ML17174A998
| ML17174A998 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 11/27/1991 |
| From: | Lougheed V, Nejfelt G, Phillips M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17174A997 | List: |
| References | |
| 50-237-91-32, 50-249-91-35, NUDOCS 9112030126 | |
| Download: ML17174A998 (11) | |
See also: IR 05000237/1991032
Text
U.S. NuCLEAR REGULATORY COMMISSION
REGION III
Reports No.
50-237/91032 (DRS); 50.-249/91035 (DRS)
Docket Nos. 50-237; 50-249
Licensee: Commonwealth Edison Company
Opus West III
1400 Opus Place
. Downers Grove, IL
60515
Facility Name:
Dresden Nuclear Power Station, Units 2 and 3
Inspection At:
Dresden Site, Morris, IL
Inspection Co~ducted: October 28 through November 12, 1991
Inspectors:
v. P. Lough9,ed
Approved By:
Inspection Summary
Section
Date '
.
11/2.t..:/91
Oat~
7
Date
Inspection on October 28 through November 12. 1991
(Reports No. 50-237/91032(DRS); 50-249/91035(DRS))
_
Areas Inspected:
This.was a special announced safety inspection
by regional based inspectors to review the events surrounding the
failure of containment purge valve 3-1601-24.
A general review
of the local leak rate test program was also performed.
Inspection module 61720 was used during this inspection.
Results:
The inspection resulted in one apparent violation
against Technical Specification 3.7.D which requires that
containment isolation valves be operable during periods of power
operation..
Du~ing the last operating cycle, valve 3~1601-24 was
inoperable because, when the valve operator was fully closed, the
valve disk was partially open.
This apparent violation is
9112030126 911127
ADOCK 05000237
G
described in section 3 of the report.
The report also discusses
an unresolved item relating to an apparent 50.59 violation where
a required Technical Specification update had not been made
(paragraph 4.A).
A licensee strength was identified in the
conduct and attitudes of the test personnel, which ~esulted in
improvements in the overall test program.
2
REPORT DETAILS
1.
Persons Contacted
Commonwealth Edison
L. Gerner, Technical Superintendent
J. Kotowski, Production Superintendent
J. Gates, Assistant Technical Staff Supervisor
K. Peterman, Regulatory Assurance Supervisor
M. Andjelic, Local Leak Rate Test Coordinator
D. Booth, Master Electrician
B. Col.ebank, Post-Maintenance Testing Coordinator
R. Geier, Master Maintenance Mechanic
J. Harrington, Nuclear Quality Programs Maintenance
Group Leader
M. Horbaczewski, Inservice Testing Group Leader
M. Korchynsky, Unit 3 Operating Engineer
D. Legler, Site Engineer *
D. Lowenstein, Regulatory Assurance Analyst
R. Stachniak, Performance Improvement Supervisor
D.
VeriPelt~*Assistant Maintenance Supervisor
G. Whitman, Inservice Inspection Coordinator*
K. Yates, Onsite Nuclear Safety Administrator
U.S. NRC
W. Rogers, Senior Resident Inspector
D. Liao, Reactor Engineer
All of the above 'were present at the exit held November 12,
1991, with the exception of Mr. Horbaczewski.
The inspectors also interviewed other licensee employees
during the course of the inspection.
2.
Licensee Action on Previous Inspection Findings
A.
B.
(Closed) Unresolved Item 237/90017-05 "Components not
in LLRT [Local Leak Rate TestJ Program":
This item was
addressed ih Inspection Reports No. 50-237/90006(DRS);
No. 50-249/90005(DRS) and a non-cited violation was
issued.
Based on the evaluation and resolution
contained in that report, this item is considered
closed.
(Closed) Unresolved Item 237/90027-09 "Adequacy of
Post-Maintenance.Testing in Regard to Failure of Torus
Purge Valve During an ILRT [Integrated Leak Rate
TestJ":
This unresolved item was resolved through
3
I
issuance of a violation.
See following _item for
closure.
c.
(Closed) Violation 237 /91006-01 "Inadeauate Post-
Maintenance Testing Resulting in Lack of Containment
Integrity for an Entire Cycle":
This violation, along
with a proposed civil penalty, was issued in
April 1991.
The licensee acknowledged the validity of
the violation and paid the civil penalty.
Their
response to the violation was acceptable~ Therefore,
this item is considered closed.
D.
(Closed) Systematic Evaluation Program Topics VI-4;
VI-6 "Installation of Proper Leak Rate Test Taps on the
Reactor Building Closed Cooling Water-System":
The
inspectors reviewed theinethodology Used by the
licensee to perform local leak rate testing of both the
supply and return on the reactor building closed
cooling water system.
The inspectors were satisfied
that the licensee was testing these penetrations in
accordance with the requirements of Appendix J.
This
item is considered closed.
E.
(Open) Systematic Evaluation* Program Topic VI-4.
"Leakage Conditions Under Which the Remote Manual
Isolation Valves on LPCI [Low Pressure Core Injection]
and Core Spray Systems Should be Isolated Are
Incorporated Into the Emergency Procedures":
At the
time of the inspection,** these conditions were not
covered in any emergency procedures, to the best of the
licensee's knowledge.
The licensee was unable to discern who had immediate
responsibility for closure of this item.
Therefore,
this item will remain open.
3.
Review of Events Surrounding the Failure of Unit 3
Containment Purge Valve 3-1601-24
A.
Sequence of Events
On December 25, 1989, a local leak rate test was
performed on penetration X-125, the Drywell/Torus Vent,
with successful results (8.49 standard cubic feet per
hour (scfh)).
This penetration was comprised of inside
containment valves 3-1601-23 and 3-1601-62, and outside
~alves 3-1601-24, 3-1601-60, 3-1601-61, and 3-1~01-63.
On January 27, 1990, a work request was generated to
repair the operator on valve 3-1601-24, an 18-inch*
butterfly valve, due to failure of Dresden Technical
Surveillance (DTS) 1600-27 "Fail-Safe Air Test~" The
4
maintenance performed involved replacement of the valve
operator piston rod.
A post-maintenance local leak
rate test was neither specified nor performed.
Maintenance was completed on February 3, 1990.
The following day (February 4, 199-0) an integrated leak
rate test was performed.
The* integrated test was
acceptable in the as-left condition, with a final
leakage rate of 1.02 wt%/day as compared to the
allowable of 1.2 wt%/day).
Unit 3 was returned to power on February 11, 1990.
At
this time known leakage from all local leakage source~
was 485.43 scfh.
Technical specification allowable
leakage was 488.452 scfh.
Approximately 19 months later, on September 9,* 1991,
Unit 3 shut down for .its next refueling outage.
A
local leak rate test was performed on penetration X-125
on September 16, 1991.
The* penetration could not be
pressurized.
On September 20, 1991 the licensee
identified that the leakage was through outboard
isolation valve 3-1601-24.
Investigation, by the
licensee, into the cause of the failure
determined
that, as part of the maintenance performed during the
previous cycle~ a new piston rod had been iristalled.
This piston rod increased the valye actuator stroke by
approximately one-eighth of an inch which resulted in
the valve disk rotating past the fully closed position
to where it was partially reopened.
The valve position
had been incorrect over the entire previous operating
cycle.
After making this determination, the licensee
immediately notified the NRC under the requirements of
10 CFR 50. 72 (b) (2) (iii).
B..
Regulatory Requirements
Drywell purge valve 3-1601-24, a containment isolation
valve, is addressed in Technical Specification
Table 3.7.1.
Technical Specification Limiting
Condition for Operation 3.7.D required that, during
power operation, all isolation valves listed in
Table 3.7.1 were to be operable.
If a containment
isolation valve was not operable, then the unit was
either to be shutdown, or the redundant isolation
valves in the line were to be isolated, with their
positions recorded daily.
Valve 3-1601-24 could not be
considered operable during the previous reactor power
operating cycle, because with the operator in the fully
closed position, the valve disk was partially open.
Because the licensee did not recognize that the valve*
was inoperable, the limiting conditions for operation,
5
c.
specified in Technical Specification 3.7.D,*were not
met.
This is an apparent violation of Technical
Specification 3. 7. D (249/91035-01 (DRS)) .*
Ro.ot Cause
The root cause was characterized by the licensee, ;in
Licensee Event Report (LER) 50-249/91-009, Revision o,
a?"*** inadequate controls were provided with the work
package ..** "
The inspectors concluded that the use of
informal communication methods, due to inadequate
procedures, contributed to the inadequate post-
maintenance testing.
A local leak rate test was not specified as a required
post-maintenance test*. following the val v.e operator
maintenance.
The maintenance workers did not convey*
back to the work analyst that they had installed a new
piston rod, nor did the work analyst inform the
technical staff of.the known scope of the repair.
The
technical staff considered a local leak rate test
following maintenance unnecessary, if the work
performed was on only the valve operator, and not on
the valve seating surface.
As this was their
understanding of the maintenance being performed on
valve 3-1601-24, they did not require a local leak rate
test.
The licen~ee identified, in their licensee event
report, two other Unit 2 valves which had not had local
leak rate test performed following operator
maintenance.
These were the reactor head cooling check
valve, 2-2.05-27, and a torus/drywell purge outboard
isolation valve, 2-1601-63.
The inspectors concluded
that a common cause, contributing to the failure to
perform post-maintenance local leak rate tests on these
valves, was that these were unanticipated maintenance
activities which came up during the last two weeks of
the outage.
The group responsible for performing local
leak rate tests did not have the opportunity to
determine the scope of the. maintenance and to identify
to the work analysts that local leak rate tests were
required, due to the informal methods of communication
being used.
During the past two and a half years,. the Quality
Assurance organization (QA), in their audits and
surveillances, neither considered whether appropriate
post-maintenance tests were specified nor if the work
analyst had suff i.cient information to determine what
the appropriate post-maintenance tests should have
been.
Rather, QA only verified that the testing that
6
D.,
was specified was that which was performed.
Consequently, QA did not provide a comprehensive
oversight of the post-maintenance testing program.
- corrective Actions
As part of their response to the event, the licensee
immediately initiated a comprehensive audit to ensure
that all containment isol_ation valves on the operating
Unit 2 had local leak rate tests administered following .
maintenance.
Additionally, those corrective actions
being implemented due to the failure of the inner
flange on Unit 2 torus purge valve, such as the .
. .
development and distribution of color coded drawings
showing the containment boundaries and isolation
valves, were reviewed and expanded.
Finally the lead
engineer for the local leak rate test program was
designated to review, on a. daily basis, all work
requests being performed.
However, the inspectors
identified that all of these corrective actions were
being done informally rather than through formal .
procedures.
For example, the initiative to perform the
daily reviews, discussed above, was specified by a
maintenance memorandum.
In the long term, the licensee committed to implement
further corrective actions, including:* (1) performanqe
of a comprehensive review on Unit 3 to ensure no local
leak rate tests -were omitted during the current outage
prior to Unit 3 startup; (2) development of a matrix
identifying all components having local leak rate. test,
inservice test, or other mandatory post-maintenance
testing requirements; (3) training, both initial and
requalification, of staff and operators concerning
local leak rate testing requirements; (4) formalization
of post-maintenance testing requirements into a
procedure; *and (5) evaluation of approaches used by
other Commonwealth Edison nuclear stations to ensure
that appropriate post-maintenance testing requirements
are included in work packages.
- The inspectors had no concerns with the proposed
long-term corrective actions.
4.
Review of Local Leak Rate Test Program
A.
Procedural Review
The licensee controlled the local leak rate test
program through Dresden Administrative Procedure (DAP)
14-05, "Leak Rate Testing Program", Revision 5.
Local
leak rate testing of individual Type B and C components
.J
7
was done in accordance with the following Dresden
Technical Surveillances:
1).
(DTS) 1600-1, Revision 14 (primary containment
valves);
2)
DTS 1600-2, Revision 6 (Bellows); DTS 1600-4,
Revision 9 (Electrical Penetrations);
3)
DTS 1600-14, Revision 9 (Personnel Access Lock);
4)
DTS 1600-15, Revision 8 (Double Gasketed Seals);
and
5)
DTS 0250-01, Revision 9, and DTS 0250--03,
Revision 3, (Main Steam Isolation Valves Dry and
Wet Tests)~
The inspectors reviewed these procedures against the
requirements of Appendix J, and the licensee's *
Technical Specifications.
The inspectors found that the procedures met all
Appendix J testing requ_irements, however, two concerns
were identified.
The first* involved the licensee's use
of long lengths of tubing to pressurize penetrations.
This was done to reduce the accumulated dose to the
staff performing the tests.
The inspectors discussed
with the licensee the potential.for a pressure drop
through the tubing, as well as the methodology used by
other sites to prevent this problem.
The licensee
committed to resolving this .concern.
The second
concern related to those tests performed against
reactor water head pressure.
The licensee committed to
- review the procedures for performing tests against a
water head, to ensure that these tests would be
- performed against the correct test pressure.
.
.
.
B.
Technical Specificat-ion/A'ppendix J Disagreement
The inspectprs identified that the licensee had changed
their airlock test methodology because of a 1982 NRR
Safety Evaluation Report which found that the
licensee's previous method of testing the airlocks did
not meet Appendix J requirements.
When the licensee
revised.the airlock test to be consistent with Appendix
J requirements, they failed to amend their Technical
Specifications. As of November 1991, the Technical
Specifications still. specified that the airlock be
tested at 10 psig with an acceptance criteria of
0.0375 La.
The licensee was performing airlock testing
at Pa, 48 psig, with an acceptance criteria of 0.05La.
8
c.
The licensee had prepared a Technical Specification
update; however, it had never
been submitted to the
Commission.
This concern is being tracked as an
Unresolved Item (237/91032-03(DRS); 249/91035-03(DRS)),
pending further review of the circumstances surrounding
the licensee's failure to amend the technical
specifications at the time of changing its testing
requirements.
Review of Testing Results
The inspectors reviewed the results of the as-found
local leak rate tests for the 1991 Unit 3 refueling.
The inspectors noted that, besides the X-125
penetration discussed above, the licensee had a number
of other penetrations which also could not be
pressurized.
One of these was the single valve pathway
on the high pressure coolant injection (HPCI) turbine
exhaust.
As a res.ult of this one valves' failure,
the
containment exceeded both the 0.6La maximum pathway
leakage requirement and the 0.75 minimum pathway
leakage requirement.
The licensee considered the large
number of failed valves to be unusual and not in
keeping with normal site practices.
D.
Because of a problem identified concerning the *adequacy
of Type B testing of bellows seal penetrations at
Quad Cities, the licensee recognized that the local
leak rate test results on Dresden Unit 3 might be
non-conservative.
The licensee performed tests on all
the Unit 3 bellows with both air and helium.
Only one
penetration showed signs of excessive leakage (main
steam line A) and was replaced during the outage.
The
licensee planned to submit a temporary waiver of
compliance on the bellows to NRR prior to unit startup,
explaining their intended course of action.
Additionally, the licensee committed to perform an
integrated leak rate test every refueling outage, on
both units, until either the bellows had been replaced
with testable bellows, or such testing was no longer
necessary.
E.
Licensee Self-Initiated Study of Containment Isolation
Valves
The licensee initiated a study of all containment
isolation valves on.Dresden in the 1989-1990 time
period.
The licensee has committed to submitting the
results of this study as a revision to the Updated
Safety Analysis Report (USAR) *Table 5.2.2.5 (i.e., the
9
table would be updated to show which valves arid
penetrations required Type B or c testing).
Several
penetrations were added to the local leak rate testing
program as a result of the study.
However, the study
also identified some penetrations where all the
containment boundaries were not be"ing properly tested
(e.g., the inboard flanges on the drywell and torus
purge valves) ..
At the time of the inspection, the
licensee stated that the study was on hold, awaiting a
meeting to be held.between the licensee and NRR.
The
date for that meeting had not been set.
5.
Unresolved Items
Unresolved items were matters about which more information
is required in order to ascertain whether they are
acceptable items*, items of noncompliance, or deviations.
An
. unresolved item disclosed during the inspection is discussed
in paragraph 4.B.
6.
Exit Interview
The inspectors met with licensee representatives (denoted in
paragraph 1) throughout the inspection.
An exit meeting was
held prior to leaving the site on November 12, 1991.
During
the exit, the inspectors summarized the scope and apparent
findings of the inspection.
The inspectors also *discussed
the likely informational content of the inspection report
with regards to documents or processes reviewed by the
inspectors during the inspection.
The licensee did not
identify any such d_ocuments or processes as proprietary*.
10
Commonwealth Edison Company *
2
test was not performed following the maintenance.
The failure to
prescribe and perform a local leak rate test as part of the post-
maintenance testing was partially due to the use of informal
communication between the maintenance groups and the local leak
rat.e test group as to when these tests were required. *
Accordingly, no Notice of Vio.lation is presently being issued for
this inspection finding.
Please be advised that the number and
characterization of apparent violations described in the enclosed
inspection report may change as a result of further NRC review.
You will be advised by separate correspondence of the results of
our deliberations on this matter.
No response regarding the
apparent violation is required at this time.
We do request; however, that you respond within 60 days of the
date of this letter to the unresolved item identified in the
attached inspection report.
Your response should include the
date when you expect to submit the Technical Specification
amendment.
In accordance with 10 CFR 2.790 of the Commission's regulations,
a copy of this letter and the enclosures will be placed in the
NRC Public Document Room.*
The responses directed by this letter are not subject to the
_clearance procedures of the Office of Management and Budget as
- required by the Paperwork Reduction Act of 1980, PL 96-511.
We will gladly discuss any questions you have concerning this
inspection.
Enclosure:
Inspection Reports
- No. 50-237/91032(DRS);
No. 50-249/91035(DRS)
See Attached Distribution
Sincerely,
Hubert J. Miller, Director
Division of Reactor Safety
RIII
RIII
RIII
RIII
--~-----SEE PREVIOUS PAGE----------------------
Loughe~g
Nejfelt
Phillips Wright
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