ML20058A727
| ML20058A727 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/16/1993 |
| From: | Stetka T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20058A723 | List: |
| References | |
| 50-313-93-31, 50-368-93-31, GL-85-22, IEB-93-002, IEB-93-2, IEIN-88-028, IEIN-88-28, IEIN-89-077, IEIN-89-77, IEIN-90-007, IEIN-90-7, IEIN-92-071, IEIN-92-71, IEIN-93-034, IEIN-93-34, NUDOCS 9312010205 | |
| Download: ML20058A727 (9) | |
See also: IR 05000313/1993031
Text
_
9
i
APPENDIX
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
i
Inspection Report:
50-313/93-31
50-368/93-31
,
Licenses:
NPF-6
-
t
Licensee:
Entergy Operations, Inc.
Route 3, Box 137G
i
Russellville, Arkansas
facility Name: Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2)
Inspection At: Russellville, Arkansas
,
L
Inspection Conducted: October 21-25, 1993
l
Inspectors:
R. Azua, Resident Inspector
.,
L. Smith, Resident Inspector
l
,
JCCN
p //(,f93
Approved:
-e
-
Thomas F. Stetha, Chief, Project Section D
Date ~
Inspection Summar_y
Areas Inspected (ANO-1 and -2): This special inspection was performed to
evaluate the circumstances concerning deficiencies discovered by the licensee
in the reactor building sumps for ANO-1 and -2.
Results (ANO-1 and -2):
One apparent violation involving three examples was identified:
(1)
The reactor building sump for ANO-1 was found.to have several
unscreened penetrations which were not reflected on installation
instructions and procedures and, therefore, not in accordance with
the facility design basis (Section 1.1).
(2)
The licensee failed to correctly translate the design basis
requirements for the ANO-1 reactor building-sump, when scupper
holes were added to design drawings during plant construction
without appropriate screening (Section 1.1).
9312010205 931116
ADOCK 05000313
G
- - _ __ _ - - _______
__
..
-2-
(3)
The reactor building sump for ANO-2 was found to have several
'
unscreened penetrations in the lower structural support which were
not reflected on installation instructions and procedures and,
,
therefore, not in accordance with the facility design basis
(Section 1.1).
Failure to identify tears in the screen mesh during sump cleaniness.
inspections is considered to be a weakness (Section 1.1).
Identification of penetrations in the ANO-2 reactor building sump while
the plant was at power resulted in the licensee requestint enforcement
discretion from the' requirements of Technical Specification 3.0.3 in
order to remain at power while repairs were being effected. The NRC
granted this request on October 22, 1993 (Section 1.1).
The licensee's Information Notice evaluations were focused on sump
debris cleaniness and, as a result, did not identify the reactor
building sump screen deficiencies (Section 1.2).
On October 25, 1993,
,
the licensee made a commitment to review a select sample of previous
Information Notices and other industry notification evaluations. This
effort will be performed to determine the quality of other notification
evaluations (Section 1.2).
1
Summar_y of Inspection Findings:
Apparent Violation 313/368/9331-01 was opened (Section 1.1)
Attachments:
Attachment - Persons Contacted and Exit Meeting
i
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
4
i
.
-3-
DETAILS
1 ONSITE RESPONSE TO EVENTS (93702)
1.1 Reactor Buildina Sump Desian Basis Deficiencies
On October 1,1993, with ANO-1 station in Refueling Outage IRll, licensee
,
personnel noted an apparent inadequacy with the reactor building sump screens.
The procedure to perform a calibration of the reactor building sump level-
instrumentation required that the integrity of the grouting, where the cabling
passes through the sump curb, be verified. During this verification, it was
noted that the conduit opening needed to be regrouted.
Licensee personnel
questioned why this regrouting was necessary considering that there were other
openings already existing in the curb. System engineering was notified of the
existence of these other openings. System engineering conducted a review of
the sump conditions and, as a result, identified the following penetrations
that did not conform to requirements:
Twenty-two (22) 6-inch diameter by 3-inch high semicircular holes, or
scuppers. These scuppers were constructed of 6-inch diameter pipe that
was split and imbedded in the sump curb at the reactor building floor
elevation. These scuppers were not screened.
Four (4) conduit / pipe penetrations into the sump which were not
'
,
adequately screened.
Two of the penetrations were in the horizontal
portion of the sump screen where a conduit and pipe penetrated the sump
screen. The north face of the sump had a single opening penetrated by
two conduits.
The sump access panel on the east end of the sump also
had a single opening penetrated by two conduits.
'
An approximately 1-inch conduit penetration in the west end of the sump
curb was not completely grouted nor screened to prevent debris from
entering the sump. The penetration was below the sump screen framework
'
and in the top of the sump curb.
Licensee system engineering identified that the penetrations did not conform
with the requirements for the reactor building sump screens as described in
,
the ANO-1 safety analysis report, and the unit's design basis Upper Level
Document ULD-1-SYS-04, which stated that the purpose of the sump screens was
to prevent the passage into the sump of any debris larger than the sump screen
mesh size of 0.132 inch by 0.132 inch. The failure by the licensee to
correctly translate the design basis requirements, as set forth in the ANO-1
safety analysis report, to the installed configuration, is considered to be an
apparent violation 10 CFR Part 50, Appendix B, Criterion III
(313/368/9331-01).
.
f
-
_
.
,
-4-
In addition to these design problems, the licensee identified two defects,
i
which appeared to be tears in the screen mesh covering the sump and some
debris in the sump.
One defect was located in the screen mesh on the sump
access panel at the east end of the sump (a 90 angle tear approximately
12 inches by 1 inch on the vertical and 14 inches by 1 inch on the
horizontal).
The other defect was located in the horizontal portion of the
sump screen at the south side of the sump (approximately a 12-inch by 1-inch
tear). The debris consisted of small amounts of duct tape and concrete dust
'
and gravel.
The licensee attributed this debris to ongoing cutage activities.
'
As a result of these findings, the licensee initiated Condition
l
Report CR-1-93-0390. The licensee stated that, although the sump screens were
in nonconformance with the design basis, the determination as to whether the
reactor building sump was operable or inoperable could not be adequately
determined without an engineering evaluation.
Based on the fact that ANO-1
was shut down for an outage, there was no immediate operability concern, thus
the reactor building sump was declared operable for the existing plant
j
conditions. As a result of this initial operability determination, in
accordance with the reportability guidance in the licensee's Administrative
Procedure 1000.104, Revision 10, " Condition Reporting and Corrective Actions,"
the licensee concluded that a report to the NRC was not required.
At the time of the issuance of the condition report, the licensee was
concerned with the status of the ANO-2 reactor building sump screens, since
ANO-2 was at full power operation. The licensee reviewed the as-built
drawings of the ANO-2 sump screens to determine if scupper holes, or any other
penetrations that might bypass the sump screens, were present. The licensee
also reviewed the pictures from the plant surrogate tour system to verify if
any conduits or pipes penetrated through the screens.
In addition, they
verified that no apparent tears in the screen mesh were visible. The licensee
also made a determination that the angle of the pictures provided sufficient
view of the sump screen base to corroborate their conclusion that no scupper
holes existed.
Finally, the licensee reviewed previous postoutage/preheatup
containment building visual inspection documentation to determine if any
anomalies had been previously noted.
None were identified.
Based on this
review, the licensee believed that the conditions affecting the ANO-1 reactor
building sump screens did not similarly affect ANO-2.
The licensee continued to review the issue concerning the operability status
of the ANO-1 reactor building sump in an effort to correct the identified
deficiencies prior to plant startup.
In accordance with Procedure 1000.104, a
corrective action plan review was required within 14 days of the initiation of
the condition report. The licensee had set a goal to have a final operability
determination completed within this 14-day period. The licensee's reviews
included a review of the debris transport calculations that would help
determine the amount of debris the sump screens would be subjected to in the
event of a loss-of-coolant accident.
In addition, the licensee was reviewing
other plant drawings and touring the reactor building to determine if other
_____
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
___
-5-
drainage mechanisms existed that could adversely affect the sump. Finally,
the licensee was developing the corrective action plan to address the
deficiencies.
While reviewing Drawing C-172, " Reactor Building Concrete Details-Sheet 2,"
the licensee noted that, although the drawing indicated the presence of the
scupper holes in question, it did not provide any indication that they needed
to be screened.
The presence of these holes on the drawing, without any
specifications that they be screened, was a failure by the licensee to
correctly translate the design basis requirements for the reactor building
sump screens into drawings.
This is another example of violation
(313/368/9331-01).
During the review of the ANO-1 reactor building drainage system, the licensee
identified that the reactor building drain headers, which empty into the sump,
were unscreened and provided several additional potential paths for debris
intrusion into the sump.
These paths included six 10-inch floor drains at
various elevations, the 2-inch drain line under the reactor cavity, several
open funnels at upper elevations which catch drainage from various equipment,
and 2-inch open drains under the rails of the refueling machine.
The licensee
included these findings in the corrective action plan being developed.
The licensee did not believe that the problems identified in the ANO-1 reactor
building drainage system were present in ANO-2.
This belief was based on the
knowledge that ANO-2 was a more recent design and, due to the fact that ANO-2
was of a different design, had different drain covers and screens in the
drains, as corroborated by a cursory review of some of the design
documentation.
On October 11, 1993, the licensee completed the repairs and modificationc on
the Unit I reactor building sump screens, curbs, and drains.
As of October 14, 1993, a final operability determination had not been made by
the engineering staff. The licensee made the decision to declare that the
reactor building sump had been inoperable due to the probability, based on the
engineering evaluation conducted to date, that the low pressure injection
system would be adversely affected due to debris migrating into the reactor
building sump following a large break loss-of-coolant accident. As the result
of this decision, a report was made to the NRC in accordance with the
requirements of 10 CFR 50.72.
To completely ensure AND-1 reactor building sump integrity, the licensee
conducted subsequent walkdowns between October 12 and 18. As the result of
these walkdowns, additional findings, which included small holes and tears in
the sump screens, and sump frame bolting deficiencies were identified.
The
licensee took action to correct these conditions.
On October 22, the licensee determined that the ANO-1 reactor building sump
had been inoperable prior to the last refueling outage.
This conclusion was
based on calculations which determined that the least amount of volume of
l
-
_ _ --- _ ----
)
.
.
_
-6-
insulation required to significantly degrade flow to the decay heat removal
outlet valves and restrict low pressure injection flow was calculated to be
. 5-1.0 f t'.
An assessment of the transport calculations for insulation that
could be postulated to break free following 'a large break loss-of-coolant
accident, indicated that approximately 15 ft of insulation could have entered
the sump.
On October 18, 1993, the licensee decided to perform a more in-depth review of -
the design documentation of the ANO-2 reactor building drainage system.
By
October 21, 1993, the licensee had identified that all but three drains in the
reactor building had the appropriate sized screens, as set forth in the design
basis requirements. The licensee was unable to confirm the presence of
screens for these remaining three drains.
In addition, the licensee
determined that two of these three drains were inaccessible while the plant
was operating at power.
Therefore, the licensee began fabricating screens for
installation on the ends of the piping from the drains in question where they
entered the sump.
Based on the uncertainty resulting from the lack of supporting data, the
licensee made the decision to make an ANO-2 reactor building entry while at
power.
The purpose of this entry was to inspect the ends of the piping from
the drains in question for the new screen installation, to inspect a random
sample of the other reactor building drains (to verify the accuracy of the
design documentation), and to perform an inspection of the reactor building
On October 22, the licensee made the decision to convene the AN0-2
safety review committee to draft Continued Safe Operation
Determination CS0-93-02. The purpose of this document would be to provide the
basis for an Enforcement Discretion Request in the event that during the tour
of the sump a condition, that would cause the sump to be declared inoperable,
was identified and had to be corrected.
On October 22, members of the licensee staff entered the reactor building
accompanied by the inspector. The drains inspected were all found to be
properly screened during the tour of the reactor building. When inspecting
l
the reactor building sump screens, the licensee identified seven penetrations
through the grout at the base of the sump screen assembly, which were
~
unscreened and, therefore, could bypass the sump screens.
In addition, the
licensee identified a flashlight within the sump structure.
The dimensions of each penetration were approximately 1 inch high by 3 inches
in length. These penetrations made the sump design contrary to the design
requirements set forth in the ANO-2 safety analysis report and design basis
i
Upper Level Document ULD-2-SYS-04, which stated that the purpose of the sump
screens was to prevent the introduction of any debris into the reactor
building sump which was larger than would be allowed to pass through the .09-
inch by .09-inch sump mesh screen.
These holes in the grout did not appear on
any of the design drawing instructions or installation procedures. This
failure by the licensee to correctly translate the design basis set forth in
..
.
. - _ _ _
_ _ _ _ _ _ _
.
_
-
-
.
.
.
.
F
-7-
,
the safety analysis report into installation instructions and procedures is
,
another example of violation (313/368/9331-01).
As a result of this finding, the licensee declared the ANO-2 s"mp inoperable
.
and placed ANO-2 in Technical Specification Action Statement 3.0.3.
This
!
'
action statement required the licensee to correct the deficiency within I hour
or begin plant shutdown so that the plant was in hot standby within the next
,
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and cold shutdown in the following 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. The licensee then
!
submitted their request for enforcement discretion, CS0-93-02, to the NRC to
allow the licensee to continue operation for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> while they repaired the
sump screens. The NRC reviewed the licensee's submittal and
granted enforcement discretion at 8:13 p.m. on October 22, 1993.
'
On October 23, at approximately 3:30 a.m., the licensee completed the sump
screen required modifications and exited Technical Specification 3.0.3
approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after they had entered it.
l
The licensee planned on completing their operability determination of the as-
found condition of the ANO-2 sump by November 5, 1993.
In addition, the
licensee evaluated the affect of the flashlight found in the sump and
..t
determined that it did not pose a clogging threat due to the existence of
vortex breaking screens locrted on the end of suction piping.
i
1.2 Generic Communications Response
i
The inspector conducted a review of the licensee's actions with respect to
previous notifications regarding reactor building sumps that the NRC had
issued to the industry.
The NRC had issued several information notices
(IN's 88-28, 89-77, 90-07, 92-71, and 93-34), a generic letter (GL 85-22), and
a bulletin (NRC Bulletin 93-02), addressing reactor building sumps, sump
screen design, and the different types of material that could become debris
and obstruct the screens during the design basis accident.
.
The main topic of most of these notices pertained to insulation materials and
the potential for these materials to block the sump screens in the event of
the design basis accident.
It appeared that the licensee's focus was
concentrated mainly on the debris concerns and not on screen design, screen.
integrity, or screen configuration.
This became evident in the licensee's
response to Information Notice 89-77, " Debris in Containment Emergency Sumps
and Incorrect Screen Configuration." The information notice identified
deficiencies in the sumps of three operating facilities, which included
designs not in accordance with plant drawings, missing screens, and gaps in
the screens which resulted in the introduction of debris'large enough to have
caused pump damage or flow degradation.
In response to this notice, the
licensee implemented a procedural change on August 6, 1990, to the ANO-1
precriticality checklist to require the inspection of sumps, sump area, and
grating for debris that may have the potential for screen inlet blockage.
Since ANO-2 already had a precriticality checklist requirement to inspect the~
sump for debris and screen integrity, no recommendations were made. The
licensee's review of the information notice did not include the need for a
-
-
-
. - -
-.
.
._
.
.
.
_
.
. _ _
.
'
.
-8-
plant specific walkdown of the sumps to address the screen integrity and
]
configuration problems identified in the information notice.
)
'The inspector discussed the finding regarding the adequacy of generic
communication followup with the licensee. As a result of these discussions,
the licensee committed to review a selected sample.of previous Information
Notices and other industry information evaluations back to 1988 to assure that
these communications were properly reviewed and resolved. The licensee stated -
that, if significant discrepancies are identified during this review, the
scope of the review would be increased.
i
1.3 Conclusions
Upon discovery of the reactor building sump deficiencies the licensee took
prompt and effective corrective actions, which involved thorough engineering
evaluations. The failure to insure that the reactor building sump designs
were in accordance.with-the design basis requirements, specified in the safety
'
analysis reports, was considered to be a violation. The licensee's failure to
- I
conduct a complete and thorough review of previous industry notifications
regarding sump screen integrity and configuration is considered to be a
3
j
weakness.
.
.
,
i
!
,
,
,
' ~ '
"-
-
.,,
_,y.
.
.
..
-
-
i
ATTACHMENT 1
1 PERSONS CONTACTED
,
1.1 Licensee Personnel
- C. Anderson, Unit 2 Operations Manager
- S. Bennett, Acting Licensing Supervisor
- J. Barrett, Quality Control Supervisor
'
S. Boncheff, Licensing Specialist
- M. Cooper, Licensing Specialist
- W. Greeson, Mechanical, Civil and Structural Engineering Manager
i
- R. Howerton, Engineering Support Manager
- R. King, Acting Licensing Director
- R. Lane, Design Engineering Director
- D. Mims, Unit 2 System Engineering Manager
- J. Taylor-Brown, Quality Control Supervisor
- J. Vandergrift, Unit 1 Plant Manager
- J. Yelverton, Vice President Operations
- Denotes personnel that attended the exit meeting.
In addition to the
personnel listed above, the inspectors contacted other personnel during this
i
inspection period.
2 EXIT MEETING
An exit meeting was conducted on October 25, 1993.
During this meeting, the
inspectors reviewed the scope and findings of the report. The licensee
acknowledged the inspection findings and did not express a position on these
-
findings.
The licensee committed to review previous NRC and industry
communications to assure proper resolution as discussed in this report.
The
licensee did not identify as proprietary any information provided to, or
reviewed by, the inspectors.
i
!
!
!
!
!
,
l
,