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Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML17265A1361997-12-23023 December 1997 Rev 2 to RCS Pressure & Temp Limits Rept (Ptlr). ML17264B1011997-08-29029 August 1997 Rev 0 to Leak-Before-Break Evaluation of Portions of RHR Sys at Re Ginna Nuclear Power Station. ML17264A8681997-04-23023 April 1997 Rev 0 to Evaluation of Ginna RCS Coolant Temp to Support LTOPs Requirements. ML17264A8511997-03-19019 March 1997 Rg&E Re Ginna Nuclear Power Plant Spent Fuel Pool Re-racking Licensing Rept. ML17264A7931997-01-31031 January 1997 Rev 1 to Final Rept, Re Ginna Nuclear Power Plant Probabilistic Safety Assessment. ML17264A6101996-09-23023 September 1996 Rev 0 to Design Analysis Operability Evaluation for 857 A/B/C Ginna Station. ML17264A6121996-09-23023 September 1996 Rev 2 to Design Analysis Ginna Station Pressure Locking Evaluation for MOVs 852 A&B. ML17309A6051996-09-13013 September 1996 Rev 2 to RCS Pressure & Temp Limits Rept (Ptlr). ML17264A6791996-05-24024 May 1996 Rev 1 to RCS Pressure & Temperature Limits Rept (Ptlr). ML17264A4001996-02-24024 February 1996 Rev 0 to RCS Pressure & Temp Limits Rept. ML17264A2791995-12-0808 December 1995 Re Ginna NPP RCS Pressure & Temp Limits Rept Cycle 25, Draft B ML17264A1051995-05-0404 May 1995 Rev 0 to Final Exam Rept for 1995 SG Eddy Current Insp at Ginna Nuclear Power Station, Dtd 950503 ML17263B0391995-04-18018 April 1995 Summary Exam Rept for 1995 SG Eddy Current Insp,Rev 0. ML17264A3411995-03-15015 March 1995 Low Temp Overpressure Analysis Summary Rept. ML17263A8351994-11-0707 November 1994 Rev 1 to Fission Product Barrier Evaluation. ML17263A8331994-10-11011 October 1994 Rev 1 to Re Ginna EALs Technical Bases. ML17263A8311994-09-26026 September 1994 Draft Rev C to Design Criteria Ginna Station Containment Structural Mods Wbs 4. ML17263A7941994-09-15015 September 1994 Safety Evaluation of Ginna SG Replacement. ML17059A3611994-07-0606 July 1994 Emergency Action Level Verification & Validation Rept. ML17263B0481994-06-30030 June 1994 Criticality Analysis of Plant Fresh & Spent Fuel Racks & Consolidated Rod Storage Canisters. ML17311A0181994-05-13013 May 1994 New York State EAL Upgrade Project Verification & Validation Rept. ML17263A8291994-03-30030 March 1994 Draft Rev a to Safety Evaluation SEV-1019, Containment Structural Mods Wbs 4. ML17263A4651993-05-17017 May 1993 Radial Displacement & Rebar Strain Measurements for EWR #5181,Rev A. ML17262B1201992-11-30030 November 1992 Re Ginna Boric Acid Storage Tank Boron Concentration Reduction Study. ML17262B0831992-07-31031 July 1992 Recommended Info for Inclusion in Section 15.6.4 of FSAR for Re Ginna Nuclear Plant. ML17262A8391992-04-30030 April 1992 Rev 0 to Summary Exam Rept for 1992 SG Eddy Current Insp at Re Ginna Nuclear Power Station. ML17262A5601991-06-18018 June 1991 Rev 1 to Design Verification Ginna Station Containment Foundation Mat Analysis. ML17262A4691991-04-25025 April 1991 Rev 0 Summary Exam Rept for 1991 Steam Generator Eddy Current Insp. ML17262A4521991-04-22022 April 1991 Control Room Heatup Analysis. ML17262A3781991-02-28028 February 1991 Nonproprietary Re Ginna Low Temp Overpressure Protection Sys Setpoint Phase II Evaluation, Final Rept ML17262A4141991-02-26026 February 1991 Safety Analysis,Ginna Station Updated FSAR Section 6.2.4 & Tables 6.2-13,6.2-14 & 6.2-15 Changes. ML17262A3681991-02-15015 February 1991 Simulation Facility Certification Rept. ML17262A4431990-10-0404 October 1990 Rev 0 to Design Analysis Ginna Station Containment Mat Design Water Level Elevation 265 ft,0 Inches. ML17262A4401990-10-0404 October 1990 Rev 0 to Design Verification Ginna Station Containment Foundation Mat Analysis. ML17262A1931990-10-0303 October 1990 Rev 1 to Safety Analysis Ginna Station Updated FSAR Table 6.2-13 Changes. ML17262A1761990-08-30030 August 1990 Voltage Simulation for Case EOF LOC4 LOCA Simulation for 50/50 Mode - Circuit 767 Details 12B Transformer Feeding Bus 12B. ML17262A1771990-07-27027 July 1990 Rev 1 to Design Analysis EWR 4525-1, Fault Current Analysis of Power Distribution Sys. ML17262A1781990-07-24024 July 1990 Rev 1 to Design Analysis EWR 4525-2, Adequacy of Electric Sys Voltages. ML17250B1761990-05-0808 May 1990 Rev 1 Summary Exam Rept for 1990 Steam Generator Eddy Current Insp. ML17261B0201990-03-14014 March 1990 Design Criteria Ginna Station Steam Generator Containment Penetration. ML17251A4811989-02-28028 February 1989 Ultrasonic Indication Sizing Technique Development. Related Info Encl ML17251A4771988-06-17017 June 1988 Rev 0 to Differential Pressure Thrust Calculation Methodology. ML17261A6571987-10-31031 October 1987 Steam Generator Tube Plugging Increase Licensing Rept for Ginna Nuclear Power Station. ML17261A5521987-07-14014 July 1987 Supplemental Rept to Dcrdr Final Summary Rept for Re Ginna Station. ML17251A4741987-04-0101 April 1987 Rev 0 to Safety Analysis,Ginna Station PORV Block Valves. ML17251A4721987-03-10010 March 1987 Rev 0 to Design Criteria,Ginna Station PORV Block Valves Replacement. ML17251A9191986-12-18018 December 1986 Rev 0 to Implementation Rept EWR 2799, Reactor Vessel Level Monitoring Sys. ML17251A6171986-03-0101 March 1986 1986 Steam Generator Eddy Current Exam Summary Rept. ML17254A7031985-12-31031 December 1985 Vols 1 & 2 to Dcrdr Final Summary Rept Program Implementation,Re Ginna Nuclear Power Plant. ML17254A6911985-12-16016 December 1985 Reinforced Masonry Wall Evaluation,Evaluation of Control Bldg Reinforced Walls. 1997-08-29
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML17265A7601999-10-0505 October 1999 Part 21 Rept Re W2 Switch Supplied by W Drawn from Stock, Did Not Operate Properly After Being Installed on 990409. Switch Returned to W on 990514 for Evaluation & Root Cause Analysis ML17265A7621999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Re Ginna Npp.With 991008 Ltr ML17265A7531999-09-23023 September 1999 Part 21 Rept Re Corrective Action & Closeout of 10CFR21 Rept of Noncompliance Re Unacceptable Part for 30-4 Connector. Unacceptable Parts Removed from Stock & Scrapped ML17265A7541999-09-22022 September 1999 LER 99-011-00:on 990823,small Tears Were Discovered in Flexible Duct Work Connector at Inlet of CR HVAC Sys Return Air Fan (AKF08).Caused by in-leakage Greater than That Assumed.Implemented Temporary Mod 99-029.With 990922 Ltr ML17265A7471999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Re Ginna Npp.With 990909 Ltr ML17265A7431999-08-24024 August 1999 LER 99-004-01:on 990412,discovered That Containment Recirculation Fan Chevron Separator Vanes Were Installed Backwards.Caused by Improper Assembly by Mfg.Moisture Separator Vanes Were Dismantled & Correctly re-installed ML17265A7341999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Re Ginna Npp.With 990806 Ltr ML17265A7291999-07-29029 July 1999 Interim Part 21 Rept Re safety-related DB-25 Breaker Mechanism Procured from W Did Not Pas Degradatin Checks When Drawn from Stock to Be Installed Into BUS15/03A.Holes Did Not line-up & Tripper Pan Bent ML17265A7181999-07-23023 July 1999 LER 99-007-01:on 990423,reactor Trip Occurred Due to Instrument & Control Technicians Inadvertently Pulling Fuses from Wrong Nuclear Instrument Channel.Setpoint Adjustments Were Completed by Different Crew of Technicians ML17265A7081999-07-22022 July 1999 LER 98-003-02:on 980904,actuations of CR Emergency Air Treatment Sys Was Noted Due to Invalid Causes.Caused by Various Degraded Components in CR RM Sys.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored ML17265A7131999-07-22022 July 1999 Special Rept:On 990407,radiation Monitor RM-14A Was Declared Inoperable.Caused by Failed Communication Link from TSC to Plant Process Computer Sys.Communication Link Was re-established & RM-14A Was Declaed Operable on 990521 ML17265A7031999-07-19019 July 1999 LER 99-S01-00:on 990617,determined That Temporary Unescorted Access Had Been Granted to Contractor Employee.Caused by Incomplete Info Re Circumstances of Individual Military Separation.Individual Access Was Revoked.With 990719 Ltr ML17265A7211999-07-19019 July 1999 ISI Rept for Third Interval (1990-1999) Third Period, Second Outage (1999) at Re Ginna Npp. ML17265A7021999-07-15015 July 1999 LER 99-010-00:on 990615,ventilation Isolation of Auxiliary Bldg Occurred When Auxiliary Bldg Gas Radiation Monitor R-14 Reached High Alarm Setpoint.Cr Operators Rest Auxiliary Bldg Ventilation Isolation Signal.With 990715 Ltr ML17265A7661999-06-30030 June 1999 1999 Rept of Facility Changes,Tests & Experiments Conducted Without Prior NRC Approval for Jan 1998 Through June 1999, Per 10CFR50.59.With 991020 Ltr ML17265A7011999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Re Ginna Npp.With 990712 Ltr ML17265A6851999-06-21021 June 1999 LER 99-001-01:on 990222,deficiencies in NSSS Vendor steam- Line Brake Mass & Energy Release Analysis Results in Plant Being Outside Design Bases Occurred.Caused by Deficiencies in W.Temporary Administrative Replaced.With 990621 Ltr ML17265A6761999-06-16016 June 1999 Part 21 Rept Re Defects & noncompliances,10CFR21(d)(3)(ii), Which Requires Written Notification to NRC on Identification of Defect or Failure to Comply. Relays Were Returned to Eaton for Evaluation & Root Cause Analysis ML17265A6661999-06-0202 June 1999 LER 99-009-00:on 990503,instrumentation Declared Inoperable in Multiple Channels Resulted in Condition Prohibited by Ts. Caused by Unanticipated High Frequency AC Voltage Ripple. Entered TS LCO 3.0.3.With 990602 Ltr ML17265A6681999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Re Ginna Nuclear Power Plant.With 990608 Ltr ML17265A6651999-05-27027 May 1999 Interim Rept Re W2 Control Switch,Procured from W,Did Not Operate Satisfactorily When Drawn from Stock to Be Installed in Main Control Board for 1C2 Safety Injection Pump. Estimated That Evaluation Will Be Completed by 991001 ML17309A6541999-05-27027 May 1999 LER 99-008-00:on 990427,overtemperature Delta T Reactor Trip Occurred Due to Faulted Bistable During Calibr of Redundant Channel.Plant Was Stabilized in Mode 3 & Faulted Bistable Was Subsequently Replaced.With 990527 Ltr ML17265A6631999-05-24024 May 1999 LER 99-007-00:on 990423,technicians Inadvertently Pulled Fuses from Wrong Nuclear Instrument Cahnnel,Causing Reactor Trip,Due to High Range Flux Trip.Caused by Personnel Error. Labeling Scheme Improved ML17265A6601999-05-21021 May 1999 LER 99-006-00:on 990421,start of turbine-driven Auxiliary Feedwater Pump Was Noted.Caused by MOV Being Left in Open Position.Closed Manual Isolation Valve to Secure Steam to Pump.With 990521 Ltr ML17265A6591999-05-17017 May 1999 Part 21 Rept Re Relay Deficiency Detected During pre-installation Testing.Caused by Incorrectly Wired Relay Coil.Relays Were Returned to Eaton Corp for Investigation. Relays Were Repaired & Retested ML17265A6441999-05-13013 May 1999 LER 99-005-00:on 990413,undervoltage Signal of Safeguards Bus During Testing Resulted in Automatic Start of B Edg. Caused by Personnel Error.Blown Fuse Was Replaced & Offsite Power Was Restored to Safeguards Bus 17.With 990513 Ltr ML17265A6431999-05-12012 May 1999 LER 99-004-00:on 990412,discovered That Containment Recirculation Fan Moisture Separator Vanes Were Incorrectly Installed,Per 10CFR21.Caused by Improper Assembly by Mfg. Subject Vanes Were Dismantled & Correctly re-installed ML17265A6381999-05-0707 May 1999 Part 21 Rept Re Replacement Turbocharger Exhaust Turbine Side Drain Port Not Functioning as Design Intended.Caused by Manufacturing Deficiency.Turbocharger Was Reaasembled & Reinstalled on B EDG ML17265A6391999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Re Ginna Nuclear Power Plant.With 990510 Ltr ML17265A6361999-04-23023 April 1999 Part 21 Rept Re Power Supply That Did Not Work Properly When Drawn from Stock & Installed in -25 Vdc Slot.Power Supply Will Be Sent to Vendor to Perform Failure Mode Assessment.Evaluation Will Be Completed by 991001 ML17265A6301999-04-18018 April 1999 Rev 1 to Cycle 28 COLR for Re Ginna Npp. ML17265A6251999-04-15015 April 1999 Special Rept:On 990309,halon Systems Were Removed from Svc & Fire Door F502 Was Blocked Open.Caused by Mods Being Made to CR Emergency Air Treatment Sys.Continuous Fire Watch Was Established with Backup Fire Suppression Equipment ML17265A6551999-04-0909 April 1999 Initial Part 21 Rept Re Mfg Deficiency in Replacement Turbocharger for B EDG Supplied by Coltec Industries. Deficiency Consisted of Missing Drain Port in Intermediate Casing.Required Oil Drain Port Machined Open ML17265A6291999-03-31031 March 1999 Rev 0 to Cycle 28 COLR for Re Ginna Npp. ML17265A6241999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Ginna Station.With 990409 Ltr ML17265A6141999-03-31031 March 1999 LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr ML17265A6131999-03-29029 March 1999 LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr ML17265A6061999-03-24024 March 1999 LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp ML17265A5661999-03-0101 March 1999 Rev 26 to QA Program for Station Operation. ML17265A5961999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Ginna Nuclear Power Plant.With 990310 Ltr ML17265A5371999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for Re Ginna Nuclear Power Plant.With 990205 Ltr ML17265A5951998-12-31031 December 1998 Rg&E 1998 Annual Rept. ML17265A5001998-12-21021 December 1998 Rev 26 to QA Program for Station Operation. ML17265A4951998-12-21021 December 1998 LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr ML17265A4931998-12-17017 December 1998 LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery ML17265A4761998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Re Ginna Nuclear Power Plant.With 981210 Ltr ML17265A4691998-11-25025 November 1998 LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR ML17265A4531998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Re Ginna Nuclear Power Plant.With 981110 Ltr ML17265A4271998-10-0505 October 1998 LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr ML17265A4291998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Re Ginna Nuclear Power Plant.With 981009 Ltr 1999-09-30
[Table view] |
Text
TELEPHONE (412) 961-0323 U. S. U. S. TOQL R QIE, INC.
An NPS Corp. Subeldiery 4030 ROUTE 8 ~ ALLISON PARK, PENNSYLVANIA 15101 K)RNADO MISSILE ACCIDENT ANALYSIS SPENT FUEL STORAGE RACKS ROCHESTER GAS AND ELECTRIC 8367-00-0005 OCTOBER 4, 1983 PREPARED BY Harold H. Waite KEENED BY Will '.
7 c .
/
If
'ATF/8 Wachter APPROVED BY
~r.~~
Engineering Manager c 8D Frank E. Witsch e40iVSOSOS BOOS>8 QGa13.~ AssQx'ance Manager PDR ADOCK 05000244 P FOR
PAGE 1 OF 16 STATEMENT OF PROBLEM:
DETERMINE THE EFFECT ON THE SPENT FUEL RACK CAUSED BY THE IMPACT OF A MISSILE, AS DEFINED BY ROCHESTER GAS 6 ELECTRIC, BEING A POLE 12.5 INCHES IN DIAMETER, 35 FEET LONG AND WEIGHING 1490 POUNDS. THIS CORRESPONDES TO THE MISSILE DEFINED IN REFERENCE ¹4 AS A 13.5 INCH DIAMETER, 35 FOOT LONG UTILITY POLE, WEIGHING 1490 POUNDS.
CONCLUSION:
THIS ANALYSIS DEMONSTRATES THAT THE IMPACT OF THE MISSILE WILL NOT IMPAIR THE INTEGRITY OF THE RACK TO MAINTAIN ITS SUBCRITICAL GEOMETRIC ARRAY.
INTRODUCTION:
A 35 FOOT LONG POLE, 13.5 INCHES IN DIAMETER AND WEIGHING 1490 POUNDS COULD BE A SUGAR MAPLE WOOD HAVING A DENSITY OF 43 LB/SQ.FT. THE POLE WOULD IMPACT THE RACK WITH THE FORCE ON THE POLE PARAI,LEL TO THE GRAIN.
PROPERTIES OF SUGAR MAPLE -12 4/i MOISTURE CONTENT- FROM REF ¹6.
MODULUS OF ELASTICITY 1.83E06 PSI MAXIMUM CRUSHING STR. PARALLEL TO GRAIN = 7830 PSI MAXIMUM SHEARING STR. PARALI.EL TO GRAIN = 2330 PSI LOAD REQUIRED TO IMBED A 0.404 IN. DIAM. BALL TO 1/2 ITS DIAM. 1450 LB.
WHEN THE POLE'MPACTS THE RACK CONSIDERABLE ENERGY Wl I L BE ABSORBED BY THE POI.E. ANYONE WHO HAS USED WOODEN MALLETS OR DRIVEN ON h PIECE OF WOOD HAS OBSERVED HOW QUICKLY THE EDGES AROUND THE ENDS WILL SPLIT AWAY AND HOW THE ENDS BECOME VERY FIBROUS.
IN THIS REPORT THE ENERGY ABSORBED BY THE POLE IS NEGLECTED' FULL SCALE TEST WOULD HAVE TO BE DONE IN ORDER TO DETERMINE WHAT ACTULLY HAPPENS TO THE RACK. A 3-D FINITE ELEMENT ELASTIC-PLASTIC ANALYSIS WOULD GIVE EXCEI.LENT INSIGHT INTO HOW THE FORCES ARE DISTRIBUTED THROUGHOUT THE RACK. HOWEVER THIS TYPE OF ANALYSIS IS LABORIOUS AND EXPENSIVE.
WACHTER ASSOCIATES PERFORMED TEST DEMONSTRATING THAT THEIR RACKS..
"'"" WOULD'ITHSTA'ND A 9000'FT-LBMISSILE 'LOAD (REF. ¹2);. THESE TESTS WERE DONE ON A SINGLE BOX WHICH WAS ADEQUATE FOR THE 9000 FT-LB L'OAD. THE IMPACT FORCE IMPOSED BY THE MISSILE UNDER INVESTIGATION IN THIS ANALYSIS IS SEVERAL TIMES GREATER TROJAN THE "WACHTER" MISSILE. THE COMPOSITE RACK AS A HONEYCOMB STRUCTURE MUST BE RECOGNIZED TO DEMONSTRATE THE ADEQUACY OF THE RACK.
PAGE 2 OF 16 VERTICAL IMPACT:
TYPICAL RACK, SHOWN ON FIG. ¹1, IS MADE UP OF 140 BOXES, 8.43 INCHES SQUARE WITH A 0.090 INCH THICK WALL. THE BOXES ARE WELDED TOGETHER TO FORM A HONEYCOMB TYPE STRUCTURE. THE INTERNAL BOXES ARE ATTACHED TO EACH OTHER BY WELDS. THE GENERAI ARRANGEMENT OF THE WELDS IS SHOWN ON FIG ¹8. THE TOTAL WELD SHEAR AREA ON ONE BOX IS.'2 EA. 1/2" DIAM. FUSION WELDS ~ 2. 36 SQ. IN.
20 EA. 2" LONG FILLET WELDS ' 7.20 SQ.IN.
(Assume fiiiet welds 0.18 " thk) h EA. 1" FILLET WELDS 0.72 SQ.IN.
TOTAL WELD AREA 10.28 SQ ~ IN.
ANY VERTICAL LOADS APPLIED TO A BOX WILL BE TRANSMITTED TO THE OTHER BOXES VIA THE WELDS. A FORCE APPLIED TO A SINGLE BOX FROM ANY DIRECTION WILL BE REACTED BY THE COMPOSITE HONEYCOMB STRUCTURE.
THE SECTION OF THE RACK ENCOMPASSED BY THE IMPACT LOAD WIIL ACT AS COLUMNS UNDER COMPRESSIVE LOADING UNABLE TO BUCKLE.
THE ELASTIC STRAIN ENERGY WILL BE:
P L/2AE (REF ¹5)
'HERE: P ~ IMPACT LOAD L ~ COLUMN LENGTH A ~ CROSS SECTIONAL AREA RESISTING THE LOAD.
E ~ YOUNG'S MODULUS THE EVALUATION OF THE EFFECT OF THE MISSILE IMPACT ON A 'RACK
~ '-'IS'ONE-USING THE STRAIN 'ENERGY METHOD DESCRIBED 'IN REFERENCE., ~
¹3, SECTION 2.8-0. AND REF ¹5 CHAPTER XI.
THE RACKS EXHIBIT ENERGY RESISTANCE:
Uu ( in-Ibs/ in. )
THIS IS FOUND US ING THE AREA UNDER THE STRESS STRAIN CURVE (STRAIN ENERGY) SHOWN IN FIG. ¹5.
PAGE 3 OF 16 Uu = (Sy + Su)Ka/2 AND U = Uu z A x L WHERE: Sy = YIELD STRENGTH Su = ULTIMATE STRENGTH Ka = ULTIMATE UNIT ELONGATION U = ENERGY IMPACTED TO THE RACK A = CROSS SECTIONAL AREA ABSORBING "U" L = LENGTH OVER WHICH "U" WILL BE ABSORBED.
USING Sy ~ 30,000 psi and Su = 70,000 psi Uu ~ 50,000 Ka ENERGY IN Ml SS I LE:
THE HORIZONTAL TORNADO VELOCITY IS GIVEN AS 132 MPH.
(132 MI/HR)<5280 FT/MI)<HR/3600 SEC) 194 FT/SEC HORIZONTAL VELOCITY ~ 0. 4 < 194 ) = 78 FT/SEC VERTICAL VELOCITY ~ 0.8<78) = 62 FT/SEC.
WHEN THE MISSILE ENTERS THE POOL THE KINETIC ENERGY IS REDUCED AS IT MOVES THROUGH THE WATER. THE KINETIC ENERGY AT POINT OF CONTACT IS FOUND FROM REF 01.
Z ~ CdAcs/W WHERE: Cd = DRAG COEFFICIENT = 1.0 Acs ~ CROSS SECTIONAL AREA ENTERING POOL W = WEIGHT OF MISSILE (LBS>
D = 12.5 in.
A = 0.852 SQ.IN.
Z ~ 0.852/1090 = 0.0006 FROM CURVES FOR 20 FT DEPTHS AND .34,.FT DEPTH KE/W EQUALS 55 AND '00 'RESPECTIVELY'.NTERPOL'ATING FOR 25 FEET, DEPTH UNDER WATER OF RACK, KE/W = 53 FT-LB/LB.
KE s <53) 1490>
< = 79,000 FT-LB.
PAGE 4 OF 16 THE VALUES OF A (AREA) AND L (LENGTH) TO BE USED IN THE EQUATION ARE ARBITRARY. OBSERVING FIGURES ¹2 AND ¹3 THE AREA UNDER THE MISSILE A SHORT DISTANCE DOWN FROM THE POINT OF IMPACT IS RESISTED BY A MINIMUM OF 24 BOX SIOES.
A = (24) (8. 43 + 8.25) <0. 09) /2 "- 18.0 SQ. IN THE LENGTH OVER WHICH THE ENERGY IS ABSORBED CAN VARY UP TO THE TOTAL I ENGTH OF 160 INCHES. AS THE LOAD MOVES DOWN THE BOX IT WI LL SPREAD OUT AT APPROXIMATLEY A 45 DEGREE ANGLE.
THE TEST BOX IN REFERENCE ¹ 2 WAS 15 INCHES LONG. IT IS REASONABLE TO ASSUME 30 INCHES AS A CONSERVATIVE LENGTH.
A x L = 540 CU. IN.
U = Uu x A x L 540 Uu 79,000< 12) /540 ~ 1755 IN-LB/CU. IN.
Ka. = 1755/50,000 = 0.035 IN/IN TOTAL DEFORMATION = 0.035(30) = 1.05 IN.
IMPACT AT PERIPHERY OF RACK:
A VERTICAL IMPACT LOADING SHOWN ON FIGURE ¹4 IS ASSUMED.
18 BOX SIDES WOULD ACT TO RESIST THE LOAD.
A = 18 ( 18/24) ~ 13. 5 SQ. IN.
U = (13.5)(30)Uu = 405 Uu Uu = 79,000(12)/405 ~ 2340 IN-LB/CU.IN.
Ka = 2340/50,000 = 0.007 IN/IN.
TOTAL DEFORMATION = (0.047)(30) = 1.40 IN.
\
PAGE 5 OF 16 HORIZONTAL IMPACT A MISSILE IMPACTING THE. RACK FROM THE SIDE WILL BE REACTED BY THE SIDES PERPENDICULAR TO THE fACE OF THE MISSILE,. THE HORIZONTAL IMPACT LOAD WIIL ACT ON THE RACK AS SHOWN GN F I GURES 06 AND 07 .
THE RACK IS FREE STANDING AS SHOWN IN FIGURE 06. HOWEVER THE SEISMIC RESTRAINTS LOCATED AT THE BASE OF THE RACK WILL TEND
'O MAKE IT BEHAVE AS A CANTILEVER BEAM WHEN IMPACTED BY A M I SS I LE AT THE TOP .
THE INITIAL VELOCITY OF 78 FPS WILL BE REDUCED AS IT TRAVELS THROUGH THE WATER. IT IS ASSUMED THAT THE VELOCITY WILL REDUCE BY AT LEAST 10 '%R TO 70 FPS.
AS THE MISSILE IMPACTS THE TOP OF THE RACK THE KINETIC ENERGY IS REDUCED TO:
2 KE = WbV (C1/(1+(We/Wb) I) /(2g)
WHERE:
Wb = WE I GHT OF MI SS I LE Wm = WEIGHT OF RACK ASSEMBLEY We = EQUI VAI,ENT WEIGHT OF THE MEMBER
- 0. 236 Wm V = INITI AL VE LOG ITY THE WEIGHT OF THE RACK, Wm, IS USED AS THE WEIGHT OF THE EMPTY RACK PLUS THE ENTRAINED WATER.
W< ent) = (8. 25) 2 (160) (62. 4) (140) /1728 55,000 lb.
W( Rack) 20,000 lb.
W(Total) 75,000 lb.
We = 0.236 x 75,000 = 17,700 lb.
KE = (1490) (7) 2 (1/C1+(17,700/1490) i) /(2g) 8800 FT-LB THE ALLOWABLE ENERGY LOAD, OR I OAD THAT CAN BE ABSORBED ELASTICALLY IS:
U = <0. 16667Sy L/E) ( I/c )
WHERE: I = MOMENT OFINERTIA OF'ACK' c = DISTANCE TO NEUTRAL AXIS I/c = IMPACT LOAD STRENGTH
PAGE 6 OF IN THE NORTH-SOUTH DIRECTION 2 = (I I/c = 489,900/(59) 140.8 SQ.IN. FROM P.7)
IN THE EAST-WEST DIRECTION I/c " = 226,720/(42.15) ~ 127.6 SQ.IN.
Umin = (0.1667)(30,000) <160)(127.9)/(28E06ai2) c.'1 00 FT-LB THIS IS LESS THAN THE KINETIC ENERGY FROM THE MISSILE IMPACT.
THE HORIZONTAL LOAD IS CARRIED THROUGH THE RACK BY .THE SIDES OF THE BOXES AS SHOWN ON FIG. ¹7. THE SECTION BETWEEN THE BOXES MUST MAINTAIN ITS STRUCTURAL INTEGRITY TO CARRY THE LOAD. THESE SECTIONS BEHAVE AS A COLUMNS WHERE:
I ~ BH /12 (8.25)<2)(0.09) /12
- 0. 124 IN.
r ~ I /A = 0. 124/0. 74
- 0. 408 8 "-(T E/ ( L / R), ( EULER ' FORMULA) 1 . 7E09 PS I THE SECTIONS BETWEEN THE BOXES WILL CARRY THE LOAD FROM ONE BOX TO ANOTHER WITHOUT BUCKI,ING.
PAGE 7 OF IG MOMENT OF INERTIA OF THE RACK
-lg d
-I I d 9d 75d a
(I 3d d
)8.25~
Ibos = (bo4 - b, ) /12, 34.81 in. 4 E4.3 Ahoy ~ <bp 2:2 bI )
t
=3.00 in.2 I -"51's +/Ad 2 Na = NUMBER OF BOXES PARALLEL TO THE AXIS ABOUT WHICH THE MOMENT OF INERTIA IS CALCULATED.
Nb = 1/2 OF THE NUMBER OF BOXES PERPENDICUI.AR TO THE AXIS ABOUT WHICH THE MOMENT OF INERTIA . IS CALCULATED.
i=<2Nb-1)(in increments of 2)
ZNat lib a boa + Ad (~(i~ )g i=1 FOR NORTH-SOUTH SEISMIC LOADING Na ~ 10, Nb = 7, In-s = 489,900 in.4 An-s = a<20) (0;.0.90) (118'.'0'2) ~
"2 212.'4 .i.n..
FOR EAST-WEST SEISMIC LOADING Na = 14, Nb = 5, I e-w = 226, 720 in.A Ae-w = (28) <0.090) (84.3) ~ 212. 0 in.2
ZOCOO LBS Ml SSI L E IMPACT ARE A INSIDE PERIPHERY QF RAC K SID E'S RES IS Tl NG IMPAC T'OAD
AG E I 0 OF l6 Fi GUPE NO 3 M I SSI LE IMPAC T A RE A INS ID E PERIPHERY OF' pgIr, SI DE S R ESI S T I NG IMPACT LOAD
AGE I I OF IG Ft GURE NO 4 MISSILE IMPACT AWE A AT P EP.IP HEPY OF RAC 8
~l-l ]
SIDE S RES IST! NG IMP AC T L 0 AD I NG
PAGE I2 OF I6 Fl Gu 8 t:- NC.5 UH I T STRESS Su Sy UNI T ST R A IN
PAGC I 3 OF IG F)GuRv NO.6 HOAtZONTA L IMPACT FC ACE
L=lG0 R E No-7 HGR IZO N T A L IMPAC T FORCE BC X
~ ~
4 P E i5 OF ij-f= iC,uRc ~O. e WF LOS I iNCH F ILLE T V/ELDS Ig DIAM FU SION WELDS r2 INCH NELDS FI L LE T
0 PAGE 16 OF 16
REFERENCES:
- 1. D.R. Miller, W.A. Williams, TORNADO PROTECTION FOR THE SPENT FUEL STORAGE POOI., APED-5696, Class 1, Nov. 1968. Atomic Power Department, G.E. SanJose, Cal
- 2. SPENT FUEL STORAGE RACKS-FUEL BOX CRUSH TEST; Report and Calculations, 4907F17, Wachter Associates, INC.,
1/2/80.
- 3. ,Orner W. Blodgett, DESIGN OF WEI.DED STRUCTURES, James F.
Lincoln Arc Welding foundation. Clev. Ohio.,1975.
NUREG-0800, V.S. Nuclear Regulatory Commission, Standard Review Plan, 3.5.1.0, MISSILES GENERATED BY NATURAL PHENOMENA., Rev. 2, July 1981.
- 5. S. Timoshenko, STRENGTH OF MATERIALS, Part I.
D. Van Nostrand Co. Inc.,N.Y.N.Y., 1955.
- 6. OW. Eshbach, M. Souders, HANDBOOK OF ENGINEERING FUNDEMENTALS, John Wiley 8 Sons, N.Y.N.Y., 1974.
Attachment C In accordance with 10CFR 50.91 this change to the Technical Specifications has been evaluated against three criteria to determine if the operation, of the facility in accordance'ith the proposed amendment would:
involve a significant increase in the probability or consequences of an accident previously evaluated; or
- 2. create the possibility of a new or different kind of accident from any accident previously evaluated; or
- 3. involve a significant reduction in a margin of safety.
As outlined below, Rochester Gas 6 Electric submits that the issues associated with this amendment request are outside the criteria of 10CFR 50.91, and therefore, a no significant hazards finding is warranted.
Attachment B presented an analysis of the dose consequences.
resulting from the impact of a tornado missile. Because of the limited deflection of the rack (1.4 inches) on impact, and the, 2 of 4 storage space configuration, it is impossible to postulate damage in excess of a total number of rods equivalent to five
'ut fuel assemblies. 'sing a conservative X/Q value, considering the assumed tornado conditions, results in a dose at the EAB of 60 rem. This is well within the guidelines of 10CFR 100 and is less than what the NRC considered acceptable for the fuel handling accident inside containment (96 rem).
Therefore a no significant hazards finding is warranted for the following reasons:
There is no increase in the probability of impact of a tornado missile on spent fuel. The consequences were evaluated to be less than what the NRC considered acceptable previously because a more appropriate X/Q value for the postulated tornado condition was used.
- 2. The possibilty of a different kind of accident is not created.
"3. There is no significant reduction in the margin of safety. The consequences of such an accident are less than previous results.