ML20213G420

Insp Rept 50-285/87-05 on 870406-10.Violations Noted:Failure to Include Criteria to Permit Determination of Acceptability in Procedures,To Identify & Retrieve Records & to Perform Design Reviews

ML20213G420
Person / Time
Site:	Fort Calhoun
Issue date:	05/08/1987
From:	Boardman J, Hunnicutt D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20213G411	List: ML20213G407 ML20213G420
References
50-285-87-05, 50-285-87-5, NUDOCS 8705180280
Download: ML20213G420 (19)

See also: IR 05000285/1987005

Text

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-* ' APPENDIX'

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,U.S. NUCL' EAR REG LATORY.. COMMISSION

REGION IV-

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-NRCLInspection Report ~ 50-285/87-05: License: DPR-40

Docket: 50-285

Licensee: Omaha Public Power District (0 PPD)

, 1623 Harney Street

' Omaha, Nebraska. 68102

Facility Name: FortCalhounStation(FCS)

Inspection At: Fort Calhoun Station, Blair, Nebraska

Inspection Conducted: April 6-10,1987

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-Inspector: h /-

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). R. Ebardman, Reactor Inspector '

Date[

Operations Section, Reactor Safety Branch

Consultant: H. S romberg, EGG Idaho, Inc./INEL-

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Approved: /' N

. M. & inicutt, Chief, Operations Section Date /

Reactoi afety Branch

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t Inspection Summary

Inspection Co'nducted April 6-10, 1987 (Report 50-285/87-05)

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Areas Inspected: Licensee maintenance program, maintenance program

. implementation, electrical maintenance, and instrumentation and control systems

-maintenance.

l Results: Within the four areas inspected, five potential violations were

identified.as follows:

1. . Failure to include in p' rocedures the criteria to' permit determination of

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acceptability (paragraphs- 3.b.1, 3.b.5, 4.c, 4.d, 5.c, and 5.q).

.2. Failure to. identify and retrieve records (paragraphs 3.b.2, 3.c.2, and

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3.d.3.a).

8705180200 870508 5

DR ADOCK 050

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3. -Failure to perform design review (paragraphs 3.b.4, 3.b.5, 3.c.3,'and

3.d.3.b).

4 .~ Failure to implement procedures (paragraphs 4.a and 4.e.1).

5. Failure to use required lubricant in Electrical Equipment Qualified (EEQ)

Limitorque valve operators (paragraph 4.1).

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DETAILS

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1. Persons Contacted

'*R.-Andrews, Division Manager, Nuclear Production

'W. Gates, Plant Manager-

• R. Mehaffey, Supervisor, General Station Engineering Electrical

• M. Core,. Supervisor, Maintenance

J. Fisicaro, Supervisor,' Nuclear Regulatory and Industry Affairs

J. Foley, Supervisor, I&C and Electrical Field Maintenance

• J. Gasper, Manager, Administrative and Training Services

• R. Jaworski, Section Manager, Technical Services

• K. Miller, Supervisor Mechanical Maintenance

• L. Kusek,'-Supervisor, Op'erations

.

• J. Tucker, GSE Electrical Engineer

• D. Munderloh, Plant Licensing Engineer

• T. McIvor, Supervisor, Technical.

• R.'Mueller, Plant Engineer

• A. Richard, Manager, Quality Assurance

T. McIvor, Supervisor Technical

• R. Drahota, Maintenance Engineer

J.-Folley, Electrical and I&C Supervisor

M. Ellis, I&C Supervisor.

R. Grimes, QC Supervisor

D. Golden, Maintenance Engineer

D. Burns, Electrical Foreman

NRC Personnel

• P. Harrell, Senior Resident Inspector

• Denotes attendance at the exit interview on April 10, 1987.

2. Licensee Maintenance Program

The NRC inspector reviewed the following procedures relating to the

maintenance program:

Procedure Revision /

Number Title Date Comments

G-17 Maintenance Order R40/10-15-86

G-62 Control of Vendor Technical R2/10-30-86 See Note 1

Information

M-2 Preventive Maintenance Program R8/4-16-85

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A-0-30 Testing of Safety-Related R3/9-30-80

Equipment

M-1 Maintenance Procedures R7/10-10-84

0-25 Electrical and Mechanical R19/8-18-86

Jumpers and Blocks Controls

G-23 Surveillance Test Program R20/9-10-86

G-17A Electrical Equipment in a Harsh R2/2-12-86

Environment

0-20 Equipment Tagging Procedure R17/9-20-85 Note 2

G-26 Maintenance Quality Control R16/10-16-85 Note 3

Program

G-56 Qualified Life Program (QLP) R6/2-12-86

SP-Vendor-1 Vendor Manual Indexing R1/10-31-85

SP-Vendor-2 Safety-Related Vendor Technical R2/1-30-87

Information Indexing and

Review

Notes: 1. Reference 3.2 of G-62 includes ANSI N45.2.9-1979 which is

not accepted by NRC. The correct revision of N45.2.9, which

is 1974, is identified in the body of G-62.

2. Sections 4.2.1.c, 4.1.8, and 4.2.1.a contains "should"

actions that appear to require "shall" instead.

3. Reference 2.5 is the 1976 transmittal of the QA Program

which is not the latest revision.

No violations or deviations were identified.

3. Maintenance Program Implementation

a. Background

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The NRC inspector reviewed the licensee's implementation of its

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maintenance program. On Tuesday, April 7, 1987, during a tour with

! the NRC Senior Resident Inspector (SRI), the NRC inspector noted

l licensee maintenance on the main feedwater system flow nozzles. The

NRC inspector reviewed the maintenance order (MO), which indicated

that retorquing of the line flange bolting would be by " slugging"

(use of a slug wrench) until the joints were tight. The NRC

inspector was concerned because this practice could result in bolt

stresses that did not meet design values. Allowable values would

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range from approximately 27,700 psi to 31,800 psi based on ASME

design stress intensity values for bolting materials in the 1965,

1968,-and 1980 (with addenda) ASME Boiler and Pressure Vessel Code

Revisions. The NRC inspector selected twn critical safety-related

applications for review. The main steam safety relief valve (MS SRV)

and pressurizer SRV line flange bolt design stress values and

resultant torque values.

b. Uncontrolled Stressing of MS SRV Line-Flange Bolts

(1) Governing Procedures

The NRC inspector requested the licensee procedure which covered

reinstallation of the MS SRVs. The licensee Procedure MP-MS-1,

Revision 13, dated March 19, 1987, " Main Steam Safety Valve

Inspection and Repair." This procedure covered torquing of the

MS SRV with the following steps:

"5.39 After offsite testing, reinstall valves with new

gaskets. Q.C. to closeout inspect piping."

No torque values were provided in Procedure MP-MS-1 to assure

that design bolt stress was achieved.

10 CFR Part 50, Criterion V, states that activities affecting

quality shall be prescribed by documented instructions and

procedures which shall include appropriate quantitative

acceptance criteria. The failure of licensee Procedure MP-MS-1,

-Revision 13, to include torque values to achieve appropriate

design bolt stress levels is a potential violation of 10 CFR

Part 50, Appendix B, Criterion V (285/8705-01). Other examples

of this potential violation are identified in Sections 3.b.5,

4.c, 4.d, 5.c, and 5.a of this report.

(2) Original Design and Installation Records

The NRC inspector requested the original design base calculation

and original installation records (" code data package") for the

MS safety relief valve line flange joint bolt stress level. The

licensee was unable to retrieve these records.

10 CFR Part 50, Criterion XVII, requires the maintenance of

sufficient records to furnish evidence of activities affecting

quality; it further requires that records be identifiable and

retrieveable. Failure to identify and retrieve the records is a

potential violation of 10 CFR Part 50, Appendix B,

Criterion XVII (285/8705-02). Other examples of this potential

violation are identified in Sections 3.c.2 and 3.d.3.a of this

report.

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-(3) Applicable Code and Specification Revisions for Piping .,ystem

Design

Fort Calhoun Station piping systems were initially designed to

USAS (now ANSI) 831.7 and ASA/USAS/ ANSI B31.1. Since Nebraska

is a code state, later piping design should have been to the

ASME Boiler and Pressure Vessel Code (BPV Code), when the BPV

Code was extended to piping systems. Licensee plant engineering

personnel were unable to identify to the NRC inspector which

revision of the code was being used by Fort Calhoun Station or,

during previous periods of time, which revisions were i

applicable. This is an unresolved item (285/8705-03) pending I

further review by the NRC inspector. .

(4) MO 16275 for Reiastallation of MS SRVs and Lack of Design

Control

The NRC inspector requested licensee M0s covering reinstallation

of the MS SRVs. The licensee was unable to retrieve any such

M0s prior to the NRC inspector leaving the site on April 10,

1987. This is another example of the licensee's inability to

retrieve records.

On April 15, 1987, via the NRC SRI, the licensee provided the

NRC inspector with a copy of MO 16275 accomplished in

March 1983, covering reinsta11ation of SRVs MS-275, 276, 277,

278, 280, 281, 282, 291, and 292.

This M0 contained instructions to torque the subject valve

in-line flange bolts 'to 750 foot pounds. The NRC inspector

questioned the basis for that torque value, but licensee

personnel did not provide it. Since design stress calculations

for these bolts was not available, the licensee was performing a

l, design function in determining these stress values.

10 CFR Part 50, Criterion III, requires design control,

including performance of design review. The licensee is been

committed to WASH 1283 in its Quality Assurance Program

Description. WASH 1283 contains ANSI N45.2.11 " Quality

Assurance Requirements for the Design of Nuclear Power Plants,"

(Draft 2, Revision 2, May 1973).

The licensee, in his latest approved Quality Assurance Program

Description (QAPD), Section A.4, is committed to Regulatory

Guide 1.64 and ANSI N45.2.11-1974 which specifies (Section 6)

that design verification is to be performed by individuals or

groups other than those who performed the original design.

There was no licensee documentation that such a design review

had been performed in determining the torque value for the MS

SRV line flange bolt stress valve.

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Failure to meet the requirements of ANSI N45.2.11 is a potential

violation of 10 CFR Part 50, Appendix B, Criterion III

.(285/8705-04). Other examples of this potential violation are

identified in Sections 3.b.5, 3.c.3, and 3.d.3.b of this

report.

(5) Deletion of a Design Requirement Without Proper Review

The torque valve provided on H0 16275 was a piping system design

parameter necessary to prevent poteritial failure of the joint

under design base (not necessarily operating) conditions.

The torque valves were marked through with the comment "cannot

be torqued." Based on licensee conversations with the NRC SRI,

the bolts were stressed in an uncontrolled manner, to an unknown

stress level, by the common plant procedure of using a slugging

wrench. Later M0s did not show torque values.

Deletion of the torque values constitutes another example of

potential violation (285/8705-01), failure to provide

appropriate acceptance criteria required by 10 CFR 50,

Appendix B, Criterion V; and potential violation (285/8705-04),

failure to meet the requirements of 10 CFR 50, Appendix B,

Criterion III, Regulatory Guide 1.64, and ANSI N45.2.11-1974,

Section 6.

c. Over Torquing of Pressurizer SRV Flanged Pipe Joints (Reactor Coolant

Pressure Boundary Integrity Joints)

(1) Findings

The NRC inspector reviewed the approved licensee

Procedure MP-PSV-1-3, Revision 15, dated February 6,1986,

" Pressurizer Safety Valve Installation." This procedure

contained detailed instructions for tightening (torquing) the

valve line flange bolts, including the required lubrication and

tightening sequence.

The NRC inspector reviewed the vendor technical manual for the

pressurizer SRVs RC-141 and RC-142. This manual contained the

same data as the licensee procedure. The manual had not had a

technical review to comply with Generic Letter 83-28 and was

classified as an uncontrolled manual by the licensee. The valve

manual procedure, A-1519-1, which contained the data apparently

used by the licensee, gave torque values for generic

applications covering different sizes of bolts and different

levels of bolt stress. The torque value used by the licensee

resulted in bolt stresses of 45,000 psi. As stated in

Section 3.A of this report, allowable stress values would be in

the range of 27,700 psi to 31,800 psi (nominal 30,000 psi),

based on the licensee identified bolting material used for this

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application (ASTM A193, Grade 87 or B16). This results in an

overstress on the bolts of 150 percent of design values.

(2) Original Design and Installation Records

The NRC inspector requested original design base calculations

and installation records for the pressurizer SRV line flange

joint bolt stress level and bolt torque.

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For a second time, the licensee was. unable to retrieve these

records. This is another example of potential violation

(285/8705-02) with regard to-10 CFR 50, Appendix B,

Criterion XVII, identification and ' retrieval of records.

(3) Lack of Design Control in Determination of Pressurizer SRV Line

Flange Joint Design Strer.s Values and Bolt Torque

The licensee was unable to provide the NRC inspector with

' documentation of their compliance with Regulatory Guide 1.64,

-ANSI N45.2.11-1974 (as discussed in Section 3.b.4) for the

determination by OPPD personnel of design bolt stress values and

resulting bolt torque for the pressurizer SRV line flange

joints. This is another example of potential violation

(285/8705-04) with regard to 10 CFR 50, Appendix B,

Criterion III, Regulatory Guide 1.64, and ANSI N45.2.11-1974,

Section 6.

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d. Reactor Containment Building Ventilation and Cooling Fan Units

The NRC inspector also reviewed the blade angles and bolt torque for

reactor containment building ventilation and cooling fan units.

(1). Review of the Vendor Technical' Manual (NRC Generic Ltr 83-28,

paragraph 2.2.2)

The NRC inspector reviewed the licensee vendor technical manual

for American Air Filter / Joy Manufacturing Company reactor

containment building ventilation and cooling fans and auxiliary

equipment. This manual is identified as "NUS-044."

Documentation of the NUS Corporation Technical Manual review

states that the manual requires lubrication of Fan Units VA-2A

and VA-28 at a 6- to 12-month' periodicity; and Units VA-12A and

VA 12-B at a 3-month periodicity. The NUS review identified

several other differences between the vendor technical manual

and the licensee preventive maintenance program, lhe licensee

has not reviewed this manual for resolution of differences, and

does not consider it a controlled manual. This will remain an

unresolveditem(285/8705-05) pending further review during a

subsequent inspection.

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(2) Findings Related to Safety-Related Fan Units VA-3A, VA-3B,

E VA-7C, and VA-7D

The section of the ir.anual containing data for Fan Units VA-3A,

VA-3B, VA-7C, and VA-7D, on page 11 states:

"During normal and emergency operation, this angle will be

degrees,' but if the fan must be tested at conditions other.

than spec values, this angle may have to be changed ta prevent

overworking of the motor with resulting burnout."

As noted above, the blade angles are not given in the technical

manual. Licensee Procedure HP-VA-7, Revision 6, dated

October 28, 1985, "VA-7C/7D Fan Blade Checks / Adjustment," shows

the blade angle as 26.5 degrees for Fan Units VA-7C and VA-70.

Licensee engineers could not provide the NRC inspector with

documentation of the design calculations or justification for

using a blade angle of 26.5 degrees. The licensee could not

retrieve construction phase design calculations and

installation / test records substantiating that this angle was

correct or any 0 PPD design review documenting the acceptability

of this angle.

Similarly, the licensee was torquing the bolts which prevent a

change in blade angle-to 450 foot-pounds, though no value was

given in the vendor technical manual. The licensee could not

retrieve construction phase design records approving this torque

or a design review showing its acceptance.

In addition, the method of adjusting blade angle in the

procedure differed from that described in the vendor technical

manual'. The method in the procedure was subject to two

interpretations that would result in different blade angles.

Two licensee engineering personnel were unable to explain to the

NRC inspector the proper blade adjustment using the procedure.

Licensee Procedure PM-VA-2, Revision 5, dated December 22, 1986,

"VA-3A-3A/3B Fan Blade Checks / Adjustment," contains the same

differences from the vendor technical manual, except that the

licensee was using a blade angle of 21 degrees.

No construction records or design reviews were retrievable to

substantiate the blade angle and torque values being used by the

licensee for this procedure.

(3) Summary

a) Failure to Identify and Retrieve Design Calculations for

Blade Angles and Bolt Torque for Safety-Reldted Fan

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Units VA-3A, VA-38, VA-7C, and VA-70

The failure of the licensee to identify and retrieve

original design calculations for blade angles on

Fans VA-3A, VA-3B, VA-7C, and VA-70 is another example of

potential violation (285/8705-02), failure to meet

10 CFR 50, Appendix B, Criterion XVII.

b) Failure to Have Documentation Substantiating Design Reviews

of Procedures Specifying Blade Angles and Bolt Torque for

Safety-Related Fan Units VA-3A, VA-3B, VA-7C, and VA-7D

Without the availability of original design calculations

for fan blade angles and bolt torques for Fans VA-3A,

VA-3B, VA-7C, and VA-70, determination of blade angles and

bolt torques was a design review. The failure of the

licensee to provide documentation of required design

reviews for blade angles and bolt torques is another

example of potential violation (285/8705-04), failure to

perform a design review as required by 10 CFR 50,

Appendix B, Criterion III, and ANSI N45.2.11-1974.

e. Vendor Technical Manuals

(1) Background

The NRC inspector reviewed licensee copies of vendor technical

manuals for the following significant safety related equipment

to determine the design maintenance base for maintenance of

these components, and review / updating in response to NRC Generic

Letter 83-28.

Turbine driven auxiliary feed pump.

Electric auxiliary feed pump.

High pressure safety injection pump.

Raw water pump.

Pressurizer safety relief valves.

Reactor containment building ventilation and cooling fan

units.

(2) NUS Review

The above manuals, except for the pressurizer safety relief

valves manual, had been reviewed recently by the NUS Corporation

for OPPD as part of the licensee's response to Generic

Letter 83-28. The NRC inspector determined that no

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documentation existed that NUS identified: (1) the vendor

recommendation for annual disassembly and inspection of the the

raw water pumps, and (2) that the manual for the reactor

containment building ventilation and cooling fan units did not

apparently contain the blade angles or bolt torque values for

the adjustable fan unit blades. Further, the NUS review did not

identify components, such as sleeve bearings in the HPSI and

auxiliary feed water pumps, or antifriction bearings in

containment building ventilation and cooling fan unit motors

that have predictable design life cycles which are less than

plant design life; and hence, need to be evaluated for periodic

bearing replacement based on plant life-cycle component

operation, including required postaccident operation.

Containment building ventilation and cooling fan units were

required to operate for approximately 1 year following a design

base event (DBE). This will remain an unresolved item

(285/8705-06) pending further review.

(3) NSSS Supplied Manuals, Lack of Review

The pressurizer safety relief valve manual did not contain

documentation that it had been reviewed in response to Generic

Letter 83-28. Licensee personnel stated that the original

program plan was for Combustion Engineering (CE) to perform the

review of vendor manuals that were in CE's original scope of

supply. Economic considerations, apparently precluded this

action, and the licensee identified no other approved,

implemented, program plan for the review and approval of vendor

manuals originally supplied by CE.

(4) Controlled Vendor Technical Manuals and Licensee Technical

Manual Review

Licensee personnel stated that a vendor technical manual could

not become a " controlled" manual until it had a technical review

by 0 PPD engineering technical service personnel. The licensee

estimated that there were approximately 600 technical manuals

for safety related equipment and components. An estimated

540 (approximately 90 percent) had been reviewed by NUS. The

remaining 10 percent were supplied by CE or were not available.

The licensee stated that no manuals had been reviewed by OPPD;

hence, there were no controlled vendor technical manuals.

(5) Licensee Position Status on Generic Letter 83-28 Regarding

Maintenance / Design Life

The licensee stated that the potential requirement for

replacement of component parts with a design life less than that

of the plant life had not been considered in their original

response to clarify Generic Letter 83-28. Additionally,

safety related preventive type maintenance activities specified

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in the vendor technical manuals, which were not presently

accomplished by Fort Calhoun Station, were not identified.

(6) Upgrading of Licensee Maintenance Practices and Procedures

The licensee could not identify a schedule, or a projected

target date, for upgrading plant safety-related maintenance

procedures and activities to meet either (1) equipment

manufacturer's specified maintenance, or (2) identifiable

required component replacement such as anticipation bearings and

electrolytic capacities, necessary to maintain the

safety-related design bases for Fort Calhoun Station, after

completion of commitments to Generic Letter 83-28. This will

remain an unresolved item (285/8705-07) pending further review

during a subsequent inspection.

4. Instrumentation Maintenance

Maintenance history, calibration records, qualification records,

procedures, work item tracking forms, and maintenance work request forms

were reviewed.

Four work activities in progress were witnessed to determine if the

technicians were following the appropriate maintenance procedures. The

activities were:

Calibration Procedure CP-B/911; SG A Wide Range Level (Revision 10).

Calibration Procedure CP-B/912; SG B Wide Range Level (Revision 4).

Calibration Procedure CP-C/112C; Reactor Coolant Loop Cold Leg

Temperature (Revision 6).

Calibration Procedure CP-C/112H; Reactor Coolant Loop Hot Leg

Temperature (Revision 6).

A number of apparent discrepancies were noted during the performance of

these activities.

a. Failure to Follow Procedures

The first discrepancy noted was that during the performance of

Calibration Procedure CP-B/911 and 912, the technicians turned a

bypass switch (RC2B) to the bypass position, which was a step that

was not provided for in the procedure. Questioning of the

technicians determined that this step was done as an extra precaution

on their part. They felt that due to the close proximity of this

module to another module having the same appearance and function

(just a different transmitter), it would be too easy to work on the

wrong module. Therefore, to ensure that no incorrect signals or

alarms resulted, they turned the controls for the second module to

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bypass, even though it was not in an approved procedure, nor had it

been reviewed as an unresolved safety issue.

The fact that steps were being performed that had not been reviewed

by the cognizant engineering group was questioned. Since the switch

repositioning had not been reviewed and approved as an authorized

change, the safety implications had not been assessed. The

technician's actions indicated that step-by-step procedural

implementation was not strictly adhered to. Failure to-implement

procedures as written is a potential violation of Fort Calhoun

Station Technical Specification 5.8.1 (285/8705-08).

b. Use'of A Defective Instrument

The second concern identified by the~ inspector was that one of the

instruments used for the performance of CP-B/911 and CP-B/912 did not

appear to be working properly in that one leg of the digital readout

was faint, or nonexistent. The technician took readings using the

seemingly defective meter without paying any apparent attention to

the malfunction. Only after questioning by the NRC inspector was the

, meter changed for one that worked properly. The technician indicated

- that since the meter did not appear to be working properly, its

calibration and operation wouid be verified. As followup to the use

of the meter, the NRC inspector questioned the technical group about

the sequence to be followed should the instrument be found defective.

It was demonstrated that a methodology existed where a record of

procedures on which this instrument was used was in place. Should

the instrument be identified as being defective the calibrations in

question could be repeated as required.

c. Lack of Specified Instrument Accuracy

The third concern was how the I&C staff had established that the

accuracy of instruments used for calibration was within the accuracy

constraints of the system on which it was being used. Questioning of

the technical and engineering staff determined that system accuracy

and calibration instrument accuracies had not been considered.

Failure to assure that instrument accuracy is adequate to meet

Technical Specification values is another example of potential

violation (285/8705-01), failure to provide criteria to permit

determination of acceptability in violation of 10 CFR 50, Appendix B,

Criterion V.

d. Failure to Control Test Connection Points

The fourth concern was that the procedures identified voltage and

current readings to be obtained but did not indicate how or where

they would be takee This practice did not assure uniformity, or

acceptability of the readings. This is an example of potential

violation (285/8705-01), failure to provide criteria to permit

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determination of acceptability in violation of 10 CFR 50, Appendix B,

Criterion V.

e. Training of Plant Personnel in Electrical Equipment Qualification (EEQ)

The fifth concern was that the personnel (electrical, I&C, and QC)

whose responsibility was.to maintain equipment qualification did not

appear to understand the requirements of EEQ. This concern was

discussed with a number of personnel from each of the applicable

organizations and.the general impression was that the people whose

responsibility was to maintain qualification had not received all of

the training necessary on qualification, how the requirements were

identified and what steps had to be performed to ensure qualification

was maintained. The NRC inspector's concern of personnel training

was discussed with the training department personnel and it was found

that the training program was under development. This will remain an

unresolved item pending further review. (287/8705-09)

f. Control of Safety-Related Requirement Parts

The sixth concern identified involved the control of qualified

replacement parts. The concern was with the method used by the

licensee for control of CQE (Critical Quality Equipment) qualified

replacement parts. The current method for storing CQE materials is

that each department (Electrical and I&C) maintains a CQE storage

area. When during the performance of a task, replacement parts are

needed, the craftperson goes to the storage area and obtains the

parts believed to be necessary and then signed in a log that the

parts were removed. The storage area was not manned constantly. The

concern included the return of parts that were removed for a job and

then found to be unnecessary. The craftpersons were questioned as to

what procedure would be followed in this case. The response in

general was that it was not clear what should be done. Discussions

with plant management indicated that a procedure existed for the

return of equipment. The procedure was not reviewed; however, it

appeared that personnel had not been adequately trained in the

control of qualified spare parts, technical approval of parts was not

specified, and shelf life controls were not identified. This will

remain an unresolved item (287/8705-10) pending further review during

a subsequent inspection.

g. Re.tew of I&C Maintenance Orders (M0s)

Eleven completed safety-related M0s were reviewed to ensure they were

completed, reviewed, stored as quality documents, all discrepancies

resolved, and the appropriate work performed.

During the completed M0 review, a concern was identified which

involved the maintenance performed on M0 No. 41 861204. This

maintenance was for replacing a potentiometer and and amplifier

module for AI-31C-CW15-B1 and then performing a Surveillance Test

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Procedure ST-RPS-7. The concern was that the procedure was not clear

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yet the procedure had been completed. The licensee researched the

files and located a signed off copy of the procedure and provided it

for review, which satisfied the NRC inspector's concern.

5. Electrical Maintenance

a. In-Process Review of Electrical Maintenance

Two maintenance action items in progress were witnessed to ensure

that the electricians performing the maintenance tasks followed the

appropriate maintenance procedures. These tasks were performance of

D. C. transfer switch surveillance testing and A. C. circuit breaker

testing. Testing was performed in accordance with the following

procedures:

CP-Tie Breaker, Tie Breaker Calibration Procedure (Revision 3).

ST-DC-3, D. C. Transfer Switch (Revision 10).

(1) Tie Breaker Calibration

During the performance of "CP-Tie Breaker" maintenance, two

concerns were identified.

(a) Use of Incorrect Procedure

The first concern involved the procedure that was used to

perform the testing. When the testing was started, it was

begun by using a " Procedure CP-Main." This procedure was

for testing main breakers instead of tie breakers. The

main breaker procedure was used until overcurrent testing

was performed, at which time the trip settings were found

to be set almost 100 seconds early on all three phases.

The electricians performing the test adjusted the setting

en one phase, when it was decided that something w s wrong.

An electrical supervisor was contacted who in turn

determined that the incorrect procedure was being used.

The correct procedure was obtained, and implemented, and

the testing was completed. Failure to implement the proper

procedure is another example of potential violation of

Fort Calhoun Station Technical Specification 5.8.1

(285/8705-08). The licensee indicated that the appropriate

procedures would be reviewed to prevent recurrence of this

problem.

(b) Use of Incorrect Vendor Technical Manual

During the performance of CP-Tie Breaker maintenance, it

was noticed that the secondary contacts on the tie breaker

had some uneven arcing marks. It was not clear that such

arcing was in accordance with the vendor's technical

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manual. Licensee personnel were requested by the NRC

inspector to provide a copy of the technical manual to

verify the requirements. Licensee personnel provided a

copy of a technical manual which was generic and not

specific to the breaker being tested. The generic manual

did not discuss the contacts and the adjustment

requirements in detail. The NRC inspector reviewed the

generic technical manual and referred the licensee

personnel to another manual which was the applicable

manual. The electricians then found the correct manual.

The contact requirements'were determined to be in

compliance. The licensee does not appear to have adequate

control of the technical manuals. This finding was

corroborated when crafts personnel (I&C and Electrical)

were questioned about the control of technical manuals.

They indicated that at times the technical manuals may not

be up-to-date, correct, or locatable. This will remain an

unresolved item (285/8705-11) pending further review during

a subsequent inspection,

b. Review of Completed M0s

Eleven completed M0s were reviewed to ersure they were completed,

reviewed, stored as quality documents, all discrepancies resolved and

the appropriate work performed. During the review, several concerns

were identified. One concern involved the review of maintenance

records that were completed during or before 1984. The review of the

completed maintenance found that in some packages the work

description was not detailed enough to determine the work performed,

not all applicable documents were included in the package, EQ

tracking documents could not be located (see M0 No. 840858), and it

was not clear that required inspections had been performed. This

concern was discussed with licensee management. The discussion

determined that the MO control procedures had been modified to

prevent these type of concerns. A number of M0s dated 1986 and 1987

were reviewed to verify that the modified procedure was adequate

enough to prevent similar types of anomalies. The review of later

M0s did not identify any further concerns,

c. Accomplishment of Meggering

Additional concerns were identified concerning meggering of

electrical equipment as follows:

A number of procedures (examples are PM-EE-VA3/7, Revision 0 and

PM-MOV-1, Revision 3) indicated that meggering was to be

accomplished; however, the instrument and/or the voltage of the

instrument was not provided.

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A number of procedures (examples are PM-EE-VA-3/7, Revision 0; i

PM-EE-1-13, Revision 5; and PM-EE-3.0, Revision 0) do not

specify meggering acceptance criteria.

A procedure for meggering reactor containment ventilation and

cooling fan unit motors was not identified during the q

inspection. Failure to identify meggering equipment required to f

provide acceptance criteria and to megger containment

ventilation and cooling unit fan motors is an example of

potential violation (285/8705-01), failure to provide criteria

to permit determination of acceptability in violation of

10 CFR 50, Appendix B, Criterion V.

d. Replacement of Safety-Related Agastat Relays

The licensee was questioned by the NRC inspector about replacing

relays at the end of their qualified life. A replacement schedule

had not been identified. The licensee is to discuss this issue with ,

Agastat and resolve it in accordance with the vendor's~ <

recommendations. This will remain an open item pending further

review during a subsequent inspection. (285/8705-12)

e. Maintenance of ASCO Solenoid Valve HEV-386-20

One concern was a task description for preventive maintenance for

maintaining qualified life on ASCO Solenoid Valve HCV-386-20. The

task description indicated that the work was to refurbish the

solenoid valve; however, the procedure that was referenced (SP-50V-1)

did not discuss refurbishing the valve but discussed exercising it.

The licensee was questioned about this discrepancy. The response

indicated that the requirement was to exercise the valve and not.

refurbish it. The refurbish description was in error. The licensee

further stated to indicate that this valve had been removed from the o

equipment qualification program.

f. Calibration of Safety-Related Relays

Relays were calibrated by System Protection, an offsite group from

0 PPD. No concerns were identified from this review.

g. Electrical Maintenance Procedures Lacking Acceptance Criteria

The NRC inspector reviewed the following electrical maintenance

procedures which did not include acceptance criteria, for the '

tightening, or verification of tightness of threaded electrical

connections such as tie bolts, terminal bounds, and mounting bolts,

and in some cases voltages for meggering.

MS-EE-4.1, Revision 1, January 1,1980, " Maintenance Standard

for 480 Motor Control Center Draw-Out Unit."

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,

PM-EE-1.1, Revision 2, March 12, 1987, " Type AM-4.16-250

Magne-Blast Circuit Breaker."

PM-EE-3.0, Revision 0, October 13, 1978, "AKD-5 Low Voltage

Swi tchgear. "

PM-EE-3.2, Revision 1, dated October 27, 1978, " Types AK-2-25

and AK-2-50 Circuit Breaker."

Failare to provide acceptance criteria is another example of

potentiai violation (285/8705-01) regarding 10 CFR 50, Appendix B,

Criterion V.

h. Lubrication of Containment Cooling Fan Motors and Limitorque Operators

The licensee was lubricating Reliance Motors in containment on an

18-month (refueling outage) in lieu of 3-6 month periodicity, and had

used MARFAK lubrication in lieu of Nebula EP-0 in Limitorque

operators in containment. This will remain an unresolved item

pending further review during a subsequent inspection. (285/8705-13)

1. Unapproved Use of Mixed Greases in Electrical Equipment

Qualified (EEQ) Limitorque Operators for Safety-Related Valves in

the Reactor Containment Building

The NRC inspector reviewed the Fort Calhoun 1987 Outage Working

Schedule of maintenance activities. Included on the schedule was the

maintenance activity " change out grease to Nebula EP-0" for the

following EEQ valves in containment:

L

'

HCV-383-4-0

F

HCV-347-0

HCV-348-0

'

HCV-331-0

HCV-383-3-0

HCV-333-0

The licensee indicated that these Limitorque operators contained a

mixture of Nabula EP-1 and Marfak greases which they considered

qualified even though such practice contradicted the specific

requirements of the applicable Limitorque Type SMB Instruction and

'

Maintenance Manual, Bulletin SM81-828. The document used as

justification was Bolt and Associates letter dated June 17, 1986,

i 11 ring Marfak with Nebula EB or (Sun Oil Company) Prestige 50 EP."

The letter basically referenced a 1961 Rock Island Arsenal Laboratory

lorhnical Report, "Compatability of Lubricating Grease." This report

covered tests of unnamed, generic type greases and concluded that:

Compatibility was determined by the extent of hardening or

softening exhibited by the binary mixture after a 10,000 stroke

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6

worker test as compared to the extent of hardening or softening

exhibited by the components when subjected to the same test.

Incompatibility can occur when greases of the same type are

mixed together. A grease made by one manufacturer may behave

differently from the same type of grease made by another

manufacturer.

It was found that grease compatibility could not be predicted

accurately from a mere knowledge of tile thickener and fluid

type, and the safest procedure to follow is to avoid mixing

greases of different types.

The letter begins with a statement that no problem exists with

Limitorque operators because of mixing, since incompatibility is

likely to result in " softening" (a de-hydrolyzation or

de-saponification'of the grease increasing its fluidity). The letter

states that this would be no problem since the grease is contained in

the gear box. Adverse chemical reactions (bearing corrosion, etc.),

reduction or loss of extreme pressure (EP) characteristics, or

separation of oil from the mixture, all of which may result from

incompatibility of the greases, were not addressed.

The letter summarizes with a paragraph which states:

"I do not wish my comments on mixing greases to be construed as

endorsing such a practice. I agree with Limitorque

(Bulletin SMBL-80) that this should not be done as a matter of

policy. It is not a good idea, particularly if the greases

mixed involve different gelling agents, as is true with the

situation at hand. The best practice is to pick an approved

grease (Nebula is the only one approved for everything) and stay

with this product down the road. If more than one supplier is

desired or as, other products are qualified, I would keep with

materials containing at least the same gelling agents."

Marfak and Nebula EP greases have different gelling agents.

Limitorque has stated that qualification of their operators in

containment is contingent upon the sole use of Exxon Nebula EP

greases.

Failure to use the required grease is a potential violation of

10 CFR 50.49 (285/8705-14).

5. Exit Interview

The NRC inspector met with licensee representatives (denoted in

paragraph 1) on April 10, 1987, and summarized the scope and findings of

inspection activities. The NRC senior resident inspector was present.

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