ML20154L296
| ML20154L296 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 09/19/1988 |
| From: | Seidle W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20154L281 | List: |
| References | |
| REF-PT21-88, TASK-A-46, TASK-OR 50-285-88-25, GL-83-28, NUDOCS 8809260141 | |
| Download: ML20154L296 (31) | |
See also: IR 05000285/1988025
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APPENDIX B
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
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NRC Inspection Report:
50-285/88-25
Operating License:
Docket:
50-285
Licensee:
Omaha Public Power District (OPPD)
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1623 Harvey Street
Omaha, Nebraska 68102
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Facility Name:
Fort Calhoun Station (FCS)
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Inspection At:
Inspection Conducted: August 8-19, 1988
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Inspectors:
W. C. Seidle, Chief Test Programs Section
(Inspection Team Leader)
J. R. Boardman, Reactor Inspector
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W. M. McNeill, Reactor Inspector
J. P. Stewart, Reactor Inspector
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T. O. McKernon, Reactor Inspector
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Approved:
[m d
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W. C. Seidle6Jchief, Test Frograms Section
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Division of reactor Safety
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Inspection Sumary
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Inspection Conducted August 8-19, 1988 (Report 50-285/88-25)
Areas Inspected: Announced special inspection of the licensee's followup to
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previous inspection findings, followup to licensee event reports (LERs),
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temporary instructions,10 CFR Part 21 reports, bulletins, and the safety
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systems outage modification inspection (SSOMI).
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Results: Within the six areas inspected, one violation was identified
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(failure to maintain records, paragraphs 7.0.6).
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1.0 Persons Contacted
Licensee
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- W. G. Gates. Plant Manager
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- S. K. Gambhir, Division Manager, Production Engineering
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- T. L. Patterson, Assistant Manager, FCS
- A. W. Richard, Manager - Quality Assurance / Quality Control
- B. Livingston, Manager, Engineering Services
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- T. J. Mcivor, Manager, Nuclear Projects
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- S. Willutt, Manager, Administrative Services
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- J. K. Gasper, Manager, Training
- H. M. Tackett, Consultant
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- M. D. Matheson, Consultant
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- C, F. Simons Onsite Licensing Engineer
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J. E. McKinley, Supervisor. Electrical Projects Control Maintenance
- J. A. Drahota, Supervisor, Maintenance Support
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- J. J. Fisicaro, Manager Nuclear Licensing and Regulatory Affairs
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- L. L. Gundrum, Nuclear Licensing Engineer
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A. J. Stepanek, Consultant
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J. L. Oyer, Senior Quality Control Inspector
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D. W. Dale, Supervisor, Quality Control
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C. N. Bloyd, Lead Special Services Engineer
J. L. Kyle, Senior Production Planner
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H. Faulhobs, Manager, Electrical Engineering
C. Brunnert, Supervisor, Operations Quality Assurance, OPPD
R. Hyde, Supervisor, Maintenance Training
J. Fluehr, Supervisor, Training
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Stone and Webster Consultants
P. A. Nelson, Engineering Assurance Engineer
B. B. Barta. Licensing Engineer
J. J. Purcell, Project Engineer
D. R. Beach, Assistant Project Manager
C. Stuart, Materials Manager
NRC
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- P. H. Harrell, Senior Resident Inspector FCS
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- T. Reis, Resident Inspector, FCS
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- Denotes those personnel attending the exit interview,
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2.0 Followup on Previously Identified Inspection Items
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?.1 Violations
2.1.1
(Closed) Violation 285/8503-02:
Failure to retrieve records of
thermal stress analysis for thermally-stressed safety-related pipe
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below 2 1/2 inch diameter. This violation was based on the fact that
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FCS stress analysis records were missing for thermally stressed
safety-related piping in the size range that included the critical
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small break loss of coolant accidtnt (SBLOCA) for combustion
engineering reactor plants.
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This violation is closed based on the licensee's planned res;e:. c
the subject pipe for thermal stresses as part of the FCS design bare
reconstitution (DBR) program.
(stressing) of main steam safety relief valve (MSSRV) per tensioning
285/8705 01:
Failure to assure pro
(Closed) Violation
2.1.2
inlet line
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flange bolting. This violation dealt with two concerns as follows:
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Failure to document a design basis for the torque specified for
bolt tightening
Failure to use W specified torque (a "slugging wrench" was
used)
In response to the specific concerns, the joints were reworked to
verified, controlled procedures.
The generic corrective action is contained in licensee program
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Project 1991, which closes this violation.
This program is
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structured to:
(1)assurethatsafety-related,aswellasother
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controlled jnints are properly made and (2) assure the specification and
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use of the correct fastener tension values. The NRC inspector
reviewed the licensee's Maintenance Precedure HP-BOLT-1 on bolting,
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Revision 0, dated June 2, 1988.
The NRC inspector discussed the
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torquing program with licensee personnel and reviewed licensee
Memorandum FC-795-87, "Guidelines for "irquing Reviews," dated
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May 17, 1987.
Generic incluston of requ) red tensioning criteria will
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be implemented by the OPPD FCS Project 1991, by licensee response to
NRC Generic letter (GL) 83-28 "Required Actions Based on Generic
Implementatien of Salem ATWS Events." and by the licensee's FCS DBR
program.
2.1.3
(Closed) Violation 285/8705-02:
Failure to have procedures for
properly tensioning bolts, for assuring the accuracy of calibration,
and for controlling the meggering of electrical circuits and
components.
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This violation involved the adequacy of maintenance procedures,
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Tensioning is covered in the resporse to Violation 285/8/05-01.
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procedures for meggering had been issued by the licensee, such as FCS
Special Procedure SP-EE-MEGGER, Revision 0, dated May 13, 1987.
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The accuracy of calibration and the precision of calibrated metrology
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requirements has bien covered by revisions to the following
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procedures *
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Standing Order (50) No. M-28, "Calibration of fest Equipment and
Plant Process Equipment used to Support the In-Service
Inspection ?f Nuclear Plant Components Program," Revision 28,
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dated July 25, 1988
SO No, M-26. "Calibration Procedures," Revision 13, dated
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July 25, 1988
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2.1.4
(Closed) Violation 285/8705-04: This violation, which aise dealt
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with tightening of the MfSRV line flange bolts, became part of
Violation 285/8705-01 when the N9 1ce of Violation was subsequently
issued. This item is closed with Violation 285/8705-01.
2.1.5
(Closed)Violr. tion 285/6724-06:
10 CFR Part 21 infomation not
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posted. This violation concerned the failure to post the latest
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revision to the 10 CFR Part 21 regulations, the licensee's
implementing procedure, and the failure to post Section 206 of the
Energy Reorganization Act (ERA) of 1974 in the generating station
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engineering offices located in the Brandeis building in downtown
Omaha.
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In response to this violation, the latest revision to 10 CFR Part 21
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regulations has been posted on bullotin boards on the sixth floor of
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the Brandets building along with Section 206 of the ERA and the
licensee's implementing procedure, "Nuclear Produ: tion Policy
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Procedure QP-12."
The postings were verified by B. Livingston.
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Manager. Engineering Services, and reported to the NRC inspector by a
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telephone call on August 11, 1988.
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The licensee issued a new procedure, NPD-QP-17-1, Revision 0,
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"Posting of NRC Cequired Dccur.wnts," on April 18,198E.
The NRC
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inspector reviewed the procedure and found that the required
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documents to be posted and those responsible for posting these
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documents in specified locations are clearly identified,
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2.1.6
(0 pen) Violation 285/8802-01:
Failure to document the use of
material as specified in the design.
This violation identified that
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during November 1985, spray paint used on containment vetilation
duct supports was not listed in the dec!gn package.
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-The licensee established the cause of this violation to be that
personnel failed to list sufficient information in the design package
beyond the words "galvanized paint" because of insufficient guidance.
The licensee attempted to identify the coating material and evaluate
it. Curing the 1988 outage, the licensee coninitted to further
evaluate the coating material. As corrective action, the licensee
has revised the design controi procedures, "Preparation of Design
Packages," GEG-3; "Station Modification Control," SO No. G-21; and
Technical Specification (TS), "Selecting, Specifying, Applying, and
Inspection Paint and Coatings," No. CTS-3. These measures should
assure that this type of problem does not occur in the future. The
NRC inspector verified that the above revisions were made and
implemented (see the comments below on corrective action in. paragraph
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7.0.5).
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2.1.7
(0 pen) Violation 285/8810-01:
Failure to promptly resolve test
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deficiencies. This violation involved the failure to evaluate
anomalies or deficiencies related to Surveillance Test ST-NZ-1
completed on May 8, 1987.
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During the followup inspection, the NRC inspector verified that the
licensee was in the process of obtaining an accident analysis based
upon operation of the containment spray system in a degraded
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condition (i.e. 12 inoperable spray valves).
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results of the analysis, a TS amendment may or may not be submitted.
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Furthermore, the licensee has committed to reviewing a sampling of
previous surveillance tests.
The licensee's present records relating
to LER 88-008 indicate that performance of Surveillance Test ST-NZ-1
is not planned for the 1988 refueling outage.
LER 88-008 states that
by the end of the 1988 refueling outage either the 12 spray heads
will be tested or based upon the request >d analysis, a change request
to the TS shall be submitted.
Since the licensee has not scheduled
the test for the 1983 outage, it follows that future action will be
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based upon the analysis findings.
This vtolation shall remain open
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pending review of the licensee's final accions.
2.1.8
(Closed) Violation 285/8810-02:
Failure to use correct qualification
level of examiner for surveillance test evaluation. This violation
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involved the surveillance test results evaluation of Surveillance
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Test ST-RLT-1.F.1 (leak test) conducted on May 29, 1987, by an
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individual qualified to a lower certification than required by ASME
Section XI, IMA-2000.
In response to the violation, the lice > se denied this violation in
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that the test
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examiners and documented on quality control log sheets as required by
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FCS S0 No. G-26A, Appendix G Paragraph 8.3.2.
During the followup
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inspection, the NRC inspector reviewed Surveillance Test
Procedure ST-RLT-1, F.1, dated May 29, 1987, 50 No. G-26A, and
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Quality Control Log Nos. 3542, 3543, 3544, and 3547, dated May 29,
1987.
The NRC inspector determined that, in fact, the licensee had
apparently violated regulations.
In particular, the licensee had
apparently violated 10 CFR 50 Appendix B, Criterion V, and licensee
procedures in that surveillance test procedures did not contain QC
signoffs. The failure to have QC signoffs in the testing procedure
is incongruent with licensee requirements in 50 No. G-26A,
paragraph 6.6.
The NRC inspector reviewed the licensee's corrective actions to
resolve the above violation and implement actions to obviate future
similar recurrences. The licensee had implemented Revision 9 to
Surveillance Test ST-RLT-1-1 on July 18, 1988, to incorporate QC
signoffs in the surveillance procedure.
Furthermore, the licensee
conducted refresher training of key personnel on the administrative
requirements of S0 No. G-26A. The NRC considers the licensee's
corrective actions in this matter to be responsive and comprehensive.
No further written response by the licensee is required.
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2.1.9
(Closed) Violation 285/8811-04:
Inaccurate information provided in
violation response.
This violation concerned the licensee's response
to Violation 285/8724-04, dated February 24, 1988, which contained
inaccurate information. The response stated that no instances of the
failure to control gas cylinders had been nmed since September M
when, in fact, an instance occurred in December 1987.
In response to this violation, the licensee issued a prompt esision
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to its February 24, 1988, response to eliminate the inaccurtte
statement.
The licensee has developed an administrative process
whereby the plant manager, or his designated alternate, assigns
responsibility for drafting responses to NRC violations and
deviations.
These assignments, which are made imediately af ter an
exit neeting, are included in the integrated regulatory requirements
log (RRL) to assure tracking of these items until the NRC inspection
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report is received. By adopting this process, the responsible
individual now has time to draft the response because the
responsibility has been clearly and promptly defined. A change has
been made to NPD Procedure G-2, "Regulatory Reouirements Log (RRL),"
to require that Form No. FC-1077, "Certification of Accuracy," be
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included with the draft response package.
This effort should provide
additional assuranTe TTiat the individuals responsible for drafting
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responses will not only have adequate time to prepare the response
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package but they will also provide accurate information.
The NRC
inspector, in a discussion with the plant manager, confirmed that the
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administrative process described above is being implemented.
The NRC
inspector reviewed NPD Policy / Procedure No. G-2 and verified that the
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procedure requires the assignee to sign Form No. FC-1077,
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"Certification of Accuracy," which is provided with the regulatory
requirement document (Ref:
paragraph 6.9.3).
The NRC inspector also
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confirmed that violations / deviations identified in NRC Inspection
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Reports 50-285/88-10, 88-12, and 88-21, which were selected at
random, could be traced back to their respective RRL sheets.
The NRC
inspector also verified that Form No. FC-1077.was included in a
response package being prepared for a violation identified in NRC
Inspection Report 50-285/88-21.
2.1.10
(Closed) Violation 285/8812-01:
Unqualified senior reactor operator
(SRO) standing shift supervisor duties.
This violation involved a
licensee's licensed operator performing duties with questionable
qualification in that the operator failed to attend a majority of
preplanned lectures.
During the followup inspection, the NRC inspector verified through a
review of training records that the cited reactor operator was
attending a majority of training sessions since the violation period.
Furthermore, corrective actions had been taken to revise Training
Procedure TAP-13 to require reactor operators to attend a minimum of
75 percent ratio of preplanned lectures.
2.2 Deviations
2.2.1
(Closed) hviation 285/8810-03:
Failure to continue implementation
of corrective actions.
This deviation involved a failurt to account
for surveillance tests as comitted to by the licensee in response to
Violation 285
11-01.
During the followup inspection, the NRC inspector reviewed the
licensee's response to Deviation 285/8810-03 and verified the
licensee's implerrentation of stated corrective actions. The NRC
inspector verified the implementation of a new surveillance test
tracking program to ensure that surveillance tests are completed on
t me.
Furthermore, the NRC inspector confirmed that the licensee has
ma<' a concerted effort to eliminate the backlog of delinquent
surveillance test reviews and that timely review of surveillance
tests are being accomplished by the responsible craft supervisors.
2.2.2
(Closed) Deviation 285/8715-01:
The Deviation consisted of the
failure to revise a surveillance procedure as stated in a licensee
eventreport(LER). On February 7,1987, LER 87-001 was issued by
the licensee, which described an event where a TS limiting condition
for operation was entered when safety-related equipment in redundant
trains was concurrently out of service. The equipment was removed
from service during the performance of Surveillance Test
Procedure ST-ESF-2.
In LER 87-001, the licensee stated that
Procedure ST-ESF-2 had been revised to designate the responsibility
for ensuring that no equipment was inoperable to the shift supervisor
prior to performing the surveillance test.
The licensee, in fact,
did not revise ST-ESF-2 until March 20, 1987.
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The NRC inspector reviewed the licenste's response to the above
deviation.
The licensee identified the nrganizational and procedural
weaknesses, which contributed to the above deviation. The licensee's
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enrrective actions reviewed included the following:
LERs are
currently being written within one group, the shift technical advisor
group; certification of the accuracy of operations incidents
information is required by 50 No. R-4, "Operating Incident Reports,"
Step 6.5.3, Form FC-1077, and licensee internal memorandum dated
December 31, 1987, (R. L. Andrews) with the attached Policy /
Procedure No. G-2 Regulatory Requirements Log (RRL).
No problems
were noted.
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2.3 Unresolved Items
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2.3.1
(Closed) Unresolved Item 285/8503-01: Apparent inability to idantify
construction records verifying as-built plant compliance with its
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licensed design bases. A specific example was documentation of the
pump curves to demonstrate operability of safety injection (SI) pumps
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as specified in PSAR Section 6.2.
During this inspection, the NRC
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inspector again requested the data necessary to prove adequate design
flow of the SI pumps.
The licensee was unable to retrieve this data.
This item is closed based on the presently identified design base
reconstitution (DBR) program for FCS.
Presentations of the scope and
schedule of this program have been made to the NRC.
2.3.2
(Closed) Ur
olved Item 285/8503-03: Design of pipe supports for
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thermally-s," ; sed safety-related pipe below 21/2 inch diamercr.
The concern ,sr proper support of thermally-stressed pipe of this
size range is discussed in the closure of Violation 285/8503-02.
This unresolved item is close( Sased on review of the subject hangers
as part of the FCS DBR program.
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2.3.3
(Closed) Unresolved Item 285/8523-01: Apparent lack of document
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controls to assure that calculations of seismic loads for
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safety-related conduit supports used in station modification.
The licensee response, which closes this unresolved item, is
contained in licensee internal memorandum from S. X Gambhir to
J. J. Fisicaro, PED-FC-88-287, dated August 5, 1988.
This memorandum
was provided to the NRC inspector by 0 PPD licensing personnel. The
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tremorandum stated that it is to provide a response to Unresolved
Item 285/8523-01.
The subject of this memorandum is seismic adequacy of
mechanical and electrical equiptrent (USI-A46), specifically, conduit
supports.
Its content is as follows:
"0 PPD's Proauction Engineering Department (PED) has calculations
and analyses for all seismic conduit supports installed during
the 1985 outage and any supports installed since then.
Licensee
Procedure GSEE-0516 and superseding licensee Standard CTS-1 were
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subsequently developed and approved for use to address standard
conduit supports and support spacing criteria.
For supports
installed prior to the 1985 outage, PED is comitted to
implement a verification program to ensure adequacy of the
supports.
"On July 29, 1988, the NRC issued a Generic Safety Evaluation
Report (SER) that endorses the parts of the Generic
Implementation Procedure (GIP) submitted by the Seismic
Qualification Utility Group (SQUG), which have been accepted for
the implementation of USI-A46. As required by the cover letter,
the NRC commits to establishing a schedule for implementation of
the SQUG verification program by October 1,1988."
2.3.4
(Closed) Unresolved Item 285/87s -03: Applicable code and
specification revisions for pi,
g system design.
The concern was
the apparent lack of document - strol of revisions of the design
codes used by OPPD for safety-i-:ated modifications and maintenance
at FCS.
Licensee Memorandum PED-FC-88-285, dated August 5, 1988, responds to,
and closes, tais concern as follows:
"The State of Nebraska has endorsed ASME Sections I, III, IV,
and VIII as acceptable to the Nebraska Code. However, it is the
NRC's understanding that the State has not taken a position on
nuclear repair programs and based on the above, has allowed the
use of ASME Section III on new systems. Therefore, the
following requirements are imposed on modifications to existing
and new nuclear piping systems which OPPD is comitted to by
law:
"10 CFR 50.55a requires licensees to develop an Inservice
Inspection Program in accordance with ASME Section XI.
10 CFR 50.55a also establishes the NRC acceptable Edition
and Addenda of the Codes (Section III and XI) for use by
licensees.
Safety-related modifications to existing piping
systems would be subjected either to ASME Section XI Repair
or Replacement rules, as applicable (e.g.,
Articles IWA 4000 and IWA 7000 of ASME Section XI,
respectively).
These articles allow modifications to
piping systems to either use the original Construction
Code (B31.1/B31.7), in whole or in part, later editions /
addenda of the same or ASME Section III
(reference IWA 4120, IWA 7210, ASME Section XI, 1986
Edition). When later editions / addenda of the Construction
Code or ASME Section III are used for ASME Section XI
replacements, reconciliation is required between the
original construction code and the code selected for the
modi fication.
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"Section XI of the ASME code'does not apply to the
installation of complete new nuclear piping systems as
delineated in ASME Section XI, IWA 1200.
These systems,
which would not modify any existing safety-related system,
would be designed and installed in accordance with ASME
Section III. The Edition and Addenda used in the
installation of new nuclear safety-related. piping systems
would be those acceptable to the NRC, as delineated in
"PED had documented in GEI-3, Preparation of Design
Packages, the ASME Codes (Sections III and XI) Edition and
Addenda, which are acceptable to the NRC.
These could be
used in existing nuclear piping systems modification work
and could be used for the design and installation of new
nuclear safety-related piping systems.
It should be noted
that NRC acceptable code editions change periodically.
For
example, 10 CFR 50.55a was updated in May 1988, and now
approves portions of the 1986 Edition and Addenda of the
1986 ASME Sections III and XI Codes. Additionally, PED is
in the process of developing a General Engineering
Instruction, ASME Section XI, Repair / Replacement Program,
to provide additional guidance to PED personnel in
preparing ASME Section XI Repairs and Replacement
modifications.
This instruction is scheduled for issuance
October 31, 1908.
"In summary the following construction codes are applicable
to the Fort Calhoun Station, which is consistent with both
federal and state requirements:
For safety-related modification work, the original
construction code, later editions / addenda thereto, or
ASME III, provided the requirements to IWA ~210 are
met.
For new safety-related systems, ASME Section III
approved revision per 10 CFR 50.55a.
Previously performed modifications will be reviewed as part
of the FCS DBR program."
2.3.5
(Closed) Unresolved Item 285/8705-05:
Weaknesses in the licensee
program for review of vendor technical manuals. This item dealt with
the adequacy of the licensees response to NRC GL 83-28, Section 2.22,
Review of Vendor Technical Information (VTI).
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This item is closed based on an enhanced technical manual review
program included in OPPD FCS Project 1991.
This review program was
identified to the NRC in OPPD Letter LIC-88-043, dated March 8, 1988,
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NRC GL 83-28. Item 2.2.2, Review of Safety Related Vendor Manuals
Received Prior to June 1985."
2.3.6
(Closed) Unresolved Item 285/8705-06: Weaknesses in the licensee
program for review of vendor technical manuals.
This item is similar
to Unresolved Item 285/8705-05 and is closed with that item.
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2.3.7
(Closed) Unresolved Item 285/8705-07:
Lack of a documented licensee
program to upgrade maintenance procedures.
This concern resulted
from the failure of the licensee to have a program to incorporate
either appropriate vendor recommended maintenance in response to NRC GL 83-28, or identifiable requirements for replacement of
subcomponent units, such as antifriction bearings having a design
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life less than plant life, as required to maintain safety-related
design bases.
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This concern has been incorporated into OPPD FCS Project 1991.
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2.3.8
(Closed) Unresolved Item 285/8705-09: Training of plant personnel in
electrical equipment qualification. Licensee Lesson Plcns 12-81-05,
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Revision 0, dated October 28, 1937, and 12-61-03, Revision 0, dated
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January 13, 1988, now include craft training in electrical equipment
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qualification requirements.
2.3.9
(Closed)UnresolvedItem 285/8705-10:
Control of safety-related (CQE)
parts. This item dealt with maintenance department control of
material after material issue from the store room.
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Maintenance Training Lesson Plans 12-81-03 ( May 12, 1988) and
12-19-0 (undated) now contain data on control of CQE material by
maintenance personnel. The FCS material program is also included in
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2.3.10
(Closed)UnresolvedItem 285/8705-11:
Control of vendor technical
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manuals. This item dealt with technical review of technical manuals
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for component specific applicability.
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This item is considered clnsed based on the licensee response to
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GL 83-28, Section 2.2.2, and OPPD FCS Project 1991.
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2.3.11
(Closed)UnresolvedItem 285/8705-13:
Periodicity of lubrication of
containment cooling fan motors.
The concern was that required
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periodicity would not include allowable operating hours plus the time
for required Post-LOCA fan operation.
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Licensee Memorandum PE0-FC-296, dated August 8, 1988, responds to,
and closes, this concern as follows:
,
m..
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"Production Engineering Department (PED) reviewed the
lubrication frequency issue relative to cor'. 'nment cooling
fans VA-3A/B and VA-7C/D.
Based on the cperatng items of the
equipment and the containment operating anvironnant, the
required lubrication frequency has been established for VA-3A/B
and VA-7C/D as 9 and 18 months respect'.vely.
These values are
documented in calculation FC-04026, "fEG-H-08 Containment Vent
Fan Qualification.
"To resolve this item, plant documents, inciuding the
Preventative Maintenance Procedure PM-EE-12, must be updated to
the new lubrication frequency for VA-2A/B.
In addition, an
engineering evaluation must be conductej to establish the
acceptability of the bearings which were lubricated on a
frequency outside the established parameters."
The NRC inspector was infomed by licensee personnel that the subject
fans would be returned to the manufacturer for overhaul during the
upcoming outage.
2.3.12
(Closed) Unresolved Item 285/8723-01:- Inspector-qualifications.
This item deals with documentation of the verification of an
inspector's employment and education.
The licensee has obtained records that show the verification of the
inspector's employment and education that were used by Ebasco during
the 1985 outage.
Tbt: NRC inspectors reviewed these records and found
no problems.
2.3.13
(Closed)Unresolveditem 285/8724-03:
Cable tray support
required.
This item deals with an observation that a cable tray
support in Poom No. 81 wus not attached to its structural member.
The licensee found, after analysis, that the support in question was
not necessary. The support, nevertheless, was reattached to its
structural member. Tne NRC inspector reviewed Calculation
No. 5100001-01-001 and the maintenance order for the reattachment (874569).
.
The NRC inspector also found that the generic issue of electrical
I
supports will be addren:d by the licensee under the Seismic
Qualification Utility Group's Generic Implementation Procedure, which
has been appror
by a NRC safety evaluation report.
'
2.4 Open items
l
2.4.1
(Closed) Open Item 285/8523-02:
Licensee control of limited life
electrolytic capacitors for safety-related functions. This item
related to electrolytic capacitors having a total design-life of 5 to
15 years.
Nonconservative failures of these capacitors might affect
reactor safety.
,
_
_ _ _ _ _ _ _ _ _ _ _ _ _
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,
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The licensee had instituted a study of design life for site-specific
electrolytic capacitors. This study is in conjunction with OPPD
FCS Project 1991. Volume III, Preventative Maintenance.
2.4.2
(Closed) Open Item 285/8705-12: Replacement of safety-related
Agastat electrical relays. This open item relateJ to the lack of a
program to replace comercial grade Agastat relays used in
safety-related applications, including the emergency diesel load
sequencing panels for engineered safeguards features' (ESF)
actuations.
Subsequent to this open item, NRC Information Notice 87-56 was
issued.
This notice identified a generic concern relat've to
comercial grade components being used beyond their design life.
The
specific components used as an example in the notice were "7000
series" Agastat comercial grade relays having a projected 2-year
qualified life. The "7000 series" relays are a one-for-one replacement
for the Agastat "2400 series." The "2400 series" was discontinued
between 1972 and 1974. The FCS Agastat "2452" relays of concern were
apparently manufactured in <>r about 1968.
The licensee comitted to
complete a review of this concern by August 31, 1988.
Implementation of the licensee program for updating and incorporating
l
VTI into maintenance procedures in response to NRC GL 83-28, as
'
contained in OPPD FCS Project 1991, is the basis for closure of this
item.
2.4.3
(Closed) Open Item 285/8713-04:
Review actions taken by the QA
department in the close out of FCS Deficiency Report DR-FC-1-87-056
!
concerning the failure to follow procedures for making changes to a
design installation package.
This item related to an improper chan
made by an engineer to a design installation package (MR-FC-83-05) ge
on
a containment storage platform and a discrepancy between installation
drawing requirements and installation procedure instructions
concerning the appropriate welding codes to be used.
The NRC inspector reviewed the licensee's closecut package for this
item and noted the following:
The licensee performed an engineering calculation to verify the
acceptability of the use of nonquality (non-CQE) material to
plug abandoned bolt holes in the platform base plates.
(May 27,
1987)
The licensee performed a 10 CFR 50.59 ualuation, which
concluded that an unreviewed safety question did not exist.
(June 4, 1987)
The licensee changed the design drawings to reflect the actual
material used to plug platform holes.
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The licensee had implemerted a revised welding. program effective
January 1988, which included improved methods for documentation
of weld design, installation, and QC inspection.
The QA department closed DR FC-1-87-056 ba:ed on the completion
of the above on October 8, 1987.
No problems were noted.
2.4.4
(Closed) Open Item 285/8812-02:
Revision to requalification program
required in TAP-13.
This open item involved the licensee's planned
revision to Training Procedure TAP-13 to change the required training
makeup period from 18 weeks to 6 weeks.
The NRC inspector verified that the licensee had revised TAP-13
effective August 3, 1988, to incorporate the planned change.
2.4.5
(Closed) Open Item 285/8812-03:
Use of contract personnel as
training instructors. This open item involved the licensee's
extensive usage of contract personnel as training instructors.
During the followup inspection, the NRC inspector conducted
discussions with key training supervisors and reviewed training
records.
In spite of a high number of contract personnel acting as
training instructors, an ongoing effort is being made to add senior
licensed individuals to the training staff.
2.4.6
(Closed)OpenItem 285/8812-04:
Need to formalize training
requirements for contract instructors.
During the followup inspection, the NRC inspector verified that the
licensee had fonnalized requirements for training staff members.
It
was verified that Training Procedure TAP-1-CRPI, Revision 3, dated
July 1,1988, requiring staff members to receive 12 days in-the-plant
familiarization had been implemented.
2.4.7
(Closed)OpenItem 285/8812-05:
Lesson plans are not being
maintained up-to-date.
During the followup inspection, the NRC inspector reviewed training
records, which statused lesson plans.
The NRC inspector noted that
the licensee was implementing revisions to lesson plans in a timely
manner.
2.5 Review of Design Base Inadequacies
The NRC inspectors reviewed the Design Base Reconstitution (DBR) System
Desigi Base Document (SDBD)-DG-112-0, covering the FCS emergency diesel
generators and associated systems. Attachment 22, Open Item 8
SDBD-DG-112-0, stated that, "the diesel generator (E) full load fuel
consumption must be determined to verify that sufficient fuel is available
on site." Site storage of EDG fuel is an operability concern.
In this
instance. Technical Specifications address fuel oil capacity.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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15
The NRC inspector was told, that there were no OPPO procedures or SWEC
procedures that required review of identified inadequate or incorrect design
base data for reportability.
By contract, the delay could be 18 months.
This item is considered unresolved pending devele ment by the licensee of
procedures for review of DBR documents for cpera'.nlity and reportability
considerations (285/8825-02).
3.0 Followup to Licensee Event Reports (LERs)
3.0.1
(0 pen) LER 88-006: Surveillance Test ST-DC-1 F.1, not performed
during January 1988.
This event involved a failure by the 11 a nsee
to perform Surveillance Test ST-DC-1 F.1, "Station Batteries" during
January 1988.
The licensee stated that the reasons for missing the
surveillance test were due to problems in the surveillance tracking
system.
During the followup inspection, the NRC inspectors verified that the
licensee had implemented a new surveillance test tracking system and
no surveillance test appeared to be delinquent.
The licensee had
furthermore committed to review past surveillance tests without the
25 percent extension allowance and submit the results of the review
to the NRC via a LER supplement,
in addition, the licensee has
submitted a TS amendment requesting a 25 percent extension for all
surveillance test intervals not presently covered under this
extension. As of the followup inspection, neither the LER supplement
had been submitted nor the approval of the TS amendment been
received.
3.0.2
(Closed) LER 88-007:
Inadvertent start of Emergency Diesel
Generator D-1 as the result of surveillance test error. This event
resulted from operator error in that the operator while unloading the
EDG allowed the load to drop too lcw before opening the breaker.
This resulted in a reverse current condition.
The EDG D-2 lock-out
relay tripped and EDG D-1 auto-started.
During the followup inspection, the NR', inspector verified the
licensee's corrective actions.
The licensee stated that planned
corrective actions included a r2 vision to the Updated Safety Analysis
Report (USAR) and the Surveillance Test Procedure ST-ESF-6. The NRC
inspector verified that the licensee had revised the USAR to identify
the EDGs as engineered safeguards components and revised the
surveillance test to incorporate instructions and signoffs for
unloading the EDG.
3.0.3
(0 pen) LER 88-008:
Failure to conduct Surveillance Test ST-NZ-1
within required interval.
This event involved the licensee's failure
to conduct the required surveillance testing on twelve of the
containment spray nozzles due to inaccessibility.
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16
The licensee has obtained the service of a consultant to perform an
accident analysis based upon a degraded containment spray system.
Based upon the analysis a TS amendment may be made. As of the
followup inspection, the licensee had not received the results of the
analysis, taken actions to amend the TS or scheduled testing of the
remaining nozzles during the 1988 refueling outage.
This LER will
remain open pending review of the licensee's final corrective
actions.
3.0.4
(Closed) LER 88-014:
Inadvertent start of Emergency Diesel
Generator D-1 during performance of surveillance test.
This LER
involved the inadvertent starting of Emergency Diesel
Generator (EDG) D-1 during performance of the monthly Surveillance
Test ST-ESF-6 F.2, Appendix E (EDG D-2).
During the postevent investigation, the licensee determined that the
EDG D-2 Lock-Out Relay 86/02 tripped due to a reverse current flow
across the output breaker.
The licensee could not find any root
cause for the event other than a possible spuricus voltage spike.
Subsequent to the event, the licensee repeated the surveillance test
and verified EDG D-2 operated properly.
The Lock-out Relays 86/D-2
and 86/D-1 are designed to trip the EDG on overcurrent, phase
differential, and reverse power across the breaker. The NRC
inspectors concluded the relays performed their intended design
function and no damage occurred to the EDG D-2.
The licensee's
actions appeared to be both complete and comprehensive.
3.0.5
(Closed)LER88-015:
Inadvertent start of the standby component
cooling water (CCW) pump during breaker testing. This event involved
the inadvertent starting of the standby CCW aump during perfonnance
of Procedure CP-AC-3B BKR. An expected brea <er mismatch occurred on
the AC-3B breaker, which resulted in an auto-start of the AC-3B
component cooling water pump.
During the followup inspection, the NRC inspectors reviewed the
licensee's corrective actions. The licensee had modified the test
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procedure to add a step requiring the operations department to
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signoff the action of placing the nonrunning redundant CCW pump, not
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tested in the pull-to-lock position and actions requiring the pump to
!
be returned to service after testing.
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3.0.6
(Closed)LER87-17:
Failure to inspect emergency diesel generators,
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This event involved the failure of the licensee to perform the annual
inspection of the emergency diesel generators within the required
surveillance period.
During the followup inspection, the NRC inspector reviewed the
licensee's actions in researching this event.
Furthermore, the NRC
inspector reviewed the licensee's surveillance tracking system and
verified that surveillance testing was being conducted on schedule.
!
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17
,
The licensee had also submitted and received NRC approval via TS
Amendment 112, dated April 19, 1988, to extend the annual EDG
inspection to each refueling period.
3.0.7
(0 pen) LER 87-29:
Failure to conduct Surveillance Test ST-DC-4
Section F.3.
This event involved the failure of the licensee to
conduct surveillance testing of the containment emergenc/ lighting
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system within the once a year TS requirement.
During the followup inspection, the NRC inspector verified the
licensee's implementation of actions directed toward the generic
problem of surveillance tests not being conducted when scheduled.
It
was further verified that Surveillance Test St-DC-4 F.3 is scheduled
to be conducted during the 1988 refueling outage.
The NRC inspector
noted that the licensee had committed to reviewing past surveillance
tests to ensure compliance with planned scheduling.
The licensee had
,
comitted to submit the review results to the NRC.
Furthermore, the
'
licensee has proposed to amend the TS to change the surveillance
period to each refueling outage versus annually.
As of the followup
inspection, the licensee had neither submitted the results of the
sampling review nor submitted a TS amendment. This LER shall remain
open pending completion of the licensee's stated comittments.
3.0.8
(Closed) LER 87-037:
Diesel generator surveillance test not in
conformance with TS.
This event involved the licensee's failure to
perform the emergency diesel generator (EDG) surveillance test on
November 11, 1987, in accordance with the TS Amendment 111, require-
ment 3.7(1)a(ii). This requirement stated that the EDGs shall be
tested at the continuous KW rating for 60 minutes. When the
surveillance test was performed on November 11, 1987, the licensee
had not incorporated the TS Amendment 111 requirements into the
surveillance procedures to ensure adequacy and timely completion.
During the followup inspection, the NRC inspector verified that the
surveillance test had been rivised to incorporate TS Amendment 111
requirements.
Further, it was verified that the licensee is pursuing
a comprehensive and continuing review of surveillance procedures to
ensure adequacy and timely completion.
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3.0.9
(Closed) LER 84-15R1:
Load over the reactor coolant system (RCS).
The polar crane was loaded with a load of 250 pounds suspended over
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the RCS with the pressurizer temperature greater than 250' F and
220 psia, violating T5 2.11(1).
The NRC inspector reviewed the licensee's Procedures 01-RC-2B,
"Reactor Coolant Vent and Leak Test Instruction" and MP-HE-1, "Polar
Crane Annually or at Refueling Inspection." The NRC inspector
verified that the licensee had revised the two procedures to add
steps and precautions to prevent the movement of the polar crane over
the RCS when pressurizer pressure is greater than 225' F.
No
L
problems were noted.
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4.0 Followup to NRC Inspection and Enforcement Bulletins
4.0.1
(0 pen)IEB88-04: Adequacy of SI pump recirculation line.
This NRC Bulletin involved the problem of hydraulic Instability or
l
impeller recirculation due to miniflow recirculation operation with
parallel pump operation through a coninon recirculation line.
Evidence suggests that operation of pumps at some point below the
best efficiency point can result in pump damage due to vibration,
excessive forces on the impeller, and cavitation.
>
l
As of the followup inspection, the licensee had responded to
l
IEB 88-04 but had requested an extension in order to compile the
requested calculations and information. This item shall remain open
pending review of the licensee's final response.
5.0 Followup to 10 CFR Part 21 Reports
5.0.1
(Closed) Evaluation of licensee's reponse to 10 CFR Part 21 reports:
87-11; 87-12; 87-13, 87-14; 87-15; 87-32; 87-33; 87-35; 87-41; 88-05;
88-06; 88-34; and 88-0155. The NRC inspector's review was a
continuation of the review documented in NRC Inspection
Reports 50-285/87-25 and 50-285/87-33.
The NRC inspector reviewed a selected sample of the avaiiable
documentation for evaluations performed by the licensee for 10 CFR Part 21 reports made by other users (licenseas) and equipment
l
suppliers and vendors.
The evaluations were performed to determine
i
the applicability of the identified problem to the safe operation of
'
the FCS facility. Based on this review, it appeared that the
licensee was performing an adequate review.
The evaluations reviewed by the NRC inspecter are listed below:
1.icensee Identification
User, Vendor, or Supp'iier
Subject
10 CFR Part 21 87-11
Indiana Electric / Terry
Improperly
Corp. machined part
for AFW turbine
speed controller
10 CFR Part 21 87-12
Virginia Electric /
Defective steel
Rockwell/ Inland Steel
with laminations
in web
10 CFR Part 21 87-13
Morrison-Knudsen/
Residual Magnetism
Square O causes 125 volt
relays to fail
(remain energized)
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19
,
10 CFR Part 21 87-14
SMUD/Limitorque
Warped limit switch
rotors
10 CFR Part 21 87-15
High humidity
effects on SPEC
200 I/V cards
10 CFR Part 21 87-32
HFA Armature binding
due to incorrect
location of stop
tabs
10 CFR Part 21 87-33
Niagara Mohawk /Agostat
GP Series relays
improper seating
10 CFR Part 21 87-35
Toledo Edison /Limitorque
Inadequate instruc-
tion to maintain
balance
10 CFR Part 21 87-41
Technology for
Model 914-1 valve
Energy (TEC)
flow monitor module
fails to reset
MAL 88-05
Exo-Sensor
Containment
Hydrogen Analyzer
Excessive
Calibration Gas
Leakage
MAL 88-06
Arizona Power & Light /
Fasteners on MOV not
Borg-Warner
in accordance with
design requirements
MAL 88-34
Nothern States Power /
Insulation Damage
on motor lead wires
RRD 88-0155
HFA Relays Latch
Engagement less than
minimum requirement
6.0 Followup to Temporary Instructions
6.0.1
(Closed) TI 2518/64R1, items 3.1.3, 3.2, 4.2.1, 4.2.2, 4.5.1;
GL 83-28 followup.
During the followup inspection, the NRC inspector verified that the
licensee had responded, as required by GL 83-28, and that the NRC had
reviewed and accepted the responses cited above.
GL 83-28 Items 2.1,
4.5.2, and 4.5.3 remain open pending receipt of the NRR Safety
Evaluation Report and NRC inspector review.
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20
6.0.2
(Closed) TI 2515/91 Item 4.1:
Reactor Trip System Reliability.
The 7:RC inspector reviewed the licensee response, the NRC's review,
and acceptance via safety evaluation report dated January 6, 1986.
No further action on this item is required.
7.0 Followup to Sa~ety Systems Outage Modifications Inspection (SS0MI)
7.0.1
During the SS0MI, the NRC inspector identified numerous observations.
These observations represent cases where it is considered appropriate
to call attention to matters that are not deficiencies or unresolved
items. As such, observations include items recommended for licensee
consideration but for which there is no specific regulatory
requirement.
The below listed SSOMI observations are hereby closed
administratively.
8529/2.3-1
8529/2.8-1
8529/2.3-2
8529/2.8-2
8529/2.3-3
8529/2.10-1
8529/2.4-1
8529/2.11-1
8529/2.4-2
8529/2.12-1
8529/2.4-3
8529/2.5-1
8529/2.5-2
8529/2.5-3
8529/2.6-1
8529/2.7-1
7.0.2
(0 pen) Deficiency (285/8529/2.5-3):
Inadequate support of seismic
instrumentation tubing near air regulators.
This SSOMI deficiency
noted that seismic supports were not installed as required by
guidelines in that the allowable span was exceeded.
The licensee found that the guidelines were not followed by its
personnel. The particular configuration was analyzed and the
excessive span was found to be acceptable.
The NRC inspector
reviewed Calculation No. FC-83-158.
The licensee also conducted
training of its personnel in regard to span requirements. The
reccrds of the training were reviewed by the NRC inspector.
(See the
coments below on corrective action in paragraph 7.0.5)
'
7.0.3
(0 pen) Violation (285/8529-!!.F.2.g (Deficiency 8529/2.5-4)):
Discrepancies were found with the installation of safety injection
tank relief valves.
The SSOMI team fcund that relief valves were
interchanged between two tanks and, as a result, the valve tags were
not correct.
The weld of the lower pipe on Tank B had a crater pit,
and the tail piece pipes for Tanks B and D had surface
discontinuities.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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21
The licensee found that the valves in question were identical and
indeeJ switched during installation because of the lack of attention
to detail by personnel. The applicable Process and Instrument
Diagram (P&ID) No. E-23866-210-130, was revised to show the as-built
configuration. The NRC inspector verified the drawing change
(Revision 36).
The licensee did analyze the surface discontinuities
and found the condition acceptable.
The NRC inspector reviewed
Calculation No. FC-84-61.
The licensee also conducted training of
its maintenance personnel in regard to the above problems.
The NRC
inspector reviewed these training records (FC-7-361-87).
The welding
coments below on corrective actions in paragraph 7.0.5) (See the
problem is discussed in this report in paragraph 7.0.4.
7.0.4
(0 pen) Violation 285/8529-II.H(Deficiency 8529/2.6-1):
The
controls of welding and nondestructiva examination were inadequate.
This 550MI finding detailed that a previously accepted weld on Safety
Injection Tank B had an unacceptable crater pit, a previously
accepted socket weld on Valve No. MS-100 was found unacceptable and
had to be repaired, and the dye penttrant inspection performed on the
component cooling water flow element was accomplished below the
temperature limits allowed by procedures.
In regard to the Safety Injection Tank B union weld, the licensee
determined that the crater pit was not sufficient to reject the weld.
A penetrant examination, which accepted the weld after the area in
question was ground, dated May 21, 1987, was reviewed by the NRC
inspector.
There appears to have been confusion about the inspection
of this weld by OPPD at the request of the SSOMI team on December 16,
1985. The SS0MI team understood that a penetrant exam had been
requested and OPPD reportedly performed a visual exam. The NRC
inspector found that there were no records of either type of exam.
(See violation below on records in paragraph 7.0.6.)
In regard to the MS-100 socket weld, the licensee found that a highly
qualified contract inspector missed the crater pit in this weld.
After reinspection, OPPD concluded the weld was unacceptable and
repaired the weld. Maintenance Order No. 857783 to repair the weld
and the report of final acceptance after repair, were reviewed by the
NRC inspector.
The training and certification records of inspection
personnel were reviewed by the NRC inspector.
The training was both
generalized and also specific in regard to this problem.
In regard to the component cooling water penetrant examination, the
licensee found that, indeed, procedure requirements were violated by
inspection personnel.
A second penetrant examination found
unacceptable indications. A maintenance order was opened, which had
the weld in question filed to remove rust before a third examination.
The third examination found the weld to have no apparent indications.
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22
!
A concern of the NRC inspector is the logic used to resolve questions
of weld quality by filing or grinding before penetrant examinations,
as was the case on both the Safety Injection Tank B and component
cooling water welds. The licensee now requires the recording of the
temperature of the examined hardware on the inspection report.
The
licensee's response letter indicated that the temperature limits
would be on the form.
This will be clarified with another revision
to the response letter to delete reference to the form in that the
limits are in the ]rocedure.
(See coments below on corrective
action in paragrap1 7.0.5.)
7.0.5
(0 pen) Violation 285/8529-II.H.2 (Deficiency 8529/2.5-1):
Inadequate welding, preparation, and inspection associated with the
replacement of Valve No. MS-100. As noted earlier (7.0.4), the
'
replacement of Valve No. MS-100 was unacceptable.
In addition,
'
during the repair the wall thickness was violated.
The licensee found that tightly adhered slag in the crater masked the
crater pit.
During the repair, the craftsman became overly
aggressive and removed too much wall.
The thinned wall condition was
evaluated.
The NRC inspector reviewed the calculations of that
1
evaluation. As noted earlier, the weld in question was successfully
repaired and finally accepted.
The licensee has also established a
program to control welding.
Specification No. CTS-4 and 50 Nos. 72,
,
^
72A, and 72B have been implemented to provide further controls of the
welding process. The NRC inspector reviewed this current program.
(See comments below on corrective action.)
The licensee's quality assurance program description states that, a
"thorough investigation and documentation of significant conditions
adverse to quality" will be required.
in review of the above items,
it was noted that the corrective actions taken to date failed to
address the potential for previous events of a similar nature.
Because of this lack of generic corrective action, the items in
paragraphs 2.1.6, 7.0.2, 7.0.3, 7.0.4, and 7.0.5 cannot be closed.
These items will remain open until corrective action is taken in
1
regard to the potential for previous events.
In particular, at a
minimum, the enumerated questions related to the following areas
should be considered:
a
Design Modification Package No MR 84-162 failed to provide
sufficient detail in regard to coatings because personnel failed
'
to follow procedures.
Corrective actions to consider:
(1) Are there other
modifications that have been completed where the packages had
insufficient information on coating materials and the impact of
such?
(1) is this problem limited to only coating information
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23
in modification packages and not other information? (3) Has the
person involved failed to identify sufficient information in
other modification packages he worked on? (4) Is there a
problem with the review process of packages?
Seismic supports for Valves YCV-1045 A and B, installed by
Modification Package MR 81-158, exceed span requirements and the
design had to be reanalized because tubing support guidelines
were not followed.
Corrective actions to consider:
(1) Are there other
modification installations of air supply) tubing where span
f
requirements may have been violated? (2 Are there any other
seismic requirements that were not followed?
(3) Have the
personnel in q)uestion installed other tubing without sufficient
supports?
(4 Is there an inspection problem with verifying
other requirements?
A change had to be made to P&IO No. E-23866-210-130 because
Valve Nos. SI 221 and SI 217 were interchanged because of lack
of attention to detail upon installation under Modification
Package No MR 84-64.
Corrective actions to consider:
(1)Couldtheinstallationcrew
have failed to attend to other details? (2) Has the crew
,
installed other modifications and errored in the same way?
(3)
'
Do other P& ids have incorrect identification information?
(4) Why wasn't this identified by QC7
'
A weld on Valve No. MS-100 installed by Modification
Package No. MR 85-042 had to be repaired after a highly
qualified inspector missed a crater pit in the original
examination,
t
'
Corrective actions to consider:
(1) What other modifications
were done where inspections were performed by the same personnel
and are these welds acceptable? (2) What other modifications
were done where the same welding procedure and/or personnel were
used and are these welds acceptable?
'
'
A penetrant examination on the component cooling water flow
transmitter installed by Modification Package No. MR 85-062 was
'
performed a second time because the personnel failed to follow
temperature limits.
Corrective actions to consider:
(1)Arethereother
modifications in which penetrant test limits could have been
violated?
(2) Are there other welds inspected by the same
personnel where procedure requirements were not followed?
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24
7.0.6
(Closed) Violation 285/8529-II.F.2.m(Deficiency 8529/2.5-7):
Weld inspections were not accomplished as required.
This SSOMI
deficiency noted that the welds of the transformer base to embedment,
controlled by Modification Package No, MR84-105, had not been
performed.
The licensee found that the planner and craftsman were not fully
aware of the inspection requirements for welding of seismic supports.
The NRC inspector found that, indeed, the "weld and test control
record" and the associated form were signed off for all six
However, there were no visual examination reports for
all six weldments. The inspection log book indicated that the
inspections of Transformer Nos. TIB 38 and 3C were performed on
December 18, 1985.
These inspections were not documented as required
by procedures.
The licensee did revise its Specification No. CTS-2
i
and Procedure No. 50 72A. Training was administered on this problem.
The NRC inspector reviewed the revised specification and procedure as
well as the training records to these revised-documents.
In review of the above and the item in paragraph 7.0.4, the NRC
inspector found that inspection records of SSOMI followup were not
maintained on Visual Inspection Form FC-182.
In particular:
(1) the
inspection of the Safety Injection Tank B union weld that took place
on or about December 16,1985; and (2) the inspection of Transformers
i
!
Nos. TIB 3B and 3C base welds that took place on December 18, 1985.
As such, the licensee's failure to follow procedures is an apparent
violation of S0 No. 26, Appendix G. Revision 26, dated August 7,
1988 (285/8825-01).
l
7.0.7
(Closed) Violation 285/8529-II.F.1.j (Deficiency 8529/2.5-6):
i
Procedures did not provide instructions for a flanged joint.
This
i
SSOMI deficiency identified that a flanged joint was leaking and
found to be out of parallel by .030 inch.
'
l
The licensee believed that the leakage occurred some time after the
joint was made up. The joint was retightened under Maintenance Order
i
No. 857887 to 60 foot-pounds.
The leakage stopped. The torque
requirements were based on vendor data.
It should be noted that
prior to the SSOMI observation, the system had been pressure tested
at 188 psi on November 11, 1985.
The NRC inspector reviewed the
j
above maintenance order and the original Modification Package
No. MR 85-62.
I
7.0.8
(Closed) Unresolved Item (8529/02.3-2):
Lack of calibration record
i
for a pressure source used for safety-related channel calibration.
This unresolved item involved the lack of evidence that the
calibration records existed for a pressure source used in the loop
calibration procedures for CP-D/102-2 (pressurizer pressure).
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.
,
.
25
During the followup inspection, the NRC inspector reviewed the
licensee's response to the SS0MI item and Operations Incident
Report 2225, dated November 29, 1985.
The licensee's records
indicated that the pressure transmitters were calibrated using both
Pressure Source Nos. 238 and 246. However, Operations Incident
Report No. 2225 discovered that even though Pressure Source No. 246
had been calibrated prior to usage, the QC verification was not
accomplished prior to usage.
Failure of QC to verify the calibration
of Pressure Source No. 246 is an apparent violation of NRC
Regulations and 50 No. M-26.
The NRC inspector reviewed the
licensee's followup and corrective actions taken. The NRC inspector
verified that the licensee had reviewed their records and QC
accomplished an after-the-fact verification on November 29, 1985, and
found the calibration to be valid.
Furthermore, the licensee implemented measures for key supervisors to
conduct refresher training of I&C technicians.
The training was
conducted to ensure test equipment is calibrated before and after
usage. Also, the training was conducted to ensure test equipment
identity is properly recorded on the calibration form at the time the
procedure is performed.
It is concluded that, in fact, the licensee
failed to follow S0 No. M-26, Section 3.3.3 in that QC verification
signoff of the Pressure Source No. 246 calibration was not
accomplished prior to usage.
However, the licensee had performed
effective and comprehensive corrective actions to resolve the cited
condition and preclude future recurrence.
7.0.9
(Closed) Violation 285/8529-II.F.1.c (Deficiency 8529/2.8-2):
Test procedure did not verify design concept under accident
conditions. This deficiency involved the inadequacy of the
licensee's test procedure to verify Modification Installation
No. MR 84-74A, fuse protection for limit switches satisfied the
intended design concept under accident conditions.
During the followup inspection, the NRC inspector verified that the
licensee had revised the procedure for functionally testing
Modification No MR-FC-84-74A.
The revised procedure functionally
tested the applicable protective fuse in a configuration which
verified the system under accident condition.
This testing was
successfully completed on December 20, 1985, in accordance with
Maintenance Order No. 857847,
7.0.10
(Closed) Violation 285/8529-!!.G.4(Deficiency 8529/2.3-6):
Calibration procedure changes without approved field changes.
This
deficiency involved the licensee's tailure to follow procedures in
that numerous instances were found in calibration procedures for
CP-X/905 and CP-X/902 where revisions were not made in accordance
,
!
.
O
I
,
.
26
During the followup inspection, the NRC inspector reviewed the
applicable calibration procedures and records and verified that the
records had been properly annotated.
Furthermore, discussions with
the licensee showed that a substantive effort is being made to
conduct initial training of personnel and conduct a continuous 2-year
refresher training for personnel in each craft area.
This refresher
training covers standing orders and their applicability to the
specific crafts.
In addition, the licensee had initiated a program
to provide both contract and licensee personnel with an infonnation
reference booklet, which references some 30 different standing orders
and the proper procedures to address.
7.0.11
(0 pen) Viol
an 285/8529-II.F.1 k; II.H.5 (Deficiency 8529/2.5-6):
Installatt
discrepancies found in installation of New Delta T
power prou.,s loop instrumentation.
This item related to
discrepancies found in wiring installation and seismic support
installation with Modification Request Package 84-140.
During the following inspection, the NRC inspactor reviewed
documentation that indicated previously identified installation
discrepancies had been corrected.
However, due to the operating
status of the plant (100 percent power level), a walkdown inspection of the
instrumentation cabinets to verify the corrected conditions was not
performed.
This violation shall remain open pending a walkdown
inspection of the corrected condition.
7.0.12
(Closed) Violation 285/8529-II.F.2.h(Deficiency 8529/2.3-4):
Training not done prior to approval of procedure change. This
violation involved the failure of the licensee to complete required
training prior to the issuance of Procedure Change 13494 to 01-FW-3.
This was contrary to the requirements of Step 3.6.1(d) of
SO No. G-30.
The licensee has emphasized the importance of ensuring that necessary
training is accomplished prior to implementing a procedure change, in
accordance with 50 No. G-30, with both training department personnel
and members of the plant review committee.
The licensee conducted a
review in 1986, and identified 2 other instances in which a procedure
change requiring pre-approval training was implemented before the
training was completed. No problems were noted.
7.0.13
(Closed) Violation 285/8529-11.1.1; 11.1.2 (Deficiency 8529/2.9-2):
Inadequate warehouse storage of safety-related material.
This
violation was related to material tags that did not agree with
material markings and documentation, critical quality element (CQE)
material that was stored in Level C areas rather than the required
Level B areas, and incomplete material certifications for 1 3/8" by 8
nuts.
.
.
.
en
,
,
,
.
27
In response to this violation, '.he licensee has taken several
corrective actions. On July 18, 1988, the licensee revised
S0 No G-22, "Storage of Critical Elements and Radioactive Material
Packaging, Fire Protection Material, and Calibration Equipment."
,
This revised procedure along with Quality Assurance Department
!
Procedure QADP-12. "Material / Service Acceptance and Receipt
Inspection," which was revised on May 18, 1988, adequately addressed
the identified deficiencies.
(The NRC inspector indicated the
licensee should consider including QADP-12 under "References" in
Section 1.3 of S0 No. G-22.)
A material verification team made up of four full-time members was
established in July 1988 to verify the information entered on tags
i
for 4200 CQE items.
The team, led by a licensee QA inspector, had
reviewed about 80 percent of the CQE items; almost 300 discrepancies
had been found. According to the operations quality assurance
supervisor, the team will complete its review on or about
September 1, 1988.
On August 12, 1988, tge licensee took possession of a recently
constructed 40,000 ft warehouse. About one-half of the warehouse
will be dedicated to storage.
Storage racks were being installed to
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2
increase the storage area to about 60,000 ft .
The warehouse was
incorporated into the protected area on August 15, 1988. The
licensee plans to complete the transfer of items from the old
warehouse to the new one by mid-September 1988.
The licensee recently hired a material control supervisor, who brings
to the job 12 years of material control experience in the nuclear
field.
8.0 Unresolved Item
Unresolved items are matters about which more information is required in
'
order to ascertain whether or not the items are acceptable, violations,
or deviations. The following unresolved item was discussed in this
report:
Paragraph
item
Subject
2.5
285/8825-02
Reportability of design base
inadequacies
9.0 Exit interview
An exit interview was conducted with FCS personnel on August 12,1988, at
the conclusion of the onsite inspection, during which the inspection
findings were sumarized.
The licensee did not identify as proprietary
any of the materials provided to, or reviewed by, the NRC inspectors
during the inspection.
.
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Criterion V of Appendix B to 10 CFR Part 50 and the licensee's approved
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quality assurance program require that the activities affecting quality be
__
accomplished in accordance with documented instructions. Visual
Examination Procedure, Standing Order No. 26, Appendix G, Revision 26
-
dated August 7, 1988, requires that a Visual Weld Examination Report
-
Form No. FC-1103 be prepired after an examination.
_
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._
of the safety system outage modification inspection followup for Safety
Inspection Tank B union weld and the TIB 3B and 3C transfonners' base
-
welds were not prepared.
_
This is a Severity Level V violation.
(SupplementI.D.)(285/8825-01)
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