HomeMy WebLinkAbout851183.tiff to Rear,
° r'To UNITED STATES
I" 9` NUCLEAR REGULATORY COMMISSION
m 3 WASHINGTON,D.C.20555
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***** November 1 K985,,,,
Docket No. 50-267 ^,r
21985 -'l
Mr. 0. R. Lee, Vice President -.
Electric Production
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
SUBJECT: FORT ST. VRAIN (FSV) TECHNICAL SPECIFICATION UPGRADE PROGRAM SCHEDULE
Dear Mr. Lee:
Enclosed is our proposed schedule for completion of the Technical
Specification Upgrade Program in accordance with commitments outlined in your
letter dated July 11, 1985 and our letters dated December 20, 1984 and
July 19, 1985. We request that you provide your comments on this schedule
within 30 days of your receipt of this letter.
The information requested in this letter affects fewer than 10 respondents;
therefore, OMB clearance is not required under P.L. 96-511.
Sincerely,
Edward J. Butcher, Acting Chief
Operating Reactors Branch No. 3
Division of Licensing
Enclosure:
As stated
cc w/enclosure:
See next page
851183
Mr. 0. R. Lee
Public Service Company of Colorado Fort St. Vrain
cc:
C. K. Millen Albert J. Hazle, Director
Senior Vice President Radiation Control Division
Public Service Company 4210 East 11th Avenue
of Colorado Denver, Colorado 80220
P. 0. Box 840
Denver, Colorado 80201 J. W. Gahm
Nuclear Production Manager
Mr. David Alberstein, 14/159A Public Service Company of Colorado
GA Technologies, Inc. P. 0. Box 368
P. 0. Box 840 Platteville, Colorado 80651
Denver, Colorado 80201
J. K. Fuller, Vice President
Public Service Company
of Colorado
P. 0. Box 840
Denver, Colorado 80201
Senior Resident Inspector
U.S. Nuclear Regulatory Commission
P. 0.Box 640
Platteville, Colorado 80651
Kelley, Stansfield & O'Donnell
Public Service Company Building
Room 900
550 15th Street
Denver, Colorado 80202
Regional Administrator, Region IV
U.S. Nuclear Regulatory Commission
Office of Executive Director
for Operations
611 Ryan Plaza Drive, Suite 1000
Arlington, Texas 76011
Chairman, Board of County Commissioners
of Weld County, Colorado
Greeley, Colorado 80631
Regional Representative
Radiation Programs
Environmental Protection Agency —
1800 Lincoln Street
Denver, Colorado 80651
Enclosure
FSV Technical Specification Upgrade Schedule
• April 1, 1985 - Licensee submittal of the upgraded FSV Technical
Specification 1st draft.
• May 21, 1985 - NRC reviewers review and mark up 1st draft
(sent to licensee).
• July 22-26, 1985 - Site visit by NRC reviewers to resolve noted
differences (developed list of utility and
NRC action items).
• September 1, 1985 - Resolution of NRC action items from site meeting
• September 15, 19851 - Licensee submittal of redrafted LCOs on helium
circulators, steam generators, and PCRV liner
cooling due to the extensive nature of the
required revision.
ö November 30, 1985 Licensee to submit the final draft of the upgraded
Technical Specifications for NRC review. To include
resolution of PSC action items.
o January 30, 1986 - Final draft reviewed by NRC and forward mark up to
the licensee (45 days).
o February 11-15, 1986 - Site visit by TSRG reviewer to resolve noted
differences and develop list of utility and NRC
action items (5 days).
o March 15, 1986 - TSRG review to resolve all NRC and licensee action items
from site visit and any new differences and proposed
changes (30 days).
o April 30, 1986 - Licensee to submit application to amend the
Specifications
(90 days from receipt of NRC final draft comments).
o May 30, 1986 - NRC issue license amendment approving FSV Technical
Specification Upgrade (30 days).
o FSV 4th Refueling - PSC implementation of the upgraded Technical
January 1, 1987 Specification (contingent on NRC approval at least
6 months prior to the refueling start date).
(S) Completed Milestone
(1) Completed September 30, 1985
EpR RE°V
J°` 41
co 0yy UNITED STATES
g NUCLEAR REGULATORY COMMISSION
N ; WASHINGTON,D.C.20555
OW
Y"o 4' October 21, 1985
Docket No. 50-267 °Ee
J
/7
\ 1Q85
Mr. 0. R. Lee, Vice President
Electric Production
Public Service Company of Colorado c"'°•
P. 0. Box 840
Denver, Colorado 80201
SUBJECT: REQUEST FOR ADDITIONAL INFORMATION - FORT ST. VRAIN BUILDING 10
Dear Mr. Lee:
We are continuing our review of your submittals dated July 20 and 25, 1984
concerning Building 10. As part of this review, we found certain material
illegible or missing. We would appreciate it if you would provide the
following information:
1. CN-1255 and CN-1332 for System 92
2. CN-1294 for System 92/45
3. Any subsequent changes in battery design as indicated on CN-1391
For each CN, please provide a legible copy with the safety evaluation, design
criteria and design analysis.
We request that you provide this information within 30 days of your receipt
of this letter.
The information requested in this letter affects fewer than 10 respondents;
therefore, OMB clearance is not required under P.L. 96-511.
Sincerely,
Edward J. Butcher, Acting Chief
Operating Reactors Branch No. 3
Division of Licensing
cc: See next page
Mr. 0. R. Lee
Public Service Company of Colorado Fort St. Vrain
cc:
C. K. Millen Albert J. Hazle, Director
Senior Vice President Radiation Control Division
Public Service Company 4210 East 11th Avenue
of Colorado Denver, Colorado 80220
P. 0. Box 840
Denver, Colorado 80201 J. W. Gahm
Nuclear Production Manager
Mr. David Alberstein, 14/159A Public Service Company of Colorado
GA Technologies, Inc. P. 0. Box 368
P. 0. Box 840 Platteville, Colorado 80651
Denver, Colorado 80201
J. K. Fuller, Vice President
Public Service Company
of Colorado
P. 0. Box 840
Denver, Colorado 80201
Senior Resident Inspector
U.S. Nuclear Regulatory Commission
P. 0.Box 640
Platteville, Colorado 80651
Kelley, Stansfield & O'Donnell
Public Service Company Building
Room 900
550 15th Street
Denver, Colorado 80202
Regional Administrator, Region IV
U.S. Nuclear Regulatory Commission
Office of Executive Director
for Operations
611 Ryan Plaza Drive, Suite 1000
Arlington, Texas 76011
Chairman, Board of County Commissioners
of Weld County, Colorado
Greeley, Colorado 80631
Regional Representative
Radiation Programs
Environmental Protection Agency
1800 Lincoln Street
Denver, Colorado 80651
J°`EpR REG0,
' UNITED STATES
NUCLEAR REGULATORY COMMISSION
m ; WASHINGTON,O.C.20555
o ;
4'
November 15, 1985
Docket No. 50-267
Mr. 0. R. Lee, Vice President / - -Electric Production
Public Service Company of Colorado DEC )
P. 0. Box 840 f 219g5
Denver, Colorado 80201 -
�r-�
SUBJECT: DISCUSSIONS ON LICENSING STATUS
_ Dear Mr. Lee:
This letter summarizes discussions held with Mr. L. Brey and Mr. M. Holmes of
your staff on October 24, 1985 concerning the status of licensing' actions.
During these discussions, two items were noted that require PSC's attention.
They are as follows:
1. In your letter dated September 18, 1985 (P-85321) on Long Term
Improvements, your response to Item b. stated: "Presently, PSC has a
program underway to requalify the CRDMs for operating temperatures of
300°F. This program is expected to be complete by January 1, 1987."
This schedule is in conflict with your commitments as contained in
Enclosure 2, Item 11 of the NRC Confirmatory Action Letter dated
July 19, 1985, which states: "Perform environmental requalification
testing of a CRDM assembly (P-85032) including the temperature sensor
epoxy (P-85195)-Sechedule 12/30/85."
If you wish relief from this commitment, you should make an appropriate
request with justification to NRC's Region IV office. Otherwise, the
dates in the Confirmatory Action Letter remain unchanged. _
2. In our letter of August 28, 1985, we requested your review and comments
on our approach concerning review of several pending applications for
changes to your Technical Specifications. An acknowledgement of your
review, even without comments, would be appreciated.
We request that you respond to this letter within 15 days of receipt.
The information requested in this letter affects fewer than 10 respondents;
therefore, OMB clearance is not required under P.L. 96-511.
Since ' ��"" "►`.'
Edward J. Butcher, Acting Chief
Division of Licensing
Office of Nuclear Reactor Regulation
cc: See next page
Mr. 0. R. Lee
Public Service Company of Colorado Fort St. Vrain
cc:
C. K. Millen Albert J. Hazle, Director
Senior Vice President Radiation Control Division
Public Service Company 4210 East 11th Avenue
of Colorado Denver, Colorado 80220
P. 0. Box 840
Denver, Colorado 80201 J. W. Gahm
Nuclear Production Manager
Mr. David Alberstein, 14/159A Public Service Company of Colorado
GA Technologies, Inc. P. 0. Box 368
P. O. Box 840 Platteville, Colorado 80651
Denver, Colorado 80201
J. K. Fuller, Vice President
Public Service Company
of Colorado
P. O. Box 840
Denver, Colorado 80201
Senior Resident Inspector
_ U.S. Nuclear Regulatory .Commission
P. 0.Box 640
Platteville, Colorado 80651
Kelley, Stansfield & O'Donnell
Public Service Company Building
Room 900
550 15th Street
Denver, Colorado 80202
Regional Administrator, Region IV
U.S. Nuclear Regulatory Commission
Office of Executive Director
for Operations
611 Ryan Plaza Drive, Suite 1000
Arlington, Texas 76011
Chairman, Board of County Commissioners
of Weld County, Colorado
Greeley, Colorado 80631
Regional Representative
Radiation Programs
Environmental Protection Agency
1800 Lincoln Street
Denver, Colorado 80651
SSINS No. : 6820
Pi£Llt CC ""4r"--:"^""S OMB No. : 3150-0011
Expiration Date: 08/01/88
n v ! ' t ;n IEB 85-03
1 ' •
I
DEC 21985 _ UNITED STATES
1 NUCLEAR REGULATORY COMMISSION
__ OFFICE OF INSPECTION AND ENFORCEMENT
GREELEY, COL,. WASHINGTON, DC 20555
November 15, 1985
IE BULLETIN NO. 85-03: MOTOR-OPERATED VALVE COMMON MODE FAILURES DURING
PLANT TRANSIENTS DUE TO IMPROPER SWITCH SETTINGS
Addressees:
All holders of nuclear power reactor operating licenses (OLs) or construction
permits (CPs) for action.
Purpose:
The purpose of this bulletin is to request licensees to develop and implement a
program to ensure that switch settings on certain safety-related motor-operated
valves are selected, set and maintained correctly to accommodate the maximum
differential pressures expected on these valves during both normal and abnormal
events within the design basis.
Description of Circumstances:
There have been two recent events, and a number of earlier events, during which
motor-operated valves failed on demand, in a common mode, due to improper
switch settings.
Event 1 Davis-Besse Plant - On June 9, 1985, the Davis-Besse Plant experienced
a complete loss of main and auxiliary feedwater. This event was described
previously in IE Information Notice No. 85-50, "Complete Loss of Main and
Auxiliary Feedwater at a PWR Designed by Babcock & Wilcox," and in NUREG-1154,
"Loss of Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June 9,
1985. " Normally open, Limitorque motor-operated auxiliary feedwater (AFW) gate
valves failed to reopen on either an automatic or manual signal from the main
control room after they were inadvertently closed during the event. While
other failures also occurred in the AFW system, the failure of these two valves
was itself enough to prevent AFW from reaching either steam generator. During
the recovery from this event, the valves were opened with the handwheels.
The results of licensee troubleshooting activities after the event led to the
conclusion that the setting for the torque switch bypass limit (torque bypass)
switch in each valve' s control circuit had not been set to remain closed long
enough to provide the necessary bypass function on valve opening with differen-
tial pressure conditions across the valve. During the event, the valves
experienced a high differential pressure after closing. The torque bypass
8511130441
Bdm-0'ZM/ 8s
IEB 85-03
November 15, 1985
Page 2 of 6
switch on both valves was improperly set, causing the torque switches to become
operable prematurely. This condition stopped valve travel before the valve
discs were fully off their seats.
The torque bypass switches were set to drop out after the valves opened to 5%
full-stroke. The 5% full-stroke setting was based on a number of handwheel
turns. In a 10 CFR 21 report, submitted subsequent to the event, Toledo Edison
Company identified two reasons why the torque bypass switch settings were not
adequate: (1) the 5% full-stroke settings were not adequate for unseating the
AFW system discharge valves with large differential pressures across the valves
and (2) the procedure for setting this switch was defective in that the 5%
full-stroke was not specified to be in addition to the handwheel turns required
to take up the motor-operator coast and backlash. The torque bypass switch
setting errors were revealed only when high differential pressure conditions
across the valves caused higher loadings. The valve failures were reproduced
during tests performed by the licensee with differential pressures applied
across the valves. During the tests, the valves operated properly when low
differential pressures were applied across them, but failed to open when high
differential pressures were applied. The valves were instrumented during these
tests to obtain signature traces of critical parameters.
Event 2, Sequoyah Plant Unit 2 - An event involving partial loss of main
feedwater occurred on May 2, 1985, at Sequoyah Nuclear Plant Unit 2 while in
Mode 2 and returning to power after a reactor trip. Feedwater was being
supplied through the main feedwater (MFW) system isolation valve bypass lines.
Operators attempted to open the MFW system isolation valves to supply water to
the steam generators; however, two of the four MFW isolation valves would not
open. The startup was discontinued and the unit was returned to hot shutdown.
During examination to determine the reason for the valve failures, the licensee
discovered that both valve stems had sheared from their discs. The discs were
found in the closed positions within the valve seats. The stems had suffered
fracture failures through approximately three-quarters of the diameter of the
shafts, in addition to stress failures of the remaining quarter. The Limitorque
motor-operators on the valves use limit switches to control valve motion in the
open direction. These MFW system isolation valves are large (18 inch diameter) ,
fast acting (154 inches per minute travel speed) valves. Because of the high
speed of these valves and the large mass of the discs, the selection of the
limit switch setpoint needs to account for the large momentum of the disc and
its continued motion after the limit switch deenergizes the valve motor-operator.
The set point was not correctly established and the disc impacted the backseat
during opening. The failure mechanism of these valves was identified by the
licensee to be impact loading of the stem on the opening stroke as a result of
the disc impacting the backseat, combined with a stress failure of the remaining
portion of the stem on the opening stroke. Main feedwater valves are not included
in the actions requested by this bulletin. The NRC is, however, continuing to
evaluate the Sequoyah event.
NRC Field Evaluation - As a part of the resolution of Generic Issue II.E.6,
"In-Situ Testing of Valves," the NRC contracted with the Oak Ridge National
Laboratory in 1984 to perform a limited study to determine the effectiveness of
signature tracing techniques in determining the operational readiness of
IEB 85-03
November 15, 1985
Page 3 of 6
safety-related motor-operated valves. It was hoped also that this study could
provide some insight as to current conditions of valve switch settings at
nuclear power plants. Signature traces of motor current, torque and limit
switch actuations and axial motion of the worm gear (an indication of operator
torque) were obtained from 36 motor-operated valves at 4 nuclear plant sites.
Although the formal technical report [NUREG/CR-4380 "Evaluation of the
Motor-Operated Valve Analysis and Test System (MOVATS) to Detect Degradation,
Incorrect Adjustments, and Other Abnormalities in Motor-Operated Valves"] has
not been issued, the current draft of the report indicates that (1) this
inspection method can be used to improve current ASME methods and (2) there
were abnormalities with nearly every valve tested.
Table 1 contains a summary of the study's findings with respect to switch
setting abnormalities. Of particular interest with respect to the events
described above is the finding that 75% of the valves had improperly set torque
bypass switches (56% of the valves had the close-to-open torque bypass switch
set so that it was opening before the valve fully unseated) and 8% of the
valves were unintentionally backseating. The abnormalities in Table 1 have not
been fully evaluated at this time, and they should not be interpreted to mean
that any abnormality resulted in an inoperable valve.
Background:
The NRC has previously identified common mode failures, on demand, of valves.
IE Circular No. 77-01, "Malfunction of Limitorque Valve Operators," reported
that on October 28, 1976, two motor-operated (Limitorque) valves located
between the refueling water storage tank and the charging pump suction at the
Trojan Nuclear Plant failed to open in response to a spurious safety injection
(SI) signal . The malfunction in both valves resulted from the torque switch in
the opening circuit becoming activated before the valves were fully off their
seats. The valves also were equipped with a torque bypass switch. Each of the
valves that malfunctioned was found to have its torque bypass switch adjusted
such that it allowed the torque switch to be operable in the circuit before the
valve was moved from its seat. The licensee's investigation revealed that in
each case the valve had been manually closed hard on its seat following a
maintenance operation. Examination by the licensee revealed similar improper
adjustments of the torque bypass switches on several other motor-operated
valves in safety-related systems.
IE Information Notice No. 81-31, "Failure of Safety Injection Valves To Operate
Against Differential Pressure," reported on September 3, 1981, that both trains
of the San Onofre Unit 1 safety injection system were found to be inoperable
when challenged to operate against differential pressure. Improperly set
switches were the principal cause of these failures. There were no adverse
consequences in this particular event because there was no accident that
required safety injection. The reactor pressure remained above the safety
injection pump's shutoff head; therefore, no actual injection of water would
have occurred if the valves had opened. However, had reactor pressure decreased
and actual injection been required, injection flow would not have been automa-
tically available as designed. These valves had been regularly tested at each
c � ,
IEB 85-03
November 15, 1985
Page 4 of 6
refueling outage, but the tests were not required to be performed with
differential pressure across these valves.
Florida Power Corporation reported an event at Crystal River Unit 3 in LER
77-9. During plant cooldown with the unit in hot shutdown, decay heat removal
valves in the decay heat removal pump suction would not open with remote
actuation. These failures were caused by pressure acting on the gate valve
discs. The valves were opened manually with the handwheels. The torque
switches were reset.
In addition to common mode valve failures on demand, there have been numerous
common mode failures discovered during testing or as a result of investigating
a single failure. NUREG/CR-2270, "Common Cause Fault.Rates for Valves,"
February 1983, contains reports of 99 common cause valve fault events from
1976 through 1980.
The NRC has previously identified other problems with motor-operated valve
switches in Bulletin No. 72-3, "Limitorque Valve Operator Failures"; IE
Information Notice No. 79-03, "Limitorque Valve Geared Limit Switch Lubricant";
Circular No. 81-13, "Torque Switch Electrical Bypass Circuit for Safeguard
Service Valve Motors"; Information Notice No. 82-10, "Following Up Symptomatic
Repairs To Assure Resolution of the Problem"; and Information Notice No.84-10,
"Motor-operated Valve Torque Switches Set Below the Manufacturer's Recommended
Value."
The failure and potential failure of Westinghouse Electro-Mechanical Division
motor-operated gate valves to close are discussed in IE Bulletin No. 81-02 and
IE Bulletin No. 81-02, Supplement 1, "Failure of Gate Type Valves To Close
Against Differential Pressure."
Copies of the above referenced NRC bulletins, circulars and information notices
can be obtained from your local public document room.
Actions for All Holders of Operating Licenses or Construction Permits:
For motor-operated valves in the high pressure coolant injection/core spray
and emergency feedwater systems (RCIC for BWRs) that are required to be tested
for operational readiness in accordance with 10 CFR 50.55a(g) , develop and
implement a program to ensure that valve operator switches are selected, set
and maintained properly. This should include the following components:
a. Review and document the design basis for the operation of each valve. This
documentation should include the maximum differential pressure expected
during both opening and closing the valve for both normal and abnormal
events to the extent that these valve operations and events are included in
the existing, approved design basis, (i .e. , the design basis documented in
pertinent licensee submittals such as FSAR analyses and fully-approved
operating and emergency procedures, etc). When determining the maximum
differential pressure, those single equipment failures and inadvertent
equipment operations (such as inadvertent valve closures or openings) that
are within the plant design basis should be assumed.
1
/1
IEB 85-03
November 15, 1985
Page 5 of 6
b. Using the results from item a above, establish the correct switch settings.
This shall include a program to review and revise, as necessary, the methods
for selecting and setting all switches (i .e. , torque, torque bypass,
position limit, overload)q for each valve operation (opening and closing).
If the licensee determines that a valve is inoperable, the licensee shall
also make an appropriate justification for continued operation in accordance
with the applicable technical specification.
c. Individual valve settings shall be changed, as appropriate, to those
established in item b, above. Whether the valve setting is changed or not,
the valve will be demonstrated to be operable by testing the valve at the
maximum differential pressure determined in item a above with the excep-
tion that testing motor-operated valves under conditions simulating a
break in the line containing the valve is not required. Otherwise, justi-
fication should be provided for any cases where testing with the maximum
differential pressure cannot practicably be performed. This justification
should include the alternative to maximum differential pressure testing
which will be used to verify the correct settings.
Note: This bulletin is not intended to establish a requirement for valve
testing for the condition simulating a break in the line containing the
valve. However, to the extent that such valve operation is relied upon in
the design basis, a break in the line containing the valve should be
considered in the analyses prescribed in items a and b above. The resulting
switch settings for pipe break conditions should be verified, to the extent
practical , by the same methods that would be used to verify other settings
(if any) that are not tested at the maximum differential pressure.
Each valve shall be stroke tested, to the extent practical , to verify that
the settings defined in item b above have been properly implemented even if
testing with differential pressure can not be performed.
d. Prepare or revise procedures to ensure that correct switch settings are
determined and maintained throughout the life of the plant.* Ensure that
applicable industry recommendations are considered in the preparation of
these procedures.
e. Within 180 days of the date of this bulletin, submit a written report to
the NRC that: (1) reports the results of item a and (2) contains the
program to accomplish items b through d above including a schedule for
completion of these items.
*this item is intended to be completely consistent with action item 3.2, "Post-
Maintenance Testing (All Other Safety-Related Components) ," of Generic Letter
83-28, "Required Actions Based on Generic Implications of Salem ATWS Events."
These procedures should include provisions to monitor valve performance to ensure
the switch settings are correct. This is particularly important if the torque
or torque bypass switch setting has been significantly raised above that required.
IEB 85-03
November 15, 1985
Page 6 of 6
1. For plants with an OL, the schedule shall ensure that these items are
completed as soon as practical and within two years from the date of
this bulletin.
2. For plants with a CP, this schedule shall ensure that these items are
completed before the scheduled date for OL issuance or within two
years from the date of this bulletin, whichever is later.
f. Provide a written report on completion of the above program. This report
should provide (1) a verification of completion of the requested program,
(2) a summary of the findings as to valve operability prior to any adjust-
ments as a result of this bulletin, and (3) a summary of data in accordance
with Table 2, Suggested Data Summary Format. The NRC staff intends to use
this data to assist in the resolution of Generic Issue II .E.6.1 . This
report shall be submitted to the NRC within 60 days of completion of the
program. Table 2 should be expanded, if appropriate, to include a summary
of all data required to evaluate the response to this bulletin.
The written reports shall be submitted to the appropriate Regional Administrator
under oath or affirmation under provisions of Section 182a, Atomic Energy Act of
1954, as amended. Also, the original copy of the cover letters and a copy of
the reports shall be transmitted to the U.S. Nuclear Regulatory Commission,
Document Control Desk, Washington, DC 20555 for reproduction and distribution.
This request for information was approved by the Office of Management and Budget
under a blanket clearance number 3150-0011. Comments on burden and duplication
may be directed to the Office of Management and Budget, Reports Management,
Room 3208, New Executive Office Building, Washington, DC 20503.
Although no specific request or requirement is intended, the time required to
complete each action item above would be helpful to the NRC in evaluating the
cost of this bulletin.
If you have any questions regarding this matter, please contact the Regional
Administrator of the appropriate NRC regional office or the technical contact
listed below.
amen M. Tay r, Director
Office of Inspection and Enforcement
Technical Contacts: H. A. Bailey, IE R. J. Kiessel , IE
(301) 492-9006 (301) 492-8119
Attachments:
1. Table 1
2. Table 2
3. List of Recently Issued IE Bulletins
Ift A.
Attachment 1
IEB 85-03
November 15, 1985
TABLE 1
Summary of Significant M0V Abnormalities
Bypass switch improperly set 75*
Incorrect thrust 50
Unbalanced torque switch 33
Valve backseating 8
High motor current 3
Torque switch abnormalities 2
Miscellaneous abnormalities 33
* Percent of valves experiencing abnormality. The total does not equal
100 percent as most valves had more than one abnormality.
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Attachment 3
IEB 85-03
November 15, 1985
LIST OF RECENTLY ISSUED IE BULLETINS
Bulletin Date of
No. Subject Issue Issued to
85-02 Undervoltage Trip Attachments 11/5/85 All power reactor
Of Westinghouse DB-50 Type licensees and
Reactor Trip Breakers applicants
85-01 Steam Binding Of Auxiliary 10/29/85 All power reactor
Feedwater Pumps facility licensees
and CP holders listed
in Attachment 1 for
action; all other
power reactor
facilities for
information
84-03 Refueling Cavity Water Seal 8/24/84 All power reactor
facilities holding
an OL or CP except
Fort St. Vrain
84-02 Failures Of General Electric 3/12/84 All power reactor
Type HFA Relays In Use In facilities holding
Class 1E Safety System an OL or CP
84-01 Cracks In Boiling Water 2/3/84 All BWR facilities
Reactor Mark I Containment with Mark I contain-
Vent Headers ment and currently
in cold shutdown
with an OL for Action
and All other BWRs
with an OL or CP for
information
83-08 Electrical Circuit Breakers 12/28/83 All power reactor
With An Undervoltage Trip facilities holding
Feature In Use In Safety- an OL or CP
Related Applications Other
Than The Reactor Trip System
83-07 Apparently Fraudulent 12/09/83 Same as IEB 83-07
Sup. 2 Products Sold By Ray Miller,
Inc.
OL = Operating License
CP = Construction Permit
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