Bulletin 74-01: Consumers Power Company
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IEB 74-01
UNITED STATES
ATOMIC ENERGY COMMISSION
DIRECTORATE OF REGULATORY OPERATIONS
REGION III
799 ROOSEVELT ROAD
GLEN ELLYN, ILLINOIS 60137
January 3, 1974
Consumers Power Company Docket No. 50-329
ATTN: Mr. Stephen H. Howell Docket No. 50-330
Vice President
1945 Parnall Road
Jackson, Michigan 49201
Gentlemen:
The enclosed Directorate of Regulatory Operations Bulletin No. 74-1, involving
two valve problems, is sent to you to provide you with informationwe recently
received from the Philadelphia Electric Company and the Wisconsin Electric
Power Company. The problems involved deficiencies identified at the Peach
Bottom, Units 2 and 3 and the Point Beach reactors. This information may
relate to the performance of certain equipment at your facilities. The
Bulletin also requests certain action on your part in this matter.
Sincerely yours,
James G. Keppler
Regional Director
Attachment:
Bulletin No. 74-1
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January 3, 1974
Directorate of Regulatory
Operations Bulletin No. 74-1
VALVE DEFICIENCIES
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 Electric 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 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. 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 facilities
and inform this office in writing within 30 days of the date of this letter
regarding the results of your determination. Also please send a copy of your
report to B.H. Grier, Assistant Director for Construction and Operation,
Directorate of Regulator 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.
Attachments:
A. Philadelphia Electric Co. Ltr dated 10-1-73 to Dr. Knuth
B. Wisconsin Electric Power Co. Ltr dated 10-29-73 to J.F. O'Leary
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ATTACHMENT A
PHILADELPHIA ELECTRIC COMPANY
2301 Market Street
Philadelphia, PA 19101
V.S. Boyer
Vice-President
October 1, 1973
Dr. D.F. Knuth, Director
Directorate of Regulatory Operations
United States Atomic Energy Commissions
Washington, D.C. 20545
Subject: Significant Deficiency Report -
High Pressure Service Water Valve Weld Failure
Peach Bottom Atomic Power Station - Units 2 & 3
AEC Construction Permit Nos. CPPR-37 and CPPR-38
File: QUAL 2-10-2 SBR No. 5
___________________________________________________________
Dear Dr. Knuth:
In compliance with 10CFR50.55, paragraph (e) attached is the Significant
Deficiency Report concerning the weld failure on the High Pressure Service
Water valve in Unit No. 2. This item was reported to AEC DRO I by telecon on
June 1, 1973.
We trust that this satisfactorily resolves this item. If further information
is required, please do not hesitate to contact us.
We appreciate your extending the time for our response to October 1, 1973 as
agreed by telecon on September 14, 1973 between our Mr. G.R. Hutt and Mr. R.
Heischmarm, USAEC DRO I.
Sincerely,
V.S. Boyer
Copy to: J.P. O'Reilly, USAEC
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SIGNIFICANT DEFICIENCY REPORT - SBR NO. 5
HIGH PRESSURE SERVICE WATER VALVE WELD FAILURE
PEACH BOTTOM ATOMIC POWER STATION - UNITS 2 & 3
AEC CONSTRUCTION PERMIT NOS. CPPR-37 AND CPPR-38
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 weld 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 weld threat
dimension caused by the installation of unauthorized shims between the yoke
legs and the mounting plate, which reduced the effective size of the weld.
Corrective Action
The failed valve is one of a series of eight valves (four in Unit 2 and four
in Unit 3). These eight valves were visually inspected and a second valve was
found to have cracks in the yoke to motor operator mounting plate weld. 1 All
eight valves were returned to the vendor for rework. The rework involved
elimination of the shims in the failed valve and the rewelding of the mounting
plates to the yoke legs with full penetration welds on all eight valves.
An investigation of similar valves (supplied by the same vendor) elsewhere in
the plant, was undertaken. A total of 108 valves were identified by the
vendor to have yoke to motor operator mounting plate construction similar to
that of the failed valve. Fifty-eight (including the above mentioned eight)
of these valves 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. The
remaining valves are Group III non-nuclear balance of plant valves.
The Vendor's weld stress analysis calculations were reviewed and a table of
acceptable weld sizes prepared.
______________________________________
1 This valve was originally reported in the interim report to have shims.
The valve was only visually inspected at that time and the cracks were
interpreted to indicate the presence of shims.
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ATTACHMENT B
WISCONSIN ELECTRIC Power Company
231 West Michigan
Milwaukee, Wisconsin 53201
October 29, 1973
Mr. John F. O'Leary, Director
Directorate of Licensing
U.S. Atomic Energy Commission
Washington, D. C. 20545
Dear Mr. O'Leary:
DOCKET NOS. 50-266 AND 50-301
FACILITY OPERATING LICENSE NOS. DPR-24 AND DPR-27
POINT BEACH NUCLEAR PLANT
BACKSEATING DISC MISLOCATION PROBLEM ON 2" DARLING VALVES
_________________________________________________________
In accordance with Section 15.6.A.3.b of the Technical Specifications for
Point Beach Nuclear Plant (Facility Operating License Nos. DPR-24 and DPR-27),
this report describes a possible generic problem with a category of 2" gate
valves installed at Point Beach Nuclear Plant. The valves in questions are
2", No. S-350 WDD welding end, outside screw and yoke, double disc gate valves
with lip seals, and are manufactured by the Darling Valve and Manufacturing
Company. The valves used at Point Beach Nuclear Plant are safety class I, ASA
series 1500 lb. valves.
An investigation of excess letdown line leakage on September 15, 1973, lead to
an inspection and subsequent repair of valve 1MOV-1299 on Unit 1 (excess
letdown system root valve) on September 26, 1973. Inspection of the valve
disclosed that its downstream seat protruded from the valve body such that if
the valve disc was fully withdrawn from the guides, as allowed by its
backseating ring, the disc could catch the "lip" of the seat ring when
reinserting. Four marks on the lip of the downstream seat ring indicated that
the disc had caught there during previous valve closings. Internal damage to
the valve consisted of a fine vertical crack at the 12 o'clock position in the
upper portion of the downstream seat ring. Two locating pins between the
upstream and downstream discs of the split disc valve were found to be
slightly bent also and some facial scratches to the down-stream disc were
evident. There was no metal loss involved in the damage.
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Mr. John F. O'Leary - 2 - October 29, 1973
Repair of the valve involved rounding the lip of the seat ring to prevent
future hangups of the disc. The thin vertical crack in the downstream seat
could not be fully lapped out during the repair. Accordingly, a manual valve
was added to the 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 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 sat
rings when the valve was fully open and backseated. The tendency for
interference 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 handing disc onto
the lip of the seat.
There are six similar 2" Darling valves in each unit at Point Beach Nuclear
Plant. In addition to the above mentioned 1299 valve, valves 270A & B
(normally open) are installed on 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. These
valves are never operated during normal pressurized and power operation. The
sixth similar valve on each unit is MOV-426 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 report are
containment isolation valves.
Valve 1MOV-427 was investigated during a Unit 1 shutdown on October 13, 1973,
after it was reported that it would not fully close remotely. manual
manipulation of the valve on September 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
1MOV-427 was capable of performing its primary function of initiating an
isolation signal for the letdown line. The slightest movement of the valve
off its backseat was found to be sufficient to activate the interlocks and
close the ACV-200 letdown isolation valves.
Measurements indicated that the discs of 1MOV-427 when
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Mr. John F. O'Leary - 3 - October 29, 1973
backseated cleared the seat rings and left the valve open to similar problems
as experienced in 1MOV-1299. Inspection showed no damage to valve 1MOV-427
other than a slight marking of the upper edge of the seat ring, similar to
that found in 1MOV-1299. Before closing up the valve, the seat ring edges
were rounded to aid in guiding the discs down between the seats. The "valve
open" limit switch was then sat for 2-1/4", 5/16" less than the maximum
backseating position of 2-9/16". Valve cycling tests were then conducted
satisfactorily.
During the same shutdown, valve 1MOV-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 telephone 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 require investigation on a schedule commensurate with the
plant operating schedules.
Valves 1MOV-1299, 2MOV-1299 and 2MOV-427 will be electrically limited
similarly to 1MOV-427. Valve 1MOV-1299 will be completely changed out during
a convenient shutdown following the receipt of a new valve. New valve stems
with backseats 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 shutdowns. 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 informed about the problems
encountered with these valves.
Very truly yours,
Sol Burstein
Senior Vice President
cc: Mr. James G. Keppler
Regional Director
Directorate of Regulatory Operations,
Region III
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