ML20199F437

Forwards Safety Sys Outage Mod Insp Rept 50-285/85-29 on 851106-08,18-22 & 1209-17.Potential Enforcement Actions Listed on App,Including Nonconforming Installations,Will Be Referred to Region IV for Appropriate Action

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: IR 05000285/1985029 ML20199F437
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

PDR

ADOCK 05000205

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Copy to RCPB, IE

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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

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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

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1623 Harney Street

Omaha, Nebraska 68102

. . .

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Omaha Public Power District

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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

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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

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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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