ML20034H841
| ML20034H841 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 02/24/1993 |
| From: | Wong H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20034H821 | List: |
| References | |
| 50-528-92-43, 50-529-92-43, 50-530-92-43, NUDOCS 9303220139 | |
| Download: ML20034H841 (25) | |
See also: IR 05000528/1992043
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U. S. NUCLEAR REGULATORY COMMISSION
REGION V
Report Nos.
50-528/92-43, 50-529/92-43, and 50-530/92-43
Docket Nos.
50-528, 50-529, and 50-530
License Nos.
Licensee
Arizona Public Service Company
P. O. Box 53999,' Station 9082
Phoenix, AZ 85072-3999
Facility Name Palo Verde Nuclear Generating Station
Units 1, 2, and 3
Inspection
Conducted
December 15, 1992 through January 19, 1993
Inspectio
Location '
Wintersburg, AZ
Inspectors
J. Sloan, Senior Resident Inspector
H. Freeman, Resident Inspector
D. Proulx, Resident Inspector (WNP-2)
F. Ringwald, Resident Inspector
D. Acker, Reactor Inspector, Region V
M. Royack, Reactor Inspector, Region V
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T. Sundsmo, Project Inspector, Region V
7/NF/qJ
Approved Bv
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F. Wong, Chi'ef
V
-Date Signed?
Reactor ProjectsSection II
Inspection Summary:
Areas Inspected: Routine, onsite, regular and backshift inspection by.the
resident insper. tors and four Region V personnel.. Areas inspected included:
review of plant activities - Units 1, 2, 3;
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surveillance testing - Units 1, 2, and 3 -
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plant maintenance - Units 1,' 2, and 3
testing of main ~ steam safety valves - Unit 2
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qualification of engineering personnel - Units 1, 2,- and 3
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emergency lightingicorrective action followup - Units- 1, 2, and 3'
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fire watch program changes - Units 1, 2, and 3'
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review of independent safety engineering (ISE) assessment 92-24 - Units
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overfill of cable trays
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Appendix R fire door problems
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qualification of electrical cable in containment penetration area
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9303220139 930226
DR
ADOCK 0500
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atmospheric dump valve post-restart item followup
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temperature compensation and corrective actions for atmospheric dump
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valve nitrogen accumulator pressure drop test
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corrective action report (CAR) 90-0010
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followup on previously identified items - Units 1, 2, and 3
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review of licensee event reports - Units 1 and 2
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During this inspection the following Inspection Procedures were utilized:
60705, 61726, 62700, 62702, 62703, 62705, 64704, 71707, 92700, 92701, 92712,
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and 93702.
Results: Of the seventeen areas inspected, one violation was identified in
Unit 2, regarding an inadequate procedure for testing main steam safety-
valves. One deviation was identified in Units 1, 2, and 3 regarding
qualification of engineering personnel and two non-cited violations, in Units
1 and 3, were identified regarding implementation of changes to the fire watch
program.
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General Conclusions and Specific Findinas:
Sianificant Safety Matters:
None
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Violations:
One cited violation - Unit 2
One non-cited violation - Unit 1
One non-cited violation - Unit 3
Deviations:
One deviation - Units 1, 2, and 3
Open Items:
Eight followup items were closed, one followup item was
left open, and one new followup item was opened..
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Strenaths Noted:
Licensee management is pursuing higher standards for the
educational qualification of engineers.
Weaknesses Noted:
Two weaknesses were identified in this inspection period.
The first involved the mishandling of vendor information
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which was observed to be a continuing problem, despite-
substantial improvements in the-programmatic controls.
The second weakness involved changes to the fire watch
program which were not well implemented.
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DETAILS
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1.
Persons Contacted
The below listed technical and supervisory personnel were 'among those
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contacted:
Arizona Public Service (APSI
R. Adney,
Plant Manager, Unit 3
J. Bailey,
Director, Nuclear Engineering
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R. Bernier,
Supervisor, Nuclear Regulatory Affairs, Technical
- T. Bradish,
Manager, Nuclear Regulatory Affairs
R. Cherba,
Manager, Quality Systems
R. Flood,
Plant Manager, Unit 2
- R. Fountain,
Supervisor, Quality Audits and Monitoring
- R. Fullmer,
Manager, Quality Audits and Monitoring
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D. Garchow,
Manager, Performance Engineering
- F.
Garrett,
Manager, Fire Protection
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- M. Halpin,
Supervisor, Operations Standards
K. Hamlin,
Director, Nuclear Safety
- W. Ide.
Plant Manager, Unit 1
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- J. Levine,
Vice President, Nuclear. Production
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D. Mauldin,
Director, Site Maintenance and Modifications
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- W. Montefour,
Senior Coordinator, Management Services
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- J. Napier,
Engineer, Nuclear Regulatory Affairs, Operations
M. Oren,
Manager, Engineering
G. Overbeck,
Director, Site Technical Support'
R. Prabhakar,
Manager, Independent Safety and Quality Engineering'-
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R. Roehler,
Supervisor, Nuclear Regulatory Affairs, Operations
- C. Russo,
Manager, Quality Control
R. Schaller,
Assistant Plant Manager, Unit 1
- J. Scott,
Assistant Plant Manager, Unit 3
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- T. Shriver,-
Assistant Plant Manager, Unit 2
R. Stevens,
Director, Regulatory and Industry Affairs
Others
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- J. Draper,
Site Representative, Southern California Edison
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- F. Gowers,
Site Representative, El Palo Electric ,
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- R. Henry,
Site Representative, Salt River Project'
Denotes personnel in attendance at the. Exit meeting held with the NRC
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resident inspectors on January 20, 1993.
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The inspectors also interviewed other licensee and contractor personnel _
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during the course of the inspection.
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2.
Review of Plant Activities -' Units 12 -2.and'3l71707and937d2)
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Unit 1
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The plant: operated at essentially 100% power through~out-the
inspection period _except for_~aldownpower:.to 65% power on January l16,
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1993 to replace a main feedwater pump power supply.- The plant
-returned to 100% power on January.17, 1993.;
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Unit 2
Unit 2 operated at- essentially 100% power throughout this inspectioni
period.
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Unit 3
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Unit 3 operated at essentially 100% power throughout this inspection.;
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period.
d.
Plant Tour
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The following plant areas at Units:1, 2, and 3 were toured by the.
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inspector during the inspection:
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Auxiliary Building-
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Control Building
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Diesel Generator Building .
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Fuel Building'
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Main Steam Support Structure-
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Radwaste Building .
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Turbine Buildings
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Yard Area and Perimeter
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The following areas were' observed during.the: tours:-
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(1) Ooeratino loos a~nd Records.
Records were-reviewed ^against:
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technical specification and administr'ative control procedure '
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requirements.
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(2) Monitorina Instrumentation - ProcessL instruments were observed '
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for correlation between channels; andsfor,conformance~ with1
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technical specification requireme'nts.
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(3)
Shift Staffina - Control room andishift st'affing were observedi
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for conformance with.10 CFR Part:50.5.4..(k), technical'
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- specifications,:and: administrative procedures.3
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Eauipment Lineuos' . Various valves-'and electrical ~ breakersLwere
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verified to be in the position'or condition ~ required by y:
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-technical specifications' and administrative. procedures forf the :
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applicable. plant mode..
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(5)
Eauipment Taaaina - Selected equipment, for which tagging
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requests had been initiated, was observed to verify that tags
were in place and the equipment was in the condition specified.
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(6) General Plant Eauipment Conditions - Plant equipment was
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observed for indications of system leakage, improper
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lubrication, or other conditions that could prevent the systems
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from fulfilling their functional requirements.
The inspector observed that fasteners were missing from several
components in Units 1 and 2, most notably from the casings of
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motors for safety injection pumps. While none of the
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deficiencies appear to impact operability, they represent
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additional indication that maintenance restoration is not-
always thorough. The licensee acknowledged these findings and
stated that walkdowns of plant systems had been initiated to
identify fastener discrepancies.
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(7)
Fire Protection - Fire fighting equipment and controls were
observed for conformance with technical specifications and
administrative procedures.
(8)
Plant Chemistry --Chemical analysis results were reviewed for-
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conformance with technical specifications and administrative
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control procedures.
(9)
Securitv - Activities observed for conformance with regulatory
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requirements, implementation of the site security plan, and
administrative procedures included vehicle and personnel
access, and protected and vital area integrity.
(10) Plant Housekeepina - Plant conditions and material / equipment
storage were observed to determine the general state of
cleanliness and housekeeping.
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(11) Radiation Protection Controls - Areas observed included control
point operation, records of licensee's surveys within the
radiological controlled areas, posting of radiation and high
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radiation areas, compliance with radiation exposure permits,
personnel monitoring devices being properly worn, and personnel
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frisking practices.
(12) Shif t Turnover - Shift turnovers and special . evolution-
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briefings were observed for effectiveness and thoroughness.
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The inspector observed the Unit 1 Operations Supervisor. brief'
one crew on upcoming emergency operating procedure (EOP).
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changes. The inspector noted that E0P 41EP-1R005, Loss of All
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Feedwater, would be revised to add a note just prior to step.
3.15 listing the conditions operators should meet prior to
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opening the main steam isolation bypass valves. One condition,
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to verify that downstream piping is near normal operating
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temperatures, appeared to be not specific and unnecessary. The
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licensee agreed and initiated Instruction Change Request 52441
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to delete this_ guidance. Operations Standards indicated that
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this change is due to be implemented by May 1993. .The
inspector encouraged the licensee to review E0Ps for other
similar instances of guidance to operators which are not
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specific, to eliminate the potential for operator confusion
during emergency operations. The licensee acknowledged the
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inspector's comment.
The inspector observed a Unit 1 Assistant Shift Supervisor
(A/SS) accompany the reactor operators (R0s) during the shift
turnover board walkdown. The inspector noted that this was. not
typical and considered it to' be a very positive step. During
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this tour the A/SS noted two instances where additional detail
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could be provided to assist the oncoming operators. The
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inspector encouraged the licensed control room supervision to
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observe R0 turnovers often enough to be aware of the'
effectiveness of operator turnovers. The licensee acknowledged
the inspector's comment.
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(13) Special Variances - The inspector reviewed the Special Variance
log in Unit 3 and observed three Special Variances that did not
appear to be appropriate. Special Variance 3-92-074, dated
July 24, 1992, dealt with a change to torque requirements for
valve SGN-V341, an atmospheric dump valve nitrogen isolation
valve, which had been replaced with a new identical valve. The
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new valve required less torque to seat fully than the worn
valve. As a result of the inspector's questions, a procedure
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change (TAPA) was submitted, and the licensee's Operations
Standards department is reviewing the issue with engineering to
determine if a permanent procedure change is appropriate. As
the condition was not a one-time special occurrence, the change
did not appear to meet the conditions for.a Special Variance.
This Special Variance was subsequently canceled.
Special Variance 3-92-328 established a vent and drain path for
a system for a normal operational configuration. As the need
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for this lineup is not a one-time occurrence, a TAPA was
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initiated and the Special Variance was canceled.
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Special Variance 3-92-052 addressed actions necessary due to a-
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leaking isolation valve. The inspector determined that no work
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document was active to correct the condition. The licensee-
determined that the condition of the valve did not require any
special actions, and canceled the Special Variance.
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The inspector concluded that the Special Variances have not
always been reviewed to determined if original conditions still
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exist, or if permanent procedure changes are more appropriate.
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The licensee responded by stating that a review of Unit 3
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Special Variances was already in progress when the inspector's
observations were made.
No violations of NRC requirements or deviations were identified.
3.
Surveillance Testino - Units 1. 2. and 3 (617261.
Selected. surveillance tests required to be performed by the Technical
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Specifications were reviewed on a sampling basis to verify that:
1) the
surveillance tests were correctly included on the facility schedule; 2) a
technically adequate procedure existed for performance of the
surveillance tests; 3) the surveillance tests had been performed at the
frequency specified in the Technical Specifications; and 4) test results
satisfied acceptance criteria or were properly. dispositioned.
Specifically, portions of the following surveillances were observed by -
the inspector during this inspection period:
Unit 1
Procedure
Description
PPS Functional Test-RPS/ESFAS Logic
CPC/COLSS Flow Verification
Unit 2
Procedure
Description
Auxiliary Feedwater Pump AFN-P01 Operability
High Pressure Safety ~ Injection Pump Operability Test
Unit 3
Procedure
Description
RCS Water Inventory Balance
No violations of NRC requirements or deviations were identified.
4.
Plant Maintenance - Units 1. 2. and 3 (60705, 62702. 62703 and 62705)-
During the inspection period,'the-inspector observed and reviewed-
selected documentation associated with maintenance. and problem
investigation activities listed below to verify compliance ~with
regulatory requirements, compliance with administrative and maintenance
procedures, required quality assurance / quality control department
involvement,. proper use of safety tags,- proper equipment alignment and
use of jumpers,-personnel qualifications, and proper retesting. The.
inspector verified that reportability for these activities was correct.
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Specifically, the inspector witnessed portions of the following
maintenance activities:
Unit 1
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HJB-F04 Essential Chilled Water Coil Bellows Seal Retest
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74' Control Building Structural Beam Firecoat Repair
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74' Control Building Scaffold Removal
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PM on Security Door J-319
Unit 2
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Preparation of Equipment for Fuel Handling Activities
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Replace UV Device on Reactor Trip Breaker
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Loop Calibration of Steam Generator 1 Level Instrumentation
Unit 3
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CEA Coil Checks 36MT-3SF15
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Replacement of Letdown Relief Valve CHV-345
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Reinstall SGB-PSV-325
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No violations of NRC requirements' or deviations were identified.
5.
Testino of Main Steam Safety Valves - Unit 2 (61726)
The inspector reviewed the-licensee's October 20, 1992, response to a-
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concern presented to the licensee by the NRC in letters dated April 7 and-
September 14, 1992. The concern related to the performance of' test
procedure 73ST-9ZZ18, Main Steam PSV Set Pressure Verification, in Unit-2
in October 1991, during which main steam safety valve (MSSV) 2SGE-PSV-574-
was adjusted'following the first test failure.
From a review of the. test
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documentation, recorded in Work Order 502437, the inspector concluded
that technically the valve was adequately adjusted and tested, with-
satisfactory results.
A note following step 8.'2.10 of procedure'73ST-9ZZl8, Revision 4, PCN 3,
specifically states that "in the event' of an initial test failure, 00 NOT'
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make any adjustments to valve setpoint. After a minimum.15 minute [ sic]
the test may be reperformed.- If results are still unacceptabl_e, .
adjustment of setpoint' is necessary (after 2 or 3 successive out-of--
tolerance lifts), Section 8.3 shall be performed and verified in Appendix
C, page 1 of 2."
Step 8.2.10 provides the direction- to perform t.he test-
until three consecutive lifts have been performed within the allowed
tolerance.
During testing of MSSV 2SGE-PSV-574 on October 8,1991, the valve was.-
tested twice with acceptable'results and the third test 1 failed.. The. .
results of the three tests showed a declining trend. At this point,- the
Test Director initiated actions to re-adjust the valve's setpoint. A.
review of the test results showed that the valve passed three consecutive-
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tests after it was adjusted; therefore, it appears that the valve'was
left in an acceptable condition.
The inspector concluded that the test procedure was not clear regarding
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the limitations on when adjustments to the valve can be made. Adjusting
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after a single failure during the testing of 2SGE-PSV-574 appeared to be
technically satisfactory based on a declining trend in test results and
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with the knowledge that the trend would continue with additional tests.
Statements in the procedure appeared to indicate the need for at least 2
test failures before adjustment of the valve; however, another statement
appears to indicate that the 2 failure criteria applies only to the first
test, if it is a failure. The inspector concluded that the procedure was
unclear and not appropriate to the circumstances; therefore, this appears
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to be a violation of 10 CFR Part 50, Appendix B, Criterion V (Violation
529/92-43-01).
The licensee subsequently revised the procedure to clarify the notes
indicating that no adjustment is to be made to a valve if it fails the
initial (as found) test. This change appears to appropriately clarify
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the intent of the procedure.
One violation of NRC requirements was identified.
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6.
Qualification of Engineerino Personnel - Units 1. 2. and 3 (71707)
The inspector reviewed the qualifications of a sample of engineering
personnel against Technical Specifications requirements and the-Updated
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Final Safety Analysis Report (UFSAR) commitments for qualification.
Technical Specifications Section 6.3.1 requires that each member of the
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unit staff meet or exceed the minimum qualifications of ANSI /ANS 3.1-1978
and Regulatory Guide 1.8, September 1975. The inspector noted that
ANSI /ANS 3.1 requires engineering degrees for only certain engineering
positions.
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The inspector noted that the licensee had upgraded the qualification
requirements for engineering personnel in early 1991, and had already
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completed revisions to the position descriptions. Additionally, the
licensee revised the UFSAR in 1992 committing personnel to meet the
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minimum requirements of ANSI /ANS 3.1 and the qualifications specified in
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the position descriptions. The licensee had recognized that some
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personnel did not meet the more stringent requirements of the revised
position descriptions, and issued memoranda:in July, November, and
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December 1992, documenting a- policy of "grandfathering" those who no
longer met the position description qualifications because of the
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revisions. The policy also -included provisions' for continued, though
restricted, progression for grandfathered personnel- in their current
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career paths. The licensee stated that the intent of the policy was that
eventually all engineering personnel would meet the revised qualification
requirements. However, the inspector noted that the exceptions and
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policy documented in the memoranda were not referenced in the current
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UFSAR, and concluded that the licensee's formal commitment was for all
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personnel to be qualified to the requirements ANSI /ANS 3.1 and their
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position descriptions.
The inspector found that the licensee did not have records to support the
educational and experience qualifications for several people.. The
licensee was in the midst af attempting to document and verify personnel
qualifications.
The inspector reviewed the quclifications of six APS engineering
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personnel whom the licensee had indicated were involved in preparing a
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Technical Specification amendment to change the acceptance criteria for
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tolerance of the setroints-for main steam safety valves, and found that
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all had the minimum educational level required by their position
descriptions. All six had engineering degrees. . Additionally, the-
position descriptions met or exceeded the commitments in the UFSAR and.
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The inspector reviewed the qualification records of a sample of 17
personnel from a list of people who supposedly did not have Bachelor of
Science (BS) degrees in engineering. The position descriptions met or
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exceeded the guidance of ANSI 3.1-1978. This sample consisted of all'
supervisors and above on the list, plus 255s of all other personnel. Of
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the 17 personnel reviewed, 14 personnel did not possess a BS in
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engineering as specified in their position descriptions. These personnel
included one Director, five Supervisors, six Senior Engineers, one
Engineer III, and one Engineer II. The licensee indicated that the
specific educational backgrounds of some of the personnel were being
reviewed for equivalency with a BS in engineering. Also, while the
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inspector did not identify any apparent lack of technical experience,' the
licensee indicated that some individuals had been . identified during its
review who lacked required experience. The inspector concluded. that: '(1)
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the 17 personnel met the qualification requirements of ANSI /ANS 3.1 in
that ANSI /ANS 3.1 did not require engineering degrees for their-
positions, and (2) the failure of the licensee to have all its personnel
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qualified in accordance with their job descriptions was a deviation from
the UFSAR commitment (DEV 528/92-43-02).
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Of the other three people reviewed, one did have a BS in engineering, and
the degrees possessed by the other two individuals met the requirements
of their position descriptions.
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In summary, the inspector concluded:
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Technical Specification requirements for personnel qualifications
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had been met.
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The licensee had . identified and was aggressively addressing
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discrepancies in the qualifications of its personnel in a
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responsible manner prior to the inspector's review.
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That the licensee's recent UFSAR revision committed to increased
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qualification requirements (above that required by Technical
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Specifications) in order to enhance the educational qualification of
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its engineering staff. This resulted in a large number of personnel
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who no longer met the qualification standard as compared to the
previous UFSAR revision.
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The inspector considers the deficieus to be related to meeting
the-commitments of the upgraded job descriptions specified in the
revised UFSAR and not related to meeting the minimum technical
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requirements for engineering personnel.
One deviation ~ from ~a UFSAR commitment was identified.
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Emeroency Liahtino Corrective Action Follow-up - Units 1. 2. and 3
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(92701)
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As part of an on-going review of emergency lighting issues, the inspector
rev.iewed the licensee's corrective actions for a past issue.
Licensee Event Report (LER) 50-528/86-059 (original issue) was issued
when preventive maintenance tests on Emergency Lighting Modular AC Power
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station batteries, for three of four safe shutdown areas, . failed to.
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supply sufficient voltage to emergency lights for an eight hour period.
LER 50-528/86-059, Supplement 1, provided an update' and identified
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corrective actions to be taken to correct the' deficiencies. LER.50-
528/86-059, Supplement 1, was closed by NRC Inspection Report 50-528,
529, 530/87-17.
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To prevent recurrence, the licensee committed to perform annual eight
hour emergency lighting tests to assure that Final Safety Analysis Report-
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commitments were met.
The inspector reviewed licensee 1991, 1992, and scheduled 1993 station
information management system (SIMS) repetitive work tasks for emergency
lighting to follow-up on the licensee's corrective actions.
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Additionally, the inspector reviewed LER 50-528/86-059,. Suyplement 1, for :
accuracy by reviewing licensee supporting closecut documentation',
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emergency lighting (QD) and essential lighting (QB) system descriptions, .
and the Holophane installation and maintenance manual for series 7XX182
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modular AC power station (Publication No. RD7-5319). .
Based on the inspector's additional review of the above documentation,
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the inspector concluded that the' licensee was performing annual eight;
hour tests of emergency lights, and that licensee information and data
presented in LER 50-528/86-059,. Supplement 1, was accurate.
No. violations of NRC requirements or deviatio'ns were identified.
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8.
Fire Watch Proaram Chanoes - Units 1. 2. and 3 (64704)
On January 1,1993, the licensee implemented changes to the Fire Watch
(FW) program, such that the security guards assumed the roving FW
function.
Several guards interviewed by the NRC resident inspectors indicated that
the on-the-job training (0JT) provided by the' licensee to familiarize
them with the FW tour route and responsibilities was not thorough. The.
guards indicated that there was little or no verification by security
supervision of their ability to perform the FW tours. The licensee's
documentation of the OJT appeared to be weak in that it did not indicate
the meaning of the guardsf or supervisors' signatures, leaving it unclear .
whether the guards' signatures docuinented acknowledgement of having been
shown the tour or acknowledgement of being able to perform the tour, and
whether the accompanying supervisors' signatures indicated verification
that the guard was able to do the tour or that the guard was merely shown
the tour. However, all guards interviewed expressed confidence in their
ability to do the tours after practicing them on their own. Following a
meeting with licensee management on December 31, 1992, the licensee
stated that the guards would be asked if they were capable of performing
the tours prior to being assigned that responsibility.
The inspectors observed some fire watch tours and determined that the
guards were able to accomplish the tours. However, an inspector noted
that some of the fire watches did not enter the tour areas sufficiently
to be able to observe the equipment in the rooms, potentially
compromising their ability to detect smoke or fire.- The inspector
observed one auxiliary building tour on January 15, 1993, where the fire
watch did not enter the 77 foot door to the east and west piping-
penetration rooms, the 120 foot coolant purification pipe chase, the low
pressure safety injection pump rooms, the high pressure safety injection
pump rooms, nor the containment spray pump rooms.
In each of these
rooms, the tour check point plaque could be seen from the doorway without
having to enter the room, and the fire watch appeared to consider the
check complete once the plaque was observed without entering the room.
The Manager, Fire Protection, concluded that the piping penetration room-
checks and the pump room checks did not meet management expectations.
In
addition, the inspector considered that procedure 14AC-0FPO4, Firewatch
Duties, step 2.7.1, was not followed in that the areas were not
sufficiently entered to determine whether there was evidence of smoke or-
fi re. This violation is not being cited because the criteria specified
in Section VII.B. of the Enforcement Policy were satisfied (NCV 528/92-
43-03). The licensee addressed this issue by promptly conducting
briefings of all security shifts by either the Manager, Fire Protection,
or the Supervisor, Fire Department. These briefings stated the
expectation for fire watches to view a majority.of the room in such a way
as to be able to reasonably detect any fire or smoke.
The inspectors confirmed that the additional fire watch responsibilities
did not result in fire watch tours or security responses being missed due
to conflicting responsibilities or insufficient time. . Prior to
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implementation, the licensee had the guards perform the fire watch tours
as practice, while conducting security drills, and determined that some
fire watches were missed during the drills. The licensee augmented the
regular security force with extra personnel to provide contingency
coverage to ensure that all fire watch and security requirements were
met, with the expectation that the need for augmentation would diminish
as guards became more familiar with the fire watch responsibilities. The
inspector considered the licensee's actions prudent and adequate.
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On January 6,1993, the inspector observed one fire watch in the Unit 3
auxiliary building during which all fire panels were not checked as
required by procedure 14AC-0FP04, "Firewatch Duties," step 3.3.4.
It was
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noted that two panels which were missed had deficiencies. .The licensee's -
Quality Audits and Monitoring department had previously identified that
the fire panels were not being checked, and issued a Quality Deficiency
Report. According to several fire watch personnel, there had been
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conflicting direction from supervision regarding whether the fire panels
were to be checked or not. The licensee instituted programmatic changes -
and deleted the requirement from the procedure. The observation of the
missed panels was made while the post orders and procedures required
checking the panels. The inspector concluded that failure to check the
panels was a violation of the licensee's procedures, and considered this
an example of a weakness in the preparation and training for the changes
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in the fire watch program. This licensee-identified violation is not
being cited because the criteria specified in Section VII.B. of the
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Enforcement Policy were satisfied (NCV 530/92-43-04).
The inspector concluded that the implementation of the changes to the
fire watch program was not thorough and could have been strengthened in
the minor deficiencies that were observed. The inspector encouraged the
licensee to continue to closely monitor implementation of the fire watch
program.
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Two non-cited violations of NRC requirements were identified.
9.
Review of Independent Safety Engineerino (ISE) Assessment 92-24 - Units
1. 2. and 3 (35702 and 40500)
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The inspector reviewed the Executive Summary for ISE Assessment 92-24,
Shutdown Safety Topical Report for 3R3. The Assessment report was issued
December 21, 1992. The goal was to incorporate recommendations from this
Assessment prior to the next refueling outage. However, as the
Assessment progressed during the outage, completed assessment elements
were issued to provide feedback to management in a timely manner.
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ISE identified several strengths, areas for improvement, and deficiencies
during the Assessment. .These were clearly presented in the Executive
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Summary. Weaknesses observed included poor tracking of lessons learned,
no tracking for recommendations of the shutdown risk assessment, and
informality of controls for containment penetrations to address shutdown
risk assessment recommendations. A Quality Deficiency Report was
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initiated documenting that some relays were calibrated under a preventive
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maintenance task instead of a surveillance test.
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Responsibility for_ addressing _ recommendations resulting from the
Assessment were assigned in the Assessment report. The inspector
discussed the status of recommended actions'with the licensee and
concluded that reasonable progress was being made to address the
recommendations.
The inspector concluded that the ISE Assessment wd:: adequate and
contributed to improvement of shutdown safety.-
No violations'of NRC requirements or deviations were. identified.
10. Overfill of Cable Travs (92701)
The inspector examined safety-related cable traps at Palo Verde that had
been identified, by the licensee, as being overfilled. To evaluate the
impact of an overfill condition, cable derating and cable tray support-
calculations were reviewed by the inspector,
a.
Cable Deratina
Cable derating is performed to ensure that operating temperatures of
power cables in overfilled cable trays do not exceed their
insulation design criteria of 90 degrees Centigrade.
-The inspector reviewed approximately 100 individual licensee cable-
tray calculations for "Derating of Cables in Trays," Calculation No'.
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13-EC-ZA-300. The reviewed calculations of overfilled cable trays
also included cable trays that are covered or had fire stops, or a
combination of all three conditions. Each condition, overfill, .
covered, and fire stops, contribute to the derating of cables,
therefore, approximately 40% of the cable trays sampled had all
three of the conditions.
The inspector did not identify any cable' tray that exceeded its
maximum watts / foot, which would-have caused the cables to exceed
their insulation design criteria value of 90 degrees centigrade.
The inspector concluded that the overfilled cable trays sampled for
derating did not exceed their insulation design criteria..
b.
Cable Trav Supports
cable tray support loading is required to be within the' limits of
the Updated Final Safety. Analysis Report (UFSAR) and IEEE 344-1978,.
" Recommended Practices for Seismic Qualification of Class IE -
Equipment for Nuclear Power Generating Stations."-
The inspector reviewed licensee data ba'se EE580, " Cable and Raceway
Tracking System," and the licensee identified cable tray' support
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analysis, to determine if any cable trays or cable tray supports
exceeded their design limits and their UFSAR commitments to IEEE 344-1978. The inspector sampled and. reviewed eight cable tray'-
supports for 25 cable trays which had overfill conditions in at
least one of-the cable trays that they-supported.
The inspector did not identify any cable tray supports which
exceeded their design load capacities. Therefore, the inspector
concluded that the cable tray supports reviewed did not exceed their
design limits.
The licensee stated that the Palo Verde cable derating-calculation and
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the cable tray support calculations were scheduled to be re-evaluated in
the design basis and calculation reverification program. These
calculations are scheduled for the 1993/1994 reveri fications.
No violations of NRC requirements or deviations were identified.
11. Appendix R Fire Door Problems (92701)
Palo Verde has a self-initiated fire door monitoring program in place to
identify Appendix R and non-Appendix R fire door problems.
The licensee's fire door inspection program included: fire department
inspections (performed daily) using procedure 14FT-9FP24~, " Fire Door
Position Verification,"; engineering inspections (performed every six ~
months) using procedure 14FT-9FP61,
" Operational Check of Appendix R -
Fire Door Closures,"; and an inspection (performed'every 18 months) using
procedure 14FT-9FP62, " Appendix R Fire Door Functional Test."
To evaluate the program the inspector reviewed . licensee material _ non-
conformance reports (MNCR's) for Appendix R fire- doors from November 1990
to November 1992.
Review of MNCR's found that for the three' units a'
total of 59 MNCR's were written for Appendix R fire doors during this
time period. The inspector reviewed six of the MNCR's.
The inspector found that:
Each MNCR could cover more than one Appendix R fire door within the
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same area.
The majority of fire door problems were for doors _ located in high
traffic areas, which would result in a higher incidence 'of, wear
induced repair problems.
The majority of the deficiencies reviewed'did not have an effect on
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the doors ability to: perform its function.
Examples of problems
were: fire door had more than three signs attached to it, signs on
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fire doors were incorrectly attached, doors were missing nameplates ~,
incorrect caulking was installed around door, and door adjustments
were incorrect.
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The inspector concluded that the licensee had an aggressive fire door
inspection program, had been aggressive in addressing Appendix R fire
door problems, and had made timely repairs and corrections to fire door
problems.
No violations of NRC requirements or deviations were identified.
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12. Qualification of Electrical Cable in Containment Penetration Area (92701]
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The inspector examined the licensees' equipment qualification enhancement
project. The project resulted from a licensee self-initiated equipment
qualification assessment.
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The inspector reviewed the following licensee Equipment Qualification
Program procedures for cables in containment penetration areas to
determine if cables installed in containment penetration areas were
quali fied:
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13-EM-035A
Electrical Penetrations
13-EM-058
600 Volt Power Cables, "Rocbestos"
600 VAC Power Cable
600 VAC Power Cable with KXL-420 Insulation Rework
600 VAC Power Cable with Fire Wall III KXL-760D XLPE
600 VAC Irradiated XLPE KXL-760G
P
13-EM-057
600 Volt Control Cables
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The cables examined by the inspector were found to be qualified in
accordance with the qualification program acceptance criteria.
In addition the inspector confirmed that as part of a licensee equipment
qualification enhancement project the licensee has included walkdowns of
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power and control cables in the containment and containment penetration
areas. The inspector reviewed eight "PVNGS Generic EQ Walkdown Data
Sheets" for Unit 3.
The inspector did not identify any anomalies or
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problems with the walkdown documentation.
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The inspector concluded that the cables in containment penetrations and
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containment penetration areas were qualified in accordance with the
equipment qualification acceptance criteria, and that the licensee was in
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the process of reverification of cable equipment qualification.
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No violations of NRC requirements or deviations were identified.
13. Atmospheric Dump Valve Post-Restart items Follow Up (92701)
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Licensee letter number 102-01294, dated June 8, 1989, from Mr. Conway,
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APS, to Mr. Martin, NRC Region V, outlined Palo Verde Units 1, 2, and 3,
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post restart items resulting from NRC Augmented Inspection Team (AIT).
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Report 50-528, 529, 530/89-13, March 3, 1989 and licensee' incident
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investigation report (IIR) 2-3-89-001, for Atmospheric Dump Valves.
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The inspector reviewed the licensee's atmospheric dump valve " Post
Restart Action Items" package. The inspector found that of the
approximately 800 individual action items, three items remain open.
These three items were record item numbers 522, 778, and 779, covered
under design change package (DCP) 1, 2, 3 FE-MA-064. Design change
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package.1, 2, 3, FE-MA-064 will permanently install equipment necessary
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to record subsynchronous oscillation (SS0)- relay quantities '(i.e.
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frequencies) on the digital fault recorder and-to allow' the display of
only the target (i.e., value at which the SSO initiated)' initiated by an
SSO relay operation. Design change package 1, 2, 3 FE-MA-064 is
presently scheduled for completion, in all three units, by the end'of
refueling outage 3R4 (April 19,1994).
The inspector concluded that the licensee had three open items left to
complete the post restart action items for the ADVs. The responsible
licensee manager stated that any changes to the schedule would be
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provided to the NRC.
No violations of NRC requirements or deviations were identified.
14. Temperature Compensation and Corrective Actions for Atmospheric Dumo-
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Valve Nitrocen Accumulator Pressure Dron Test (92701)
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NRC Inspection Report 50-528, 529, 530/90-23 identified that the licensee
had taken corrective action to correct temperature compensating formulas
in surveillance test procedure 41ST-1SG05, "ADV Nitrogen Accumulator Drop .
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Test "
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Surveillance test 41ST-ISG05 is performed to verify that nitrogen leakage '
from the ADV nitrogen accumulator will not drop pressure below Technical
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Specifications Section 3.7.1.6 limits, when the. accumulator is isolated-
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from the nitrogen and instrument air supply headers.
Below the Technical Specifications Section 3.7.1.6 limit of 400 psig, the ADV will be
inoperable since there may not be sufficient nitrogen accumulator-
pressure to operate the ADV per design conditions as stated in UFSAR
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paragraph 10.3.2.2.4.
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The inspector reviewed licensee Engineering Evaluation. Report (EER) 90-
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59-139 for temperature compensation in evaluating "ADV Nitrogen
Accumulator Drop Test,"l41ST-ISG05. Calculational temperature
corrections were required to accurately compensate.for pressure changes,
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in the ADV nitrogen accumulators, resulting from changes in ' ambient
temperatures during a two hour ADV nitrogen. pressure drop tests.
Additionally the inspector reviewed a listing of Condition
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Report / Disposition Requests (CRDRs), Quality Deficiency Reports-(QDRs)',
and Material Non-conformance Reports (MNCRs), for theLatmospheric dump
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valves. The inspector.noted that of the 31 deficiency documents issued
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against'the ADVs, for the' period from December 1990 to December 1992, 13
items were associated with the nitrogen pressure drop test. The safety
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analyses for Palo Verde take credit for two of the four ADVs. -The
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inspector also noted that licensee corrective actions have been
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successful in reducing the number of ADV nitrogen drop test failures over
the period examined. Corrective' actions that have been -implemented to
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reduce test failures are:
Addition of a double check valve down stream of the nitrogen header
isolation valve to assure that the nitrogen accumulator is
constantly pressurized, site modification SMSG-0025,
Addition of a filter in the nitrogen supply line_ to the nitrogen
accumulator to ensure nitrogen is filtered to 3 microns (absolute),
Correction of nitrogen regulator (from the accumulator to the ADV
control and operation) calibration as left acceptance criteria
(preventative maintenance tasks Change S123-JSG-002).
The licensee stated that the number of problems with the nitrogen system
for the atmospheric dump valves had been identified and was-being tracked
under CRDR 920329 for evaluation. CRDR 920329 was scheduled for
completion by March 5,1994.
The inspector concluded that EER 90-59-139 accurately evaluated
temperature compensation for the ADV nitrogen accumulators, that the
correct temperature compensation formulas have been implemented in 41ST-
ISG05, 42ST-2SG05, and 43ST-ISG05, that filters had been added to the
nitrogen supply system to preclude foreign material clogging, and that
subsequent ADV pressure drop test problems were being identified and
tracked in accordance with licensee procedures.
No violations of NRC requirements or deviations were identified.
15. Corrective Action Report (CAR) 90-0010 (92701)
Licensee corrective action report (CAR) 90-0010 identified that certain
administrative controls had not been sufficient to ensure that equipment
design output information was being incorporated into plant configuration
documents (information ia the station information management system
(SIMS) equipment database was not being translated'into the preventative
maintenance program). This problem resulted in preventative maintenance
(PM) work orders (WO) potentially being issued and work ~ being performed
at a lower quality classification than that required for the component.
In CAR 90-0010, the licensee stated that equipment databases in SIMS had
been updated from the equipment database to show proper quality group
classifications and that tasks would be routinely updated to i .clude
correct quality classifications. CAR 90-0010 was closed on August 8,
1991 indicating that the corrective actions were complete and that SIMS
task items had been updated from updated equipment databases. The
inspector verified that the licensee's verification efforts for CAR 90-
0010 had been completed.
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The inspector. reviewed a-listing of equipment in the SIMS program which
did not have a quality classification listing. ' The listing indicated '
that approximately 3300 task items in SIMS did not have quality
classification listings. The inspector found that approximately 10% of
the 3300 task items which did not have a quality classification were
listed as being quality related work items. Quality classification and
quality related work items appear as two independent fields on the SIMS.
The inspector concluded that either:
The licensee had not performed an adequate QA check of quality
classifications in SIMS prior to closure of CAR 90-0010, or,
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Licensee corrective actions to continuously update quality
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classifications from the equipment database were not always being
completed.
In response to the inspector's findings the licensee issued Condition
Report / Disposition Request (CRDR) 92-0745 and stated that they would
perform an evaluation to determine if any preventative maintenance or
work was performed at a lower quality level than the equipment
classification. This item will be followed as Unresolved Item 50-
528,529,530/92-43-05).
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No violations of NRC requirements or deviations were identified.
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16. Followup on Previously identified items - Units 1. 2. and 3 (62700 and -
92701)
a.
Unit 1
(1)
(0 pen) Followup Item 528/92-41-01: Reactor Trio due to Known
Defective Neoative Seauence Relay - Unit 1 (92701 and 62700)
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This item involved the failure of a negative sequence relay due to a-
known defect which resulted in a reactor trip. The known defect was
published in GE Service Advice Letter (SAL) 189.1 dated October-1,
1987, and had been-issued to the licensee on October 12, 1987. Four
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questions were asked in inspection report 528/92-41-01. .The answers
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to these four questions were addressed during a discussion with
personnel from the Vendor Engineering Group on January 7,1993.
During this discussion the inspector was notified that the vendor
technical manual (VTM) consolidation had already occurred for VTM
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H045-0001 which included the negative sequence relay, and yet GE SAL
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189.1 was not in the consolidated VTM. The licensee stated two
reasons why GE SAL 189.1 was not in the newly consolidated VTM. The
first was due to internal administrative errors which prevented the
initial SAL from being incorporated.into the technical manual when a
design change installed this model negative sequence relay. .The
second resulted when the vendor was re-contacted during the
consolidation effort in September 1991, and the GE representative
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did not identify SAL 189.1 as applicable to this model negative
sequence relay.
The licensee further stated during the discussion on January 7,
1993, that during the evaluation following the reactor. trip on
December 8,1992, the licensee's initial contact with the. vendor
resulted in an answer that no applicable vendor information related
to the negative sequence' relay. . Subsequent telephone calls to the
vendor identified four applicable SALs.
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The answers to the four questions raised in NRC Inspection Report
50-528/92-41 are as follows:
- The VTM consolidation program attempts to capture all SALs and
does incorporate those that the VTM program is aware of where
applicable;
- Not all critical systems are currently being incorporated; 75% of ~
all quality (Q) class critical system components, 55% of all quality
augmented (QAG) class critical components, and 40% of all not
quality related (NQR) class critical system components are currently
complete. This represents all the critical system components that
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can be addressed by'the VTM consolidation program without
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identifying and verifying more vendor 'model numbers _ for installed
plant equipment. For all systems, the current VTM consolidation
effort encompasses 90% of all Q class components, 60% of all QAG-
class components, and 60% of all NQR class components. The licensee-
is evaluating the need and effort required to complete more of this
program and is scheduled to complete this evaluation by March 1,
1993;
- The current Vendor Engineering Program addresses equipment
deficiencies identified by the vendor, and Condition
Report / Disposition Request (CRDR) 9-2-0743 addresses future design
changes to ensure that all available vendor information -is addressed -
prior to the installation of the design change;
- The VTM consolidation program does not verify installed equipment.
A model number verification program is in progress to verify the
Station Information Management System (SIMS) model number with the
design documents. The licensee determined that the effort required
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to verify SIMS model numbers with installed plant equipment was not
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justified. The plant construction was based on design documents,
and the model number verification has assured that the SIMS model
number matches the design documents. Each work order. compared the
installed equipment model number with the SIMS model number and very.
few discrepancies have been identified. Vendor information that was
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in the old VTM prior to the consolidation was assumed to have been
evaluated, but any vendor 'information that was not in the old VTM
was evaluated for applicability and corrective action as needed.
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The inspector. concluded that the licensee's answers to these
questions appeared appropriate. The inspector further concluded
that two areas of vulnerability existed in the present VTM. program.
The first area involved licensee internal communication to ensure
that vendor information'was appropriately identified and
incorporated into the VTM. The second area involved the quality of
information obtained from vendors, particularly during the vendor
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re-contact every three years. The inspector noted that the first
area of vulnerability was being addressed using training, meetings,
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and other forms of internal communications. The inspector
acknowledged that the second area of concern was not fully under
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licensee control, but to the extent that the licensee does have
control, the inspector encouraged the licensee to exercise it with -
vendors to ensure that the best information is available.
The
inspector asked the Supervisor, Vendor Engineering, what corrective
action is planned to address the vendor information-related concerns
raised by this reactor trip event. The Supervisor responded saying
the corrective action will be finalized by March 1,1993.. This item -
will remain open pending a review of these corrective actions.
(2)
(Closed) Followuo Item 528/90-25-03: Failure to Perform an Annual
Review of Palo Verde Pre-Fire Strateoies Manual
NRC Followup Item 528/90-25-03 identified that the licensee had
apparently failed to perform an annual review of the Palo Verde Pre-
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Fire Strategies manual (PFSM) in accordance with Palo VerdeL
administrative procedure 01AC-0AP02 in 1988.
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In response to this concern the licensee provided APS internal
memorandums, 229-00048, dated February 16, 1988, requesting- ..
engineering review of the pre-fire strategies manual, and 167-02213,-
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dated April 18, 1988, documenting engineering's completion of the
review of the pre-fire strategies manual.
The inspector reviewed the licensee memoranda, noted above, and
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determined that review of the Palo Verde pre-fire strategies manual-
had been completed in accordance with Palo Verde administrative,
procedure 01AC-0AP02.
In addition, the inspector reviewed licensee
records that documented licensee reviews of the Palo Verde pre-fire
strategies manual in 1989, 1990, 1991, and 1992. The licensee
stated that the annual review of the Palo Verde pre-fire strategies.
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manual is being tracked by corrective action report (CAR)90-014.
The inspector concluded that the licensee had. performed the complete
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annual review of the Palo Verde pre-fire strategies manual, in 1988,-
and was continuing to perform a yearly review in accordance with
administrative procedure 01AC-0AP02. This item is closed.
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(3)
(Closed) Followup Item 528/90-25-04:' Failure to Provide Oualified-
-Outdoor Emeroency Lichting to Support Safe Shutdown Operation:
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NRC' Followup Item 528/90-25-04 identified that the licensee had
failed to provide emergency lighting of approved design for outdoor
use to support safe shutdown. .NRC inspectors found that Emergi-Lite
units, installed in the Palo Verde' main steam support structure -
(MSSS) breezeway were not tested and approved for outdoor, " wet or
damp" locations, as required by Article 410-4'of NFPA 70-1975 and
the facility Operating License.
Licensee documentation during the inspection period of. Inspection
Report 50-528/90-25 stated that the Emergi-Lite units were
acceptable for outdoor use in accordance with NRC Branch Technical
Position (BTP) 9.5-1 and the National Electric Code. _NRC inspector
review of the documentation found that licensee documents did note
support the qualification the Emergi-Lite units for outdoor use in
accordance with (BTP) 9.5-1 and the National Electric Code.
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In response to the finding, and the NRC inspector's interpretations,
the licensee committed to replace the Emergi-Lite Units with
qualified Holophane Modular Power Stations (MPS) and fluorescent
lite fixtures.
Installation of the modular power stations and the
fluorescent lites was completed by August 1990 under. design change
package (DCP) 1, 2, 3, FE-QD-25.
Installation of the MPS lighting
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fixtures was verified by an NRC inspector and documented in NRC
Inspection Report 50-528, 529, 530/91-30.
The inspector reviewed licensee design change package, DCP 1, 2, 3,;
FE-QD-25, and concluded that the licensee had installed outdoor
qualified lighting in Units 1, 2, and 3 MSSS breezeways.
The inspectar concluded that, based on the inspector's review of DCP:
1, 2, 3, FE-QD-25, and the verification of . installation of qualified
emergency lighting in the MSSS breezeway. this' item is closed.
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(4)
(Closed) Unresolved item 528/91-30-02: Root Cause of Battery
Failures
NRC Unresolved Item 50-528/91-30-02' identified an NRC concern that
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the licensee may not have been performing timely engineering-
evaluations for licensee identified battery deficiencies, of
Holophane and Exide emergency lights, after. corrective actions to
upgrade and improve the emergency lighting' systems. -The corrective-
actions were taken in response to previous NRC identified
deficiencies in qualification of emergency lights.
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In response to the concern, the licensee provided the NRC: a
licensee generated " Emergency / Essential Lighting Systems Asr.9ssment
Report," dated February 1992; " Calculation for Holophane Emergency
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Lighting Unit Availability Study," 13-NC-QB-200; and " Calculation
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for Exide Emergency Lighting Unit Availability Study," 13-NC-QD-201. -
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The above documents provided information for calculated known and
estimated unavailability for lights, calculations of unit'(emergency
lighting) and plant availabilities, and work order graphs for"
emergency lights. The calculations and. assessment concluded that~
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availability of Holophane units increased from 93.8 percent to' 97.9
percent (after site modification 1, 2, 3, SM-QD-007 upgraded
Holophane unit capacities), and that Exide inverter-availabil.ity
increased from 77 percent to 82.7 percent.(after Design Change
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Package 3-XE-QD-026 replaced deficient batteries). There was a
trend of improved performance for the lighting units. The licensee
was monitoring Exide inverter availability under Condition
Report / Disposition Request-(CRDR) 9-2-0012 to further improve
performance.
The inspector concluded, based on the NRC review and evaluation of
licensee calculations and assessment, that the licensee was
monitoring, trending, and performing engineering evaluations 'of
emergency lighting failure rates and availability, and that
availability of emergency lighting units was increasing. - The
licensee was reviewing ways to further improv'e performance. This
item is closed.
b.
Units 1. 2. and 3
(1)
(Closed) Followup Item 528/92-17-02: Feedwater Isolation Valve
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(FWIV) 4-way Valve Failures - Units 1. 2. and 3 (92701)
This item addresses the root causes of failure (RCF) and'the safety
significance of internal oil' leakage resulting from- the recurring .
failures of FWIV 4-way valves due to o-ring problems. -The inspector
reviewed Condition Report / Disposition' Request _(CRDR) .1-2-0393, which
documents the licensee's evaluation of failures of 4-way_ valves for-
FW1Vs 3JSGAUV0174 ("M" valve), IJSGBUV0132 ("M" valve),- and:
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2JSGAUV0174 ("N" valve).
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The licensee determined that a common--deficiency in all these
failures was that-the clearances:between the body seal port cavity
and the seal in'the'"P" ports was excessive. . This was the root
cause of failure (RCF) in two of the failures.
In the 4-way valve
for IJSGBUV0132, the RCF was'the existence of a depression on the
body surface, causing insufficient compression of the 0-rings on ;the
bottom metal surface, allowing undesired bypass flows. The licensee -
obtained concurrence _on the RCF from the supplier, Anchor-Darling.
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Based on the identified RCF, the licensee analyzed all. installed 4-
way valves and determined that only _four were at risk because
insufficient data was available to support their longevity and/or-
acceptability.
(Those valves which had undergone the enhanced
dedication process implemented by Anchor-Darling in 1990 were not~
considered at risk.) Because o-ring failures of valves in the "N"
application would be immediately annunciated in the control room,
only the one "at risk" valve in the."M" application was immediately
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replaced. . Actions were initiated to check ~ all 4-way. valves- in
storage'and in service at the earliest opportunity.
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The licensee also evaluated the safety significance .of the internal
leakage' associated with the failures. .The licensee determined that
the "M" valve failures would result in the inability of the valves
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to fast close to perform their safety ~ function, and therefore
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considered the failures safety-significant. However, the redundant
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isolation valves in series with the' failed FWIVs were operable,. so -
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that the safety function was not-lost.
Additional corrective actions initiated by the licensee included
followup with the supplier regarding 10 CFR~ Part 21 notifications,.
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developing a critical dimension list for use in dedication of the
valves, initiating a. comprehensive disassembly, inspection, and .
reassembly instruction, and reassessing Anchor-Darling's dedication
program.-
Based on this review of the licensee's evaluatics and corrective
actions, this item is' closed.
7
No violations of NRC requirements or deviations _were identified.
-<
17.
Review of Licensee Event Reoorts (LER) - Units 1 and 2 (92700 and 92712)
Through direct observations, discussion with licensee personnel, or
review of-the records, the following LERs were closed.-
a.
Unit 1
-
(1)
92-05,
Revision L1
" Technical. Specification ' Action Missed
While Containment Isolation Valve
!
LER 528/92-05, Revision LO, reported that the licensee had.
found Unit I check valve SIE-V133. incorrectly assembled. .The
licensee determined that the bonnet and disc. assembly had been
,
installed backwards.
.
An inspector determined, as~ noted in NRC Inspection Report 50 -
,
528/529/530/92-23, that LER 528/92-05, Revision LO, did not
'
assess associated Technical Specification and ASME Section XI
requirements, did'not provide any required corrective actions,
and incorrectly stated that the problem with SIE-V133 was an
isolated occurrence.
l
,
The licensee issued LER 528/92-05, Revision L1, to address the
inspector's concerns. Revision L1 addressed Technical
Specification and ASME Section XI requirements, immediate and
followup' corrective' actions, and identified .similar check valve
assembly problems.
22
!
,
_
.;
.
.
The inspector reviewed Revision L1, discussed long range
corrective actions with 'the licensee's check valve engineer,
and reviewed maintenance records for similar check valves.
The inspector did not identify any maintenance records
indicating additional check valve problems similar to SIE-V133.
The inspector concluded that the licensee had adequately
addressed the NRC concerns with LER 528/92-05, Revision LO,
based on review of LER 528/92-05, Revision L1; discussions with
the licensee's check valve engineer; and review of maintenance
records. Based on the above, LER 528/92-05, Revision L1 is
closed.
(2)
92-16,
Revision LO
" Reactor Trip caused by Main Turbine
Trip"
This event was reviewed in NRC Inspection Report 50-528/92-41,
.
paragraph 7.
The only additional issue raised by the review of-
,
this LER was the fact that the vendor technical manual (VTM)
!
consolidation project for the negative sequence relay was
_
!
complete on September 30, 1991, and issued as VTM H045-0001.
At the time of the trip, this VTM did not contain General
t
Electric (GE) Service ,tivice Letter (SAL) 189.1 as discussed in
paragraph 10.a.(1) of this report nor several other applicable-
GE SALs. This fact was not addressed in this LER.
Based on
the above and the continuing review of inspection followup item
.
!
528/92-41-01, this LER is closed.
b.
Unit 2
(1)
92-05,
Revision L1
" Missed Technical Specification
Surveillance for Charging Pump (ASME)"
,
(2)
92-06,
Rev'ision LO
" Reactor . Trip During CEDM MG Testing"
No violations of NRC requirements or deviations were identified.
-!
18. Exit Meetinq (71707)
An exit meeting was held on January 20, 1993, with licensee management
and the NRC resident inspectors during which the observations and
conclusions in this report were generally discussed.
The licensee did not identify as proprietary any materials provided to or
reviewed by the inspectors during the-inspection.
.-
23
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L_