HomeMy WebLinkAbout841146.tiff " UNITED SEAT' S
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NUCLEAR REGULATORY COMMISSION
REGION IV
^.JI 611 RYAN PLAZA DRIVE, SUITE 1D00
'1-,) „° ARLINGTON. TEXAS 76011
JUN 2 2 19S4
Docket: 50-267 Err f'""T't
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JUN 251984 ?
Mr. 0. R. Lee, Vice President
Electric Production GREELgy
Public Service Company of Colorado ' cQLo.
P.O. Box 840
Denver, Colorado 80201
Dear Mr. Lee:
We have completed our review of the information you provided, concerning steam
generator tube inspections at Fort St. Vrain (FSV) , by letter dated January 20,
1984. The results of our review are contained in the attached Safety
Evaluation (Enclosure 1). We have concluded that there is no practical
nondestructive examination method for examining the steam generator tubes, and
that primary coolant moisture monitoring and secondary coolant radiation
monitoring provide adequate indication of a leak so that corrective action can
be initiated. We do not agree, however, that the previous two tube leaks were
necessarily random in nature. While we do not propose the implementation of
scheduled or unscheduled inservice inspections of the steam generator tubes, we
request that your proposed post leak examination program, as committed to in
your January 20, 1984 submittal , be included in the plant' s Technical
Specifications (TS).
In addition, to including the examination program in the Technical
Specifications, we request that your application include provisions to
establish, implement, and maintain a secondary coolant chemistry program similar
to that shown in Enclosure 2. (Other program type requirements have been
requested for fuel element surveillance and post-accident sampling by earlier
correspondence). We further request that the application be submitted within
60 days of your receipt of this letter.
If you have any questions on our Safety Evaluation or on the content of the
requested TS, please contact the NRC Project Manager -
P. Wagner at (817) 860-8127.
841146
Public Service Company -2-
of Colorado
Since this request relates solely to FSV, OMB clearance is not required under
P. L. 96-511.
Sincerely,
Eric H. Johnson, Chief
Reactor Project Branch 1
Enclosures:
1. Safety Evaluation
2. Model TS
cc:
C. K. Millen Chairman, Board of County Commissioners
Senior Vice President of Weld County, Colorado
Public Service Company Greeley, Colorado 80631
of Colorado
P. 0. Box 840 Regional Representative
Denver, Colorado 80201 Radiation Programs
Environmental Protection Agency
David Alberstein, 14/159A 1860 Lincoln Street
GA Technologies, Inc. Denver, Colorado 80203
P. 0. Box 85608
San Diego, CA 92138 Don Warembourg
Nuclear Production Manager
Public Service Company of Colorado
P. 0. Box 368
J. K. Fuller, Vice President Platteville, Colorado 80651
Public Service Company
of Colorado Albert J. Hazle, Director
P. 0. Box 840 Radiation Control Division
Denver, Colorado 80201 Department of Health
4210 East 11th Avenue
Denver, Colorado 80220
G. L. Plumlee
NRC Senior Resident Inspector Kelly, Stansfield & O'Donnell
P. 0. Box 640 Public Service Company Building
Platteville, Colorado 80651 Room 900
550 15th Street
Denver, Colorado 80202
Darrell G. Eisenhut, Director
Division of Licensing
Office of Nuclear Reactor Regulation
U. S. Nuclear Regulatory Commission
Washington, D. C. 20555
3,F RE(;0(1 O
UNITED STATES
8 NUCLEAR REGULATORY COMMISSION Enclosure 1
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l WASHINGTON, D. C.20555
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SAFETY EVALUATION
OF PUBLIC SERVICE COMPANY OF COLORADO REPORT
SUPERHEATER TUBE LEAKS IN THE STEAM GENERATORS
OF THE FORT ST. VRA1N HIGH TEMPERATURE
GAS COOLED REACTOR" JANUARY 1984
Background
On December 8, 1982, a secondary side to primary side leak was discovered
in the economizer-evaporator-superheater section of the B-2-3 module in
the Loop 2 steam generator of the Fort St. Vrain nuclear plant. The leak
was assumed to have developed following a reactor scram transient which
occurred on September 30, 1982.
The leak elevation was located and the leaking tube was identified as
one of three tubes connected to subheader "M" in the affected module.
Based on leak rate results the hole (leak) was on the order of a 0.003
inch diameter orifice. The plugging operation involved the removal of
the 3 tubes connected to subheader "M" from service out of 54 tubes in
the affected module. In the plugging operation, sections from the
feedwater lead in and the steam lead out tubes were removed and both
ends of each tube were capped. A section of the steam generator tube,
alloy 800 grade 1 and a section of the feedwater tube, carbon steel
SA 210 type A-1 which were removed to perform the plugging operation
were sent to General Atomics for laboratory examinations.
- 2 -
In their report, dated January 1983 and entitled "Metallurgical Examina-
tion of Tubes Removed from Fort St. Vrain Steam Generator B-2-3" , General
Atomics presented the results of visual examinations and metallurgical
examinations of the steam generator and feedwater tube sections.
Visual examination of the alloy 800 steam generator tube revealed an
apparent thin oxide film on the exterior (gas side) and a thin coating
on the inside of the tube; there was no evidence of corrosive attack.
The feedwater tube section had uniform corrosion, as anticipated, with
no evidence of anomalous degradation.
Metallurgical evaluation included metallographic mounting of specimens
for microstructural examination, microhardness measurements and
energy dispersive analysis (EDAX) for determining the composition of
the oxide or corrosion films.
The oxide film on the alloy 800 steam generator tube consisted primarily
of Fe-Cr-Ni oxide and had an average thickness of 0. 008 inch with no
microscopic evidence of pitting, cracking or erosion/corrosion damage.
The microstructure was fine grained with evidence of cold work, primarily
in bend sections but microhardness measurements did not suggest any extensive
work hardening. At 1000 X magnification, the microstructure was considered
typical for alloy 800 grade 1 and the grain boundaries were observed to be
free of significant carbides precipitation indicating no degree of
sensitization.
- 3 -
The feedwater tube corrosion film was magnetite with thicknesses ranging
from 0. 010 to 0. 040 inch and averaging 0.021 inch (the tube wall thick-
ness is 0. 165 inch minimum). The EDAX analysis of the magnetite indicated
iron as Fe3 04 and some silicon and copper were also detected. The
presence of copper with some oxygen and chlorides in the system suggests
the reason for the thick magnetite growth on the feedwater tube.
Microstructurally, the feedwater tube was ferritic/pearlite and fine
grain, typical for type SA 210 carbon steel . .
GA concluded that the tube sections of both the 800 alloy steam generator
tube and ferritic steel feedwater tube are in good condition although the
thick magnetite film on the feedwater tube suggests that it may be
necessary to chemically clean the tubes in the future. It was recom-
mended that an effort be made to reduce the copper, oxygen and chloride
content in the feedwater to control magnetite growth. Based on these
examinations, the licensee concluded that the actual cause of the leak
could not be determined and it appeared to be random in nature.
The staff didn't concur that the tube leak was a random occurrence and
recommended that in the event further leaks occur some form of NDE be
conducted to assess the extent of damage.
- 4 -
Discussion
In response to the staff' s concern regarding the ability to conduct
non-destructive examinations of the steam generator tubes in the event
of future tube leakage, the licensee submitted the referenced report
and accompanying documentation with their January 20th, 1984 letter.
In the reference report, Public Service Company of.Colorado and GA
Technologies evaluated the two (2) tube leaks in the Fort St. Vrain
steam generators. The first leak occurred •in November 1977 and the
second in December 1982. Both leaks occurred near the bottom of
superheater 2; 1977 in loop 1 and 1982 in loop 2. Both leaks were
found at or near a floating tube support plate at about the same
elevation.
In order to determine the cause of the tube leaks, the licensee con-
sidered all potential factors including residual stresses in the tube
bends, weld joint defects, vibration stresses causing fatigue, water
chemistry, corrosion, wear, cold springing, low cycle fatigue, crack
propagation and loss of tube sleeves and wedges.
The licensee concluded that there is no evidence that any of the
above factors were responsible for tube degradation and leakage. However,
the coincidental locations of the two tube leaks at the support plate
- 5 -
location raise the remote possibility that the sleeve/wedge assemblies at
these support locations were missing or became loose whereby vibrations
due to tranverse flow across the tube bundle could have caused the tube
leaks. Until additional leaks occur at similar locations, the described
degradation mechanism cannot be verified and the tube leakage cause can
therefore be considered unknown.
The licensee concludes that the ability to perform quantitative NDE
on the steam generator tubes would be desirable in order to determine
whether degradation occurred in the steam generator tubes. However,
Fort St. Vrain steam generator tubing is generally inaccessible for
tubing inspection due to lack of physical access to the tubing area
and unit configuration. There currently is no method available for
inspecting steam generator tubes without removing steam generator
modules from the prestressed concrete reactor vessel (PCRV). The tubes
are not accessible from the primary side due to the shroud design which
surrounds the tubes and cannot be inspected internally using current
technology, because of the tube design (helical tube bundles , varying
tube I. D. and 90° turns at the tube to header or subheader junctions).
Although the PCRV was designed with provisions for removal and replace-
ment of steam generator modules, it would be a difficult, costly and
time consuming task. Furthermore, the method has not been demonstrated
- 6 -
nor is equipment available to do the job. Therefore, non-destructive
examination of the steam generator tubes is considered impractical at
this time and cannot be used to verify tube integrity.
The only areas where NDE is practical are not wholly representative of
the tube leak area. These areas are external to the PCRV. The subheader
tubes in these areas are made accessible for NDE in the process of capping
the subheaders containing the leaking tube(s).
Immediately following each future tube leak, the licensee proposes to perform
a metallographic examination of specimens taken from the accessible sub-
header tubes that are connected to the inaccessible tubes which contain
the leak. The results of these examinations will be compared to those
obtained from the specimens taken from the tubes that are connected to
the previous tube leaks. The licensee will also evaluate the size and
elevation of all future tube leaks to determine if additional evidence
or circumstances can help to identify a cause or trend in the degradation
of the tubes of the Fort St. Vrain steam generators.
Conclusions
The staff find that since both tube leaks were similar in magnitude and
located at or near a tube support plate, they may not be random in nature.
However, the staff agrees that there is no practical NDE method for
- 7 -
examining the steam generator tubes due to inaccessibility, helical
configuration, 900 turns and varying tube inside diameters along
the tube length. In view of the fact that the calculated flaw size
in the leaking tube was only 0.003 inch diameter in a 0.205 inch
thick tube wall , we do not believe that tube rupture was imminent
or that structural integrity of the tubes has been imoared. Further-
more, since through-wall tube penetration results in secondary (water)
to primary (helium) inleakage, we do not believe that there is any risk
to the health and safety of the public where tube leakage occurs;
there is however, an economic penalty for the licensee. Based on
these conclusions, 'the staff does not recommend imposing or
implementing scheduled or unscheduled inservice inspection of the
steam generator tubes but recommends continuation of primary side
moisture monitoring and radiation monitoring of the secondary coolant
system as a means of initiating corrective action in the event of
steam generator tube leakage. In addition, future post-leakage
evaluations proposed by the licensee are acceptable.
Enclosure 2
ADMINISTRATIVE CONTROLS
PROCEDURES AND PROGRAMS (Continued)
c. Secondary Water Chemistry (PWRs only)
A program for monitoring of secondary water chemistry to inhibit
steam generator tube degradation. This program shall include:
1. Identification of a sampling schedule for the critical
variables and control points for these variables,
2. Identification of the procedures used to measure the values
of the critical variables,
3. Identification of process sampling points, which shall include
monitoring the discharge of the condensate pumps for evidence
of condenser in-leakage,
4. Procedures for the recording and management of data,
5. Procedures defining corrective actions for all off-control
point chemistry conditions, and
6. A procedure identifying (a) the authority responsible for the
interpretation of the data, and. (b) the sequence and timing
of administrative events required to initiate corrective
action.
•
ALL STS 6-15
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