ML17328A466
| ML17328A466 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/27/1990 |
| From: | Pierson R Office of Nuclear Reactor Regulation |
| To: | Miraglia F, Murley T, Russell W NRC |
| References | |
| NUDOCS 9010050288 | |
| Download: ML17328A466 (62) | |
Text
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Docket No. 50-315/316 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 September 27, 1990 MEMORANDUM FOR:
T. Murley" F. Miraglia*
W. Russell, ADT" J. Partlow, ADPR" D. Crutchfield S.
Varga C.
Rossi J.
Richardson A. Thadani B. Grimes F.
Congel J.
Roe R.
Wessman G.
Lainas J. Zwolinski B. Boger W. Travers F. Gillespie E. Butcher W. Lanning P.
- McKee, EAB/Acting Chief T. Gody, Sr.,
LPEB W. Bateman, EDO Operations Center""
L. Reyes, RII""
Cox)h )k )hl THRU:
Robert C. Pierson, Director Project Directorate III-1 Division of Reactor Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation FROM:
Timothy G. Colburn, Senior Project Manager Project Directorate III-1 Division of Reactor Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation
SUBJECT:
DAILY HIGHLIGHT -
FORTHCOMING MEETING WITH Indiana Michigan Power
- Company, D.C.
Cook Units 1 and 2
DATE & TIME:
October 4,
1990 9:30 a.m. - 5:00 p.m.
LOCATION:
One White Flint North 11555 Rockville Pike Rockville, MD Room 16Bll PURPOSE:
To discuss recent SSFI inspection results at D.C.
Cook Nuclear Power Plant - see attached agenda r
9010050288 900927 1
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PARTICIPANTS":
NRC UTILITY J. Zwolinksi W.
Lanning R.
Pierson T. Colburn P. Koltay S.V. Athvale M. Phillips, RIII D. L.
T. 0.
M. P.
S. J.
J. R.
P. G.
M. S.
J. C.
Williams Argenta Alexich Brewer Anderson Schoepf Ackerman Jef frey Timothy G. Colburn, Senior Project Manager Project Directorate III-1 Division of Reactor Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation cc:
See next page
" Meetings between NRC technical staff and applicants or licensees are open for interested members of the public, petitioners, intervenors, or other parties to attend as observers pursuant to "Open Meeting Statement of NRC Staff Policy,"
43 Federal Receister
- 28058, 6/28/78.
PARTICIPANTS":
NRC UTILITY J. Zwolinksi W.
Lanni ng R.
Pi erson T. Colburn P. Koltay S.V. Athvale M. Phillips, RIII D. L.
T.0.
M. P.
S. J.
J.R
~
P. G.
M. S.
J AC.
Williams Argenta Alexich Br ewer Anderson Schoepf Ackerman Jef frey cc:
See, next page
/s/
Timothy G. Colburn, Senior Project Manager Project Directorate III-1 Division of Reactor Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation
" Meetings between NRC technical staff and applicants or licensees are open for interested members of the public, petitioners, intervenors, or other parties to attend as observers pursuant to "Open Meeting Statement of NRC Staff Policy,"
43 Federal
~Re ister 28058, 6/28/78.
DISTRIBUTION Docket File NRC 8 Local PDRs PD3-1 Reading T. Colburn W.
Lanning OGC (MS15B18)
E. Jordan G.
Brimes Receptionist (White Flint)
NRC Participants ACRS(10)
(MSP315)
GPA/PA (MS2G5)
V. Wilson L. Thomas J. Clifford (MS 17G21)
J.
Stransma (RIII)
S.
Meador OFC: PD3-1/PM NAME
- TCelburP DATE
- g /4'7/90
- SM r
- RPierso
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- PD3-1/LA: PD-31/D OFFICIAL RECORD COPY Document Name:
DC COOK MEETING NOTICE 10/4
0 I
IJ I
Nr. Gerald B. Slade Palisades Plant Palisades Plant cc:
N. I. Nilier, Esquire Sidley 8 Austin 54th Floor One First National Plaza Chicago, Illinois 60603 Nr.
Thomas A. NcNish, Secretary Consumers Power Company 212 West Nichigan Avenue
- Jackson, Nichigan 49201 Judd L. Bacon, Esquire Consumers Power Company 212 West Nichigan Avenue
- Jackson, Nichigan 49201 Reg iona l Admini str a tor, Reo ion I I I U.S. fluclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 Jerry Sarno Township Supervisor Covert Township 36197 N-140 Highway Covert, Nichigan 49043 Office of the Governor Room 1 - Capitol Building Lansing, Nichigan 48913 Nr. David J.
Vandewalle Director, Safety and Licensing Palisades Plant 27780 Blue Star Nemorial Hwy.
Covert, Nichigan 49043 Resident Inspector c/o U.S. Nuclear Regulatory Commission Palisades Plant 27782 Blue Star Nemorial Hwy.
Covert, Nichigan 49043 Nuclear Facilities and
- Environmental Nonitoring Section Office Division of Radiological Health P.O.
Box 30035 Lansing, Nichigan 48909 Gerald Charnoff, P.C.
Shaw, Pittman, Potts 5 Trowbridge 2300 N. Street, H.W.
D.C.
20037 Nr. David L. Brannen Vice President Palisades Generating Plant c/o Bechtel Power Corporation 15740 Shady Grove Road Gaithersburg, Naryland 20877
D.
C.
Cook Units j. and 2 SSFIL Inspection Meeting Agenda Introduction
Background
Design Verification Technical Issues (I) Thermal Relief Valve Sizing (2) Cable Sizing (3) Circuit Breaker Fault Current Rating (4) Inverter Voltage qualification (5) Diesel Generator Load Study (6) Battery Testing Conclusions M.
P. Alexich M.
ST Ackerman P.
G.
Schoepf P.
G.
Schoepf J.R.
Anderson T. 0. Argenta
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 0
MEMORANDUM FOR:
James A. Norberg, Chief Mechanical Engineering Branch Division of Engineering Technology THRU:
FROM:
SUBJECT:
David Terao, Section Chief Component Technology Section Mechanical Engineering Branch Division of Engineering Technology Shou-nien Hou Component Technology Section Mechanical Engineering Branch Division of Engineering Technology MEETING MINUTES - D.C.
COOK THIMBLE TUBE WEAR The staff met with the licensee for the Donald C.
Cook facility (Indiana Michigan Power Company) and representatives from Mestinghouse Electric Company
'1 11 1991 The urpose of this meeting was to discuss the root cause of thimble tube wear in the Mestingh'ouse reactor vessel and leakage which occurred in Unit 2 and the licensee's plans to mitigate the tube wear concerns.
Enclosure 1 is the list of attendees.
BACKGROUND:
A thimble tube leak occurred at D.C. Cook, Unit 2,,on June 18, 1990.
This was the third thimble tube leakage incident at this plant.
The first two incidents occurred in Unit 1 in 1983 and 1988.
Since then, all thimble tubes in both 1
d Th tubes in Unit 2 were replaced in December 1988.
This since 1978 latest leakage incident was unexpected because Unit 2 had operated si with no leaks of the original tubes for ten years.
However, the leak occurred with the new tubes even before the end of the first fuel cycle.
A meeting was held between the licensee and NRC on August 29, 1990, to discuss the incident.
At that time, the licensee was investigating the problem and agreed to provide NRC with the results of their root cause evaluation and corrective action plans by ear ly 1991.
The purpose of the present meeting was t
d th's information to the staff.
Enclosure 2 provides a copy of the meeting presentation slides.
Highlights of the snformation presente
~
~
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d at the meeting are summarized below.
ROOT CAUSE EYALUATION:
Following the Unit 2 leakage incident, the licensee performed eddy current inspections of the thimble tubes and found significant wear in several tubes.
The wear was detected primarily in the lower core plate area.
Two tubes, including the failed tube and another tube with severe
- wear, were sent to
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APR 2g 1S9l llestinghouse hot cell facilities for metallurgical investigation and evaluation.
The examinations included surface examinations, metallographic examinations, detailed wear morphology evaluations and hardness measurements.
The results were documented in Mestinghouse report MT-MNA-035(91), "Metallurgical Investigation of Wa11 Thinning and Leakage of the BMI Thimble Tubing at D.C.
Cook Unit 2," January 1991.
A copy of this report was provided to the staff.
A photograph of the wear scar of the failed tube is shown in Enclosure 3.
The licensee concluded that the scar shape matched the bottom nozzle of the fuel assembly.
The wear was primarily one-sided.
The examination indicated lack of cold work or deformation.
Flaking was seen near the area that had worn through.
The results suggested that the wear mechanism was low-impact fretting wear.
The licensee stated that flow induced vibration was the root cause of the problem.
However, the licensee could not provide a sound explanation for the cause of the accelerated wear among the newly installed tubes.
l/EAR MITIGATION PLAN The licensee plans to implement a fix to reduce wear rate during the 1992 refueling outage.
The licensee is in the process of evaluating bids from several vendors.
The proposed fixes involve both flow reduction and wear resistant thimble tube designs.
The licensee has reviewed industry activities and feels that any reactor core modifications to reduce the flow-induced vibration have no guarantee of success.
They have considered installing larger tubes but concluded that this is not feasible for their plant.
Enclosures:
As stated Original sigtgd 4yp Shou-nien Hou Component Technology Section Mechanical Engineering Branch Division of Engineering Technology cc:
J.
E. Richardson B.
D. Liaw C.
Y. Cheng L. B. Marsh T.
G. Colburn D. DeGrassi, BNL Distribution:
Central Files EMEB Files JNorberg DTerao DET: 8 DET:EME SHou DTerao f/z.g/91 g /L3/91
APR 83 ASS) Westinghouse hot cell facilities for metallurgical investigation and evaluation.
The examinations included surface examinations, metallographic examinations, detailed wear morphology evaluations and hardness measurements.
The results were documented in Westinghouse report MT-MNA-035(91), "Metallurgical Investigation of Wall Thinning and Leakage of the BMI Thimble Tubing at D.C.
Cook Unit 2," January 1991.
A copy of this report was provided to the staff.
A photograph of the wear scar of the failed tube is shown in Enc'losure 3.
The licensee concluded that the scar shape matched the bottom nozzle of the fuel assembly.
The wear was primarily one-sided.
The examination indicated lack of cold work or deformation, Flaking was seen near the area that had worn through.
The results suggested that the wear mechanism was low-impact fretting wear.
The licensee stated that flow induced vibration was the root cause of the problem.
However, the licensee could not provide a sound explanation for the cause of the accelerated wear among the newly installed tubes.
WEAR MITIGATION PLAN The licensee plans to implement a fix to reduce wear rate during the 1992 refueling outage.
The licensee is in the process of evaluating bids from several vendors.
The proposed fixes involve both flow reduction and wear resistant thimble tube designs.
The licensee has reviewed industry activities and feels that any reactor core modifications to reduce the flow-induced vibration have no guarantee of success.
They have considered installing larger tubes but concluded that this is not feasible for their plant.
Or.'g'.rJ stgmdby;"
Shou-nien Hou Component Technology Section Mechanical Engineering Branch Division of Engineering Technology
Enclosures:
As stated cc:
J.
E. Richardson B.
D. Liaw C.
Y. Cheng L. B. Marsh T.
G. Colburn D. DeGrassi, BNL Distribution:
Central Files EMEB Files JNorberg DTerao DET: B SHou
</ /xJ/91 DET:EME DTerao g />3/91
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Attachment 1
Attendance List
Thimble Tube Meeting April 11, 1991 (9:00 12 noon)
Attendance Sheet Name Timothy G. Colburn Wm. Pegg C.E. Carpenter Alberto Verteramo Doug Malin Steve Brewer Daniel Merkovsky Karl J. Toth Richard B. Bennett Thomas A. Georgantis Christine M. Vertes John C. Hoebel Geoff Hornseth Eve Fotopoulos Shou-nien Hou Giuliano DeGrassi Tad Marsh David Terao Or 'anization NRC/NRR/PD3 1 NRC/NRR/PD3 1 NRC/NRR/PD3 1 X&M-D.C. Cook AEPSC AEPSC Westinghouse AEPSC AEPSC AEPSC Westinghouse Westinghouse NRC/NRR/
Mats.& Chemistry SERCH Licensing/
Bechtel NRC/NRR/Mech.
Eng.
BNL/NRC NRC/NRR/PD3 1 NRC/NRR/EMEB Phone (301) 492-1341 (301) 492-1362 (301) 492-1347 (616) 465-5901 (614) 223-2050 (614) 223-2020 (412) 256-6344 (614) 223-2038 (614) 223-2065 (614) 223-2059 (412) 374-5683 (412) 374-3355 (301) 492-3303 (301) 417-3094 (301) 492-0793 (516) 282-2949 (301) 492-1340 (301) 492-3317
Attachment 2
Licensee Presentation Slides
Indiana Michigan Power Company Donald C. Cook Nuclear Plant Thimble Tube Wear Status Report Presentation to the U. S. Nuclear Regulatory Commission Roekville, Maryland April 13, 1991 T. A. Georgantis R. B. Bennett
AGENDA Introduction Cook Nuclear Plant Experience Root Cause Evaluation Wear Mitigation Plan Summary
Cook Nuclear Plant Thimble Tube Ex erience Date Event October 18, 1983 Unit 1 Thimble Tube J-1 failed.
Thimble tube isolated.
Replaced during next refueling outage.
April 1985
.First industry attempt to eddy current thimble tubes unsuccessful.
Date April/May 1988 June 1988 Event Eddy current inspection of Unit 2 thimble tubes.
Large amount of wear found.
Decision made to replace all thimble tubes in both units.
October 24, 1988 Unit 1 Thimble Tube J-1 failed again.
Tube isolated.
Date Spring 1989 Event All Unit 2 thimble tubes replaced.
0.313" diameter tubes utilized.
Westinghouse upgraded product.
Summer 1989 All Unit 1 thimble tubes replaced.
0.313" diameter tubes also utilized.
N IIC4 ~ ~ C ~ CH T
Date June 18, 1990 Event Unit 2 Thimble Tube C-7 failed.
Tube isolated.
SummerlFall 1990 Unit 2 eddy current inspection revealed accelerated thimble tube wear.
10 tubes replaced, 19 tubes repositioned.
C-7 and A-9 shipped to hot cell facility for root cause determination.
Date Fall 1990 Event Unit 1 eddy current inspection also revealed accelerated thimble tube wear.
29 thimble tubes repositioned.
March 1991 Unit 2 Thimble Tubes C-7, A-9, and two others isolated.
Wear Summar Wear gp 60% to 90%
30% to 60%
30%
No Indications Unit 2 C~cle 6
0 Unit 2 C cle 7
19 25 Unit C cle 0
26 27 Corrective Action:
REPLACE ALL REPL>60%(10)
REPL)66%(0)
REPOS)30%(19)
REPOS>30%(29)
NIIC< 0 6H.C NT
Root Cause Evaluation Wear at lower core plate area seen by eddy current examination Flow induced vibration confirmed as root cause by hot cell examination
Flow Geometry
Hot cell metallurgical examination Surface examination Metallographic examination Detailed wear morphology Hardness measurements
Axial Wear Profile Fuel Bot tom Nozzle Thimble Tube Both Materials Stainless Steel
Radial Wear Profile d ~ 0.448" NOZZLE 240 d
0.210" d
M 0.313"
Other Results Material met original specifications ow impact fretting wear indicated Circumferential wear scars Lack of cold workideformation Flaking seen near wear throu h
Wear Mitigation Plans Corrective Action During 1992 outages install a fix to reduce wear rate
Survey of Industry Tubes "wear in" Wear dependent on (at least)
-Tube size Lower internals Flow rate Fuel type Several plants have one or two atypical tubes
Cook Wear Characteristics The Unit 2 wear is lower internal dependent The support penetrations are worse than cruciform penetrations This pattern is not seen in other units The Unit 1 wear not as pronounced as Unit 2 It is probably similar to other four loop 15x15 units The only other data is Zion 1
Current Status We are evaluating bids from several vendors Both flow reduction and/or wear resistant features are offered.
Outage contracts are required by June 15, 1991
Summary The problem is vibration Both thick and thin tubes wear Corrective action technologies are available:
Core plate modifications have no guarantee of success Most promising is a wear resistant coating This should greatly improve tube lifetime while minimizing the risk of increasing the vibrations 1992 Implementation
Attachment 3
Fhotograph of Wear Scar
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l UNITEDSTATES t
NUCLEAR REGULATORY COMMISSION WASHINGTON,D. C. 20555 April 1, 1991 MEMORANDUM FOR:
Russ L. Bywater, Jr.,
Reactor Engineer Division of Reactor Projects Region III FROM:
Timothy G. Colburn, Sr. Project Manager Project Directorate III-1 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation SUBJECT'RR OPERATING'REACTOR ASSESSMENT INPUT FOR CHAIRMAN CARR'S VISIT TO D.
C.
COOK NUCLEAR PLANT, UNITS 1
AND 2 f
The operating performance of the D.
C.
Cook Nuclear Plant has shown some improvement over the past year.
A generally conservative operating philosophy, above-average performance indicators, and a robust, forgiving plant design have resulted in a Category 1 rating during the most recent SALP.
Equipment problems which have plagued the plant include thimble tube vibration-induced fretting.
A 4KV breaker explosion last July killed one worker and injured three others.
Boron precipitation and resultant loss of flow in one-of-three (two-of-three required) boric acid injection paths has been a recent problem and a 30-gallon-
'er-minute leak in a Unit 1/Unit 2 charging line crosstie valve has resulted in the plant being placed in a 60-day limiting condition for operation (LCO).
The licensee has requested a Temporary Waiver of Compliance from Region III in order to use a freeze seal in high pressure piping to allow repair of this valve at power.
The licensee's Security, Fitness-for-Duty and Emergency Preparedness Programs are considered strong.
Additionally, they are considered to possess significant corporate resources and exhibit a visible management presence on-site.
The Vice President-Nuclear, Mr. Milton P. Alexich, retired effective April 1, 1991, and has been replaced by Mr. Gene Fitzpatrick.
The licensee has an impressive training facility and on-site simulator and uses them judiciously.
Although the material condition of the plant is considered improving, as evidenced by reclamation of 20,000 square feet of contaminated area since the Unit 1 steam generator replacement project w'as completed, progress is slow.
Numerous green poly "catch-basins" have been installed beneath valve assemblies to trap potential or actual leaks and direct them to appropriate sumps.
Weak communications and interface between the corporate and plant staffs, narrow root cause
- analyses, and the decision to defer design basis reconstitution beyond their scheduled individual plant examination
( IPE) completion date have contributed to a Category 3 in Engineering and Technical Support.
The licensee is just beginning to initiate a systems engineer program, which while still in its infancy, holds promise to help resolve the above concerns and is also responsive to staff concerns from Maintenance Team and Safety System Functional Inspections conducted dur ing the last year.
Contact:
T. Colburn X21341
Russ L. Bywater, Jr.
The licensee's schedule for maintenance improvement has been deliberately slow.
Past maintenance practices focused on corrective, rather than preventive maintenance, and were of insufficient quality to prevent 25 percent rework being required.
- However, programs have been initiated, which when fully implemented, should help address staff concerns.
The licensee has had a long and checkered history with Appendix R compliance including material false statements and other serious violations dating back to 1982.
The licensee was recently issued a Notice of Violation and civil penalty totaling
$ 150,000 for Appendix R violations found in September 1990.
Recent modifications to the plant include steam generator replacement on Unit I and sleeved steam generator tubes on Unit 2.
The licensee has several significant licensing actions submitted for staff review.
These include a request for license extension to recoup lost generating years during the licensing process, a significant modification to the Section 6.0 administrative technical specifi-
- cations, and the temporary waiver of compliance mentioned previously.
The licensee is conducting a Level III probabilistic risk assessment (PRA) for internal and external events in response to Generic Letter 88-20 for development of an IPE for severe accident vulnerabi lities.
The licensee plans completion by December 1991.
cc:
J. Zwolinski B. Clayton Timothy G. Colburn, Sr. Project Manager Project Directorate III-I Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Russ L. Bywater, Jr.
The licensee's schedule for maintenance improvement has been deliberately slow.
Past maintenance practices focused on corrective, rather than preventive maintenance, and were of insufficient quality to prevent 25 percent rework being required.
- However, programs have been initiated, which when fully implemented, should help address staff concerns.
The licensee has had a long and checkered history with Appendix,R compliance including material false statements and other serious violations dating back to 1982.
The licensee was recently issued a Notice of Violation and civil penalty totaling
$ 150,000 for Appendix R violations found in September 1990.
Recent modifications to the plant include steam generator replacement on Unit I and sleeved steam generator tubes on Unit 2.
The licensee has several significant licensing actions submitted for staff review.
These include a request for license extension to recoup lost generating years during the licensing process, a significant modification to the Section 6.0 administrative technical specifi-
- cations, and the temporary waiver of compliance mentioned previously., The licensee is conducting a Level III probabi listic risk assessment (PRA) for internal and external events in response to Generic Letter 88-20 for development of an IPE for severe accident vulnerabilities.
The licensee plans completion by December 1991.
DISTRIBUTION:
PDIII-I r/f PShuttleworth TColburn
- g. ~~(i~hi B. Clayton Timothy G. Colburn, Sr. Project t1anager'roject Directorate III-I Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation D
NAME,:
PSh ttl orth DATE: Rll I~l
- TColburn/tg
- LBMarsh Document Name:
NRR OPERATING REACTOR
Russ L. Bywater, Jr. The licensee's schedule for maintenance improvement has been deliberately slow.
Past maintenance practices focused on corrective, rather than preventive maintenance, and were of insufficient quality to prevent 25 percent rework being required.
- However, programs have been initiated, which when fully implemented, should help address staff concerns.
The licensee has had a long and checkered history with Appendix R compliance including material false statements and other serious violations dating back to 1982.
The licensee was recently issued a Notice of Violation and civil penalty totaling
$ 150,000 for Appendix R violations found in September 1990.
Recent modifications to the plant include steam generator replacement on Unit 1 and sleeved steam generator tubes on Unit 2.
The licensee has several significant licensing actions submitted for staff review.
These include a request for license extension to recoup lost generating years during the licensing process, a significant modification to the Section 6.0 administrative technical specifi-
- cations, and the temporary waiver of compliance mentioned previously.
The licensee is conducting a Level III probabi listic risk assessment
( PRA) for internal and external events in response to Generic Letter 88-20 for development of an IPE for severe accident vulnerabi lities.
The licensee plans completion by December 1991.
DISTRIBUTION:
Timothy G. Colburn, Sr. Project t1anager Project Directorate III-1 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation PDI II-I r/f PShuttleworth TColburn J, ~4()~le I B. Clayton D
NANE PS ttl orth
- TColburn/tg DATE: 9/l IRl: 4(i(A/
- LBHarsh Document Name:
NRR OPERATING REACTOR
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UNITEDSTATES t
NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 March 29, 1991 Docket Nos. 50-315 and 50-316 MEMORANDUM FOR:
L. B. Marsh, Director Project Directorate III-1 Division of Reactor Projects III/IV/V FROM:
SUBJECT:
DATE 5 TIME:
LOCATION:
PURPOSE:
Timothy Colburn,'roject Manager Project Directorate III-1 Division of Reactor Projects III/IV/V FORTHCOMING MEETING WITH DONALD C.
COOK PLANT April 11, 1991 9:00 a.m. - 12:00 p.m.
U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockvi 1 le, Maryland 20852 Room 88-11 DISCUSS ROOT CAUSE OF D.C.
COOK THIMBLE TUBE WEAR AND PLANS TO MITIGATE THIMBLE TUBE CONCERNS (AGENDA ATTACHED)
- PARTICIPANTS:
NRC T. Colburn L. Marsh W. Pegg S.
Hou R.
Hermann
~Ut 1 1 it S. Brewer R. Bennett T. Georgantis D. Malin A. Verteramo C. Vertes D. Mervkowski cc:
See next page Timothy Colburn, Project Manager Project Directorate III-1 Division of Reactor Projects III/IV/V
- Meetings between NRC technical staff and applicants or licensees are open for interested members of the public, petitioners, intervenors, or other parties to attend as observers pursuant to "Open Meeting Statement of NRC Staff Policy,"
43 Federel Reqister 28058, 6/28/78.
NRC RK @55II<R ~~PV
DISTRIBUTION J
L Docket File~
HRC a Local PDRs PD31 Reading T. Nurley/F. tliraglia J. Partlow B. Boger J. Zwolinski T. Colburn W. Pegg A. Chaffee, EAB OGC E. Jordan B. Grimes Receptionist (White Flint)
NRC Participants ACRS (10)
GPA/PA E.
Tana Y. Overton L. Plisco R. Lobel,'(Region III Plants)*
J. Strasma, Region III**
P. Shuttleworth cc:
Licensee/Applicant Im Service List
Mr. Milton Alexich Indiana Michigan Power Company Donald C.
Cook Nuclear Plant CC:
Regional Administrator, Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 Attorney General Department of Attorney General, 525 h'est Ottawa Street Lansing, Michigan 48913 Township Supervisor Lake Township Hall Post Office Box 818 Bridgman, Michigan 49106 Al Blind, 'Plant Manager Donald C.
Cook Nuclear Plant Post Office Box 458 Bridgman, Michigan 49106 U.S. Nuclear Regulatory Commission Resident Inspectors Office.
7700 Red Arrow Highway Stevensville, Michigan 49127 Gerald Charnoff, Esquire Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.LJ.
lJashington, DC 20037 Mayor, City of Bridgman Post Office Box 366 Bridgman, Michigan 49106 Special Assistant to the Governor Room 1 - State Capitol Lansing, Michigan 48909 Nuclear Facilities and Environmental Monitoring Section Office Division of Radiological Health Department of Public Health 3500 N. Logan Street Post Office Box 30035 Lansing, Michigan 48909 Mr. S.
Brewer American Electr ic Power Service Corporation 1 Riverside Plaza
- Columbus, Ohio 43216
V
APRIL 11, 1991 MEETING DONALD C.
COOK NUCLEAR PLANT THIMBLE TUBES AGENDA INTRODUCTION
" HISTORY OF THIMBLE TUBES THIMBLE TUBE PROBLEM ROOT CAUSE ANALYSIS PLANS TO MITIGATE THIMBLE TUBE WEAR CONCERNS
MAR 2 6 egg~
MEMORANDUM FOR:
FROM:
SUBJECT:
Hubert J. Miller, Director Division of Reactor Projects, Region III Thomas M. Novak, Director Division of Safety Programs Office for Analysis and Evaluation of Operational Data AEOD INPUT FOR THE VISIT OF CHAIRMAN CARR TO THE D.
C.
COOK SITE ON APRIL 10, 1991 Per your staff's request, enclosed is the AEOD input to the briefing package for Chairman Carr's visit.
Our input consists of a summary analysis of operational experience at D.
C.
Cook Units 1 and 2 from January 1,
- 1990, using various data sources (e.g.,
10 CFR 50.72 and 50.73 reports).
If you need any additional information, please contact Bennett M. Brady (FTS 492-4499) or John Crooks (FTS 492-4425) of my staff.
ppiginal Signed bY:
Thomas M. Novak, Director Division of Safety Programs Office for Analysis and Evaluation of Operational Data
Enclosure:
As stated cc: w/enclosure M. Ring, RIII R. Lobel, DEDRO DISTRIBUTION
'PAB RF TNovak DSP RF JRosenthal EJordan JCrooks DRoss MHarper VBenaroya FManning WJones BBrady
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Review of Operational Experience for Chairman Carr's Visit to the D.
C.
Cook Site on April 10, 1991 Division of Safety Programs Office for Analysis and Evaluation of Operational Data
Summary of Operational Experience at D. C.
Cook Since January 1,
1990 I.
Overview Using the AEOD screening criteria, AEOD identified two LERs as important events at D.
C.
Cook Units 1 and 2 during the review period.
No event at either unit was identified as a Significant Event for the Performance Indicator (PI) Program.
II.
Abnormal Occurrences For the period beginning January 1,
1990,.there have been no Abnormal Occurrences and no Appendix C items ("Other Events of Interest" ) reported for the D. C.
Cook site.
III.
Other 0 erational Data LER Reviews Using its screening criteria, AEOD's review of the D. C.
Cook site operating experience identified two events that concerned both units as events of particular importance for the period from January 1,
- 1990, through mid March 1991.
(The last LER reviewed for Unit 1 was LER 91-002 with an event date of January 1,
1991; the last LER reviewed for Unit 2 was LER 91-001 with an event date of January 7,
1991.)
Both of the important events involved equipment design errors.
The events are summarized below:
LER 315/90-008 LER 315/90-010 "10 CFR [50] Appendix R Deficiencies Resulting in Potential for Loss of Auto Start of Service Water Pumps Due to Incorrect Implementations of Design Change."
The isolation circuitry for the low pressure auto start switch for the Essential Service Water pumps was found to be installed incorrectly.
Fuses of the same size were installed for both this start circuit and other upstream circuits.
Since the fuses weren't size coordinated to blow out sequentially, they could, if a pressure switch short developed, blow out simultaneously defeating the auto start capability for the Essential Service Water pumps.
The same condition was found for the isolation relay circuit for the low header pressure auto start switch for the Component Cooling Water pumps, but this condition would not render those pumps inoperable.
The cause of these problems was equipment design errors.
(Event Date:
June 6,
1990.)
"10 CFR 50 Appendix R Deficiencies Resulting in Potential Loss of Local Shutdown Indication Panel Function."
The routing of cables for the Unit 1 and Unit 2 Local Shutdown Indication panels was not in compliance with 10CFR50, Appendix R.
It, was
discovered that a fire in any one of several different areas through which the cables feeding these panels pass could have resulted in loss of power to both panels.
The cause of these problems was equipment installation design errors.
(Event date:
August 24, 1990.)
Performance Indicator Pro ram Reviews The Performance Indicator Program did not identify any events as Significant Events from January 1,
- 1990, through December 31, 1990.
IV.
Performance Indicator Data Performance Indicator (PI) data extending through December 1990 are attached.
Note:
o The PI safety system actuations are a specific subset of all ESF actuations
- emergency core cooling system actuations and emergency power (diesel generator) actuations in response to a
dead bus.
FIGURE 4.22 CQOI<
89-1 to 90-4 Hbcotor end'rder Plant 6 rttr Movng Average
~---.
Critical HOurS 6 Ouorter Moving Average (long term trends) 0 2
X$)gl I. Automalfc Scrams While Critical 89-1 S9 2 89 5 89 4 90-1 90-2 90-3 90-4 Year - Ouorler 2500 ~
2250 9 mo~
u50 0 1500 o50 5 eo*
250 500 c0 250 0
.s 0
O +2
- 2. Safely System Acluations I
I 89-1 89-2 69 S 89 4 90 1 90-2 90-3 90 4
Year - Ouorler
~xo I
IFA 2
- erp, usA 5 kxo P
.V QH 250 0
- 3. Slgttiticant Events
- 4. Safely Sysi.em Failures 4 g'5 wb 0
p-05 d8
~ 29 Og 2 ors:
I I
~
I 89-1 Ss-2 89-5 89 4 90-1 90-2 90 3 90-4 Yeor - Ouorter
- 5. Forced Oulagc Rate (x) 89-1 69-2 69-5 69 4 90-1 90-2 90-5 90-4 Year - Ouorter
- 6. Equipmcnt F'orccd Outages/
1000 Commercial Critical Hour
-"P SO rr. ~
1P 60 8
~ 0 20 8y u p4 r
I 69-1 89-2 69-5 89-4 90 1 90-2 90 5 90-4 Year - Quarter I
I I
89 1 89-2 69-3 69 4 90 1 90 2 90 5 90<<4 Year - Ouoeter 200 fe eo yF go 4J
- 7. Radiation Exposure
'I/
sot I5 89-1 89-2 89-5 89-4 90-1 90-2 90 3 90-4 Year - Ouorter 2500 2250 5 2000 u50 ~0 500 vr Q50 5 eo ~
250 v
500 250 0
hiAINT DESIGN E UIP VAIL 6.
Long Term Cause Code Trends Att LXlt Cause Codes Through 00-8 ADhilN LIC OPER OTHER I'ER
l
FIGURE 4.22 COOK 1
Per forrnonce Indicators Short Term Trends Declined Improved Deviotions from Older Plont Long Term Meons Below Avg. Above Avg.
- l. Automolic Scroms While Criticol 1.5
- 2. Safety System Acluotions
- 3. Significont Evenls 08
- 4. Sofely System Foifures 0.3 SSr Cornrorers to pwrt srecns
- 5. Forced Outoge Rote 0.4
- 6. Eouipment Forced Outoges
/1000 Commercial Critlcol Hour t2
- 7. Couse Codes (Aff LERs) 7o. Acsvinstrotne Control Procrem 7O. Lkensed Operator P corrm 7e. Otner Personnel Oror 7rS Scsntenrnee Prccwn 7e. Oes'~rrtnstcoot'<<rnrrrcoAoten Protrem 7l. Ccripfnere rotrre as
-2.5
-1.5
-0.5 0.5 1.5 Ocviotions tforn Older Plant t.o~
Term Meters (Meaecred rr Standred Oev'fetmS)
(6 Qtr. Avg end 90-4) 2.5
-2 5 -15>>05 05 1.5 2 5 Oev'etre from Ccrrent 6 Otr. Plant Means (Meosrred
'et Stondord Oevtottons)
(2 Qlr. Avg end 90-4)
~
HOlE: Cause Code Avgs end 90-3 53
FIGURE 4.23 COOK 2
89-1 to 90-4 tnbedtOr Le end:
Ohter P~t 6 qtr Mov'%vs ";,;
~----
C11ticot Hours 6 Quarter MOInng Averege (tong Term TrenCs)
'r 0
- l. Automatic Scrams While Critical
%c r
I 89 1 89 2 89 5 89-a 90-1 90-2 90-5 90-a Yeor - Ovorter 2500 225O 5 2000 I 1750 O
1500 Q50 h 1XO
- 75O vo 500
'00 250 0
a
.O
>F5 II O y2
- 2. Salety System Actuatlons r/
/
I I
89-1 89 2 89-5 89-1 90-'1 90"2 90-5 90-a Yeor - Ovorter II 1 50 ew *
&to 0
III aI C IS 4I sr'
- 3. Significant Events
- 4. Safety System Failures a
r as 5 bF O +
aI rL I
89 1 89-2 89 5 89-a 90 1 90-2 90 5 90 4
Yeor - Ovorter I
89-1 89-2 89-) 89
~ 90 1 90 2 90-5 90 Year - Ovorter 80 S. Forced Outage Rate (s)
F 8
ms
- 8. Equipmcnt forced Outages/
1000 Commercial Critical %lour 17 89-'1 89-2 89-5 89- ~ 90-1 90 2 90-5 90-a Yeor Ovor ter 1.71 II V
O.as I
I I
I I
I 89-1 89-2 89 5 89 4 90-'1 90-2 90 5 90-a Yeor - Qvorter
- 7. Radiation Exposure 2500
- 8. Long Term Cause Code Trend all tN Cause Codes Throueh 90 8
~F 100 50 VI NA 89-1 89 2 89 5 89-4 90 1 90-2 90 5 90-4 Yeor Qvorter 2250 5 mo~
1750 O
5CO al C50 1000 ~
75O g
500 250 0
ADMIN hiAINT LIC OPER OTHER PER DESIGN E UlP FAIL 54
FIGURE 4.23
'OOK 2
Per formonce indicators Short Term Trends Declirted Improved Deviotions from Older Plant Long Term Meons
'elow Avg. Above Avg.
- l. Automotic Scroms White Criticol
-0.4
-1,3
- 2. Solety System Actuotions 0.4 0,4
- 3. Signilicont Events 0.4
- 4. Sofety System Foilures 0.7 0.6 ssr ccmpcscd to prrtt Mere
- 5. Forced Outoge Rote 0.1
- 6. Equipment Forced Outoges r/10~00 Cornmerciol Crittcol Hour
-0.7
- 7. Couse Codes
{All LERs) ro, scsrfnstrottve Contrcr Proerem rtr. Lecrecd operator protrem rc. Otncr pe scrrec Error rd, tres'ntensnce Prccrem re. Erescrr/4stosot'em/rcortcot<<rn protrcrn Tf. EcSr'omens ro&e ar as ar or ar
-2,5 - I.S
-0.5 0.5 LS 2.5 Oe~ottons from Odcr Ptont L~
Term Sseons tkceoscred Ln StoNford Oe crtens)
(6 Otr. Avg end cC-4)
-2.5 1.5 0.5 0.5 1.5 2.5 Devet<<ms from Ccrrent 6 Otr. Ptont Ltcons (sccosceed ss Stonckrd Oc etens)
(2 Qtr. Avg end 90-4)
NOTE: Corse Code Avgs end 90-3
TABLE 8 ~ 22 COOK 1 PX EVENTS FOR 90-1 NONE
, PX EVENTS FOR 90-2 POWER:
83 SSF 06/19/90 LERB 31590008 50.72M:
GROUP:
ESSENTIAL SERVICE llATER SYSTEH GROUP SYSTEH:
ESSEHTIAL SERVICE WATER SYSTEH OESC:
BECAUSE OF A DESIGN ERROR, THE ESW PUHPS DID NOT HEEL APPENDIX R SEPERATION CRITERIA. THE DESIGN CHANGES HAD BEEN PROPERLY PREPARED, BUT TNE CHANCES WERE NOT PROPERLY IHPLEHENTED OM TNE DESIGM DRAWINGS.
SSF PWR HlST:
GROUP SYSTEH DESC PZ EVENTS FOR 90 3
NONE PX EVENTS FOR 90-4 11/02/90 LERII 31590015 50.72II:
19771 EVEHT DISCOVERED DURING COLO SHUTDOWN.
CONTAINHEHT AND COHTAINHENT ISOLATION GROUP REACTOR CONlAINMEHT BUILDING LLRT LIHITS FOR TYPE B AND C CONTAINHEHT LEAKAGE WERE EXCEEDEO DUE To DFGRADATIOM OF ISOLATION VALVE SEATING SURFACES+
TYPE 89.1 89-2 89-3 89-4 90.1 90-2 90-3 90-4 SCRAHS > 15X POWER/1000 CRITICAL HOURS SCRAHS <-"15X POWER TOTAL SCRAHS SAFElY SYSTEH ACTUATIOMS SIGNIFICAHT EVEHTS SAFETY SYSTEH FAILURES fORCED OUTAGE RATE (X)
EQUIPS FORCED OUTAGES/1000 COHHERCIAL NRS CRITICAL HOURS COLLECT IVE RADIATIOH EXPOSURE CAUSE CCOES:
ADHINISTRAT IVE I.ICEMSED OPERATOR OTHER PERSONNEL HAIHTENANCE DESIGN/INSTALLATION/FABRICATION EQUIPHENT FAILURE 0.55 0.00 1
0 2
0 0
0 0
0 0
0 1
0 0.00 0.00 1810 0
95 138 2
3 I
0 1
2 4
5 2
0 0
0 0.00 0.00 0
0 0
0 0
0 0
0 1
0 4
0 0.00 0.00 2151 2209 10 10 3
1 0
1 3
0 4
1 0
0 0
0 O.OO O.OO 0
0 0
0 0
0 0
0 0
0 0
0.00 0.00 2097 2183 15 6
\\
2 0
1 1
0 1
1 1
0 0 0.00 0.00 0
0 0
0 0
0 0
0 0
I 0
0 o.oo o.oo 2208 457 154 NA 2
NA 0
NA 1
MA 2
NA 4
NA 0
NA 46
TABLE 8'3 COOK 2 SSF GROUP SYSTEH DESC SSA DESC PI EVENTS FOR 90-1 01/10/90 LER¹ 31690002 50.72¹:
17530 POMER:
0
- MAIM STEAN ISOLATIOH VALVES GROUP
- HAIH STEAN ISOLATIOH VALVES
- ALL FOUR HAIN STEAM ISOLATION VALVES HAY HAVE BEEN INOPERABLE DURING PLANT POMER OPERATIOM.
EXCESSIVE CONDENSATE ACCUHULATIOH OH THE VENT SIDE Of THE HSIV OPERATING PISTON CAUSED THE VALVE CLOSING TIMES TO EXCEED THE TECHHICAL SPECIFICATIOM LIMIT.
01/12/90 LER¹ 31690001 50.72¹:
17535 POMER:
0
- A TECHNICIAM 'MAS PERFORHIHG A TINE DELAY RELAY CALIBRATIOH MHEH A LEAD THAT MAS LIFTED (PER
'ROCEDURE)
CANE IM COHTACT MITH A DIFFEREHT TERMINAL, CAUSIHG A LOSS OF EMERGENCY BUS AND DIESEL START.
PI EVENTS FOR 90-2 SCRAM 06/11/90 LER¹ 31690004 50.72¹:
18681 POMER:
85 OESC:
ROO 'H8'S SUSPECTED TO HAVE DROPPEO CAUSING A HIGH NEGATIVE FLUX RATE REACTOR SCRAM.
THE EXACI'AUSE IS UNKNOMM.
SSF GROUP SYSTEM DESC POMER:
90 06/19/90 LER¹ 31590008 50.72¹:
ESSEHTIAL SERVICE MATER SYSTEM GROUP ESSENTIAL SERVICE
'MATER SYSTEM BFCAUSE OF A DESIGH ERROR, THE ESM PUMPS OIO NOT MEET APPENDIX R SEPERATIOM CRITERIA. THE DESIGN CHANGES HAO BEEN PROPERLY
- PREPARED, BUT THE CHANGES MERE HOT PROPERLY IMPLEMENTED ON THE DESIGN DRAMINGS.
PZ EVENTS FOR 90-3 NONE PZ EVENTS FOR 90-4 SCRAM 12/12/90 LER¹ 31690012 50.72¹:
20048 PMR HIST: PIKER OPERATIONS AT 100X OESC:
REACTOR TRIPPEO ON LOM SG LEVEL FOLLOMING RECEIPT OF A SPURIOUS THRUS'I BEARING TRIP ALARM ASSOCIATED MITH THE NEST HAIH FEEDMATER PUMP.
SCRAM 12/15/90 LER¹ 31690013 50.72¹:
20083 PMR HIST: REDUCING POMER AT 35X OESC:
REACTOR TRIPPED FROM 35X OH A TURBIHE TRIP DUE TO A HlsADJUSTED ATMS TRIP SYSTEM SET POINT.
TYPE 89-1 89.2 89.3 89.4 90-1 90.2 90.3 90.4 SCRAHS
> 15X POMER/1000 CRITICAL HOURS SCRAHS <= 15X POMER TOTAL SCRAHS SAFETY SYSTEM ACTUATIOHS SIGNIFICANT EVENTS SAFETY SYSTEM FAILURES FORCED OUTAGE RATE (X)
EOUIP.
FORCED OUTAGES/1000 CCHHERCIAL HRS CRITICAL HOURS COLLECTIVE RADIATION EXPOSURE CAUSE CODES:
ADMIN I STRAT IVE LICCNSED OPERATOR OTHER PERSONHEL HAIHTEHAHCE DESIGN/INSTALLATION/FABRICATION EQUIPMENT FAILURE 0.00 0.00 0
0 0
0 0
0 0
0 0
0 0,
0 D.OD 0.00 395 1863 95 138 2
2 1
2 3
2 8
5 2
0 0
0 0.47 0.00 0
0 1
0 0
0 1
0 0
0 5
0 0.47 0.00 2114 2209 10 10 2
1 0
0 1
0 2
2 0
1 0
0.00 0.48 0
0 0
1 1
0 0
1 1
17 3
0.00 0.48 1695 2093 15 6
1 1
0 0
2 1
2 1
2 2
0 0.00 1.71 0
0 0
2 0
0 0
0 0
0 0
13 0.00 1.71 0
1170 154 MA 1
NA 0
MA 1
MA 2
'NA 4
NA 0
NA 47