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Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML17059A3611994-07-0606 July 1994 Emergency Action Level Verification & Validation Rept. ML17311A0181994-05-13013 May 1994 New York State EAL Upgrade Project Verification & Validation Rept. ML20196D3011988-10-31031 October 1988 Reactor Vessel Matl Surveillance Program for Indian Point Unit 2 Analysis of Capsule V ML20155H2541988-09-30030 September 1988 Rev 2 to Indian Point Unit 2 (NRC Bulletin 88-008 Thermal Stresses in Piping Connected to RCS) Indentification of Unisolable Piping & Determination of Insp Locations ML20153F9161988-04-28028 April 1988 Changes,Tests & Experiments - 1986 ML20151T4681988-01-31031 January 1988 Experimental & Finite Element Evaluation of Spent Fuel Rack Damping & Stiffness ML20151U9971986-01-31031 January 1986 Status Rept:Indian Point Special Proceeding Decision CLI-85-06 ML20211G4111985-07-0303 July 1985 Effective Max Beta Energy & Determination of RO-2,2A Response Factors ML20094E0071984-07-26026 July 1984 Addl Info Obtained from Ny State on New York Pirg Petition Which Suppls FEMA Rept to NRC ML20094E0021984-07-26026 July 1984 Evaluation of Addl Issues Raised by NRC Re New York Pirg Petition to Suspend Operation of Indian Point,Units 2 & 3 ML20024B3061983-04-30030 April 1983 Determining Portion of Regulatory Burden on Indian Point 3 Capital & O&M Expenditures 1976 - 1986:10-Yr Cost Projection Also Included. ML20065J6771982-09-14014 September 1982 Testing & Repair of Siren Portion of Alert & Notification Sys for Indian Point Nuclear Power Generating Plant ML1003215451981-10-31031 October 1981 Evaluation of Indian Point Station Unit No 2. ML0934307941981-10-31031 October 1981 Evaluation of Indian Point 3 Nuclear Power Plant. ML20063G0531981-10-26026 October 1981 Evaluation of Prompt Alerting Sys for Indian Point Nuclear Power Station. NUREG/CR-2655, Evaluation of Prompt Notification Sys... & NUREG/CR-2654, Procedures for Analyzing Effectiveness for Alerting Public Encl ML20003C1221981-01-31031 January 1981 Feasibility Study for Mod of Containment Cooling Sys. ML19340E7091980-11-30030 November 1980 Technical Evaluation of the Susceptibility of Safety- Related Sys to Flooding Caused by Failure of Noncategory I Sys for Indian Point 2. ML1002718271980-09-30030 September 1980 Thermal - Radiation Matls Application Data. One Oversize Drawing Encl.Aperture Card Is Available in PDR ML19331B8381980-08-31031 August 1980 Compliance Study in Response to NRC 800211 Confirmatory Order Re Ucs Petition to Suspend Operation.Demonstrates Methods by Which Safety Rules & Regulations Are Implemented. Financial & Reporting Requirements Not Addressed ML19305A7421979-09-28028 September 1979 Safety Evaluation Re Changes to Tech Specs for Mgt Titles ML1009802731979-01-31031 January 1979 Support Sys Redesign for Piping Runs on North Half of Containment. ML1002002351978-12-31031 December 1978 Evaluation of ECCS Per 10CFR50.46 & App K of 10CFR50. Describes Major Reactor Coolant Sys Pipe Ruptures on Westinghouse Model ML1009802761978-04-30030 April 1978 Rept 03-00101,Revision 0,Piping Sys Hydraulic Shock Suppressor Removal Study, & Rept 03-00102,Revision 0, Piping Sys Hydraulic Shock Suppressor Analysis for Removal. ML20079N3161977-07-31031 July 1977 Near-Field Effects of Once-Through Cooling Sys Operation on Hudson River Biota ML1002002391977-06-30030 June 1977 Reactor Vessel Material Surveillance Program,Analysis of Capsule T Submitted in Support of Amend 1 to Application for Amend to License DPR-26.Vessel Matl Surveillance Capsule Was Tested & Evaluated ML20084B1221973-03-30030 March 1973 Review of Replies to AEC Questions by Con Ed on Indian Point Unit 2 Safety & Relief Valve Installation Reanalysis ML20084B2751972-10-11011 October 1972 Investigation & Consequences of Steam Generator Nozzle Debris Plug Assembly Being Inadvertently Left In-Place During Operation of Rcs 1994-07-06
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217H8501999-10-14014 October 1999 Safety Evaluation Supporting Amend 197 to License DPR-64 ML20206U2551999-02-0909 February 1999 Safety Evaluation Supporting Amend 187 to License DPR-64 ML20236Y1571998-08-0303 August 1998 Part 21 Rept Re ASTM A351,GR. CF8 Matl at Indian Point Being Out of Specifications in Molybdenum & Chromium.Cause & Corrective Actions Are Not Stated ML20236V4281998-07-13013 July 1998 Safety Evaluation of TRs WCAP-14333P & WCAP-14334NP, PRA of RPS & ESFAS Test Times & Completion Times. Repts Acceptable ML20236T5511998-06-24024 June 1998 Consolidated Edison Co of Ny,Indian Point Unit 2,Drill Scenario Number 1998C ML20248B2371998-03-31031 March 1998 Revised Monthly Operating Rept for March 1998 for Indian Point Station Unit 2 ML17264A9381997-07-10010 July 1997 Deficiency Rept Re Potential Safety Hazard Associated w/FM-Alco 251 Engin,High Pressure Fuel tube-catalog: 4401031-2 in Which Dual Failure Mode Exists.Caused by Incorrect Forming Process ML18153A1431997-06-10010 June 1997 Part 21 Rept Re Possible Machining Defect in Certain Stainless Steel Swagelok Tube Fitting Bodies.Facilities Have Been Notified About Possible Problem ML20210E3591997-03-27027 March 1997 Part 21 Rept Re Sorrento Electronics Inc Has Determined Operation & Maint Manual May Not Adequately Define Requirements for Performing Periodic Surveillance of SR Applications.Caused by Hardware Failures.Revised RM-23A ML1005008001997-02-28028 February 1997 Conditional Extension of Rod Misalignment TS for Indian Point 3. ML20115J3981996-07-22022 July 1996 Interim Part 21 Rept Re 3/4 Schedule 80 Pipe Furnished to Consolidated Power Supply.Investigation Revealed Only One Nuclear Customer Involved in Sale of Matl ML20096E5101995-12-31031 December 1995 Resubmitted Rev 13 to QA Program ML17059A3611994-07-0606 July 1994 Emergency Action Level Verification & Validation Rept. ML17311A0181994-05-13013 May 1994 New York State EAL Upgrade Project Verification & Validation Rept. ML20029C7801994-03-31031 March 1994 Monthly Operating Rept for Mar 1994 for Indian Point Unit 1. W/940415 Ltr ML20056E5171993-08-31031 August 1993 Technical Review Rept, Tardy Licensee Actions ML20062J2281993-07-23023 July 1993 Consolidated Edison Co of Ny Indian Point Unit 2,Drill Scenario 1993 ML20118A2681992-12-31031 December 1992 Consolidated Edison Co of Ny Indian Point,Unit 2 Exercise Scenario,1992 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 ML20096H2301992-05-21021 May 1992 Special Rept:On 920504,south Side Lower Electrical Tunnel Detection Sys 8 Taken Out of Svc for Mod to Reposition Detection Sys Run of Conduit.Detection Sys Declared Operable on 920520 After Mod Completed & Sys Retested ML20079F9181991-05-31031 May 1991 Structural Evaluation of Indian Point,Units 2 & 3 Pressurizer Surge Lines,Considering Effects of Thermal Stratification ML20059E9461990-08-31031 August 1990 Nonproprietary Rev 2 to Indian Point 2 Tube Fatigue Reevaluation ML20059G2011990-07-31031 July 1990 Final Rept on Steam Generator Insp, Repair & Restoration Efforts During 1990 Midcycle Insp ML20058K4151990-06-30030 June 1990 Steam Generator Insp,Repair & Restoration Program Presentation to Nrc ML20058K4121990-06-30030 June 1990 Status Rept,Indian Point Unit 2 Mid-Cycle Steam Generator Insp Presentation to Nrc ML20043A4891990-05-30030 May 1990 Nonproprietary Indian Point Unit 2 Steam Generator Insp, Repair & Restoration Program. ML20042F1441989-12-31031 December 1989 New York Power Authority Annual Rept for 1989. W/900430 Ltr ML19332B9371989-11-30030 November 1989 Nonproprietary Info Presented to NRC Re Indian Point Unit 2 Steam Generator Secondary Side Loose Objects. ML19332D6661989-10-31031 October 1989 Nonproprietary Rev 2 to Indian Point Unit 2 Steam Generator Girth Weld/Feedwater Nozzles Rept Spring,1989 Outage. ML20247J8161989-07-31031 July 1989 Safety Evaluation for UHS Temp Increase to 95 F at Indian Point Unit 3 ML20248B3171989-06-30030 June 1989 Rev 1 to Nonproprietary WCAP-12294, Indian Point Unit 2 Steam Generator Girth Weld/Feedwater Nozzles Rept,Spring 1989 Outage ML20248D3631989-06-30030 June 1989 Rev 1,to Indian Point Unit 3 Reactor Vessel Fluence & Ref Temp PTS Evaluations ML20247N5331989-05-31031 May 1989 Nonproprietary Indian Point Unit 2 Steam Generator Girth Weld/Feedwater Nozzles Rept Spring,1989 Outage ML20247J8071989-05-31031 May 1989 Containment Margin Improvement Analysis for Indian Point Unit 3 ML20247G5171989-04-30030 April 1989 Monthly Maint Category I Rept Pages from Monthly Operating Rept for Apr 1989 for Indian Point ML20244C3311989-04-10010 April 1989 Safety Evaluation Supporting Amend 137 to License DPR-26 ML20248F4211989-03-31031 March 1989 NSSS Stretch Rating-3,083.4 Mwt Licensing Rept ML20235V5931989-03-0202 March 1989 Special Rept:During Cycle 6/7 Refueling Outage Scheduled from Feb-May 1989,openings Will Be Made in Plant Penetration Fire Barriers in Order to Install Various Mods. Fire Watches Posted & Fire Detection Tests Completed ML20248F3001988-12-31031 December 1988 10CFR50.59(b) Rept of Changes,Tests & Experiments Completed in 1988 ML20246E2711988-12-31031 December 1988 Con Edison 1988 Annual Rept ML20196D3011988-10-31031 October 1988 Reactor Vessel Matl Surveillance Program for Indian Point Unit 2 Analysis of Capsule V ML20155H2541988-09-30030 September 1988 Rev 2 to Indian Point Unit 2 (NRC Bulletin 88-008 Thermal Stresses in Piping Connected to RCS) Indentification of Unisolable Piping & Determination of Insp Locations ML20154M5661988-08-31031 August 1988 Monthly Operating Rept for Aug 1988 for Indian Point Station Unit 2 ML20154M5691988-07-31031 July 1988 Revised Page to Monthly Operating Rept for Jul 1988 for Indian Point Station Unit 2 ML20151N7881988-07-31031 July 1988 Rev 1 to Charpy Toughness & Brittle Transition Temp Characterization of HAZ High Hardness Zone of Indian Point Unit 2 Steam Generator Girth Weld by Gleeble Weld Thermal Cycle Simulation ML20006B4741988-05-31031 May 1988 Nonproprietary Indian Point Unit 2 Evaluation for Tube Vibration Induced Fatigue. ML20151N7821988-05-31031 May 1988 Fracture Sensitivity Study of Girth Weld 6 Repaired Configuration,Indian Point Unit 2 ML20153F9161988-04-28028 April 1988 Changes,Tests & Experiments - 1986 ML20151T4681988-01-31031 January 1988 Experimental & Finite Element Evaluation of Spent Fuel Rack Damping & Stiffness 1999-02-09
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CORPORATE STANDARO THERMAL - RADIATION MATERIALS APPLICATION DATA TR.44763BB Description Flame retardant glass mat reinforced polyester laminated composition.
Application: DH breakers, terminal boards, radomes, insulation barriers.
Thermal Data This item is a glass fiber reinforced polyester composite having the properties of a NEMA GPO-2 laminate. Figure 1 shows the 40 year life temperature for GPO-2 laminates as a function of the end of life criterion. If 50% retention is chosen as the end of life criterion, the 40 year life temperatures are 1250C for flexural strength and 980C for dielectric strength.
Radiation Data **
Glass polyesters are among the more radiation resistant materials.
Their resistance will vary with the ratio of glass to polyeste ard manufacturing variations. In general, radiation effects literature indicates that most glass polyesters are not signi ficantly 8 affected (4 25% change in any parameter) for doses 4$ 5 x'10 rads. A radiation effects study Sf a similar glass polyester indicates that after a dose of 10 rads, the flexural and tensile strengths have degraded <25%.
- Report #77 IC6 CBSEL M32 Westinghouse Electric Corporation COPPoRATE STANODARS,. & CENTER *~*kt?§ 1* USA Sep, 1980 e202o9o363 8202o4 PDR ADOCK 0500026 P PDR.
Curve 693090-A 140 C-,
0 a
I-30 40 50 60 70 80 End of Life Criterion, %of Initial Strength Fig. I - Forty year life temperature vs end of life criterion for GPO-2 laminate TR-44763BB, page 2
TEFBANKLIN INSTITUTE UU U RESEARCH LABORATORIES THE BENJAMIN FRANKLIN PARKWAY
- PHILADELPHIA. PENNSYLVANIA 19103 TEL£PHONC (215) 44S-1000 MEC1-LANICAL AND NUCLEAR ENGINEERING DEPARTMENT May 14, 1971 Mr. R. L. Korner Electronic Tube Division Westinghouse Electric Corp.
Elmira, New York 14902
Dear Mr. Korner:
The electrical penetration which you provided was mounted on a test chamber that is used to simulate the environment that results from an accident in a nuclear power station containment building. In particular the temperature, pressure, humidity and chemical spray conditions are provided with chamber. The radiation exposure is done elsewhere.
Your electrical penetration was used to provide electric power connection for 6 simulated "loss of coolant accident." The primary test being the one for Westinghouse Nuclear Energy Systems Division, Pittsburgh. This test is documented in FIRL final report F-C2709 (Jan. 1970) which is property of Westinghouse.
The penetration was used to provide power connection for a series of tests, with your permission, before it deteriorated and had to be removed. The evaluation of material and/or equipment under test in each particular case is proprietary information of that sponser so we cannot supply you with copies of reports. However, the exposure and performance of your penetration is tallied in the table which I have compiled and attached.
Very truly yours, L. E. WITCHER Research Engineer Atch.
ELECTRICALPENETRATION PERFORMANCE HISTORY FIRL Final Report TEST ENVIRONMENT PENETRATION PERFORMANCE F-C2709, Jan. 1970 hours 55 psig steam 285°F Satisfactory performance in hours 20 psig steam 220*F every respect days 5 psig air 1620F 100% RH F-C2737, Apr. 1970 hours 82 psig steam 320 0 F Insulation of Pigtail indicated days 5 psig steam 228°F signs of decay-could not megger hours of Boric Acid Spray or hipot with penetration in circuit of item under test.
F-C2750, Mar. 1970 hours 51 psig steam 295°F Use for supplying power to item days 6 psig air 160'F 100% RH under test - 480-10 amps hours of Boric Acid Spray F-C2857, Sept. 1970 hours 50 psig steam 276*F Use for supplying power to item days 164 psig air 100% RH under test 480-10 amps hours of Boric Acid Spray F-C2870, Sept. 1970 1 min. 55 psig steam 275°F Use for supplying power to item I
decreasing to 9 psig over 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> under test 3U 480-39 amps 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 15.5 psig air 208°F 100% RH F-C2896, Dec. 1970 3 days 12 psig steam 240*F Moisture neted in annular section 3 days 7 psig air 160* 100% RH of penetration removed from test Boric Acid Spray 3 days chamber
Clarification Of the leak rate' testing during and after exposure.
The canister tested has two seals approximately 9 feet apart. The outboard seal remains at a relatively low temperature of 107'F. As both seals are considered in determining leakage thru the penetration, the same leak rate established before the test, applies during and after the test since the outboard seal is not affected by the test.
-6 This value was found to be less than 1IX 10. std. cc/sec. (helium) prior to test and therefore applies duri ng and after the test.
The fact that the internal canister pressure 14.2 PSIG, remained at the same value before and after the test verified that the inboard seal did not leak within the limits of the accuracy of the gage reading.
6.0 JUSTIFICATION THAT THE PROTOTYPE UNIT TESTED ACTUALLY REPRESENTS THE INSTALLED EQUIPMENT The installed penetration is shown in Drawing D-E 2198 and is-the same as the specimen tested except for the following:
(1) Test specimen was 2" larger in diameter.
(2) Test specimen was 16" longer.
These deviations are inconsequential to performance. The test specimen therefore can be considered representative of the actual Indian Point No. 3 design.
Dwg,. E2198 7.0 IDENTIFICATION OF MATERIALS USED FOR TYPES 1, 2, 3 (Ref. E2198)
NO. PART MATERIAL
- I. Container Stainless Steel A312 Type 304L
- 3. Monitor Ring Carbon Steel A516 Grade 70
Cajon Fitting #16 PBW-A-85W Carbon Steel
- 5. Header Stainless Steel A240 Type 304
- 6. Shroud (inner) Stainless*Steel A312 Type 304
- 7. Potting Dow Corning Silicone Compound 96-052
- 8. Potting Epoxy - Minnesota Mining & Manufacturing XR 5136
- 9. Spacer Flame Retardant glass mat reinforced polyester laminated composition P.D.S. #44763BB
- 10. Monitor Tube Copper 1N. Pressure Gage Steel
- 12. Sleeving Silicone Rubber Fiberglass Ben-Har 1151
- 13. Sleeving Raychem heat shrink RNF 100
- 14. Conductor Seal Metal Ceramic Seal #16 AWG to #10 AWG Size 55-193
- 16. Shroud-Outer Stainless Steel A240-304 Only Items marked with asterisk (*) have pressure retaining function.
8.0 QUALIFICATION DATA FOR TYPE6