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Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data ML20211B5971999-07-31031 July 1999 Cycle 20 Voltage-Based Repair Criteria 90-Day Rept ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20196F4081999-06-23023 June 1999 Revised Pages 71,72 & 298 to Rev 7 of NSPNAD-8102, Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units ML20202J7711999-02-0404 February 1999 Simulator Certification Rept for Prairie Island Plant Simulation Facility,1998 Annual Rept ML20202G3761999-01-31031 January 1999 Non-proprietary Rev 7 to NSPNAD-8102-NP, Prairie Island Nuclear Power Plant Reload SE Methods for Application to PI Units ML20196D7341998-11-20020 November 1998 Third Quarter 1998 & Oct 1998 Data Rept for Prairie Island Isfsi ML20198P0571998-09-0303 September 1998 Rev 1 to 95T047, Back-up Compressed Air Supply for Cooling Water Strainer Backwash Valve Actuator ML20236R6481998-07-15015 July 1998 Metallurgical Investigation & Root Cause Assessment of Part Length CRDM Housing Motor Tube Cracking at PINGP Unit 2 - Preliminary Summary Rept ML20247E2671998-05-0505 May 1998 Rev 0 to Pingp,Units 1 & 2,Pressure & Temp Limits Rept (Effective Until 35 Efpy) ML20216H0341998-03-31031 March 1998 Cycle-19 Voltage Based TSP Alternate Repair Criteria 90-Day Rept ML20217D2041998-03-13013 March 1998 Rev 1 to 28723-A, Intake Canal Liquefaction Analysis Rept for Pingp,Welch,Mn ML20236P9801998-03-12012 March 1998 Rev 0 to 97FP02-DOC-01, Compliance Review of 10CFR50,App R, Section Iii.O RCP Lube Oil Collection Sys ML20216D0911998-03-0606 March 1998 Rev 0 to Prairie Island Generating Plant,Units 1 & 2, Pressure & Temp Limits Rept ML20236Q0021998-02-28028 February 1998 Final Rept, Prairie Island Fire IPEEE Technical Assessment ML20216D1171998-02-28028 February 1998 Non-proprietary Final,Rev 2 to Prairie Island,Units 1 & 2, COMS Setpoint Analysis ML20198N8401998-01-0606 January 1998 Non-proprietary Version of Prairie Island Unit 1 Cycle 19 Thimble Deletion Study ML20212B7941997-10-20020 October 1997 Final Rept for Maximum Allowable Primary to Secondary Leak Rate Using ML20141F2771997-06-30030 June 1997 Non-proprietary Rept NSPNAD-97002, Slb Methodology ML20141A9911997-03-0707 March 1997 Rev 1 to Prairie Island Unit 2 45 F Ultrasonic Exam Rept (3rd 10yr) ML20141A9791997-03-0707 March 1997 F Ultrasonic Exam Rept (3rd 10yr) ML20135A9451997-02-28028 February 1997 Spent Fuel Pool Dilution Analysis ML20135A9481997-02-28028 February 1997 NSP Prairie Island Units 1 & 2 Spent Fuel Rack Criticality Analysis Using Soluble Boron Credit ML20136F3921997-02-28028 February 1997 Response to NRC Addl Info Request ML20138G7791996-12-18018 December 1996 Rev 20 to Operational QA Plan ML20132F2951996-12-12012 December 1996 Rev 0 to NSPLMI-96001, Prairie Island Nuclear Generating Plant Ipeee ML20115E8921995-12-31031 December 1995 FERC Form 1:Annual Rept of Major Electric Utils,Licensees & Others for NSP (Minnesota) & (Wisconsin) ML20094P0351995-11-20020 November 1995 NSP PINGP Units 1 & 2 USNRC USI A-46 Resolution Seismic Evaluation Rept ML20094P0451995-11-15015 November 1995 Rev 0 of Vol I of VI of NSP PINGP Units 1 & 2 USNRC USI A-46 Resolution Ssel & Relay Evaluation Rept ML20199A0251995-03-31031 March 1995 Firezone R Single & Multi-Conductor High Temperature 600 Volt,14 AWG-4/0 Awg Specification ML20078N9531995-01-26026 January 1995 Simulator Certification Rept for Prairie Island Plant Simlation Facility 1994 Annual Rept ML20058M8411993-08-18018 August 1993 Plant,Sbo/Esu Project, Design Rept, Rev 2 ML20056H0451993-04-30030 April 1993 Transient Power Distribution Methodology ML20045C1841993-02-28028 February 1993 Criticality Analysis of Prairie Island,Units 1 & 2 Fresh & Spent Fuel Racks. ML20248L4081992-09-0101 September 1992 Annual Rept Review of Meteorological Data 1991 Monticello & Prairie Island Stations ML20092A6671992-01-14014 January 1992 App B Suppl to Generic Licensing Topical Rept EDR-1, Summary of Plant Specific Crane Data Supplied by Ederer Inc for NSP Co Praire Island Nuclear Generating Plant Units 1 & 2 Auxiliary Bldg Crane ML20086T5631991-12-20020 December 1991 Rev 1 to Prairie Island Nuclear Generating Plant Station Blackout/Electrical Safeguards Upgrade Project ML20092A6791991-12-18018 December 1991 Design Criteria for Lifting Sys for TN-40 Dry Storage Cask at Praire Island ML20067B7211991-02-28028 February 1991 Revised Prairie Island Large Break LOCA Calculation Using WCAP-10924,Vol 1,Addenda 4 Model Changes ML20067C9301991-02-0101 February 1991 Simulator Certification Rept for Prairie Island Plant Simulation Facility ML20058J8881990-10-15015 October 1990 Design Rept for Facility Station Blackout/Electrical Safeguards Upgrade Project ML20011D6991989-12-12012 December 1989 Addendum 1 to, Sacm Diesel Generator Qualification Rept. ML20248D6491989-09-29029 September 1989 Rev 0 to Societe Alsacienne De Consts Mecaniques De Mulhouse Diesel Generator Qualification Rept ML20245L2891989-05-25025 May 1989 Used Molded Case Circuit Breakers Installed in Safety- Related Applications, Addendum 1 to Safety Evaluation 252 ML20245G7711989-04-28028 April 1989 Untraceable Molded Case Circuit Breakers Installed in Safety-Related Application, Safety Evaluation 265 ML20248J8881989-03-31031 March 1989 Criticality Analysis of Prairie Island Units 1 & 2 Fuel Racks ML20245L2631988-11-11011 November 1988 Used Molded Case Circuit Breakers Installed in Safety- Related Applications, Safety Evaluation 252 ML20154Q8981988-09-21021 September 1988 Criteria for Determining Justification for Continued Operation When Encountering Major Discrepancies in 'As-Built' Safety-Related Equipment ML20151Q2201988-08-31031 August 1988 Demonstration of Conformance of Prairie Island Units 1 & 2 to App K & 10CFR50.46 for Large Break Locas ML20150F2771988-06-30030 June 1988 Rev 1 to Safety Evaluation of Increased Fq & Fdeltah 1999-09-30
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217G4461999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Pingp.With ML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data ML20217A1691999-09-22022 September 1999 Part 21 Rept Re Engine Sys,Inc Controllers,Manufactured Between Dec 1997 & May 1999,that May Have Questionable Soldering Workmanship.Caused by Inadequate Personnel Training.Sent Rept to All Nuclear Customers ML20216E7151999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Pingp,Units 1 & 2. with ML20211D3981999-08-24024 August 1999 Safety Evaluation Supporting Requested Actions to Licenses DPR-42 & DPR-60,respectively ML20211C2531999-08-0404 August 1999 Unit 1 ISI Summary Rept Interval 3,Period 2 Refueling Outage Dates 990425-990526 Cycle 19 971212-990526 ML20210Q4891999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Pingp,Units 1 & 2. with ML20211B5971999-07-31031 July 1999 Cycle 20 Voltage-Based Repair Criteria 90-Day Rept ML20209J1131999-07-15015 July 1999 Safety Evaluation of Topical Rept NSPNAD-8102,rev 7 Reload Safety Evaluation Methods for Application to PI Units. Rept Acceptable for Referencing in Prairie Island Licensing Actions ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20209F9811999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Prairie Island Nuclear Generating Plant,Units 1 & 2.With ML20196F4081999-06-23023 June 1999 Revised Pages 71,72 & 298 to Rev 7 of NSPNAD-8102, Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units ML20195G5181999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Prairie Island Nuclear Generating Plant,Units 1 & 2.With . Page 3 in Final Rept of Incoming Submittal Was Not Included ML20207B5931999-05-26026 May 1999 SER Accepting Licensee Proposed Alternative to ASME Code for Surface Exam (PT) of Seal Welds on Threaded Caps for Unit 1 Reactor Vessel Head Penetrations for part-length CRDMs ML20196L2501999-05-13013 May 1999 Rev 0 to PINGP Unit 1 COLR Cycle 20 ML20206L6191999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Pingp,Units 1 & 2. with ML20205N1081999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Pingp,Units 1 & 2. with ML20205Q5101999-03-15015 March 1999 Inservice Insp Summary Rept Interval 3,Period 1 & 2 Refueling Outage Dates 981109-1229 Cycle 19,970327-981229 ML20207J6951999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Prairie Island Nuclear Generating Plant ML20202J7711999-02-0404 February 1999 Simulator Certification Rept for Prairie Island Plant Simulation Facility,1998 Annual Rept ML20202G3761999-01-31031 January 1999 Non-proprietary Rev 7 to NSPNAD-8102-NP, Prairie Island Nuclear Power Plant Reload SE Methods for Application to PI Units ML20207L2811999-01-31031 January 1999 Revised Monthly Operating Repts for Jan 1999 for Pingp,Units 1 & 2 ML20202J1731999-01-22022 January 1999 Safety Evaluation Concluding That NSP Proposed Alternative to Surface Exam Requirements of ASME BPV Code for CRD Mechanism Canopy Seal Welds Will Provide Acceptable Level of Quality & Safety ML20206P7861998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Prairie Island Nuclear Generating Plant.With ML20205H0561998-12-31031 December 1998 Northern States Power Co 1998 Annual Rept. with ML20198J6441998-12-17017 December 1998 Rev 0 to PINGP COLR Unit 2-Cycle 19 ML20206N2731998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Prairie Island Nuclear Generating Plant,Units 1 & 2.With ML20196D7341998-11-20020 November 1998 Third Quarter 1998 & Oct 1998 Data Rept for Prairie Island Isfsi ML20155K6301998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Prairie Island Nuclear Generating Plant,Units 1 & 2.With ML20154H4061998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Prairie Island Nuclear Generating Plant.With ML20202J7991998-09-30030 September 1998 Non-proprietary Version of Rev 3 to CEN-629-NP, Repair of W Series 44 & 51 SG Tubes Using Leaktight Sleeves,Final Rept ML20198P0571998-09-0303 September 1998 Rev 1 to 95T047, Back-up Compressed Air Supply for Cooling Water Strainer Backwash Valve Actuator ML20153B0761998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Prairie Island Nuclear Generating Plant.With ML20237A3961998-08-11011 August 1998 Safety Evaluation on Westinghouse Owners Group Proposed Insp Program for part-length CRDM Housing Issue.Insp Program for Type 309 Welds Inadequate from Statistical Point of View ML20237A8171998-08-0505 August 1998 SER Related to USI A-46 Program GL 87-02 Implementation for Prairie Island Nuclear Generating Plant,Units 1 & 2 ML20236X8531998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Prairie Island Nuclear Generating Plant ML20236R6481998-07-15015 July 1998 Metallurgical Investigation & Root Cause Assessment of Part Length CRDM Housing Motor Tube Cracking at PINGP Unit 2 - Preliminary Summary Rept ML20236R0771998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Prairie Island Nuclear Generating Plant ML20249A5751998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Prairie Island Nuclear Generating Plant ML20247G7011998-05-31031 May 1998 Metallurgical Investigation & Root Cause Assessment of Part Length CRDM Housing Motor Tube Cracking at Prairie Island Nuclear Generating Plant,Unit 2 ML20248M0561998-05-31031 May 1998 Rev 5 to CEN-620-NP, Series 44 & 51 Design SG Tube Repair Using Tube Rerolling Technique ML20247E2671998-05-0505 May 1998 Rev 0 to Pingp,Units 1 & 2,Pressure & Temp Limits Rept (Effective Until 35 Efpy) ML20247G2921998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for Prairie Island Nuclear Generating Plant ML20217M6901998-04-29029 April 1998 Safety Evaluation Accepting Methodology for Relocation of Reactor Coolant Sys P/T Limit Curves & LTOP Sys Limits Proposed by NSP for Pingp,Units 1 & 2 ML20216C6361998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Prairie Nuclear Generating Plant Units 1 & 2 ML20216H0341998-03-31031 March 1998 Cycle-19 Voltage Based TSP Alternate Repair Criteria 90-Day Rept ML20217D2041998-03-13013 March 1998 Rev 1 to 28723-A, Intake Canal Liquefaction Analysis Rept for Pingp,Welch,Mn ML20236P9801998-03-12012 March 1998 Rev 0 to 97FP02-DOC-01, Compliance Review of 10CFR50,App R, Section Iii.O RCP Lube Oil Collection Sys ML20248L3931998-03-10010 March 1998 ISI Summary Rept Interval 3,Period 1 & 2 Refueling Outage Dates 971018-971212 Cycle 18,960303-971212 ML20216D0911998-03-0606 March 1998 Rev 0 to Prairie Island Generating Plant,Units 1 & 2, Pressure & Temp Limits Rept 1999-09-30
[Table view] |
Text
r-s .
Attachment A
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Docket No. 50-306 LER R0-80-03
% 5/23/80 SG-80-04-030 NON-DESTRUCTIVE FXAMINATION OF PRAIRIE ISLAND 2 STEAM GENERATOR TUBING .'
i A. V. Klein Materials Technology Westinghouse Nuclear Technology Division April 25, 1980 Introduction Recent eddy current examinations of Prairie Island Unit 2 steam generator tubes have revealed indications in a number of tubes as summarized below. The affected tubes werc located primarily in the periphery of the bundle and the indications were located at the first and second tube support intersections on the cold leg side. The statistics are as follows:
Steam Generator 21 Steam Generator 22 No. of peripheral tubes inspected: 642 1026 No. of tubes with Indications at 1st TSP 31 40 No. of tubes with Indications at 2nd TSP 7 57 The Indications ranged from less than 20% wall thickness to one at approximately 75%, with the majority in the less than 20% category. The indications were Interpreted as some form of OD wall loss, located in the center of the support plate intersection, but no more definitive information could be obtained from the eddy current signals.
To determine the actual cause of the indications, a representative tube, R34C79 of S/G 22 was removed on Janu:ry 27, 1980, for detailed non-destructive and
_ l 1
destructive examination. This document describes the tube removal and presents 1 the results of the non-destructive examination performed by Westinghouse. The results of the destructive evaluation are reported separately in Vestinghouse Research Report 80-5D9-PRARI-RI, " Examination of Prairie Island 2 Inconel Alley 600 Steam Generator Tube," R. G. Aspden, dated March 10, 1980.
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cxhibited wall loss, the next four intersections had slight discolorations and possibly local deposits and the seventh exhibited sh7tiow circumferential striations. All other areas, including tie 6" length from the tubesheet crevice, exhibited a tight, dull, black corrosion oxide typical of service exposed Inconel 600 tubing, with no extraneous deposits.
B. Radiography - All tube support Intersection zones were radiographed using the double wall technique. The first support zone exhibited %0.010" maximum
- ' ell reduction while the second support zone had approximately 0.020"-0.025" wall loss. The 3rd, 4th, 5th, 6th, and 7th showed no detectable wall loss. Standards shot with each exposure revealed that 1 mil of wall reduction could be detected.
C. Eddy Current Examination - All segments containing tube support plate Intersections were eddy current tested using 400, 200, and 100 kHz multi-frequency techniques. The results were very similar to those obtained from radiography. The first TSP Intersection revealed a 20% indication, the second two indications of 40 and 60%, and the remaining intersections had no evidence
, of wall loss. A simulated tube support plate section was fitted over the tube at the second TSP intersection location and the signal obtained was similar to that observed in the in plant inspection. When the support plate simulant was placed over the first TSP intersection, however, there was no discernable tube indication because of the small amount of wall penetration, %20%. ,
l D. Profilometry - Detailed profilometry of the OD at the affected first and second intersections generally supported the other findings; the maximum OD ;
reduction agreed with the X-ray and E/C estimates of wall loss.
l C ONCLUS IONS The results of the Initial non-destructive examination indicated local OD wall loss in the center of the first and second TSP intersections. All other areas examined were normal and free of any evidence of distress. The eddy current, I
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- 1. Tube Removal Access for remot ing peripheral tubes from the primary side is limited by the curvature of the channel head. In order to remove the selected tube, it was necessary to part the tube, within the tubesheet, at an elevation approximately 6" below the secondary side. This operation was performed from the primary side.
The secondary side was then entered and the tube was again parted below the first anti-vibration bar support at the top of the tube bundle. The bent end was straightened, an inside diameter gripper was inserted in the tube, and the tube was hydraulically raised vertically through the tube support plates. Only 1000 pounds force was required to break the tube loose (a low load indicating freedom from restrictions in the tubesheet crevice and tube support plate intersections).
As the tube was withdrawn manually hend over hand above the top (seventh) tube support plate, it was cut into random lengths approximately 30" long to aid in the removal and subsequent packaging and shipment. Care was taken to avoid cutting the tube near the tube support plate intersections. Also, small paint dabs were placed on each section to indicate its orientation with regard to the wrapper and shell.
After visually examining the tube sections at the plant, they were cut into somewhat shorter segments, packaged, and trucked to the Forest Hills Site for the non-destructive examination.
The ends of each cut segment were numbered sequentially, starting at the top (U-bend) end. Figure 1 describes the Idsntification of the seven relevant
_ sections, each about 12" long across each of the tube support plate intersections.
1
- 2. Non-destructive Examination of Tube R34c79 A. Visual Examination and Photography - All tube sections containing support plate intersections were photographed at 4 rotations (0 , 90 ,180 , 270 ) .
As shown in the attached photographs, only the first and second intersections I
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radiographic and profilometric results were all consistent, Indicating about a 0.010" local wall reduction at the first and a 0.025" local wall reduction at the second TSP intersection. The laboratory eddy current examinations generally reproduced and confirmed the in plant examination results.
The cause of the observed local wall reduction could not be determined from the non-destructive examination. That required destructive metallography and micro-analytical examinations which were subsequently performed and reported separately In Westinghouse Research Report 80-5D9-PRARI-R1, " Examination of Prairle Island 2 Inconel Alloy 600 Steam Generator Tubes," R. G. Aspden, dated March 10, 1980.
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