ML20211N890
| ML20211N890 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 07/02/1986 |
| From: | Harrell P, Hunter D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20211N872 | List: |
| References | |
| 50-285-86-14, NUDOCS 8607110067 | |
| Download: ML20211N890 (8) | |
See also: IR 05000285/1986014
Text
APPENDIX 8
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report:
50-285/86-14
License:
Docket:
50-285
Licensee:
Omaha Public Power District
1623 Harney Street
Omaha, Nebraska 68102
Facility Name:
Fort Calhoun Station
Inspection At:
Fort Calhoun Station, Blair, Nebraska
Inspection Conducted:
May 1-31, 1986
Inspector:
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[ g 'InspectorP. H. R&rr'ellf' Serptfr Resident ReactorDate /
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Approved:
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R. Hunt er, Chief, Project Section B,
Date
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eactor) Projects Branch
Inspection Summary
Inspection Conducted May 1-31, 1986 (Report 50-285/86-14)
Areas Inspected:
Routine, unannounced inspection of operational safety
verification, maintenance, surveillance, plant tours, safety-related system
walkdowns, followup on previously identified items, followup on a licensee
event report (LER), and followup on a 10 CFR Part 21 report on Valcor valves.
Results:
Within the eight areas inspected, one violation was identified
(failure to maintain cable trays in accordance with design documentation,
paragraph 6).
8607110067 860703
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DETAILS
1.
Persons Contacted
- W. Gates, Plant Manager
C. Brunnert, Operations Quality Assurance Supervisor
M. Core, Maintenance Supervisor
D. Dale, Quality Control Inspector
J. Fisicaro, Nuclear Regulatory and Industry Affairs Supervisor
J. Foley, I&C and Electrical Field Maintenance Supervisor
M. Kallman, Security Supervisor
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- L. Kusek, Operations Supervisor
- T. McIvor, Technical Supervisor
R. Mueller, Plant Engineer
- G. Roach, Chemical and Radiation Protection Supervisor
J. Tesarek, Reactor Engineer
- S. Willrett, Administration Services and Security Supervisor
- Denotes attendance at the monthly exit interview.
The inspector also contacted other plant personnel, including operators,
technicians, and administrative personnel.
2.
Followup on Previously Identified Items
(Closed) Open Item 8425-01:
Isolation capability of the feed regulating
valve bypass valve.
This item is closed based on actions taken by the licensee.
See
paragraph 3 for a detailed discussion of this item.
(Closed) Unresolved Item 8526-01:
Component cooling water pump A piping
vibration.
The licensee has installed wooden wedges between the piping and the piping
floor penetration to minimize the vibration.
Subsequent to installation
of the wedges, the licensee took readings to verify that the vibration
levels were within the allowable limits.
The NRC inspector reviewed the
results of the vibration readings taken by the licensee and confirmed that
all readings were within specifications.
Readings were also taken for the
B and C pumps, and no problems were noted.
The licensee has performed an
engineering evaluation to verify that the wooden wedges installed adjacent
to the piping will not adversely affect system operation.
3.
Licensee Event Report Followup
Through direct observation, discussions with licensee personnel, and
review of records, the following event report was reviewed to determine
that reportability requirements were fulfilled, immediate corrective
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action was accomplished, and corrective action to prevent recurrence had
been accomplished in accordance with Technical Specifications (TS).
LER 85-009 described a circuit that did not comply with the single failure
criteria for plant protection systems. The circuit was designed to shut
the main steam isolation valves in the event of a main steam line break.
During a review of this circuit by the licensee, the licensee noted that a
single relay provided the signal to shut both main steam isolation valves.
In the event of a main steam line break concurrent with the failure of the
relay to assume its accident position, the automatic closure of both main
steam isolation valves would not occur.
The licensee has modified the
logic to eliminate the single failure concern and has also reviewed other
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modifications designed during the time frame of this design to verify no
other safety-related systems were installed using relays subject to a
. ingle failure.
The result of the licensee review indicated that no other
designs contained the same type of problem.
The NRC inspector reviewed
the documentation associated with the licensee activities.
No problems
were noted and we have no further questions of this matter at this time.
>
No violations or deviations were identified.
4.
Followup on a 10 CFR Part 21 Report on Valcor Valves
In a letter to the NRC dated March 17, 1986, the licensee reported that
problems had been encountered with the operation of 2-inch, Series V526
Valcor valves.
The licensee noted that two valves in the charging and
volume control system would not operate on an open signal.
The licensee performed an inspection of the valves and found that the disc
guide assembly springs had failed.
When maintenance personnel opened the
valve, the springs were found to be in numerous pieces.
Laboratory
testing found that the springs had failed due to hydrogen embrittlement.
The licensee replaced the springs and the valves tested satisfactorily.
The replacement spring used to repair the valve was the same type of
spring that had originally failed in the valve.
The licensee performed an
engineering evaluation and determined that the same spring will allow
proper operation until the next refueling outage currently scheduled for
March 1987.
The licensee intends to replace the springs during the
outage.
The licensee inspected other Valcor valves to determine if the spring
problem exists in another system.
No other problems were noted with the
Valcor valves.
Based on inspection of the valves and the results of the
failure analysis, the licensee has concluded that the failure will most
likely occur in an application where the valve will contact primary
coolant.
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Based on the documentation reviewed and discussions with licensee
personnel, it appeared that the licensee has taken the appropriate
corrective action for ensuring that all potential valve failures were
addressed.
No violations or deviations were identified.
5.
Operational Safety Verification
The NRC inspector conducted the reviews and observations of selected
activities to verify that facility operations were performed in
conformance with the requirements established under 10 CFR, administrative
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procedures, and the TS.
The NRC inspector made several control room
observations to verify:
Proper shift staffing.
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Operator adherence to approved procedures and TS.
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Operability of reactor protective system and engineered safeguards
equipment.
Logs, records, recorder traces, annunciators, panel indications, and
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switch positions complied with the appropriate requirements.
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Proper return to service of components.
Maintenance orders initiated for equipment in need of maintenance.
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Appropriate conduct of control room and other licensed operators.
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No violations or deviations were noted.
6.
Plant Tours
The NRC inspector conducted plant tours at various times to assess plant
and equipment conditions.
The following items were observed during the
tours:
General plant conditions.
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Equipment conditions, including fluid leaks and excessive vibration.
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Plant housekeeping and cleanliness practices including fire hazards
and control of combustible material.
The physical security plan implementation in accordance with the
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station security plan.
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Adherence to the requirements of radiation work permits
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Work activities performance in accordance with approved procedures
During tours of the auxiliary building, the NRC inspector noted that
cables and cable trays were not being maintained in accordance with basic
installation and design documentation.
The following discrepancies were
identified:
a.
Drawing 11405-E-60, " Reactor Auxiliary Building Tray Conduct Layout
Plan," required in Note 17 that solid covers be installed on cable
trays.
The NRC inspector noted that trays in various locations in
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Rooms 69 and 57, and in the east and west switchgear rooms did not
have the tray covers properly installed.
The cables in the trays
contained safety-related circuits.
b.
Drawing 11405-E-151, " Cable and Conduit Schedule Notes," required in
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Note 18 that control and instrument cable be tied down in a neat
configuration after installation in trays.
The NRC inspector noted
that safety-related control and instrumentation cables installed in
vertical trays EA 5-4 and EB 5-4 were not tied down.
Without being
secured, the cable was protruding into the personnel pathway causing
a potential problem with snagging the cable.
c.
Drawing 11405-E-151 also required in Note 20 that 600-volt power
cable not exceed a 40 percent cable tray fill criteria.
The NRC
inspector noted that the cable exceeded this limit at tray
location 21-5.
These discrepancies are examples of the failure to maintain cable tray
installations in accordance with design and installation documentation.
This is an apparent violation.
(285/8614-01)
As a result of this NRC inspection finding, the licensee has initiated
walkdowns of the plant safety-related cable trays to verify that the
installation was in accordance with the requirements.
The control and
instrumentation cable in trays EA 5-4 and EB 5-4 has been secured to
prevent snagging.
The NRC inspector also noted during plant tours that manual valves in
branch lines between the outside of the containment wall and the first
automatic isolation valve were not locked.
The valves were located in
branch lines in the component cooling water supply and return lines for
the nuclear detector well cooling units.
The NRC inspector notified the
licensee of the situation and discussed the status of containment
isolation as required by Criterion 57 of Appendix A to 10 CFR 50.
The
licensee locked the two valves in question as a prudent action while
reviewing the status of the applicability of Criterion 57 for the
Fort Calhoun Station (FCS).
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Subsequent to this discussion with the licensee, tne NRC inspector also
noted that valves in branch lines between the outside containment wall and
the first isolation valve in the chemical and volume control system were
not locked.
The NRC inspector discussed the applicability of 10 CFR Part 50, Appendix A,' Criterion 55 with the licensee.
The licensee locked
the valves with circular handwheels.
Due to there being no convenient
method for locking valves with T-handles, the licensee performed an
evaluation and determined that a shut valve with an installed cap was
sufficient.
This action was taken because the licensee felt it was the
prudent thing to do while reviewing the applicability of Criterion 55 for
the FCS.
The licensee has yet to determine whether or not 10 CFR Part 50,
Appendix A, Criteria 55 and 57 is applicable to the FCS.
The NRC
inspector discussed this item with the NRR Project Manager (PM) for the
FCS.
This item will remain unresolved pending further review by the NRC
to determine the specific applicability of Criteria 55 and 57 of
Appendix A to 10 CFR 50 to the FCS.
(285/8614-02)
The NRC inspector also reviewed the Updated Safety Analysis Report (USAR)
in an attempt to determine the applicability of Criteria 55 and 57.
The
NRC inspector noted that the discussion provided in Section 5.9.5 of the
USAR did not discuss the criteria applicability.
However, the NRC
inspector did note that the discussion provided in Section 5.9.5 and
Table 5.9-1 did not reflect the actual plant installation.
For example,
Table 5.9-1, in conjunction with Figure 5.9-19, indicated that the steam
generator secondary side drains are two air-operated valves in series.
The actual installed configuration in the plant was two manual valves in
series. The licensee stated a review of Section 5.9.5 and Table 5.9-1 in
conjunction with Figure 5.9-19 would be performed to verify that the
information provided was correct.
This is an open item pending review of
the USAR information during a subsequent inspection.
(285/8614-03)
7.
Safety-Related System Walkdowns
The NRC inspector walked down accessible portions of the following
safety-related emergency diesel generator 1 and 2 systems to verify system
operability.
Operability was determined by verification of selected valve
and switch positions.
The systems were walked down using
Procedures 0I-DG-1, Revision 19, OI-DG-2, Revision 19, and the drawings
noted below:
Fuel oil system (Drawing M-262, Revision 6)
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Air start system (Drawing B120F07001, Revision 12)
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Lubricating oil system (Drawing B120F03001, Revision 1)
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Jacket cooling water system (Drawing B120F04002, Revision 1)
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During the walkdowns, the NRC inspector noted minor discrepancies of an
editorial nature between the drawings, procedures, and plant as-built
conditions.
None of the conditions noted affected the TS operability or
safe operation of the system.
Licensee personnel stated that the noted
minor discrepancies would be corrected.
No violations or deviations were identified.
8.
Monthly Maintenance Observation
The NRC inspector reviewed / observed selected station maintenance
activities of safety related systems and components to verify the
maintenance was conducted in accordance with approved procedures,
regulatory requirements, and the TS.
The following items were considered
during the reviews / observations:
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The limiting conditions for operation were met while systems or
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components were removed from service.
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Approvals were obtained prior to initiating the work.
Activities were accomplished using approved maintenance orders (MO)
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and were inspected, as applicable.
Functional testing and/or calibrations were performed prior to
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returning components or systems to service.
Quality control records were maintained.
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Activities were accomplished by qualified personnel.
Parts and materials used were properly certified.
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Radiological and fire prevention controls were implemented.
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The NRC inspector reviewed / observed the following maintenance activities:
Overhaul and repair of motor-driven fire water pump (MR FC-83-69).
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Inspection of wiring on Limitorque motor-operated valves (M0 862062).
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No violations or deviations were noted.
9.
Monthly Surveillance Observation
The NRC inspector observed selected TS required surveillance testing on
safety-related system 3 and components.
The NRC inspector verified the
following items during the tasting:
Testing was performed using ap, coved procedures.
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Test instrumentation was calibrated.
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Limiting conditions for operation were met.
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Removal and restoration of the affected system and/or component were
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accomplished.
Test results conformed with TS and procedure requirements.
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Test results were reviewed by personnel other than the individual
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directing the test.
Deficiencies identified during the testing were properly reviewed and
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resolved by appropriate management personnel.
The NRC inspector witnessed the following surveillance test activities:
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Component cooling water pump quarterly test (ISI-CC-3-F.1).
Emergency diesel generator monthly test (ESF-6-F.2).
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Control element assembly check (CEA-1-F.4).
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No violations or deviations were identified.
10.
Exit Interview
The NRC inspector met with Mr. W. G. Gates (Plant Manager) and other
members >f the OPPD staff at the end of this inspection.
At this meeting,
the ins actor summarized the scope of the inspection and the findings.
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