ML20149E766
| ML20149E766 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 01/08/1988 |
| From: | Jaudon J, Tapia J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20149E756 | List: |
| References | |
| 50-382-87-29, IEB-87-002, IEB-87-2, NUDOCS 8801140012 | |
| Download: ML20149E766 (6) | |
See also: IR 05000382/1987029
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APPENDIX
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U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
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NRC Inspection Report:
50-382/87-29-
Operating License: NPF-38
Docket:
50-382
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Licensee: - Louisiana Power.& Light Company (LP&L)
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317 Baronne Street
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New Orleans, Louisiana 70160
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Facility Name:
Waterford Steam Electric Station, Unit 3 (WSES)
Inspection At:
Taft, Louisiana
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Inspection Condacted:
November 30 through December 4, 1987
Inspector:
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J. /1.)Tapi'a, Projpit ' Engineer, Project
Date
St< tion A, DiviMon of Reactor Projects
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Approved:
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J. P' Jaudon, ';hief, Project Section A
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Division of Reactor Projects
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Inspection Summary
Inspection Conducted November 30 through December 4, 1987'(Report 50-382/87-29)
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Areas Inspected:
Routine, announced inspection of previously identified items
(one violation and two unresolved items), and implementation of licensee
actions taken in response to NRC Compliance Bulletin No. 87-02.
Results: Within the two areas inspected, no violations or deviations were
identified.
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8801140012 880108
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ADOCK 05000382
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. DETAILS
1.
Persons Contacted
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WSES
- G. Wuller, Operational Licensing Supervisor
T. Garrets, Nuclear ~ Services Manager
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- R. Bennet, Supplier Audits Quality Assurance (QA) Supervisor
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B. Toups, QA Representative
Partek-Laboratories
T.'Blanchard, Marketing Manager
R. Sutton, Chief Engineer
- Denotes those present at the exit interview.
In addition to the above personnel, the NRC inspector. held discussions
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with various operations, engineering, technical support, and
administrative members of the licensee's staff.
2.
Followup of Previously Identified Items
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(Closed) Unresolved Item (382/8602-01):
Procedure for Documenting
Evaluations of Events - During an inspection in January 1986, the NRC
inspectors found entries in the shift supervisor and reactor operator logs
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for the 7 a.m. to 3 p.m. shift on January 8, 1986, indicating that control
element assembly (CEA) No. I had dropped:with the reactor at approximately
40 percent power.
The logbooks indicated that the electrical breaker was
found open, was reclosed, the dropped CEA' recovered, and power ascension
continued.
The NRC' inspectors questioned licensee management at the time
of the event on the extent of the evaluation of the incident _since no
evidence of a formal. evaluation could be identified.
It was explained
that althoughLan evaluation had been performed by the assistant plant
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manager for operations and maintenance, the operations superintendent, and
the maintenance superintendent, it had not'been documented.
The plant
manager committed to conduct a review to determine if a weakness existed
in the licensee's event reporting and evaluation program.
Since the
dropped CEA and ensuing actions did not violate Technical Specification
requirements, the dropping of the CEA did not constitute a reportable
event.
Nevertheless, the pending results of the licensee's review were
considered an unresolved item.
This item was subsequently addressed in June 1986 (NRC Inspection
Report 50-382/86-13).
During that inspection, the NRC inspector reviewed
the results of the licensee's evaluation of the event reporting and
evaluation program.
The inspection disclosed a memo from the plant
manager to various members of plant management on the subject of
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determining accountability when investigating and reporting events and
also disclosed various event rcoorts documented by memos to file,
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including one for the dropped CEA incident.
The NRC inspector noted that
it appeared that an independent review.of event reports would be
appropriate in some instances and that the threshold for initiating a
report was not clear in all instances.
Also,-the method of documentation
required was not always clear (e.g. , potential reportable event, condition
identification work authorization, quality' notice, memo to file, etc.).
Based on these observations, the unresolved item remained open.
In response to these additional observations, the licensee. established an
Event Analysis and Reporting Group consisting of a Senior Engineer and two
Associate Engineers and having primary responsibility for the
investigation, documentation, and closure of Potentially Reportable
Event (PRE) reports.
Administrative Procedure No. UNT-6-010,. Revision 4,
dEvent Notification and Reporting," provides the instructions to all plant
personnel for reporting conditions potentially adverse to quality.
The
procedure also references Administrative Procedure Nos. UNT-5-002,
Revision 7, "Condition Identification," and QP-015-001, Revision 3,
"Nonconformance and Corrective Actions." These three documents p; vide
the threshold and define the appropriate document to obtain resolution of
reported conditions.
In addition, Procedure No. UNT-6-010 now requires
approval of PRE reports by the Plant Operations Review Committee.(PORC).
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Discussions with the head of the Event Analysis and Reporting Group
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disclosed pending changes, as yet unfinalized, in the overall method for
reporting problems.
This licensee representative indicated that a generic
problem report modeled after the Institute for Nuclear Power
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Operations (INP0) Human Performance Evaluation System is forthcoming.
Based on the actions taken to date, the unresolved item concerning the
procedure for documenting evaluations of potentially reportable events is
considered closed.
(Closed) Unresolved Item (382/8606-01):
Component Cooling Water Valve
Checklist and Drawing Comments - During March 18-21, 1986, NRC inspectors
reviewed the component cooling water system standby valve lineup and
performed a walkdown of accessible portions of the system to verify
operability.
As a result of that inspection, discrepancies were noted
between the valve lineup procedure and the affiliated drawings.
Although
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all main flow path valves were correctly identified in the valve lineup
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procedure and were found correctly positioned during the walkdown, the
noted discrepancies were considered an unresolved item.
Previously, on
January 9, 1986, the NRC inspectors performed a walkdown of the Essential
Services Chilled Water System and generated comments siinilar to the
discrepancies which are the subject of this unresolved item (see NRC
Inspet. tion Report 50-382/86-02, paragraph-8).
At the time of the January
inspection, the Assistant Plant Manager for Operations and Maintenance and
the Operations Superintendent responded to the NRC inspectors' comments by
describing an upgrading program for retagging all plant valves and
reviewino and updating each safety-related valve line-up checklist.
During
following subsequent inspections, the.NRC inspectors augmented
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discrepancies noted during ESF system walkdowns:
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NRC Inspection Report No.
ESF System Walked Down
50-382/86-11
Auxiliary Component Cooling Water
50-382/86-15
High Pressure Safety Injection
50-382/86-17
Low Pressure Safety Injection
50-382/86-29
Emergency Feed Water
50-382/87-01
The upgrading program completion and correction of the deficiencies
identified in the listed NRC inspection reports were reviewed during this
inspation.
Although the upgrading program was at one point expanded to
include the labeling of all electrical breakers, only the quality of valve
tagging and associated valve line-up procedures were addressed. .The
results of the upgrading program indicate that all noted discrepancies
were corrected and that the program encompassed all safety-related
systems.
In order to ensure that the quality of valve tagging remains
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current, Procedure Nos. PMI-315, "Instructions for Dispositioning Detailed
Construction Packages," and PE2-006, "Administrative Procedure - Plant
EngineerinC Station Modification," require that all components be properly
tagged before any modification is tested and placed in service.
Based on
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the implementation of these procedural requirements and on the results of
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the valve line-up checklist walkdowns, this matter is considered closed.
(Closed) Violation (382/8634-01):
Failure to Quantify Containment
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Isolation Valve Leakage - During an inspection conducted December 8-11,
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1986, it was datermined that the lucal leak rates for six valves required
to be tested had recorded as-found leak results of "off-scale" with no
recorded value, so that the continuous running sum of local leak rates was
not known and the required comparison with the Appendix J criterion could
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not be made for the as-found condition.
This was identified as a
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violation of the requirements of Appendix J,Section III.C.3, which
states, in part, "The comt'ined leakage rate for all penetrations and
valves subject to Type B and C tests shall be less than 0.6 La."
At the
time that the local leak rate testing was being performed, the plant was
in Mode 5, cold shutdown.
The six valves in question had as-found leak
rates which exceeded the capability of the test equipment.
Prior to
performing maintenance on the six valves, the local leak rate testing
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equipment was modified to provide a greater flow measuring capability.
Corrective action was initiated by maintenance on all six valves and tne
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appropriate retests were conducted.
Of the six valves identifitd as
having leak rates of "off scale," numerical values were identified for
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five of these valves.
The sixth valve, a 24-inch containment vacuum check
valve (CVR-102) could not be pressurized with the modified test unit, thus
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no leak rate could be established.
In accordance with the requirements of
Appendix J, it was assumed that the leakage for Valve CVR-102 would result
in exceeding the 0.60 La limit.
Accordingly, per the requirements of
Appendix J,Section V.B.3, Valve CVR-102 is the subject of a separate
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accompanying summary report to the "Reactor Containment Building
Integrated Leak Rate Test" report.
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In order to prevent recurrence, Surveillance' Procedure No. PE-5-002,
"Local Leak Rate Test (LLRT)" was modified to require that, "After each
individual test.a cumulative leakage total shall be updated to reflect the
new as-found leakage."
In addition, the procedure now requires that if
the leakage-is higher.than the current measuring capabilities, then it
will be assumed that the.0.6 La limit has been exceeded and the applicable
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portions of Technical Specification 3.6.1.2 will be invoked.
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review of the actions _taken in response to the Notice of Violation, this
matter is considered closed.
3.
Response to NRC Compliance Bulletin No. 87-02:
Fastener Testing to
Determine Conformance With Applicable Mate"ial Specifications
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The licensee's ongoing actions taken in response to NRC Compliance
Bulletin No. 87-02 were addressed during this inspection.
The bulletin
was issued to request that licensees review their receipt inspection
requirements and internal controls for fasteners and determine, through
independent testing, whether fasteners in storage' meet the required
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mechanical and chemical specification requirements. . The bulletin also
required the participation of an NRC inspector-in the selection of the
sample for testing.
The NRC inspector's examination of ongoing actions included reviewing the
licensee's receipt inspection program and procedures for safety-related
and nonsafety-related fasteners to determine what characteristics are
inspected and in reviewing the maintenance and warehousing procedures for
the issuance and control of fasteners.
The following site quality
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procedures, including historical revisions, were reviewed:
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QI-010-006, "Materials Receipt Inspection"
QI-010-002, "Material Storage Inspection"
QI-004-001, "Site Review of Procurement Documents"
QI-004-002, "Records Review Checklist"
UNT-8-044, "Requisition and Return of Items to Stores"
The following QA procedures, including historical revisions were also
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reviewed:
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QAP-208, "Procurement Administration"
QAP-250, "Materials Receipt Inspection"
The NRC inspector participated in a meeting held onsite between licensee
representatives and personnel from Partek Laboratories.
This meeting was
held to discuss bulletin requirements and the testing schedule.
The NRC
inspector verified that the testing laboratory was on:the licensee's
qualified suppliers list and reviewed the results of the licensee's audit
which was performed to provide the basis for the testing laboratory's
qualification.
It was agreed that LP&L would conduct a surveillance of
all ongoing testing.
The following safety related sample site was decided
upon after discussion of the Bulletin requirements and plant fastener
usage:
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Type
_ Quantity'to be Tested
Approximate Percent of Use
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A193, Grade B7
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A193, Grade B8
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A325', Type 1
24
45'
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A490
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A307
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The total sample of 54 safety-related fasteners was selected in. increments
of three to provide one chemical, one hardness, and one tensile test.
For
nonsafety-related fasteners, six lots of four fasteners each were selected
for a sample site of 24.
No tensile testing is required for
nonsafety-related fasteners.
With the NRC inspector participating, the samples selected for testing
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were pulled from warehouse stock.
The samples were bagged by lot-and
identified by type and quantity.
The NRC inspector confirmed that tF.e
sample taken was properly tagged.
The results of all. tests, togeth.:r with
supporting information, will be reviewed during a subsequent insp2ction.
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4.
Exit Interview
The inspection scope and findings were summarized with those persons
indicated in paragraph 1 at the end of this. inspection.
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