IR 05000348/1990003

Insp Repts 50-348/90-03 & 50-364/90-03 on 900423-27 & 0514. No Violations Noted.Major Areas Inspected:Design Control Activities,Engineering & Technical Support Provided to Operations & QA Audit of Design Engineering Activities

ML20043G807
Person / Time
Site:	Farley
Issue date:	06/06/1990
From:	Jape F, Peebles T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20043G803	List: IR 05000348/1990003 ML20043G802
References
50-348-90-03, 50-348-90-3, 50-364-90-03, 50-364-90-3, NUDOCS 9006210152
Download: ML20043G807 (28)

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UNIT E] STATES

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NUCLEAR REGULATORY COMMISSION

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Report Nos.: 50-348//90-03 and 50-364/90-03 Licensee: Alabama Power Company 600 North 18th Street Birmingham, AL 35291-0400 Docket Nos.:

50-348 and 50-364 License Nos.:

NPF-2 and NPF-8

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Facility Name:

Farley I and 2 Inspection At:

Alabama Power Company, Corporate Offices at Inverness Center, Birmingham, Alabama i

Inspection Conducted:

April 23 - 27, 1990, and May 14, 1990 Inspector:

Mem b d. oat kM7 O

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F. Jape (/ /

Date Signed

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Team Members:

K. Poertner C. Smith M. Thomas Contractors:

J. Klein W. Lundgren Approved by:

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_4-4 14 T. Peebles,'Ch'ief ~

Date Signed Operations Branch Division of Reactor Safety

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SUMMARY

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Scope:

This special, announced inspection was conducted at the corporate office in the areas of design control activities, engineering and technical support provided to operations, QA audits of design engineering activities, and review of licensee's self-initiated activities.

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Results:

In the areas inspected,- violations or deviations were not identified.

The licensee's NS staff is primarily'a project management organization responsible

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for administering all offsite engineering support.

Reviews of PCNs performed by the NS staf f are not performed to verify the technical adequacy of plant modifications prepared by outside engineering organizations.

The licensee maintains a qualified, experienced, and stable staff of contractors through its contractor design organization, to provide engineering and technical support to FNP. The interface between APCo and its contractor design organizations is good.

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9006210152 900606 i

PDR ADOCK 05000340 i>

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I QA audits of the design organizations were thorough and in-depth.

The

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inspectors' conclusions.of the PCNs reviewed were generally positive.

The

number of NS staff personnel having FNP experience prior to being assigned to I

the corporate is a strength. Some of the NS staff either currently hold or

have held a SRO license at FNP.

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The inspectors noted that much of the communications between FNP and APCo corporate offices was informal.

The inspectors considered the licensee's j

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timeliness in responding to various engineering support activities to be

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weak.

There was a weakness in the lack of ', involvement by the design organizations in identifying acceptance criteria and post modification testing i

requirements for PCNs. The training for NEL personnel is not very formal and.

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appears to be weak in the areas of 10 CFR 50.59 safety evaluations and PCN

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reviews.

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. REPORT DETAILS 1.

Persons Contacted Licensee Employees R. Fucich, Manager Nuclear Administration

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1. • J. Garlington, General Manager Nuclear Support

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1. • D. Jones, Manager Nuclear Engineering

D.-Mansfield, Manager Nuclear Maintenance Support

• D. McCoy, Engineer, Nuclear Engineering and Licensing

• J. McGowan, Manager Safety Audit and Engineering Review

1. • B. McKinney, Manger Nuclear Engineering and Licensing

• B. Moore, Manager Licensing

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1. L. Stinson, Assistant General Manager Plant Support, FNP

1. • L. Troutt, Engineer, Nuclear Engineering and Licensing J. Woodard, Vice President Nuclear Generation R. Woodfin, Engineer, Nuclear Engineering and Licensing

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Other licensee employees contacted during this inspection included

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engineers and administrative personnel.

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Other Organizations

• N. Antonio, Electrical Engineering Group Manager, SCS

• S. Arora, Mechanical Group Supervisor, Bechtel

• R. Conry, Quality Assurance Manager, SCS

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G. Daniels, Senior Designer, SCS

• T. DiPerna, Electrical Group Supervisor, Bechtel A. Gaylor, Project QA Engineer, SCS

• T. Higgins, SCS W. Hill, Manager Project Support and Administration, SCS'

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• F. Kuester, Project Engineering Manager, SCS

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l D. ' Lloyd, Mechanical Engineering Group Manager, SCS l

• J. Love, Project Engineer, Bechtel i

• R.

Lyon, SCS R. Ponder, Senior Engineer, SCS

• Attended exit interview

1. Attended telephone conference call on May 14, 1990 Acronyms and initialisms used throughout this report are listed-in the last paragraph.

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2.

Corporate Design Engine'. ring Nuclear support staf f duties involving design-engineering activities have

l been assigned to the Manager-Nuclear Engineering and Licensing (NEL). He

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is responsible for administering all off-site engineering support, which includes design changes and engineering studies.

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Additionally, he provides technical direction and administration of off-site engineering support not assigned to other sections.

The Manager NEL and his staf' do not perform reviews of PCNs to verify the technical adequacy of plant modifications prepared by outside engineering organizations, a.

Design Authority Organizations There are two desis i authorities that provide technical support to FNP by way of preparation, review, and approval of PCNs. Additonally,

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engineering support authorizations are used by APCo to initiate engineering studies which do not result in hardware changes to the plant and which may be performed by either organization. The scope of the responsibilities of both design authorities are specified in procedure 220100.4-100, Processing of Production Change Requests, Production Change Notice Revision Requests, Production Change Notices, Work Completion Notices and Proposed Production Change Requests, Revision 13. Appendix E of the procedure lists Category A structures and systems, assigns system design responsibility to Bechtel Power Corporation for Category A items with support design responsibility provided by SCS. For Category B items, SCS has been assigned system and support design responsiblity.

An interface document, contract number BHM-86-001, dated June 30, 1986, describes the interface procedures used for implementing the "/.greement for Engineering and Technical Services",

between SCS and Bechtel Power Corporation, b.

Staffing (1) SCS has a dedicated project team that provides engineering and technical suppor t to FNP.

The engineering organization

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consists of the following disciplines:

civil, electrical, mechanical, and instruments and controls. General administrative support to coordinate the efforts of the Farley Project team are provided by a project support and administrative staf f.

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Other off project support available to the project team are i

delineated in procedure 220100.2-0, Section 2, Engineering I

Organizations, Revision 6.

Management of engineering resources required to provide an adequate level of engineering and technical support for FNP is l-accomplished via Resources Assurance Team support described in l

SCS intracompany memo, dated July 19, 1989, _ from Doug Dutton, to:

Cost Center Managers, Resources Assurance Team

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Facilitators, Resources Assurance Team members.

Personnel

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L within the technical dist.iplines have a combined engineering I

and design experience of 13.1 years of which 9.5 is nuclear related.

The project tear consists of a total of 31 persons holding degrees.

Stability in project staffing level is demonstrated by SCS ma m ements established organizational j

goals concerning the limited use of contract employee y l

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(2) Bechtel Power Corporation provides engineering and technical support to FNP through a dedicated project team.

The project consists of the following disciplines:

mechanical process, electrical and I&C, civil and plant design.

The percentage of degreed engineers within the technical disciplines is 89 percent, with at least 12 employees having professional engineering licenses. Additionally, technical personnel have an average of 14.2 years employment with Bechtel Power Corporation, and an average cf 7.2 years employment on the Farley Project.

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staffing level consists of approximately 75 personnel with an anticioated peak in staffing level of 80 for 1990.

The inspectors reviewed the number of PCRs/ESs assigned to Bechtel Power Corporation during 1990 and discussed the work process employed for completion of assigned tasks.

Factors addressed during this discussion included prioritization of

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work, establishment of safety classifications for PCNs, procurement considerations, and design interfaces.

The inspectors concluded that a qualified, experienced and'

stable staff of contractors is maintained to provide engineering and technical support to FNP.

c, Communications The interface document to contract number BHM-86-001 requires that a correspondence routine be established by SCS for correspondence to and from Bechtel for the Farley project.

SCS Inc. Correspondence Routine for Farley Nuclear Plant, dated January 2,1990, was

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developed to satisfy this interface requirement.

Additional methods established for communicating needed design information across

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external-design interfaces were reviewed by the inspectors, i

Specifically, procedure 220100.4-100, was reviewed to verify the

l adequacy of the interface controls for design-engineering activities involving buth SCS and Bechtel Power Corporation.

No deficiencies were identified during this review, t

Procedure GO-NG-11, Design Change and Design Control, Revision 9,

establishes guidelines for the NS staff concerning design changes or new designs for FNP. This procedure specifies the administrative controls to be used for preparation, review, and approval of PCRs.

Administrative controls have been established to ensure that appropriate design basis information is transmitted to SCS and Bechtel Power Corporation via the PCRs. Completed PCNs prepared by l

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SCS or Bechtel Power Corporation are checked by NS staff project engineers using the guidance of Figure 6, Procedure GO-NG-11, PCN Review Checklist.

The inspectors were informed that this review of the PCN package is not intended to verify the technical adequacy of

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the design change.

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The extent of the design authorities involvement and communications with the site appears adequate during PCN development.

Monthly design change status meeting > are held to discuss and plan design, engineering, and related procurement activities, Although a formal process does not exist for the involvement of the design authorities during PCN installation, support is provided to the plant when requested by assignment of design engineering personnel to the site.

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The inspectors concluded that requirements for communications and interface design controls had been established for PCNs.

The extent of engineering and technical support provided by the design authorities for (1) initial station problem identification i

and evaluation for operability concerns and (2) installation and l

post-modification testing of PCNs appears to be weak.

Very little

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design engineering support to the operating unit is provided at the l

front end and the back end of the design engineering process.

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example, the use of ES by the site is not specifically addressed in procedure GO-NG-34.

Neither is design engineering support for immediate operability concerns or real time operating events addressed.

l An informal process used by the NS staff for disposition of ES is l

descrioed in paragraph 2.g. of this report.

General observations related to post-modification testing identified with PCNs are addressed in paragraph 2.f. of the report, d.

Vendor Document Control An evaluation of the licensee's vendor document control program was performed to verify that administrative controls had been

established for processing:

(1) vendor technical manuals / drawings I

required because of PCNs and (2) unsolicited vendor technical

manuals / drawings transmitted by the OEM for installed plant

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equipment. The following documents were reviewed during this effort:

Procedure G0-NG-22, Procedure for Nuclear Maintenance Support, Conduct of Operations, Revision 7 Procedure FNP-0-AP-4, Control of Plant Documents and Records, Revision 14 L

Procedure FNP-0-AP-65, FNP Operating Experience Evaluation Program, Revision 5 The inspectors determined that vendor technical manuals are handled l

by the NS staff in accordance with the requirements of the See-In-Program specified in procedure GO-NG-22.

Additionally, requirements have been established for site group managers to review

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vendor manuals to determine the effects on procedures, programs, and l

activities under their cognizance.

Revisions to existing vendor i

L manuals / drawings and additions of new vendor manuals / drawings are

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accomplished by completing the vendor manual / drawing revision form

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and transmitting it to the document control supervisor.

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e Information concerning revised vendor manuals / drawings received for installed plant equipment are transmitted to the design engineering organization by the document control center.

The design authority processes the revised vendor manual / drawing; issues a general revision to the affected document; and if necessary, provides comments on the vendor manual / drawing revision form that is returned to document control.

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The programmatic controls established for processing new vendor technical manuals and revised unsolicited vendor technical m6nuals

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appears to be adequate.

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Within this area no violations or deviations were identified.

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Setpoint Controi Program I

The inspectors evaluated the licensee's setooint change control program to verify the adequacy of design controls which ensure that original design objectives of safety systems are being met.

The following documents were reviewed during this effort.

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Procedure 220100.4-4, Preparing and Reviewing Design Calculations Procedure 220100.4-103, Process Instrumentation and Control

Setpoint, Revision 2 I&C Design Criteria for Joseph M. farley Nuclear Plant, Units

I and 2, Revision No. 1

I Farley Nuclear Plant Unit No. 1 Setpoint Document B-175968 Revision 17 Fadey Nuclear Plant Unit No. 2 Setpoint Document B-205968 Revision 14

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Review of the above objective evidence and discussions with SCS and

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Bechtel Power Corporation personnel revealed that setpoints changes are controlled as design changes.

PCRs which involve setpoint changes require preparation of a PISC to support the determination or r

verification of related setpoints in PCNs.

Setpoint configuration control is assured by revisions to setpoint documents to incorporate any changes in the setpoint data.

During performance of the SSSA on the Service Water System, licensee management determined that instrument setpoints were contained in multiple documents. To ensure better setpoint configuration control, a plan was developed for consolidation of setpoint documents. This plan is described in APCo intra-company correspondence, NDS-90-1555, dated March 7,1990. It involves control and use of the FNPIMS data base as the primary source of setpoint information with the exception i

I of MOV setpoints. This effort is estimated to require 18 months for completion.

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The inspectors determined that a standard design guide for performing i

instrument loop accuracy caTculations was nct available for use by i

SCS I&C staff.

Discussions with Lesign engineering personnel and review of objective evidence revesled that existing I&C design criteria specifies the correct methodology for performing the calculations.

The design authoritics agreed that a step-by-step precedure governing the technical aspects of the calculation does not exist.

However, it was their position that existing procedures and guidance ensures correct performance of the calculations.

Additionally, the inspectors were informed that evaluations of setpoint change controls were being performed in response to NRC IN 89-68 Evaluation of Instrument Setpoints During Modifications.

Within this area no violations or deviations were identified, f.

plant Modification Review The inspectors reviewed production change notices (PCN's),

calculations, procedures, and also held discussions with licensee personnel to assess the effectiveness of the engineering and technical support provided to FNP.

The results of the mechanical inspection were primarily positive.

However, improvements could be made in the areas of PCN reviews, formality in correspondence, and the interface between bechtel, Westinghouse and the licensee.

These concerns and observations are identifind in the discussion which follows.

1.

PCN B-87-1-4353, Revision 10, Replacement or Removal of Excess Flow Check Valves.

This PCN provided the authorization for the removal of two 3 inch

excess flow check valves from the Chemical and Volume Control System (CVCS) letdown li r.e of Farley Nuclear Plant Unit I and for an increase in the high letdown flow alarm setpoint.

The PCN was initiated as a result of maintenance problems experienced with these valves and the plant's inability to obtain safety-related replacement

parts.

A production change request was issued to Bechtel to provide a reliable replacement or an acceptance for the removal of the valves.

Due to problems in establishing a valve supplier to provide an exact replacement for the existing valves, Bechtel pursued

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the option related to the elimination of these valves.

In the course of developing this PCN, Bechtel interfaced with Westinghouse, the original designer of the CVCS.

Westinghouse provided acceptance of the modification provided the following precautions and monitoring checks were conducted:

Regenerative Heat Exchanger Shell Side Outlet Temperature does not exceed 380 degrees F (High Alarm 365 degrees F).

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Letdown divert on high temperature setpoint is not approached (to protect demineralizer resin, approximately 135 degrees F).

Letdown Heat Exchanger Tube Outlet Temperature can be controlled to approximately 115 degrees F.

Low pressure letdown line control valve (PCV-145)

continues to maintain adequate back pressure downstream of the letdown orifices (approximately 257 psig setpoint).

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Delta P across the demineralizer should be 1 25 psid.

The inspectors requested the licensee to demonstrate how these precautions were addressed.

The licenste provided the inspectors with documentation which indicated that the first four precautions were implemented.

However, the last precaution was not implemented in the operators logs or procedures.

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Westinghouse included this precaution to prevent potential failure of the demineralizer screen under high flow conditions.

The licensee committed to ravise the appropriate operating procedures to

monitor the pressure drop across the demineralizer, In addition, the licensee indicated that the FSD which will be prepared for the CVCS will be expected to identify the 25 psid as the bounding value for differer.tial pressure across the demineralizer.

The inspectors made the following observations in reviewing this PCN:

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The mechanical calculation referenced on Page R-2 of the PCN did not indicate on the cover sheet the applicable PCN and the appropriate revision level of the PCN for which the calculation was prepared.

This is a requirement of Procedure 220100,4-4, Revision #8, dated l

May 3, 1989.

The calculations only referenced an Engineering Study

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l number, b)

The PCN referenced Revision 0 of Farley Mechanical Calculation #25.9 (Bechtel Calc.

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531-5-7597) whereas the calculation was Revision 1,

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The above comments were considered minor and were brought to the attention of the licensee for their consideration.

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PCN B-89-1-5770, Revision 0, MOV 8701B and 8702A Removal of NAMCO Limit Switches.

This PCN provided the design for removal of the stem mounted limit switches and their associated circuits from motor operated valves (MOVs) 8701B and 8702A of Farley Nuclear Plant Unit 1.

The PCN was initiated because the function of the limit switches was eliminated due to the removal of interlocks which was accomplished by PCN 89-1-5748.

l No deficiencies were identified within this PCN package.

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PCN 87-2-4714, April 26, 1989, Replacement of Limitorque Motor Operators.

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This PCN provided the design to replace SMB-0-15 limitorque motor actuators on Farley Nuclear Plant, Unit 2 with SB-0-15 Limitorque motor actuators for valves MOV 8130A & B, MOV 8131A & B, MOV 8820A & B, and LCV 115B & D.

The PCN was initiated because the installed motor operators had brake assemblies for which no replacements could be obtained from Limitorque.

Also, analysis of test data in response to IE Bulletin 85-03 revealed that the MOV's exhibited high inertial thrust after torque switch trip while seated.

The PCN was prepared by Bechtel.

Bechtel's acceptance of the replacement motor operators was based on Westinghouse's evaluation and approval of this change, since Westinghouse is the original supplier of the valve

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assemblies.

In the p #:ess of reviewing the documents to be revised, as listed in the

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<spectors expressed a concern that the seismic analysis for the valvt

.,mblies were not identified in this listing.

The licensee and Becht_.esponded that the seismic analyses were in Westinghouse's possession and under their control. The list of documents to be revised, as indicated in the PCN, only reflected the documents under the

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l-responsibility and control of Bechtel. The inspectors considered this to

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be a weakness in the interface between Bechtel and Westinghouse.

l At the request of the inspectors the licensee telephoned Westinghouse to determine the status of the seismic analyses.

Westinghouse informed the licensee that new seismic analyses were developed for the eight valve assemblies, to reflect the new motor operators. The old seismic analyses L

were marked as superceded.

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The licensee is aware that there is a weakness in the interface

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between Bechtel and Westinghouse, as a result of findings identified in a Quality Technical Audit conducted by the licensee of Bechtel in 1989.

Based on the audit's findings, Bechtel has revised their Project Work Request Procedure EDPI 3.16-17 to further clarify the interface responsibilities between themselves, and Westinghouse.

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Calculation 1.6, Revision 2, Charging / Safety Injection Pump NPSH, This calculation was performed to determine the available net positive l

suction head (NPSH) to the high-head safety injection pumps for suction flow from the refueling Water Storage Tank for Units 1 and 2.

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i The inspectors' review of this calculation found it to be thorough, complete and satisfactory, k

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Calculation 30.0?, Bechtel CALC. No. 513-55-7597, Revision 0, Main Steam Line Break (0 percent Power Cases) with 10 Seconds MSIV Bypass Line Valve Closure Time.

i This calculation was performed to analyze the effects of increasing the isolation time, from five seconds to 10 seconds, for the main steam

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isolation bypass line valves.

The calculation referenced Engineering Study (ES) 89-1380.

The inspectors reviewed this study and had the i

following observations:

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Westinghouse submitted safety evaluation check list (SECL) No.89-009, to Alabama Power Company.

The SECL was not signed (only names were typed-in) nor dated and was submitted informally ty a typed note without any date.

This finding was brought to manapment's attention for resolution.

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The inspector also reviewed the following modification packages:

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B-88-1-4763, Eliminate BIT Bypass Line, Revision 1 -

This modification eliminated the BIT bypass line to prevent potential leakage through BIT bypass valve 8911.

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B-88-2-5434 Replace MOV 3406 Motor Operator, Revision 2 -

This modification replaced the existing two f t.- lb motor on MOV 3406 with a qualified five ft.-lb motor.

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B-86-2-3796, Replace Charging Pump Cooling Water Supplies, l

Revision 28 - This modification replaced the existing service

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water supplies to the gear oil and bearing oil coolers of the chargining pumps with component cooling water supplies.

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During review of the PCN packages, and discussions with APCo and design'

organization personnel, the inspectors determined that the applicable design organizations have very little involvement in the identification l

of acceptance criteria or post modification testing requirements for PCNs.

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APC0 management stated that this function is considered FNP's responsibility.

The appropriate design organization provides support when requested by FNP. The inspectors stated that the design organizations should be more involved in identifying PCN acceptance criteria. The design organizations are more knowledgeable of the design requirements for the PCNs assigned to them. Guidance should be provided by the designer which will demonstrate

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that the PCN has been installed in accordance with design requirements, the I

design change will accomplish its intended function, and no other safety features were adversely affected.

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The inspectors consider the lack of involvement by the designer in identifying PCN acceptance criteria or post modification testing requirements to be a weakness.

No violations or deviations were identified during review of the above plant modification packages.

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Engineering Studies The APCo nuclear support staff is responsible for the resolution of ES.

The majority of these are initiated at the site and sent to NS for review and resolution.

After receipt of the ES they are assigned to either the engineering staff of Bechtel or SCS for performance of detailed design engineering tasks, i.e., studies, calculations, reports, investigations, etc.

Upon completion, the specific aeliverable is forwarded to APCo NS staff for review and subsequent transmittal to the plant, if necessary.

APCo tracks the status of each ES via a Schedule and Budget Report issued by the NS department.

It appears that ES contribute a significant majority

of the engineering tasks for which APCo is responsible.

The review of the specific electrical ES, that were identified on the Budget Report, concentrated on the following three areas for FNP:

(1) DC Electrical System (2) Emergency Diesel Generator Loading Analysis (3) AC Electrical System Specific observations and comments for each system are addressed in the following paragraphs.

(1) DC Electrical Systems The initial plant battery sizing calculation, E-13, was performed in 1971.

In 1987, calculation E-95 was developed to address the station blackout event.

This calculation verified that the original load profile in calculation E-13 was conservative.

The inspector determined that various ambient temperatures have been mentioned in both calculations, and in engineering correspondence between the design engineering organization and APCo as a result of additonal ES.

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The inspector questioned if APCo had base 11ned the DC system such that any future load additions could be properly evaluated against the actual loading profile.

This baseline document should also address the correct temperature correction factors.

The inspectors noted that on March 16, 1989, the Design Adequacy Review Committee (DARC) had indicated that a model of the DC system would be illustrated on a controlled drawing such that revisions and deletions per design changes would be factored into the model.

Per the DARC meeting minutes, this would then result in an as-built status of the DC loads such that the design adequacy of the DC system can be assured.

With regard to the current status of this controlled drawing, the inspectors were informed that the draft load profiles for the safety-related auxiliary building batteries had been issued to FNP site for review. On April 26, 1990. FNP indicated by memorandum that the site had completed its review of the draft load profiles and found them to be acceptable, Bechtel was to be requested to issue the drawings by June 25, 199.

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The inspector review the APCo resoonse to NRC Information Notice 89-16, entitled " Excessive Voltage Drop fn DC Systems," dated February 16, 1989.

As a result of a plant request cated, April 5,1989, APCo requested Bechtel to review the concerns addressed in IN 89-16.

This request was made on June 16, 1989.

Subsequent correspondence between APCo, Bechtel

and specific equipment vendors was reviewed.

Specific direction was l

1ssued to the plant on April 6, 1990, advising them of the current status of the engineering response and identifying the need for field-surveillance testing of six circuits.

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This is an example of lack of a timely response to an engineering support activity. The issue has been identified for 15 months and has not been

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formally resolved as of this inspection.

The response or analysis of existing plant conditions with respect to engineering concerns should be promptly addressed.

Based on the above, the following action plan was proposed by APCo:

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Issue the baseline DC load profile document as a controlled

design document by June 1990.

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Revise calculation E-13 to identify what portions of the calculation are superseded by calculation E-95 by June 1990.

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Develop a method for assessing and documenting small loads by June 1990.

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Factor the results of ES-89-1468 into calculation E-95 and the load profile summaries by June 1990.

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Establish and document the design temperature to be used for IEEE 485 battery sizing.by June 1990.

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Revise all documents, as necessary, to incorporate the established temperature by July 1990.

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During the upcoming refueling outages, data will be taken during the SI/LOSP testing to resolve the potential voltage margin issue.

The action items described above will provide a consolidated design basis for the safety-related auxiliary building batteries based on the current industry methodology contained in IEEE 485. The committment to resolve this issue will be tracked as IFI 50-348,364/90-03-01, DC l.oad Profile.

(2) Emergency Diesel Generator (EDG) Loading Analysis

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l On April 4,1990, APCo reviewed Revision 2 of the EDG Loading Analysis Calculation No. 42. The calculation contained the following statements.

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In addition, this study identifies those miscellaneous loads which are not required for safe shutdown.

APCo should take whatever action is required to insure that the miscellaneous loads not recuired for safe shutdown, and indicated as 'normally off', will be inoperable during an LOSP".

(Revision 3, Sheet 3,Section III, Design Basis for the Miscellaneous Loads).

In all of the analyzed situations Diesel Generator 2C is greatly underloaded (10.9 percent of its continuous rating).

The diesel generator manuf acturer recommends that the diesels be loaded to 50 percent of their continuous rating.

Diesel generator 2C must therefore be loaded manually or tripped in order to comply with the manufacturer's recommendation.

(Revision 3, Sheet 4,Section IV, Conclusion).

The inspector asked ApCo what action had been taken to ensure that the l

miscellaneous loads and emergency diesel generator 2C are operated as required.

APCo noted that the statement concerning operation of the miscellaneous loads, not required for safe shutdown, was intended to reflect the fact that assumptions were made for the calculation regarding the status of various miscellaneous loads such as motor operated valves, room sump pumps, motor heaters, room heaters and ventilation fans (depending upon the season). These assumptions were based upon discussions held at'

meetings attended by FNP and Southern Company Services personnel during i

the preparation of the load study.

The only specific restriction placed

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upon operation of individual miscellaneous loads involved the Train A

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room heaters at the service water structure,, This restriction was added l

in Revision 1 of the calculation which was transmitted by Bechtel to

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SCS and APCo during March 1980. Sheet 155 of the Revision 1 calculation

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l contained a note requiring that pump roam heaters D E, J and K be "normally off (inoperable during LOSP) when both units are in operation".

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In order to meet this requirement, the plant initiated a tagging order to I

ensure that the required heaters were not loaded on the MCC.

This is still maintained under Tagging Order number 81-496-0 pending formal notification j

that the restriction has been determined to be no longer required (as reflected by Revision 3 of the calculation).

During evaluation of the NRC's concern, APCo determined that the calculation also contains the following statements concerning alignment

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of the shared Train A motor control centers:

J With both units in operation, the shared MCC's IF and 1K should i

always be powered from Unit 1.

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l The shared MCC IX should always be powered from Unit 2, when both

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units are in operation, due to the lube water booster pump A required for the operation of the Unit 2 Train A river water pumps.

(Revision 3, Sheet 2,Section I, Design Basis for Operation of the Shared MCC's - Train A).

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Review of the current System Operating Procedures for the 600, 480 and

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208/120 Volt AC Electrical Diitribution System, FNP-1/2-SOP-36.3

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revealed that there are no precautions in the procedures to prevent alignment of these MCC's contrary to the assumptions of the calcultion.

Plant personnel were requested to check the current alignment of the MCC's and found that MCC's IF,1K and IX were aligned to Unit 1.

As e result of this review, the following actions have been taken or will be taken to address these concerns:

1)

MCC IX has been realigned to be powered from Unit 2, which is the normal alignment per procedure FNP-1-SOP-36.3.

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A precautionary note will be added to procedure FNP-1-SOP-36.3 to indicate that aligning MCC IX to Unit I with both units at power may cause the 2000-hour rating of Emergency Diesel t

Generator 1C to be exceeded. This will provide for consistency

between the MCC alignment and the assumptions in the calculation during dual unit operation.

Followup on these commitments will be tracked as IFI 50-348,364/90-03-02,

EDG Loading Conditions.

With regard to the EDG loading question, current operating procedures (normal and emergency) instruct the operator to avoid operation of the diesels for extended periods in an unloaded condition.

However, the 50 percent threshold is not specifically contained in the procedures.

Formal correspondence between the equipment vendor and APCo placed the 50 percent loading recommendation in the context of regular surveillance or maintenance testing of the diesel. This is documented by the original purchase proposal for the diesels in letters dated December 18, 1981, January 24, 1984 and July 26, 1984, from the vendor to APCo.

The July 26, 1984, letter stated that diesel operation at a minimum of 50 percent load or 600 degrees F minimum cylinder exhaust temperature is recommended to prevent oil accumulation in the exhaust manifolds.

The project engineer currently responsible for the EDGs could remember no conversations in which he was involved where minimum loading of the diesels under emergency conditions was mentioned as a concern by Colt Industries.

The operating manuals for the EDGs were updated in 1989 to incorporate vendor information notices received since the original issuance of the manuals. No vendor information letter concerning this issue was included in the updated manuals.

The only guidance provided by the vendor in the current operating manuals is as follows:

The unit is designed for emergency starting and operation with the keepwarm system operable and normally maintained room temperature.

Common anse practices with large equipment operation should be followed when exercising the unit below normally maintained temperatures.

These practices include:

start engine at 900 RPM speed setting, apply partial load when synchronous speed is reached, and do not operate at rated speed over long time periods without load on the unit. These practices extend the unit service lif *

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This guidance has been incorporated into the System Operating Procedure,

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FNP-0-SOP-38.0 and the surveillance procedures for exercising the EDGs.

These procedures ensure that the EDGs are loaded at their rated load for a minimum of one hour.

The stated reason for this restriction in the I

operating procedure is to minimize the potential for exhaust header fires.

However, this is primarily a concern when the engine is operated at below normal temperatures as stated in the operating manual.

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The inspector requested that APCo obtain clarification of the minimum

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loading requirements under emergency operations and upon receipt of this information, revise the plant procedures, if required.

Followup of this question will be tracked by IFI 50-348,364/90-03-03, Clarification of EDG Loading Restriction.

On December 2,1986, APCo issued ES 86-765 to Bechtel to review the electrical design deficiencies identified in NRC Information Notice 86-70 and to determine if the same, or similar, deficiencies exist in the design of the FNP EDGs..The Bechtel response to the ES (letter AP-14230 dated February 22,1988) stated that the deficiencies identified in the Information Notice do not exist in the design of the FNP EDGs and that

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l adequate administrative controls exist to ensure that such deficiencies L

will not occur as a result of future design modifications performed by Bechtel or SCS. This letter did note that load center transformer losses (low Kw values) had not been directly accounted for in the EDG load evaluations; however, it was Bechtels engineering judgement that, I

conservatism in the EDG loading calculation compensates for these losses.

l Bechtel requested direction from APCo with regard to the necessity of revising the calculations to directly include these losses.

Bechtel was l

not directed to include these losses at that time. On March 21, 1988, APCo informed FNP of the Bechtel evaluation results, l

It was determined from further discussions with APCo engineering personnel that Bechtel had been requested via letter No. AP-17086, dated January 12, 1990, to re-evaluate the feasibility of adding new loads -

l automatically onto the EDG. These loads were identified as instrument l

air compressors, pressurizer heater groups and a Unit 2 containment cooler.

I In order to perform this re-evaluation, the EDG load study must be revised.

The above APCo response indicates that the current loading of EDG IC is now 2999 kw plus an incremental value for the transformer losses.

This

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was determined as a result of Bechtel evaluating all possible design combinations for alignment of the shared motor control centers (MCC)

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The bounding case was MCCs IF, IK, and IX connected to Unit I during a dual unit loss of offsite power with a LOCA l

occuring on Unit 2.

This would result in the continuous rating of 2850 i

being exceeded by approximately 149 kw, but it would be less than the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating of 3100 kw.

Upon completion of the ongoing EDG

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re-evaluation study, APCo should ensure that all calculations are current and consistent with plant documents. This will ensure that all subsequent load modifications or changes to the AC System and EDGs are co-ordinated, documented properly, and available for subsequent evaluation as required.

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(3) AC Electrical System

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This system is under the control of Southern Company Services.

During discussions with cognizant SCS engineers, it was determined that SCS is

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in the process of completing a detailed analysis of the system. The draft response indicated that they were deleting the assumption " Starting of a large Non IE motor concurrent with SIAS."

This assumption was in the original analyses in response to NRC, but due to voltage restrictions on their system, this restricted load expansion.

Therefore, SCS has performed a (draft) safety evaluaton that addressed the acceptability of not including the simultaneous start of a condensate pump for the updated load study.

The NRC issued a letter in 1979 to all licensees that addressed the

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adequacy of station electric distribution system voltages.

The letter provided guidance to licensees for performance of voltage drop calculations.

One guideline indicated that consideration to the starting of large l

non-safety related electric loads, such as a condensate pumps. should be I

included in the plant analysis.

Each licensee was required to perform a l

review of the electrical system in accordance with the letter and to

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provide a copy of the system analyses to the NRC.

The licensee has indicated that the revised analysis may not be submitted to the NRC. The

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question of whether or not the revised analysis should be submitted to the NRC will be tracked as IFI 50-348,364/90-03-04, Revised AC Load Analysis.

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Other ES Observations In addition to the above three areas of electrical engineering the team noted the following observations:

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Timeliness of Responses Engineering Study 89-1570, entitled "Possible Submergence of

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i Electrical Circuits Located Above Flood Level Because of Water i

Intrusion and Lack of Drainage" was issued to the respective design organization on February 5,1990.

This was in response to NRC Information Notice 89-30, High Temperature Environments at Nuclear Power Plants, dated March 15, 1989. The inspector noted

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that the completion date was identified as May 1,1990, and inquired as to its status.

APCo advised that the designer has the response in the review l

cycle and that the new target date for completion is l

June 15, 1990.

APCo further stated that the original target l

date was based on designer input at the time of the ES issuance and was based on the designer workload at that time.

The designer is now currently addressing higher priority items which have diverted resources away from this ES.

APCo stated that there are

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no requirements for addressing Information Notices within any l

particular time limitation.

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One observation that was identified during the licensee's SS$A, was that the as installed condition of several motor-operated valves, that are located in valve boxes, may not be adequate to ensure operation of these valves if the motor operators are submerged.

The response by APCo determined that it was not necessa ry for the valves to be equipped with submersible operators.

The inspector expressed concern that the investigation associated with ES-89-1570 is not receiving appropriate attention in light of the fact that a similar concern with regards to submergence was raised and resolved during the SSSA.

This example and the chronology associated with the resolution of this and other issues identified previously in this report, e.g., response to IN-89-16 and 86-70, indicate that issues are not always receiving the appropriate attention necessary to provide timely responses.

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Interface between Bechtel and Southern Company Services The design engineering responsibilities for F2 are split

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between Bechtel and SCS.

Tasks associated with the AC power distribution system are performed by SCS while tasks associated with the EDGs are performed by Bechtel.

In review of various Engineering Studies and the associated documentation, the team found that co-ordination between the two organizations was well defined and accurate. With tasks assigned to different organizations proper co-ordination in the exchange of design information is a key interface, i.

Minor Departure from Design Program The inspectors held discussions with licensee management concerning the MDD program.

These discussions were a followup to IFI 50-348, 364/89-17-01, and previous concerns over the MDD program.

The inspectors expressed the following concerns with the licensee's implementation of MDDs.

The first concern involves the level of control during implementation of MDDs.

There are no requirements for either on-site or off-site engineering

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review of MDDs prior to implementation.

The inspectors expressed concern that the current level of approval

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authority for MDDs at FNP may not have adequate technical knowledge or access to design basis information in all areas that could be affected by the various MDDs.

The current process involves utilizing a checklist during the preparation and review of MDDs. This process does not require the reviewer to document areas considered in reaching conclusions during the safety evaluation screening process.

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The second concern involves the scope of the changes which can be made under the MDD program.

The administrative controls are such that design changes can be made under a MDD.

The inspectors reviewed the following modification packages that had originally been implemented via the minor departure process and PCNs were required to obtain a drawing change resulting from MDDs:

89-1-6256 This modification package revised drawing U-258363

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(Main Steam Hangers) to show the modification of a pipe support frame for the main steam safety valve discharge piping.

86-1-3931 This modification package jumpered/ bypassed the

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manual cut off switch for the air heater located on the penetration room filtration system.

89-1-6296 - This modification package documented the re-routing and capping of the CCW drain line from the RHR heat exchanger downstream of valve 01P17V0124B.

89-1-6288 - This modification package documented the acceptability of one mounting screw being missing from the post LOCA sump cover grating cage due to pipe interferences that made installation of the screw impossible.

The inspectors reviewed the associated minor departures and determined l

that the acceptability of the minor departure was determined prior to installation and return to service via interaction with the associated vendor or design authority, i

The inspectors considered some of the MDDs to be design changes and were beyond the scope of what the licensee considers as MDDs.

Licensee

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management stated that they did not consider the changbs made under MDDs to be design changes, and therefore, the controls established for design changes do not apply to MDDs.

The inspectors informed the licensee that IFI 50-348,364/89-17-01 would remain open pending further review by the NRC of the concerns relating to MDDs.

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Management Support The Farley Project Final Approval Matrix, Revision 6,

dated

August 3, 1989, documents final approval within the SONOPC0 Project for activities being performed by Farley Project and FNP personnel.

Job authorizations such as manning levels and organizational changes require approval at the Senior Vice President level or higher.

For procurement activities, the GMNS can approve project blanket order releases up to $250,000; and project change order requests up to $100,000 O&M and $500,000 capital.

The VPN Farley Project can approve blanket order releases up to $1,000,000; and change order requests up to $1,000,000 for O&M and $3,000,000 for capital.

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amounts from the O&M and capital budgets have to be approved at the Senior Vice President level or higher.

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Licensee management's efforts in improving site engineering support is evident in the establishment 6f-the Maintenance - Engineering Support Group at FNP.

Licensee personnel stated that addition of this group was approved at the Executive Vice President level.

Staffing for this group is included in the FNP 1990 budget.

FNP managemont is currently in the

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process of filling the various positions within the group.

Licensee

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personnel stated that this group was established, in part, due to NRC concerns over engineering support for maintenance, identified in NRC Inspection Report 50-348,364/89-10; and the licensee's self assessment.

l of its maintenance program.

l ApCo management's policy is to encourage its technical staff and managers to become active in professional societies and nuclear industry groups.

l APCo's industry 4 ;aciation and participation includes EPRI, NUMARC, WOG, i

etc.

The inspc' ', concluded from reviewing the 1989 capital and O&M budget expenditures and from discussions with management personnel concerning FNP's 1990 Capital budget, that APCo is committed to supportin0 engineering needs and enhancing the safety of the plant.

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Quality Assurance j

The inspectors reviewed selected 1989 QA audits of design control activities.

Audit findings, responses to the findings and reaudit reports were also

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reviewed.

The following audit reports were reviewed:

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(1) Quality Technical Audit Report, Bechtel Power Corporation File:

ENG 11-3-FAX-BE

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(2) Quality Technical Reaudit Report, Bechtel Power Corporation I

File:

ENG 11-3-FAX-BE J

c (3) -Quality Technical Audit Report, SCS Farley On-Project Activities

File:

ENG 11-3-FAX-i (4) Quality Assurance Reaudit Report, SCS Farley Project

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File:

ENG 11-3-FAX i

(5) QA Audit of. Westinghouse, Concentrating on Farley, Vogtle, and Pooled Inventory Management (PIM) Project Activity - Audit Report, August 14-17, 1989 i

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File:

EGN 11-3-W (6) Quality Assurance Audit Report, SCS Nuclear Operational Support Of f-Project Activities, Farley, Hatch, and Vogtle Projects File:

ENG 11-3-FAX, ENG 11-3-ilAX, ENG 11-3-V0X (7)

SAER Audit of Nuclear Support - Nuclear Engineering and l

Licensing SAER File: A35.94.16 Except for the SAER audit of Nuclear Engineering and Licensing, all of the other audits were performed by SCS QA on behalf of APCo's corporate SAER group.

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The audits of the aplicable organizations reviewed are performed annually,

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This annual audit frequency exceeds the TS frequency of every two years.

The reaudits were performed to review implementation of corrective actions taken for the open audit findings and audit follow-up items. In addition to SCS and SAER QA personnel, the audit teams also consisted of engineering personnel from SCS and APCo, as applicable, who served as technical specialists.

The audit teams reviewed activities and modification packages for technical and programmatic compliance.

The audits reviewed by the inspectors were considered to be thorough, in-depth, and effective in identifying problem areas.

Various requirements of the licensee's Operational Quality Assurance Program as implemented by the design authorities were reviewed by

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The inspectors determined that the record retention

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time for SCS quality compliance audit reports is six years as delineated in procedure 220100,5-100, Handling of QA Records, Table o

5-100-1, which is in agreement with ANSI N45,2,9 (1974).

TS Section 6.10.21, specifies record retention for the duration of the unit operating license.

A discussion was held with licensee mangement and SCS management regarding the difference in record I

retention requirements.

It was determined that the APCo's records j

are sufficiant to satisfy the requirement as stated in the TS.

l However SCS committed to revise their procedure for QA audit report

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records from six years to life of plant.

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In addition to the QA activities performed, the licensee also established

a Design Adequacy Review Committee (DARC).

The function of the DARC was

to provide independent review and assessment of design activities.

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Licensee personnel stated that ApCo committed to INPO to conduct DARC activitier >

' vear, consisting of quarterly' meetings, in order to i

provide a et

,iiaependent review of the design preparation and

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independent wofication processes. A DARC meeting was held during each

quarter o 1980 The inspectors reviewed the minutes from each of the y

quarterl.s JARC meet 1'

and the 1989 DARC recommendations.

The insp(*',

c 1ereu the DARC meetings to be thorough and detailed in their -

• - '. w ;ichnical adequacy of design activities.

Licensee. m o a aed that although the commitment was to conduct

DARC acth '

%r sne year, they plan to continue with the DARC meetings,

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because the meetings have been beneficial in identifying potential weaknesses

in -design control activities. The inspectors considered this a positive indication of licensee management's support of efforts to improve the quality of design control activities, 1.

Training The inspectors held discussions with licensee management concerning training for NEL personnel. ApCo procedure G0-NG-13, Revision 6, Documentation of Education, Training, Experience and Qualifications for Farley Project Nuclear Support Personnel. The inspectors reviewed this procedure and the training records for selected individuals within NEL.

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i In general..the inspectors noted that training requirements were not very formalized or specific with regard to minimum requirements for NS personnel below manager level.

For example, per APCo procedure GO-NG-11, Revision 9,

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Design Change and Design Control, PCRs and PCNs are assigned to project i

engineers within the Nuclear Engineering Section for processing and tracking the design changes during design development. Project engineers are required to review the PCRs and PCNs and complete the applicable

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GO-NG-11 checklists.

The inspectors noted that neither GO-NG-11 or any other GO-NG procedures are specificrJ ?y rererenced in GO-NG-13 as required training for NS personnei.

lw m gectors discussed this matter with licensee personnel

who statk that c similar finding was identified during the 1989 SAER

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audit A %uel training on G%NG-11 was implemented in 1989 for NEL personnel as a result of the sad audit finding.

The inspectors noted that the training docume'1ted for the individuals reviewed did not include training on 10 CFR 50.59 safety evaluation

preparation or review.

The inspector also noted that GO-NG-13 did not I

provide any guidance for 10 CFR 50.59 training.

The inspector asked

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licensee personnel if there was a list of designated or qualified 10 CFR 50.59 safety evaluation preparers and reviewers within either NEL i

or NS.

Licensee manageme.it stated that there is not a list of 10 CFR 50.59

qualified preparers or reviewers, nor is there formal training covering q

10.CFR 50.59. The inspectors considered the lack of 10 CFR 50.59 training

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to be a weakness since NEL personnel are involved in 10 CFR 50.59 l

preparation and review relating to TS changes, PCR and PCN reviews, and

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procedure changes. NEL management stated that there has not been a formal

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10 CFR 50.59 training program because of limited involvement by NEL

personnel in 10 CFR 50,59 related activities.

The TS changes performed by NEL are minor in scope, and the PCR and PCN reviews are not performed to determine technical adequacy of the applicable PCRs and PCNs.

i Licensee personnel stated that the corporate training records reviewed by

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the-inspectors did not reflect training NEL personnel may have received at FNP.

For example, several NEL managers and engineers hold current SRO licenses or have received SRO training at FNP.

These individuals were involved in 10 CFR 50.59 evaluations when they were assigned to FNP.

It was further stated that in addition to the annual training required by G0-hG-13, managers within NEL may also identify additional training

requirements for their staffs. In some cases, this has included systems i

training at FNP which would not be reflected in the corporate training

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records. The licensee stated that a task force has been formed within SON 0PC0 to develop a training program covering 10 CFR 50.59 based on the NUMARC document NSAC/125, Guidelines for 10 CFR 50.59 Safety Evaluations.

The inspectors acknowledged the licensee's explanation of additional training received by NEL personnel. The inspectors noted that the number of individuals within NEL who have SRO training or have previous experience at FNP is a strength for NEL. The inspectors still consider the lack of training on 10 CFR 50.59 for those involved in 10 CFR 50.59 reviews and the lack of a formalized training program to be a weakness.

The actions being taken by the licensee in this area should resolve this weakness.

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Reporting Procedures GO-NG-10, Corrective Action provides direction for the establishment of a system to monitor and control follow-up action required to correct non-compliancies identified within the area of responsibility of the Manager-Nuclear Maintenance, Manager-Nuclear Administration,

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Manager-Nuclear Engineering and Licensing and to provide guidance to the Nuclear Support Staff concerning the processing of corrective

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action reports generated at the plant. site.

The engineering

l organization is responsible for the disposition and resolution of NRC'

Bulletins and 10 CFR part 21 reports and is involved in reportability determinations if a request is made by plant site personnel.

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engineering organization does not trend LER reports or plant parameters, however this function is accomplished at the site and the engineering

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organization would assist this effort if requested, n..

Configuration Management

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Recognizing the need for self-assessment in the area of configuration management due to increased industry and NRC attention in this area,' the licensee initiated a self-initiated safety system

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assessment (SSSA) of the service water system.

The purpose of the SSSA was to evaluate the effectiveness of the configuration management program and to identify areas of the program that could be improved.

The SSSA on the service' water system was conducted May 22-June 23, 1989, and the final report was issued-on September 15, 1989.

The SSSA identified five functional areas of weakness and five weaknesses in the area of configuration. control associated with the service water system.-

The functional weaknesses consisted of.

incomplete, out-of-date, and inconsistent design bases; inattention to-equipment indirectly related to-the-system;-incomplete calibration andL testing program; inconsistent definition and application of

" safety-related" and lack of understanding of system design bases by licensee personnel. The configuration control weaknesses consisted-of communications breakdowns between design and operations, heavy

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reliance on operator actions rather than design, incomplete

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documentation revisions resulting from plant changes, FSAR inconsistencies and unsupported statements, and I&C drawing inconsistencies.

The SSSA

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also identified strengths in the areas of physical implementation of design changes, procurement for design changes, operator knowledge and awareness of the system configuration and operation.

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The inspector reviewed the scope and status of the licensee's configuration mangement project. The licensee's program consists of the following action plans:

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- Prepare Functional System Descriptions (FSDs)

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Review FSAR to identify requirements 3.

Revise the Q List 4.

Issue Configuration Management Manual 5.

Consolidate Setpoint Document

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Identify documents to be revised or voided for Design ~ Changes

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Trend Maintenance Work Requests

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Establish Outbuilding Cleanliness Program 9.

Prepare Attendent Equipment. List 10. ~ Establish Preventative Maintenance Criteria The configuration management project is scheduled to take five years to complete and is presently scheduled to be completed by December-1994. The licensee does not consider that the weaknesses

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identified by the SSSA to be indicative of a breakdown in the configuration control program. The configuration management project was initiated to enhance the configuration control program and:

resolve the weaknesses identified by the SSSA.

The purpose of the FSD program is to define the functional design requirements and to consolidate existing system and component functional

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design requirements in an understandable and easy - to use format.

The FSDs will document design functional requirements for all plant modes-as well as accident conditions. The licensee plans to develop FSDs for the following systems:

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Service Water Component Cooling. Water Residual Heat Removal / Low Head Safety Injection Containment Isolation Electrical Distribution Diesel Generators Control Room Ventilation Reactor Protection System Containment Spray Chemical.and Volume Control /High Head Safety Injection / Accumulators Auxiliary Feedwater Post Accident Sampling System Instrument Air These systems were selected based on the following criterion:

I Provide a safety related function i

Significant design changes have been implemented since original design Systems with a history of problems in the nuclear industry Systems which APCo perceives the most benefit for plant ~

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safety and reliability Systems that could potentially identify new programatic design issues not identified during the past SSSA Systems that afford control of radiation or radioactivity

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either to the public or in plant

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-i The FSDs will be prepared by the original design organization which performed the major detailed design for the construction and operational modifications to the system.

These organizations consist of Bechtel, Westinghouse and Southern Company Services.

The licensee has developed a procedure for preparation of the functional system descriptions.

As part of the FSD preparation process, design source documentation

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inc'uding the FSAR will be reviewed and descrepancies, inconsistencies-or neomplete design requirement information will be identified and processed for resolution as outlined in the procedure for preparation of she FSDs.

The licensee also plans to perform an SSSA after the com;1etion of each FSD to verify the adequacy of the FSD.

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As of this inspection, pilot FSDs were being developed on the service water, component cooling water, and residual heat removal / low head safety injection systems. These pilot FSDs are scheduled to be issued in draft form for review and comment by the end of May 1990, with

issuance of Revision 0 scheduled for August 1990.

Revision 0 and

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each subsequent revision of a FSD will be processed and controlled by the same procedures that control design drawings.

The purpose of the FSAR review program is to identify design requirements and design guidelines contained in the FSAR and to correct any errors or inaccuracies.

The licensee is presently _

establishing guidelines for conducting the reviews and plans to have

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the reviews performed by vendor personnel.

This portion of the program is scheduled for completion in 1991, however, FSAR reviews will continue as part of the FSD program and completion of this program is not scheduled until 1994.

The Q list program will generate a computerized document utilizing the existing component data base and will contain a list of all safety related equipment on a component basis.

This program is

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scheduled for completion in 1991, for all systems not included in the L

FSD program. The Q list will be updated for the systems included in the FSD program as the FSD for each system is completed.

i The purpose of the configuration management manual progam i s.to-document the existing processes which control the configuration of

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the plant and the transfer or communication of information among the different organizations involved in the processes.

The effectiveness of the process. and communication will also be evaluated for potential enhancements. This program is scheduled for completion in April 1991.

The purpose of the setpoint document program is to generate a single, centrally controlled document that identifies setpoints for the Farley Nuclear Plant.

Presently the setpoints are contained in several documents and drawings. The licensee is presently reviewing the -program for approval and implementaiton.

This program is scheduled for completion in 199 d?

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'The identification of documents to be revised or voided for Design

Changes involves two activities, the voiding of identified outdated drawings and the evaluation of the existing process for identifying drawings to be revised or voided as a result of design changes. The

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l licensee has already identified and voided approximately _450 functional control diagrams that had not been maintained current but

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still existed as controlled drawings in the document control system.

The licensee is presently in the process of evaluating _ the

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L existing process.

This program is presently scheduled to be

completed by the end of 1990.

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The purpose of the Maintenance Work Request trending program is to develop and -implement a computerized maintenance work request trending program to-identify the root causes or repetitive f ailures or problems.

The licensee is presently in the process of obtaining contractor support so that the program can be developed and

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implemented.

This program is scheduled to be completed in 1992.

The outbuilding cleanliness program consists of a one time clean-up of all outbuildings to improve the material condition of the buildings and the equipment located in them and to increase management's attention to the material condition of equipment. The one time' clean-up effort has been completed and increased management attention is-ongoing.

The purpose of the _ attendant equipment list program is to clearly define and document the equipment which must perform a specified function in order for the supported equipment or system to be considered operable, as defined in the Technical-Specifications.

The licensee is presently evaluating the draft program description and plans to initiate a pilot program sometime in 1990.

The purpose of the PM criteria program is to clearly define and document the criteria for the PM program, establish guidance for

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application of the criteria in each maintenance area and to evaluate the existing PCN process for identifying PM tasks. The licensee is presently in the process of establishing this program.

The configuration management project appears comprehensive in scope and the programs appear to be on schedule as far as completion dates, however, the programs are in the early stages of implementation and development.

3.

Exit Interview The inspection scope and results were summarized on April 27, 1990, and further discussed by telephone on May 14, 1990, with those persons indicated in paragraph 1.

The inspectors described the areas inspected

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and discussed in detail the inspection results. Although reviewed during this inspection, proprietary information is not contained in this report.

Dissenting comments were not received from the licensee.

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The following IFIs' were identified during this inspections:

IFI 50-348,364/90-03-01, DC Load Profiles,-paragraph 2 g.-(1)

IFI 50-348,364/90-03-02, EDG Loading Conditions, paragraph 2 g. (2)

IFI 50-348,364/90-03-03, Clarification of EDG Loading Restriction,

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paragraph 2 g. (2)

IFI 50-348,364/90-03-04, Revised AC Load Analysis, paragraph 2 g. (3)

4.

Acronyms and Initialisms l-APCo Alabama Power Company l-BIT Boron Injection Tank CFR Code of Federal Regulations CVCS Chemical and Volume Control System

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DARC Design Adequacy Review Committee i

EDG Emergency Diesel Generator EPRI Electric Power Research Institute

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ES Engineering Study FNP Farley Nuclear Plant FNPIMS Farley Nuclear Plant Information Management System FSAR Final Safety Analysis Report FSD Functional System Description GMNS General Manager Nuclear Support I&C Instrumentation and Control IFI Inspector Follow-up Item IN Information Notice INPO Institute of Nuclear Power Operation LOCA Loss of Coolant Accident LOSP Loss of Offsite Power MCC Motor Control Center MOD Minor Departure from Design MOV Motor Operated Valve MSIV Main Steam Isolation Valve NEL Nuclear Engineering and Licensing p

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NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission NS Nuclear Support-NSAC Nuclear Safety Analysis Center NUMARC Nuclear Management and Resources Council 0&M Operations and Maintenance OEM Original Equipment Manufacturer PCN Production Change Notice PCR Production Change Request PISC Process Instrumentation Setpoint Checklist PM Preventative Maintenance QA Quality Assurance SAER Safety Audit and Engineering Review-

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Southern Company Services

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SECL Safety Evaluation Check List SR0 Senior Reactor Operator SSSA.

Self-Initiated Safety System Assessment

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Technical Specifications i-VPN Vice President Nuclear WOG'

Westinghouse Owners Group:

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