ML20054A305
Text
.4' 3)st h Gentlemen:
The enclosed Information Notices; No. 82-09, " Cracking in Piping of Makeup Coolant Lines at B&W Plants," and No. 82-10, "Following up Symptomatic Repairs to Assure Resolution of the Problem," are forwarded to you for information.
No written response is required.
If you desire additional information regarding these matters, please contact this office.
Sincerely,
~~ kri QM~
nald C. Haynes
[LRFgionalAdministrator
Enclosures:
IE Information Notice No. 82-09 with 1 Attachment IE Information Notice No. 82-10 with 1 Attachment CONTACT:
S. D. Ebneter (215) 337-5283 l
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8204150365 820331 PDR ADOCK 05000003 PDR O
LIST OF HOLDERS OF NUCLEAR POWER REACTOR OPERATING LICENSES AND CONSTRUCTION PERMITS RECEIVING IE INFORMATION NOTICES 82-09 AND 82-10 Baltimore Gas and Electric Company Docket Nos. 50-317 ATTN: Mr. A. E. Lundvall, Jr.
50-318 Vice President, Supply P. O. Box 1475 Baltimore, Maryland 21203 Boston Edison Company M/C Nuclear Docket No. 50-293 ATTN:
Mr. William D. Harrington Senior Vice President, Nuclear 800 Boylston Street j
Boston, Massachusetts 02199 Connecticut Yankee Atomic Power Company Docket No. 50-213 ATTN: Mr. W. G. Counsil Senior Vice President - Nuclear Engineering and Operations P. O. Box 270 Hartford, Connecticut 06101 Consolidated Edison Company of Docket Nos. 50-03 New York, Inc.
50-247 ATTN:
Mr. John D. O'Toole Vice President - Nuclear Engineering and Quality Assurance 4 Irving Place New York, New York 10003 Duquesne Light Company Docket No. 50-334 ATTN:
Mr. J. J. Carey Vice President Nuclear Division P. O. Box 4 Shippingport, Pennsylvania 15077 GPU Nuclear Corporation Docket No. 50-219 ATTN:
Mr. P. B. Fiedler Vice President and Director Oyster Creek Nuclear Generating Station P. O. Box 388 Forked River, NJ 08731 4
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2 Maine Yankee Atomic Power Company Docket No. 50-309 ATTN:
Mr. John H. Garrity Senior Director Nuclear Engineering and Licensing 83 Edison Drive Augusta, Maine 04336 GPU Nuclear Corporation Docket No. 50-289 ATTN: Mr. H. D. Hukill Vice President and Director of TMI-1 P. O. Box 480 Middletown, Pennsylvania 17057 GPU Nuclear Corporation Docket No. 50-320 ATTN:
Mr. J. J. Barton Acting Vice President and Director of TMI-2 P. O. Box 480 Middletown, Pennsylvania 17057 Niagara Mohawk Power Corporation Docket No. 50-220 ATTN:
Mr. T. E. Lempges Vice President Nuclear Generation 300 Erie Boulevard West Syracuse, New York 13202 Northeast Nuclear Energy Company Docket Nos. 50-336 ATTN:
Mr. W. G. Counsil 50-245 Senior Vice President - Nuclear 50-423 Engineering and Operations Group P. O. Box 270 Hartford, Connecticut 06101 Philadelphia Electric Company Docket Nos. 50-277 ATTN:
Mr. S. L. Daltrof f 50-278 Vice President Electric Production 2301 Market Street Philadelphia, Pennsylvania 19101 Power Authority of the State of New York Docket No. 50-286 Indian Point 3 Nuclear Power Plant ATTN:
Mr. J. C. Brons Resident Manager P. O. Box 215 Buchanan, New York 10511 i
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3 Power Authority of the State of New York Docket No. 50-333 James A. FitzPatrick Nuclear Power Plant ATTN:
Mr. Corbin A. McNeill, Jr.
Resident Manager P. O. Box 41 Lycoming, New York 13093 Public Service Electric and Gas Company Docket Nos. 50-272 ATTN:
Mr. Richard A. Uderitz 50-311 Vice President - Nuclear Mail Code T15A P. O. Box 570 Newark, New Jersey 07101 Rochester Gas and Electric Corporation Docket No. 50-244 ATTN:
Mr. John E. Maier Vice President Electric and Steam Production 89 East Avenue Rochester, New York 14649 Vermont Yankee Nuclear Power Corporation Docket No. 50-271 ATTN: Mr. Robert L. Smith Licensing Engineer 1671 Worcester Road j
Framingham, Massachusetts 01701 Yankee Atomic Electric Company Docket No. 50-29 ATTN:
Mr. James A. Kay Senior Engineer-Licensing 1671 Worcester Road Framingham, Massachusetts 01701 l
Duquesne Light Company Docket No. 50-412 ATTN:
Mr. E. J. Woolever Vice President Nuclear Construction Robinson Plaza Building No. 2 Suite #210, PA Route 60 Pittsburgh, Pennsylvania 15205 i
Jersey Central Power & Light Company Docket No. 50-363 ATTN:
Mr. J. T. Carroll Acting Director Oyster Creek Nuclear Generating Station P. O. Box 388 Forked River, New Jersey 08731 p--,---
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4 Long Island Lighting Company Docket No. 50-322 ATTN:
Mr. M. S. Pollock Vice President - Nuclear 175 East Old Country Road Hicksville, New York 11801 Niagara Mohawk Power Corporation Docket"No. 50-410 ATTN: Mr. Gerald K. Rhode Vice President System Project Management c/o Miss Catherine R. Seibert 300 Erie Boulevard, West l
Syracuse, NY 13202 Pennsylvania Power & Light Company Docket Nos. 50-387 ATTN:
Mr. Norman W. Curtis 50-388 Vice President Engineering and Construction - Nuclear 2 North Ninth Street Allentown, Pennsylvania 18101 Philadelphia Electric Company Docket Nos. 50-352 ATTN:
Mr. John S. Kemper 50-353 Vice President Engineering and Research 2301 Market Street Philadelphia, Pennsylvania 19101 Public Service Electric & Gas Company Docket Nos. 50-354 ATTN:
Mr. T. J. Martin 50-355 Vice President Engineering and Construction 80 Park Plaza - 17C Newark, New Jersey 07101 Public Service Company of New Hampshire Docket Nos. 50-4'43 ATTN: Mr. W. C. Tallman 50-444 Chairman and Chief Executive Officer 1000 Elm Street Manchester, New Hampshire 03105
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DESIGNATED ORIGIliAL I
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NUCLEAR REGULATORY COMMISSION 0FFICE OF INSPECTION AND ENFORCEMENT
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WASHINGTON, D.C.
20555 March 31, 1982 IE INFORMATION NOTICE N0. 82-09:
CRACKING IN PIPING OF MAKEUP COOLANT LINES AT B&W PLANTS Description of Circumstances:
On January 21, 1982, Crystal River Unit 3 commenced shutdown to investigate an unidentified 0.9-gpm primary leak.
During power reduction the leak rate increased to about 1.0 gpm and the plant proceeded to hot standby conditions.
A visual inspection inside the reactor building at this time revealed the leak was associated with a 2h-inch check valve (MOV-43) in the makeup line to the 26-inch reactor coolant (RC) loop A inlet line.
This line is used for normal makeup of reactor coolant but is also part of the redundant high-pressure injec-tion system. After the insulation was removed from the affected valve a 1400 circumferential crack in the check valve body near the valve-to-safe end weld (i.e., valve end toward RC inlet nozzle) was found.
The leak was nonisolatable and the plant promptly proceeded to cold shutdown conditions in accordance with plant technical specifications.
The check valve was removed and liquid penetrant testing (LPT) was performed on the accessible inside diameter (ID) surfaces including 5 inches into the 2h-inch line on the inlet side of the affected valve.
This inspection dis-closed an extensive network of heat-check type cracks around the safe end ID surface. A similar condition was observed inside the valve body from the i
discharge side up to the disc seat area.
The valve inlet side and connecting i
piping were not affected.
The most severe cracking in the safe end appeared to have penetrated up to 25 percent of the wall thickness.
A visual inspection also revealed the thermal sleeve inside the high-pressure injection (HPI) nozzle was loose and showed evidence of wear in areas of contact.
Some cracking of the thermal sleeve was also observed.
l As a result of the Crystal River 3 findings, Duke Power Company initiated a radiographic examination of the RC inlet nozzle connections on the two HPI i
lines used for normal makeup at Oconee Unit 3 to determine the thermal sleeve conditions.
This examination disclosed that in one of the makeup nozzles the thermal sleeve was loose, the four thermal sleeve retaining button welds on the safe end side were missing, and the thermal sleeve was slightly displaced in the upstream direction of flow.
Action was tnen taken to remove the pipe extension to replace the affected themal sleeve.
Further findings and expanddd inspection as a result of this action are summarized below.
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IN 82-09 March 31, 1982 Page 2 of 3 Investigation and Findings:
A.
Crystal River A, metallurgical investigation of the affected valve body indicated two crack initiation sites.
One was inside on the valve body at a machine mark (i.e., weld counterbore area) and one was on the outside diameter (00) at the valve-to-weld transition (geometrical discontinuity).
The cracks progressed through the wall on a slightly different plane and merged about mid-wall of the valve body.
Scanning electron microscope examination of the fracture features disclosed the cracks propagated transgranularly and exhibited clearly defined grain structure striations characteristic of cyclic fatigue failure.
Cracks in the thermal sleeve and safe end sections exhibited similar fracture morphology.
No evidence of corrosion interaction from chemical attack was identified.
During the design phase, Babcock and Wilcox (B&W) performed the stress analysis on the primary system up to the affected check valve which is the design code (USAS B31.7-USAS B31.1) interface boundary.
Gilbert Associates, as architect-engineer, perfonned the balance of plant design.
The B&W design calculations for the HPI lines included a pipe section that was not installed during plant construction.
The potential thermal discontinuity at this point is believed to be partly responsible for the cracking and is currently being evaluated by both organizations.
Based on the above findings, the mode of cracking was tentatively attri-buted to thermal cycle fatigue.
However, the synergistic thermal-hydraulic effects contributing to the failure mechanism are yet to be determined.
Contributing factors being investigated include operational design limits and setpoints with regard to makeup water temperature and flow rate, minimum bypass flow, and system thermal-hydraulic parameters around the HPI ilozzle used for makeup.
B.
Oconee When the pipe extension at Oconee 3 was removed to gain access to the thermal sleeve in order to repair it, liquid penetrant testing (LPT) disclosed cracks on the ID surfaces of the makeup /HPI pipe extensior & 1 nozzle safe end.
Crack features were similar in nature to those fourd '
Crystal River.
Reportedly, the cracks penetrated up to 20 percent of the thickness of the pipe wall.
The other makeup nozzle assembly was examined by radiography and a special ultrasonic testing (UT) technique developed by B&W for this purpose.
tio indication of cracking or degraded thermal sleeve conditions was observed.
Further UT and radiographic testing (RT) of the two remaining HPI nozzle assemblies indicated a loose thermal sleeve in one of the nozzles (flozzle 381).
IN 82-09 March 31, 1982 Paga 3 of 3 At Oconee 2, results of the UT and RT indicate the thermal sleeve in one of the makeup nozzles may be loose and the retaining button welds on the safe end side are missing.
Cracking was also found in the safe end and pipe extension.
The other makeup nozzle showed no indications of a degraded thermal sleeve or cracking.
Examination of the two remaining HPI nozzle assemblies indicated a loose thermal sleeve (i.e., retaining weld buttons missing) in one and a crack in the rolled area of the other nozzle thermal sleeve.
At Oconee 1, examination of the four HPI nozzle penetrations to the RC loop inlet line showed no evidence of degradation.
Discussion:
In B&W design plants the line(s) for normal makeup of reactor coolant are also part of the redundant high pressure injection system.
These plants do not have a regenerative heat exchanger in the makeup coolant circuit.
Therefore, during operations, the potential exists for the makeup coolant temperature to be much lower than the reactor coolant temperature in the loop.
Fluid temperature fluctuations resulting from mixing in the HPI nozzle coupled with hydraulic effects are thought to be primary contributors to the cracking problem at Crystal River and at the Oconee plants.
Although the cracking location is within the scope of the LOCA (loss-of-coolant accident) safety analysis, the existence of cracking in an area not routinely included in the program of ISI represents an unacceptable challenge to system integrity.
An evaluation of the cracking problem and its resolution has been requested of the B&W Regulatory Response Group.
Pressurized-water reactor systems of the Combustion Engineering and Westinghouse designs do have a regenerative heat exchanger in the makeup coolant line which is a separate, dedicated system.
During normal power operation the makeup coolant enters the nozzle at temperatures on the order of 500-1500F below the temperature of the reactor coolant loop respectively.
However, tr.ansients may occur in which the makeup flow rate is greater than the letdown flow rate.
Depending on the frequency and duration of these transients, the makeup coolant might not be heated to the expected temperature.
Therefore, the potential may exist for large temperature fluctuations in the makeup nozzle to cause problems similar to those discussed above.
Past experience has shown similar thermal fatigue problems with nozzle-thermal sleeve assemblies in other systems of both BWR (NED0-21821, 1978) and PWR (WCAP-7477 and NED0-9693-1980) designs.
This IE information notice is provided as an early notification of a potentially significant matter that is still under review by the NRC staff.
If NRC evalua-tion so indicates, further licensee action may be requested.
In the interim, we expect that licensees will review this information for applicability to their facilities.
No written response to this information notice is requested.
If you need additional information, please contact the Regional Administrator of the appropriate NRC Regional Office.
Attachment:
Recently Issued IE Information Notices
Attachment IN 82-09 March 31, 1982 I
RECENTLY ISSUED l
IE INFORMATION NOTICES Information Subject Date Issued to Notice No.
Issued 82-08 Check Valve Failures 03/26/82 All holders of a power on Diesel Generator reactor OL or CP Engine Cooling System
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82-07 Inadequate Security 03/16/82 All holders of a power 1
Screening Programs reactor OL or CP 82-06 Failure of Steam 03/12/82 All holders of a power
)
Generator Primary Side reactor OL or CP Manway Closure Studs 82-05 Increasing Frequency 03/10/82 All holders of a power of Drug Related reactor OL or CP Incidents 82-04 Potential Deficiency of 03/10/82 All holders of a power Certain AGASTAT E-7000 reactor OL or CP Series Time-Delay Relays 82-03 Environmental Tests of 03/04/82 All holders of a power Electrical Terminal reactor OL or CP Blocks 82-01 Auxiliary Feedwater Pump 02/26/82 All holders of a power Rev. 1 Lockout Resulting from reactor OL or CP Westinghoue W-2 Switch Circuitry Modification 80-32 Rev. 1 Clarification of Certain 02/12/82 All holders of a materials Requirements for Exclu-license, a Part 50 license, sive Use Shipments of or a fuel facilities Radioactive Materials license 81-37 Rev. 1 Unnecessary Radiation 02/9/82 Selected holders of a Exposures to the material license Public and Washers During Events Involving Thickness and Level Measuring j
Devices 82-02 Westinghouse NBFD 1/27/82 All holders of a Relay Failures in povar reactor Reactor Protection OL or CP Systems at Certain Nuclear Power Plants
oggicnAID ORIGIMAL cortifica By g,! b h -
SSINS:
6835 Accession No.:
8202040126 IN 82-10 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 March 31, 1982 IE INFORMATION NOTICE N0. 82-10:
FOLLOWING Up SYMPTOMATIC REPAIRS TO ASSURE RESOLUTION OF THE PROBLEM Discussion:
There have been a number of instances in which licensees have attempted to correct valve problems by treating the symptoms rather than the underlying cause.
This failure to distinguish between the problem and its symptoms has resulted in recurrence of the problem and further damage to or destruction of the valve or operator.
Symptomatic repairs provide for a return to operability without addressing the underlying problem, earning the label " quick and dirty fixes." The industry jargon recognizes not only that the immediate needs are met, but also that the underlying problem remains to be corrected.
It is this second point which is emphasized: the underlying problem remains to be corrected.
Specifically, valves that leak beyond Technical Specification limits have been restored to operability by allowing additional stem travel.
The direct result of this symptomatic solution has been damage to or destruction of the valve or operator.
One licensee has routinely backseated valves with Limitorque operators using the full motor torque in order to stop stem packing leakage, resulting in damage to the valve backseat.
Bypassing the open limit switch allowed over-l travel in the open direction resulting in binding of the stem in the stem nut.
Because the unthreaded portion of the stem galled in the threads of the stem nut, the threads deformed and the nut cracked.
The valves involved were 600 psi class, Anchor 10-inch integral backseat gate valves with pressure seal bonnets.
Stem and body were type 316 stainless steel.
On a larger scale, in a survey of Licensee Event Reports (LER) for 1978 through 1980, 444 valve operator events were identified for 66 plants. Of these, 193 were identified as motor operator events.
Corrective actions which involved torque switches comprised the largest single corrective action group.
The principal means of corrective action identified was the adjustment of torque switch setting.
This solution was applied to valves in similar service and,
IN 82-10 March 31, 1982 Page 2 of 2 repetitively, to the same valve at several plants.
This indicates that the problem was not being corrected.
The second major corrective action group was limit switch adjustments.
This was a common solution to problems involving valve operation within a time limit.
The cause for the problem is repeatedly given as instrument drift which is undoubtedly true as far as it goes; however, repetition of the events points to the need for a wider ranging solution which will prevent recurrences and improve system reliablity.
During the survey period, there were 16 reported instances of motors replaced in motor operators in the high pressure coolant injection (HPCI), reactor core isolation cooling and residual heat removal systems of boiling water reactors.
This was the third largest corrective action group.
The damage that required replacement of some of the 16 motors resulted from thermal overload protection being bypassed and may be another indication that the underlying valve problem was not corrected.
The common thread in the events as reported by LERs surveyed is the repetition of the problem or the solution, either of which can indicate that a symptomatic repair has been made.
Symptomatic repairs become of concern to the NRC where they impact upon the reliability of the system and where they may adversely affect the health and safety of the public.
When considering the solution to a valve problem, it must be recognized that a symptomatic repair may cause damage to the valve or operator which could impair the safety function of the system to which it is applied.
Consideration should be given to the kind of damage that can occur as a result of the repair and the consequences should a valve fail in a nonconservative direction. A mechanism should exist to identify and resolve the underlying problem when symptomatic repairs are applied.
When the possibility exists for degradation of a safety system as a result of a temporary symptomatic repair to restore operability, prudence dictates a closer surveillance of the system so affected.
This information notice is provided as notification of a potentially significant matter.
It is expected that recipients will review the information for applica-bility to their facilities.
No specific action or response is required at this time.
If you have any questions regarding this matter, please contact the Regional Administrator of the appropriate NRC Regional Office.
Attachment:
Recently Issued Information Notices
._.=_ _
l Attachment TN 82-10 March 31, 1982 RECENTLY ISSUED IE INFORMATION NOTICES 4
Information Subject Oate Issued to Notice No.
Issued 82-09 Cracking in Piping of 03/31/82 All holders of a power Makeup Coolant Lines at B&W Plants 82-08 Check Valve Failures 03/26/82 All holders of a power on Diesel Generator reactor OL or CP Engine Cooling System 82-07 Inadequate Security 03/16/82 All holders of a power Screening Programs reactor OL or CP 82-06 Failure of Steam 03/12/82 All holders of a power Generator Primary Side reactor OL or CP Manway Closure Studs 82-05 Increasing Frequency 03/10/82 All holders of a power j
of Drug Related reactor OL or CP j
Incidents 82-04 Potential Deficiency of 03/10/82 All holders of a power Certain AGASTAT E-7000 reactor OL or CP Series Time-Delay Relays 82-03 Environmental Tests of 03/04/82 All holders of a power Electrical Terminal reactor OL or CP Blocks 82-01 Auxiliary Feedwater Pump 02/26/82 All holders of a power Rev. 1 Lockout Resulting from reactor OL or CP Westinghoue W-2 Switch Circuitry Modification 80-32 Rev. 1 Clarification of Certain 02/12/82 All holders of a materials Requirements for Exclu-license, a Part 50 license, sive Use Shipments of or a fuel facilities Radioactive Materials license 81-37 Rev. 1 Unnecessary Radiation 02/9/82 Selected holders of a Exposures to the material license Public and Washers Ouring Events l
Involving Thickness and Level Measuring Devices l
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