ML19305A777
| ML19305A777 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 03/07/1980 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Nichols T SOUTH CAROLINA ELECTRIC & GAS CO. |
| References | |
| NUDOCS 8003180321 | |
| Download: ML19305A777 (2) | |
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UNITED STATES 8
NUCLEAR REGULATORY COMMISSION n
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REGION 11 o
101 MARIETTA ST N.W., SUITE 3100 o
ATLANTA, GEORGIA 30303 In Reply Refer To:
MAR 0 71980 RII:JPO G595D
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South Carolina Electric and Gas Company Attn:
M. C. Johnson, Vice President Special Services and Purchasing Post Office Box 764 Columbia, South Carolina 29218 Gentlemen:
This Information Notice is provided as notification of a potentially significant matter.
It is expected that recipients will review the infor-mation for possible applicability to their facilities.
No specific action or response is requested at this time.
If further NRC evaluations indicate the need, an IE Circular or Bulletin will be issued to request specific licensee actions.
If you have questions regarding this matter, please con-tact the Director of the appropriate NRC Regional Office.
LSincerely,
/
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/ " James P. O'Reilly Director
Enclosures:
1.
IE Information Notice No. 80-08 w/its Enclosure 2.
List of Recently Issued IE Information Notices l
800318032.1
14 i
South Carolina Electric and Gas Company :
cc w/ encl:
T. B. Conners, Jr.
Conners, Moore and Corber-1747 Pennsylvania Avenue, N.W.
Washington, D.C.
20006 A. A. Smith Quality Assurance Post Office Box 8 Jenkinsville, South Car-lina 29065
- 10. S. Bradham, Manager Nuclear Operations Post Office Box 8 Jenkinsville, South Carolina 29065 a
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a SSINS: 6870 UNITED STATES Accession No.:
NUCLEAR REGULATORY COMMISSION 7912190689 0FFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 March 7, 1980 IE Information Notice No. 80-08 THE STATES COMPANY SLIDING LINK ELECTRICAL TERMINAL BLOCK Description of Circumstances:
On July 19, 1979, the Consumers Power Company notified the Nuclear Regulatory Commission of a defect found in the sliding link eltetrical terminal block manufactured by the States Company, a subsidiary of Multi Amp Corp. The defective terminal blocks were found at the Midland plant.
The connection between the two slotted bars on the terminal block is made by a U-shaped sliding link and spacer located between the two bars.
The top of the U-shaped link and the spacer are drilled and the bottom of the link is threaded to accept a 8-32 screw. When the screw is tightened it binds the link, spacer and bar together to make electrical connection. Loosening the screw and sliding the link from between the bars breaks the connectioa. The purpose of the link is to provide easy insertion of test instruments, etc.
into the circuit.
The defect, which has been identified in 5% of the' terminal blocks checked, occurs in the form of a crack between the threaded screwhole and the side of the U-shaped link.
When the screw is tightened the crack widens and a poor or intermittent electrical connection can result. A defective link is impossible to cinch tightly in place and is difficult to detect visually.
Enclosure I shows the States Company terminal block. The defect, a crack in the bottom portion of the metal U-shaped link, is displayed in the exploded view of the terminal block assembly. These# terminal blocks are widely used in the nuclear industry and may be used as permanent installations in safety related systems. The defective mechanical connection can cause an electrical circuit malfunction.
This Information Netice is provided to inform licensees of a potentially significant matter.
It is expected that recipients will review the informa-tion for applicability to their facilities. No written response to this IE Information Notice is required. However, the reporting requirements as set forth in the regulations must be met.
If you require additional information regarding this matter, contact the Director of the appropriate NRC Regional l
Office.
1
Enclosure:
Graphic Display of Terminal Block
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ENCLOSURE 1 TO IE INFORMATION NOTICE 80-08 i
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Exploded View of States Company Terminal Block
a IE Information Notice No. 80-08 Enclosure g
March 7, 1980 RECENTLY ISSUED IE INFORMATION NOTICES Information Subj ect Date Issued To Notice No.
Issued 80-08 The States Company Sliding 3/07/80 All holders of an OL or Link Electrical Terminal Block a CP 80-07 Pump Shaft Fatigue Cracking 2/29/80 All Light Water Reactor Facilities holder power reactor OLs and cps 80-06 Notification of Significant 2/27/80 All holders of Reactor Events OLs and to near term OL applicants 80-05 Chloride Contamination 2/8/80 All licensees of nuclear of Safety Related Piping power reactor facilities and applicants and holders of nuclear power reactor cps 80-04 BWR Fuel Exposure in 2/4/80 All BWR's holding a Excess of Limits power reactor OL or CP 80-03 Main Turbine Electro-1/31/80 All holders of power Hydraulic Control System reactor OLs and cps 80-02 8X8R Water Rod Lower 1/25/80 All BWR Facilities End Plug Wear holder power reactor OLs or cps 80-01 Fuel Handling Events.
1/4/80 All holders of power d
reactor OLs and cps 79-37 Cracking in Low Pressure 12/28/79 All power reactor OLs Turbine Discs and cps i
79-36 Computer Code Defect in 12/31/79 All power reactor OLs Stress Analysis of Piping and cps Elbow 79-35 Control of Maintenance 12/31/79 All power reactor facilities and Essential Equipment with an OL or CP 79-34 Inadequate Deuign of 12/27/79 All holders of power reactor Safety-Related Heat OLs and cps
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. January 3, 974 O.
Ig Directorate of Regulatory Operations Bulletin 74-1
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VALVE DEFICIENCIES i
Information was recently received from the Philadelphia Electric Company and the Wisconsin Electric Power Company concerning two types of deficiencies relating to valves.
The deficiency identified by the Philadelphia flectric Company at the Peach Bottom, Units 2 and 3 facilities related to weld failures between the valve yoke and the motor operator mounting plate in valves supplied by the Walworth Company. A full description of the deficiency is provided in Attachment A.
The second deficiency, identified by the Wisconsin Electric Power Company at the Point Beach plant, involved a backseating disc mislocation problem on two inch Darling valves. Full details are provided in Attachment B.
In light of the above information, you are requested to determine whether similar valves are installed or scheduled to be installed in your h
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h facilities and inform this office in writing within 30 days of the date of this letter regarding the results of your deturmination.
Also please send S copy of your report to B. H. Grier, Assistant Director for Construction and Operation, Directorate of Regulatory Operations, USAEC, Washington, D. C.
20545. In the event such valves are identified, you are requested to determine whether those identified valves have the deficiencies described and if so, to inform us in your letter of the corrective action-planned and the date of scheduled completion of that corrective action.
-i Attachments:
A.
Philadelphia Electric Company Letter, dated October 1,1973 to Dr. Knuth B.
Wisconsin Electric Power Company Letter, dated October 29, 1973 to J. F. O' Leary O
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Ph1LADELPHIA ELECTRIC COMPANY 2301 MARKET STREET
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P:'!LADELPHIA. PA.19101 1215) 841-4 Soo ncs-entseosur October 1, 1973 Dr. D. F. Knuth, Director Directorate of Regulatory Operations United States Ato:ic Energy Con =ission Washington, D.C.
20545
Subject:
Significant Deficiency Report -
High Pressure Service Water Valve Veld Failure Peach Bottom Atomic Power Station - Units 2 & 3 AEC Construction Permit Nos. CPPR-37 and CPPR-38 File: CUAL 2-10-2 SDR No. 5
Dear Dr. Knuth:
In compliance with 10CFR50 55, paragraph (e) attached is the s
Significant Deficiency Report concerning the veld failure on the High Pressure Service Water valve in Unit No. 2.
This item was reported to v
AEC DRO I by telecon on June 1, 1973 We trust that this satisfactorily resolves this item. If further infor=ation is required, please do not hesitate to contact us.
We appreciate your extending the time for our response to October 1, 1973 as agreed b telecon on September 14, 1973 between f
our Mr. G. R. Hutt and Mr. R. Heischnann, USAEC DRO I.
l Sincerely, 1
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J. P. O'Reilly, USAEC Attachment A t
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HIGH PRESSURE SERVICE WATER VALVE 11F'~' FAILURE PEACH BOTTOM ATOMIC POWER STATION - UNITS 2 & 3
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AEC CONSTRUCTION PERMIT NOS. CPPR-37 AND CPPR-38 i
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_ Description of Deficiency During a routine walk-thru of Unit No. 2 plant by the licensees operating personnel, a 12 inch - 300 pound motor operated globe valve in the High Pressure Service Water line on the discharge side of one Residual Heat Removal heat exchanger was discovered to have experienced a veld failure.
The failure occurred between the valve yoke and the motor operator mounting plate.
The reason for the failure has been identified as insufficient fillet veld throct dimension caused by the installation of unauthori::ed shims between the yoko legs and the mounting plate, which reduced the effective si::e of the veld.
l Corrective Action l
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The failed valve is one of a series of eight valves (four in
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Unit 2 and four in Unit 3).
These eight valves were visually inspected and a second valve uns found to have cracks in the yoke to motor operator mounting plate veld.1 All eight valves were returned to the vendor for rework.
The rowork involved elimination of the chims in the failed valve and the revelding of the counting plates,to the yoke legs with full pene-q tration velds on all eight valves.
i An investigation of similar valves (supplied by the same vendor) elsewhere in the plant, was undertaken.
A total of 108 valves were iden-tified by the vendor to have yoke to motor operator count struction similar to that of the failed valve.
Fifty..eight including the above mentioned eight) of these /alves are nuclear valves classified as Group II as defined by Figure A.2.1 of Appendix A of the Peach Bottom Atomic Power Station FSAR.
halance of plant valves.
The remaining valves are Group III non-nuclear 3
The Vendor's veld stress annlysis calculations were reviewed and a table of acceptable veld si::es prepared.
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1 This valve was originally reported in the interim report to have shims.
The valve was only visually inspected at that time and l
the cracks were interpreted to indicate the presence of shims.
1 Attachment A Page 2 of 2 L
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October 29, 1973 Directorate of LicensingMr. John F. O' Leary, Directo Washington, D. C.U. S. Atomic Energy Ccmmissior.
20545
Dear Mr. O' Leary:
DOCKET NOS.
FACILITY OPERATING LICENSE NOS50-266 AND 50-BACKSEATING DISC MISLOCATION PROBL ON 2" DARLING VALVES Technical Specifications for Point bin accordance with Sec E
Operating License Nos. DPR-24 and DPReach Nuclear Plant L
...b of the stalled at Point Beach Nuclear Planta possible gen
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(Facility y of 2" gate valves in-are 2" disc ga,te v2.ives with lip sealsNo. S-350 hTD welding end, Darling Valve and Manufacturing Co
, and are manufactured by thee screw a j
,U Point Beach Nuclear Plant are safet mpany.
lb. valves.
The valvos used at F,>
y class I, ASA series 1500 An investigation of excess letdown li September 15, 1973, lead to an insp ne leakage on of valve IMOV-1299 on Unit 1 (excess letdection and subsequen I
on Septerber 26, 1973.
the. valve disc was fully withdrawn frits downstre Inspection of the valve disclosed thown system at e valve body such that if by its backses
.ng ring, the disc could catch the " lip" of theom th seat ring when reinserting.
stream seat ring indicated that the disc h d Four marks on the lip of the dow previous valve closings.
a n-of a fine vertical crack at the 12 o'clo kInternal damage to the val portion of the downstream seat ring the upstream and downstream discs of c
e Two locating pins betweenpositio found to be slightly bent also and scm the split disc valve were down stream disc were evident.
e facial scratches to the in the damage.
There was no metal loss involved f~\\
Attachment B Page 1 of 3 I
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nr. uann r. u b ty October 29, 1973 gs Repair of the valve involved rounding the lip of the
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seat ring to prevent future hangups of the disc.
The thin ver-g
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tical crack in the downstream seat could not be fully lapped out during the repair.
Accordingly, a manual valve was added to the j
system downstream of 1MOV-1299 to back up the root valve.
Valve 1MOV-1299 thereby remains effective and operable as a remotely controlled root shutoff valve, but is considered not totally capable of effecting completely tight shutoff without some through-leakage.
At the time, measurements indicated that the location l
of the backseating ring on the valve stem was too low but this could not be assuredly determined.
If such was the case, this would allow the split discs to fully clear the seat rings when the valve was fully open and backseated.
The tendency for inter-ference to occur between the downstream disc and seat during valve closing could be expected to increase if there was flow through the valve, creating a differential pressure-which could swing the loose hanging disc onto the lip of the seat.
There are six similar 2" Darling valves in each unit i
at Point Beach Nuclear Plant.
In addition to the above mentioned 1299 valve, valves 270A & B (normally open) are installed en the reactor coolant pump seal return lines.
These valves are. rarely operated in the life of the plant.
Also, valves 598 and 599 on the reactor coolant system drain line are of this type.
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operation.
The sixth,similar valve on each unit is MOV-427 on the normal letdown line.
The function of valve 427 is to close in the' event of low pressurizer level and, in closing, cause the closure of the containment isolation valves 200 A, B and C, via an interlock.
None of the Darling valves described in this re-port are containment isolation valves.
Valve IMOV-427 was investigated during a Unit 1 shut-down on October 13, 1973, after it was reported that~it would not fully close remotely.
Manual manipulation of the valve on Septem-ber 28, 1973, had shown that at approximately one-half shut and again just prior to closing, the valve operation became sticky.
Tests were conducted at that time to verify that IMOV-427 was capable of performing its primary function of initiating an iso-lation signal for the letdown line.
The slightest movement of the valve off its backseat was found to be sufficient to activate the t
interlocks and close the AOV-200 letdown isolation valves.
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Measurements indicated that the discs of 1MOV-427 when i
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backseated cls red the seat rings and le.c the valve apen to simi-
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lar problems as e::pericz ed in 1MOV-1299.
Inspection showed no
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damage to valve IMOV-427 other than a slight marking af the upper edge of the seat ring, similar to that found in 1MOV *293.
Before s
closing up the valve, the seat ring edges were rounded te aid in
_ guiding the discs down between the seats.
The " valve c an" limit u
switch was then set for 2-1/4", 5/16" less than the maximum back-seating positien of 2-9/16".
Valve cycling tests were then con-ducted satisfactorily.
4 During the same shutdown, valve IMOV-270B was cycled manually with no evidence of stickiness or disc hangup.
At the completion of repair of 1MOV-427, on October 13, 1973, it was concluded from measurements taken, operating experience and tele-phone discussions with the valve manufacturer, that, indeed, a dimension error could exist with respect to backseat locations on the stem.
With these confirmations, it was concluded that all twelve valves of this type would recuire investigation on a sche-dule commensurate with the plant operating schedules.
Valves IMOV-1299, 2MOV-1299 and 2MOV-427 will be
. electrically limited similarly to IMOV-427.
Valve IMOV-1299 i
will be completely changed out during a convenient shutdown fol-lowing the receipt of a new valve.
New valve stems with back-
. seats located so that full opening of the valve will not permit the discs to lose the guide effect of the seats have been ordered and will be fitted in the remaining valves at convenient shut-h(m,/.)
downs.
The service of the 598, 599 and 270A & B valves is such that it is not considered necessary to change the stems of these valves until the next refueling shutdown of each unit.
The nuclear steam supply system supplier has been in-formed about the problems encountered with these valves.
Very truly yours, Sol Burstein Senior Lice President I
cc:
Mr. James G. Keppler Regional Director Directorate of Regulatory Operations, Region III Attachment B Page 3 of 3 e
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Directorate of Regulatory Operations Bulletin 74-A
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dated January 3, l'974
- gO Distribution:
Chief, Field Supp* ort and Enforcement Branch Deput.y Director, RO Assistant Director for Construction and Operation, RO (3)
Assistant Director for Radiological, Environmental, and Materials Protection, RO Director, Office of Operations Evaluation, RO Assistant for Plans and Programs, RS Director, DL Deputy Director, DL Deputy Director for Technical Review, DL Assistar. Director for B~wT 'r. sL (3)
Assistant Director for PWR4, % (3)
Assistant Director for Operating Reactors, DL (3)
Assistant Director for Environmental Projects, DL (3)
R. F. Fraley, ACRS p
__en Files DR Central Files cc: Letter adaressed to Licensee w/enci w/o distribution and address list O
NSIC R. L. Shannon,'DTIE (TIC)
Chief, Regulatory News Branch, IS PDR Local PDR l
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November 27, 1973 Directorate of Regulatory Operations Bulletin No. 73-6 I
INADVERTENT CRITICALITY IN A BOILING WATER REACTOR We recently received an abnormal occurrence report from the Vermont Yankee Nuclear Power Corporation relating to an inadvertent criticality incident that was experienced at their Vermont Yankee facility. A copy of this report is being sent to you under separate cover to provide you with pertinent details of this event.
At the time of the inadvertent criticality incident, the reactor vessel and primary containment heads were removed, ths refueling cavity above the reactor vessel was flooded, control rod friction tests were in progress, the rod worth minimizer was bypassed, and core verification had been 11 progress. As a result of the incident, no measurable radioactiv:.ty was released, no fuel damage resulted and no personnel exposures were experienced. The incident is currently under review and evaluation by the Regulatory Staff.
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t.ction requested by this bulletin is contained in Section A.
A.
Action Requested by Licensees l
In light of this occurrence you are requested to take the following actions. Upon completion of these actions you are requested to inform this office in writing, within 45 days of the date of this letter, of the status of each item identified in each paragraph and subparagraph listed below:
1.
Procedural Review a.
Control Rod Drive Operating and Testing Procedures (1) Conduct a review of your control rod drive operating and testing procedures to determine that approved procedures exist for all operations and tests.
(2) Verify that appropriate prerequisites are includuo in the procedures to require testing of associated 2.nter-locking and protection features before control rad f
testing is permitted.
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(3) Assure that prerequisites and detailed instructions are provided that demonstrate compliance with technical specification requirements and design bases.
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b.
Bypass Installation Procedures (Jumpers or Lifting of Leads)
Assure that existing bypass installation procedures have been conservatively reviewed for technical adequacy and for afainis-strative controls.
c.
Radiation Protection Procedures Assure that procedures for access control and personnel accountability in areas subject to accidents are current.
d.
Shift Transition Procedure (Turnover)
Assure ; I complete and detailed procedures are in effect that provide instructions for a proper and conservative turnover of shif t responsibilities. Such procedures must include requirements for communicating the status of all safety related equipment and conditions.
2.
h agement Controls Assure that your management contr'ols that are in effect provide for
[ 'T qualified technical and administrative reviews and approvals of
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temporary circuitry changes and temporary off-normal plant conditions.
This review should assure that the responsibilities and requirements associated with the review and approval, installation, verification, removal, and subsequent testing of temporary circuitry changes and temporary off-normal plant conditions are clearly delineated in station procedures, are understood by the station staff, and are being properly implemented.
3.
Licensed Operator Performance Assure that management provides the necessary opportunities and time so that operators are adequately trained to carry out their assigned responsibilities.
In particular, confirm that shift crew members are provided special training for safety related activities that are infrequent, complex, or have unusual safety significance.
If you have any questions concerning this request, please contact this office.
Enclocsre:
A0 No, 7331 - Ltr dtd 11/14/73 (separate cover) f'~'s l
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September 13, 1976 IE Circular No. 76-03 RADIATION EXPOSURES IN REACTOR CAVITIES DESCRIPTION OF CIRCUMSTANCES:
-On March 18, 1976, an employee at the Zion station received a "whole body" radiation dose of 8 rems or more upon entering the cavity beneath the reactor vessel during a refueling outage.
On April 5, 1976, a similar reactor cavity entry at Indian Point resulted in a 10-rem whole body dose to a licensee employee.
A similar entry on October 5,1972, caused a 5-rem dose to a Point Beach employee.
These three overexposures appear to have been caused by failure to appropriately control entry into high radiation areas, failure to conduct adequate surveys and failure to compensate for exposure rate variations that can occur in various areas in power reactors, e.g., the cavity beneath the reactor vessel. With the incore thimbles and detectors inserted into the core, radiation levels in the cavity appear to be low.
With the thimbles or detectors with-drawn into the cavity, however, exposure rates of hundreds or possibly thousands of roentgens pre h;ai 4.cn exist. Overexposures can occur in seconds.
All three overexposure events involved entry into potentially high radiation areas without surveys and/or special controls over equip-ment which could cause transients in the exposure rate.
ACTION TO BE TAKEN BY LICENSEES:
While the three exposures above occurred at pressurized water reactors, similar situations could develop at other types of reactors, e.g.,
pneumatic irradiation equipment areas (research reactors) and traveling incore probe equipment areas (boiling water reactors).
Accordingly, holders of power, test and research reactor operating licenses are to complete the following:
1.
Perform a thorough review of plant areas and operations to identify high radiation areas, both continuous and transient, as defined in 10 CFR 20.202(b).
i-2.
Verify that entryways into high radiation areas are conspicuously posted and locked or otherwise controlled in such a manner as to explicitly identify the nature of the hazard, appropriately I
control entry, and require adequate pre-entry surveys, t
3.
Ensure that radiation protection procedures and radiation pro-tcction training and retraining programs specifically address the matter of control of and access to such areas and initiate appropriate retraining of all plant personnel, L
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IEbircularNo.76-03 September 13, 1976 4
4.
Ensure that the procedures governing personnel entry into all i
nctual or potential high radiation areas permit such entry only after appropriate management review and approval so that conditions t
within the area are known and not subject to change while the area is occupied, 5.
Periodically audit whatever controls result from items 1-4, above, to ensure their continued effectiveness, and 6.
Confirm by written rep 3) within 60-days that the actions for items 1-4 above, have been or are being taken.
A record, detailing findings, actions traen, and actions to be taken, should be retained for review by NRC during the next radiological safety inspection.
This request for information was approved by GAO under a blanket clearance number B-180225 (R0072); this clearance expires July 31, 1977, t
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August 17, 1976 IE Circular No. 76-02 e
RELAY FAILURES-WESTINGHOUSE BF (ac) AND BFD (de) RELAYS DESCRIPTION OF CIRCUMSTANCES:
During testing of Westinghouse BFD relays, the Point Beach nuclear power plant experienced malfunctions with two relays in the reactor trip system.
The malfunctions were caused by the pin that connects the plunger to the operating head rubbing against the contact block. Although the coils were fully energized the relay contacts remained in the deenergized position.
A similar malfunction occurred in one of a set of relays undergoing accelerated aging tests at the Westinghouse Beaver facility.
The malfunction relating to pin misalignment may be common to both BF (ac) and BFD (de) relays. Portions of a Westinghouse service letter containing information about these relays are attached to this circular.
Further instructions regarding this relay problem can be obtained from Westinghouse Nuclear Service Division, Pittsburgh, Pennsylvania 15230.
ACTION TO BE TAKEN BY LICENSEES AND PERMIT HOLDERS:
The following actions should be taken w1.th respect to all Westinghouse BF (ac) and BFD (de) relays in safety related systems:
1.
Describe the action taken or that you plan to take to verify that normally energized relays in safety related systems are in fact operable and that the relay contacts are in the energized position.
2.
Describe the action taken or that you plan to take to verify that normally deenergized relays in safety related systems operate properly when energizt' and that the relay contacts are in the energized position.
Reports for facilities with operating licenses should be submitted within 60 days after receipt of this circular, and reports for facilities with construction permits should be submitted within 90 days after receipt of this circular. Your report should include the date when the above actions were or will be completed.
Reports should be submitted to the Director of the NRC Regional Office and a copy should be forwarded to the NRC Office of Inspection and Enforcement Division of Reactor Inspection Programs, Washington, D. C. 20555.
Approval of NRC requirements for reports concerning possible generic problems has been obtained und.er 44 U.S.C 3152 from the U. S. General Accounting Office.
(GAO Approval B-180255 (R0072), expires 7/31/77)
Attachment:
j Extract from Westinghouse Service Letter:
i BF (ac) and BFD (de) Relays l
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EXTRACT FROM WESTINGHOUSE SERVICE LETTER:
BF & BFD RELAYS During the testing of certain Westinghouse BFD relays at an operating nuclear power plant, two relays in the reactor trip system were found to have malfunctions. Although the coils were fully energized, the relay contacts remained in e.he deenergized position.
It was determined that, in both cases, the pin that connects the plunger to the operating head was rubbing against the contact block. This rubbing action resulted in friction that impeded the plunger movement when the relay coil was energized thereby preventing contact movement. The malfunctioning relays were immediately replaced. When dissassembled it was found that the relays would operate normally when the pin was centered in the plunger.
Coincidently, Westinghouse (Beaver) the relay manufacturer, experienced a similar malfunction in one of a set of similar relays which are currently undergoing accelerated aging tests.
Westinghouse (Beaver) and Westinghouse (NES) are currently investigating this situation in detail as it applies to both BF (ac) and BFD (de) models.
Con-sideration is being given to various means by which the pin could be held cap-tive, thereby precluding further pin misalignment, if.such a course of action becomes necessary.
RECOMMENDED ACTION Visually inspect normally energized relays to verify that such relays are in fact picked up.
Observe the performance of normally deenergized relays during normal periodic testing.
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July 28, 1976 IE Circular No. 76-01 CRANE HOIST CONTROL - CIRCUIT MODIFICATIONS DESCRIPTION OF CIRCUMSTANCES:
In response to NRC concerns about the potential for, and con-sequences-of, dropping a spent fuel shipping cask or other heavy load, Commonwealth Edison modified the hoist control system for the fuel cask handling cranes at their Dresden Units 2 and 3 and Quad-Cities Units 1 and 2 to provide additional hoist redundancy and slow speed boist capability. The original design utilized a General Electric "magspeed" hoist control system.
In this sys-tem which includes two electro-mechanical brakes in series, spring force holds the brakes engaged while DC solenoids, energized when the hoist motor is energized, disengage the brakes.
The modification which added the slow speed hoist capability included installing additional contactors in the brake solenoid power circuit to energize the solenoids when the low speed hoist motor was
-ergized.
The original hoist control system design utilized a single Size 2 DC contactor (two contacts in series) in the solenoid circuit.
ra The design modification added a circuit in parallel with the
- original DC contactor which utilized four AC rated Size-1 single - - - --~
contacts in a series-parallel array to distribute current carrying and interrupting burden.
Initial experience with the modified hoist control system at Dresden showed that the circuit interrupting capacity of the series-part.llel array was marginal. On several occasions when the low speed motor was stopped in the lowering mode, the solenoid circuit contacts arced resulting in power being supplied to the solenoida long enough so that the load 'ropped some distance before the brakes engaged. Over travel of as much as 15 inches was reported, but no damage to hoist or load was found.
The crane manufacturer's representatives have advised the NRC that the proposed corrective action is to install a single Size 2 DC contactor (two contacts in series) with are suppressors, the same as originally provided in-the General Electric design, in place of the added four AC rated contacts. The original contactor -
in the normal speed control circuit has shown satisfactory service since initial operation of the plant in 1969.
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IE Circular No. 76-01 July 28, 1976 ACTION TO BE TAKEN BY LICENSEE:
1.
Determine and report to this office within 90 days the following information:
(a) Have you made, or do you plan to make modifications to the hoist control for your installed cranes similar to the described modiff cations?
(b) If such modifications have been made, or are planned, identify changes required in brake power and control circuitry?
(c) What steps have been taken or are planned, to provide assurance that brake power contactors are adequate for the service?
2.
If modifications are planned, provide the schedule for completion and a brief description of your plans for design review and functional testing.
Your response should be submitted to the Director of this Office, with a copy to the Director, Division of Reactor Inspection Programs, Office of Inspection and Enforcement,.
U. S. Nuclear Regulatory Commission, Washington, D. C.
20555.
Approval of NRC requirements for reports concerning possible generic problems has been obtained under 44 U. S. C. 3152 from the U. S. General Accounting Of fice.
(GAO Approval B-180255 (R0072), expires 7/31/77).
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