in Component of Dsrv & Dsr Enterprise Standby DG Sys,Regarding Potential Problem W/ Subcover Assembled Atop Power Head ML20057F2261993-09-30030 September 1993 Safety Evaluation Supporting Exemption Request from Requirements of 10CFR50.54(q) for License NPF-82 ML20056C7181993-07-14014 July 1993 SE Supporting Amend 10 to License NPF-82 ML20045B3551993-06-11011 June 1993 LER 93-001-00:on 930429,refueling Jib Crane Moved in Vicinity of Spent Fuel Pool Using vendor-supplied Lifting Eye in Violation of NUREG-0612.Caused by Failure to Identify Crane as Heavy Load.Meetings held.W/930611 Ltr ML20045C8881993-06-0808 June 1993 Vols 1 & 2 to Refueling Jib Crane 1T31-CRN-008A Incident Root Cause Analysis. W/One Oversize Encl ML20044C1181993-02-28028 February 1993 Shoreham Nuclear Power Station Updated Decommissioning Plan. ML20128P6451993-02-28028 February 1993 Snps Decommissioning Project Termination Survey Final Rept for Steam Turbine Sys (N31) ML20128P7431993-02-19019 February 1993 Rev 3 to 93X027, Nuclear QA Surveillance Rept ML20127H2301993-01-15015 January 1993 Part 21 Rept Re Potential Defeat in Component of Dsrv & Dsr Enterprise Standby DG Sys.Starting Air Distributor Housing Assemblies Installed as Replacement Parts at Listed Sites ML20126B0421992-12-17017 December 1992 Final Part 21 Rept Re Potential Problem W/Steel Cylinder Heads.Initially Reported on 921125.Caused by Inadequate Cast Wall Thickness at 3/4-inch-10 Bolt Hole.Stud at Location Indicated on Encl Sketch Should Be Removed ML20125C7161992-12-0707 December 1992 Part 21 Rept Re Possibility for Malfunction of Declutching Mechanisms in SMB/SB-000 & SMB/SB/SBD-00 Actuators. Malfunction Only Occurs During Seismic Event.Balanced Levers May Be Purchased from Vendor.List of Affected Utils Encl ML20127P5861992-11-23023 November 1992 Followup to 921005 Part 21 Rept Re Potential Defect in SB/SBD-1 Housing Cover Screws.Procedure Re Replacement of SBD-1 Spring Cover Bolts Encl.All Fasteners Should Be Loosened & Removed.List of Affected Utils Encl ML20128B9641992-10-31031 October 1992 Rev 0 to Shoreham Decommissioning Project Termination Survey Plan ML20118B4391992-09-11011 September 1992 Part 21 Rept Re Degradation in Abb Type 27N Undervoltage Relays Used in Electrical Switchgear.Recommends That Users Review Applications Requiring Exposures Greater than 1E03 Rads TID W/Time Delay Function Option ML20099H5781992-07-31031 July 1992 Rev 4 to Shoreham Defueled Sar ML20114A6311992-07-28028 July 1992 Shoreham Decommissioning Plan ML20101K5791992-06-25025 June 1992 Long Island Power Authority Shoreham Decommissioning Project,Shoreham Nuclear Power Station,Technical Rept on Water Processing & Water Mgt Activities for Reactor Pressure Vessel & Wet Cutting Station ML20094L1271992-03-13013 March 1992 Amend 1 to Part 21 Rept 159 Re Potential Defect in Power Cylinder Liner.Initially Reported on 920115.Caused by Liner/ Block Fit & Localized Matl Microstructure.All Drawings & Specs Revised to Address Matl Design Requirements ML20082M5081991-08-26026 August 1991 Rev 3 to Shoreham Defueled Sar PM-91-125, Monthly Operating Rept for Jul 1991 for Shoreham Nuclear Power Station1991-07-31031 July 1991 Monthly Operating Rept for Jul 1991 for Shoreham Nuclear Power Station PM-91-112, Monthly Operating Rept for Jun 1991 for Shoreham Nuclear Power Station1991-06-30030 June 1991 Monthly Operating Rept for Jun 1991 for Shoreham Nuclear Power Station PM-91-075, Monthly Operating Rept for Apr 1991 for Shoreham Nuclear Power Station1991-04-30030 April 1991 Monthly Operating Rept for Apr 1991 for Shoreham Nuclear Power Station ML20024G7171991-04-22022 April 1991 LER 91-001-00:on 910324,RB Normal Ventilation Sys (Rbnvs) Outboard Exhaust Valve Closed for No Apparent Reason.Cause Inconclusive.Sys Restored to Normal Lineup & Rbnvs Outboard Valve Will Be Stroked on Routine basis.W/910422 Ltr SNRC-1806, Revised Pages 2 & 6 to Encl a of 10CFR50.59 Annual Rept for 19901991-04-15015 April 1991 Revised Pages 2 & 6 to Encl a of 10CFR50.59 Annual Rept for 1990 PM-91-058, Monthly Operating Rept for Mar 1991 for Shoreham Nuclear Power Station1991-03-31031 March 1991 Monthly Operating Rept for Mar 1991 for Shoreham Nuclear Power Station PM-91-037, Monthly Operating Rept for Feb 1991 for Shoreham Nuclear Power Station1991-02-28028 February 1991 Monthly Operating Rept for Feb 1991 for Shoreham Nuclear Power Station PM-91-016, Monthly Operating Rept for Jan 1991 for Shoreham Nuclear Power Station1991-01-31031 January 1991 Monthly Operating Rept for Jan 1991 for Shoreham Nuclear Power Station SNRC-1797, 10CFR 50.59 Annual Rept of Facility Changes,Procedure Changes,Tests & Experiments for Jan-Dec 19901990-12-31031 December 1990 10CFR 50.59 Annual Rept of Facility Changes,Procedure Changes,Tests & Experiments for Jan-Dec 1990 SNRC-1794, Shoreham Nuclear Power Station Annual Operating Rept,19901990-12-31031 December 1990 Shoreham Nuclear Power Station Annual Operating Rept,1990 SNRC-1799, Lilco 1990 Annual Rept1990-12-31031 December 1990 Lilco 1990 Annual Rept ML20069Q3901990-12-31031 December 1990 Shoreham Nuclear Power Station Decommissioning Plan. (Filed in Category P) ML20028H0231990-09-28028 September 1990 LER 90-007-00:on 900907,unplanned Actuation of ESF Sys Occurred During I&C Surveillance Test.Caused by Inadequate procedure.SP44.650.16 Revised to Require That Leads Lifted & Individually separated.W/900928 Ltr ML20056A2001990-07-31031 July 1990 Safety Evaluation Supporting Amend 6 to License NPF-82 PM-90-097, Monthly Operating Rept for June 1990 for Shoreham Nuclear Power Station1990-06-30030 June 1990 Monthly Operating Rept for June 1990 for Shoreham Nuclear Power Station ML20055E3911990-06-25025 June 1990 Safety Evaluation Supporting Amend 5 to License NPF-82 PM-90-083, Monthly Operating Rept for May 1990 for Shoreham Nuclear Power Station1990-05-31031 May 1990 Monthly Operating Rept for May 1990 for Shoreham Nuclear Power Station 05000322/LER-1987-0351990-05-16016 May 1990 LER 87-035-02:on 871221,880106 & 0330,high Energy Line Break Logic Isolations of RWCU & Main Steam Line Drain Valves Occurred.Caused by Problems W/Temp Monitoring Units. Grounding Scheme Changed & Transformers Rewired 05000322/LER-1986-0321990-05-16016 May 1990 LER 86-032-01:on 860728,RWCU Isolated on High Differential Flow Sensed by Steam Leak Detection Sys While Placing Filter Demineralizers in Operation.Cause Not Determined. Operating Procedures Revised to Monitor RWCU Sys 05000322/LER-1987-0091990-05-16016 May 1990 LER 87-009-01:on 870203,full Reactor Trip Occurred Due to Perturbation in Ref Leg.Caused by Spurious Low Level Reactor Pressure Vessel Water Level Signal.Existing Level & Pressure Transmitters Replaced W/Newer Models 05000322/LER-1989-0031990-05-16016 May 1990 LER 89-003-01:on 890310,emergency Diesel Generator (EDG) 102 Manually Shutdown During 18-month Surveillance Test Due to Failure of EDG Output Breaker.Cause Not Determined. Replacement Breaker Installed in Cubicle 102-8 05000322/LER-1985-0591990-05-16016 May 1990 LER 85-059-01:on 851219,half Reactor Trip,Full NSSS Shutoff Sys Isolation & Reactor Bldg Standby Ventilation Sys Initiation Occurred Due to Loss of Power to Reactor Protection Sys Bus B.Assembly Breaker Reset 05000322/LER-1988-0151990-05-16016 May 1990 LER 88-015-02:on 880916,seismic Monitoring Instrumentation, Including Peak Acceleration Recorders,Removed from Svc for More than 30 Days Due to Corrosion on Scratch Plates.Cover Gasket Replaced & Thermal Barrier Mount to Be Installed 05000322/LER-1988-0031990-05-16016 May 1990 LER 88-003-01:on 880322,unplanned Automatic Initiation of Reactor Bldg Standby Ventilation Sys Side a Occurred During Deenergization of Relay.Caused by Close Placement of Relay Terminals.Wiring Inside Electrical Panels Reworked 05000322/LER-1986-0141990-05-16016 May 1990 LER 86-014-01:on 860305,full Reactor Trip Occurred Due to Momentary False Low Vessel Level Signal,Causing Hydraulic Pressure Spike in Ref Leg A.Bourton Tube Type Pressure Transmitter Replaced W/Rosemount Model 1153 05000322/LER-1987-0121990-05-16016 May 1990 LER 87-012-01:on 870504,uplanned Actuation of ESF Occurred. Caused by Technician Loosing Footing & Accidently Hitting Outside Cover of Level Switch.Permanent Ladders & Platforms Installed at Head Tank Level Switches 05000322/LER-1988-0171990-05-16016 May 1990 LER 88-017-01:on 881025,discovered That Seismic Monitoring Instrumentation Returned to Svc Prior to Verifying Sys Operability & Special Rept Not Written.Caused by Personnel Error.Surveillance Engineer Reassigned 1997-05-01
[Table view] |
Text
.
g* e-ADDITIC;AL I?!FCPS.ATIC?!
01 JUSTIFICATICI OF V.AEK II LEAD PLA!7f SRV LOAD DEFI?!ITIC thy 15,1979 Stone & k'ebster 1
2268 227 7006010zos
4 I TABI2 0F CC .TP.T3 Section Par.e
- 1. Introducticn 1
- 2. Low Frequency Piping Results 2 2.1 Response to IIFE Request !!o.1 2 2.2 Response to IIRC Request flo. 2 3 3 Cenclusion 6 1
2268 228
- 1. INTRCDUCTIC!:
The report entitled " Justification of thrk II Lead Plant CRV Lcad Definiticn" submitted to the NRC en thrch 30,1979 (EURC-374) demonstrates the ccnservatism present in the Ra=shead Lead Definition. The concerns expressed by the "RC regardinC SRV bubble phasing and frequency character-istics are addressed and furthermore, ccmparicens of the effects on the plant are made with those resulting frca a ccncervatively constructed load definitien tased en the KWU T-Quencher. This T-Quencher is the actual discharge device installcd in Shoreha: and other lead plants.
This report presents additienal informaticn specifically requested in a recent cctrunication (1) with the URC based upon a review of the thrch 30 report.
The two specific NRC requests of Shoreham were the following:
- 1. Shoreham to cctplete Table 2 of the SRV load repcrt by 3
including the same information for the lcw frequency piping systets presented in Table 3
- 2. Sucaittal by Shoreham concerning a discussicn of odal participaticn fer piping analysis. Detailed analysis results are rcquired for the low frequency systems presented in Table 3 concerning hcw much sedal participatien there is for the fundamental codes.
This data is required in order to resolve uncertaintics about 1cw frcquency piping systen responses.
(1) Telephcne ccenunicaticn cn April 27,19?? betueen NRC, CaaE, c&W asa L.
2268 229
_2
- 2. Low Frecuency Pining Results 2.1 Resoonse to fiRC Recuest '!o. 1:
Table 1 presents the support load and pipe stress comparisons at three (3) selected locaticns for the four (4) low frequency subsystems identified in table 3 of the original report. Since it is impractical to present detailed infernation for all piping locations, the follcuing selection criteria, in order of priority, were used:
- 1. Locaticns of high stress.
- 2. Locaticns representing all types of piping cc=ponents.
3 Locatiens evenly distributed along piping subsystem.
It is emphasized that the results at tnese selected locatiens are repre-sentative of the higher stress locations. Results at low stress locaticns 1
are relatively unimportant to plant safety and therefore net presented here (see section 6.3 of the original repert).
2268 230
_3_
2.2 Resnonse to "RC Recuent No. 2:
This section contains a discussien of todal participatien in the dynamic response of piping systems to SRV 1 cads. Particular censideratien is given to systemr which have fundamental frequencies low enoug+ to have one or more mcdes in the frequency range in which the T-Quencher (TQ)
ARS may be greater than the Ratshead (RH) ARS.
The modal response of a culti-degree of freedca systen subject to dynamic support motion (such as a piping system subject to SRV actuation building vibraticns) depends en two basic parameters. The first is the medal partici-paticn facter which is a function of physical characteristics, i.e.,
gecretry and mass distributicn. The second parameter is related to the emplitude (G's) cf the support acceleration at the cedal frequency, i.e.,
the ARS value. Fcr a typical piping systen with cceplex geometry (three g
dimensicnal pipe rcuting, cultiple bends, nuncrcus interier pipe supports unevenly spaced, branch lines, etc. ) and many concentrated tasses (valves, reducers, elbows, tees, equipment, etc.) ' the amplitudes of the redal participation factors are quite varied with significant facters associated with many nodes frem the very lev to the very high frequencies. Large modal responses will cecur at codes with both significant todal participatien factors and significant ARS amplitudes. The tctal response is contributed to by many =cdes of ecmnarable significance ever a wide rance of frequencies.
For the complex piping systens in the Shoreham Nuclear Power Staticn the centri-bution frca the fundamental (lcuest frcquency) =cde is of no special significance.
It may be among the test or least i portant dependirq cn t'.e po rtmeters discussed above. 77/0 77i LJi LL00
Tables 2 through 6 present the mede-by-acde contributions to resultant bending =ctents (acting en the pipe cross secticn) in typical piping systems when subject to SRV building vibratiens. These systems include one ' higher' frequency system (Table 2) for comparisen of behavier with the four ' low' frequency syste=s (Tables 3 through 6). The data presented on these tables is based upon the actual building amplified response spectra (ARS) for both the RH and TQ loads. These building ARS are applied simultaneously in the plant N-S, E-W and vertical directions and vary from elevaticn to elevaricn within the reactor building. Figure 6-1 presented in the original report frc= which the NRC has requested additicnal' information is identified as an " Idealized ARS" and was prescnted in crder to show the majcr features of response. It should not be uscd to make quantitative ec.rrarisons for specific piping systems even though it dces 1
reflect overall trends. Inspecticn cf the tabulatcd results ccafirms the follcuing important points:
- 1. Many modes contribute significantly to the total response regardless of the fundamental frequency, the locaticn cf piping, or the discharge device.
- 2. The centributien frca the fundamental mcde is of no special consequence.
3 For the system with all frequencies above 7 HZ (Table 2) all medal respcnses frca the TQ 1 cad are less than the medal respenses from the RH load.
4 For systems with lower frequencies all modal responses below 7 HZ frca the IQ lead are greater than thcce frc the Rh lead, while all modal respenses above ahcut 15 HZ are higher frca the RH lead. Frc=
7 to 15 HZ results are comparable.
2268 232
5 Systems with the lowest fundamental frequencies have cnly a small percent cf their significant =cdes belov 7 HZ.
- 6. Because many =cdes centribute to the total response, the RH lead provides conservative results even when a few modes occur in the frequency range where T-Q response'is the greater.
2268 233 1
CQ CLUSIrt!
The detailed analysis results presented in this report demenstrate that the
'H load provides conservative results even when the fundamental mode of a piping system is lcv encuch that a few modes occur in the frequency range where the TQ ARS is greater than the RH ARS. The primary reascn is that so many modes centribute to the total system response that the centributicn from those in the lov frequency range is only a small part of the total.
It is therefore concluded that piping systems designed to ramshead load can adcquately accc =cdate the T-Quencher lead. This document further reinforces the justification of using ranshead SRV load definiticn as the 1hrk II lead plant SRV lead definition for plant component design.
The additional information presented in this report is intended to resolve uncertainties regarding the respense of leu frequenc3 piping systems and complete the dccutentation involving the Lead Plant SRV Ramshead Lead Definition as delineated in the 1brch 30,1979 submittal.
2268 234
TATILE 1 TYPICAL REGUI.TL A~ . :.ECTED I C ATIC:0 PAitT 1 JUi PL..T LLAD C"'?An f.1:
CR'. a y PIPI:!G II' N RH+CBE IM 5RV77+CEE GUBSYGD! C Q '.P G ;E'iT (lb or ft-lb) (lb or ft-lb) SR'.TO Centrol Rcd Restraint 65 42 35 1.44 Drive 1265 Resultant Fcree Restraint 85 16 13 1.40 Resultant Force Restraint 110 34 33 1.09 Resultant Fcree Closed Lccp Restraint 95 1 ,81 5 1,801 1.C8 Cooling Water Resultant Fcree 031 Restraint 175 2,041 2,013 1.18 Resultant Fcree Anchor 5 3,1 31 3,052 1.17 Resultant Menent Reactor Eater Ancher 184 649 424 1.91 Cleanup Resultant Mcnent 013 Snubter 186 567 417 1.40 Axial Fcree Snubbcr 623 338 279 1.44
.v91 r rca ,
I'ain Stcan Anenor 340 147,CCO 135,0C0 1.11 25CO Rcsultant :'.cnent Rest:nint 365 33,CCO 3 30,CCO 1.12 Resultant Fcree Enubter 4CG 13,CCO 10,3CO 1 31 y;,1 pm ,n TABLE 1 PART ? FRIt'7Y E0 E.d.. !"T E" C Y C M AR i d:
PIPII'G I;+5RVgg+ CEE : -dR'.t;t 0EE 5Rt. : g (psl)
(pslj3
,m,.,.,-, ,, dy ...'.m^
LU J LI Ui nci v , ss J L . . u. 1,., r . . n ,
i Centrcl Red Valve 1 8,076 7,631 1.C8 Drive 1265 Eltow 50 4,353 4,176 1.07 Run 185 6,65s 6,355 1.17 Closed Lcep Eltcw 10 4,720 4,639 1.07 Cooling Water Tee 205 3,078 2,932 1.16 031 Run 275 5,7C6 4,645 1.38 Reacter '.:ater Yalve 151 8,2C6 7,319 1.50 Cleanup Run 919 7,795 7,280 1.39
- 4r ,
C 1 s,
..t . ; c t r n/
.i c , w, - ,, y , :. < i..
- hin attc- . u '. e ...> d,, > c. , .. :: s. .
I 25c0 Eltcw 415 8,393 E,783 1.27 Run 440 13,161 12,973 1.22 m/0 7'D"
(. O LJ
TALIE 2 FW 301 MODE-EY-: TIE C"-"RIntr K?
REstiLT A!;T ;" "r?;T.,
REDi CFR 65 TEE 67 FlN.'J 8 ' l Rii TQ hii TQ Rii T's FREQUE!CY (ft-kip) (ft-kip) (ft-kip) (ft-kip) (ft-kip) (ft-kip;l MODE (HZ) l l
1 7.1 1.5 0.9 0.6 0.4 0.9 C.5 l 2 90 8.2 8.1 16.6 16.4 12.2 12.0 1.2 1.2 4.4 43 2.9 2.9 3 94 (2) 6.3 45 {
4 11.9 10.3 74 4.5 3.2 ,
-- - 13 7 - - - 4.4--- ----34 -- --
4.0
--- - - -3.1-- - - - 3.2. , - - --
2.c- - -
5 7 19 6 6.9 47 37 25 5.1 34 8 22.7 10.7 7.3 7.5 5.2 2.0 1.4 9 24 4 36 3.2 1.9 1.7 2.3 2.C i 10 26.3 4.7 2.6 4.6 2.6 0.9 05 i 11 26.5 0.8 05 0.7 0.4 0.1 0.1 !
12 28.6 2.5 1.6 8.7 55 4.5 2.8 13 30 3 59 33 6.4 3.6 10.4 59 14 32 3 (3) 2.0 1.4 1.2 0.8 3.2 2.?
15 33.3 7.9 51 13.2 8.6 8.8 5.7 I 16 34 5 6.7 38 0.6 03 9.7 5.4 i 17 37.0 2.9 2.3 1.8 1.4 0.3 0.2 l 18 38 5 2.8 2.2 2.9 23 4.6 3.6 i 19 40.0 4.1 30 4.5 33 57 4.1 I 20 40.0 1.8 1.2 0.6 0.4 0.9 C.6 21 43 5 3.8 2.9 1.0 0.6 1.3 0.o 24 58.8 1.4 0.9 1.8 1.2 0.8 0.e I 25 58.8 0.6 0.4 03 i 0.2 0.2 0.2 (4) j SRSS 31.1 18.7 29.1 22.3 25.2 18.6 i Above 4
(5)
Cenbine l All 35.6 22.3 36.7 28.0 32.2 23.6 Modes I
!!cten:
(1 ) Lev frcquency ncdes telcw 7 He, TQ> RH.
(2) Intermediate frcquenc:< ncdes 'cetuecn 7 to 15 He, T;~ RH.
(3) High frcquency mcdes ateve 15 "c, EH) T;.
(4) For refercnce only, centritutions frcm cther =cdes are lecc significant.
(5) Basis for streno calculaticn, ccnents are ccabincd by Reg. Guide 1.92 Orouping Methcd.
2263 236
TABLE 3 CRD 1265 t'0DE-FY "'ZE Cf ':TR E.UTIC'l RESULTA:.7 :CE:.T5
'. . L'. E 1 E!!'* . . 'O EU" 1 M _
FREQUE::CY ici TL Rii TQ RH K MODE (H2) (ft-lb) (ft-lb) (ft-lb) (ft-lb) (ft-lb) (ft-lb) ,
1 2.9 0.0 1.1 0.0 0.0 0.0 4.5 l 2 51 (1 ) 1.0 9.0 0.0 2.4 0.0 4.5 i 3 _.6. 6_ _ _ . ., _ 2 . 2 _ _ .,__9.0_____ 1.0 2.2 2. 0_ _ _ _ _9 . 9. . !
8.0 2.2 0.0 0.0 3.6 :; .c 4--- 2.4 5 10.0 7.3 15.3 2.0 35 2.2 %9 !
6 11.8 (2) 7.1 19.0 2.0 4.4 1.4 4.1 !
7 12.5 - - . . -- 12.2 .- -
. 23.1----. -
32 4.5 2.0 2.2 I 5 g_
9 16.7 2.0 1.5 1.0 1.1 1.C 1.1 10 19.2 1.4 1.5 0.0 0.0 1.0 1.1 11 25.0 0.0 0.0 0.0 0.0 3.0 1.9 I 12 27.0 11.2 5.6 1.0 1.1 3.2 2.?
13 27.8 3.6 1.5 0.0 0.0 7.0 2.2 14 32.3 (3) 11.9 6.1 0.0 c.0 1.0 1.1 15 37.0 11.2 8.9 1.0 1.1 0.0 0.C i 16 38.5 12.5 9.4 1.0 1.1 1.4 0.C 17 47.6 1.0 0.0 0.0 0.0 7.0 5.5 18 50.0 1.0 1.1 0.0 1 0.0 30 2 . .'
19 50.0 2.2 1.1 1.0 1.1 2.0 2.2 20 58.8 3.3 2.4 1.0 1.1 1.0 0.0 .
21 62.5 20.3 16.8 9.1 7.8 1.0 1.1 22 66.7 4.2 2.2 1.0 1.1 0.0 0.0 .
(4) sRss !
Above 51 .4 46.0 12.7 12.0 2C.9 13 5 .
(5)
Combine All 53.8 50.8 13.6 12.9 22.8 19.6 i Modes 77/0 tLJJ 777 LJJ Notes:
See Table 2 for Fcotnotes.
TABLE 4 CLCU 031 "0EE-EY "0TE CC""RIFJTIC" RESIJLTn!.T l'DE:TL ELEC'.i 10 TEE 205 RU:. 275 i FREQUE'.'OY RH T; RH TQ F3i T:.
MODES (HZ) (ft-lb) (ft-lb) (ft-lb) (ft-lb) (ft-lb) (f t-lb '; }
1 3.6 10. 61. 4. 19. 4. 19.
2 4.3 12. 85 5 32. 7. 45 3 4.9 (1) 11. 113 11. 122. 17. 183 4 55 21 . 288. 6. 83 5 76.
-. -1c5-- 71 0 . 9 55
- ---6.1 g
-- -- 7.--- ---45 - - - - - - - - --- ---
--g---
7 8.9 153 286. 64. 120. 21 . 39 8 98 36. 37 44 46. 46. l.7 .
9 10.0 65 70. 19 20. 28. 29 10 11.2 (2) 11 2. 101. 24. 28. 12. 13 11 11.8 283 350. 16. 20. 22. 23.
12 12.7 Sc. 63 51 . 58. 21. ?5.
13 14.1 301. 274 138. 125 24. 22.
14 14.8- 210.
141. 102. 69. 35 23.
--~~
g5 ~ 17
~ 75 ' ' ~~7.657~ ~ ~ ~256.-- ~ ~ ~ 9 .- - 6. 3 2.
16 18.2 245 180. 2. 2
- 2. 5 3 17 19.2 6. 4 1041. 711. 1220. 267.
292. l 18 24.0 (3) 2. 1. 740. 507. 425 -
22 30.3 4 3 6C8. 446. 2698. 1977.
24 32.7 1. 1. 17. 13 60. !.S .
(4)
SRSS l' Above 736. 734 1433 1233 3018. 2190.
l (5) i Combine I All 1C04 935 1463. 1262. 3C55 2220.
Medes Notec:
See Table 2 for Footnotes.
')}{}}
. TABI.E 5 RKCU 013 :/00E-EY L'rJ.r, Cel."RIEUTIC::
RE.SULT A :T :" :'1:;T3 RU:: 919 'EE,CCih f:'-
!AI/.E 151 1 RH TQ i RH T; RH T-FREQUE"CY MODES (HZ) (ft-lb) (ft-lb) (ft-lb) (ft-lb) (ft-lb) (ft-lb) l 1 4.9 0. 2. 6. 83 6. 83
- 0. O. O. O. O.
2 6.5 O.
3 6.6 (1 ) C. 1. 10. 43 10. 44.
6.8 0. O. O. O. O. O.
4 5-~~ 7.7 1. 1. ~ - 63 58. 17. 17. ~R
- 6. 11. O. O. O. O. I 6 7.9 8.0 0. O. O. O. O. O.
7 8 9.0 4 4 44. 53 28. 34.
0 11.1 1. 2. O. O. O. O. l 10 11.4 (2) 3 3 O. O. O. C. l 11.8 76. 147. O. 1. O. O. l 11
- 17. C. O. O. O.
12 12.5 14 l 12.8 19. 23 . O. O. O. O. I 13
- 8. 22. 21. 13. 11.
_15 . _14.1 _ ._ _ 8.__ ~~--~ ~~~ ~~
g-- _ .
, _ _ _ _ 507 - ~ ~ jI5 ~~~ 13 13 ~ ~ ~4 .~' i 18 16.1 138. 61. 79 34. 24. 11. ;
20 16.8 8. 3 66. , 26. 30. 12.
19.4 138. 7/. 15 8. 3 1.
23 25 21.6 381. 19C. 23 11. 2. 1.
27 22.9 71. 44. 1C8. 66. 4. 2. i 28 23.0 100. 111. 73 51. 20. 14. ,
29 24.1 235 169 254 184 15 11.
30 25.0 37. 25 29 20. 1CS. 73 25.0 (3) 61 . 42. 24 15 63 43 31 32 26.6 110. 8/. . 188. 143 8. 7. !
35 28.6 291. 198. 18. 12. 2. 1. i' 37 31.3 6. 3 42. 28. 11. 7.
39 32.3 10. 4 22. 14. 12. 7.
42 37.0 78. 57. 45 33 21. 15 44 38.5 19 14. 6. 4 29 24 47 43.5 68. 60. 5 4. 2. 1. ;
48 43.5 174. 125 21 . 15 5 3 I 51 47.6 78. 65 3 2. 2. 1. l 57 55.6 2. 1. 7. 4 55 37. i l
(4 SRSS Above 664 437. 378. 286. 155 143 (5)
Combine 938. 61 5 466. 331.. 213 1 '/3 I
All Modes n- eo O7O l E ' d '##
I;ot en : See Table 2 for Footnctes.
. TABLE 6 MS 25CO l'CDE-EY-Mai>E CU.TRIELTIO:
RESULTA?:T " :i :T.:,
VA L'. E J /. 5 E'? "f. 41 5 P.C: uC F?@ UE';CT Rh 4 RH T 's Rh T ';.
i MODE (liZ) (ft-kip) (ft-kip) (ft-kip) (ft-kip) (ft-kip) (ft-cip; 1 _5._1 _( 1_) _ 0.2 0.8 1.1 4.1 2.8 1C.9 i 2 13.5 4.6 53 3.8 4.3 3.3 3.5 !
3 14.5 (2) 3.5 6.0 0.6 1.0 0.3 C.5 4 15.2 12.4 20.0 0.9 1.5- - - -- 0. 6--- - - -
0.o -
3___ _176.7 4.5 3.3 18.4 13.4 14.9 10.9 ;
6 17.9 39.3 23.7 03 0.2 0.2 0.1 7 20.8 2.1 1.2 9.2 52 93 53 !
8 23.8 8.2 5.2 90 5.7 10.2 6.5 i 9 26.6 36.5 17.7 2. 7 1.3 3.6 1.7 1 10 29.4 22.2 31.4 1.3 1.8 0.4 0.6 11 30 3 (3) 1.9 2.1 1.2 1.4 1.6 1.8 l 12 35.7 3.7 2.7 22.0 16.2 15.3 11.3 13 37.0 6.8 5.4 20.3 16.0 14.3 11.3 i 14 43 4 0.9 0.6 2.9 2.3 27.7 21.4 ,
15 43.5 0.1 0.1 1.3 1.1 44.2 37.7 !
16 43.5 0.8 0.6 3.1 i 2.5 27 3 22.3 17 45.5 0.8 0.6 6.4 5.5 2.6 2.2 18 52.6 2.5 2.2 15.0 12.8 2.1 1.E 19 55 5 23 2.1 15.1 12.3 3.8 3.1 i SRS Above 61.1 49.2 44.0 34.1 66.2 54.6 (5) ,
Combine l All 63.5 51 .8 48.4 38.0 105.7 86.7 !
Modes i Notes: oqrn 7An
(.LU0 LtU See Table 2 for Fcotnotes.}}