ML20199F437
| ML20199F437 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 03/19/1986 |
| From: | Taylor J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | Reznicek B OMAHA PUBLIC POWER DISTRICT |
| Shared Package | |
| ML20199F443 | List: |
| References | |
| NUDOCS 8603280211 | |
| Download: ML20199F437 (9) | |
See also: IR 05000285/1985029
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UNITED STATES
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NUCL2AR REGULATORY COMMISSION
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WASHINGTON, D. C. 20555
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Dockst No. 50-285
March 19, 1986
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Omaha Public Power District
ATTN: Mr. Bernard W. Reznicek
President and Chief Executive Officer
1623 Harney Street
Omaha, Nebraska 68102
Gentlemen:
SUBJECT: SAFETY SYSTEM OUTAGE MODIFICATION INSPECTION (INSTALLATION
AND TEST) 50-285/85-29
This letter conveys the results and conclusions of the installation and test
portions of the Fort Calhoun Station Safety Systems Outage Modification
Inspection conducted by the NRC's Office of Inspection and Enforcement. The
inspection team was composed of personnel from the NRC's Office of Inspection
~and Enforcement, Region IV and consultants. The inspection took rlace at the
Fort Calhoun Station and at your offi.ces in Omaha, Nebraska. This inspection
was part of a trial NRC program being implemented to examine the adequacy of
licensee management and control of modifications performed during major plant
outages.
The purpose of the installation and test portions of the Trial Safety Systems
Outage Modification Program was to examine, on a sampling basis, installation
and testing of plant modifications accomplished during the September 1985-
January 1986 outage at Fort Calhoun Nuclear Station. This portion of the trial
program concludes the inspection program at Fort Calhoun.
Reports forwarding
the results of the design inspection and the vendor inspections have already
been issued. The applicable report numbers are provided in Section 3 of the
report.
Section 2 of the report is the detailed discussion of the installation and
testing inspection. The effort was hardware and test oriented and centered
around 18 modifications accomplished during the outage.
Particular attention
was directed toward adequacy of installation procedures, conformance of the
modifications to requirements, adequacy of functional tests, material control,
and safety-related maintenance activities.
Section 1 of the report is a summary of the results of the inspection and the
conclusions reached by the team. The most significant concerns identified were
examples in the areas of:
(1) lack of engineering safety evaluations for
design changes, (2) nonconforming installations, (3) inadequate quality
control, and (4) inadequate material control.
8603280211 860319
ADOCK 05000205
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PDR.
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Omaha Public Power District
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March 19, 1986
During this inspection the NRC inspection team performed a preliminary review
of your planned corrective actions for the significant findings which has been
identified at an interim status briefing on October 8, 1985 for the design
portion of the Safety Systems Outage Modification Inspection.
In addition,
Section 1.4 of this report discusses corrective actions for specific items
which the OPPD r,epresentatives indicated, during the exit meeting of December
18, 1985, would be. corrected prior to plant startup following the outage.
The Appendix to this letter contains a list of potential enforcement actions
which are based on the deficiencies identified during the installation and
testing inspection.
These will be reviewed by the Office of Inspection and
Enforcement and the NRC Region IV office for appropriate action. At the
completion of that review, the Region IV office will issue any enforcement
actions resulting from the installation and testing inspection, as well as from
the earlier design and vendor inspections.
In addition, Region IV will monitor
your corrective actions relating to those enforcement actions.
In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosures
will be placed in the NRC Public Document Room.
No reply to this letter is
required at this time.
You will be required to respond to these findings
after a decision is made regarding appropriate enforcement action.
Should you have any questions concerning this inspection, please contact
me or Mr. James Konklin (301-492-9656) of this office.
Sincerely,
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mes M. Tay
, Director
ffice of I pection and Enforcement
Enclosures:
1.
Appendix, Potential Enforcement Actions
2.
Inspection Report 50-285/85-29
cc w/ enclosures:
-See next page
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Omaha Public Power District
-3-
March 19, 1986
cc w/ enclosure:
Harry H. Voigt, Esq.
LeBoeuf, Lamb, Leiby & MacRae
1333 New Hampshire Avenue, N.W.
Washington, D.C.
20036
Mr. Jack Jensen,' Chairman
Washington County Board
of Supervisors
Blair, Nebraska 68023
Metropolitan Planning Agency
ATTN:
Dagnia Prieditis
7000 West Center Road
Omaha, Nebraska 68107
Mr. Phillip Harrell, Resident Inspector
U.S. Nuclear. Regulatory Commission
P. O. Box 309
Fort Calhoun, Nebraska 68023
Mr. Charles B. Brinkman, Manager
Washington Nuclear Operations
C-E Power Systems
7910 Woodmont Avenue
Bethesda, Maryland 20814
Regional Administrator, Region IV
U.S. Nuclear Regulatory Commission
Office of Executive Director
for Operations
611 Ryan Plaza Drive, Suite 1000
Arlington, Texas 76011
Mr. William C. Jones
Vice President, Nuclear Production,
Production Operations, Fuels, and
Quality Assurance and Regulatory
Affairs
Omaha Public Power' District
j
1623 Harney Street
Omaha, Nebraska 68102
. . .
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Omaha Public Power District
-4-
M reh 19, 1986
DISTRIBUTION:
DCS (Docket No. 50-285)
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APPENDIX
POTENTIAL ENFORCEMENT ACTIONS
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As a result of the NRC Trial Safety Systems Outage Modification Installation
and Test Inspections at Fort Calhoun during November 6-8, November 18-22,
and December 9-17, 1985, the following items are being referred to Region IV
as, Potential Enforcement Actions.
Section references are to the detailed
inspection report.
1.
10 CFR 50.59 requires that safety evaluations be accomplished for
temporary or permanent design changes to the facility to determine
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whether an unreviewed safety question exists or whether a change to
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the Technical Specifications is involved.
Contrary to the above, the NRC inspectors found that'the licensee's
procedures for accomplishing engineering safety evaluations were not
effectively implemented in that;
a.
No safety evaluations had been accomplished for installation-of
lead shielding on safety related piping for at least the last two
^
and one half years (Section 2.2.1),
b.
No safety evaluation was available for a design change involving
a penetration through a fire barrier which had been completed for
several years (Section 2.2.2).
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c.
Safety-related electrical jumpers had.been installed as long as 18
months without documeated safety evaluations (Section 2.2.3).
2.
10 CFR 50, Appendix B, Criterion IX, as implemented by QAM Section'10,
requires that measures be established to assure special processes,
including, welding and nondestructive testing are accomplished using
qualified procedures in accordance with applicable codes, standards,
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specifications, criteria and other special requirements.
Contrary to th9 above, the NRC inspectors found the licensee's program
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for control of welding and nondestructive examination was inadequate
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in that:
a.
A standard flat plate 90 fillet weld procedure was used to
accomplish skewed fillet welds, plug welds, pipe boss attachment
welds and seal welds in two modification packages installed during
this outage (Section 2.6.2).
b.
An unacceptable crater pit and other surface discontinuities were
found in previously accepted welds on SIT relief valve union
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installations (Sections 2.5.3 and 2.6.1).
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c.
A safety-related nonisolable. socket weld was accepted by dye penetrant
_ inspection when, in fact, the weld was unacceptable both visually
and by subsequent dye penetrant inspection for a modification package
installed during this outage (Sections 2.5.1 and 2.6.1).
d.
Dye penetrant inspections were tcund to have been accomplished,
and accepted, at surface temperatures below the minimum allowed
by procedures when, in fact, the welds were unacceptable by
reinspection above the r;inimum temperaiore for a modification
package installed duriag the outage (Sectior. 2.6.1).
e.
Welds on seismic conduit supports and installation of the conduits
and supports did not conform to the installation procedure design
details for a modification package installed during the outage
(Section 2.5.5).
3.
Fort Calhoun Technical Specificatic,ns, Section 2.19(8) requires that
a continuous fire watch be posted and backup fire suppression equipment
be provided when the Haloa fire suppression system is disabled in the
switchgear room.
Contrary to the above, the NRC inspectors found this requirement was
not implemented by the licensee when no continuous fire watch or backup
fire suppression equipment was provided in the switchgear room from
December 6-10, 1985 with the Halon fire suppression system disabled
(Section 2.4.2).
4.
10 CFR 50, Appendix B, Criterion XIII, as implemented by QAM Section 14,
requires that measures be established to control the storage of materials,
provide proper protection, and provide correct environmental conditions.
Appendix A to the Fort Calhoun Updated Safety Analysis Report (USAR)
commits OPPD to ANSI N45.2.2-1972, " Packaging, Shipping, Receiving,
Storage and Handling of Items for Nuclear Power Plant." ANSI N45.2.2
requires stored materials to be adequately protected, to be located in the
correct storage environment according to material quality, and to be
properly identified with quality assurance acceptance tags.
Contrary to the above, the NRC inspectors found the licensee's program
for control of material in storage to be inadequate in that:
a.
Safety-related cable was found damaged in a temporary safety-
related storage area (Section 2.9.1).
b.
Level B safety-related material was found stored in a Level C
storage area for up to 19 months (Section 2.9.2).
c.
Safety-related inaterial was found with identification tags that
did not agree with material markings or other material documentation
(Section 2.9.2).
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d.
Material was found in a temporary safety-related storage area without
quality assurance . acceptance tags -(Section 2.9.1).
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e.
Nonsafety-related material was found stored in a safety-related
storage area (Section 2.9.1).
f.
Quality control surveillances of temporary safety-related storage
areas were not accomplished on the required monthly basis (Section
2.9.3).
5.
10 CFR 50, Appendix B, Criterion VI, as implemented by QAM, Section 7,
requires. that measures. be established -to control issue of procedures and
drawings'and that changes to these documents be reviewed and approved by
authorized personnel and distributed to the location of the prescribed
quality activity.
Contrary to the above, the NRC inspectors found that 'the licensee's
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document control programs were not effectively implemented in that they:
a.
Failed to adequately control drawings used for construction
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(Section 2.3.1 and 2.3.6).
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b.
Failed to adequately control field changes to installation
procedures (Section 2.3.2).
c.
Failed to adequately control field changes to calibration procedures
(Section 2.3.7).
d.
Failed to provide .the required review of a change to an operating
procedure prior to implementation (Section 2.3.3).
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Failed to provide training associated with a procedure change prior
to implementing the change (Section.2.3.4).
6.
10 CFR 50, Appendix B, Criterion XVI, as implemented by QAM Section 17,
requires that measures be established to assure that identified deficien-
cies adverse to quality are promptly identified and corrected.
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Contrary to the above, the NRC inspectors found that:
a.
An adequate program for control of installation of lead shielding
was not implemented after inspections by INP0 in 1982 and 1984 that
identified deficiencies in the program, and after issue of an IE
Information Notice in 1983 addressing the installation of lead
shielding (Section 2.10.1).
b.
No program existed for resolution of discrepancies identified by the
System Acceptance Committee for those plant modifications accepted
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for system operation by the committee with outstanding discrepancies
(Section 2.10.2).
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7.
10 CFR 50, Appendix B, Criterion V, as implemented by QAM Section 6,
requires that activities affecting quality be described by documented
instructions, procedures or drawings and be accomplished.in accordance
with these instructions, procedures and drawings..
Fort Calhoun . Technical Specifications, Section 5.8.1, requires that
written procedures be established that' meet or exceed the minimum require-
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ments of ANSI N18.7-1972, " Administrative Controls and Quality Assurance
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for the Operational Phase of Nuclear Power Plants," Section 5.3.
ANSI
N18.7Property "ANSI code" (as page type) with input value "ANSI</br></br>N18.7" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., Section 5.3, requires that activities affecting safety be described
,
by written instructions, procedures or drawings, and be accomplished in
accordance with these instructions procedures ~or drawings.
Contrary to the above, the NRC inspectors found that:
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a.
The installation procedure for replacement of a nonisolable safety-
related valve did not provide sufficient work step detail to assure
ddequate Conduct of safety-related maintenance activities (Section
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2.4.1).
Associated problems were identified with completed valve
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replacement accomplished during this outage (Section 2.5.1).
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b.
The installation procedure for installation of safety-related seismic
instrumentation tubing did not provide installation criteria for the
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tubing or seismic supports and did not reference the applicable
Stone and Webster' guideline for installation of seismic tubing and
supports (Section 2.4.1).
The support requirements specified -in the
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guideline were vio' 2.ted a number of times for one modification
package installed during the outage (Section 2.5.2).
c.
Ar installation procedure which included makeup of a flanged joint
d o not provide instructions or provide reference to another instruc-
tion for proper make'up of a flanged joint (Section 2.4.1). Discrepan-
cies were identified with the completed flange installation accom-
plished during this outage (Section 2.5.4),
d.
Safety-related cables were tie-wrapped to nonsafety-related cables
in two electrical panels for cne modification package installed
during the outage (Section 2.5.5).
e.
Procedures did not provide adequate instructions for installation of
air accumulator tanks, adequate instructions for protection of SIT
relief valve 0-rings during welding, adequate inspection requirements
for welding of 4160/480 volt transformer bases, adequate criteria for
inspection of cable splices, adequate requirements for verifying
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acceptance during a battery charger load test, or adequate require-
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ments for testing of fuse protection for limit switches (Sections
2.4.1, 2.5.6, 2.8.1 and 2.8.2).
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f.
Instances in which~ procedure requirements were not followed included
passing a QC hold point prior.to drilling stud holes through the
battery room wall, using other than Level III inspectors to review
and approve procedures, tagging out breakers without documented
-shift supervisor review, installing pipe unions. incorrectly to SIT
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relief valves, and incorrectly identifying installed valves
(Sections 2.4.2 and 2.5.3).
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