ML20214M693
| ML20214M693 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 08/19/1986 |
| From: | Jape F, Long A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20214M657 | List: |
| References | |
| 50-369-86-19, 50-370-86-19, TAC-61512, TAC-61513, NUDOCS 8609110226 | |
| Download: ML20214M693 (14) | |
See also: IR 05000369/1986019
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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REGloN 11
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101 MARIETTA STREET.N.W.
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ATLANTA, GEORGI A 30323
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Report Nos.:
50-369/86-19 and 50-370/86-19
Licensee: Duke Power Company
422 South Church Street
Charlotte, NC 28242
Docket Nos.:
50-369 and 50-370
License Nos.:
Facility Name: McGuire 1 and 2
Inspection Conducted: July 7-11, 1986
Inspector:
M*-
R. Long
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Date Signed
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Approved by:
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F. Jape, Section Chief
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Dafe Signed
Test Programs Sections
Engineering Branch
Division of Reactor Safety
SUMMARY
Sccpe: This routine, unannounced inspection was in the area of evaluation of
fuel assembly damage by baffle jetting impingement.
Results: One violation was identified - failure to control materials entering
primary coolant system.
(See paragraph 6)
8609110226 860828
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REPORT DETAILS
1.
Persons Contacted
Licensee Employees
- N. Atherton, Compliance, Associate Chemist
- S. Copp, Maintenance, Planning Engineer
- J. Day, Licensing
- E. Estep, Project Services Engineer
- J. Foster, Health Physics
- J. Goodman, Quality Assurance
- B. Hamilton, Superintendent of Technical Services
- M. Hatley, Mechanical Maintenance, Associate Engineer
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- M. Kitlan, Jr. , Reactor Engineer
L. Kunka, Technical Services
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- D. Lampke, Operations Fuel Handling
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- S. LeRoy, Technical Specialist
B. Mcdonald, Health Physics
- R. Michael, Station Chemist
- D. Rains, Maintenance Superintendent
- M. Sample, Integrated Scheduling Superintendent
T. Saville, Design Engineer, Nuclear Engineering
- R. Tomonto, Land Engineer, Nuclear Engineering, Design Engineering
,
Other licensee employees contacted included engineers, technicians, security
force members, and office personnel.
Westinghouse Employees:
R. Meyer, Pittsburgh
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0. Correal, Pittsburgh
- J. Roth, McGuire
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NRC Resident Inspectors
- W. Orders, Senior Resident Inspector, McGuire
W. Bradford, Senior Resident Inspector, Farley
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- Attended exit interview on July 11, 1986.
2.
Exit Interview
,
The inspection scope and findings were summarized on July 11, 1986, with
those persons indicated in paragraph 1 above.
The inspector described the
areas inspected and discussed in detail the inspection findings.
The
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following new items were identified:
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Violation 369/86-19-02,
Failure to control materials entering the
primary coolant system (See paragraph 6).
IFI 369/86-19-01, Followup corrective actions on fuel assembly damage.
(See paragraph 5)
The licensee was informed in a phone call on July 29, that the first item
will be a violation.
Although proprietary information was reviewed by the inspector while at the
'
site, no proprietary information has been included in this report.
3.
Licensee Action on Previous Enforcement Matters
This subject was not addressed in the inspection.
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4.
Unresolved Items
Unresolved items were not identified during this inspection.
5.
Loose Fuel Pellets in Reactor from Damaged Assembly (92705)
Significant damage to a Unit 1 fuel assembly was discovered during the
refueling outage following Cycle 3.
The fuel had been loaded and the core
loading verification completed, when a fuel pellet was found on the baffle.
The damaged assembly was then identified. Several fuel pins in the assembly
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were breached, with an estimated 50 to 150 fuel pellets loose. About 40 of
these pellets were lying inside the assembly.
The damage to the fuel assembly is thought to have resulted from vibrations
and fuel pin rotation induced by water-jetting through the baffle joints.
The driving force for baffle jetting is a pressure differential across the
baffle joints due to reactor coolant flowing downward on the outer surface
and flowing upward on the core side of the baffle plates.
Two types of
baffle gaps related to fuel failures are (1) center-injection joints, where
the direction of the impinging flow is perpendicular to the outer row of
fuel pins and (2) corner-injection joints, where water flows parallel to the
,
-outer row of pins adjacent to the baffle plate. The fuel damage at McGuire
occurred in a corner-injection locatien.
Fuel assembly damage from baffle jet impingement was identified at the
Trojan Nuclear Station, Farley Nuclear Plant, Point Beach, and at other
reactors both within the United States and overseas.
At approximately 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br /> on Thursday, June 26, McGuire Reactor Group
members noticed small cylindrical objects on the baffle as they were
completing their visual verification of the core loading.
A previously
scheduled camera inspection of the baffle plates for loose parts or debris
(as required by procedure MP/0/A/7150/43 for Reactor Vessel Upper Internals
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Removal and Replacement) was then performed. The licensee discovered what
appeared to be four whole fuel pellets and two or three crushed pellets
lying on the baffle next to core location P-3.
Additional camera
inspections were made and the route for transferring fuel assemblies between
tne core and the spent fuel pool was checked. The route from the core to
the upender was devoid of pellets, but four or five more fuel pellets were
observed at the upender area at approximately 0000 on June 27. These events
were reported to the NRC at 0020 that morning.
On Friday, June 27, the licensee and Westinghouse determined a plan of
action to verify that the objects were indeed fuel pellets, to retrieve and
store the pellets, to identify the damaged assembly or arsemblies, and to
determine the cause of the failure.
By 0400 on June 28, a Westinghouse
underwater vacuum system was used to remove the objects and catch them in a
filter bag.
Radiation readings of the pellets confirmed them to be
irradiated fuel. Radiation levels of greater than 1000 R/hr were measured.
The NRC was notified of this confirmation at 0354.
Fuel assembly D03, which occupied position P-3 during Cycle 3, was suspected
to be the source of the fuel pellets. The three fuel assemblies surrounding
D03 were removed to create an open water area to allow for a camera
inspection of the assembly suspected of being damaged.
The assembly was
verified by camera inspection to be damaged. The top six to eight inches of
fuel rod number 16 of the 17 x 17 fuel pin array was missing, and the rod
was bent outward and toward rod 1 along the fuel assembly face.
A decision was made to move the assembly to the spent fuel pool for thorough
examination. There was a temporary delay in doing this becausa the damage
to the assembly interfered with lifting it into the fuel mast, and rod
bowing prevented ' it from being fully seated.
Assembly D03 was then
relocated to the spent fuel pool, using a net enclosure as a pellet catcher.
On July
1,
a high magnification video inspection of the assembly was
performed. The inspector viewed a video tape of the fuel assembly damage.
The licensee provided the inspector with the following written summary of
.
the results:
a.
Fuel rod 16 on face 1 protrudes frcm the assembly plane through a torn
section of grid 7 and is bent outward and toward rod 1 along the fuel
assembly face.
The top section of the rod is missing. This rod is
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also severed at several grid locations including 4, 5 and 6.
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b.
Grids 3-7 are damaged at the rod 16 location, showing a tear or
separation due to vibratory wear which is about 1/8 inch wide.
The
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grid failures are in line with the rod 16 grid springs.
c.
The cladding on rod 16 appears worn (flattened) on the outside between
grids 3 and 4.
This edge would have been in contact with the baffle
plate.
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.d.
Rod 15 is in place with the top section of the rod missing from just
below grid 8.
e.
Rods 14 and 17 are likely failed. Rods 15 and 16 are relocated on the
bottom nozzle and have indications that the end plugs are worn,
probably due to vibration.
f.
A fuel rod plenum spring is lodged in grid 8 at rod 16 location and
protrudes through the top nozzle and is in contact with the assembly
hold down spring.
g.
Numerous loose pellets and pellet debris are visible between fuel rods
and at grid locations.
Debris is predominantly between the outer two
rows of rods on face 1 with some visible between rows 2 and 3.
A more detailed summary of the damage, including an illustration, was
provided by the licensee after the inspection was concluded, and has been
appended to this report as Attachment 1.
A team was formed by the licensee to identify the cause of the problem, to
determine and perform repairs to eliminate the problem, and to assure all
pellets and fragments are removed from the system.
The peripheral location of the assembly made baffle jetting impingement a
likely cause.
The damage to the assembly was characteristic of baffle
jetting. On July 4, the core was unloaded and at approximately 1200 on
July 6, all 16 corner injection baffle joints had been measured and compared
to an acceptance criterion of 0.003" provided by Westinghouse. Thirteen of
the joints met the acceptance criterion at all elevations. Three joints
exceeded the 0.003" criterion in at least one elevation. Core location A-5
had one reading between 0.005" and 0.007" at 5' down from the top of the
core. Core location E-1 had a reading of 0.005" at 10' down from the top of
the core. Location P-3, where the damage to assembly D-3 occurred, had gap
measurements exceeding the acceptance criterion at 7 of the 12 elevations
measured.
The licensee and Westinghouse concurred that the cause of the
damage was baffle jetting impingement.
After identifying the cause of the problem, additional actions were taken to
assess the extent of baffle jetting damage to the core.
By approximately
1400 on July 6, all 16 fuel assemblies that resided in corner injection
baffle locations had been inspected by camera in the spent fuel pool.
Sipping or ultrasound testing was not performed.
No visually apparent
damage was found other than minor damage on D08 (core location R-11) to the
edge of the grid at the Face 4/ Face 3 corner. (Location R-11 had no baffle
gaps which measured 20.003"). The assembly was judged acceptable for reuse.
Although the center injection baffle joints are bolted at McGuire and,
therefore, not expected to cause problems, eight assemblies in Cycle 3
center injection locations were also inspected.
In addition, assemblies C56
and DOS which occupied position P-3 in the first two core cycles were
inspected.
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5
The licensee consulted both B&W and Westinghouse regarding repairs to
eliminate the problem. . Both vendors concurred that the preferred fix would
be replacing some outer row fuel pins with solid stainless steel in
assemblies in high-jetting locations. At the time of this inspection, the
tentative plan for preventing fuel assembly damage was to replace a four pin
by two pin section of the assembly in location P-3 and a three pin by two
pin section of the assembly in location A-5 with stainless steel rods.
These are the only two locations where a gap of greater than 0.003" was
observed in the upper portions of the fuel assemblies (where the
differential pressure across the gap is highest). If baffle jetting should
occur in these locations, the water will impinge on the steel rods and not
lead to fuel rod failure. This type of modification has been previously
licensed and successfully used at other plants. Additional repairs, such as
an up-flow modification to the baffle, will be evaluated, based on the
results of Cycle 4.
Westinghouse concurs that an up-flow modification is
not necessary at this time.
The action plan to insure all pellets and debris have been removed from the
system included video examination of the coolant system, reactor vessel, and
internals.
Following the full core unload on July 4, a video examination
was performed which included the area under the lower core plate.
This
examination identified some debris which included pieces of cotter pins and
a spring, but no intact pellets. Two towels were also found under the lower
core plate (see paragraph 6).
4
The reactor building upender and vessel area (including under the core
plate) were vacuumed and all debris removed.
The reactor building cavity
(including the deep end) was drained and all debris removed. The spent fuel
pool racks and the area under the racks was examined for debris, with none
found.
The spent fuel pool transfer area was searched for pellets. No
pellets were found, but other debris was noted and scheduled for vacuuming.
In addition to the video-inspections, radiation surveys were used to search
for fuel debris. The reactor building cavity, including the deep end, was
drained and surveyed by Health Physics.
Health Physics also surveyed the
reactor coolant system piping and filters, the residual heat removal system
piping and components, the spent fuel pool letdown lines including filters,
the refueling water storage tank piping, and the reactor building letdown
lines. No radioactivity indications of fuel debris were found in these
surveys.
The licensee has estimated that the amount of cladding which was breached
might have released up to 150 pellets from the fuel.
Approximately 40
pellets were observed with the video inspection to remain in the assembly.
About ten pellets, including pellet debris, were retrieved from the baffle
and upender.
This leaves up to 100 pellets unaccounted for. The licensee
believes that the video inspections and health physics surveys have been a
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sufficient search. The licensee also believes that this number of pellets in
the system will not present safety problems. (More pellets than this were
unaccounted for in previous baffle jetting fuel damage at other plants.)
Westinghouse concurs that this amount of missing fuel material is accept-
able.
The licensee has not been able to verify that all hard debris such as
cladding fragments have been accounted for.
An upper plenum spring is
known to be missing.
The licensee believes the amount of such debris is
within tolerable levels. Westinghouse concurs with this decision.
The licensee will be submitting a 50.59 evaluation of the safety impact of
the missing pellets and other metal debris.
As previously described, the fuel assembly damage was found because some
fuel pellets fell onto the baffle and were observed. The damage was not
observed during the refueling because of poor visibility.
The licensee's
procedures require inspection of selected fuel assemblies during each
refueling.
Four of the 24 assemblies in baffle jet locations were
inspected. Assembly D03 from location P-3 was not one of the assemblies
inspected. The decision to inspect only four assemblies was allowed by
licensee procedures and was reasonable based on the considerations described
in the paragraphs below. The four assemblies selected and inspected were in
double impingement locations, which would normally be worst case.
The
location P-3, where the fuel damage occurred, had only corner-injection
impingement.
Prior to the initial startup of both McGuire units, modifications were made
to the baffle which were believed to be adequate to prevent jet impingement
problems.
In accordance with Westinghouse Field Change Notice (FCN)
No. DAPM-10609, edge bolts were installed along the full length of the
baffle plate on all the center injection joints. All baffle joints (both
center and corner injection) were then inspected and the gap dimensions
compared to the acceptance criteria attached to the FCN. Corner injection
joints, which physically cannot be bolted, were peened in accordance with
the Westinghouse procedures to within the acceptance criteria. Peening is a
mechanical deformation of the edges of the metal plates to close the gaps.
All gaps exceeding 0.003" were peened, and most of the joints were peened to
within 0.001".
It was assumed that the corner injection joints, (the type
lending to the fuel damage at McGuire) would close with the heatup of the
reactor.
On this basis, no baffle jetting impingement problems were
anticipated.
The inspector reviewed the FCN and noted that it did not
recommend any additional baffle impingement related surveillance.
At the end of Unit 1 Cycle IA, the licensee inspected selected faces of the
assemblies in all 24 baffle jet locations, in accordance with procedure
TT/1/A/9100/80.
No baffle jetting damage was observed.
Assemblies in 8
baffle jet locations in Unit 1 Cycle 2 were inspected on all faces
(PT/0/A/4150/14). In Unit 2, all 24 assemblies were inspected on selected
faces at the end of Cycle 1, and 20 assemblies were inspected following
Cycle 2 (PT/0/A/4550/24).
No damage was observed in these inspections.
Based on these inspections and the large number of non-jet location
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assemblies to be inspected in the PIE program, a reasonable decision was
made to limit the jet-location inspections following Cycle 3 to the four
assemblies expected to be worst case.
Westinghouse wrote the licensee in June of 1982 concerning the status of
baffle jetting problems. The letter stated that " Experience to date seems
to indicate that baffle peening in conjunction with the edge bolt
modification is adequate to prevent baffle jetting..." The letter goes on
to state, referring to Cycle 1A, "The absence of any change in fission
products in the RCS chemistry at McGuire 1 provides us with continuing
evidence that baffle jetting is not occurring in that plant." The "up-flow"
modification is described, but the letter states, "It is not expected that
plants equipped with edge bolts, such as McGuire, will require this
additional modification." While the inspector was on site, a licensee
representative contacted Westinghouse to verify for the inspector that they
had sent the licensee no subsequent correspondence contradicting the
information described in the June 1982 letter.
Early on, the licensee had initiated further efforts f.o keep abreast of the
generic baffle jetting program.
Licensee representat'ves contacted Trojan
and Farley Nuclear Plants to obtain information on baffle jetting problems
encountered at those units. A file of material on the subject had been
maintained by the licensee in case McGuire were to someday have problems in
that area. The inspector looked at some examples of information from the
file and concurred that the licensee had been following the generic issue.
The licensee received IE Information Notice No. 82-27: Fuel Rod Degradation
Resulting from Baffle Water-Jet Impingement, August 1982.
The inspector
reviewed this document and found no recommendations which the licensee could
have acted upon which would have prevented the current fuel damage event.
Another very significant reason the licensee did not suspect jet impingement
problems, and thereby inspected only a sample of the jet-location
assemblies, is the large margin between the plant's coolant radioactivity
levels and the Technical Specification Limits.
The coolant isotopes monitored and trended at McGuire included Iodine-131,
,
Iodine-133, and the ratio of I-131 to I-133. The I-131 dose equivalent term
was also monitored. Coolant levels of other isotopes were also measured and
recorded, but it was not required by plant procedures that they be trended
and evaluated.
This had never been shown to be necessary, and is not
required by Technical Specifications.
McGuire's Technical Specifications limit the specific activity of the
coolant to <1.0 microcurie / gram Dose Equivalent (DE) I-131 and <100/E
microCuries/ gram gross specific activity.
Iodine-131 DE is normally the
limiting parameter.
Steady state Iodine-131 DE levels for Cycle 3 were
typically on the order of 0.03 to 0.04 microCuries/ gram, a factor of 30
below Technical Specification Limits.
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Thus, coolant activity was not considered by the licensee to be indicating
any serious problems.
The licensee was aware that coolant activities in Unit 1 were higher than in
Unit 2.
This was not considered to be significant because the Unit 1
activity levels had been somewhat high since Cycle 18. On one instance in
November of 1985, Iodine-131 DE following a trip exceeded the Technical
Specification limit.
This was attributed by the licensee to a four-month
period of continuous operation prior to the trip. The inspector reviewed
the Incident Investigation Report No. 1-85-56, and determined that the
Iodine level had returned within limits well within the allowances of the
Technical Specification Action Statement. The incident was reported to the
NRC as required.
In retrospect, the radio chemistry characteristics at McGuire 1 did in fact
show indication of gross fuel failures.
The licensee, however, did not
associate the chemistry data with baffle jetting damage because, as outlined
above, baffle jetting was not expected to be causing a problem and coolant
activity levels were relatively low.
Five groups reviewed the chemistry data at McGuire. The station chemist draws
the samples and compiles coolant activity data.
His only requirement in
plant procedures for interpreting trends in the data is that he alerts
management to any significant changes in gross coolant activity.
The
radiochemistry data is then reviewed onsite by the Performance Group, and at
the corporate office by both the Fuels and Chemistry Groups. At the end of
each cycle, Westinghouse analyzes the chemistry data for that cycle and
provides the licensee with a report on the status of fuel integrity in that
core.
The inspector reviewed chemistry data for Cycle 3 provided by the corporate
office.
A code package from Combustion Engineering had been used to
estimate the extent of fuel leakage based on coolant activity history. The
code indicated "one leaker" and no obvious clues to gross fuel failure
damage were seen by the licensee.
Although the licensee noted a low
Iodine-131/133 ratio, indicative of open defects, this was not considered
unusual relative to previous coolant activity levels. The average Cycle 3
steady-state values of I-131 and I-133 were relatively low (.008
microCuries/ml and .0584 microCuries/ml, respectively).
The inspector reviewed a package of coolant activity level data compiled by
the station chemist after the fuel damage had been discovered. The data
package contained plots which trended various isotopes beginning with the
initial startup of both units.
As previously discussed, the data had not
been trended in this way before because it was not required by procedures
and no need for such a study had been identified. A review of the features
of the curves in the package indicated signs of the fuel failures occurring.
Particularly indicative of fuel failure were levels of Neptun'um-239, and
levels of I-134, which are not normally trended.
The station chemist
compared I-131 level data and Iodine ratio data to a Westinghouse figure
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entitled " Fuel Defect Characterization." Although the Iodine level was not
unusually high and the ratio was not unusually low, the combination of the
two parameters on the graph indicated fuel pellets in the coolant system.
The inspector contacted Westinghouse to determine the origin and intended
use of this " Fuel Defect Characterization" figure. It was learned that the
curve was derived using typical plant data and is for use as a rough
indication of the nature of defects. The curve had been given to licensee
personnel as a handout in a radio chemistry class conducted by Westinghouse.
The inspector also learned from Westinghouse that their analysis of Cycle 3
coolant activity levels had shown gross fuel defects with significant
continuous exposure of pellets to the coolant. The results of the analysis,
however, had not been completed and released to the licensee before the end
of Cycle 3 refueling, so the licensee was not alerted to possible baffle
jetting damage.
At the time of this inspection, Duke had not yet determined what, if any,
changes will be made to their operating procedures based on information
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gained from this event. However, the inspector was given a description of
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an internal commitment the licensee is making at INP0's recommendation.
Duke's " Fuel Reliability Goal and Action Plan" gives steady-state goals for
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Dose Equivalent Iodine of <0.04 Micro Curies for 1986 and 0.01 Micro Curies
for 1990. If steady-state I-131 activity exceeds 0.05 Micro Curies during
the cycle, Ultrasonic Testing (UT) is to be performed for detection of
individual leaking rods.
(I-131 dose equivalent at the end of Unit 1
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Cycle 3 was about .035).
The specific scope of the UT is to be partially
determined by radiochemistry analysis.
The examination will be used:
(1) to determine failure mechanisms, (2) for reconstitution decisions, and
(3) to correlate with radiochemistry data for better predictive models.
The licensee also expressed the intention to trend and evaluate Iodine-134,
Neptunium-239, and other isotopes as necessary to assess fuel integrity
during plant operation.
l
During the next outage of of both McGuire Units, Duke will
Perform a complete inspection of the appropriate face of each fuel
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assembly presently operating in a corner-injection core location.
Measure the baffle gap on all 16 corner injection joints.
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In summary, no actions were identified by the inspector which were required
of the licensee, but not performed, that would have prevented this event.
At no time was the Cycle 3 core close to Technical Specification limits on
steady state coolant activity.
Corrective actions for the baffle jetting fuel assembly damage will be
tracked as Inspector Followup Item 369/86-19-01, Followup corrective actions
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on fuel assembly damage.
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6.
Discovery of Towels in Reactor Vessel
What appeared to be a towel was observed by the licensee beneath the Unit I
lower core plate on July 8,1986. When the object was removed from the
reactor vessel on July 25, it was found to actually be two terry cloth
towels, each about 12" by 12".
The size of the towels and the location
where they were found indicate they might have come in through one of the
cold legs. The licensee is conducting an investigation to determine how the
,
towels got into the reactor coolant system.
The towels were found as the licensee performed a special inspection of the
area under the lower core plate for debris from a damaged fuel assembly (see
paragraph 5). If this special inspection had not been performed, the towels
would probably not have been discovered and the plant would have started up
with the two towels in the reactor vessel.
Allowing the towels to get into the reactor coolant system is a violation in
the area of housekeeping.
10 CFR 50, Appendix B, Criterion II requires
adequate cleanliness control in components affecting quality and safety.
Duke Power Company Topical Report Duke-1-A, Quality Assurance Program,
commits to conform to ANSI Standard N45.2.3-1973 in the area of
housekeeping.
This standard is implemented in Station Directive 3.11.0
which requires control of all tools, equipment, materials, and supplies that
are used in Zones I, II, and III to prevent the inadvertent inclusion of
deleterious materials or objects in critical systems. Appropriate control
measures are required, such as logging items entering the clea'nliness zones.
The undetected entry of the two. towels into the reactor coolant system shows
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a failure to comply with Station Directive 3.11.0, which requires accounting
of all objects entering cleanliness Zone II. Before the towels were
discovered, the licensee was not aware that they were missing and the area
under the lower core plate is not normally inspected. The foreign materials
were fortuitously discovered rather than discovered through established
procedures to control reactor coolant system cleanliness.
The failure to comply with Station Directive 3.11.0 will be identified as
Violation 369/86-19-02, Failure to control materials entering the primary
coolant system.
A related finding in the area of housekeeping was recently documented as
Inspector Followup Item 369, 370/86-18-01, where the inspector noticed
refueling personnel working over the reactor vessel and failing to tie off
their safety glasses.
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ATTACHMENT 1
SUMMARY OF DAMAGE TO FA 003
IN CORE LOCATION P-03
Damage was limited to three fuel rods on Face-1 of this assembly.
A review of the high magnification (5 rods per pass) video inspection, provides
the following observations of fuel assembly D03.
Fuel Rod #17
This rod was limited to damage at only one location, below grid #3 (second
zircaloy grid from the bottom). This damage was observed to be a thru-wall
defect caused by fretting.
Fuel Rod #16
This rod experienced the largest degree of degradation.
A review of the
videotape identifies a loss of cladding corresponding to approximately the top
eight inches. The fuel rod spring was found lodged between the holddown spring
and the upper grillage.
The remaining portion of the rod (~12.5 inches) between the top inconel grid and
the first intermediate zircaloy grid (Grid #7) was found bent out from the
assembly towards fuel rod #1.
The tubing was split longitudinally and sheered
off to a point. Portions of the cladding were identified behind Row 1 on Face 1.
The displacement of the rod from Face 1 was found to be less than 2.5".
Thru-wall cracks were observed at grids 6, 5, and 4, with the rod at both grid 5
and 6 completely torn.
Below grid #3, the fuel rod was observed to have a flat surface along the
cladding 00. This damage is expected to have corresponded to contact with the
baffle wall.
Fuel rod #16 is seated on top of the grillage on the lower nozzle. The endcap is
worn, up to the point of the heat affected zone (HAZ) on the weld surface.
Below grid #2, swirling lines are observed along the clad.
Fuel Rod #15
Fuel Rod #15 was missing approximately the top 1.5" of the cladding (such that
the top of available rod material was seated below the inconel grid). The fuel
rod spring was not visible.
Between spacer grid 7 and 8, the cladding was
displaced and located adjacent to Rod #14 in rows 1 and 2 and Rod #15 in row 2.
Below grid #7, circumferential wear was observed.
-
f
Fuel Rod #15 is seated on bottom of the grillage on the lower nozzle. This rod
is worn approximately 1/8" to the chamfer.
,
, - -
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Attachment 1
13
Spacer Grid
With the exception of spacer grids #1 and 8, damage occurred on all grids
between Rods-16 and 17. Specifically grids 4-7 were torn completely through the
outer grid strap. Grid #3 was torn approximately 75% through and grid #2 was
missing only the dimple spring.
Debris
Fuel pellet debris was identified above spacer grids 1, 5, 6, and 7 along Face 1.
A review of face 4 identified lodged fuel pellets above-spacer grid #1, 5, 6, and
'7, between fuel rods 1 and 3.
(Some of this debris was observed from both face 1
and 4.) Portions of the fuel cladding were found behind fuel rods 6 on face #1.
FUEL ASSEMBLY D-03
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