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Category:ARCHIVE RECORDS
MONTHYEARML17292B6881999-04-15015 April 1999
Quality Dept Audit Rept for WNP-2 Emergency Preparedness Program.
ML17292B6591998-10-0606 October 1998
Calculation Mod Record CMR-98-0243, Fracture Mechanics Evaluation of N1A Nozzle Safe End.
ML17164A8881998-06-22022 June 1998
Rev 0 to EC-RISK-1065, Assessment of Common Cause Failure Probabilities for Use in Susquehanna Ipe.
ML17164A8871998-06-22022 June 1998
Rev 0 to EC-RISK-1063, Evaluation of Operator Actions for Application in Susquehanna Individual Plant Examination.
ML20195F8421998-06-0202 June 1998
Revised Design Calculation DC-5957, GL 96-06 Calculations
ML17292B6581998-05-29029 May 1998
Rev 0 to ME-02-98-04, Fracture Mechanics Evaluation of N1 Safe End.
ML17292B6151998-05-26026 May 1998
Rev 0 to Calculation ME-02-98-04, Fracture Mechanics Evaluation of N1 Nozzle Safe End.
ML17292B3681998-04-10010 April 1998
Audit 298-008, WNP-2 Emergency Preparedness Program.
ML20217B5921998-03-0606 March 1998
Rev 1 to L-001443, Reactor Water Cleanup High Flow Isolation Error Analysis
PLA-4652, Partially Withheld Package Consisting of Util Response to NRC RAI Re EA 97-1391998-02-26026 February 1998
Partially Withheld Package Consisting of Util Response to NRC RAI Re EA 97-139
ML20217M6611998-02-20020 February 1998
Rev 2 to L-001166, Post LOCA CR Auxiliary Electric Room & Offsite Dose
ML20202H3051998-02-10010 February 1998
Pp&L,Inc Corporate Audit Svc, Investigation of Radwaste Control Room Offgas Panel Alarm Testing by Auxiliary Sys Operators at Susquehanna Ses, for Job Number 739459-2-98
ML20203E9021998-02-0303 February 1998
Rev 2 to L-001337, Containment Liner Leak Chase Channel Assessment
ML20203J7501997-12-0909 December 1997
Plant Issues Matrix
IA-97-451, Plant Issues Matrix1997-12-0909 December 1997
Plant Issues Matrix
ML20197A4991997-12-0808 December 1997
Rev 1 to L-001378, Calculated Probability for Sustained Boiling in LaSalle Fuel Pools
ML20202H3651997-12-0101 December 1997
Pp&L,Inc Corporate Audit Svc (Interim Rept), Investigation of Allegations That Plant Control Operators Did Not Perform Certain Alarm Test, for Job Number 739459-1-97
ML20249C6391997-11-26026 November 1997
Rev 0 to Design Calculation 5922, NUREG 0619 RPV Feedwater Nozzle Crack Growth Reevaluation
ML20199K7361997-11-21021 November 1997
Rev 0 to Calculation L-001443, Reactor Water Cleanup High Flow Isolation Error Analysis
ML20199K7111997-11-19019 November 1997
Rev 1 to Calculation L-001420, Unit 1 RWCU Room Setpoint Margin Analysis & Loop Accuracy
ML20199K6111997-11-13013 November 1997
Rev 1 to Calculation L-001324, Area Ambient & Differential Temperature Design Basis Calculations for Reactor Coolant Leak Detection
ML20199K5871997-11-0707 November 1997
Rev 1 to Calculation L-001281, RWCU Areas Temperature Response Due to High Energy RWCU Fluid Leakage
ML20199L4091997-11-0303 November 1997
Rev 0 to L-001337, Containment Liner Leak Chase Channel Assessment
ML20202H3411997-10-15015 October 1997
Pp&L,Inc Corporate Audit Svc (Interim Rept), Investigation of Allegations That Mgt Misrepresented Alarm Test Info to NRC, for Job Number 739459-97
ML20217C4051997-09-12012 September 1997
Rev 1 to L-001119, Vc/Ve Mixed Air Relative Humidity Calculation
ML20217C3911997-06-12012 June 1997
Rev 1 to L-001166, Post LOCA Control Room,Auxiliary Electric Equipment Room & Offsite Doses
ML18026A4761997-04-0202 April 1997
Rev 0 to Sensor Response Time Values for Select RPS & MSIV Isolation Functions.
ML17292A8391997-03-31031 March 1997
Audit 297-005, WNP-2 Emergency Preparedness Program.
ML20134P1221996-11-27027 November 1996
Press Release IV-96-60, NRC Staff Proposes $100,000 Fine for Apparent Violations at WNP-2
ML20134M2121996-11-22022 November 1996
Press Release 96-165, NRC Begins First of Series of Design Insps at Nuclear Power Plants
ML17158B8871996-11-19019 November 1996
Rev 1 to Single Failure Analysis for GE Supplied Instruments Connected to Class 1E Circuits.
ML20134G0321996-11-0808 November 1996
Press Release IV-96-59, Nrc,Supply Sys Officials to Discuss Performance at WNP-2
ML18026A4681996-10-30030 October 1996
Rev 2 to Inadvertent Reactor Vessel Injection Resulting from Spurious Operation of HPCI or RCIC Sys.
ML18026A2791996-10-30030 October 1996
Rev 4 to App R Safe Shutdown Path 2 Analysis for Fires in Control Room Fire Zones.
ML20128Q3171996-10-18018 October 1996
Press Release III-96-64, LaSalle Nuclear Station Rated Acceptable in Three Areas, Good in Fourth Area in Latest NRC Assessment Rept
ML20134D2261996-10-17017 October 1996
Press Release IV-96-56, Nrc,Supply Sys Officials to Discuss Apparent Violations at WNP-2
ML18017A2911996-10-0202 October 1996
Rev 0 to EC-ENVR-1025, ISFSI Fuel Cycle 40CFR190 Offsite Dose Calculations.
ML18017A0531996-10-0202 October 1996
Rev 0 to EC-ENVR-1026, SSES Maximum Offsite Dose Rate from ISFSI & Other Fuel Cycle Sources.
ML17292A6101996-07-12012 July 1996
Rev 0 to MOV Pressure Locking Calculation.
ML17292A6111996-07-12012 July 1996
Rev 0 to Pressure Locking Evaluation for RCIC-V-31.
ML20116B6141996-07-0303 July 1996
Rev 2 to Evaluation of H Pratt 54-Triton Xl Butterfly Valve w/H3BC Operator for Seismic Loads
ML20116B5921996-07-0202 July 1996
Rev 2 to Evaluation of 36-125 Lb Crane Ferrosteel Wedge Gate Valve for Seismic Loads
ML20116B5891996-07-0101 July 1996
Rev 1 to Evaluation of 54 Steel By-Pass Line - Lake Screan House. Drawings Encl
ML20210E3201996-06-11011 June 1996
Vol I,Rev 0 to Cchvac Concern Resolution Task - Duct Evaluation of 4316 Sys Series
ML17158C1601996-05-23023 May 1996
Rev 1 to EC-RADN-0531, Secondary Containment Isolation Setpoints.
ML17292A1481996-04-0303 April 1996
WNP-2 Emergency Preparedness Program Audit.
ML20100P8981996-02-29029 February 1996
Main Steam Tunnel Temp Response for Linear Leakage Ramp Rate Sensitivities
ML20100L3821996-02-28028 February 1996
Methodology & Acceptance Criteria
ML20100P9061996-02-28028 February 1996
Rad Monitor Dose Rates Due to Postulated Steam Tunnel Leakage
ML20100L4071996-02-28028 February 1996
Rev 0 to Calculation L-000190 for LaSalle
1999-04-15
[Table view]
Category:PHOTOGRAPHS & SLIDES
MONTHYEARML20091K2771984-02-24024 February 1984
Slide Presentation Entitled, Industry Evaluation of Operating Shift Experience Requirements, Presented at 840224 Meeting Before Commissioners in Washington,Dc
ML20079F1601984-01-13013 January 1984
Viewgraph Presentation Entitled, Independent Design Review
ML17276B6251982-09-28028 September 1982
Slide Presentation Entitled, Corporate Organization & Mgt. Organizational Charts Encl
ML20023C0531982-09-17017 September 1982
Viewgraphs Entitled Insp Cg & E Catalytic,Inc (Ci) Presented at 820810-0917 Audit Meetings
ML18026A3891982-04-30030 April 1982
Slides of Engineering Design Control Program Presented at 820301 Meeting W/Nrc
ML20058E8081981-12-11011 December 1981
Slide Presentation Entitled Six-Inch Pipe Weld QA Presentation, from 811113 Meeting W/Util
ML20078P1921981-10-28028 October 1981
Viewgraphs Entitled Mitigation of Intergranular Stress Corrosion Cracking
ML20079E0811981-10-21021 October 1981
Slide Presentation to NRC Caseload Forecast Panel,811021-23. Related Info Encl
ML17276A1801981-09-15015 September 1981
Presentation on Number of OBE Fatigue Cycles for BWR NSSS Design (Except Piping).
ML17138B7091980-12-18018 December 1980
Slide Presentation Re Seismic & Hydrodynamic Loads
ML19337B4401980-09-18018 September 1980
Viewgraphs of, Structural Modeling Refinement & Development of Improved Analytical Procedures.
ML19337B4391980-09-18018 September 1980
Viewgraphs of, Improved Design Load Spec for Single Valve Discharge.
ML19337B4361980-09-18018 September 1980
Viewgraphs of, Caorso Single Valve Tests,Evaluation of Pressure Data Base.
ML19337B4331980-09-18018 September 1980
Viewgraphs of SRV Discharge Loads Improved Definition & Application Methodology to Mark II Containments.
ML19337B4351980-09-18018 September 1980
Viewgraphs of, Improved SRV Discharge Load Definition.
ML20213D3281980-09-18018 September 1980
Viewgraphs Entitled, Safety/Relief Valve Discharge Loads, Improved Definition & Application Methodology to Mark II Containments
ML19347C2751980-09-0303 September 1980
Slide Presentation from 800903 Meeting Re Coil Properties Used in soil-structure Interaction Analyses
ML19338E7431980-08-28028 August 1980
Viewgraphs Entitled, Emergency Response Facilities.
ML19321A3801980-06-13013 June 1980
Slide Presentation Entitled, Lead Plant Generic T-Quencher Methodology.
ML19320A5101980-05-14014 May 1980
Viewgraph from near-term OL Licensing Resolution Group 800514 Meeting W/Nrc
ML19312E8541980-05-0808 May 1980
Slide Presentation Entitled, NRC Caseload Forecast Panel Meeting, Re Status of Facility Const
ML19318A4261980-04-10010 April 1980
Slide Presentation Entitled, Safety/Relief Valve Discharge Into Suppression Pool.
ML19312D5351980-02-21021 February 1980
Slide Presentation Entitled, Zimmer NSSS Sqrt Re-Evaluation Status Rept.
ML17272A8871980-02-0606 February 1980
Slide Presentation Entitled Sacrificial Shield Wall: Analysis,Design,Const of Sacrificial Shield Wall.
1984-02-24
[Table view] |
Text
_ - _ _ _ _ _ _ - - _ _ - _ _- __ .
EM ME3 l 5/14/80 O
'! NTOL LICENSING RESOLUTION GROUP i.
i MEETING WITH NRC MAY 14, 1980 f, .
. 9:00 A.M.**
PHILLIPS BUILDING BETHESDA, MARYLAND f
t B00625y
. AGENDA 5/14/80 l l
- 1. - REVIEW PURPOSE ,
~
o
SUMMARY
OF MAY 1 MEETING l ORGANIZATION .
INITIAL ASSESSEMENT-OF FEASIBILITY INITIAL NRC RESPONSE
- 2. OBJE'TIVES C OF MAY 14 M$ETING
^
e FINALIZE NRC/ UTILITY INTERFACE e AFFIRMATION OF LEAD PLANT SER APPROACH JOINT REVIEW 0F ISSUES (CATEGORIZE /PRIORITIZE/ RESPOND)
COMMON CLOSURE BASES e HOW IT SHOULD WORK!
(GROUP MECHANICS)
~
- 3. FUTURE" ACTIONS e WORKING GROUP MEETINGS APPLY EXISTING CLOSURE BASES REVIEW REMAINING OPEN ISSUES RSB I & CB MARK II TMI GENERIC ISSUES OTHER BRANCH ISSUES e LEAD PLANT SER ISSUANCE
'TMI SUPPLEMENT 4
= . , ~ , . - w -.. -
- v. - . - . . , . , , c - - - ~ ,.-. -.--y..-r -w r--=---y
BWRNT0LLICENSINGACTI0ffGROUp [-
COMMONWEALTH EDIS0N COMPANY .
LASALLE COUNTY - 1 CINCINNATI GAS & ELECTRIC COMPANY ZIMMER DETROIT EDISON COMPANY FERMI - 2 .
LONG ISLAND LIGHTING COMPANY .
SHOREHAM-PENNSYLVANIA POWER & LIGHT COMPANY SUSQUEHANNA - 1 WASHINGTON PUBLIC POWER SUPPLY SYSTEM . WNP - 2 O q E
N ?O _ ORGAN ZATL ON I
I CHAIRMAM -
NRC
. NT0t. EXECUTIVE 5 -------- R6 VIEW /
i D)R ECTOR (9 Mpp6AL i- ( C. Rf ED , Cico) u
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, : 0 I
"NTol, C00RolNATO R N RC coo RDIM AToR
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I I COMMON i OTHER OTHER
! I uTii.iTY/G6 NRC 3 t
! ! t. - sensnic, --- . GEMERIC .}
pRoggggs PROGRAMS
( M Wll3 TMI) _
(MK ll TMD 3 l _
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NTOL N RC -PM
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! AP?LICAOT APPLICRMT ,
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PLANT l NTOL NRC-PM UNigcG' APPLICANT ---- *-~~~~~ APPLICAMT ISSU65 i
I L
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APPLICAW -- - - - - - ~ - - - - - - - - AppLicawT x < x JL i .
i ._ __ _ _ _ _ _ _ _ __- _
4 -M,
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SUMMARY
LIsT'0F'NSS5LICENSiNGiSSUE5- .
NUMBER RESOLVED NUMBER FOR ON AT WHICH GENERIC FOTENTIAL-
. TYPE OF NUMBER.0F LEAST ONE RESOLUTION IS .. TM.I ..
FEASIBLE EXPANSION- EXAMPLES ISSUE ISSUES DOCKET DEGRADED GRID VOLTAGE CATEGORY 1 8 -- '2 DRAWING AND SCHEMATIC
, PLANT SPECIFIC 23
' CONTRADICTIONS ISSUES RPS MG SET PROTECTION CATEGORY 2 10 29 11 ODYN REANALYSIS BWR DESIGN GENERIC 40 -
ISSUES
' ANTICIPATED TRANSIENTS CATEGORY 3 -
WITHOUT SCRAM (ATWS) 4 12 L ENVIRONMENTAL 4
UNRESOLVED GENERIC 15
' QUALIFICATION (NUREG-051 SAFETY ISSUES i .
.e
COMPLETE SER's
- 1. ZIMMER IS A COMPLETED MODEL
- 2. EXCEPT FOR OPEN ITEMS; ALL PLANTS READY FOR BULK OF.SER ,
- 3. FINISHOPENITEMSFORLEADNT0LPLANT-FILLIN COMMON ITEMS FOR OTHER 5
- 4. TAKE EACH PLANT IN LINE BEYOND LEAD P NT
~
A. COMPLETE PLANT SER B. APPLY COMMON RESOLUTIONS TO REMAINING SER's
- 5. LAST PLANT SHOULD HAVE MINIMUM UNIQUE ITEMS REMAINING AFTER 5 COMPLETED. .
- 6. EACH PLANT SUPPORTS THOSE BEFORE IT AND RECEIVES SUBSEQUENT REDUCTION OF ISSUES-e O
O
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.E N . N PO PT T E UIT UA TR V OR SO I O C RN R E L EF G GU CE BF '
C E F - C A .
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o TN4 70 AWV1d cVB7 (Tdb W1 h) ~~
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bf N OV/0/N6' $UNN* WU$$Af 5 m ua a c.e ~~.
./Alf0M8UY.RfJ04liEa 7fft/? YAI 5979 .
QUESTTON 212.74 fff/fesWS/Gd$$ ((JC1ffED ffff,f[ ~
~
"In analyzing anticipated operational transients, the applicanc has taken credit for plant operating equipment which has not been shown ro be reliabic as required by General Design Criterion
- 29. The staff has discussed the application of this equipmenc generically with General Electric. Based on these discussions, it is the staff's understanding that the most limiting transient that takes credit for this equipment is the excess feedwater event.
Further, it is the staff's understanding that the only plant operating equipment that plays a significant role in citigating this event is the turbine bypass system and the Level 8 high water level trip (closes turbine stop valves). -
"In or$er to assure an acceptable level of performance, it is the staff's position that th'is equipment be identificil in the plant Technical Specifications eith regard to availability, set points, and surveillance testing. The applicant must submit his plan for implemehti'ng this requirement along with any system r.odifications that may be required to fulfill the requirements."
RESPONSE
In discussions between CE and the NRC on November 20 and 21,1978, GE q
V reported on the results of transient analysis when performed to design-basis cccident conditions assu:ptions, and ' equipment avail-abilities, that failure to give credit to the L8 Turbine Trip and the Main Turbine By-Pass system could respectively result in ACPR's of 0.02 and 0.08. In no manner could these postulated accident events res' ult in unacceptable impacts on the health and safety of the public .
as CDC criteria,,No. 29 requires. :
4 L8 Tech Spec
~
The L8 instrumentation is already subject to technical specifications requirements associated with the HPCS. Since the NRC issues the fa-cility tech specs such a requirement can be acccc::'odated by the present design.
Main Turbine By-Pass System Tech Spec The turbine bypass sys'em t and stop valves 3re furnished'with the main turbine generator by Westinghouse and have' exhibited high reliability ,
on existing nuclear and fossil fueled operati6n units. .:
Normal CG&E operating procedures require that the valves be func- .
tionally exercised weekly. This will ensure valve operability and provide adequate. assurance -that the valves.will operate when required. .-
The feedwater LSD uill be submitted in Revicion 58. This drawing will 57 indicate the testability of the control circuits for the turbine bypass. '
valves.
Q212*70~1 l%SS-l9 f ffggggg .
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