ML19339B551
| ML19339B551 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 10/08/1976 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Groce R YANKEE ATOMIC ELECTRIC CO. |
| References | |
| NUDOCS 8011070217 | |
| Download: ML19339B551 (1) | |
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Docket No. 50-29 Yankee Atomic Electric Company
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Attention:
Mr. Robert H. Groce j
Licensing Engineer f
20 Turnpike Road
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Westborough, Massachusetts 01581 s
j Gentlemen:
If there The enclosed Circular 76-05 is foruarded to you for action.
are any quettions related to your understanding of the actions required, 7
please contact this office.
U Sincerely,
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3 James P. O'Reilly R
Director
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Enclosure:
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IE Circular 76-05 il l-
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NUCLEARREGULATORYCOMMISSIO
'dFFICE OF INSPECTION MiD ENFORCEMENT
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IE Circular No. 76-05 N
DATE:
October 8, 1976
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Page 1 of 4 f
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[i llYDRAULIC Si!0CK AND SWAY SUPPRESSORS - MAINTENANCE OF BLEED My AND LOCK-UP VELOCITIES ON ITT GRINNELL'S MODEL NOS. - FIG. 200 h
AND FIG. 201, CATALOG Pil-74-R f.
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DESCRIPTION OF CIRCUMSTANCES:
E1 Recent information has become availabic related to improper lock-up and bleed rates of certain ITT Grinnell shock and sway suppressors (snubbers).
The control block or valve block on the snubbers involved is one which
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contains the mechanisms for a " dual orifice type" snubber valve arrange-j ment.
These control blocks or valve blocks can be identified by four locking nuts on the surface of the block.
Thesc nuts lock the stem of the valve control into position.
Rotation of the stem modifies the orifices uhich control the lockup and bleed rates.
The mits in ques-tion, at this time, are those snubbers which have control block or valve block serial numbers from B-0001 through B-2000.
These snubbers include 1 1/2 inch diameter (3 kip capacity) through 6 inch diancter (72 kip capacity) units, p
ITT Grinnell first identified this ptoblea in connection with their own testing program during the last two weeks of August 1975.
The origin of the problem vas traced by Crinnell to November of '974 when a design change (known as Revision B) was initiated on Model Nos.- Figure 200 and g
l, Figure 201, Catalog Pii-74-R snubbers, along with the addition of a Grinnell design specification for lockup and bleed velocities.
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[J new rates were to be 8 in/ min. plus or minus 2 in/ min. for lockup and 4 in/ min. plus or ninus 1 in/ min, for biced.
At that time production testing of completed snubbers included within the above series of serial
%w numbers was completed on the basis of "go" or "no-go" testing and did net specifically determine the lockup and bleed rates for each snubber.
N Grinnell has determined that all snubbers involved should lock under seismic events as designed, but perhaps at a lower velocity.
Addi-D tionally, the bleed rate has been determined to exist at a value as low as 1/8 in/ min.
This could result in increased piping stresses depending on the specific design.
ith the uncertainty of specific lock-up and bleed rates, there is the ' possibility of adverse effects on the piping systems or components the snubbers were designed to protect.
It has been stated, however, that to the best of Grinnell's knowledge, there were no lock-up and bleed criteria defined by their customers for the j
plants identified as having utili::ed snubbers from the B-0001 through B-Q 2000 series.
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DATE:
October 8, 1976 Page 2 of 4 t
Even though lock-up and biced rates on snubbers with adjustable orifices are now determined prior to shipment from the manufacturer, there is i
always the possibility that subsequent redoval or tightening of the
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locking nuts on the control' block or valve block will modify the orifice
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setting, thus affecting lock-up and bleed rates.
At least two models of g
Crinnell's snubber control valves use some type of seal on the locking 4
washers and in some cases these may have been replaced in the field to g
change the seal material or correct small leaks.
In these cases, if Q
proper procedures were not followed to maintain the stem portion in its 4t:
original position, the orifice setting could change.
The tightening of y
a lock nut itself could result in a change to the orifice.
1TT Grinnell g
makes the following statement in the printed material furnished with 4
cach snubber assembly: " Adjustment of snubber valve requires equipment
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DO NOT ATTEMPT TO RESET ADJ.USTMENTS IN THE FlELD."
Y The recommended corrective action outlined by ITT Grinnell is to replace 4
the entire valve assembly (2 required per snubber) known as the " barrel" in the case of the lockup control, and the pins or stems (2 required per f' ".
snubber) used to control the bleed rate. ITT Grinnell has prepared documents which prescribe the replacement procedures.
The replacement parts are precalibrated.
Therefore, field calibration of the entire snubber unit will not be necessary if the replacement parts are in-stn11ed in accordance with the prescribed procedures.
The tolerances on the precalibrated pins or stems associated with the bleed mechanism will exceed the original value of 4 inches per minute plus or minus 1 inch /
4 min.
The new biced rate will be 4 inches / minute plus or minus 2 inches /
minute.
Corrective action could also consist of recalibration of the existing 1
snubbers with adjustments made to bring the operating characteristics of
'j the units back within the tolerences specified by ITT Grinnell.
Moni-toring of relative ratation of the lock washer and stem could in the y
future be carried out if position marks were provided on the parts after m
I calibration.
1TT Crinnell.has identified the following facilities which have. licenses under 10 CFR 50 which received snubbers within the group defined by the f
bounding serial number.
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DATE:
October 8, 1976 SN Page 3 of 4 Q1
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!! umber of Grinnell Snubbers f/
Plant Model Nos.-Fig. 200 & Fig. 201 h
fs Farley Nuclear Plant 239 f
Millstone Point Unit #2 18.5
't.q North Anna 155 y
Davis Besse Unit #1 111 4
D. C. Cook 78 Calvert Cliffs 66 T
Deaner Valley 53
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Diablo Canyon 41 Ginna 8
jy rewaunec 2
y Peach Bottom 2
Browns Ferry 1
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ITT Grinnell has also informed the utilities who own these facilities regarding this matter; however, some of the series may be in use or 1
intended for use at other reactor facilities as a rest.lt of utilities loaning or selling units to other utilities or by the fact that sub-contractors or other suppliers might be involved.
I ACTION TO BE TAKEN SY LICENSEE:
^1 For all power reactor facilities with an operating license or a con-
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struction permit using or intending to use the above described snubbers on safety related systems or components:
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Deternine if any of the subject snubbers are installed or scheduled to be installed in safetyrelated systems at your facility.
s a.
Provide a list containing each snubber identified and J
t indicate the piping system or component with which it is associated.
I b.
Indicate for each snubber its history of testing and test results as well as its maintenance history.
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c.
For each snubber list the current required lockup and biced rates and what rates the snubbers were
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specified to meet by the original construction /pur-p chase specification.
Indicate the organization, by y
name, that was responsible for the preparation of the construction / purchase specification.
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et IE Circular 76-05 DATE:
October 8, 1976 Page 4 of 4 d.
If replacement parts are utilized instead of recali-bration of the snubbers, provide an assessment of the effect of a bleed rate of 2 inches / minute which repre-sents the new louer bound.
2.
For snubbers within the above describe'd series that do net nave documentation available defining the lockup and bleed velocities, provide a description of the actions to be taken to substantiate these characteristics.
- Also, provide the schedule to complete these verifications and and modifications, if needed.
3.
Review maintenance and modification procedures for the above described series of snubbers where such actions could affect the performance characteristics of the l
Such actions include but are not limited to the disassembly of snubbers, replacement of seals, changing hydraulic fluid and the changing of any valve springs. Modify the procedures, as appropriate, to assure that any future maintenance or modifications will not change the performance characteristics.
4.
Provide a report which contains the responses to items 1, 2 and 3 above within 90 days of the date of this Circular.
Reports should be submitted to the Director of the NRC Regional Of fice and a copy should be for-N' warded to the NRC Of fice of Inspection and Enforcement, Disision of Reactor Inspection Programs, Washington, D. C.
20555.
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l Approval of NRC requirements for reports concerning possible generic
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problems has been obtained under 44 U.S.C. 3152 from the U.S. General j_'
Accounting Office.
(GAO Approval B-180255 (R0062), expires 7/31/77).
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