IR 05000267/1987004

Insp Rept 50-267/87-04 on 870126-30.No Violations Noted. Major Areas Inspected:Program to Maintain Qualification of Electrical Equipment Per 10CFR50.49.Potential Enforcement/ Unresolved Items & Open Items of Program Listed

ML20205T579
Person / Time
Site:	Fort Saint Vrain
Issue date:	03/27/1987
From:	Hubbard G, Potapovs U NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
To:
Shared Package
ML20205T521	List: IR 05000267/1987004 ML20205T514
References
50-267-87-04, 50-267-87-4, IEIN-86-053, IEIN-86-53, NUDOCS 8704070330
Download: ML20205T579 (15)

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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Report Nos.:

50-267/87-04 Docket No.:

50-267 License No.:

OPR-34 Licensee:

Public Service Company of Colorado Post Office Box 840 Denver, Colorado 80201-0840 Facility Nane:

Fort St. Vrain Nuclear Generating Station Inspection At:

Platteville, Colcrado Inspection Conducted:

January 26-30, 1987

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Inspector: G.T. Hubbard, Equipment Qualification & Test Engineer

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Also participating in the inspection and contributing to the report were:

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U. Potapovs, Chief, Equipment Qualification Inspection Section (EQIS), IE

A.R. Johnson, Reactor Inspector, RIV D.E. Nonran, Peactor Inspector, RIV J.E. Bess, Reactor Inspector, RIV P. Shemanski, Sr. Electrical Engineer, NRR M. Yost, Sr. Engineer, Idaho National Engineering Laboratory (INEL)

J. Hanek, Sr. Engineer, INEL J. Grossman, Member of Technical Staff, Sandia National Laboratories r

Approved by:

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[ MhtS 3 - 2~7* @ 7 U. Potapovs, Chief, E05,IE Date

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S 8704070330 870403 PDR ADOCK 05000267 G

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INSPECTION SUMMARY Inspection on January 26-30, 1987 (Inspection Report No. 50-267/87-04)

Areas Inspected: Announced inspection to review the licensee's implerentation of a program for establishing and maintaining the qualification of electric equipment within the scope of 10 CFR 50.49.

Results: The inspection determined that the licensee has implemented a program to meet the recuirements of 10 CFR 50.49, except for the deficiencies listed below.

Peport Item Name Paragraph Number Potential Enforcement / Unresolved Items:

1.

Qualification of Rockbestos Coaxial 4.E.(1)

50-267/87-04-03 Cable 2.

Qualification of Raychem Splices 4 F.(1)

50-267/87-04-06 Open Items:

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Region RIV Review of FSV EQ 4.B.

50-267/87-04-01 Procurement Program Implementation 2.

Region RIV Review of FSV EQ 4.C.

50-267/87-04-02 Maintenance Progran Irplementation 3.

ASCO Solenoid Valve Conduit 4.E.(2)

50-267/87-04-04 Installation 4.

Upgrade of Cable Binders To 4.E.(3)

50-267/87-04-05 10 CFR 50.49 5.

Limitorque Valve Operator 4.F.(0)

50-267/87-04-07 Oualification Binder Revision Concerning Vertical Conduit Entrances

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DETAILS

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Persons Contacted:

1.1 Public Service Company of Colorado (PSC)

• R.D. Williams, Jr., Vice President Nuclear 0perations

• D. Warembourg, Manager, Nuclear Enaineering

• H.L. Brey, Manager, Nuclear Licensing and Fuels (NLF)

• M.E. Neihoff, Nuclear Design Manager / EQ Project Coordinator

• S. Marquez, Nuclear Engineering Division (NED) E0 Coordinator

• J.R. Ressy, Staff Assistant

• G.S. Bates, Supervisor, Nuclear Projects

• D.T. Brown, Instrumentation and Controls (18C) Supervisor

• M. Lehr, Supervisor Quality Assurance (QA) Engineering

• J. Hunter, Shift Supervisor

• D. Johnson, NED Site Supervisor

• R. Gappa, NED Site Engineer

• B.G. Barta, NED Site E0 Coordinator

• C. Bomberger, NLF

• D. Goss, Nuclear Licensing-NRC Coordinator

• F.J. Borst, Support Services Manager

• J. Eggebroten, Superintendent, Tech Services Engineering

• C.H. Fuller, Station Manager

• F.J. Novachek. Tech / Administrative Services Manager

• D. Weber, Staff Assistant to Station Manager

• R. Sargent, Assistant to Vice President Nuclear Operations

• J. Waalot, NED

• L. Marquez, Engineer

• M.J. Ferris, QA Operations Manager

• R.L. Craun, Nuclear Site Engineering Manager

• F.W. Tilson, Supervisor, Nuclear Projects

• J. Varina, Consultant l

• J. Selan, Nuclear Licensing

• P.F. Tomlinson, Manager, QA Division

• R.A. Doyle, CA Engineer

• T.C. Prenger, QA Services Manager

• L. Pierce, Media Representative

• M.H. Holmes, Nuclear Licensing Manager J.W. Gahm, Manager, Nuclear Production Division G.V. Reigel, Supervisor Site Engineering K. Dvori, Engineer

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1.2 PSC Contractors

• M. Spisak, Senior Project Engineer, Sargent and Lundy (S&L)

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• M. Capello, Consultant, T.D. A.

• M.C. Annon, Consultant, ISC Engineering Associates

• R. Berg, Consultant, TERA

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• R. Frantz, Consultant, Proto-Power Corporation

• R.J. Atkisson, Lead Engineer, Consultant, Proto-Power Corp.

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B.M. Loy, Senior Computer Engineer, S&L J.W. Anderson, Engineering Project Manager, Wyle Labs J. Gleason, Director Nuclear Engineering, Wyle Labs J. Sinnappan, Supervisor, S&L l

1.3 NRC

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• R.E. Ireland, Chief, Engineering Section, RIV

• R.E. Farrell, Senior Resident Inspector P. Michaud, Resident Inspector

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PURPOSE:

The purpose of this inspection was to review the licensee's implementation

of a program meeting the requirements of 10 CFR 50.49 for Fort St. Vrain Nuclear Generating Station.

3.

BACKGROUND:

In the fall of 1985, PSC requested that the NRC grant Fort St. Vrain Nuclear Generating Station (FSV1 an extension of the November 30, 1985, deadline for meeting the ecuipment qualification (EQ) requirements of 10 CFR 50.49. This request was documented in PSC letters dated September 24, October 22, and November 22, 1985. The extension request was granted in Commission Memorandum and Order, dated November 26, 1985.

This extension allowed operation of the plant until May 31, 1986, with reactor power level limited to 35%. At that time the plant was shut down and modifications for compliance with 10 CFR 50.49 were becun. Included in the Commission Memorandum and Order was the condition that operation of the plant after May 31, 1986, would only be allowed when the FSV EQ program was completed and approved by the Director, NRR, and the Director, IE.

Based on PSC submittals dated Pay 12, June 20, July 28, and July 31, 1986, l

a final E0 Safety Evaluation Report (SER) was issued for FSV on

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I September 10, 1986.

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FINDINGS:

The NRC inspectors examined the licensee's program for establishing the qualification of electric equipment within the scope of 10 CFR 50.49. The program was evaluated by examination of the licensee's qualification docu-mentation files, review of procedures for controlling the licensee's E0 activities, verification of the adecuacy and accuracy of the licensee's 10 CFR 50.49 couipment list, and examination of the licensee's progran for maintaining the qualified status of the covered electric equipment.

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Based on the inspection findings, which are discussed in more detail below, the inspection team determined that the licensee has implemented a program to meet the requirements of 10 CFR 50.49, although some deficiencies were identified, j

The licensee's EQ program is defined in FSV procedure G-16. " Environmental Qualification of Electric Equipment (EO)," issue 1, dated January 19, 1987.

This procedure reflects the PSC EQ Policy as defined in Nuclear Operations Nuclear Policy No. 28, " Equipment Oualification of Electric Equipment (EQ)

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Policy," issue 1, dated January 7,1987, and defines the FSV E0 program controls established with guidance from the Nuciear Operations Nuclear

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Guidelines No. 28.1, " Guidelines for Environmental Qualification of Electric Equipment Program," issue 1, dated January 7, 1987.

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The NRC inspectors' review of G-16 determined that it identifies the

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requirements for implementation of the EQ program and defines the

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responsibilities of the various organizations. The procedure discusses the following topics in defining EQ program requirements: (1) accident

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parameters which nest be considered when establishing equipment qualifi-cation; (2) the environmental qualification master equipment list (MEL);

(3) environmental qualification binders; (4) the EQ maintenance program; (5) training; (6) plant modifications and design document changes; (7) EQ coordinators; (8) operations; and (9) materials management (including procurement). In addition to procedure G-16, other procedures are used at FSV to implement the EQ program. These procedures are divided into the three program categories of nodifications, procurement, and maintenance.

The inspectors' review of the procedures in these areas is discussed in the following paragraphs:

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A.

EQ Modification Procedures The NRC inspectors' examination and review of modification program

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procedures and their implementation was accomplished by review of the

.ollowing procedures.

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Q-3, " Design Control System," issue 10, dated October 24, 1986

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l G-9, " Controlled Work Procedures," issue 6, dated December 5,

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1984 ENG-1, " Control of Modifications and Documentation Changes,"

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issue 11, Januar." 5, 1987

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ED-100, " Change Notice Preparation and Document Control,"

issue 2, dated January 5, 1987

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DD-CQD-2, " Design Change Directive for Electrical Component

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Qualification," issue 2, dated January 19, 1987 NDMAP-4, "EQ Binder Review and Acceptance," issue 3, dated January 12, 1987 The inspectors reviewed the modification progran procedures to determine that adequate procedures ard controls had been established to meet the requirements of 10 CFR 50.49. Areas of the progran procedures reviewed included methods of procedural control and their effectiveness for:

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Establishing harsh environmental conditions at the location of equipment through engineering analysis and evaluation.

Establishing, evaluating, and maintaining EQ docunentation.

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Controlling plant modifications, including installation of new and replacement equipment, and providing for updating replacement eouipment.

The inspectors concluded that the procedures meet EQ requirerents and the E0 modification program was well planned and implemented.

B.

E0 Procurement Procedures The inspectors determined that the procurement systen at FSV is implemented through the following documents which were reviewed by the inspectors.

0-4, " Procurement Systen," issue 9, dated October 24, 1986 0-7, " Control nf Procured Materials, Enuipment, and Services," issue 9, dated October 24, 1986 P-5, " Materials Control," issue 9, dated October 24, 1986 ENG-16, "NED Procurement Document Evaluation," issue 3, dated January 5, 1987 ED-76, " Environmental Qualification Procurement Document Review," issue 1, dated January 12, 1987

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Tha procurement system at FSV, including E0 procurements, is established by procedure 0-4 which defines the requirements and responsibilities for preparation, review, and approval of all PSC generated procurement documents.

It also identifies the interfaces of the organizational divisinns performing procurement activities.

The program implementation is acccmplished through the remaining procedures listed above.

In particular, ENG-16 provides requirements for the Nuclear Engineering Division (NED) evaluation of procurement documents issued for purchase of equipment, parts, and services for use at FSV. ED-76 is used.in conjunction with ENG-16 to provide requirements for the NED review of procurement documents for purchase of equipment and parts in EQ applications to verify that purchases meet the requirements of 10 CFR 50.49.

As a result of the inspection it was concluded that a program is in place to control E0 purchases.

Since only a limited review of E0 purchase orders was performed during the inspection, a subseouent RIV inspection will be made to review the overall implementation of the EQ procurement system. Completion of this inspection is considered an Open Item, 50-267/87-04-01.

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C.

E0 Maintenance Program The inspectors' review of the E0 maintenance program determined that it was an integral part of the overall FSV maintenance program which is implemented by administrative procedure SMAP-27, " Preventative and

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Corrective Maintenance Equipment Review," issue 3, dated October.31, 1986. Maintenance identified by the equipment manufacturer, vendor, or others which reauires specific tasks and frequencies ~ to maintain eauipment in a qualified condition are defined in the procedure as

" Required Periodic Maintenance" which must be accomplished. Each EQ Binder (oualification file) is reauired to be reviewed for necessary maintenance which is entered into the plant. maintenance program where scheduling is performed.

Administrative procedure SMAP-25, " Equipment Qualification Review of Station Service Requests," issue 2, dated December 26, 1986, provides guidelines for the technical review of equipment qualifi-

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cation " Station Service Requests" (SSRs) to ensure that equipment qualification is maintained. An SSR is recuired for each maintenance

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action and the SMAP-25 review ensures that each action is reviewed to determine the potential effect on qualified eouipment.

In addition to these two procedures, the following procedures were reviewed by the inspectors.

SMAP-29, " Station Service Request Work Pl?rning," issue 1, dated April 11, 1986 P-7, " Station Service Recuest Processing," issue 11, dated

June 2, 1986 P-12, " Plant Maintenance," issue 2, dated June 4, 1986 RP-A-02, NBD Instrument Calibration and Maintenance Program," issue 2, dated July 7, 1986 Based on the procedure reviews and personnel interviews, the NRC inspectors concluded that the licensee has a program in place to maintain the qualified status of qualified equipment.

However, since only a limited sample of maintenance procedures. (SSRs) and records for selected equipment covered in the file reviews and plant walkdown inspections were examined, a subsequent inspection

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will he made by RIV to evaluate overall EQ maintenance program

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implementation.

Completion of this EQ maintenance _ implementation inspection is Open Item 50-267/87-04-02.

D.

10 CFR 50.49 Master Eauipment List (MEL)

The NRC inspection team reviewed the FSV 10 CFR 50.49 MEL, dated January 26, 1987, which lists the safety-related (SR) electric equipment located in a harsh environment. The MEL for FSV identifies electric equipment within the scope of: (1) paragraph (b)(1) of

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10'CFR 50.49 concerning equipment relied upon to remain functional during and following FSV design basis accidents (DBAs); (2) paragraph I

(b)(2) of 10 CFR 50.49 concerning nonsafety-related equipment whose failure under harsh environmental conditions could prevent. satis-factory accomplishment of. safety functions; and (3) paragraph (b)(3)

of 10 CFR 50.49.concerning post-accident monitoring (PAM) instru-mentation and equipment which requires qualification to meet the i

intent of Regulatory Guide (RG) 1.97. :The DBAs postulated for FSV

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line break (HELB); (2) DBA-1, are.(1) the worst case high energy (3) DBA-2,. rapid depressurization

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permanent loss of forced cooling; and secondary closures of one prestressed

due to failure of. primary (PCRV) penetration; and (4) maximum concrete reactor vessel-

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kl creditable accident (MCA), break in helium purification, regeneration line outside the PCRV. Only the HELB and DBA-2 accidents apply to

l equipment listed on the MEL.. Equipment in the reactor, building,

qualified for a HELB, is also qualified for DBA-2,'since it is enveloped by the HELB. Equipment in the turbine building is not exposed -to DBA-2; therefore, it is only qualified.to HELB conditions.

DBA-1 does not produce a harsh environment;-therefore, it does not affect oualification of equipment (FSAR Section 14.10). The MCA, i

using the CONTEMPT-G code, does not exceed 104 degrees Fahrenheit i

(F) in the reactor building (FSAR Section 14.8) and is a mild environment.

l The basis for the FSV MEL is defined in station procedure SR-6-0,

"Model for Equipment Required for Safe Shutdown Cooling Following a HELB," issue 1, dated January 25, 1987. SR-6-0 provides a model to identify major plant components that are required to function for safe shutdown cooling following a HELB.

In addition, SR-6-0 provides a basis to identify instrumentation which is required

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for PAM following a HELB. SR-6-0 also provides a means of identifying

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safe shutdown cooling and PAM equipment as a result of FSV station

modifications. The methods used by FSV for preparing design changes /

i station modifications and analyzing the original and subsequent SR-6-0 information are controlled by Section 4 of design directive i

procedure DD-CQD-2. The 10 CFR 50.49 MEL is a subset of the FSV l

SR-6-2 safety-related component data base and SR-6-8 safety-related component (subtier) data bases.- The MEL is extracted from these data

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bases using EQ sorting constraints.

Safety-related data bases are controlled and revised in accordance with administrative procedure

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These activities are further controlled by FSV station procedures ENG-1 and ED-100. Additional control is defined in procedure

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ED-2, " Nuclear Engineering Division (NED) Controlled CN Foms,"

issue 3, dated December 1, 1986. Changes to'the MEL are primarily assigned to NED and Nuclear Licensing and Fuels (NLF), who are also responsible for the controlling, handling, distribution, and docu-

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ment storage of design modifications, coordination, approval, construction, and closecut of Change Notices (CNs).

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.h During the Quality Assurance (QA) Audit of the Environmental Qualification Program (PSC Report No.: QAA-1402-86-09) performed'

in November 1986, by third party industry experts, two deficiencies requiring corrective action were identified with regards to the MEL (Corrective Action Requests.(CARS) 193 and 195). A lack of. contrcl f additions and deletions to the MEL (regarding equipment additions /

.odeletions), in that their design basis were apparently unverified, was identified. However, the MEL in force during the QA audit used procedure SEMAP-4 to control additions and deletinns. The SEMAP-4 MEL was frozen on July 28, 1986, and transferred to FSV's safety-related components l lists (SR-6-2 and SR-6-8) and was classified as a design document. The SEMAP-4 MEL was deleted and the SR-6-0 MEL was issued on January 26, 1987. Changes to the new SR-6-0 MEL are now being controlled through the CN process defined in ENG-1 and ED-100 which reflect additions /deletiens to the MEL concurrently with system design modifications. Also procedure E0-75,'" Preparation of Engineering Evaluations (EEs)," issue 3, dated January 23, 1987, has been revised to clarify the preparation / revision to EEs which are

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used as design inputs impacting the MEL. CAR-193 and CAR-195 have been closed out satisfactorily to meet the requirements of ANSI N.45.2.11-1974, Sections 4.1 and 10, which require _ traceability of the modified design'back to the-source of design input.

In crder to-test the completeness of the current SR-6-0 MEL, the NRC inspection team _ selected several systems from the SR-6-0 model for safe shutdown cooling, to verify items'in the MEL.

Primary components.

were identified from the simplified flowpath diagrams and descriptions found in.Section 5.5 of the SR-6-0 document. Also the inspectors

reviewed FSV emergency operating procedures (1) 0 POP XII, " Recovering from Actuation of SLRDIS," issue 1; and (2) EP-0, " Steam Leak of Inadvertent lifting of Steam Relief Valve," issue 51. OPOP XII and

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EP-Q were reviewed by the inspectors to identify components in the systems that were on the MEL.

All of the components / equipment' selected for review were on the MEL.

However, one week prior to the inspection, the licensee had identified 12 local indicating lights and two limit switches, required to be removed /added to the MEL. These items were-in the licensee's' process for corrective action during the inspection..

In addition to the documents discussed above, the following documents were reviewed by the inspection team during the evaluation of the MEL.

MEL, dated January 26, 1987 (sorted by equipment tag number)

MEL, dated January'26, 1987-(sorted by manufacturer and model number)

Nuclear Policy No. 28, " Environmental Qualification of Electric Equipment (EO) Policy,"' issue 1, dated January 7, 1987 s

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O Nuclear Guidelines No. 28.1, " Guidelines for Environmental s

Qualification of Electric Equipment Program," issue 1, dated January 7, 1987 G-16, " Environmental Qualification of Electric Equipment (EQ)," issue 1, dated January 12, 1987 GA Technologies Document No. 908849, "FSV-EQ and Offsite Radiation Doses for 100% Power," no change GA Technologies to PSC Letter No. GP-2700, " Design Bases Accident No. 2 - Probability and Consecuent Reactor Buildinn Radiation Levels," dated January 24, 1986 The inspection team's evaluation determined that the licensee had established a 10 CFR 50.a9 List and had implemented procedures to

maintain the list in an up-to-date status.

No deficiencies in the licensee's list were identified during the review.

E.

Environmental Qualification Files The NRC inspectors examined the files to determine the qualified status of equipment within the scope of 10 CFR 50.49.

In addition to comparing plant service conditions with quali#ication test conditions, the inspectors selectively reviewed areas such as required post-accident operating time as compared to the duration of time the equipment has been demonstrated to be qualified; similarity of tested equipment to that installed in the plant (e.g.,

insulation class, materials of components of the equipment, tested configuration compared to installed configuration, and documentation of both); evaluation of adequacy of test conditions; aging calculations for qualified life and replacement interval determination; effects of decreases in insulation resistance on equipment performance; adequacy of denonstrated equipment accuracy; evaluation of test anomalies; and applicability of EQ problems reported in IE Information Notices (ins)/ Bulletins and their resolution.

The inspectors reviewed 37 qualification files associated with 37 equipment items for the plant. At FSV the documentation which supports qualification is contained in qualification documentation files called binders. The inspectors found that the birders reviewed were readily

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auditable and documented cualification of the eouipment except for the deficiencies described in the following paragraphs.

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(1) During the inspectors' review of binder EQ-FSV79, Rockbestos coaxial cable, the inspectors determined that first generation coaxial cable was installed in the plant. Since this generation cable has displayed shortina characteristics due to thermal expansion problems above identified temperature thresholds and was the subject of a 10 CFR Part 21 report several years ago,

FSV was asked to explain the conclusion that the cable was qualified for its FSV envircoment. Qualification in the binder

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was based on Rockbestos qualification report QR-6802 and the Rockbestos "LD-LE" coaxial cable similarity letter, which were not applicable to the FSV applications of first generation cable.

During the inspection, FSV found a test report in file EQ-FSV38 FSV which included first generation cable in the test program.

also had a letter from Pockbestos which defined the temperature thresholds for the onset of thermal expansion problems and they were perfonning a thermal lag analysis to show.that the cable temperatures remained below the damage threshold. The inspectors reviewed the additional information and found that it would be acceptable to support qualification.

Even though the licensee was able to support oualification of the first generation coaxial cable in the plant this item is identified as Potential Enforcement / Unresolved Item 50-267/

87-04-03, since qualification was not established when the initial review of the binder was made by the NRC inspectors.

(2) During tM plant walkdown the inspectors noted that ASCO solenoid valves were installed in the plant without conduit seals on the conduits leading to the solenoids on the valves. Since this installation was different than most qualified installations in other nuclear plants, the inspectors reviewed binder EQ-FSV47 for ASCO solenoid valves relative to valve installation require-ments. This review identified that Section 5.3 of the test program, " Summary of Results and Conclusions," specified that ASCO solenoid valves must be installed using a properly sealed vented conduit / junction box system. The maintenance and sur-veillance requirements; however, specified installation with a 1/4-inch weep hole. The licensee was able to show the inspectors, from photographs of the configuration used during the qualifi-cation testing, that the tested configuration and the actual installed configuration are identical. Additionally the licensee demonstrated to the inspectors that solenoid valve failures occurring during the qualification tests were of no significance in the qualification of FSV valves. This was justified by the short duration FSV accident temperatures (less than 1 minute above 212 degrees F) and the low accident pressures (less than 1 psig) when compared to the time (areater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 13 days)andenvironmentalconditions(265degreesForgreaterand 24 psig or greater) of the test failures. The licensee agreed to revise binder EQ-FSV47 to clarify the sealed vented conduit /

junction box system and how it relates to the FSV installation.

Completion of this clarification is Open Item 50-267/87-04-04, which will be reviewed during a future NRC inspection.

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l (3) The inspectors' review of binder-E0-FSV54, Rockbestos Firewall III cable, determined that even though documentation in the binder would support qualification to 10 CFR 50.49, cualification was claimed only to the D0R Guidelines. Discussions with the licensee determined that cable binders EO-FSV50, 51, 64, and 77 i

also contained documentation for 10 CFR 50.49 qualification;

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however, cualification was also only claimed to the Guidelines.

Since some of the cable in these binders was used in recent plant E0 activities and could be used in any future plant E0 modifi-cations, the inspectors suggested that the qualification of these cables should be upgraded as required by 10 CFR 50.49. Based on the discussions, the licensee agreed to change these five binders to reflect qualification to 10 CFR 50.49. Completion of these binder upgrades is considered Open Item 50-267/87-04-05, which will be reviewed during a future NRC inspection.

(4) The inspectors review of E0-FSV02, Masoneilan electropneumatic transducers and valve positioners model 8005, 8006, and 2012, determined that the test report contained statements that the samples passed the functional tests during the qualification pr6 gram; however, no copies of the data sheets were included.

The licensee contacted the test lab to see if data sheets were available and was informed that the data was destroyed in a fire in 1980. When this was learred, the licensee provided the inspectors an additional test report on the model 8012 from Wyle Labs which had auditable test results. While the inspectors felt that qualification was supported by the original data, they recommended that the licensee enhance this file by incorporating the Wyle report in the binder.

(5)

In addition to the above recommendation on the Masoneilan binder above, the inspectors recommended other file improvements or clarificaticns to the licensee during the inspection.

In all of these cases the inspectors considered that cualification had been adequately supported and the recorrendations were only for binder enhancement. Durirg the inspection a number of these recommendations were incorporated into change notices (CNs) and the inspectors reviewed the CNs and found them to be acceptable.

Other recommendations not incorporated into CNs prior to the end of the inspection were agreed to by the licensee in " Speed Letters" submitted to the inspectors. While the NRC does not plan to review incorporation of the " Speed Letter" information into the appropriate binders or other appropriate documentation, it is expected that the licensee will complete the actions described in speed letters for the following binders.

EQ-FSV02 Masoneilan Electropneumatic Transducers and Valve Positinners EQ-FSV16 Limitorque Valve Operators E0-FSV20 Westinghouse Relays

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EO-FSV82 National Wire and Cable, PVC/PVC EQ-FSV86 Revere Thermocouple Cable, EPDM/HYPALON C00-034680 Loop Inaccuracy Calculation for Cable IR (6) ~IE Information Notices and Bulletins: The inspectors' review of the qualification binders determined that the licensee had established means to evaluate ins / Bulletins relative to qualified equipment. The inspectors determined that ins /

Bulletins are addressed in Section C, " Analysis & Conclusion Section," of the binders. No deficiencies in the licensee's handling of ins / Bulletins relative to qualified eouipment were identified during the inspection.

F.

Plant Physical Inspection The plant physical inspection consisted of the examination of 25 types of qualified safety-related electric equipment installed in the plant.

The inspectors examined characteristics such as mounting configuration, interfaces, model number, environment, and physical condition. During the walkdown the following deficiencies were identified.

(1) The NRC inspectors performed a walkdown inspection of approxi-mately 22 Raychem splices in the following systems:

Limitorque Valve Operators Thermocouple and Instrumentation Junction Boxes Reliance and Westinghouse Motors ASCO Solenoid Valves The inspection consisted of an examination of splices for con-formance with provisions described in the FQ oualification documentation with special attention to ensure that conditions described in IN 86-53 (supplemented by IE Temporary Instruction 2500/17) were not present or were noted.

During the inspection it was found that several splices did not comply with provisions of the Raychem qualified splices in that the splice sealing sleeve did not overlap the shim sleeve. This condition existed on both WCSF-N and N-PKV splices.

Following the walkdown the licensee showed the inspectors a preliminary copy of a Wyle Labs test report, 17883-1, which had tested configurations similar to those found during the walkdown.

However, the report had not received a final review and release by the testing organization ner had it been documented as a basis for qualification in the binder. The inspectors' review of the prelininary report deter-mined tFat it should support oualification of the observed configurations.

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Even though the licensee was able to provide preliminary data which should support qualification of the observed Raychem configurations, this item is identified as Potential Enforcement /

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Unresolved Item 50-267/87-04-06, since qualification (based upon approved test data) was not established in the binder at the tire of the inspection.

(2) During the walkdown of four Limitorque valve operators, the inspectors identified that field cables entered some operators through unsealed conduits with long vertical runs. The limit switch compartment was mounted horizontally in a downward position and had no means for drainage in the event water entered the compartment through an open conduit. The licensee was questioned concerning this installation configuratier and he was able to demonstrate cualification of the installation, since the limit switch compartments are not subject to water accu-nulation through the conduit by either flooding, fire suppression spray, or steam. However, the licensee agreed to update the qualification binder so there would be no question that the Limitorque operators were cualified with the conduit entrarce configurations observed. Completion of this binder update will be reviewed during a future NRC inspection and is identified as Open Item 50-267/87-04-07.

(3) During the walkdown, the inspectors also identified the ASCO solenoid valve conduit installation problem discussed in paragraph 4 E.(2).

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