HomeMy WebLinkAbout851167.tiff OEfkP NEON, UNITED STATES
W NUCLEAR REGULATORY COMMISSION
_
REGION IV
t
°r 011 RYAN PLAZA DRIVE,SUITE 1000
i " O ARLINGTON,TEXAS 70011
JUL 121985
In Reply Refer To:
Docket: 50-267 WELD COMITY rii1, ";c 3;
NERD
JULi 61985 '
Public Service Company of Colorado
�
�iJ'
ATTN: 0. R. Lee, Vice President ,
Electric Production SRMILET. (O.
P. 0. Box 840
Denver, Colorado 80201
Dear Mr. Lee:
Our consultants at EG&G Idaho, Inc. have reviewed your June 13, 1985 (P-85201)
submittal related to the control rod drive and orifice assembly (CRDOA)
bearings. A copy of their evaluation report is enclosed for your information
and comment.
Based on our review of the information you have provided and the enclosed
evaluation, we have concluded that the installed bearings are acceptable
replacements for the original design. In order to ensure continued acceptable
bearing performance, we request that the test programs discussed in P-85201 be
completed and that bearing performance be factored into your improved CRDOA
surveillance and preventative maintenance program.
We have also completed our review of your June 7, 1985 (P-85195) submittal
related to the epoxy used to attach temperature sensors in the CRDOAs. We
agree that this material appears to be adequate for the attachment of the
sensors. It is our understanding that the epoxy will be included in the
upcoming environmental qualification program. We recommend that the epoxy be
inspected for deterioration as part of the CRDOA preventative maintenance
program.
If you have any questions on this subject, please contact us.
Sincerely,
E. H. Johnson, Chief
Reactor Project Branch
Enclosure:
Evaluation of CRDOA Bearings
cc w/Enclosure: (cont. on next page)
851167
Li ,, 10 °l I„IRs
Public Service Company of Colorado -2-
cc:
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
EVALUATION CRDOA BEARING REPORT FROM
PUBLIC SERVICE COMPANY OF COLORADO
Bert L. Barnes
July 1 , 1985
EVALUATION OF CRDOA BEARING REPORT FROM
PUBLIC SERVICE COMPANY OF COLORADO
Summary
An independent review of the Public Service Company of Colorado letter
report indicates the conclusions reached within the report to be generally
valid and that the replacement bearings are statistically and operationally
as good as the original CRDOA bearings. These conclusions are based upon
an independent review of technical literature on ball bearing failure, dry
molybdenum disulfide lubrication, ball bearing materials, bearing
clearances and tolerances, life testing, life calculations, contact
stresses, duty cycles and loadings. Where enough data was provided.
calculations described within the Reference 1 report were independently
spot checked and found to be correct. The Public Service Company of
Colorado CRDOA Report provides positive support for the NRC Regulatory
approval for all proposed changes to the CRDOA hardware described by this
report.
Object
The object of this report was to review the Public Service Company of
Colorado report (Reference 1 ) to determine if the Fort Saint Vrain control
rod device and orifice assembly replacement ball bearings were equivalent
to or at least as good as the original bearings used prior to January 1985.
Introduction
This letter report was prepared under the provisions of the NRC
Form 189 for FIN A6701, Task Order R. to provide technical assistance to
the NRC Region IV office.
The report (Reference 1 ) reviewed herein summarily describes actions
taken by Public Service Company of Colorado and their subcontractors (GA
Technologies, Industrial Tectonics, Inc. , and SKF Industries) to justify
the use of replacement ball bearings differing only slightly from the
original ball bearings for the Fort Saint Vrain control rod drives and
orifice assemblies.
The items discussed below follow the format and chronological order of
the Reference 1 report. Page and paragraph numbers cited refer to this
report.
Commentary Review
All the bearing raceways and some of the balls are 440C stainless
steel (pg. 2, paragraph 3). This material is recognized as an excellent
ball bearing material (Reference 2), though it is not as commonly used as
E52100 type steels. The 440C materials have an upper limit on hardness of
Rockwell C60 (200°F temper), while the E52100 steels have a Rockwell C64
hardness upper limit (200°F temper) . Reference 2 (pg. 433-438) generally
indicates that bearing load capacity and fatigue life are both very
1
sensitive to hardness. A few points of increase to hardness can easily
double the load capacity of a given material and maximum fatigue life is
also obtained with the materials of highest hardness. However, in the
presence of moisture and oxidizing contaminants, 440C bearing materials may
be superior (Reference 2, pg. 324) to the E52100 steels because of
corrosion and oxidation resistance. In either case. 440C is an excellent
material for this application. The tungsten carbide balls, 17-4 PH and
nitralloy 135M cages are also generally recognized (Reference 2) as being
good materials for these applications. Though Reference 1 was generally
lacking in detail , the sintered bronze lubricant reservoirs are expected to
be equivalent to the "Bearite" material used earlier in the original
designs.
The choice of dry molybdenum-disufide powder, (pg. 2, paragraph 3) as
a lubricant is also excellent for this application, given that the high
radiation level and operating temperature make the use of common oil and
grease lubricants impractical . The applicable temperature range for this
lubricant exceeds 600°F (Ref. 2, pg. 212 and pg. 234, Ref. 9, pg. 26.
Ref. 3, pg. 145-146, Ref. 4, pg. 163-164) . Molybdenum disulfide appears to
be the best choice in dry lubricants for temperatures around 200°F to
250°F; for higher temperatures, Pb0 and CaF2 exhibit superior
performance. It should be noted that molybdenum disulfide lubricants are
not all the same; depending upon the binder material and other additives,
the performance of this lubricant can be variable (Reference 2, pg. 212) .
The Reference 1 report gives no details regarding the specific molybdenum
disulfide lubricant used and how it compares with the original lubricant.
However, the testing program does at least partially verify that the
lubricant used was of high grade and applicable for this service.
The dimensional differences between original and replacement bearings.
including reductions in the number of balls, do not appear to contribute to
altering the reliability of the control rod drive mechanisms (pg. 3.
paragraph 1 ). Though the dimensional data given in Appendix A of
Reference 1 are incomplete, the tolerances and clearances for corresponding
parts in the original and replacement bearings are very similar . A
comparison of ISO Tolerance Tables (starting on page 228 of Ref. 5)
indicates that there are no significant differences in the original and
replacement bearings. This conclusion is also supported by data from
Reference 6 on bearing tolerances, though, as might be expected, the
tolerances for instrument bearings listed in Reference 6 are in general
tighter than those used for these dry lubricated ball bearings.
Though Appendix A (Reference 1 ) does contain limited data on
comparative surface finishes for the retaining rings and cages , no such
data is listed for the balls and raceways. Reference 2, pg. 436-438
indicates that bearing life (fatigue) is very much a function of surface
finish. Better surface finishes in general exhibit significantly longer
fatigue life. Ball bearing rolling resistance, and in particular the
"breakaway torque" for dry lubricated ball bearings, are very much a
function of ball and raceway surface finish. Not all surfaces on a ball
bearing are critical ; the roughest surfaces have about a 32 finish while
ball and raceway finishes are generally much better than this . A
2
comparison of replacement and original ball bearings with regard to this
important factor is not possible without this data. Also, a comparison of
the actual critical surface finishes for original and replacement balls and
cages is more important than a comparison of the specified values. In this
regard, Reference 1 is deficient.
The finish specifications given for the retaining ring (Part
Number SLR-01210-222, Appendix A, pg. 2) are at best very very rough. The
63 and 32 outside diameter surface finishes for the cage (Appendix A.
pg. 3) are also relatively rough for critical parts such as these.
However, the cage outside diameter surface finish is probably relatively
unimportant compared to that of the ball pockets. All parts which touch
the ball bearings themselves and the inner and outer raceway critical
surfaces should have RMS surface finishes of 4.0 or better. In particular,
the cage ball pockets, the sintered bronze lubricant reservoirs, and other
surfaces of the cage which rub or roll against the inner or outer races
should have RMS surface finishes of 4.0 or better, but certainly no rougher
than RMS 10 at the very worst. The retaining ring specified surface finish
(63 RMS original, 125 RMS replacement) is probably not critical . The
actual surface finish is probably much better than this. However, lacking
the lack of surface finish data for other critical parts is an error of
omission for a design parameter which strongly influences bearing life and
'breakaway torque", the primary failure mechanism of this entire control
rod device. A 125 RMS surface finish is typically what results from a
rough lathe turning or roughing cut on a milling machine with no surface
polishing or sanding.
In order to facilitate comparison of original and replacement
bearings, all critical surfaces should have a listing of both specified and
actual as-measured surface finishes. While it may be that the replacement
bearings have critical surface finishes equal to or better than the
original bearings, Reference 1 provides no evidence of this.
Reference 1 (pg. 3, paragraph 3) indicates the design parameters
considered critical in assessing bearing performance to be, applied loads,
internal geometry, materials, and lubrication. In this application in
particular, contamination should also be considered as a critical
parameter. The failure of the control rod drives has been suspected of
being associated with or linked to the presence of water and water vapor in
and around these bearings. Reference 2, pg. 215-216 indicates that
contaminants are generally undesirable. Some contaminants can
significantly increase the wear rate of the bearing though no specific
reference was found in the literature search to water or water vapor
causing premature failures. In this respect, both original and replacement
bearings must be considered equally resistant to contamination.
The 80/20 figure cited (pg. 4, paragraph 2) for assumed duplex bearing
loading distribution seems like an excellent choice. The actual load
distribution is probably much better than this. However, lacking more
specific data, this assumption is probably conservative.
3
The radial load data (pg. 4, paragraph 2) is impossible to evaluate
without more data. However, the loads listed seem reasonable by
superficially considering the geometry of the parts, the loads applied to
the cable by the control rods.
and
x h the re ct ond at coft st fnt
duiratio
each
ach stage
esof the, pg.gear box. If the diagram (Appendix
these loads could be verified. The accuracy of the fatigue life
calculations is directly related to the accuracy of the bearing loadings.
In similar fashion, the validity of the testing program is dependent on
accurately modeling the actual bearing loads.
The bearing operational life cycle data (pg. 5, paragraphs 1 and 2) is
presented in sufficient detail that the numbers can be verified by
calculation. Verification calculations indicate all the numbers calculated
within these paragraphs to be both reasonable and correct.
The general worth and applicability of the fatigue life calculations
described in Reference 1 (pg. 5, paragraph 3 to pg. 8, paragraph 4) are
deserving of comment.
1 . The fatigue life of a bearing is related to a subsurface material
failure and as such is not the surface failure mechanism observed
for these bearings. This statement is supported by the
observation that the bearings with the lowest predicted life
(pg. 8 tabular L-10 lives) have never failed in service. In
fact, the only bearings that have failed (due to increases in
surface roughness) have fatigue lives of between three and four
orders of magnitude greater than the bearings that are predicted
to be the shortest lived. Fatigue life calculations are worth
while to facilitate comparison of original and replacement
bearings, but fatigue failures apparently have little or nothing
to do with the failures observed for the control rod drive
bearings . The fatigue life of original and replacement bearings,
while not exactly equal, are roughly equivalent (pg. 8. tabular
data).
2. The results of the fatigue life calculations are based upon the
commonly used 10% failure criterion and as such are statistically
highly variable. To quote from Reference 2, pg. 167:
"If a number of similar bearings are tested to fatigue
at a specific load, there is a wide dispersion of life
among the various bearings. For a group of 30 or more
bearings, the ratio of the longest to the shortest life
may be of the order of 20 or more. A curve of life as
a function of the percent of bearings that failed can
be drawn for any group of bearings. For a group of 30
or more bearings, the longest life would be of the
order of four or five times the average life. The term
life, as used in bearing catalogs, usually means the
life that is exceeded by 90 percent of the bearings.
This is the so-called 8-10 or 10-percent life. The
10-percent life is one-fifth the average of 50-percent
life for a normal life-dispersion curve".
4
Considering that the 37 control rod drives each employ 14 ball
bearings, three of which are acknowledged to be critical, the
probability of an early fatigue, life failure significantly
differing from the mean predicted life is very high. If the
predicted fatigue lives were not so long (51 years minimum
continuous operation), this would justify concern. Even if the
fatigue life were reduced by a factor of 100 for the 111
(37 units x 3 resulting bearings/unit)
shortestfa critical motor bearings, the
tiguelife would exceed 1000 years.
3. The statements made (pg. 8, paragraph 2) regarding the
applicability of fatigue life calculations for oil lubricated
bearings not being directly applicable o dry
film
lubricated
dies
bearings are true. Reference 2, (pg. 383, paragraph b)
this statement by saying:
"All experimental evidence obtained to date indicates
that the inverse cubic relation between load and life.
which was found to exist for point contact with
conventional bearing materials with mineral-oil
lubrication, is approximately true for other materials
and lubricants and for bench-type fatigue testers used
for studying the effects of different variables on
rolling-element-bearing fatigue".
4. The Hertzian contact stress criterion of 368 KSI being one-third
the Brinnel hardness number was not verified by a review of
NASA-SP38 (Reference 2). This is a large reference; perhaps the
author missed finding the words to verify this. Also, since no
data was given for bearing material hardness, it was not possible
to verify that 368 KSI was one-third the Brinnel hardness
number. The 368 KSI contact stress cited seemed reasonable when
compared to Hertzian stresses described in Reference 2 (pg. 383
to 384). It should be noted that calculated Hertzian contact
stresses are not particularly sensitive to applied radial loads.
Minor errors in the calculation of the applied radial loads
(pg. 4, paragraph 2) will not result in significant errors for
calculated Hertzian contact stresses because the contact stress
varies as the cube root of the normal force (radial load). The
predicted fatigue life, on the other hand, is very sensitive to
errors in predicting radial loads. The predicted fatigue life of
a ball bearing varies inversely as the cube tabulardata) onthe
the
load.
Because
the lowest predicted fatigue life (pg.
order of 51 years of continuous operation,rati onl ) sd rror
n, y gross
ses in
the radial load prediction (pg. 4.
reason for alarm. However, as stated earlier, the radial loads
were not verifiable from the data provided in Reference 1 . If
for any reason the actual loads are later found to be higher than
the predicted loads, these calculations should be repeated.
The physical testing programs described (pgs. 9 to 11 ) seem
reasonable. However, the conclusions drawn from a small sample size can be
5
very misleading and very inaccurate. The quote cited earlier from
Reference 2, pg. 161 provides ample evidence that the life of seemingly
identical ball bearings is statistically highly variable. To test two or
three bearings of a given type and then to predict or even hint that 111
3 per drive unit and 37 drive units) of these bearings will last as long is
not prudent. This is probably the biggest single factor that makes the
test results inconclusive.
The 30 oz-in. bearing torque criterion (pg. 9, paragraph 2) described
seems to be a questionable choice for a failure criterion. The shim motors
each employ three bearings; a cumulative (3 bearings) frictional bearing
torque exceeding 15 in.-oz constitutes a failure of the control rod drive
mechanism. For this reason, a more realistic failure criterion might be
based upon the total frictional torque from any three bearings exceeding
15 in.-oz. Any conclusions drawn on the basis of tests with a two bearing
30 in.-oz failure criterion are inconclusive. The point in time when the
cumulative bearing torque from two bearings exceeded 2/3 x 15 in.-oz =
10 in.-oz might also be meaningful .
The reasoning behind the choice of 65 lb and 15.3 lb radial loads was
not clear (pg. 10, paragraphs 3 and 4) . The table on page 4 lists loads of
50.8 lb and 12.2 lb as being closest to the test loads for the shim motor.
Though the failures experienced in shim motor bearings are not directly
related to fatigue life failures, the fact that fatigue life is known to
vary inversely with the cube of the radial load should provide motivation
to very carefully select the test loads. Perhaps the differences in these
figures is an indication of additional static loading not included in the
load table on page 10. The Reference 1 report provides no explanation for
these differences.
Some tests did use a helium environment and select other operational
duty cycle test parameters which were conservative (more demanding) than
the actual duty cycles. However, no tests were done at elevated
temperatures duplicating that of the real operational environment of 200°
to 250°F. The lubrication of the bearings has been acknowledged as being
critical to bearing life. Though molybdenum disulfide lubrication is
capable of much higher temperatures (exceeding 600°F for some
applications), certainly life tests run at room temperature are not
necessarily applicable to operational temperatures of 200° to 250°F. Also,
prolonged temperatures of 250°F will reduce the hardness of the 440C
bearing races if a temperature of less than 250°F was used to temper these
parts when originally heat treated. As stated earlier, ball bearing life
has been shown to be directly related to material hardness. A few points
reduction in hardness can significantly reduce ball bearing life. In this
report, both original and replacement bearings are expected to be equal .
The lack of including both radiation and pressure effects in the test
series seems justified because the author has found no evidence to indicate
that bearing life is sensitive to these parameters.
6
Conclusions
This author is in general agreement with all the conclusions from
physical testing (pg. 11, paragraphs 2 to 4). In particular, this author
believes the replacement bearings are roughly equivalent to the original
bearings. Though the test results are inconclusive, and some of the test
parameters and criteria are questionable, the author believes the
replacement bearings to be equivalent to the original and suitable for use
in the control rod drives.
7
REFERENCES
1 . D. W. Warembourg, Public Service Company. Colorado, letter report to
E. H. Johnson, USNRC, Region IV, Evaluation of CRDOA Bearings.
June 13, 1985.
2. E. E. Bisson and W. J. Anderson. Advanced Bearing Technology,
Cleveland, Ohio, NSA SP-38, Lewis Research Center. 1964.
3. P. Freeman, Lubrication and Friction, London. Whitefriares Press Ltd. ,
1962.
4. F. P. Bowden and D. Tabor, The Friction and Lubrication of Solids,
Clarendon, Oxford University Press, 1958.
5. A. Palmgren, Ball and Roller Bearing Engineering, Philadelphia, third
edition, S. H. Burbank and Co. Inc.. 1959.
6. Anti-Friction Bearing Manufacturers Association, Inc. , "AFBMA
Standards for Instrument Ball Bearings.' Section 12, Revision 2,
AFBMA, New York. October 1966.
7. Plant Engineering Training Systems. Unit 4, Bearings Lubrication,
Edited by Sayre, Clifford R. , Technical Publishing Company,
Barrington, Ill . , 1970.
8. Plant Engineering Training Systems, Unit 2, Lubrication, Edited by
Sayre, Clifford R. , Technical Publishing Company, Barrington, Ill . ,
1970.
9. FMC Corporation, Link-Belt Bearing Division, Bearing Technical
Journal , Indianapolis, Indiana, First Edition, 1970.
8
,.P no
?Jost �4r UNITED STATES
W" it NUCLEAR REGULATORY COMMISSION
I i
; REGION IV
•• -g e? 611 RYAN PLAZA DRIVE,SUITE 1000
e° ARLINGTON,TEXAS 78011
JUL 1 0 1985 ,11119 COUNTY COMMISSIONERS
Docket 50-267 D Cif-PP��R \
JUL 161985
Mr. 0. R. Lee, Vice President �e � r•
Electric Production
Public Service Company of Colorado
P.O. Box 840
Denver, Colorado 80201
Dear Mr. Lee:
As a result of the problems encountered during the testing of the emergency
diesel generator sets (EDGs) on December 18, 1984, the NRC has reevaluated the
Fort St. Vrain (FSV) emergency electrical systems. The results of our review
are contained in the enclosed Safety Evaluation (SE). We have concluded that,
even with the potential EDG single failure and independence problems
identified in the SE, FSV can be operated safely for an interim period while
corrective measures are being pursued.
Your commitment to provide an evaluation of the above problems, within 90 days
of your June 14, 1985 (P-85208) submittal , will be confirmed in the listing of
various commitments that will be issued with the authorization to restart FSV.
In addition to the safety questions addressed in the enclosed SE, the question
of plant operation in possible nonconformance with the established licensing
basis (FSAR) must be examined from a regulatory compliance viewpoint. We have
discussed this matter with members of your staff and are awaiting additional
information.
If your have any questions on this matter, please contact the NRC Project
Manager, P. Wagner, at (817)860-8127.
Since the reporting requirement relates solely to FSV, OMB clearance is not
required under P.L.96-511.
Sincerely,
ti: ea. o
E. H. Johnson, Chief
Reactor Projects Branch
Enclosure:
SE on EDGs
cc: (see next page)
I 1 _ 1_ 4 I lY I4 C
cc:
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
0fl. mewfr UNITED STATES JUL 0 5 1985
O
* NUCLEAR REGULATORY COMMISSION
ft
•
i REGION IV
011 RYAN PLAZA DRIVE,SUITE 1000
fib ARLINGTON,TEXAS 76011
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
PUBLIC SERVICE COMPANY OF COLORADO
FORT ST. VRAIN (FSV) NUCLEAR GENERATING STATION
DOCKET NO. 50-267
EMERGENCY ELECTRICAL POWER SYSTEM
INTRODUCTION
By task interface agreement (TIA) No. 85-02, Rev. 1, dated February 5, 1985 the
NRR staff was requested to review the subject design with reference to the
problems encountered during the testing of the emergency diesel generator sets
(EDGS) on December 18, 1984. The staff has reviewed FSV's latest revision of
the FSAR and electrical schematic drawings of the emergency power systems. The
review focused primarily on the compliance of the design with the redundancy,
independence and single failure criterion established in the FSAR.
BACKGROUND
Emergency electrical distribution system at FSV is a 3 bus (two redundant and
one swing bus) system with two 100% load capacity EDGs. Each EDG has two
tandem engines each rated to 1/2 of the generator output capacity. If only one
of the two engines operates in one EDG system, the other redundant EDG must
also be operative to supply the shutdown load with at least one of its engines
operating. The intended logic at FSV is to start both EDGs simultaneously and
let the first EDG with rated voltage, frequency and 100% output (both engines
operating) be connected to its assigned 480 volt bus together with the swing
bus. The first generator on line assumes sequence "A" loading which is
sufficient for an orderly shutdown and continued maintenance of the plant in a
safe shutdown condition. The second generator, if available with rated voltage
and frequency, will assume sequence "B" loading.
On December 18, 1984, with the reactor shutdown and the PCRV depressurized, the
loss of offsite power and turbine trip semiannual surveillance test was
initiated by blocking one EDG (EDG-A) to intentionally make the other EDG
(EDG-B) first on line and assume sequence A load. Due to the nonavailability
of one of the two engines with EDG-B, this logic could not be completed and
breaker did not close. The logic should have made the intentionally blocked
EDG-A as the second generator in line and should have closed its supply breaker
to initiate sequence B loading on EDG-A. The EDG-A breaker also did not close,
thus causing loss of both redundant emergency power supplies to the essential
buses. EDG-A failure to supply power was attributed to the inadvertent trip of
exhaust temperature switches on both engines of EDG-A due to the loss of
instrument power to these switches. This event necessitated a review of the
FSV emergency electric system to establish the following:
(1) Independence and redundancy of the onsite AC power supply distribution
system and the safety loads to perform their safety function.
JUL 0 5 198:
- 2 -
(2) Reasons for EDG-B's inability to get connected to the bus when only one of
its two engines failed and the other was available to supply 1/2 of the
designed capacity of EDG-B.
EVALUATION
The FSAR maintains in Section 8.2.5.1 that the AC and DC power systems in FSV
design are each redundant systems; the onsite power supplies are completely
independent and meet the single failure criterion. Our review of FSV's onsite
electric system drawings (EDG breaker and bus tie breaker schematic diagrams
and auxiliary tripping relays control diagrams), on sample basis revealed the
following information.
1. Automatic closure of one redundant EDG breaker is dependent on the
operation of components associated with the other redundant EDG.
Interlocks from one division providing permissive in the breaker close
circuitry of the other could potentially prevent the required operation of
both circuits and render both emergency power supplies incapable of
performing their safety function.
2. Each redundant EDG should be capable of supplying 100% of its rated power
when both engines operate and 50% of its rated power when only one is
operative. The EDG breaker should close for rated voltage and frequency
irrespective of whether one engine is operating or both. FSV design (EDG
breaker schematic) indicates that the breaker will not close if one of the
two engines is inoperative.
Both identified discrepancies, were explained to the licensee in detail in a
meeting held on May 16, 1985. Based on the available information, it is the staff's
understanding that the automatic operation of the redundant EDG circuit
breakers is dependent on each other which is contrary to the FSAR requirement.
This discrepancy could potentially render both emergency power supplies
incapable of automatically performing their safety function. However, in case
the EDG breaker fails to close automatically, manually operated electrical
control breaker closing circuitry is available in FSV design to initiate
closure of the breaker immediately after identifying the failure of the
automatic circuitry. The licensee confirmed that the manual circuitry is not
affected by the failure of the automatic breaker closure circuitry.
Our review of the FSAR indicates that besides the EDGs, the FSV design includes
an alternate means of providing electric power for cooling the reactor, in case
both offsite power and EDGs are not available. This power source is capable of
operation, independent of disruptive faults or events, such as a major fire in
the congested cable areas. This system is named "Alternate Cooling Method
(ACM)" and manually started to restore 480V electric power at the control
terminals of the required safety equipment within two hours (1 to 2 hours).
The ACM power source is a non Class 1E, 4160 volt, 60 Hz AC diesel generator,
rated at 2500 kW (equal to the combined rating of both EDGs) and is located
JUL 05 198E
- 3 -
away from the existing plant structure. The ACM is designed to accomplish the
following functions by means of local manual initiation:
(a) To maintain the reactor subcritical using Reserve Shutdown System.
(b) To resume PCRV liner cooling, thereby cooling the core and the PCRV.
(c) To allow depressurization of the PCRV through the helium purification
system.
(d) To establish operation of the Reactor Building Exhaust System and
radiation monitoring of the exhaust effluent to the atmosphere.
Additionally, the ACM can power the plant lighting, fire pumps, service water
pumps and plant ventilation system.
The staff reviewed Section 8.2.8.5 of the FSAR and noted that in the unlikely
event when both EDGs are not available coincident with the loss of offsite power,
and the ACM power is restored to the emergency equipment by manual means
within two hours, then adequate core cooling and depressurization of primary
coolant system can be achieved maintaining integrity of the core and the PCRV.
CONCLUSION
The staff's evaluation of the FSV EDG electric system has identified some
discrepancies in EDG breaker control logic regarding independence of the
redundant EDG system. However, the inherent capability of the PCRV and
core with an alternate non-Class lE power source (ACM) provides an added
assurance of safe shutdown capability.
In the interim, until the licensee proposes any necessary modification in the
EDG breaker automatic control circuitry, manually operated switches are
available to override the automatic control circuit failure and close the
breakers to provide power to operate equipment necessary for safe shutdown of
the plant. This can readily be accomplished well within the time frame
available to prevent damage to the reactor.
It is the staff's conclusion, as it relates to the emergency power system prob-
lems identified, that the plant operation may resume without undue risk to the
health and safety of the public. However, the licensee should establish a sche-
dule, without undue delay, for the review and resolution of the potential single
failure and independence problems for EDGs identified in this report.
Date: JUL 05 1985
Principal Contributor:
I. Ahmed, DSI
y
SSINS No. : 6835
IN 85-50
UNITED STATES WEIR COUNTY COMMISSIFERS
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMEN 0
WASHINGTON, D.C. 20555 JUL'1 8
July 8, 1985
GREELEY. COLO.
IE INFORMATION NOTICE NO. 85-50: COMPLETE LOSS OF MAIN AND AUXILIARY FEEDWATER
AT A PWR DESIGNED BY BABCOCK & WILCOX
ADDRESSEES:
All nuclear power facilities holding an operating license (OL) or construction
permit (CP).
Purpose:
This information notice is being provided to inform licensees of a significant
reactor operating event involving the loss of main and auxiliary feedwater at
a pressurized water reactor. Information in this notice is preliminary and was
obtained from the special NRC fact finding team which is investigating the
event. A complete report of findings will form the basis for further communi-
cations or actions related to this event. The NRC expects that recipients
will review this notice for applicability to their facilities. Suggestions
contained in this notice do not constitute NRC requirements; therefore, no
specific action or written response is required.
Description of Circumstances:
On June 9, 1985, the Davis-Besse plant was operating at 90% power with Main
Feedwater Pump 2 in manual control because problems in automatic had been
experienced. A control problem with Main Feedwater Pump 1 occurred, and it
tripped on overspeed. Reactor runback at 50% per minute toward 55% power was
automatically initiated. Nevertheless, 30 seconds later, the reactor tripped
at 80% power on high pressure in the reactor coolant system.
One second after reactor/turbine trip, one channel of the Steam and Feedwater
Rupture Control System (SFRCS) was automatically initiated due to a spurious
signal indicating low water level in Steam Generator 2. Both Main Steam
Isolation Valves (MSIVs) closed. Three seconds after the actuation, the SFRCS
automatically reset. Closing of the MSIVs isolated the turbine of the operating
main feedwater pump from its source of steam. The pump continued to supply
feedwater to the steam generators for a few minutes as it coasted down.
Four and a half minutes after reactor trip, water level in the steam generators
began to fall from the normal post-trip level which is 35 inches. After MSIV
closure, steam release to atmosphere continued to remove decay heat. One minute
later, Channel 1 of SFRCS actuated when the water level in Steam Generator 1
actually reached the SFRCS setpoint at 27 inches (See Figure 1). SFRCS started
Auxiliary Feedwater Pump 1 and initiated alignment of it to Steam Generator 1.
8507080156
IN 85-50
July 8, 1985
Page 2 of 4
Within seconds after automatic initiation of Channel 1 of SFRCS, the operator
actuated both channels of SFRCS; however, he inadvertently actuated both SFRCS
channels on low steam pressure instead of low water level . When an SFRCS
channel is actuated on low steam pressure, a rupture of the steam line associated
with that channel is presumed to have occurred. The SFRCS closes the steam
generator isolation valves, including a valve in the auxiliary feedwater line,
and aligns the auxiliary feedwater pump to the other steam generator. Because
both channels had been manually actuated on low steam pressure, both steam
generators were isolated from both auxiliary feedwater pumps. Five seconds
after the operator' s inadvertent actuation of both channels on low steam
pressure, SFRCS Channel 2 received an actual low water level actuation signal .
Because low pressure initiation takes precedence, alignment of the auxiliary
feedwater pumps remained unchanged. At six minutes into the event as both
auxiliary feedwater pumps were accelerating, they tripped on overspeed.
In summary, all main feedwater had been lost, both steam generators were isolated
from feedwater and were boiling dry, all auxiliary feedwater pumps were tripped,
pressure of the reactor coolant system was rising, and reactor coolant system
temperature was increasing.
Within one minute after the operator' s inadvertent actuation of the SFRCS on
low steam pressure, the mistake had been recognized and the SFRCS had been
reset. If equipment had performed in accordance with system design requirements,
the operator' s error might not have had a significant impact on the event.
The auxiliary feedwater isolation valves should have reopened automatically,
but the valves did not reopen. The operator then tried to reopen the valves
from the main control panel , but the valves would not reopen. Operators were
dispatched to locally start the auxiliary feedwater pumps, open the auxiliary
feedwater isolation valves, start the nonsafety-related motor-driven startup
feedwater pump, and valve it to the system.
Pressure and temperature in the reactor coolant system continued to rise
because there was not sufficient water in the steam generators to provide an
adequate heat sink. At 13 minutes after reactor trip, reactor coolant system
pressure reached 2425 psig, and the Pilot Operated Relief Valve (PORV) opened
three times to limit the pressure rise. On the third lift, the valve remained
open. The operator closed the PORV block valve and reopened it two minutes
later after the PORV had closed.
Approximately 16 to 18 minutes after reactor trip, the operators had the startup
and auxiliary feedwater pumps running and the valves aligned. Water levels were
beginning to rise in the steam generators. Reactor coolant temperature reached
a maximum of 594° F and then started to decrease to normal . Refilling of the
steam generators caused the reactor coolant system to fall to 1716 psig and
about 540°F before returning to normal (See Figure 2).
At 30 minutes after reactor trip, plant conditions were essentially stable.
IN 85-50
July 8, 1985
Page 3 of 4
Discussion:
For several minutes after reactor trip, the steam generators were unable to
cool the reactor coolant system adequately.
The first problem contributing to this event was the loss of all main feedwater
due to closure of the MSIVs. The licensee' s hypothesis, based on information
from Babcock & Wilcox, is that turbine trip caused a pressure transient upstream
from the turbine stop valves which caused the outputs of the redundant steam
generator level instrumentation channels to oscillate widely for several
seconds. The licensee believes that this caused a spurious low level actuation
of SFRCS which closed the MSIVs.
Three additional problems contributed to this event by affecting the availability
of both trains of the auxiliary feedwater system. The first occurred when the
reactor operator pressed the wrong SFRCS buttons. The second occurred when
both auxiliary feedwater pumps tripped on overspeed. The third occurred when
both auxiliary feedwater isolation valves did not reopen when SFRCS was reset.
Control buttons for the SFRCS are arranged in two vertical columns. Each
column of buttons controls one SFRCS channel . The operator should have pressed
the fourth button from the top in each column. Instead, the operator pressed
the top buttons causing isolation of both steam generators.
Both auxiliary feedwater pumps are driven by Terry turbines which tripped on
overspeed early in the event. When this occurred, steam was being supplied to
the turbines via crossover lines, which are longer than the normal supply lines
and include long horizontal runs. The licensee believes that significant
condensation may have occurred in the crossover lines. Further, the licensee
believes that the quality of steam arriving at the turbines may have been
affected significantly by the configuration of the crossover lines and may have
caused the overspeed trips.
The auxiliary feedwater system isolation valves have Limitorque motor operators.
The motor operators have torque switches which prevent overtorquing of the
valves by disconnecting power to the motors. When the valves are being opened,
additional torque is required to overcome friction while the gates are being
unseated and while a significant pressure differential may exist across the
gates. During the initial part of the opening stroke, the torque switch in the
motor operator is bypassed by a bypass switch so that full motor torque is
developed if necessary. The licensee believes that these bypass switches went
off bypass too early. The valves did not reopen until an operator unseated
them by hand.
IN 85-50
July 8, 1985
Page 4 of 4
No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact the Regional
Administrator of the appropriate NRC regional office or this office.
alEdward Jordan, Director
Divisio of Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
Technical Contact: R. W. Woodruff, IE
(301) 492-4507
Attachments:
1. Figure 1 - Steam Generator 1 Level and Pressure
2. Figure 2 - RCS Temperature and Pressure
3. List of Recently Issued IE Information Notices
Attachment 1
IN 85-50 '
LB83 SG 1 SU RANGE LVL, 983 (IN) July 8, 1985
o i.S' 50 75' /00 /JS 4-0 /75 200 say .2-SO
t I 1
P932 SG 1 OUT SIM PRESS.PT12B2 PSIR
600 ESo Tan 750 goo $Sb 900 9.522 ioao /ctso //co
— -
0 T
cn
•tsi el
_. ...........o a ...........e.40 9
5 �1
! . . ! .
• IHT
4 .
FIGURE I : STEAM GENERATOR I LEVEL
auD PRESSURE
Attachment 2
IN 85-50
July 8, 1985
P725 RC LOOP I HLG WR PRES5.SFRS CH 3 • PSIR
/We /Soo 1600 M» /S00 /900 .lee NOo faoo K. •4t'O
f
T709 ' RC RVG NR TEMP ', F
5.10 453 d 5+0 .SS° $6o 37o a a sy0 boo 610 620
C'-44
0
--1
- - - —y-
0%- i
ham-. -0 _ ........... .............
3
AN
. - _ : ---I
3 1+
as
b
FIGURE z: RCS TEMPERATURE:
I u r, oor_ CC '0
Attachment 3
IN 85-50
July 8, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
85-49 Relay Calibration Problem 7/1/85 All power reactor
facilities holding
an OL or CP
85-48 Respirator Users Notice: 6/19/85 All power reactor
Defective Self-Contained facilities holding
Breathing Apparatus Air an OL or CP, research,
Cylinders and test reactor,
fuel cycle and
Priority 1 material
licensees
85-47 Potential Effect Of Line- 6/18/85 All power reactor
Induced Vibration On Certain facilities holding
Target Rock Solenoid-Operated an OL or CP
Valves
85-46 Clarification Of Several 6/10/85 All power reactor
Aspects Of Removable Radio- facilities holding
active Surface Contamination an OL
Limits For Transport Packages
85-45 Potential Seismic Interaction 6/6/85 All power reactor
Involving The Movable In-Core facilities holding
Flux Mapping System Used In an OL or CP
Westinghouse Designed Plants
85-44 Emergency Communication 5/30/85 All power reactor
System Monthly Test facilities holding
an OL
85-43 Radiography Events At Power 5/30/85 All power reactor
Reactors facilities holding
an OL or CP
85-42 Loose Phosphor In Panasonic 5/29/85 All power reactor
800 Series Badge Thermo- facilities holding
luminescent Dosimeter (TLD) an OL or CP
Elements
OL = Operating License
CP = Construction Permit
MEMORANDUM
��DF cow
„„, DIVISION OF DISASTER John P.Byrne
FQ _ EMERGENCY SERVICES — - DIRECTOR
* �' Gy * Camp George West HELD CONNI�CO COMMISSIONERS
` •- • Golden,Colorado 80401
*7876 (303)273-1624 1
JUL 151985
DATE: 10 July 1985
GREELEY. COLO.
TO: (See Distribution)
FROM: John P. Byrne, Di ct r, Division of Disaster Emergency Services
SUBJECT: Summary Report f 98 rt St. Vrain Exercises
1. Attached is the summary report for the 1985 Fort St. Vrain full scale
exercise for your review and action as necessary.
2. The next full scale exercise will be conducted in 1987. In the interim,
we will conduct periodic functional drills and exercises to sustain
emergency preparedness.
3. While F0SAVEX-85 is still fresh in our minds, please forward any changes or
revisions to the Radiological Emergency Response Plan to Bruce Smith
as soon as possible.
1 attachment
F0SAVEX-85 Exercise Report
•
/js
cs
COLORADO •
DEPARTMENT OF
PUBLIC SAFETY
Distribution
Office of the Governor
Department of Administration
Department of Public Safety
Department of Health
Department of Agriculture
Department of Highways
Department of Social Services
Department of Law
Department of Military Affairs
Department of Local Affairs
Department of Natural Resources
Division of Disaster Emergency Services
Division of Colorado State Patrol
Weld County Commissioners
Weld County Sheriff
Weld County Emergency Management Director
Weld County Health Department
Weld County Agricultural Extension Service
National Weather Service
Federal Emergency Management Agency
Environmental Protection Agency
Food and Drug Administration
Department of Energy
American Red Cross
Salvation Army
Public Service Company of Colorado
OVCD%�
MEMORANDUM
- --v t -DIVISION OFDISASTER __ _ John P.Byrne
�$ EMERGENCY SERVICES DIRECTOR
N J�
* �� 44., *) Camp George West
+ • • *- Golden,Colorado 80401
1876# (303)273-1624
DATE: 1 July 1985
TO: John P. Byrne, Director, DODES •
FROM: Bruce N. Smith, Exercise Directoz] 21.
Summary Report for 1985 Fort St. Vrain Exercise
SUBJECT:
1. The 1985 full scale exercise to test major off—site functional elements
of the Colorado State Radiological Emergency Response Plan (RERP) for
the Fort St. Vrain Nuclear Generating Station was conducted on 18 June
1985. The exercise short title was FOSAVEX-85 and was a joint
participation exercise with the Public Service Company (PSC) of
Colorado. The objectives of the off-site exercise were to exercise and
evaluate the following:
a. Notification procedures and public warning dissemination
systems, to include activation of NOAA Weather Radio Network and
passage of an exercise warning message to the Emergency Broadcast
System (EBS) lead station.
b. Deployment of response elements and their direction and
control, to include activation of the State elements of the Forward
Command Post (FCP) and establishment of a controlled area and
traffic control points.
c. Radiological monitoring, assessment and reporting procedures
from field deployed Health Department monitoring teams through the
FCP to the State Emergency Operations Center (EOC) .
d. Operational communications and associated direction and control
procedures at all levels.
e. Food and agricultural control procedures.
f. Dissemination of official, coordinated, pertinent and accurate
information to the news media and general public.
c s
COLORADO
DEPARTMENT OF
PUBLIC SAFETY
2
This report is based upon the comments and observations of the official
state exercise observation_and evaluation team, review of exercise logs
and message traffic files, and discussions held during the post-
exercise critique.
2. Overall, the exercise response was excellent and was indicative of
sustained, on-going emergency preparedness training programs at all
levels. In particular, the level of seriousness and professionalism
displayed by exercise participants at all locations was exemplary--as
one observer commented, "If you didn't have prior knowledge of an
exercise, you would think an actual emergency existed". No major
deficiencies were noted and the capacity to protect the health and
safety of the public in the vicinity of the plant was clearly
demonstrated. Contrary to what is often encountered, communications at
all levels during FOSAVEX-85 were exceptional.
3. As in every successful exercise, areas were identified where
improvements are possible. These "lessons learned" are the most
valuable results of the exercise because, with proper follow up and
training, they lead to improved emergency response capabilities. The
following observations and comments are provided to assist in such
improvements:
a. State Emergency Operations Center (EOC)
(1) The initial notification sequence was timely and efficient,
due primarily to the rapid and correct reaction of the Weld
County Communications Center.
(2) Access processing at the EOC was degraded because of
outdated organizational access rosters which required excessive
delays and frequent escorts.
(3) The exercise scenario was forced to be excessively
artificial because of existing regulations. Scenario time
compression forced a too rapid transition to a General Emergency
category. PSC and DODES will once again request a determination
from the Nuclear Regulatory Commission (NRC) and the Federal
Emergency Management Agency (FEMA) as to whether a General
Emergency category could ever be reached at Fort St. Vrain,
given the radioactive inventory and reactor design.
(4) EOC activation was smooth and efficient but some agencies
required by the REAP to report to the EOC were notified but did
not report.
(5) Television and radio broadcasts were not monitored to keep
track of public information being disseminated to assure
accuracy
(6) There is a need for more effective and detailed information
exchange between PSC and State agencies when emergency classes
change.
(7) If NRC is going to interact with State agencies concerning
3
off-site activities, this needs to be formally recognized and
incorporated into the RERP.
b. Forward Command Post (FCP)
(1) Rapid and efficient access was also a problem at the FCP
because of outdated access rosters and the frequent need for
individual, personal validation by the FCP Director. A positive
aspect was the superb job of access control done by Weld County
Deputy Sheriff Park -- if you weren't cleared, you didn't enter,
period.
(2) The FCP message center function needs improvement in the
area of physical movement of message traffic between the FCP and
the communications van.
(3) A larger scale map of the plant vicinity, with clearly
marked road identifications, would improve the efficiency of
plotting and reading.
(4) Non-essential vehicle traffic was allowed to pass through
the alley outside the FCP and parking was not fully controlled.
Both factors created unnecssary congestion and impeded efficient
personnel movement.
(5) The establishment and control of traffic control points by
the Weld County Sheriff's Department was especially effective.
(6) The physical layout of the FCP needs to be changed to take
full advantage of video monitor equipment and to eliminate
duplicate posting of the same data. Key staff elements,
particularly the Department of Health and DODES, need to be more
effectively located to enhance interaction and reduce
distractions from peripheral activity.
(7) It was significant and appropriate that the Department of
Health senior representative continued to request field monitor
readings at specific locations in the controlled area even after
being informed that the on-site leak was contained and emergency
class downgraded.
c. Field Monitor Activity
(1) After initial problems with frequency selection, the field
monitor direction, control and reporting radio network
functioned extremely well. To my knowledge, this is the first
time that this capability has been effectively demonstrated and
represents a significant step forward.
(2) Field monitor team vehicles need to be identified in some
manner so as to be readily recognized as emergency vehicles.
(3) Even though marked improvement was evident, more practice
in proper radio communication procedures would be beneficial.
(4) Some method (vests, hats, etc.) should be used to identify
field monitor team members to expedite assembly, briefing and
dispatch. Improvement is also needed in issuing controlled area
access badges to team members. Signs should be used to identify
team assembly area to minimize confusion.- -
(5) Better maps are needed for field monitor teams.
(6) More actual demonstration of the use of monitoring
equipment and sample collection procedures is needed as well as
collection of air sampling equipment filters and aggregation and
transport of samples to the laboratory for analysis.
(7) Most monitoring equipment issued was overdue for
calibration and personal dosimeters were not readily available
for individual issue to monitor team members.
(8) Greater attention is needed relative to emergency
worker/monitor exposure control procedures and status briefings
to same on the radiological emergency.
/js
J``¢M neau4J UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION IV
•
c e 611 RYAN PLAZA DRIVE,SUITE 1000
Y�'44 � .coq ARLINGTON,TEXAS 78011
JUL 12 1985
Docket: 50-267
!VD CBDDT3 t rz SSrR RS
Public Service Company of Colorado r ;L
ATTN: O. R. Lee, Vice President a1- 16935
� I
Electric Production
P. O. Box 840
Denver, Colorado 80201 clRca,car.czranm,
Dear Mr. Lee:
We have reviewed the various responses you have submitted related to our
October 16, 1984, "Preliminary Report Related to the Restart and Continued
Operation of the Fort St. Vrain Nuclear Generating Station" and the subsequent
meetings. Our review of the germane submittals is presented in the enclosed
Safety Evaluation (SE).
Our review indicates that numerous issues have been resolved. There are,
however, additional issues which must be resolved prior to plant restart and
some longer term issues for which commitments for resolution must be acceptable
prior to restart. Some of these issues are addressed by the enclosed SE;
others will be the subject of separate correspondence.
If you have any questions on this matter, please contact the NRC Project
Manager - P. Wagner at (817) 860-8127.
Sincerely,
E. H. Johnson, Chief
Reactor Project Branch
Enclosure:
Safety Evaluation
cc:
Mr. D. W. Warembourg, Manager
Nuclear Engineering Division
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
(cont. on next page)
�,� me Ho (8S
Public Service Company of Colorado -2-
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 194
Platteville, Colorado 80651
L. Singleton, Manager, Quality
Assurance Division
(same address)
Colorado Radiation Control Program Director
ENCLOSURE
tRR REG& UNITED STATES
W° °� NUCLEAR REGULATORY COMMISSION
L REGION IV
w
611 RYAN PLAZA DRIVE,SUITE 1000
�'►� po ARLINGTON.TEXAS 70011
SAFETY EVALUATION BY THE NUCLEAR REGULATORY COMMISSION
FORT ST. VRAIN NUCLEAR GENERATING STATION
PUBLIC SERVICE COMPANY OF COLORADO
DOCKET NO. 50-267
Introduction
As a result of the failure of six Control Rod Drive Mechanisms (CRDMs) to
automatically insert their associated neutron absorber material (2 control rod
strings per CRDM) following a reactor scram on June 23, 1984, the NRC issued a
confirmatory action letter (CAL). The CAL, dated June 26, 1984, required
Public Service Company of Colorado (PSC or the licensee) to take various
actions and to maintain the Fort St. Vrain (FSV) reactor in the shutdown
condition until the NRC authorized restart. This failure to insert problem, in
conjunction with other areas of concern, prompted the Director of the Office
Nuclear Reactor Regulation, NRC, to direct an overall conduct of operations
evaluation of FSV. A team of NRC personnel and NRC consultants from the Los
Alamos National Laboratory conducted audits of the FSV facility on July 9
through 11, 1984 and August 1 through 3, 1984. The results of these audits are
contained in a "Preliminary Report Related to the Restart and Continued
Operation of the Fort St. Vrain Nuclear Generating Station" (Assessment Report)
which was transmitted to PSC by letter dated October 16, 1984. This safety
evaluation presents the NRC findings related to the various PSC submittals and
meetings in response to the Assessment Report.
Other areas of concern which require resolution prior to plant restart (e.g. ,
the concrete reactor vessel prestressing tendon system problems and the
emergency electrical systems operation) are discussed in separate
correspondence.
Evaluation
The Assessment Report required the completion of a number of items prior to the
restart of FSV and additional long term items following restart. PSC responded
to the Assessment Report requirements by letter dated January 4, 1985,
(P-85003) with commitments to resolve the NRC concerns. Numerous meetings and
submittals expanded and clarified those commitments. Our evaluation of the PSC
commitments and corrective actions is contained below and follows the numbering
sequence used in the Assessment Report.
A. Items Required to be Completed Prior to Restart
1. Actions Required for Control Rod Problems
a. "Ensure that future scram signals will result in all rods
automatically being inserted into the core. The licensee must
identify the failure mechanism and take corrective action for
-2-
the rods that did not scram; or if the cause cannot be
positively identified through examination or analysis of the
drive mechanisms and the circumstances of the failure, other
compensatory measures must be taken to provide assurance of
reliability of control rods. These measures could reasonably
include refurbishing all drive mechanisms. Regardless, of any
other measures taken to remedy the failure to scram problem
prior to reactor restart, PSC must outline and commit to
periodic inspection/preventive maintenance and surveillance
programs for control rod drives and associated position
instrumentation.
b. "Implement procedures to prevent overdriving the control rods
past the rod-in limit.
c. "One 20-weight percent and one 40-weight percent reserve
shutdown hopper should be functionally tested to assure that the
reserve shutdown capability is fully available.
d. "Until the long term corrective actions are completed, the
licensee should develop a procedure that will require a reactor
shutdown under conditions where purge flow is lost or when high
levels of moisture exists in the coolant.
e. "Implement a procedure for recording representative samples of
CRDM temperatures at all operating conditions until continuous
recordings capability is available."
By letter dated January 31, 1985, PSC provided the plans to resolve
the Control Rod problems. Our review of this, and supplemental
information, is described in Attachment 1, Safety Evaluation of
Control Rod Drive Mechanisms and Reserve Shutdown Systems.
Attachment 2 is a copy of the Technical Evaluation performed by our
consultant at the Los Alamos National Laboratory.
There are, however, a number of items which require additional
resolution or which have occurred since the issuance of the
Assessment Report.
As discussed in the Safety Evaluation (SE) , Attachment 1.
The licensee must provide a commitment to operate the plant
within the CRDM temperature limits accepted by the NRC. The
temperature limits cannot be changed without NRC approval of new
temperature limits or alternative methods of assuring CRDM
operability; and
-3-
The licensee must provide a commitment to submit an improved
- CRDM surveillance and preventative maintenance program within
six months of plant restart.
In addition, the following issues must be resolved:
The acceptability of the replacement ball bearings used in the
- CRDM refurbishment; and
The acceptability of the epoxy used to attach the CRDM
- temperatures sensors.
By letter dated June 14, 1985 (P-85199) PSC stated that the interim
Technical specifications (TS) contain a requirement that the CRDM motor
temperatures are monitored to ensure the temperatures are within acceptable
limits. (See PSC letter P-85180, dated June 7, 1985, and the discussion
in Item 3 below. ) In addition, P-85199 committed to submit an improved
CRDM surveillance and preventive maintenance program within six months of
plant restart. By letters dated June 7, 1985 (P-85195) and June 13, 1985
(P-85201) PSC provided information on the epoxy and the bearings,
respectively. Our review of this information is in progress and will be
the subject of a subsequent SE.
Additional information concerning the CRDM position instrumentation and
procedures to prevent overdriving is contained in Item 3 below, "Actions
Required for the upgrade of TS" , and in Attachment 3.
2. "Actions Required to Correct Weaknesses Noted in the Area of Overall
Conduct of Operations
"In the area of overall conduct of operations, the staff confirmed
the deficiencies noted in various Region IV inspection reports and in
the last two SALP reports. The staff has concluded that PSC must
develop a comprehensive program for identifying the underlying causes
for the deficiencies and for applying corrective measures. This
program should be conducted by a third party consulting organization
and should be aimed at reviewing the PSC management structure and
practices relative to the operation of FSV with emphasis on
correcting deficiencies noted in the various Region IV inspection
reports, the last two SALPs and programmatic weaknesses identified in
Section 4 of this report. PSC should submit the scope and schedule
for this program prior to reactor restart."
The scope of this management review program was further discussed
with the licensee in a meeting on November 14, 1985. (The summary
of this meeting is contained in IE Inspection Report 50-267/84-32
dated May 22, 1985. Based on the information contained in the
-4-
assessment report and the further understanding gained from this
meeting, the licensee commissioned the NUS Operating Services
Corporation to perform this review. By letter dated February 28,
1985 (P-85066), PSC provided a copy of the NUS Report, "An Analysis
and Evaluation of the Management of Nuclear-Related Activities of
the Public Service Company of Colorado" together with their response
to that report. PSC provided additional information in their March
29, 1985 (P-85107) submittal.
The staff reviewed these submittals and requested additional details
regarding PSC's proposed actions for Sections 4.2.5 "Conduct of
Operations" and 4.2.6 " Maintenance Practices" of the Assessment
Report. This additional information is contained in the licensee's
letter dated May 22, 1985 (P-85178). Based on our review of these
submittals, we find that PSC has carried out a management review
that meets the requirements of the Assessment Report.
To provide the means of implementing the corrective actions needed
to address the recommendations made in this management review, PSC
has developed a Nuclear Performance Enhancement Program. This
program is described in the March 29, 1985, letter referenced
above. Additional details were provided to the NRC on May 31, 1985,
during the SALP management meeting held with PSC at the Fort St.
Vrain site. The status of implementation of this program was
described to the staff in a meeting on June 17, 1985.
The Nuclear Performance Enhancement Program is a proactive management
scheduling and followup device that includes all of the significant
findings of the NRC Assessment Report, the NUS management audit, NRC
inspection findings and company generated findings. Each of the
items is assigned a project manager and a schedule for completion is
established in accordance with the licensee's overall priority
scheme. Progress against this schedule is determined bi-weekly and
management is kept advised in order to ensure that any schedule
changes are agreed to by management.
The staff has determined that the Nuclear Performance Enhancement Program
has the structure and capability of carrying out the corrective actions
that are necessary. If properly supported by PSC management, many of the
deficiencies noted by the NRC in licensee performance should be adequately
addressed.
3. Actions Required for the Upgrade of Technical Specifications
a. "A high priority effort should be undertaken to review and propose
revisions to the existing Technical Specifications to reduce the
likelihood of operator error and/or misinterpretation and correct
omissions. The staff has determined that a schedule should be
-5-
developed by PSC which will reflect completion of the review,
revision, and submittal of the proposed Technical Specifications by
April 1, 1985.
b. "To improve control rod and reserve shutdown reliability, the licensee
shall propose the following changes to Technical Specifications, and
implement interim procedures until the Specifications are approved:
(1) A Weekly control rod exercise surveillance program for all
partially or fully withdrawn control rods;
(2) A Limiting Condition for Operations defining control rod
operability and the minimum requirements for rod position
indication; and
(3) A Limiting Condition for Operations and a corresponding
surveillance test to define and confirm reserve shutdown
system operability."
By letter dated April 1, 1985 (P-85098) PSC provided a submittal titled
"Upgraded Technical Specifications for Acceptance Review" in fulfillment
of item a. above. Our evaluation of this submittal will be the subject
of a future SE.
By letter dated March 15, 1985 (P-85089) PSC submitted "Draft Technical
Specifications to Improve Control Rod Reliability." Our review of this
submittal determined that numerous problems existed and a meeting was held
on May 3, 1985 to discuss those problems. The results of this meeting are
documented in our letter dated May 28, 1985.
By letter dated June 7, 1985 (P-85180) PSC resubmitted the TS, together
with a commitment to implement those requirements through the use of
interim procedures prior to restart. Our evaluation of this resubmittal
indicated that some improvement and clarification was necessary to ensure
resolution of our concerns. PSC has agreed to submit information to
satisfy our concerns; we will evaluate this information prior to restart.
As discussed in Attachment 3 - "Safety Evaluation Related to Control Rod
Position Instrumentation", additional operability and surveillance
requirements are needed to ensure safe operation. Specifically, the
following issues should be incorporated:
Additional TS and procedures for determining control rod
- full-in position. (See Section 3.4 of Attachment 3. )
Additional surveillance tests on rod position indications.
— (See Section 3.4 of Attachment 3. )
-6-
Initiation of the backup shutdown system if rod full-in
position indication cannot be verified within 1 hour.
(See Section 3.4 of Attachment 3.)
_ Additional procedural control to prevent inward overtravel
of control rods. (See Section 3.2 of Attachment 3.)
The proposed TS (P-85180) discussed above provide addition controls and
limitations on the operation and testing of the control rods and their
indication. In addition, plant operating procedures were revised to
prevent inward overtravel of the control rods. (SOP 12-01, Issue 15 which
was submitted in letter P-85040 dated January 31, 1985). Some problems
have been identified which PSC has agreed to resolve in the submittal
discussed above.
4. Actions on Continued Water Ingress
"In the area of continued water ingress the staff has determined that
PSC must develop a plan to carry out any of those modifications recommended
by the PSC "Moisture Ingress Committee" that are determined by PSC to have
a high potential to significantly reduce the frequency and severity of
upsets involving injection of circulator bearing water into the helium
coolant. Any significant reduction would clearly reduce the frequency of
plant transients; improve the reliability of overall plant operations;
and might, if has an effect, improve the performance of control rod
drives. This plan should include a status report to the NRC as part
of the annual report on the progress in implementing modifications."
By letter dated January 24, 1985 (P-85022) PSC outlined some of the
modifications which have been completed, are presently under way, and
are planned to be implemented to control the water ingress problem.
Since this submittal did not contain sufficient detail to resolve our
concerns, PSC agreed, during a meeting at FSV on February 21, 1985,
to provide additional information.
Additional detail was provided in PSC letter, P-85082, dated March
12, 1985. Our review of this submittal and other germane information will
be included in a future SE.
B. Actions Required Following Restart
The Assessment Report, in addition to requiring that certain items be
completed as a condition for plant restart, required some longer term
actions. Specifically, the Assessment Report stated:
"In addition to the above items required for restart, the staff noted
several weaknesses that should be corrected on a longer term basis. The
licensee must submit schedules within 60 days of restart for completing
these items."
The January 4, 1985 (P-85003) PSC letter provided the required schedules,
NRC review of the schedules resulted in a meeting on January 15, 1985, to
-7-
discuss differences. The results of the meeting are contained in our
January 17, 1985, letter to PSC.
The items listed in the Assessment Report for which action is required
following plant restart are:
a. "Provide continuous recording of a representative sample of CRDM
temperatures at all operating conditions to provide part of the data
necessary for the longer term program noted below (Section 2).
b. "Determine whether compensating design and/or operational
modifications are needed to minimize moisture ingress to the CRDM
cavities and minimize temperatures in the vicinity of the rod drives.
In the event that temperatures recorded during plant operation prove
to be higher than those for which the assembly was initially
qualified, take immediate steps to perform environmental
requalification testing of a CRDM assembly or hold temperatures to
that for which the CRDM has been qualified (Section 2).
c. "The present Watt-meter testing of the shim motor during drive-in and
drive-out is not a reliable method to verify full insertion or
withdrawal of control rods. This test should be refined or an
alternative, reliable test for control rod position verification,
must be developed (Section 3).
d. "Investigate a design change to provide a positive stop on the CRDM
position indicator potentiometer shaft to prevent overtravel (Section
3) and provide the results to the NRC.
e. "Conduct an integrated systems study to resolve rod position
indication, maintenance and operability questions (Section 3).
f. "Establish procedures for verification and sign-off by the
Maintenance Quality Control (MQC) of key steps in Technical
Specifications surveillance procedures (Section 5).
g. "Establish a procedure for review and concurrence by the QA organi-
zation of safety-related procedures and changes thereto (Section 5).
h. "At the time of the audit, the MQC group was reviewing each completed
surveillance procedure. The staff concluded that this practice
should continue.
i. "A review by the QA organization of the content and adequacy of the
Technical Specification procedures is important, and the staff has
determined that this should be implemented."
These items were all discussed in the January 4, 1985 (P-85003) submittal
and some were clarified in our January 17, 1985, letter. In addition,
-8-
some of the items are further discussed as follows:
a. Continuous recording of CRDM temperatures as, discussed in PSC letters
P-85032 dated January 30, 1985, and P-85199 dated June 14, 1985, has
been implemented through the use of multipoint recording devices
which will provide frequent monitoring.
b. A discussion of the modifications planned to control moisture and
purge flow to the CRDMs and plans to qualify a CRDM to higher
temperatures is contained in PSC letter P-85032 dated January 30,
1985.
c. A discussion of the wattmeter test is contained in PSC letter P-85040
dated January 31, 1985. (The discussion of the wattmeter test in
Attachment 3 to this SE indicates that further improvement is
necessary. )
d. A discussion of the positive stop to prevent overdriving is contained
in P-85032 dated January 30, 1985 and was discussed during our
February 21, 1985, meeting.
The final resolution of these long term issues will be the subject of a
future SE.
Summary
All of the issues presented in the NRC's Assessment Report have been
addressed by PSC as discussed above. Most of the near term issues have
been satisfactorily resolved as have some of the long term issues. The
remaining issues and additional issues requiring resolution prior to
restart will be (or have been) the subject of separate SEs.
Since there have been numerous commitments by PSC to provide various
documents and/or implement various programs or procedures, we will
confirm these commitments in writing in connection with authorization of
plant restart.
Attachments:
1. SE on CRDM and RSS
2. TER on CRDMs, RSS, and Tendons
3. SE on Control Rod Position Indication.
rJ400NEou4r Attachment 1
0 UNITED STATES
NUCLEAR REGULATORY COMMISSION
w ' WASHINGTON,D.C.20555
r4' .0
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
FORT ST. VRAIN NUCLEAR GENERATING STATION
PUBLIC SERVICE COMPANY OF COLORADO
AND RESERVE SHUTDOWN SYSIEM
Docket No. 50-267
1.0 Introduction and Background
On June 23, 1984, a failure occurred at the Fort St. Vrain Station (FSV)
in which 6 of 37 control rod pairs failed to insert on the receipt of a
scram signal . As a result of this incident and other problems at FSV,
the Director of the Office of Nuclear Reactor Regulation asked his staff
to assess several aspects of operations at FSV. The resulting
assessment (Reference 1) defined actions required of the licensee before
the plant could be restarted. It also defined longer term issues for
the licensee to address.
Subsequent to the June 23, 1984 event, on November 5, 1984, a reserve
shutdown hopper failed to fully discharge during a routine Surveillance
Test. Further meetings were held with the licensee on November 28
and 29, 1984 and on January 15, 1985. During the January 15 meeting,
additional commitments made by the licensee and requirements
for restart were clarified (References 2 and 3). Key commitments made by
the licensee included refurbishment of all control rod drive mechanisms
(CRDMs) and replacement of all reserve shutdown system (RSS) material .
In addition, the licensee provided a number of submittals in response to
the restart issues addressed in the assessment report. These submittals
are referenced in the Technical Evaluation Report (TER) (Reference 4),
which is enclosed with this evaluation. The TER was prepared by Los
Alamos National Laboratory under contract to the NRC.
This evaluation has been performed by the Office of Nuclear Reactor
Regulation in partial fulfillment of Item 1 of Technical Interface Agreement
(TIA) 85-01 dated January 2, 1985. This TIA was formulated in response to
a request from Region IV (Reference 5). The key issues covered in this
evaluation are:
- Control Rod Drive Mechanism Failures
- CRDM Refurbishment Program
- CRDM Temperature Recording and Requalification
- CRDM Surveillance and Preventative Maintenance
2.0 Evaluation of Individual Issues
Restart of the Fort St. Vrain Nuclear Generating Station is dependent on
satisfactory evaluation of several major issues identified in the
assessment report (Reference 1). The following evaluations are based on
the enclosed TER (Reference 4).
- 2 -
2.1 Control Rod Drive Failure Mechanisms (Reference 4 - Section 2.1)
Failure of the control rods to insert on a scram signal has been
attributed to two mechanisms. The first failure mechanism is the
accumulation of debris in both the CRDM motor bearings and gear
mechanism. This debris could eventually bind the mechanical drive and
prevent control rod insertion. The second failure mechanism is the
potential effect of temperature and moisture on the CRDMs. Loss of CRDM
purge flow can result in the CRDM binding due to differential expansion
effects as well as a change in properties of the molybdenum disulfide
lubricant.
Since the failure mechanism(s) has not been definitely identified, the
licensee has proposed multiple measures in response to each failure
mechanism. First, each of the 37 CRDMs will be completely refurbished
prior to restart.
Second, the temperature of the CRDMs will be monitored. Third, backup
source of purge flow capable of two hours of operation will be
installed. Fourth, the licensee will institute a program to qualify the
CRDMs for a temperature of 300°F.
We find that the measures proposed by the licensee address each of the
proposed failure mechanisms. However, we conclude that further work
must be done to assure that the CRDMs are operable over the remaining
plant life. These long term issues will be discussed in Sections 2.3
and 2.4. We conclude the work to establish the CRDM failure mechanism
is adequate and an acceptable basis for plant restart.
2.2 CRDM Refurbishment Program (Reference 4 - Section 2.2)
In view of the mechanical deterioration of the CRDM motor bearings and
gear trains, the licensee has elected to perform a complete
refurbishment of the CRDMs. Additionally, since the licensee also found
deterioration in the CRDM cables and the Reserve Shutdown System (RSS)
material , he elected to replace these items with new materials.
CRDM Refurbishment
The licensee's program for refurbishment of the CRDMs involves inspection,
testing, refurbishment or replacement of all major components. We find
that the refurbishment process is thorough and represents an effort to
restore the CRDMs as closely as possible to as-new conditions.
Furthermore, the licensee has proposed a testing program to establish
the acceptability of the CRDMs following refurbishment. This program
involves a multiple step testing program, with eventual tests in the
reactor core. Certain aspects of this testing procedure are still in
development and cannot be reviewed at this time.
We conclude that the licensee's mechanical refurbishment program is
adequate and provides an acceptable basis to support plant restart.
- 3 -
However, we also conclude that an improved testing and surveillance program
should be formulated to assure continued CRDM operability during plant
operation. We have discussed these long term items in Sections 2.3 and 2.4.
Other Refurbishment Items
The licensee has elected to replace the control rod cables which were found
to be failing from stress corrosion. The licensee has selected different
materials for portions of this refurbishment to improve the resistance of
these materials to future stress corrosion failures. We have reviewed the
replacement materials proposed by the licensee and found them acceptable.
We recommend continued investigation by the licensee of the sources of
chlorine in the reactor and its potential effects on other reactor components.
The licensee has also elected to completely replace the materials contained
in the RSS hoppers. The replacement was in response to the failure of a hopper
to discharge due to formation of boric acid crystals on the RSS material .
The licensee selected a new material with improved properties to reduce
the likelihood of future boric acid crystals. The licensee has also
taken measures to reduce the possibility of moisture ingress into this
system.
We have reviewed the licensee's program for replacement of the control
rod cables and the RSS materials and found it acceptable.
2.3 CRDM Temperature Recording and Requalification (Reference 4 - Section 2.3)
The licensee has proposed to upgrade the CRDM temperature measuring
systems to provide continuous records. He intends to monitor weekly the
CRDM temperature, in all operating conditions. We have reviewed this
monitoring program and find it is acceptable for steady state operation.
However, we recommend it be supplemented with a transient monitoring
program to provide an increased monitoring frequency during transient
events that could lead to high CRDM temperatures.
The licensee has proposed that the CRDMs are currently qualified to
272°F, and recommends plant operation with an administrative temperature
limit of 250°F. In addition, the licensee is pursuing a program to
requalify the CRDMs to temperatures of 300°F. Our evaluation of the CRDM
qualification based on test data only supports an average operating
temperature of 215°F. The licensee has indicated that the 215°F temperature
limit would impact on plant operation.
We conclude that pending the receipt of new qualification data, the
temperature of the CRDMs be limited to 215°F as an administrative
limit. In the event that 215°F is exceeded, continuous monitoring of
the affected CRDM must be initiated and the results reported to the NRC
on a monthly basis. The CRDMs, until requalified, should not be
operated at higher than 250°F. We require that the licensee provide
a commitment to operate within these temperature limits until further
requalification test data is available or provide another method of
assuring CRDM operability. These temperature limits cannot be changed
- 4 -
without NRC approval of new temperature limits or alternative methods of
assuring CRDM operability.
2.4 CRDM Surveillance and Preventative Maintenance (Reference 4 - Section 2.4)
The licensee has proposed a program of surveillance and preventative
maintenance to assure continued operability of the CRDMs and RSS during
plant operation. The surveillance program is an interim program, based
mainly on weekly ten inch control rod drops. Our evaluation of this
surveillance program is that it may not provide the detailed information
needed to predict potential failures of the CRDMs to insert on a scram
signal over the long term. The licensee is cognizant of the limitations
of the current tests, and is developing more thorough surveillance tests.
We conclude that the licensee's proposed interim surveillance procedures
provide an acceptable basis for plant restart. However, we require the
licensee to submit an improved (long term) surveillance program within
six months of plant restart for NRC review.
The licensee has also proposed a preventative maintenance program for
the CRDMs and RSS. Certain of these components are normally removed from
the reactor for refueling. At each refueling, the systems and their com-
ponents would be thoroughly inspected and refurbished as needed. Additionally,
as more sensitive surveillance tests are developed, surveillance data could
be used to determine additional candidates for preventative maintenance
activities. We conclude that the licensee's preventative maintenance
program is an acceptable basis for plant restart. However, the licensee
must provide information on the interaction between improved surveillance
program (noted above) and the preventative maintenance activities. This
information should be provided with the revised surveillance program.
3.0 Conclusions on Restart Issues
We have evaluated the licensee's proposed programs to address the issues
covered by this evaluation and related commitments made to the NRC (References
2 and 3). We find that these issues, as discussed in Section 2.0, have been
adequately resolved to allow plant restart with the exception of the following:
1. The licensee must provide a commitment to operate the plant within
the CRDM temperature limits accepted by the NRC. These limits
cannot be changed without NRC approval of new temperature limits
or alternative methods of assuring CRDM operability.
2. The licensee must provide a commitment to submit an improved CRDM
surveillance and preventative maintenance program within six months
of plant restart.
- 5 -
These open issues should be resolved prior to plant startup.
4.0 Long Term Issues
Continued plant operation beyond one refueling cycle should be
contingent on resolution of the longer term items identified in the
Assessment Report of October 16, 1984 (Reference 1). We request that
the licensee commit prior to restart to resolution of these long term
items as outlined in the Assessment Report (Page viii).
Enclosure:
LANL TER
Date: May 21, 1985
Principal Contributor:
K. Heitner
REFERENCES
1. TIA 85-01 dated January 2, 1985
2. Memorandum from J. R. Miller and E. Johnson to F. Miraglia dated
February 5, 1985
3. Preliminary Report Relating to the Restart and Continued Operation of
Fort St. Vrain Nuclear Generating Station, Docket No. 50-267,
October 16, 1984
4. Evaluation of Control Rod Drive Mechanism and Reserve Shutdown System
Failures and PCRV Tendon Degradation Issues Prior to Fort. St. Vrain
Restart. LANL FIN No. A-7290, March 12, 1985 (Enclosed)
Attachment 2
Evaluation of Control Rod Drive Mechanism and
Reserve Shutdown System Failures,
and PCRV Tendon Degradation Issues
Prior to Fort St. Vrain Restart
NEC Fin No. A-7290
March 12, 1985
Los Alamos National Laboratory
•
Deborah R. Bennett, Q-13
Gerald W. Fly, Q-13
L. Erik Fugelso, Q-13
Robert Reiswig, MST-6
Stan W. Moore, Q-13
Responsible NRC Individual and Division
J. R. Miller/ORB3
Prepared for the
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555
DISCLAIMER
Tnis report was prepared as an account of work sponsored by
an agency of the United States Government. Neither the
United States Government nor any agency thereof, or any of
their employees, makes any warranty, expressed or implied,
or assumes any legal liability or responsibility for any
third party's use, of any information, apparatus, product
or process disclosed in this report or represents that its
use by such third party would not infringe privately owned
rights.
- i -
Table of Contents
1.0 Background
1.1 Assessment Report Restart Issues
1.2 PCRV Tendon Restart Issues
1.3 Purpose of the Technical Evaluation
2.0 Control Rod Drive and Orifice Assemblies
2.1 Failure Mechanisms
2.1.1 Motor Brake Malfunctions
2.1.2 Reduction Gear Mechanism Malfunctions
2.1.3 Motor and Motor Bearing Malfunctions
2.2 Refurbishment Program
2.2.1 CRDOA Refurbishment
2.2.2 Control Rod Cable Replacement
2.2.3 Reserve Shutdown System Material-Related Failure
2.2.4 Purge Flow and Seal Replacement
2.3 CRDM Temperature Recording and CRDM Requalification
2.4 CRDM Preventive/Predictive Maintenance and Surveillance
2.4.1 CRDM Preventive/Predictive Maintenance
2.4.2 CRDM Interim Operational Surveillance
3.0 Moisture Ingress Issues
4.0 PCRV Post-Tensioning Tendon System
4.1 Tendon Accessibility, Extent of Known Degradation and Failure
Mechanism
4.2 Tendon Corrosion Corrective Measures
4.3 PCRV Tendon Interim Surveillance
4.4 PCRV Structural Calculations by Los Alamos National Laboratory
5.0 Conclusions
6.0 References
- ii -
Evaluation of Control Rod Drive Mechanism and
Reserve Shutdown System Failures,
and PCRV Tendon Degradation Issues
Prior to Fort St. Vrain Restart
1.0 Background
On June 23, 1984, following a moisture ingress event resulting in a
loss of purge flow to the Control Rod Drive Mechanism (CRDM) cavities, 6
of 37 control rod pairs in the Fort St. Vrain (FSV) High Temperature Gas-
Cooled reactor failed to insert on a scram signal. Subsequently, all six
control rod pairs were successfully driven into the core.
In July, 1984, an assessment team consisting of Nuclear Regulatory
Commission (NRC) personnel from Headquarters, Region III and Region IV,
and their technical consultant, Los Alamos National Laboratory, conducted
an on-site review of the Control Rod Drive Mechanism failures, overall
conduct of plant operations, adequacy of technical specifications and a
review of the continued moisture ingress problem. An additional plant
visit in August, 1984, reviewed CRDM instrumentation anomalies.
1.1 Assessment Report Restart Issues
The results of both assessments were reported in the "Preliminary
Report Related to the Restart and Continued Operation of Fort St. Vrain
Nuclear Generating Station"1, in October, 1984. The report concluded
that Fort St. Vrain should not oe restarted until modifications and/or
other corrective actions had been taken, or until all control rod drive
mechanisms had been inspected and refurbished to provide reasonable as-
surance that the control rods would insert automatically on receipt of a
scram signal. More specifically, and as included in this technical eval-
uation, Reference 1 required Public Service Co. of Colorado (PSC) to com-
plete the following, prior to restart:
a. The licensee must identify the CRDM failure mechanism(s) and
take corrective actions, or, if the mechansm(s) cannot be posi-
tively identified, take other compensatory measures to provide
assurance of control rod reliability, which could reasonably
include refurbishment of all CRDMs.
- 1 -
b. The licensee must outline and commit to periodic inspection,
preventive maintenance and surveillance programs for control
rod drive mechanisms and associated position instrumentation.
A change in the Technical Specifications shall be proposed to
implement a weekly control rod exercise surveillance program
for all partially or fully withdrawn control rods. A Limiting
Condition for Operation should define control rod operability,
and the minimum requirements for rod position indication.
c. The licensee must functionally test one-20 weight % boron and
one-40 weight % boron hopper from the Reserve Shutdown System
(RSS), to assure the full availability of the RSS, prior to
restart. The licensee must outline and commit to periodic in-
spection, preventive maintenance and surveillance programs for
Reserve Shutdown System material. A change in the Technical
Specifications shall be proposed to implement the RSS surveil-
lance program. A Limiting Condition for Operation should define
and confirm the operability of the Reserve Shutdown System.
d. The licensee should develop a proceaure requiring reactor shut-
down when high levels of moisture exist in the primary coolant,
or when CRDM purge flow is lost.
e. The licensee should implement a procedure for recording repre-
sentative samples of CRDM temperatures at all operating condi-
tions, until continuous recording capability is available.
f. The licensee should implement procedure to prevent overdriving
the control rods past the "Rod-In" limit.
g. The licensee must develop a plan to implement any modifications
recommended by the PSC Moisture Ingress Committee that are
determined, by PSC, to have a high potential for significantly '
reducing the severity and frequency of moisture ingress events.
1.2 PCRV Tendon Restart Issues
As a result of previously identified tendon degradation in the Pre-
stressed Concrete Reactor Vessel (PCRV) post-tensioning system, PSC must
complete the following, as comfirmed by Reference 2, prior to restart:
- 2 -
a. The licensee should submit documentation evaluating the mechan-
ism(s) causing corrosion on and failure of the PCRV tendon
wires, and corrective measures to eliminate further tendon
degradation, thereby assuring the continued structural integ-
rity of the PCRV and its post-tensioning system.
b. The licensee should propose and implement a tendon surveillance
program that determines the extent of current tendon degrada-
tion in the PCRV, and that systematically monitors the rate of
tendon corrosion.
1.3 Purpose of tne Technical Evaluation
This document provides a technical review of tne restart issues
identified above, and the corrective measures and/or actions proposed by
licensee, based on the licensee's January 31, 1985 submittals (References
given as used in this document), and the meeting between the licensee and
NRC at the FSV plant site on February 20-22, 19d5, as transcribed in
References 3, 4 and 5.
2.0 Control Rod Drive and Orifice Assemblies
This section includes a review of CHDM failure mechanisms, Control
Rod Drive and Orifice Assemblies (CRDOA) refurbishment, CRDM temperature
recording and requalification testing, CRDM preventive/predictive main-
tenance and surveillance.
2.1 Failure Mechanisms
The failures of control rod pairs to scram, under various operating
conditions, has been documented since 1982,6,7 and are as noted in
Table 1 by region, CRDOA number and CRDM purge flow subheader (total of 8
purge flow subheaders).
- 3 -
Table 1. Control Rod Failures
Date 2/22/82 6/23/84 1/14/85
Region 7 28 6 7 10 14 25 28 28 31 32
CRDOA # 18 44 29 18 14 25 7 44 36 17 15
CRDM Purge 1 1 6 1 7 2 5 1 1 2 3
Subheader #
High moisture content in the primary coolant and loss of purge flow
were common modes during the 2/22/82 and 6/23/84 events. Substantial
descriptions and operating characteristics of the drive motor, friction
brake and dynamic braking, the reduction gear mechanism, the cable drum
and cable, and the bearing lubricant are provided in Reference 6. The
licensee reviewed those CRDM components that could have caused the fail-
ures to scram, and postulated various failure mechanisms that could have
interacted on each component, as described below.
2.1.1 Motor Brake Malfunctions
During a scram, the motor brake is de-energized and released, thereby
freeing the motor rotor shaft and gear train assembly to rotate under the
torque applied by the weight of the control rods. In the motor brake
assembly, failure of the scram contactor to de-energize dc power to the
electromagnet was discounted because the operator had removed the brake
fuses following the CRDM failures to insert the control rod pairs.
According to the licensee, electromagnetic remanence and reduced
spring constant in the brake spring plungers (due to elevated tempera-
tures) were eliminated as possible failure mechanisms. Some corrosion
and rust was identified on the brake disks of CRDOAs 25, 18 and 29. How-
ever, the disks of a CRUM motor brake assembly with "discoloration and
whatever surface variations"3' p.149, could not be made to stick in an
elevated temperature helium environment with high moisture content (test
T-228). The licensee concluded that the motor brake was not instrumental
in the failures to scram.
Los Alamos agrees with the licensee that the motor brake assembly
was probably not related to the CRDM failures.
- 4 -
2.1.2 Reduction Gear Mechanism Malfunctions
The reduction gear train is driven by the motor rotor shaft, and
rotates the cable drum with a gear ratio of 1150 between the motor and
drum. The condition of the reduction gear mechanism was postulated by
the licensee to potentially contribute to a failure to scram through gear
tooth or bearing damage, by the presence of large particulate matter pre-
venting gear rotation, and/or the presence of particulate matter in the
gears or gear bearings reducing the gear train efficiency--i.e., the
torque transmitted from the gear train to the motor rotor shaft might
have been insufficient to overcome the friction of the motor bearings.
The licensee stated that no major damage has been identified on sev-
eral inspected reduction gear mechanisms, even though some wear and debris
were observed. The licensee's analyses indicated that particulates with
a size of 0.030 inches in diameter or greater, and with a comparable
material composition as the reduction gear mechanism (implying comparable
hardness), would be required to inhibit gear or gear bearing rotation.
Analyses of CRDOA debris8 showed the presence of rust, molybdenum di-
sulfide and traces of silicon particles, which are relatively soft mate-
rials. The average particle of 0.020 inches was uniform in size, and
tended to be smaller than that thought to inhibit rotation, even though
rust particles on the order of 0.0625 to 0.125 inches were scraped off
the ring gear pinion housing of CRDOA 18. However, the presence of debris
in the gears and gear bearings tended to support the licensee's case of
reduced gear train efficiency when sensitivity studies indicated that the
motor bearings were only three times more sensitive to debris than the
first pinion gear mesh of the reduction gear assembly, and 500 times more
sensitive to debris than the cable drum bearings.
Los Alamos agrees with the licensee that the presence of debris ,
especially in the first pinion gear mesh and the gear bearings, could
reduce the efficiency of the reduction gear train, and thereby contribute
to CRDM failures.
2.1.3 Motor and Motor Bearing Malfunctions
During a scram, the motor is de-energized and does not directly con-
tribute to the scram process, even though it operates as an induction
generator. However, because 1b-20 inch-ounces of resisting torque on the
- 5 -
motor rotor shaft can forestall scram,9 the friction from the motor
bearings can be a significant contributor to the failure to scram. Pos-
sible contributions to increase the friction include debris in the bear-
ing race, wear on the bearing ball or race, and changes in the lubricant
properties during adverse conditions.
The licensee reported that debris was observed in the bearing races
of CRDOAs 7, 18 and 44, "roughness in rolling the bearing balls was noted
in virtually all of the unrefurbished bearings examined",6 and minor
race wear was identified. Reference 8 verified that the major debris
constituents could be attributed to the motor bearing materials (which
includes bearing balls, races, and other bearing components), whereas
minor constituents were indicative of the motor itself. The analysis
provided little evidence to support the theory that debris had been
"washed" into the bearing races. The licensee also determined, because
of the relatively close bearing tolerances and because rod weight alone
might not produce sufficient "crushing force" to deform bearing particu-
late, that bearing operation could be reduced with the presence of par-
ticulate matter. Tne licensee therefore concluded that internally gener-
ated wear byproducts in the CRDM motor bearings contributed significantly
to the failures to scram.
Los Alamos agrees with the licensee that increased friction in the
motor bearings, caused by the presence of internally generated debris,
could have been a likely contributor to the failures to scram. Los Alamos
also agrees with the licensee that the "wash in" theory of debris into
the motor bearing races is not supported.
Los Alamos contends that the loss of CRDM purge flow allowed primary '
coolant with high moisture content to enter the CRDM cavity. An indepen-
dent literature search indicates that the dry film lubricant, molybdenum
disulfide, MoS2 , experiences an increase in its coefficient of fric-
tion in the presence of moisture38. Therefore, the increased frictional
coefficient of the lubricant on the motor bearings, MoS2, may have also
contributed to the CRDM failures by resisting motor rotor shaft rotation.
2.2 Refurbishment Program
The cause of the failures to scram could be attributed to several
mechanisms such as reduced reduction gear train efficiency, internally
- 6 -
generated debris in the motor bearings causing increased friction on the
motor rotor shaft, and possibly an increased frictional coefficient in
the dry film lubricant in the presence of moisture. Because the CRDM
failure mechanism cannot be specifically delineated, and because of CRDM
cable failures, the licensee has undertaken a refurbishment program, in-
volving the CRDM motors and reduction gear mechanisms, on all 37 CRDMs.
The licensee reported that the CRDM refurbishment process and a testing
program will ensure the ability of the control rods to scram under oper-
ating conditions.
In addition, the licensee has elected to replace the control rod
cabling and other connecting hardware in light of recently identified
stress corrosion problems, to replace the Reserve Shutdown System material
due to the discovery of material "bridging" during hopper discharge, ane
to install seals around certain penetrations into the CRDM cavity to
mitigate the effects of primary coolant ingress by natural circulation.
2.2.1 CRDOA Refurbishment
The licensee has proposed complete refurbishment of all Control Rod
Drive and Orificing Assemblies to ensure that the CRDOAs will perform
their intended safety functions, and to avoid potential operability prob-
lems that could limit plant availability. As specified in Reference 10,
the following major components are to be inspected, tested, refurbished
or replaced, as necessary:
1. Control Rod Drive (200) Assembly--shim motor and brake assembly,
bearings, reduction gears, limit switches/potentiometers.
2. Orifice Control Mechanism--orifice control motor, bearings,
potentiometer, gears, drive shaft and nut, drive shaft housing.
3. Control rod clevis bolts.
4. Reserve Shutdown System--boron balls, rupture disks, DP switch.
Design modifications include the replacement of control rod cables,
cable end fittings, and cable clevis bolts, the installation of new purge
seals into the CRDM cavity, and the installation of RTDs (Resistance Tem-
perature Detectors) in all CRDOAs--the impact of tnese design changes
will be evaluated later in this report.
Each CRDOA will undergo the following series of scram tests in the
refurbishment process6: a pre-refurbishment, in-core full scram test; a
- 7 -
pre-refurbishment full scram test in the Hot Service Facility (HSF); a
scram test with refurbished reduction gear mechanism and unrefurbished
shim motor, using dummy weights; a full scram test using a "standardized"
motor, using dummy weights; a scram test with completely refurbished 200
assembly, using dummy weights; a post-refurbishment, full scram test in
the HSF; and finally, a post-refurbishment, full in-core scram test.
As designated by the licensee in Reference 6, back-EMF voltage meas-
urements from the shim motor will be taken for the series of scram tests
conducted before, during and after refurbishment, and should define the
CRDM operating characteristics. From the back-EMF voltage measurements,
the licensee states that they can generate the following information--
voltage versus time, frequency versus time, voltage versus frequency,
acceleration versus time, torque versus time, peak angular velocity, time
to peak back-EMF and angular velocity, average torque on motor rotor dur-
ing acceleration to peak velocity, maximum torque on motor rotor each 10
second interval, maximum deviation of torque values each 10 second inter-
val, and gear train efficiency.
The licensee has proposed a CRDOA refurbishment acceptance criterion,
taking into account the results of the back-EMF voltage measurements and
the resulting calculations of acceleration and torque such that6:
1. Tne minimum calculated average torque during acceleration to
peak velocity will be 17.0 inch-ounces; this value corresponds
to an average acceleration to peak velocity of 98.83 radians/
second2.
2. The maximum torque calculated during "steady-state" will be 7.0
inch-ounces.
According to the licensee, final acceptance of a refurbished CRDOA will
be based upon the results of its in-core full scram test.
Los Alamos agrees with the mechanical refurbishment of all CRDOAs,
as the program is currently being implemented by the licensee. In par-
3, pp. 174-75
titular, the replacement of shim motor bearings is con-
sidered essential to the refurbishment process. However, the current
program of mechanical refurbishment alone cannot ensure CRDOA
operability.
- 8 -
From the documentation presented by the licensee and reviewed earlier
in this section, Los Alamos believes that the proposed back-EMF testing
and acceptance criteria have potential in providing a data base from which
control rod operability might be determined. But, an element of uncer-
tainty, as to CRDOA operability based on back-EMF testing, is introduced
because the test method and interpretation of its results are still in
the developmental stages, and because in-core full scram testing of re-
furbished CRDOAs has not yet taken place.
Los Alamos recommends that the back-EMF testing method continue to
be developed, that the further collection of back-EMF information be used
in preparing a statistical data base for possibly defining CRDOA opera-
bility, and that more attention be paid to the initial, start-up scram
characteristics of the CRDOA, in developing a better understanding of
break-away torque effects. In line with Region IV's increased inspection
of the refurbishment process, we suggest a review, by Region IV, of all
testing results pertaining to CRDOA refurbishment acceptability, after
in-core testing is complete, but prior to startup. As an additional
method to ensure CRDOA operability during scram, a procedure requiring
control rod run-in is recommended.
As a post-startup item, Los Alamos recommends that a final determina-
tion be made as to the suitability and acceptability of back-EMF testing
in defining CRDOA operability.
2.2.2 Control Rod Cable Replacement
In September, 1984, the control rod cable on CRDOA 25 was severed in
several places during an investigation of a slack cable indication.11
A subsequent metallurgical examination12 of the austenitic 347 stainless
steel cable indicated that the cable surface was pitted and cracked, that
the delta-like material cracks were typical of stress corrosion cracks,
and that the fracture surfaces were brittle in nature. Further investi-
gation revealed that the 347 SS cable material was susceptible to stress
corrosion when under the existing stressed conditions, and in the presence
of chlorides and moisture.
Tne potential sources of the chlorides in the primary coolant con-
tributing to the chloride stress corrosion are reviewed in Reference 13.
The licensee states that the chlorine occurs as two different species--HC1
- 9 -
gas and a salt; the sources of the gas species include the fuel rods ,
H-327/H-451 graphite, PGX/HLM graphite and the Ti sponge, whereas the
sources of the salt species include the ceramic insulation, concrete and
water, all to varying degrees.
As part of tne overall CRDOA refurbishment program, the licensee
elected to replace the control rod cable witn Inconel 625, which is con-
sidered resistant to chloride stress corrosion, and has increased strength
and fatigue properties over the former 347 SS. Cable components and con-
necting hardware that were made from materials susceptible to stress cor-
rosion, and are being replaced with materials more resistant to stress
corrosion include:
Component Material
1. Cable and rod portion Inconel 625--high strength
of the ball end and resistance to oxidation
2. Anchor, set screw Martensitic steel-high
strength, ability to be
nitrided, resistance to
oxidation
3. Spring, connecting bolt Inconel X-750--high yield
strength, resistance to
oxidation.
Drawing numbers and material information are available in Reference 12.
A safety analysis of tne material changes in tne reactor control rod drive
and orificing assembly, which are classified as Class I, Safety Related
•
and Safe Shutdown components, is included in Reference 14.
Los Alamos metallurgical analyses on a sample of the corroded control
rod cable15 also indicate pitting on the cable surface, ductile and
brittle fracture surfaces, and to a lesser degree than the licensee ,
cracking indicative of stress corrosion cracking. Qualitative measure-
ments confirm the presence of chlorine on fracture surfaces. Therefore,
Los Alamos agrees tnat chloride stress corrosion contributed to the de-
graded condition of the control rod cable. The Los Alamos analysis also
observed that a certain particle removed from between the individual cable
strands of the Los Alamos sample had a "shaved" appearance, and was
- 10 -
identified as a 7000 series aluminum alloy--the licensee noted that the
control rod cable drum is constructed of 7075 aluminum alloy4; p.20,
and that no excessive drum wear had been noted.
Los Alamos agrees that the licensee's recommended material changes
tend to improve the overall resistance of the CHDOA cable components and
connecting hardware to chloride stress corrosion. However, Los Alamos
also recommends a continued analysis into the sources of the chlorine and
its effects on other reactor components, especially components potentially
subjected to high chlorine concentrations such as the bottom plenum or
other areas where water could accumulate.
2.2.3 Reserve Shutdown System Material-Related Failure
In November, 1984, during the required testing of a 20 weight $
boron and a 40 weight % boron hopper in the Reserve Shutdown System, only
half of the RSS material in CRDOA 21 (40 weight $ boron) was discharged.
The licensee's examination of the undischarged material revealed that the
4C boronated graphite balls had "bridged" together through a crystal-
line structure on the ball surfaces. Analyses on the crystalline material
indicated that it was boric acid.16 The formation of the boric acid
crystals was caused by moisture reacting with residual boric oxide in the
RSS material. It was concluded that the moisture had entered the RSS
hopper through the CRDOA vent/purge line by "breathing", and/or by water
contamination in the helium purge line.
In Reference 16, the licensee proposed a threefold corrective action
to the RSS material problems. First, new RSS material, manufactured by
Advanced Refractory Technologies (ART) in late 1984 and early 1985, has
an order of magnitude less residual boric oxide in the B4C material,
and will be installed in all RSS hoppers as part of the overall CRDOA
refurbishment program. No effort will be mane to use ART blended RSS
material currently in stores4' p.32 unless NRC is notified. Second, an
expanded HSS material surveillance program, which will be incorporated
into the Technical Specification, will test one 20 weight % boron hopper
and one 40 weight $ boron hopper during each refueling outage, and will
include visual examinations for boric acid crystal formations, chemical
analyses of RSS material for boron carbide and leachable boron oxide con-
tent. Third, efforts will be mace to mitigate or eliminate the ingress
- 11 -
of moisture into the RSS hoppers by installing a knock-out pot, moisture
elements, and a back-up helium source for the main CRDOA purge and Reserve
Snutdown System purge lines.17 Each knock-out pot will be equipped
with a sight glass and a high level alarm in the Control Room.
Los Alamos concurs that the crystalline structures on the surface of
the B4C RSS balls is meta-boric acid,18 most probably formed by mois-
ture reacting with leachable boric oxide in the B4C material. In light
of the new RSS material to be used, the increased surveillance efforts ,
and measures to mitigate the ingress of moisture in the ASS hoppers, Los
Alamos agrees that the refurbished RSS should be able to reliably perform
its function.
2.2.4 Purge Flow and Seal Replacement
Just prior to the June 23, 1984 event when 6 of 37 control rod pairs
failed to insert on a scram signal, a high moisture content in the primary
coolant resulted in the loss of purge flow into the CRDM cavities. The
loss of purge flow may have allowed the additional ingress of moist pri-
mary coolant into the CRDM cavities, resulting in mechanisms that may
have contributed to the CRDM failures. Because the exact CRDM failure
mechanism has not been determined, and to alleviate the possibility of
purge flow loss and/or hign moisture content in the primary coolant con-
tributing to future CRDM failures, the licensee has proposed several cor-
rective measures19 as part of the overall CRDOA refurbishment program.
To provide an accurate measure of the purge flow into the CRDM cavi-
ties, the licensee has proposed tnat new flow indicators with a range of
0-20 scfm be installed on each helium purge line, providing local indica-
tion, remote indication in the Control Room, and an alarm in the Control
Room to indicate low flow conditions.20 A minimum of 8 bypass lines
(one line serviced by each of tne 8 purge flow subheaders) will be in-
stalled prior to restart. Tne licensee intends to install the flow
instrumentation4, pg 3-4 on tnese subheaders when the devices are
available.
As mentioned in section 2.2.3, to reduce tne possibility of moisture
ingress into tne CRDM cavities via the helium purge lines, the licensee
will install a knock-out pot, moisture elements and a back-up helium
source for the main CRDOA purge and RSS purge lines, prior to criticality
- 12 -
following the fourth refueling outage4, pg 5. The knock-out pots will
be equipped with a sight glass and a high level alarm in the Control Room.
The helium trailer, which will act as the back-up source of dry helium
for purge, can provide helium at a rate of 7.4 acfm (4.5 lbms/hr per pen-
etration at 700 psig) for approximately 2 hours.i7
To mitigate the ingress of primary coolant, which could contain
moisture, into the CRDM cavity, seals will be installed on four large
flow passages into tne CRDM cavity--the two passages in the reserve shut-
down tube holes, and the two passages over the eye bolts that penetrate
the floor of tne CRDM cavity.21 Cover plates with integral gaskets
will also be installed on the four access openings on the lower CRDM
housing. Thermal and mechanical analyses22 have determined that the
seal additions will not interfere with the R6S performance under the in-
fluence of mechanical , thermal or seismic loadings. Tne flow calculations
in Reference 22 conclude tnat addition of tne mechanical seals to the RbS
pressure tubes and the lifting eyebolts will reduce naturally convective
ingress of primary coolant into the CRDM cavity from a flow rate of 0.68
acfm to less than 0.006 acfm. Additional calculations have confirmed
that the seals are able to withstand both a design basis slow depressuri-
zation transient and a design basis rapid depressurization transient.
The licensee has proposed a procedure in Reference 23 tnat basically
requires reactor shutdown in the event CRDM purge flow is lost, or if
high moisture content is present in tne primary coolant.
Los Alamos agrees with tne efforts of tne licensee in monitoring the
flow and moisture content of tne helium purge into the CRDM cavities, in
restricting the ingress of moisture into the CRD cavities via the purge •
lines, and in providing a back-up source of helium in case of purge flow
loss. From the review of the provided documentation, Los Alamos agrees
that the addition of seals and coverplates with integral gaskets will
indeed mitigate the ingress of primary coolant and moisture into tne CRD
cavities through penetrations.
In addition, Los Alamos believes tnat the procedure requiring reactor
shutdown with loss of purge flow or high moisture levels in the primary
coolant fulfills the requirements of the assessment report1. Tne
licensee defines "high moisture levels" in Reference 23.
- 13 -
2.3 CRDM Temperature Recording and CRDM Requalification
Tne lack of direct measurements of CROM temperatures curing tne June
23 event, and during steady state and other transient operating condi-
tions, has prompted the installation of Rills to monitor tne CRDM cavity
closure plate (ambient) , orifice valve motor plate and control roc drive
motor temperatures. Strip chart recorders will continuously record tne
tnree temperatures for each CRDM,24 and will provide a CALM operating
temperature data base. The old data collection surveillance procedure25
4, p.60
will be modified to collect data on a continuous basis. Tne
licensee intends to install the permanent recorders prior to
restart.4, p'57
The licensee postulates in Reference 26 that "the maximum temperature
rating of tne Drive mecnanism which might innibit the scram function is
272°F", and in monitoring CRDM temperatures "the maximum temperature
rating of 272°F should not be exceeded curing power operation".
Tne licensee has also proposed a CRDOA requalification testing pro-
gram that is designee to establisn a temperature at which the CRDOA is
qualified for operation.27 The helium test environment will be operated
at 2o0°F, 260°F, 270°F, 2o0°F, 290°F and 300°F with a goal of qualifying
all CRDOA components for 300°F operation. Results of tne requalification
testing are anticipated by tne end of 1965.
Los Alamos agrees tnat the placement of CRDOA tnermocouples, and tne
continuous data monitoring at all operating conditions is sufficient to
provide a CRDOA temperature data base during steady state and transient
operating conditions.
In addition, Los Alamos believes tnat the CRDOA is currently only
qualified to operate up to 215°F based on the original mechanical CRDOA
qualification tests, an NRC recommendation,28 and previous Los Alamos
calculations.29 Tne licensee's argument tnat the CRDOA is qualified
for 272°F operation based on analytical calculations4, p.49 is not sub-
stantiated. Tnerefore, Los Alamos recommends tnat CRDOA operation be
limited to 215°F until mechanical requalification supports a higher oper-
ating temperature.
- 14 -
2.4 CRDM Surveillance and Preventive/Predictive Maintenance
The licensee has proposed a set of preventive/predictive mainte-
nance tests and surveillance inspection procedures that are intended to
monitor the performance of the CHDOAs and to determine the overall opera-
bility of the CRDOAs during reactor operation. Initial development of
these operating tests are considered part of the CRDOA refuroishment pro-
gram, and will utilize the data base and resultant trends formulated dur-
ing refurbishment.
2.4.1 CR111 Preventive/Predictive Maintenance
Tne licensee's CRDOA preventive/predictive maintenance program is
proposed in Reference 30. According to the licensee, the normal preven-
tive maintenance (PM) program will be implemented on a refueling basis
rotational cycle for CRDOAs that would normally be removed for refueling,
unless the predictive maintenance (PDM) program indicates tne need for
more frequent maintenance. The PM program would emphasize the mecnanical
examination and refurbishment of the shim motor/brake assembly, the drive
train, control rod cable, reserve shutdown system, position potentiom-
eters, limit switches, orifice drive motor assembly, orifice drive lead
screw, assorted seals, ;valves, electrical components, bolts and the ab-
sorber string.
On the other hand, the predictive maintenance techniques would be
used to monitor the most important aspect of CRDOA performance--the
"scram capability"--by determining the shim motor/brake and gear train
performance. The tests proposed in the PDM program include wattage
requirements, back-EMF voltages, delivered torque at tne motors, scram
times, rod drop rates and torques to rotate motor/brake assemblies. Cer-
tain aspects of the PUM program would be implemented on a weekly basis to
determine scram capability and temperature performance during power oper-
ation. The licensee has also proposed that testing information be
acquired during reactor shutdown for trending purposes.
Los Alamos concurs with the proposed preventive maintenance pro-
gram as outlined by the licensee, on the assumption that data acquired
during reactor operation will show that predictive maintenance tecnniques
can be used to detect a reduction in CRDOA performance. Tne PDM testing
tecnniques are closely linked to tne techniques that are being used for
- 15 -
the acceptance criteria in the refurbisnment program, and will tnerefore
be dependent on the suitability and acceptability of back-EMF testing for
determining CRDOA operability, as discussed in section 2.2.1.
2.4.2 CRDM Interim Operational Surveillance
Tne licensee's CRDOA interim surveillance program is proposed in
Reference 31. The surveillance tests are scheduled on a weekly basis,
using a 10" rod drop method on all withdrawn and partially inserted con-
trol rods, except the regulating rod.s' pg 82 The surveillance tests
will obtain data for analysis and long term trending, exercise the rod,
test selected circuitry, verify FSAR (Final Safety Analysis Report)9
assumes scram times, and confirm control rod operaoility. In addition,
CitDOA temperature and purge flow information will be collected.
For a fully withdrawn rod, analog and digital position
information will be obtained, "Rod-Out" lights will be verified on,
"Rod-In" aria "Slack Cable" lights will oe verified off, and the rod will
be dropped approximately 10" by de-energizing the brake, while back-EMF
data are obtainea for future trending. The "Rod-Out" light indication
will be verified off, and analog and digital information will be compared,
with an acceptable deviation of 10 inches between position indications.
The rod will then be withdrawn to the full out position, so that analog
ana digital positions can again be ootained. Control rods that are par-
tially or fully inserted will undergo variations of this method.
Quarterly surveillance tests are intended to supplement weekly sur-
veillance information, and to verify redundancy of selected control roe
position limit switches. Refueling snutdown surveillance will acquire
the same information as the weekly and quarterly tests, except full stroke
insertion tests will be performed.
The operaoility acceptance criteria, according to the licensee, will
be based on distance and time rod drop data used to calculate a conserva-
tive average full lengt❑ scram time. A CRDOA will be considered inoper-
able if it does not meet the maximum scram time of 160 seconds as defined
in the FSAR9. Such an indication would warrant back-EMF testing in
confirming scram operability.
Los Alamos agrees that the basic surveillance methodology is suffi-
cient to exercise the control rod, verify FSAR scram times, and to test
- lo -
selected circuitry. However, references to 272°F as the maximum CRDOA
operating temperature are still considered inappropriate as discussed in
section 2.3, and a 10 inch deviation is not considered acceptable between
digital and analog position indications--such a deviation coulo inadver-
tently lead to control rod overdrive through a misinterpretation of rod
position. Also, the back-EMF testing methods and interpretation of the
results are still in the developmental stages, and an engineering deter-
mination of tne suitability and acceptability of tnis testing metnod in
determining continued CRDOA operability will need to be mace before the
licensee can finalize tnis portion of tne surveillance program.
3.0 Moisture Ingress Issues
The licensee nas submitted32 a listing of the issues considered,
and actions taken, by the FSV Improvement Committee (formerly the FSV
Moisture Ingress Committee) in significantly reducing the frequency and
severity of moisture ingress events. The issues were divided into four
categories:
1 . Issues currently under consideration by tne Fort St. Vrain Im-
provement Committee.
2. Circulator Auxiliary System modifications yet to be completed
prior to startup.
3. Circulator Auxiliary System modifications to be completed prior
to startup, provided material availability and schedule permits.
4. Items identified by tne Moisture Ingress Committee wnicn are
installed and operational.
Los Alamos believes that a listing of intended and installed modifi-
cations does not provide any indication as to what any given modification
really is, wny tney contribute to tne reduction in potential for moisture
ingress events, nor which improvements will substantially reduce the
severity and frequency of moisture ingress events. Tne licensee nas com-
mitted to submit a more explanatory version of tne actions to mitigate
moisture ingress, prior to restart
4, pg 80
4.0 PCRV Post-Tensioning Tendon System
In tne spring of 1984, during scheduled PCRV tendon surveillance,
tendons with corroded and broken wires were found. Since that time, the
- 17 -
licensee has evaluated the corrosion mechanism, has performed lift-off
tests on selected tendons to determine their load-carrying capability,
and proposed corrective actions and an increased surveillance procedures.
4.1 Tendon Accessibility, Extent of Known Degradation ana Failure
Mechanism
The licensee, in determining the extent of tendon corrosion in the
PCRV, determined what fraction of the tendons were available for visual
examination ana lift-off tests. Tne tendon system is subdivided into
four major groups: the 90 longitudinal (vertical) tendons have 169 wires
per tendon; the 210 circumferential tendons in tne PI:tV sidewail have 1)2
wires per tendon, and the 50 circumferential tendons in botn the top ano
bottom heaas have 169 wires per tendon; tne 24 bottom cross-heap tendons,
and 24 top cross-nead tendons nave 169 wires per tendon. Of the four
groups, the licensee states the following accessibility33:
Tendon Group Both Ends Acces. One End Acces. Neither
End Acces.
Longitudinal
Visual 20 69 1
Lift-off 0 74 16
Circumferential
Visual 2d1 2
Lift-off 236 62 12
Bottom cross-heau
Visual 20 4 0
Lift-off lb 4 4
Top cross-neaa
Visual 17 7 0
Lift-off 16 6 2
Tne number of tendons witn Known broken wires as identifies in tue
licensee's 1904 surveillance,34 include° 10 longitudinal tendons with 1
to 22 broken wires, 2 circumferential tendons witn 2 and 15 broken wires,
8 bottom cross-Head tendons with 1 to 19 broken wires, and no top cross-
head tendons with broken wires. In some cases, tne total number of cor-
rodeo, broken wires include wires broken during lift-off tests, or during
retensioning.
- 18 -
The results of 74 longitudinal lift-off tests35 indicated that
tendons with identified broken wires generally had a slightly smaller
lift-off value than intact tendons. Thirty lift-off tests on circumfer-
ential tendons snowed little change in lift-off value. Some of the fif-
teen bottom cross-head tendon lift-off tests showed a definite reduction
in lift-off value for tendons with multiple wire breaks. The value of
tne lift-off test on one top cross-head tendon was nominal. All lift-off
test values exceedeu tne minimum limits.
The licensee conducted metallurgical investigations into the cause
of the corrosion, and determined that microbiological attack on tne tendon
NO-OX-IL CM organic grease caused the formation of formic and acetic
acids,34,36 Tne acids, in conjunction with moisture in tne tendon tube,
vaporized and recondensed on the cooler portions of the tendons--in this
case, usually toward the tendon ends. Tne acidic attack resulted in re-
duced cross-sectional wire area, stress corrosion cracking, localized
tensile overload and wire breakage.
Los Alamos believes, based on the documentation presented by the
licensee, that microbiological attack of the tendon grease and the resul-
tant formation of acetic and formic acids, in the presence of moisture,
is a probable cause for tne currently observed tendon corrosion, and has
led to the subsequent wire breakage through tensile overload. However,
Los Alamos believes that the extent of known tendon corrosion, breakage
and previous suryeillance have not been clearly defined by the licensee.
Los Alamos therefore recommends that a complete map oe mace that lists
each tenoon, its visual examinations and lift-off values, and tne number
ano location of corroded and broken wires. An indication of tne degree
of wire corrosion would also be desiraole.
4.2 Tendon Corrosion Corrective Measures
The licensee evaluated several methods for arresting the corrosion
process,34,36 including the use of ozone as a biocide to kill the micro-
organisms, the use of an alkaline grease which should not be conducive to
microbiological growth, and the use of an inert blanket consisting of
nitrogen gas. The licensee's consultants found that the nitrogen atmo-
sphere arrested the growth of the microbes in the NO-OX-ID CM organic
- 19 -
grease34,35, and eliminated the oxygen wnich is necessary for the cor-
rosion process to continue. based on these results, and as a snort term
action, the licensee has proposed that nitrogen blankets be establisneu
on the longitudinal and bottom cross-Head tendons. Long term actions
would include further investigations into tne corrosion process and ar-
resting techniques, and the possible installation of additional load cells
in monitoring the PCRV behavior.
Los Alamos believes that the use of a nitrogen blanket to halt tne
corrosion process may be suitable, but difficult to implement as proposed.
The tendon tubes are not likely to oe leaktight, and maintaining an inert
gas atmosphere at a set over-pressure may prove difficult. Consideration
might be given to maintaining an intermittent or continuous purge flow
through the tendon tubes, as needed, rather than to maintaining a speci-
fied overpressure. However, Los Alamos recommends that initially the
nitrogen be purged through the individual tendon tubes to remove as much
moisture as possible, and tnat gas samples be used to monitor moisture
and oxygen reduction. Further investigation into the long term effects
of a nitrogen blanket on tendons, tne corrosion process and currently
available corrosion acids are also recommended.
4.3 PCRV Tenoon Interim Surveillance
Because the total extent of tendon corrosion in tne PCRV is unknown,
because the rate of existing corrosion is unknown, and because tne use of
a nitrogen blanket as an arrest to tne corrosion process is an unknown,
the licensee has proposed an interim surveillance program designed to
address eacn of tnese issues.5, pp•164-7• The interim tendon surveil-
lance program would include increased visual and lift-off surveillance
for three years, or until effective corrosion control Has been estab-
lished. Two populations of tendons would be inspected--a population of
tendons that have not been previously identifies as being corroded, and a
control population with known corrosion. On a six-month frequency, visual
surveillance of both tendon ends, when accessible, would include:
- 2U -
Tendon Group No. of New Tendons No. of Control Tendons
Longitudinal 24 6
Circumferential 13 3
Bottom cross-head 6 2
Top cross-head 1 1
• Lift-off tests would be performed on two frequencies--an to month
frequency for the population of new tendons, and a b month frequency for
the control population. The number of tendons for lift-off will include:
Tendon Group No. of New Tendons No. of Control Tendons
Longitudinal 12 3
Circumferential 13 3
Bottom cross-head • 3 1
Top cross-head 1 1
As an acceptance criteria, tne licensee proposed that, based on vis-
ual examinations, a mandatory engineering evaluation be conducted on any
tendon tnat has 20% of its wires broken. For any tendon that has only
one accessible end, the mandatory engineering evaluation will be con-
ducted when any tendon has ioy of its wires broken. The control tendon
population will include those tendons with tne worst known corrosion with
ready accessibility.
Los Alamos agrees that the increased tendon surveillance program of
the nature proposed by the licensee will provide more information on the
extent of corrosion in the PCRV by inspecting new tendons each surveil-
lance, and at the same time, monitor the rate of corrosion with the con-
trol tendon population. Tne increased surveillance should also determine
the effectiveness of the nitrogen blanket in arresting corrosion, or any
other corrective measure the licensee may propose. Los Alamos recommends
that the licensee submit an outline of the intended mandatory engineering
evaluation, which should include all lift-off, load cell and relaxation
data incorporated into a safety evaluation. The licensee should define
the extent of tne visual and lift-off testing procedures, ano could use
US/NRC Regulatory Guide 1.3537 for guidance.
- 21 -
4.4 PCRV Structural Calculations by Los Alamos National Laboratory
The PCRV tendons are intended to apply sufficient compression in tne
concrete to balance or exceed the circumferential and vertical tension in
the concrete tnat results from the internal pressure. A combined analyt-
ical and numerical study39 was undertaken by Los Alamos National
Laboratory to evaluate the evolution of these stresses, both to the ini-
tial prestressing and to subsequent partial and total rupture of tnese
tendons. At the stress levels anticipated in the concrete, and for the
anticipated operating life span of tne PCRV, tne concrete benavior was
modeled as a linear viscoelastic solid with tne creep strain varying pro-
portionally witn tne logarithm of time at constant stress tnroubnout the
projected reactor lifetime.
A one-dimensional model of a long concrete column of rectangular
cross-section, with an embedded prestressing tendon along the length, was
used to evaluate the concrete and steel stresses as well as tne nold-down
and lift-off forces. Tnese were evaluated for the intact tendons and the
degraded tendons. Tne degree of tendon degradation is described tnrough
the ratio of tne number of unbroken strands to the original numoer of
strands. Initial time of rupture was varied from the time of initial
prestressing to 400 days after emplacement. Tne formulation led to an
integral equation, which was solved numerically. The hold-down forces
decayed approximately with tne logarithm of time and for both the extreme
observed degradation (21 broken strands) and for a more extreme case (40
broken strands) , the hold-down force still exceeded the minimum safety
design requirements.
In addition, several finite element calculations, using the finite
element code hDNSAP-C, were made to evaluate complete tendon failure in a
60° sector of the Fort St. Vrain PCRV. This code has an extensive
material library of constitutive relations to model the various properties
of concrete, together with a specialized element model to simulate pre-
stressing tendons. Two rows of vertical and an arc row of circumferential
tendons were incorporated in the model as a baseline calculation. Tne
tendons were prestressed to 705, of the ultimate and an internal pressure
of 775 psi was applied (this pressure is tne internal pressure of tne
nelium coolant in the HTGR) and the creep of the concrete and slow decay
of tne tenuon stresses were evaluated out to 30,000 days. Then, three
- 22 -
cases wherein one tendon was removed at one day were evaluated. First
the middle vertical tendon in the outer row and in line with the outer
buttress was removed. Second, an inner vertical tendon opposite the
thinnest portion of the PCRV wall was removed. Finally, an inner layer
circumferential tendon at micheight was removed. Stress redistributions
at 300 days after ruptures were calculated and snifts of the remaining
tendon loads to accommodate the broken tendon were calculated. Regions
of local tensile and snear stress in the concrete portion of tne PCRV
were identified and related to overall structural integrity.
With all tendons present, the mean vertical stress was about -7b0
psi , the radial stress decreased from the applied internal pressure of
-705 to about -1200 psi at the ring of circumferential tendons and the
tangential stress ranged from -2400 psi at tne inner wall to about -2200
psi at the same place. Removal of a vertical tendon reduced the mean
axial stress by about +40 psi, the local tangential stress by -10 psi and
did not materially affect the radial stress. Removal of a circumferential
tendon reduced the mean tangential stress by +30 psi and the local axial
stress by -80 psi. The vertical hold-down force from zero days tnrough
30,000 days decreased linearly and remained above the prescribed safety
limit, as did the circumferential hold-down force.
Comparison of the analytical solution and a small finite element
problem simulating the analytical proolem was made to verify the visco-
elastic creep models and the tendon element in the NONSAP-C code. Excel-
lent agreement for stresses, strains and nold-down forces was ootained.
5.0 Conclusions
Los Alamos concludes that tne licensee, Public Service Co. of
Colorado, has made a conscientious effort to address all of the restart
issues listed in the assessment report.1 The refurbishment program on
all CRLOAs provides confidence in CRDOA operability during reactor opera-
tion and the ability to scram, even if the exact "failure to scram" mecn-
anism has not been defined. Questions concerning the reliability of the
back-EMF testing procedure on the shim motor/brake assembly in determining
control rod operational acceptability still exist, but further method
development, more experience with result interpretation, and in-core
- 23 -
testing may alleviate the questions. Until CRDOA operability can defin-
itely be ascertained with these methods, we recommend that the licensee
have backup measures such as rod run-in following scram.
Control rod caole and connecting hardware material replacement, along
with replacement of tne Reserve Shutdown System material, serve to rectify
the material problems brought on by corrosive mechanisms. '
In light of chloride stress corrosion problems, Los Alamos also
recommends that all reactor components exposed to the primary coolant be
reviewed for susceptibility to chloride attack, especially the PCRV liner.
Review should continue into the source of cnlorine and methods to elimi-
nate its generation and presence.
The effects of purge flow loss have not been determined to be in-
strumental in CRDOA failures to scram, yet the licensee has committed to
maintaining purge flow by external means, and to reducing the effects of
primary coolant naturally convecting into the CRDOA cavity with extra
seal installation.
Even though current qualified CRDOA operating temperatures are very
much in question, the licensee is in the process of requalifing the mech-
anism for temperatures more in line with those anticipated during reactor
operation.
From a mechanical standpoint, CRDOA preventive/predictive mainte-
nance procedures are certainly reasonable, but like the proposed surveil-
lance program, they are dependent on back-EMF testing methods wnicn are
still in the developmental stages.
Evaluation of moisture ingress corrective measures was difficult due
to the lacx of information with which to understand the measures taken.
The licensee has committed to submit a more explanatory version of the
actions to mitigate moisture ingress prior to restart.
Tne extent of PCHV tendon degradation is not well known, even if tne
licensee may nave determined the cause of the corrosion. Further investi-
gation into arresting measures is definitely required, especially because
the nitrogen blanket technique may be so difficult to employ. However,
the interim surveillance program should provide information on the degree
and rate of corrosion, in addition to establishing a tendon wire loss
acceptance criteria. The tendon acceptance criteria should ensure PCHV
margins to safety.
- 24 -
b.0 References
1 . "Preliminary Report Related to the Restart and Continue° Operation
of Fort St. Vrain Nuclear Generating Station," Docket No. 50-267,
Public Service Co. of Colorado, October, 1964.
2. "Review of Dallas Meeting (1/15/85) and Restart Committments",
letter from Martin, NRC/Reg IV, to Lee, PSC, 1/17/85.
3. "Fort St. Vrain Meeting, NRC-PSC, February 20, 1985," Volumes I, II
and III, recorded and transcribed by Koenig & Patterson, Inc.
4. "Fort St. Vrain Meeting, NHC-PSC, February 21, 1985," Volumes I and
II, recorded and transcribed by Koenig & Patterson, Inc.
5. "Fort St. Vrain Meeting, NRC-PSC, February 22, 1985," Volumes I and
II, recorder and transcribed by Koenig & Patterson, Inc.
6. "Engineering Report on CRDOA Failures to Scram-Control Rod Drive and
Orifice Assemblies," PSC submittal P-85037, 1/31/85.
7. "Failure of Three CRDOAs to SCRAM," PSC submittal P-85029, 1/28/85.
8. "bearing Deoris Analysis," PSC submittal P-85017, 1/18/85.
9. "Fort St. Vrain Nuclear Generating Station, Updated Final Safety
Analysis heport," Public Service Co. of Colorado.
10. "CRDOA Refurbishment Program Report," PSC submittal P-85040-2,
1/31/85.
11. "Control Rod Drive Cable Replacement," PSC submittal P-85032-2,
1/20/85.
12. "Control Roo Drive Cable Replacement Report," GA Technologies
Document 907622, Attachment 1 to PSC submittal P-85032-2, 1/31/85.
13. "Investigations into Sources of Chloride in FSV Primary Circuit,"
PSC submittal P-66036, 1/31/85.
14. "Safety Analysis Report--Change in Material of the FSC Control Rod
and Orifice Assemblies," Attachment 2 to PSC submittal P-85032-2,
1/31/85.
15. "FSV Control Rod Cable Metallurgical Examinations," draft report
from Los Alamos National Laboratory, 3/85.
16. "Report on Reserve Shutdown Absorber Material," PSC submittal
P-85027, 1/28/85.
17. "Moisture Control in CRDOA Purge Lines," PSC submittal P-85032-9,
1/20/85.
- 25 -
18. "FSV Reserve Snutdown System Material Metallurgical Examinations, "
draft report from Los Alamos National Laooratory, 3/05.
19. "CRDOA Moisture/Purge Flow," PSC submittal P-85032-6, 1/20/85.
20. "Modifications to CNDUA Helium Purge Supply," PSC suomittal
P-85032-8, 1/20/85.
21. "Control Rod Drive Cavity Seals," PSC submittal P-85032-7, 1/20/85.
22. "FSV CRD Cavity Seals Design Report," GA Tecnnologies Document
907604, Attachment 1 to PSC submittal P-05032-7, 1/20/85.
23. "Operations Order No. 84-17 Describing Operator Actions Upon a Loss
of Purge Flow and or Detection of Hign Moisture Levels in Primary
Coolant," PSC submittal P-85040-8, 1/31/85.
24. "CND Temperature and Helium Purge Flow Recorders," PSC submittal
P-85032-3, 1/20/65.
25. "Current CRD Temperature Data Collection Procedure Wnich Requires
Station Manager Notification Upon Discovery of a Measured CRD
Temperature in Excess of 25u°F," PSC submittal P-85040-9, 1/31/85.
2b. "Control hod System Operaoility Evaluation Report," PSC suomittal
P-85040-1, 1/31/85.
27. "CNDUA Mechanism Temperatures Environmental Requalification," PsC
submittal P-85032-1, 1/20/85.
2d. Letter from Robert A. Clark, Cnief, ORu3, to 0. R. Lee, PSCo. ,
December 2, 1982.
29. Meier, K. , "Fort St. Vrain Reactor Control Rod Drive Mechanism Over-
Temperature Problem," Los Alamos National Laboratory, 1982.
30. "CRDOA Proposed Preventive/Predictive Maintenance Program Report, "
PSC submittal P-85040-3, 1/31/05.
31. "ChDOA Interim Surveillance Program Report," PSC submittal P-85040-5,
1/31/85.
32. "FSV Improvement Committee Actions," PSC submittal P-85022, 1/24/85.
33. "Tendon Accessibility Report," PSCo. letter from Warembourg, PSC, to
Jonnson, NRC/Reg IV, PSC submittal P-84523, 12/14/84.
34. "Lab Report No. 52--Examination of Failed Wires from Fort St. brain
Unit No. i," PSC submittal P-o4543-4, 1/24/85.
35. "Liftoff Tests," Attachment 1 to "Engineering Report on Fort St.
Vrain Tendons," PSC submittal P-84543, 12/31/84.
- 26 -
`0,,II,EGV Attachment 3
• . 9! UNITED STATES
W • NUCLEAR REGULATORY COMMISSION
�� '1 WASHINGTON,D.C.20555
Y�
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
RELATED TO CONIROL ROD POSITION INSTRUMENTATION
TO FACILI1Y OPERATING LICENSE NO. DPR-34
PUBLIC SERVICE COMPANY OF COLORADO
FORT ST. VRAIN
DOCKET NO. 50-267
1. INTRODUCTION
This Safety Evaluation concerns control rod position instrumentation
(RPI) failures that were recognized subsequent to the partial ATWS event
at Fort St. Vrain on June 23, 1984. The NRC issued an Assessment Report
[Reference 1], which addressed the ATWS event and the RPI failures, and
included recommended actions to be taken at Fort St. Vrain before the plant
is restarted. The objective of this evaluation is to review the licensee's
response regarding the following issues: the causes of the RPI failures,
the methods to prevent further failures, operability and surveillance
actions for the RPIs, and the adequacy of the licensee's proposed backup
method for verification of rod full-in position.
The submittal by the licensee in response to the Assessment Report takes the
form of a series of reports dated about January 31, 1985 [References 8-15].
Numerous additional documents were reviewed, including background information
about control rod drive design and operation, supplied by General Atomic
Company, NRC reports, and Public Service Company of Colorado (PSC) related
submittals [1-7, 16-18]. Evaluations and conclusions are provided based on
the licensee's response to questions and the information contained in the
licensee's submittals.
2. BACKGROUND
On June 23, 1984, the Fort St. Vrain plant experienced a partial ATWS event,
in that 6 of 37 control rod pairs failed to insert into the reactor core in
response to an actual scram signal . Subsequently, on July 31, 1984, PSC
reported 11 RPI anomalies to the NRC. The anomalies included improper
analog and digital indications of rod position, faulty slack cable,
erroneous rod full-in and full-out limit indications.
During August 1984, we visited the plant to investigate the RPI failures
and related problems. We met with the licensee again at the plant site on
November 28, 1984, at which time the licensee indicated the general steps
that were being taken regarding the instrumentation problems. On
January 31, 1985, the licensee provided documentation of its evaluations
and efforts relating to restart. These documents were reviewed by the
staff and its consultant, Franklin Research Center. Questions arising
from the review were addressed by the licensee during a transcribed meeting
at the plant site on February 22, 1985.
- 2 -
In order to establish the scope of this review, we have revisited the
Assessment Report, the Region IV letter of January 16, 1985 to PSC, and
the PSC response of January 28, 1985 [1, 13, 17]. The following restart
items are within the scope of this review.
o Determination of RPI Failure Causes and Corrective Action
o Procedures for Prevention of Driving Control Rods Beyond
the Full-in Limit
o RPI Limiting Conditions for Operation
o Position Instrumentation Surveillance Program
During this evaluation, we determined that revisions of the plant procedures
are necessary to assure that appropriate action is taken upon indication of
mispositioned rods following a scram. Additionally, during the February 22, 1985
meeting with the licensee, the Wattmeter test was reclassified as a restart issue.
3. EVALUATION
The evaluation is divided into the following sections:
Page
3.1 Rod Position Instrumentation (RPI) Failures and Corrective Actions 2
3.2 Prevention of Inward Overtravel of Rods 4
3.3 RPI Minimum Performance Requirements 5
3.4 RPI Surveillance Requirements
3.5 Backup Reactor Shutdown Procedure 9
3.6 Backup Full-in Position Verification Test 9
(Wattmeter)
Each section includes the conclusions regarding that topic.
3.1 ROD POSITION INSTRUMENTATION (RPI) FAILURES AND CORRECTIVE ACTIONS
Subsequent to the reactor trip of June 23, 1984, the licensee identified 11
control rod position instrumentation (RPI) failures. The failures included
simultaneous rod full-in rod and rod full-out indications, full-out switch
lights remaining lit, indications of partial rod withdrawal , no position
signals, disparity between analog and digital rod position information, and
a slack-cable indication. The failures caused conditions in which the
operators could not determine the position of certain rods based upon the
existing indicators. As described in the October 16, 1984 Assessment Report
[1], for one rod pair the analog and digital RPIs indicated it to be at the
40-inch withdrawn position. Simultaneously, the full-in position indicator
for this rod was not operable. When asked if the instrumentation should be
believed, PSC personnel responded that it did not believe the installed RPIs,
but had verified to its satisfaction that the control rod was fully inserted.
- 3 -
The results of the licensee's investigation into the causes of these failures
and his proposed corrective actions are included in Attachment 4 of PSC
letter P-85032 [12], dated January 30, 1985. This report suggests that the
failure causes for both the switches and the potentiometers were primarily
mechanical in nature and were not electrically induced. The licensee
confirmed this general conclusion during the February 22, 1985 meeting.
Limit Switch Failures
The limit switches that indicate full-in or full-out positions are roller
plunger-type microswitches. The licensee stated that the 30° slope of the
cams that operate the limit switches is too abrupt, causing a high lateral
force between the plunger and the shaft tube that eventually results in
switch failure. This failure cause may have been compounded by increased
friction due to pitting of the plunger and shaft tube from the effects of
moisture.
Our evaluation of the licensee's description of the failure cause indicates
that it should be expected, given the design of the limit switch. Further,
the failure mechanism is consistent with our previous experience with such
switches. We conclude that the failure cause identified by the licensee
is credible.
The licensee proposes to change the slope of the limit switch cams to
15° to reduce the lateral force. However, these cams are not expected
to be installed at the time of return to service.
During the present refurbishment of the CRDs, the operation of the switches
will be checked and those switches which are not operational will be replaced
in kind. The level of deterioration of the operational switches will not
be evaluated. Therefore, potentially degraded but operational switches
will be returned to service.
Potentiometer Failures
There are two separate failure mechanisms associated with the analog and
digital position potentiometers: one relates to internal damage of the
potentiometers and one relates to external damage of the associated coupling.
The internal failure mechanism for the potentiometers is travel beyond the
10-1/2-turn capability. The greatest potential for such overtravel is at
the full-in limit, especially if the full-in limit switches have failed.
Review of the CRD design indicates that damage may result from an inward
overtravel of as little as six inches. There is no mechanical stop that
interrupts CRD travel before potentiometer failure might occur. The analog
and digital potentiometers share a common drive shaft so that simultaneous
failure may occur if both potentiometers exceed their overtravel limits.
The licensee proposes to replace the 10-1/2-turn potentiometers with 15-turn
potentiometers so that overtravel of the drive will not cause internal
damage to the potentiometers. However, the 15-turn potentiometers are not
expected to be available for installation at the time of the scheduled plant
restart.
- 4 -
The external failure mechanism associated with the potentiometer is caused or mechanical
interference
couplingconbthecommon driveen the cams shaft the liformit switches
the potentiometers.
During an inward overtravel , one of the limit switch cams can strike the
coupe nt ce has wn that,thereby breaking
occurs, coupling.ithispos a
possible fortheeanalognindicatoro if
to agree
with the digital indicator, and for both indicators to track rod motion, but
both indicators can be grossly inaccurate (i.e. , offset) with respect to the actual cations have been licenseeototion of eliminatee rod damageair. No caused by theiinterference betweenotheed by the
cam and
the coupling.
Prior to startup, already-damaged potentiometers will be replaced, as will
damaged multi-jaw couplings that connect the potentiometer shaft to the CRD
gear train. The replacement potentiometers will be 10-1/2-turn devices, not
15-turn devices.
3.2 PREVENTION OF INWARD OVERTRAVEL OF RODS
Since modifications that prevent overtravel or otherwise preclude damage
as a result of overtravel will not be in place at the time of restart,
procedures must be instituted that preclude intentional inward overtravel
of the control rod pairs and reduce the probability of such damage.
In Attachment 7 to PSC letter P-85040 [14], the licensee provided a revision
of page 17 of 20 of Procedure SOP 12-01, that describes actions to be taken
if the "in limit" light is not received when a rod is believed to be fully
inserted. The previous practice had been to drive the rod further inward in
search of limit switch actuation. In the revised procedure, the operator is
directed to first withdraw the rod a few inches, trying to obtain the full-in
position indication. The procedure then directs the operator to move the rod
switches inward until rod motion automatically stops, or an indication of zero
inches is attained, whichever occurs first.
Use of the revised procedure will help assure that rod position indications
remain available for use. However, no direction or suggestion is given to
perform a lamp test or to verify the analog/digital position indication
prior to this special rod movement. It is recommended that these steps
be added to this procedure.
It is also unclear that this particular procedure, SOP 12-01, governs all
conditions in which manual rod inward travel in the vicinity of the full-in
position may be performed. The licensee should modify all applicable
procedures to similarly reduce the likelihood of inward overtravel of the
rods near the full-in limit.
- 5 -
3.3 RPI MINIMUM PERFORMANCE REQUIREMENTS
Control rod full-in limit indications are "important to safety" because
they are the primary means of verifying that all the CRDs have fulfilled
their reactor scram safety function. Immediately following a plant trip,
total reliance on the startup range nuclear instrumentation may be inadequate.
The analog position indicators are "important to safety" because they
provide the operator with continuous position information for all the rods
during operation and following a scram.
In contrast, however, the digital position indicators and the full-out
position indicators are not as important, but provide information that
is useful for accurate position data, fine control of reactor power,
and as operational conveniences. They may be used a backup indicators,
in the event of a loss of the analog or full-in limit indications.
The NRC Assessment Report recommended the Technical Specification Limiting
Conditions for Operation be established to define the minimum performance
requirements for the RPIs.
The licensee has concluded that a specific limiting condition for operation
(LCO) governing rod position instrumentation is not necessary. The basis
of this conclusion is that other LCOs indirectly require instrumentation
to be operable. In Attachment 1 to PSC letter P-85040 [14], the licensee
discussed LCOs 4.1.2, Operable Control Rods; 4.1.4, Partially Inserted Rods;
and 4.1.8, Reactivity Status that require surveillance tests that in turn
require the use of position indicators. The licensee's reasoning is that
lack of position indication will preclude performance of a surveillance
test and therefore an LCO will not be met.
The use of indirect LCOs as a requirement for position indicator operability
does not focus attention on the importance of the RPIs, does not provide
definitive action statements for partial losses of instrumentation, and
does not ensure timely resolution of indication failures. In addition,
the use of indirect LCOs is not consistent with the Technical Specifications
for other operating plants. In summary, we do not agree with the licensee's
conclusions.
In light of the possibility of an ATWS event, coupled with the propensity
of the position indications to fail , the following concepts related to
RPIs must be implemented via procedures prior to restart and via a new
Technical Specification LCO soon after restart:
At startup and during operation, the analog rod position indicator
and full-in rod limit indicator for each control rod pair shall
be operable.
- 6 -
If an analog indicator is inoperable, operation may continue provided
that one of the following conditions is met: (1) when the rod is fully
inserted, the full-in limit indication is operable or the full-in
position has been established by an independent means of verification;
(2) when the rod is in a mid-position, rod position is continuously
indicated by an operable digital position indicator and the full-in limit
indtheifull out and rthe efull-in; or )position when elimit rod iindicators s in the uare operable.11-out on,
If the full-in indicator is inoperable, operation may continue provided
that one of the following conditions is met: (1) when the rod is fully
inserted, the full-in position has been established by an independent
means of verification; or (2) when the rod is in a mid or full-out
position, both the digital and analog indicators are operable and are
known to be accurate at the full-in position, and a digital indicator is
continuously indicating the rod's position.
In addition to the new LCO described above, a modification to LCO 3.1.2,
Operable Control Rods, is necessary. This modification must incorporate
the RPI concept that rod all be red newpLCOble areif notts associated
met.
Without such operability requirements, the operators may tend to lose faith
in the position indications to the extent that actual failure of the rods
to attain the full-in position may be inappropriately considered to be a
faulty indication rather than a valid condition. The LCOs described above
are necessary whether or not the proposed modifications are made to the
indicator mechanisms. However, the LCO is particularly important for the
present restart since the modifications that will reduce the failure rate of
the indicators will not be in place. With the required LCO in effect at the
time of restart, restraints will be placed on reactor operation to allow safe
operation should position instrumentation failure occur.
3.4 RPI SURVEILLANCE REQUIREMENTS
The NRC Assessment Report recommends that the licensee establish Technical
Specification Surveillance Requirements regarding the RPIs. In Attachment 5
of PSC letter P-85040 [14], the licensee outlines the proposed interim
surveillance program for the control rod drives. This program includes
partial verification of rod position indication operability at weekly and
quarterly intervals during operation and during each refueling outage.
For fully withdrawn rods, the proposed weekly surveillance verifies the
oerabiit s and betweenithe analog and ity of the ldigital-out mindications rto be no r mores the difference
than 10 inches.
- 7
However, the program outline does not indicate that the analog and digital
readings must correlate to the full-out or full-in position, nor does it
state that the indicators must correctly track the rod position when the
rod is exercised (i.e., partially inserted and then withdrawn).
For partially inserted rods, the weekly surveillance will verify that the
analog and digital position indications are within 10 inches of each other.
Again, the proposal does not state that the indicators must appropriately
track rod motion.
For fully inserted rods, there is no proposed weekly surveillance, because
other Technical Specification requirements apparently preclude movement
of fully inserted rods during power operation.
The proposed quarterly surveillance adds one further test to the weekly
surveillance. For rods in the full-out position, the operability of each
of the redundant full-out limit switches will be verified. No additional
quarterly surveillance is proposed for the RPIs associated with partially
inserted and full-in rods.
The proposed refueling outage surveillance includes verification of the
operability of each redundant full-out and full-in limit switch and
comparison of the analog and digital position indications during a full
travel scram of the analog and digital indications during a full travel
scram of the rod. While not directly stated, it is assumed that the
accuracy of the analog and digital position indicators will also be
confirmed at the full-out and full-in positions during these tests.
The RPI surveillance actions proposed by the licensee are worthwhile and
appropriate. However, the proposed surveillance program does not encompass
all of the necessary aspects of RPI operability verification. To be
consistent with the position established in Section 3.3, that the full-in
limits and the analog indicators are "important to safety," appropriate
surveillance actions are needed for these devices. Further, if the digital
indicators are to be used as backup indications, some surveillance actions
for these devices are also appropriate.
The surveillance tests proposed by the licensee do not address the need to
verify operability of the full-in limit switches prior to outward movement
from the full-in position. Nor do they verify the accuracy of the digital
and analog position indicators at the full-in position prior to the first
outward rod motion. Also, the proposed surveillance program description
does not provide sufficient detail to determine if the operability of analog
and digital position indicators will be evaluated for partially and fully
withdrawn rods.
- 8 -
Since the failure mechanisms identified by the licensee for the full-in
switches and potentiometers are associated with overtravel of a rod near
the full-in limit, it is appropriate to verify that these position indicators
are operable before or at the next outward movement of the rod from the
full-in position. This suggests the need for verification test of these
RPIs at each reactor start-up. Verification of the operability of the
full-in limit switches and the accuracy of the analog RPIs may only be
performed when the rods are full-in, as they are at or prior to start-up.
Furthermore, simple verification checks at this time would have minimal
adverse impact on plant operations.
The following concepts should be incorporated into the surveillance program
prior to reactor restart. The surveillance requirements that result from
these concepts are in addition to the surveillance actions proposed by the
licensee and are necessary to assure that instrumentation is operable and
reasonably accurate at reactor start-up, during operation, and at the time
of a scram.
Prior to each reactor start-up, each full-in limit indicator should be
verified as operating when the rod is full-in and operable by virtue of
the change in the indication when the rod is withdrawn a short distance.
Alternatverified t
hefirst time dur rod full-in ingioron afteristart-uppthatiairodmis withdrawn be
from the full-in position.
Prior to each reactor start-up, or during the first outward motion of a
rod, the analog position indicator should be shown to be acceptably accurate
at the full-in position and be shown to respond appropriately when the rod
is withdrawn a short distance. The accuracy requirement should be such that
returning the rod to an "0" position as indicated by the analog RPI will not
result in overtravel that could cause damage to the potentiometers and the
associated coupling. The digital indication should likewise be shown to
be operable and accurate at the full-in position.
During each weekly surveillance during power operation, the reasonableness
of the analog RPI should be verified by comparing the change in analog indi-
cation with the direction and time duration of the rod travel . The analog
and digital position indications should agree with each other within a
predetermined amount. If a larger difference is observed, the licensee
should, conservatively, assume that the analog indicator is the inoperable
channel , unless it can be proven to be accurate and operable by another means.
The combination of the surveillance actions proposed by the licensee and
the additional surveillance actions stated herein will provide the
necessary assurance that the RPIs are operable and accurate.
r
- 9 -
3.5 BACKUP REACTOR SHUTDOWN PROCEDURE
At the time of a reactor scram, it is essential that the reactor operator
have confidence in the control rod position indications and take conservative
actions based upon those indications. Therefore, when a reactor scram
occurs, the reactor operator must be able to verify that the rods are
full-in or must take appropriate alternate shutdown actions.
In view of our conclusions regarding the corrective actions to prevent
future RPI failures, we have determined that additional action is necessary.
Plant procedures should specify that, if after a reasonable but conservative
period (e.g. , one hour), more than one rod pair cannot be verified as being
at the full-in position, these rods must be assumed not to be fully inserted
and the backup shutdown system is to be initiated. Such actions are
necessary to assure adequate shutdown. Furthermore, immediately following a
plant trip, total reliance on the startup nuclear channels may not be
adequate. Therefore, the Fort St. Vrain procedure must be revised, as
necessary, to include the following concepts:
Following each reactor scram, each rod pair shall be
verified to be at the full-in position by one of the
following means:
1. the agreement of the analog position indication
and the full-in position indications; or
2. the agreement of the analog and digital position
indications (that were known both to be operable
prior to the scram and to be accurate at the full-in
position); or
3. the use of an independent rod position verification
method (e.g. , Wattmeter test).
If more than one control rod pair cannot be verified to be
fully inserted at the end of one hour, the backup
reactivity control system must be initiated. Rods that
were known to be fully inserted into the reactor prior
to the scram may be excluded from the above considerations.
3.6 BACKUP FULL-IN POSITION VERIFICATION TEST
The NRC Assessment Report of October 16, 1984 [1], concluded that the
installed RPI system may be inadequate to determine rod positions under
adverse conditions. The licensee has been using the Wattmeter test as an
independent verification of full-in positions. However, as a result of
our review, we determined the test was too judgmental to provide convincing
verification. Therefore, the licensee was requested to refine its means
of verifying that the control rods are fully inserted.
- 10 -
The licensee's Wattmeter tests analyze the electrical power required
by the motor when the control rod pair is moved a short distance in the
vicinityletter
fthe llinposition
.
[hese tests were described in AtThe electrical power requirements at the full-in position differ from
other positions due to the
unwinding
the last 10 inches, th
cable
the
e point of
rum. As
the cable unwinds for approximately
connection of the cable to the drum changes from the horizontal
to a,vertical s(6 i or'clock)re the positionwherent rthen the momentrum armis isgzero.
Because of this change in moment, the power requirements for the
motor in both the inward and outward directions near the full-in
position are different from other positions of the rod. On an outward
pull from the full-in position, the transient power peak is lower and
of shorter duration than on outward pulls from other positions, and
the power peak is followed by a dip in power below the steady-state
level , which also does not occur at other positions. On an inward
position isereached.full-inp This ositirise tdoes he onot occur atwer rises iother positions ghtly as the of
the rod.
The licensee has sufficiently described and demonstrated the phenomena
related to the electrical power changes. The uniqueness of these
phenomena provides an acceptable concept for verifying the full-in
rod position.
The key areas of concern continue to be the difficulty of interpretation
of the wattmeter charts and the degree to which judgment is required in
order to properly interpret the results. The acceptance criteria
proposed by the licensee require interpretation of graphical results
at or beyond the levels of precision and resolution of good engineering
practice. The wattmeter used is a 1000-watt meter coupled to a stripchart
recorder. The smallest division on the chart paper is equal to 20 watts.
The procedure requires interpretations of differences of 4-6 watts.
The procedure also requires interpretation of the decay time of the motor
starting transient. The non-standard definitions of decay time stated in
the licensee's procedure could allow greatly differing interpretations
of the same transient. Under the definitions, one interpretation of the
decay time may be as much as twice that of another interpretation for the
electrical transient.
The licensee proposes to use a wattmeter that has a 1000-watt scale and
uses paper with a 500-division scale. The written procedure alternately
notsewatts and ally surescale whethervwattssorlmost scaleidivisionseisly, such c that one is
correct.
- 11 -
4.0 SUMMARY
We have evaluated information provided by the licensee regarding: (1)
control rod position instrumentation (RPI) failure mechanisms and corrective
actions; (2) procedures to prevent inward overtravel ; (3) Technical
Specifications relating to RPI operability; (4) RPI surveillance program,
(5) backup reactor shutdown procedures; and (6) the backup full-in position
verification (wattmeter) test. The conclusions relating to these concerns
are summarized below.
The licensee has determined that the failure of the limit switches is caused
by the steep slope of the limit switch cams, and that two failure causes are
associated with the potentiometers. In the first, the potentiometer is forced
beyond its 10-1/2 turn overtravel limit and is damaged during inward overtravel .
In the second, the shaft coupling for the potentiometers is struck by a limit
switch cam during inward overtravel of the rod. The licensee has proposed a
modification to the limit switch cam to eliminate the limit switch failure
mechanism and has proposed to use a 15-turn potentiometer to eliminate the
forcing of the potentiometer beyond its internal travel limit. However, these
modifications will not be implemented prior to the scheduled restart, and no
modification has been proposed to address the second cause of the potentiometer
failures.
The licensee has modified Procedure SOP 12-1 to preclude manual inward
overtravel of rods following a reactor scram. It is not clear that this
procedure covers all conditions in which manual inward motion of a control
rod could occur. Therefore, the licensee should confirm that all procedures
in which rod travel beyond the full-in limit could occur contain appropriate
cautions and controls to reduce the likelihood of damage to potentiometers
and their couplings.
Contrary to the recommendation in the NRC Assessment Report, the licensee
believes that no additional limiting conditions for operation (LCO) for
RPIs are necessary and proposed to rely upon indirect LCOs. The use of
indirect LCOs does not provide definitive actions to be taken upon the loss
of RPIs and is inconsistent with the Technical Specifications of other
reactors. Therefore, the LCOs described in Section 3.3 for RPI operability
are necessary.
The licensee has proposed surveillance actions for the RPIs for use during
reactor operation and during refueling outages. These surveillance actions
are appropriate and worthwhile. However, they do not cover verification
of operability of the RPIs at each startup of the reactor. The failure
mechanisms for all RPIs (except the full-out limit RPI) are associated with
travel near the full-in limit. Therefore, it is appropriate to verify
operability of these RPIs at the next outward motion of the rod (i.e. , at
startup). The combination of the licensee's proposed surveillance program
and the additional surveillance stated in Section 3.4 combine to provide
adequate assurance of RPI operability.
- 12 -
When a scram occurs, the operator must believe the rod position indicators.
If multiple control rod pairs are not indicated as being full-in, the
operator must take timely action to initiate the backup reactor shutdown
system. The modifications to the operating procedures described in
Section 3.5 will assure this action.
The wattmeter test proposed by the licensee is an adequate method of
verifying the rod full-in position provided that the ease of interpretation of the he inter-
polation decreased sthroughttheluseloffa reliance
moreappropriate ter-
choice of
wattmeter range and recorder speed.
Date: May 17, 1985
Principal Contributor:
J. T. Beard, DL
REFERENCES
1. Letter from H. R. Denton, NRC, to R. F. Walker, PSC. Preliminary
Report Related to Restart and Continued Operation of Fort St. Vrain
Nuclear Generating Station Docket No. 50-267, Public Service of
Colorado, October 16, 1984.
2. PSC Presentation and Handout Notes from Meeting at Fort St. Vrain,
November 28, 1984.
3. Letter from 0. R. Lee, PSC, to E. H. Johnson, NRC Region IV, Attachments 1
and 2, Observations on Reworked Control Rod Mechanisms, August 28, 1984.
4. PSC History of Observations on 44 CRDOA, Fuel Handling Procedure Work
Packet, FHPWP, August 2, 1984.
5. Memorandum from J. T. Beard, NRC, to J. Miller, NRC. Subject:
Executive Summary, FSV Assessment Report, September 11, 1984.
6. Rod Control System Equipment I-9303, Operation and Maintenance Manual
93-I-1-327, E-115-265, (Rev. 3), Prepared for PSC, Unit 1, Fort St. Vrain,
Colorado, Gulf General Atomic, 1979.
7. Installation, Operation, and Maintenance Manual for the Control and
Orificing Assembly for the Fort St. Vrain Reactor 12-D-4-63, FCN-3849,
GA-9806 (Rev.) Prepared for PSC of Colorado Fort St. Vrain Nuclear
Generating Station, General Atomic, May 1977.
8. Letter from 0. R. Lee, PSC, to E. H. Johnston, NRC Region IV. Subject:
Comments on Index No. 3 (January 15, 1985 Meeting Minutes) , January 28,
1985.
9. Letter from H. L. Brey, PSC, to E. H. Johnson, NRC Region IV. Subject:
Description of CRDM Failures on January 14 and 16, 1985 (Work Reports) ,
January 28, 1985, P-85029.
10. Letter from 0. R. Lee, PSC, to E. H. Johnson, NRC Region IV. Subject:
Creation of FSV Improvement Committee and Items of Consideration,
January 24, 1985, P-85022.
11. Letter from D. W. Warembourg, PSC, to E. H. Johnson, NRC Region IV.
Subject: Results of CRDOA Debris Analysis Test Report Attached,
January 18, 1985, P-85017.
- 2 -
12. Letter from 0. R. Lee, PSC, to E. H. Johnson, NRC Region IV. Subject:
Transmittal of Technical Reports, January 30, 1985, P-85032.
Attachment 1: CRODA Mechanism Temperatures Environmental Requalification
Attachment 4: Control Rod Instrumentation
Attachment 5: Cable Anchor Welding
Attachment 6: CRDOA Moisture/Purge Flow
13. Letter from R. D. Martin, NRC Region IV to 0. R. Lee, PSC. Subject:
Minutes of Meeting of NRC Region IV and PSC on January 15, 1985.
Commitments by PSC on Plant Upgrades, January 16, 1985.
14. Letter from PSC (author unknown) to R. Martin, NRC Revion IV.
Subject: (See Attachments), January 31, 1985, P-85040.
Attachment 1: Control Rod System Operability Evaluation Report
Attachment 5: Control Rod Drive and Orificing Assembly Interim
Surveillance Program
Attachment 6: Wattmeter Use to Determine Inserted Absorber String
Position
Attachment 7: Page 17 of SOP 12-01, Discussion of No "In Limit" Light
on Fully Inserted Rod.
15. Letter from D. W. Warembourg, PSC, to E. H. Johnson, NRC Region IV.
Subject: Cover Letter for Attachment, January 31, 1985, P-85937.
Attachment: PSC Report No. EE-12-0010, 'Failures to Scram - Control
Rod Drive and Orifice Assemblies."
16. Letter from D. W. Warembourg, PSC, to E. H. Johnson, NRC Region IV.
Subject: High-Pressure Scram and Ensuring Control Rod Automatic
Insertion Failures P-84227, July 23, 1984.
17. Letter from 0. R. Lee, PSC, to E. H. Johnson, NRC Region IV, January 28,
1985. Subject: Response to Reference 13.
18. NRC Memorandum from J. T. Beard, ORAB, to E. Johnson, Region IV. Subject:
Fort St. Vrain -- Design Weakness in Reactor Scram System, March 4, 1985.
19. Letter from J. R. Buchanan, Oak Ridge National Laboratory, to
Frederick J. Hebdon, NRC, February 14, 1985.
Hello