HomeMy WebLinkAbout851157.tiff ,o_to REc 4J UNITED STATES
W '9. NUCLEAR REGULATORY COMMISSION
I 3 REGION IV
011 RYAN PLAZA DRIVE,SUITE 1000
ARLINGTON,TEXAS 78011
MAX 17 1985
In Reply Refer To:
Docket: 50-267
Public Service Company of Colorado Mqy
ATTN: 0. R. Lee, Vice President
Electric Production 211 1
P. 0. Box 840
Denver, Colorado 80201
Dear Mr. Lee:
SUBJECT: DRAFT TECHNICAL EVALUATION REPORT (TER) FOR SALEM ATWS ITEM 1.2
(GENERIC LETTER 83-28)
Re: Fort St. Vrain Nuclear Generating Station (FSV)
The staff has completed a preliminary review to assess the completeness and
adequacy of licensee responses to Generic Letter 83-28 Item 1.2. For FSV,
your response was found to be incomplete in one of the areas evaluated.
The enclosed TER provides a technical evaluation representing the staff's
initial judgement of the areas evaluated. Pen and ink changes have been made
to the TER to make the wording consistent with our approach.
In order to preserve our present review schedule, we would appreciate your
cooperation in obtaining additional information that will permit us to
complete our review. It would appear that the needed information on your
facility could be obtained by telephone conference on May 31, 1985. We will
arrange an appropriate time.
Sincerely,
E. H. Johnson, Chief
Reactor Project Branch 1
Enclosure:
DRAFT TER on Salem
ATWS Item 1.2
cc w/enclosure: (cont. on next page)
851157
X -
�,.d- /t- � =-3 /lc" ��/ /,l
Public Service Company of Colorado -2-
Mr. D. W. Warembourg, Manager
Nuclear Engineering Division
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
Mr. David Alberstein, 14/159A
GA Technologies, Inc.
P. 0. Box 85608
San Diego, California 92138
Kelley, Stansfield & O'Donnell
Public Service Company Building
550 15th Street, Room 900
Denver, Colorado 80202
Chairman, Board of County Comm.
of Weld County, Colorado
Greeley, Colorado 80631
Regional Representative
Radiation Programs
Environmental Protection Agency
1860 Lincoln Street
Denver, Colorado 80203
Mr. H. L. Brey, Manager
Nuclear Licensing/Fuels Div.
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
J. W. Gahm, Manager, Nuclear
Production Division
Fort St. Vrain Nuclear Station
16805 WCR 19}
Platteville, Colorado 80651
L. Singleton, Manager, Quality
Assurance Division
(same address)
Colorado Radiation Control
Program Director
•
SAIC-85/1514-2
REVIEW OF LICENSEE AND APPLICANT RESPONSES
TO NRC GENERIC LETTER 83-28
(Required Actions Based on Generic Implications of
Salem ATWS Events), Item 1.2
"POST-TRIP REVIEW: DATA AND INFORMATION CAPABILITIES" FOR
FORT ST. VRAIN, UNIT 1 (50-267)
Technical Evaluation Report
Prepared by
Science Applications International Corporation
1710 Goodridge Drive
McLean, Virginia 22102
Prepared for
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555
Contract No. NRC-03-82-096
FOREWORD
This report contains the technical evaluation of the Fort St. Vrain,
Unit 1 response to Generic Letter 83-28 (Required Actions Based on Generic
Implications of Salem ATWS Events), Item 1.2 'Post Trip Review: Data and
Information Capabilities."
For the purposes of this evaluation, the review criteria, presented in
part 2 of this report, were divided into five separate categories. These
are :
1. The parameters- monitored by the sequence of events and the time
history recorders,
2. The performance characteristics of the sequence of events
recorders,
3. The performance characteristics of the time history recorders,
4. The data output format, and
5. The long-term data retention capability for post-trip review
material .
All available responses to Generic Letter 83-28 were evaluated. The
plant for which this report is applicable was found to have adequately
responded to, and met, categories 2, 3, 4 and 5.
The report describes the specific methods used to determine the cate-
gorization of the responses to Generic Letter 83-28. Since this evaluation
report was intended to apply to more than one nuclear power plant s ecific _ cev',eta
regarding how each plant met (or failed to meet) the iF are not ecite-i -
presented. Instead, the evaluation presents a categorization of the
responses according to which categories of requirements are satisfied and
which are not. The evaluations are based on pecific criteria (Section 2)
derived from the requirements as stated in the eneric letter.
review
C yet ecic
TABLE OF CONTENTS
Section Page
Introduction 1
1. Background 2
2. Review Criteria 3
3. Evaluation 9
4. Conclusion 10
5. References 11
G. SoePatt gae d ec,u►he cT Pat. Tc.Cec o 1 . . • . . . . - is
•
INTRODUCTION
SAIL has reviewed theAsubmittals prepared in response to Generic Letter
83-28, item 1.2 "Post-Trip Review: Data and Information Capability'.
The submittal5 (see references)
contained sufficient information to determine that the data and information
capabilities at this plant are acceptable in the following areas.
• The sequence-of-events recorder(s) performance charac-
teristics.
• The time history recorder(s) performance characteris-
tics.
• The output format of the recorded data.
• The long-term data retention, record keeping, capa-
bility.
However, the data and information capabilities, as described in the
submittal , either fail to meet the review criteria or provide insufficient
information to allow determination of the adequacy of the data and
information capabilities in the following area.
• The parameters monitored by both the sequence-of-events
and time history recorders.
1
1. Background
On February 25, 1984, both of the scram circuit breakers at Unit 1 of
the Salem Nuclear Power Plant failed to open upon an automatic reactor trip
signal from the reactor protection system. This incident occurred during
the plant startup and the reactor was tripped manually by the operator about
30 seconds after the initiation of the automatic trip signal. The failure
of the circuit breakers has been determined to be related to the sticking of
the under voltage trip attachment. Prior to this incident; on February 22,
1983; at Unit 1 of the Salem Nuclear Power Plant an automatic trip signal
was generated based on steam generator low-low level during plant startup.
In this case the reactor was tripped manually by the operator almost coinci-
dentally with the automatic trip. At that time, because the utility did not
have a requirement for the systematic evaluation of the reactor trip, no
investigation was performed to determine whether the reactor was tripped
automatically as expected or manually. The utilities' written procedures
required only that the cause of the trip be determined and identified the
responsible personnel that could authorize a restart if the cause of the
trip is known. Following the second trip which clearly indicated the
problem with the trip breakers, the question was raised on whether the
circuit breakers had functioned properly during the earlier incident. The
most useful source of information in this case, namely the sequence of
events printout which would have indicated whether the reactor was tripped
automatically or manually during the February 22 incident, was not retained
after the incident. Thus, no judgment on the proper functioning of the trip
system during the earlier incident could be made.
Following these incidents; on February 28, 1983; the NRC Executive
Director for Operations (EDO), directed the staff to investigate and report
on the generic implications of these occurrences at Unit 1 of the Salem
Nuclear Power Plant: The results of the staff's inquiry into the generic
implications of the Salem Unit incidents is reported in NUREG-1000, 'Generic
Implications of ATMS Events at the Salem Nuclear Power Plant." Based on the
results of this study, a set of required actions were developed and included
in Generic Letter 83-28 which was issued on July 8, 1983 and sent to all
licensees of operating reactors, applicants for operating license, and
construction permit holders. The required actions in this generic letter
consist of four categories. These are: (1) Post-Trip Review, (2) Equipment
2
Classification and Vender Interface, (3) Post Maintenance Testing, and (4)
Reactor Trip System Reliability Improvements.
The first required action of the generic letter, Post-Trip Review, is
the subject of this TER and consists of action item 1.1 "Program Description
and Procedure" and action item 1.2 'Data and Information Capability." In
the next section the review criteria used to assess the adequacy of the
utilities' responses to the requirements of action item 1.2 will be
discussed.
2. Review Criteria
The intent of the Post Trip Review requirements of Generic Letter 83-28
is to ensure that the licensee has adequ `e procedures and data and
information sources to understand the caus nd progression of a reactor
trip. This understanding should go beyond a simple identification of the
course of the event. It should include the capability to determine the root
cause of the reactor trip and to determine whether safety limits have been
exceeded and if so to what extent. Sufficient information about the reactor
trip event should be available so that a decision on the acceptability of a
reactor restart can be made.
The following are the review criteria developed for the requirements of
Generic Letter 83-28, action item 1.2:
The equipment that provides the digital sequence of events (S0E) record
and the analog time history records of an unscheduled shutdown should pro-
vide a reliable source of the necessary information to be used in the post
trip review. Each plant variable which is necessary to determine the
cause(s) and progression of the event(s) following a plant trip should be
monitored by at least one recorder Ouch as a sequence-of-events recorder or
a plant process computer for digital parameters; and strip charts, a plant
process computer or analog recorder for analog (time history) variables].
Each device used to record an analog or digital plant variable should be
described in sufficient detail so that a determination can be made as to
whether the following performance characteristics are met:
3
• Each sequence-of-events recorder should be capable of detecting
and recording the sequence of events with a sufficient time
discrimination capability to ensure that the time responses asso-
ciated with each monitored safety-related system can be ascer-
tained, and that a determination can be made as to whether the
time response is within acceptable limits based on
FSAR Chapter 15 Accident Analyses. The
recommended guideline for the SOE time discrimination is approxi-
mately 100 msec. If current SOE recorders do not have this time
discrimination capability the licensee or applicant should show
that the current time discrimination capability is sufficient for
an adequate reconstruction of the course of the reactor trip. As
a minimum this should include the ability to adequately recon-
struct the accident scenarios presented in Chapter 15 of the plant
FSAR.
• Each analog time history data recorder should have a sample inter-
val small enough so that the incident can be accurately
reconstructed following a reactor trip. As a minimum, the
licensee or applicant should be able to reconstruct the course of
the accident sequences evaluated in the accident analysis of the
plant FSAR (Chapter 15). The recommended guideline for the sample
interval is 10 sec. If the time history equipment does not meet
this guideline, the licensee or applicant should show that the
current time history capability is sufficient to accurately recon-
struct the accident sequences presented in Chapter 15 of the FSAR. _
• To support the post trip analysis of the cause of the trip and the
proper functioning of involved safety related equipment, each
analog time history data recorder should be capable of updating
and retaining information from approximately five minutes prior to
the trip until at least ten minutes after the trip.
• The information gathered by the sequence-of-events and time
history data collectors should be stored in a manner that will
allow for retrieval and analysis. The data may be retained
in either hardcopy (computer printout, strip chart output, etc.)
or in an accessible memory (magnetic disc or tape). This
4
information should be presented in a readable and meaningful
format, taking into consideration good human factors practices
(such as those outlined in NUREG-0700).
• All equipment used to record sequence of events and time history
information should be powered from a reliable and non-
interruptible power source. The power source used need not be
safety related.
The sequence of events and time history recording equipment should
monitor sufficient digital and analog parameters, respectively, to assure
that the course of the reactor trip can be reconstructed. The parameters
monitored should provide sufficient information to determine the root cause
of the reactor trip, the progression of the reactor trip, and the response
of the plant parameters and systems to the reactor trip. Specifically, all
input parameters associated with reactor trips, safety injections and other
safety-related systems as well as output parameters sufficient to record the
proper functioning of these systems should be recorded for use in the post
. trip review. The parameters deemed necessary, as a minimum, to perform a
post-trip review (one that would determine if the plant remained within its
design envelope) are presented on Tables 1.2-1 through 1.2-3. If the appli-
cants' or licensees' SOE recorders and time history recorders do not monitor
all of the parameters suggested in these tables the applicant or licensee
should show that the existing set of monitored parameters are sufficient to
establish that the plant remained within the design envelope for the appro-
priate accident conditions; such as those analyzed in Chapter 15 of the
plant Safety Analysis Report.
Information gathered during the post trip review is required input for
future post trip reviews. Data from all unscheduled shutdowns provides a
valuable reference source for the determination of the acceptability of the
plant vital parameter and equipment response to future unscheduled shut-
downs. It is therefore necessary that information gathered during all post
trip reviews be maintained in an accessible manner for the life of the
plant.
5
Table 1.2-1. PWR Parameter List
SOE Time History
Recorder Recorder Parameter / Signal
x Reactor Trip
(1) x Safety Injection
x Containment Isolation
(1) x Turbine Trip
x Control Rod Position
(1) x x Neutron Flux, Power
x Y. Containment Pressure
(2) Containment Radiation
x Containment Sump Level
(1) x x Primary System Pressure
(1) x x Primary System Temperature
(1) x Pressurizer Level
(1) x Reactor Coolant Pump Status
(1) x x Primary System Flow
(3) Safety Inj.; Flow, Pump/Valve Status
x MSIV Position
x x Steam Generator Pressure
(1) x x Steam Generator Level
(1) x x Feedwater Flow
(1) x x Steam Flow
(3) Auxiliary Feedwater System; Flow,
Pump/Value Status
x AC and DC System Status (Bus Voltage)
x Diesel Generator Status (Start/Stop,
On/Off)
x PORV Position
(1): Trip parameters
(2): Parameter may be monitored by either an SOE or time history recorder.
(3): Acceptable recorder options are: (a) system flow recorded on an SOE
recorder, (b) system flow recorded on a time history recorder, or (c)
equipment status recorded on an SOE recorder.
6
Table 1.2-2. BWR Parameter List
SOE Time History
Recorder Recorder Parameter / Signal
x Reactor Trip
x Safety Injection
x Containment Isolation
x Turbine Trip
x Control Rod Position
x (1) x Neutron Flux, Power
x (1) Main Steam Radiation
(2) Containment (Dry Well ) Radiation
x (1) x Drywell Pressure (Containment Pressure)
(2) Suppression Pool Temperature
x (1) x Primary System Pressure
x (1) x Primary System Level
x MSIV Position
x (1) Turbine Stop Valve/Control Valve Position
x Turbine Bypass Valve Position
x Feedwater Flow
x Steam Flow
(3) Recirculation; Flow, Pump Status
x (1) Scram Discharge Level
x (1) . Condenser Vacuum
x AC and DC System Status (Bus Voltage)
(3)(4) Safety Infection; Flow, Pump/Valve Status
x Diesel Generator Status (On/Off,
Start/Stop)
(1): Trip parameters.
(2): Parameter may be recorded by either an SOE or time history recorder.
(3.) : Acceptable recorder options are: (a) system flow recorded on an SOE
recorder, (b) system flow recorded on a time history recorder, or
(c) equipment status recorded on an SOE recorder.
(4): Includes recording of parameters for all applicable systems from the
following: HPCI, LPCI, LPCS, IC, RCIC.
7
Table 1.2-3. HTGR Post-Trip Review Parameter List
SOE Time History
Recorder Recorder (2) Parameter/Signal
X Reactor Trip
X Turbine Trip
X Control Rod Position
X (1) X Neutron Flux,Power
X (1) X Reactor Building Temperature
X Reactor Building Radiation
X (1) X Primary System Pressure
X Core Region Outlet Temperature
X (1) X Circulator Inlet Temperature
X (1) Circulator Speed
X (1) Circulator Seal Malfunction
X (1) Circulator Bearing Water Malfunction
X (1) Circulator Drain Malfunction
X (1) Circulator Penetration Overpressure
X Primary System Flow
X (1) X Primary System Moisture
X MSIV Position
X RSIV Position
X (1) X Main Steam Pressure
X (1) X Main Steam Temperature
X (1) X Hot Reheat Steam Pressure
X (1) X Hot Reheat Steam Temperature
X (1) Hot Reheat Steam Activity
X Steam Generator Level
X (1) Steam Generator Penetration Overpressure
X Feedwater Flow
X Steam Flow
X Emergency Auxiliary Feedwater System;
Flow, Valve Status
X AC and DC System Status
X Diesel Generator Status
X Primary Relief Valve Block Vlye Position
(1) Trip parameters.
(2) Parameter may be monitored by either an SOE or time history recorder.
8
3. Evaluation
The parameters identified in part 2 of this report as a part of the
review criteria are those deemed necessary to perform an adequate post-trip
review. The recording of these parameters on equipment that meets the
guidelines of the review criteria will result in a source of information
that can be used to determine the cause of the reactor trip and the plant
response to the trip, including the responses of important plant systems.
The parameters identified in this submittal as being recorded by the
sequence of events and time history recorders do not correspond to the
parameters specified in part 2 of this report.
The information provided in the submittal indicates that the equipment
used to monitor the digital and analog parameters meets the minimal
requirements set forth in part 2 of this report. The sequence of events and
analog time history recorders are powered from a non-interruptable power
supply. The monitoring characteristics are all within the guidelines of the
review criteria.
The data and information recorded for use in the post-trip review
should be output in a format that allows for ease of identification and use
of the data to meet the review criterion that calls for information in a
readable and meaningful format. The information contained in this submittal
indicates that this requirement is met.
The data and information used during a post-trip review should be
retained as part of the plant files. This information could prove useful Cfae‘io
during future post-trip reviews. Therefore one requirenenfpresented in
part 2 of this report is that information used during a post-trip review be
maintained in an accessible manner for the life of the plant. Information
contained within this submittal indicates that this criterion will be met.
4. Conclusion
The information supplied in response to Generic Letter 83-28 indicates
that the current post-trip review data and information capabilities are
adequate in the following areas:
9
1. The SOE recorders meet the minimum performance characteristics.
2. The time history recorders meet the minimum performance character-
istics.
3. The recorded data is output in a readable and meaningful format.
4. The information recorded for the post-trip review is maintained in
an accessible manner for the life of the plant.
The information supplied in response to Generic Letter 83-28 does not
indicate that the post-trip review data and information capabilities are
adequate in the following area:
1. As described in the submittal , sufficient analog and digital
parameters are not recorded for use in the post-trip review.
It is possible that the current data and information capabilities at this
nuclear power plant are adequate to meet the intent of these requiremonts,re% Eer;a
but were not completely described. Under these circumstances, the licensee
should provide an updated, more complete, description to show in more detail
the data and information capabilities at this nuclear power plant. If the
information provided accurately represents all current data and information
capabilties, then the licensee should either show that the parameters
currently recorded will enable the licensee to determine that the reactor
trip progressed within the design limits of the Safety Analysis Report
accident analysis, or detail future modifications that would enable the
licensee to meet the intent of the evaluation criteria.
10
REFERENCES ..
NRC Generic Letter 83-28. "Letter to all licensees of operating
reactors, applicants for operating license, and holders of construction
permits regarding Required Actions Based on Generic Implications of
Salem ATWS Events." July 8, 1983.
NUREG-1000, Generic Implications of ATWS Events at the Salem Nuclear
Power Plant, April 1983.
Letter from O.R. Lee, Public Service Company of Colorado, to D.G.
Eisenhut, NRC, dated November 4, 1983, Accession Number 8311090032 in
response to Generic Letter 83-28 of July 8, 1983, with attachment.
11
SUPPORTING DOCUMENT FOR TELECON
Fort St. Vrain
1. Parameters recorded: Unsatisfactory
The following parameters are not recorded: Reactor Building
Temperature, Primary System Flow, RSIV Position, Hot Reheat Steam
Pressure, Hot Reheat Steam Temperature, Hot Reheat Steam Activity, Steam
Generator Level , and Primary Relief Valve Block Valve Position
2. SOE recorders performance characteristics: Satisfactory
Plant computer: 2msec time discrimination with a non-interruptible
power supply
NOVA Sequence of events:
3. Time history recorders performance characteristics: Satisfactory
Plant process computer: parameters are sampled every 5 seconds for 6
minutes before the trip until 10 minutes after the trip.
Strip charts are also used.
4. Data output format: Satisfactory
SOE data output includes time of event, event descriptor and sensor ID
Analog data output includes times, parameter name and value, and sensor
ID
5. Data retention capability: Satisfactory
Data is retained for the life of the plant.
12
SSINS No. : 6835
N16%{?t 5;„H plement 1
UNITED STATES
NUCLEAR REGULATORY COMMISSION (
OFFICE OF INSPECTION AND ENFORCE MAY
CD if
WASHINGTON, D.C. 20555 i 1985
1'
May 14, 1985
coLo
IE INFORMATION NOTICE NO. 84-55, SUPPLEMENT 1: SEAL TABLE LEAKS AT PWRs •
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or a
construction permit (CP).
Purpose:
This supplement to Information Notice (IN) 84-55 provides additional information
concerning seal table leaks and failures of compression type mechanical fit-
tings. A correction also has been included in this supplement concerning the
use of the term "SWAGELOK" fitting. It is expected that recipients will review
the information for applicability to their facilities and consider actions, if
appropriate, to preclude a similar problem occurring at their facilities. How-
ever, suggestions contained in this supplement do not constitute requirements;
therefore, no specific action or written response is required.
Background:
IN 84-55 described two events that led to primary reactor coolant leaks at the
seal table. The leaks were caused by failure of the mechanical seals during
maintenance under high-pressure conditions. These events occurred at Sequoyah
1 and Zion 1. INPO Significant Event Report (SER) 43-84 and Supplement 1 to
SER 43-84 also discuss these two events as well as the Trojan event (LER
84-014) that occurred on September 13, 1984.
Further review of the Sequoyah event by the Tennessee Valley Authority (TVA)
determined the exact cause of the high-pressure seal failure. Their review
showed that the design of the as-supplied cleaning tool had been modified by
Sequoyah personnel and that these modifications produced a tool that would
transfer force to the thimble tube high-pressure seal at a level sufficient to
cause separation of the seal assembly. The original cleaning tool , supplied by
Teleflex, Inc. , consisted of a drivebox and flexible plastic tube that connected
to the upper flare fitting at the end of the thimble tube. Use of the handcrank
on the drivebox moved the cleaning tool into or out of the thimble tube. Excessive
force could not be transmitted to the fitting of the high-pressure seal because
the flexible plastic tube would be the weakest component subjected to stress.
However, because of the flexibility of the plastic tube, the workers had difficulty
feeding the cleaning device through the drivebox into the tube. During 1978 or
1979, plant personnel , to compensate for this difficulty, had modified the tool
by a metal extension sleeve that threaded onto the thimble tube flare fitting.
This modified tool was remodified on several occasions because of missing pieces
or further attempts to improve tool stability.
'5130027
7
IN 85-55, Supplement 1
May 14, 1985
Page 2 of 4
The tool that was used during the Cycle 3 refueling consisted of a base that
slipped over the thimble tube assembly at the seal table and a metal extension
piece that threaded on to the thimble tube flare fitting to which the drivebox
was attached. This tool was lost during containment cleanup activities before
plant startup. When the decision was made to clean the tubes at power, another
tool was fabricated with a slightly shorter base, which did not mate flush
with the upper extension piece. Shims were used to try to correct this problem,
but they appear to have still allowed some fulcrum effect causing the seal to
fail .
In addition to the events discussed in IE IN 84-55, several other similar events
occurred during the past year. These are briefly described below.
1. On October 23, 1984 an unisolatable reactor coolant leak occurred at
Catawba Nuclear Station, Unit 1 (LER 84-18). The unit was in hot standby
with reactor coolant system (RCS) pressure at 1500 psig and temperature at
440°F. The leak occurred when the stainless steel conduit containing an
incore thermocouple separated from the mechanical compression tube
fitting. An approximate 5 gpm leak rate occurred and a total of 12,000
gallons of coolant from the RCS leaked to the containment floor and equip-
ment sump.
Evaluation of the failure indicated that the conduit had not been fully
inserted into the fitting and the fitting had been tightened only 13/20
turns. The required number of turns is 1 1/4 turns. Another thermocouple
also was found to be loose. The cause of the failure was evaluated as a
construction/installation deficiency.
2. Two events occurred at the Rancho Seco Nuclear Plant. The first occurred
on April 20, 1984, when a 3/8 stainless steel sensing line blew out of a
compression fitting under system pressure (2200 psig) following routine
recalibration of a pressurizer level transmitter. The second event
occurred on July 31, 1984. A leak was noticed on a steam generator level
transmitter sensing line fitting. An attempt to tighten the fitting wor-
sened the leak. Before attempting to further tighten the fitting the
technician tapped the sensing line with a wrench, whereupon the stainless
steel line blew out of the fitting.
Both of the fittings involved were installed during or after a 1983
refueling outage. A decision was made to inspect 1444 compression fittings
installed during that period. The inspection revealed that approximately
3% of the tube fittings were improperly made up as a result of poor work-
manship. A few instances of improperly oriented or missing ferrules were
found as gross errors , but the majority of the deficiencies found concerned
the improper location of the ferrule with respect to the end of the tube.
These problems were attributable to poor workmanship in the cutting of the
tubing and also to potential spring back of the tubing after firm seating
in the fitting during initial assembly.
IN 85-55, Supplement 1
May 14, 1985
Page 3 of 4
3. On October 25, 1983, a dewater and air drying operation was being conducted
at 39 thimble locations on the seal table at D. C. Cook Unit 1. The reactor
was at about 50% power throughout the 2-day effort. At the conclusion of
the maintenance effort, all guide tube connections were checked and found
satisfactory except for one. This fitting had apparently not been properly
assembled because the middle high-pressure seal came loose causing the
thimble to rise approximately 5 inches above the seal table before it
stopped. Clamps were reconnected to prevent any further rising of the
thimble out of the core.
Discussion:
Compression fittings are used throughout nuclear plants in primary, secondary,
and auxiliary systems. In some applications , failure of a fitting may be
nothing more than a nuisance. In other applications serious events could
follow the failure of one of these fittings; however, past failures of fittings
cannot be traced to any one single source. The examples cited indicate that
the failures are varied and can include the following: improper maintenance,
improper installation, improperly designed tools , and improper materials.
In the Sequoyah event, modifications made to the cleaning tool were not
controlled adequately by existing formal programs. There were no records that
any of the modifications to the tool had been technically evaluated or tested
to determine their effect on the thimble tube seal . Lack of such controls
appears to have been a significant precursor to the event that occurred.
In the Trojan, Rancho Seco, D. C. Cook, and Catawba events, the immediate cause
was improper installation of the fittings coupled with corrective maintenance
on the fittings while the RCS was hot and pressurized. Compression fittings
exhibit a mode of failure involving pullout of the tubing from the fitting when
subjected to an axial load. Tensile tests have demonstrated this mode of
failure can occur when the ferrule is improperly located on the tube during
initial assembly. When this occurs , it can cause a substantial reduction in
the safety factor of the fitting. Thus , maintenance on the fitting under these
conditions has led to a substantial number of fitting failures.
Some events involving the failure of compression fittings have become known
because of the circumstances that followed the fitting failure. However,
because of the varied applications of the compression fitting, many failures
of fittings , or problems associated with the fittings, probably go unreported.
Because of the number of severe events that have occurred during the past year,
it is appropriate that positive steps be taken to reduce the number of similar
events from occurring.
In this regard, some utilities have revised procedures on the assembly and
maintenance of these fittings to address the problems mentioned above. They
also have upgraded training efforts on the assembly and maintenance of the fit-
tings and have enacted stricter controls on when maintenance can be conducted
(e. g. , restricting maintenance on fittings used in hot, pressurized systems).
Additionally, Westinghouse is developing fitting inspection guidelines and
detailed acceptance criteria.
4N
IN 85-55, Supplement 1
May 14, 1985
Page 4 of 4
In the original IE information notice, reference was made to the failure of
"SWAGELOK" fittings. It was found that at Sequoyah and Trojan the compression
fittings used were hybrid fittings assembled using parts from other vendors,
including Crawford Fitting Co. Brand names of other compression fittings
include SNO-Trik, Gyrolok, Megalok, Ringlok, Unilok, Wadelok, etc. Apparently,
many of the parts are used interchangeably. In the case of Sequoyah, "Gyrolok"
and "Swagelok" parts were used to construct the fittings. Therefore, use of
the name "SWAGELOK" fitting in the above applications was inaccurate. A
"SWAGELOK" fitting has a nut, front and back ferrule, and body, all manufactured
by Crawford. In fact, Crawford cautions against interchanging parts of tube
fittings made by other manufacturers with 'SWAGELOK" tube fitting parts.
No specific action or written response is required by this information notice.
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.
EdWard L.' Jordan, Director
Division of -Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
Technical Contact: D. Powell , IE
(301) 492-7155
Attachment: List of Recently Issued IE Information Notices
Attachment 1
IN 84-55, Supplement 1
May 14, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
85-20 Motor-Operated Valve Failures 5/14/85 All power reactor
Sup. 1 Due To Hammering Effect facilities holding
an OL or CP
85-36 Malfunction Of A Dry-Storage, 5/9/85 All licensees
Panoramic, Gamma Exposure possessing gamma
Irradiator irradiators
84-52 Inadequate Material 5/8/85 All power reactor
Sup. 1 Procurement Controls On facilities holding
The Part Of Licensees And an OL or CP
Vendors
85-35 Failure Of Air Check Valves 4/30/85 All power reactor
To Seat facilities holding
an OL or CP
85-34 Heat Tracing Contributes To 4/30/85 All power reactor
Corrosion Failure Of Stainless facilities holding
Steel Piping an OL or CP
84-84 Deficiencies In Ferro- 4/24/85 All power reactor
Rev. 1 Resonant Transformers facilities holding
an OL or CP
85-33 Undersized Nozzle-To-Shell 4/22/85 All power reactor
Welded Joints In Tanks And facilities holding
Heat Exchangers Constructed an OL or CP
Under The Rules Of The ASME
Boiler And Pressure Vessel
Code
85-32 Recent Engine Failures Of 4/22/85 All power reactor
Emergency Diesel Generators facilities holding
an OL or CP
85-31 Buildup Of Enriched Uranium 4/19/85 All uranium fuel
In Ventilation Ducts And fabrication licensees
Associated Effluent Treatment
Systems
OL = Operating License
CP = Construction Permit
SL S No. : 6835
IN 85-20,,,,Supplement 1
UNITED STATES r
NUCLEAR REGULATORY COMMISSION J r
OFFICE OF INSPECTION AND ENFORCEMENT 't Mq
WASHINGTON, D. C. 20555 i 21119857
May 14, 1985 GRcccevco`o
IE INFORMATION NOTICE NO. 85-20 SUPPLEMENT 1: MOTOR-OPERATED VALVE FAILURES
DUE TO HAMMERING EFFECT
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or a
construction permit (CP).
Purpose:
This supplement to IE Information Notice (IN) 85-20 is provided to alert
recipients of additional information on a potentially significant problem
pertaining to motor-operated valve failures caused by the hammering that may
result when a fully closed (opened) valve continues to receive a close (open)
signal at the valve operator. It is expected that recipients will review the
information for applicability to their facilities and consider actions, if
appropriate, to preclude a similar problem occurring at their facilities.
However, suggestions contained in this notice do not constitute NRC require-
ments; therefore, no specific action or written response is required.
Background:
IN 85-20 described a number of valve failures at the Dresden Nuclear Power
Station Unit 2 and Quad Cities Nuclear Power Station Unit 1 sites. It identi-
fied a process by which the relaxation of torque on a closed valve would lead
to repeated attempts to further close the valve as long as the valve operator
continued to receive a valve-close demand signal . Such a continuing signal
would occur if the plant operator held the control switch in the closed posi-
tion or an emergency signal (i . e. , containment isolation, etc. ) was present.
Discussion:
Subsequent conversations with several utilities and Limitorque Corporation, the
manufacturer of the motor operators used on the valves identified in the
information notice, have clarified the conditions under which this hammering
effect can occur. The majority of valve motor operators manufactured by
Limitorque use a self-locking worm-and-worm gear to drive the valve stem. In
such installations, when the valve is fully closed, the Belleville washer is
compressed and the torque switch opens , stopping the motor. The geometry of
the worm gear tooth form prevents the worm from moving when the motor stops.
Thus, the torque is maintained on the valve, the torque switch remains open,
and hammering is prevented.
8505130038
IN 85-20, Supplement 1
May 14, 1985
Page 2 of 2
However, Limitorque also produces an actuator which uses a low-ratio worm gear
that is not self-locking. This is generally associated with high speed valve
applications. In this type of actuator, the worm may be repositioned by the
energy stored in the compressed Belleville washers, which are located at one
end of the worm. Because the operator essentially uses the position of the
worm as an indication of torque, this movement of the worm closes the torque
switch contacts. Thus , if a valve-close demand signal is still present, the
motor will restart and attempt to further close the valve.
Table 1 provides a list of worm gear ratios that are not self-locking. Conver-
sations with Commonwealth Edison confirmed that the valves identified in the
information notice had high-speed operators on them.
Conversations with several utilities who were previously aware of this problem
indicated that their normal design practice was to close on torque unless the
valve had a high-speed operator. In such cases, they closed on valve position.
However, now the current philosophy is to use a combination of torque and
position to ensure closure while preventing hammering. Limitorque Corporation
indicated that actuators without self-locking ratios and with a motor brake,
generally do not experience this hammer effect. As noted in the information
notice, Quad Cities was not aware of this in their change to the operator logic
until after they removed the motor brake from the valves.
No specific action or written response is required by this information notice
supplement. If you have any questions about this matter, please contact the
Regional Administrator of the appropriate NRC regional office or this office.
Edwar/. Jordan Director
Divisio of Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
Technical Contact: R. J. Kiessel , IE
(301) 492-8119
Attachments:
1. Table 1, "Worm Gear Ratios Which Are Not Self-locking"
2. List of Recently Issued IE Information Notices
At .chment 1
IN 85-20, Supplement 1
May 14, 1985
Page 1 of 1
TABLE 1 WORM GEAR RATIOS WHICH ARE NOT SELF-LOCKING
MODEL
SMB OR SB WORM GEAR
SIZE RATIO
000 18 2/3 : 1
00 19 : 1
0 18 2/3 : 1
1 14. 5 : 1
2 13. 3 : 1
3 10. 3 : 1
16 : 1
4 12 2/3 : 1
19 : 1
5 none
Attachment 2
IN 85-20, Supplement 1
May 14, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
85-36 Malfunction Of A Dry-Storage, 5/9/85 All licensees
Panoramic, Gamma Exposure possessing gamma
Irradiator irradiators
84-52 Inadequate Material 5/8/85 All power reactor
Sup. 1 Procurement Controls On facilities holding
The Part Of Licensees And an OL or CP
Vendors
85-35 Failure Of Air Check Valves 4/30/85 All power reactor
To Seat facilities holding
an OL or CP
85-34 Heat Tracing Contributes To 4/30/85 All power reactor
Corrosion Failure Of Stainless facilities holding
Steel Piping an OL or CP
84-84 Deficiencies In Ferro- 4/24/85 All power reactor
Rev. 1 Resonant Transformers facilities holding
an OL or CP
85-33 Undersized Nozzle-To-Shell 4/22/85 All power reactor
Welded Joints In Tanks And facilities holding
Heat Exchangers Constructed an OL or CP
Under The Rules Of The ASME
Boiler And Pressure Vessel
Code
85-32 Recent Engine Failures Of 4/22/85 All power reactor
Emergency Diesel Generators facilities holding
an OL or CP
85-31 Buildup Of Enriched Uranium 4/19/85 All uranium fuel
In Ventilation Ducts And fabrication licensees
Associated Effluent Treatment
Systems
85-30 Microbiologically Induced 4/19/85 All power reactor
Corrosion Of Containemnt facilities holding
Service Water System an OL or CP
OL = Operating License
CP = Construction Permit
Hello