ML19350A811
| ML19350A811 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Byron, Braidwood, Quad Cities, Zion, LaSalle  |
| Issue date: | 03/04/1981 |
| From: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Reed C COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 8103160894 | |
| Download: ML19350A811 (1) | |
Text
,p Ker t-UNITED STATES 8 *' 3.,.,
j NUCLEAR REGULATORY COMMISSION
-r REGION 111 Q
[
799 ROOSEVELT ROAD f
o' GLEN ELLYN, ILLINOIS 60137
- ..+
March 4, 1981 Gentlemen:
IE Bulletin No. 81-01 has been revised to change the identification of the power reactor facilities with crustruction permits to which the bulletin is applicable. This change does not affect your requirements with respect to the bulletin. As a result, the enclosed revised pages 5 and 6 are forwarded to you for information.
Should you have any questions regarding this bulletin revision, please contact this office.
Sincerely, j.
7,
.L 2
. )_y L[
s L L '* w *
[/JamesG.Keppler l
Director
Enclosure:
IE Bulletin No. 81-01, Revision 1 810316089Y1 Q
/
Docket Nos. 50-10, 50-237, 50-249; 50-254, 50-265; 50-295, 50-304; 50-374; 50-454, 50-455; 50-456, 50-457 Commonwealth Edison Company ATIN:
Mr. Cordell Reed Vice President Post Office Box 767 Chicago, IL 60690 cc w/ enc 1:
g, f,%/,b g
J. S. Able, Director g
of Nuclear Licensing y
(((p D. J. Scott, Station 3
Superintendent 7
N. Kalivianakis, Plant T
f4IE 9
40IO8/A.
N Superintendent (3
h K. L. Graesser, Station A
Nbg ff Superintendent A
L. J. Burke, Site
\\s//
Construction Superintendent 4
ed T. E. Quaka, Quality Assurance Supervisor R. H. Holyoak, Station Superintendent B. B. Stephenson, Project Manager V. I. Schlosser, Project Manager R. E. Querio, Station Superintendent Gunner Sorensen, Site Project Superintendent R. Cosaro,p:ndent oject Superi -
'J. F.
dac, Station erintendent tral Files AD/ Licensing AD/ Operating Reactors AEOD Resident Inspectors, RIII PDR Local PDR NSIC TIC Dean Hansell, Office of Assistant Attorney General Myron M. Cherry
SSINS No.:
6820 Accession No.-
8005050075 IEB 81-01 n
D
~3 UNITED STATES NUCLEAR REGULATORY COMMISSION 3o AE OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 January 27, 1931 IE Bulletin No. 81-01:
SURVEILLANCE OF MECHANICAL SNUBBERS Description of Circumstances:
Several instances of failures of mechanical snubbers supplied by International Nuclear Safeguards Corporation (INC) have been identified that indicate passible deficiencies in these snubbers.
A summary of the failures that have occurred is provided below:
1.
On August 9, 1974, the Tennessee Valley Authority sutmitted event
. port BFA0-50-260/741W identifying 11 of 14 INC Model MSVA-1A snubbers that were found inoperable on Browns Ferry Nuclear Power Station Unit 2 and subsequently identified 5 of 14 inoperable units on Browns Ferry Nuclear Power Unit No 3.
All of these units were found to be frozen, and the cause was attributed to a failure to lubricate the parts during assembly.
The failed snubbers were replaced with new units produced by the same manufacturer.
2.
On April 12, 1976, le St. Lucie Plant Unit 1 facility of Florida Power and Light Corporatit i submitted event report No. 50-335-76-9 wherein five INC Model MSVA-1 snu bers were identified as inoperable because they were found to be frozen.
The failures were caused by oxidation on the internals and oy improper assembly.
All INC mechanical snubbers were replaced with units pr]duc'.d by another manufacturer.
3.
On April 8, 1977, Iowa Electric Light and Power Company submitted event report No. 77-23 for the Duane Arnold Energy Center facility that identi-fied 13 INC Model 1MSVA-1 Type AS snubbers to be frozen; the cause of failure was attributed to large amounts of interior oxidation.
The units were replaced with those produced by another manufacturer.
4.
On December 5, 1979, personnel from the Nuclear Regulatory Commission visited Department of Energy (DOE) facilities at Richland, Washington, to obtain information on DOE experience with INC snubbers at the Fast Flux Test Facility (FFTF).
The DOE owned FFTF was equipped with more than 4,000 mechanical pipe restraints (snubbers) supplied by INC.
In 1978, FFTF examined more than 800 of these mechanical snubbers by removing them from their installation and found that 43, or about 5% of those examined, were frozen.
The plant was_still under construction so the snubbers had seen no service and had been subjected to only normal construction environments for 1 to 2 years.
Attachment:
Recently issued IE bulletins
IEB 81-01 January 27, 1981 Page 2 of 6 Tests were conducted on three operable snubbers by installing them on a Hanford Engineering and Development Laboratory (HEDL) process line.
The three snubbers were subjected to flow-induced low-amplitude vibration (0.003 inches or less).
These snubbers were of both the combined carbon steel and stainless steel construction and the all stainless steel con-struction.
Detailed test data are not available to the NRC at this time.
However, all three snubbers froze af ter being subjected to the vibration for periods of 3 to 30 days.
The failure modes on all units inspected and tested involved a number of different mechanisms leading to the freezing of the snubbers.
Following disassembly of some of the snubbers, inspections showed the failures were caused by improper assembly; overheating of internal components caused by we'. ding (during fabrication); and sensitivity of the design to dirt, corrosion, and inadequate or excessive lubrication.
DOE concluded that there were generic deficiencies in the design of the snubbers of this specific manufacturer for application to the FFTF facility and for pipes subjected to vibration.
All INC mechanical snubbers in FFTF have been replaced-with snubbers produced by another manufacturer.
5.
On May 31, 1980, Georgia Power Company reported eight INC snubbers located on instrument and drain lines at Edwin I. Hatch Nuclear Plant Unit 1 were identified as inoperable (LER 321-80-55).
The cause of the failures was identified as internal corrosion that causea a frozen condition.
In an attempt to free a snubber (750 pound capacity), forces up to 1500 pounds were applied in both the " extend" and " retract" directions and the snubber did not move.
The inspection of INC snubbers was completed at the Hatch facility and, on June 30, 1980, NRC received a supplemental report that 45 of the 61 snubbers that had been inspected on 01it I had been identi-fied as inoperable and three of the 42 snubbers ttat were inspected on Unit 2 were inoperable. All inoperable snt.bbers were replaced prior to startup of the affected unit.
Some were replaced with mechanical units produced by another manufacturer, some were replaced with later model INC snubbers, and three were replaced with rigid restraints.
Plans are being made to replace.all INC snubbers during upcoming refueling outages.
Analyses are also being' performed on the piping affected by the locked up snubbers.
In addition to INC snubber failures, failures of mechanical snubbers by another manufacturer are identified below:
1.
On September 7, 1979, Public Service Electric and Gas Company reported the failure of three Model PSA-3 mechanical snubbers manufactured by Pacific Scientific Company that were located on, rr,.1 feedwater line of Salem Nuclear Generating Station Unit-1 (LER 79-54).
These three snubbers could not be rotated cround their spherical rod end bearings.
The snubbers were removed and inspection revealed that the lead screw and traveling nut assembly, which translates linear to rotational motion, had failed.
The snubbers no lont,er provided seismic shock restraint under this condition.
These snubbers are directly upstream of the nuclea-Class II piping boundary and are included in the stress calculations for the seismic analysis of the nuclear portion of the main feedwater piping. _ Failure of the snubbers
IEB 81-01 January 27, 1981 Page 3 of 6 appeared to result from a force many times greater than the design load of the snubbers.
This force was either an extreme shock load or occurred when the snubber was in the fully retracted condition.
The snubbers were replaced with units produced by the same manufacturer.
2.
On April 10, 1979, Consumers Power Company reported a failure of eight Model PSA-3 Pacific Scientific snubbers at their Big Rock Point Nuclear Plant facility (LER 79-017/03L-0).
The cause of the failure was improper installation in that a spherical washer was omitted from the transition tube.
3.
On March 15 and June 11, 1979, Florida Power and Light reported failures of Pacific Scientific Company mechanical snubbers on main steam and feedwater systems at Turkey Point Plant Units 3 and 4 (LER 79-006/03L-0 and 79-009/03L-0 respectively).
The cause in both cases was attributed to excessive loading.
The nature of the above mechanical snubber failures is to prevent the piping systems, to which they are attached, from moving freely during the normal thermal heat up and cool down associated with plant operations.
Restraining this thermal motion results in higher than normal stresses which, if high enough and repeated frequently enough, can lead to a premature fatigue failure of the piping system.
These mechanical snuboers have been installed for a number of years without any NRC requirements for periodic surveil ance to determine their condition.
As a result, their current condition is unknown to NRC and therefore NRC is requesting a prompt examinatico of all mechanical snubbers installed to date.
Because of the high percentage of failures discovered with the INC snubbers, the time frame for their examination is the shortest and additional opera-
-bility tests are called for.
Actions to be Taken by Licensees of Operating Reactors:
1.
Within 30 days of the issuance date of this bulletin, all normally accessible
- INC mechanical snubbers installed on safety-related systems or in storage shall be visually examined and tested as follows:
a.
Perform a visual examination for damage and, without causing the system to be inoperable except as permitted by the facility technical specifications, verify that the snubbers have freedom of movement by performing a manual test over the range of the stroke in both com-pression and tension.
b.
Perform an operability test to confirm that (1) actit. tion (restraining action) occurs in both compression and tension and (2) the drag forces are-within the specified range in both compression and tension.
The tests shall be performed on all snubbers in storage and on a representative sample (10% of the total of this type of snubber in use in the plant or 35, which ever is less) of the
""Normally accessible" refers to those areas of the plant that can be entered during reactor operation.
IEB 81-01 January 27, 1931 Page 4 of 6 normally accessible snubbers that are in service and can be indiv-icually removed without causing the system to be inoperable, except as permitted by the facility technical specifications.
For each snubber that does not meet the test acceptance criteria, an additional representative sample (as defined above) of this type of snubber shall be tested.
For each of these additional snubbers that do not meet the 2est acceptance criteria, another representative sample of this type of snubber shall be tested. This cycle shall be repeated cntil no more i
failures have been fcund or until all snubbers of this type have been tested.
The samples
.:,uld be made up of snubbers representing the various sizes.
Snubbers which have been examined and test'd in a manner comparable c.
to Items la and lb above within the lass
< months may be exempted.
d.
If any failures are identified in Itt la or Ib above, take corrective action and evaluate the effect of the failure on the system operability pursuant to the facility technical specifications for continued operation.
If failures are identified in Items la and Ib above, and if INC snubbers e.
are known to be located in any inaccessible areas, a plant shutdown shall be performed within 30 days after the discovery of the first inoperable snubber and inspections conducted in accordance with Item 2a and 2b below, unless justification for continued operation has been provided to the NRC.
2.
Visually examine and test all inaccessible INC mechanical snubbers installed on safety related systems at the next outage of greater than five days duration as follows:
Visually examine and manually test all inaccessible snubbers as a.
described in Item la above.
b.
. Perform an operability test on a representative sample of inaccessible i
snubbers as described in Item Ib above.
Snubbers which have been examined and tested in a manner ccmparable c.
to Items 2a and 2b above within the last six months may be exempted.
d.
If any failures are identified in Items 2a or 2b above, take corrective action to evaluate the effect of the failure on system operability pursuant to the facility technical specifications for resuming operation.
3.
Provide a schedule for an inspection program covering mechanical snubbers produced by~other manufactures. 'As a minimum, this inspection program shall:
Include all snubbers _ installed on safety-related systems; a.
b.
Include the visual examination and manual. test described in Item la above for all snuboers;
~
i IEB 81-01, Rev. 1 March 4, 1981 Page 5 of 6 c.
Snubbers which have been examined and tested in a manner comparable to Item 3b above within the last twelve months may be exempted:
d.
Require the corrective action and evaluations described in Items Id and 2d above; and e.
Be completed prior to the completion of the next refueling outage.
Plants which are currently in a refueling outage should perform the visual examination and manual tests of inaccessible mechanical snubbers before resumption of operations unless some other basis for assurance of snubber operability is provided to the NRC.
4.
Submit a report of the results of the inspections, testing and evaluation requested in Item 1 to NRC within 45 days of the issuance date of this bulletin.
Report the results of the inspections, testing and evaluation requested in Item 2 within 30 days after the inspection and testing have been completed.
The response to Item 3 shall be submitted within 60 days of the issuance date of thi Bulletin.
The results of the inspections 4
performed for Item 3 shall De submitted within 60 days after the completion of the inspection.
1 The reports shall contain the following:
a.
A description of the visual examinations and tests performed.
b.
Number of snubbers examined and tested.
Grouping by manufacturer name, model number, and size is acceptable.
c.
Number of failures identified; manufacturer name, model number, size, mode.of failure, cause of failure, corrective action, snubber. location, effect of fai re on plant and system safety, and justification for i.
continuing os resuming operation.
d.
The-above information shall also be provided for the snubbers-exempted by Items 1c, 2c, and 3c above.
. Actions to be Taken by the Following Licensees Holding Construction Permits:
Diablo Canyon Nuclear Power Plant Units 1 and 2;-Grand Gulf Nuclear' Station, R1
+
Unit 1; LaSalle. County Station, Unit 1; Virgil C. Summer Nuclear Station, R1 Unit-1; and Susquehanna Steam Electric Station, Unit 1 shall perform the R1 following:
~
R1 i
1.
After preoperational and/or hot functional testing and preceding fu.i loading, visually _ examine and test the mechanical. snubbers installed on safety-related systems as follows:
a.
'For all snubbers perform a visual examination for damage and verify that.the snubbers have fre' dom of movement'by performing a manual test over the range of the stroke in both compression and~ tension.
r 1
4 IEB 81-01, Rev. 1 March 4, 1981 c
Page 6 of 6 b.
For INC snubbers, perform an operability test to confirm that (1) activation (restraining action) occurs in both compression and tension and (2) the drag forces are within the specified range in both compression and tension.
The tests shall be performed on a representative sample (10% of the total of this type of snubber in use in the plant or 35, which ever is less).
For each snubber that does not meet the test acceptance criteria, an additional representative sample (as defined above) of this type of snubber shall be tested.
For each of these additional snubbers that do not meet the test acceptance criteria, another representative sample of this type of snubber shall be tested.
This cycle shall be repeated until no more failures have been found or un".il all snubbers of this type have been tested.
The samples should be rnade up of snubbers that represent the various sizes.
If any failures are identified in Items a or b above, take corrective c.
action prior to fuel loading.
2.
The schedule for the inspections and tests requested in Item 1 above, shall be submitted within 60 days of the issuance date of this bulletin.*
The R1 results of the inspections, testing, and evaluation requested in Item 1 shall be reported to NRC within 30 days after the inspection and testing have been completed.
The reports shall contain +he following:
A description of che visual examinations and tests performed.
a.
b.
Number of snubbers examined and tested.
Grouping by manufacturer name, model number, and size is acceptable, Number of failures identified; manufacturer name, model number, c.
size, mode of failure, cause of failure, corrective action, and snubber location.
Reports, signed under oath or affirmation, under the provisions of Section 182a of the Atomic Energy Act of 1954, shall be submitted to the Director of the appropriate NRC Regional Office and a copy shall be forwarded to the Director of the NRC Office of Inspection and Enforcement, Washington, D. C.
_20555.
If you desire additional information regarding this matter, please contact the IE Regional Office.
Approved by GA0 B-180225 (581003) expires December 31, 1981.
"The "issurance date of this bulletin" shall be considered to be the date R1 of issuance of revision 1 for-the following licensees holding construction R1 permits:
Diablo Cnayon Nuclear Power Plant, Unit 2; Grand Gulf Nuclear R1 Station, Unit 1; LaSalle County Station, Unit 1; and Susquehanna Steam R1 Electric Station, Unit 1.
R1
Attachment IEB 81-01, Rev 1 March 4, 1981 RECENTLY ISSUED IE BULLETINS Bulletin No.
Subject Date Issued Issued To 80-17, Failure of Control Rods 2/13/81 All BWR facilities Supplement 5 to Insert During a Scram with OL or CP 81-01 Surveillance of 1/27/81 All power reactor Mechanical Snubbers facilities with OL
& to specified facilities with CP 80-25 Operating Problems with 12/19/80 All BWR facilities Target Rock Safety-Relief with OL & specified Valves at BWRs near term OL BWR facilities & all BWRs with a CP Supplement 4 Failure of Control Rods 12/18/80 To specified BWRs to 80-17 to Insert During a Scram with an OL & All at a BWR BWRs with a CP 80-24 Prevention of Damage 11/21/80 All power reactor Due.to Water Leakage facilities with Inside Containment OL or CP ~
(October 17, 1980 Indian Point 2 Event) 80-23
-Failures of Solenoid 11/14/80 All power reactor Valves Manufactured by facilities with Valcor Engineering OL or CP Corporation
~
80-22 Automation Industries, 9/11/80 All radiography-Model 200-520-008 Sealed-licensees Source Connectors 80-21 Valve-yokes supplied by 11/6/80=
All light water Malcolm Foundry Company, Inc.
reactor facilities with OLs or cps Supplement 3 Environmental. Qualification 10/24/80 All power' reactor to 79-018 of Class 1E Equipment facilities with an OL OL = Operating License CP = Construction Permit