ML19326D070
ML19326D070 | |
Person / Time | |
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Site: | Arkansas Nuclear |
Issue date: | 02/21/1973 |
From: | Hall W, Newmark N NATHAN M. NEWMARK CONSULTING ENGINEERING SERVICES |
To: | |
Shared Package | |
ML19326D066 | List: |
References | |
NUDOCS 8004300696 | |
Download: ML19326D070 (2) | |
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Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML20195D1991999-05-28028 May 1999
[Table view]Probabilistic Operational Assessment of ANO-2 SG Tubing for Cycle 14 ML20199F7401998-11-16016 November 1998 Rev 9 to ANO-1 Simulator Operability Test,Year 9 (First Cycle) ML20155H7161998-07-15015 July 1998 Rev 1 to 96-R-2030-02, Revised Reactor Vessel Fluence Determination ML20248J9901998-05-26026 May 1998 Rev 0 to SIR-98-055, Risk Evaluation & Element Selection in Support of ASME Code Case N-560 ANO-1 ML20217N9101998-03-27027 March 1998 Rev 0 to SIR-98-026, Svc History & Susceptibility Review, Risk Evaluation & Element Selection for Svc Water Sys at ANO-2 ML20217N8771998-03-27027 March 1998 Rev 0 to SIR-98-011, Evaluation of Damage Mechanisms for Svc Water Sys at ANO-2 ML20236J1331998-02-16016 February 1998 Rev 1 to 97-R-2015-01, PPS Response to Failure of DC Bus ML20203D9841997-11-30030 November 1997 ANO Commitment Change Summary Rept for Sept 1995-Nov 1997 ML20211N4631997-10-0707 October 1997 SG Tube Integrity Evaluation ML20154B7501996-11-30030 November 1996 Rev 0, Review of Degradation Mechanisms in EPRI Risk- Informed Inservice Evaluation Procedure ML20135E8141996-11-30030 November 1996 Pressure-Temp Limits for 32 Efpy ML20135D2431996-10-16016 October 1996 Rev 1 to ANO Unit 2 Leak Rate Model for Circumferential Cracks in SG Tubes ML20135D2591996-10-16016 October 1996 Rev 0 to ANO Unit 2 Projected End of Cycle 12 Circumferential Crack Population ML20135D2481996-10-16016 October 1996 Rev 1 to Probability of Burst Model for ANO-2 Tts Circumferential Cracks ML20133M1641996-09-0909 September 1996 Revised OTSG Iga Sizing Technique App H Qualification ML20116F3021996-07-31031 July 1996 Steam Generator Operational Assessment ML20135D2551996-04-11011 April 1996 ANO Unit 2 SG Tubes Evaluation of Bursh Pressures W/Circumferential Flaws Present ML20135D2621996-03-31031 March 1996 ANO Unit 2 SG Tubes:95/95 Mechanical Properties ML20101R5381996-02-28028 February 1996 LTOP for 21 Efpy ML20101H1111996-02-28028 February 1996 Input to Items 2A & 2C of NRC Question on Relief Request for Insp of Transition Piece to Bottom Head Weld of Reactor Vessel at Arkansas Nuclear One - Unit 1 ML20101H1261996-02-22022 February 1996 Flaw Acceptance Stds for Arkansas Nuclear One Unit 1 Reactor Pressure Vessel Weld Insps ML20095C2431995-10-31031 October 1995 ANO Unit 2 Simulator Performance Testing Rept ML20091R6571995-08-31031 August 1995 ANO Unit 2 SG Tube Circumferential Cracking Repair Limit Evaluation ML20082E7541995-03-24024 March 1995 Failure Analysis Rept 3707R MSSV Valve Serial Number BM-02978 Nonconforming Matl - Lower Adjusting Ring Pin at Wyle Lab ML20080H6631995-02-28028 February 1995 ANO Unit 2 SG Circumferential Cracking Evaluation ML20077S0081995-01-17017 January 1995 Simulator Performance Testing Rept ML20063L7551994-02-0101 February 1994 Rev 0 to Twentieth Yr Physical Surveillance of ANO Unit 1 Containment Bldg ML20057D0741993-09-21021 September 1993 ANO Unit 2 Cycle 10,Proposed Changes to Incore Detector Ts ML20057D0711993-09-21021 September 1993 Cycle 10,COLSS & Cpcs Addressable Constants for Reduction of Operable Incore Detectors ML20056E2331993-08-31031 August 1993 Evaluation of Thermal Stratification Effects on Shutdown Cooling Line for Ano,Unit 2 ML20067D3311993-08-31031 August 1993 Technical Evaluation Rept,Evaluation of B&Wog Pressurizer Surge Line Thermal Stratification Program to Address NRC Bulletin 88-11 ML20056E2291993-08-17017 August 1993 SG Degraded Tube Analysis Per Reg Guide 1.121 ML20056D3111993-01-29029 January 1993 Analysis Allowing Extended Operation W/One PPS Channel in Bypass for ANO Unit 2 ML20101K9861992-06-23023 June 1992 Response to NRC Generic Ltr 92-01 for Arkansas Nuclear One - Unit 2 ML20097G0851992-06-0808 June 1992 Rev 0 to ASME Code Case N-481 Evaluation of Arkansas Nuclear One Unit 1 Reactor Coolant Pumps ML20092C9851991-12-31031 December 1991 Technical Rept, Monitoring Hydrogen Gas in Containment During Early Phases of Severe Accident ML20087A4731991-12-20020 December 1991 Rev 0 to Summary of Entergy Position Re Deferral of Permanent Nozzle Repair for Arkansas Nuclear One,Unit 1 Pressurizer Level Tap Nozzle, Engineering Rept ML20081K9331991-05-31031 May 1991 Final Rept on Low-Temperature Overpressure Protection Analysis for ANO 2 for 21 Effective Full Power Years ML20073K8451991-03-31031 March 1991 Entergy Nuclear Performance Rept,Mar 1991 ML20028H6441991-01-17017 January 1991 ANO Unit 1 Simulator Certification. ML20076B3501990-12-31031 December 1990 Criticality Analysis of Arkansas Nuclear One,Unit 1 Fresh Fuel Rack ML20235N6101989-02-14014 February 1989 Summary of Engineering Evaluation of ANO-1 High Pressure Injection Sys Backflow Condition ML20235T2121989-02-14014 February 1989 Addendum to Amend 6 to SAR for Arkansas Nuclear One Unit 2 ML20206F4041988-10-31031 October 1988 Request for Exemption from ATWS Rule 10CFR50.62 for Arkansas Nuclear One Unit 2 ML20207B2901988-07-0101 July 1988 Vols I & II of Rept, Fifteenth Yr Physical Surveillance of Arkansas Nuclear 1 - Unit 1 Primary Reactor Contatinment Bldg ML20154F6091988-07-0101 July 1988 Ten-Yr Visual Tendon Surveillance of Arkansas Nuclear One, Unit 2 Primary Reactor Containment Bldg,Surveillance Rept ML20151N3001988-04-30030 April 1988 Primary Reactor Containment Integrated Leakage Rate Test, Final Rept ML20147C6491987-12-0303 December 1987 One,Unit-II,Simulator Operability Test ML20147C6291987-11-18018 November 1987 Revised Abstract:Performance Test Development,Conduct & Evaluation ML20238C3071987-08-28028 August 1987 Justification for Continued Operation 1999-05-28 Category:TEXT-SAFETY REPORT MONTHYEARML20217L8931999-10-31031 October 1999
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(First Cycle) ML20195B4801998-11-0707 November 1998 Rev 20 to ANO QA Manual Operations 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 0CAN119808, Monthly Operating Repts for Oct 1998 for Ano,Units 1 & 2. with1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Ano,Units 1 & 2. with ML20197H0741998-10-29029 October 1998 Rev 1 to Third Interval ISI Program for ANO-1 ML20155C1351998-10-26026 October 1998 Rev B to Entergy QA Program Manual ML17335A7641998-10-22022 October 1998 LER 98-004-00:on 980923,inadvertent Actuation of Efs Occurred During Surveillance Testing.Caused by Personnel Error.Personnel Involved with Event Were Counseled & Procedure Changes Were Implemented.With 981022 Ltr ML20154J2471998-10-0909 October 1998 SER Accepting Inservice Testing Program,Third ten-year 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N A T H A N M. NEWMARK 8004300 [
CON 3OLTING ENGINEERING SERVICES 1114 CIVIL ENGINEERING BUILDING URBANA. ILLINOIS 618ol 21 February 1973 FINAL REPORT ON FINAL SAFETY ANALYSIS REPORT FOR ARKANS AS NUCLEAR ONE -- UN IT I ARK /NS AS POWER AND LIGHT COMPANY AEC Docket No. 50-313 by N. M. Newmark and W. J . Hall After our review of the FSAR, including Amendments through 29, we believe that the design of the Arkansas Nuclear One -- Unit 1 plant can be considered adequate in terms of provisions for safe shutdown and containment for a Design Basis Earthquake of 0.20g transient maximum horizontal ground acceleration and capable otherwise of withstanding the effects of an Operating Basis Earthquake of half this intensity.
Our review was based on consideration of the design criteria and resul ts of analysis as presented by the applicant in the FSAR and the answers to questions accompanying the FSAR for the site foundations, dynamic analysis of reactor buildings and other Class I structures, piping, vertical earthquake excitation, buried piping, stress limits, Class i equipment in Class li s tructures , reactor internals, cri tical items of control and instrumentation,
. penetrations , and the design of the reactor l ining.
,m .....
-- 2
~ The procedures-employed in the design as described in the FSAR are
' in accordanc'e with'.. the state-of-the-a rt. We believe that the design will incorporate an acceptable _ range of margins of safety to resist the design seismic hazard considered.
.