ML20057B818

Insp Rept 50-461/93-10 on 930706-0824.Violations Noted.Major Areas Inspected:Engineering & Technical Support & Mgt Activities

ML20057B818
Person / Time
Site:	Clinton
Issue date:	09/16/1993
From:	Oneal D, Pegg W, Shafer W, Walker H, Yin I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20057B807	List: ML20057B806 ML20057B816 ML20057B818
References
50-461-93-10, NUDOCS 9309240037
Download: ML20057B818 (27)

See also: IR 05000461/1993010

Text

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U. S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-461/93010(DRS)

Docket No. 50-461 License No. NPF-62

Licensee: Illinois Power Company

500 South 27th Street

Decatur, IL 62525

facility Name: Clinton Power Station

Inspection At: Clinton Power Station, Clinton, IL

Inspection Conducted: July 6 throuah August 24, 1993

Inspectors: , [

H. A. Walker, Team Leader

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D. M. O'Neal

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R. D. Pegg W Date

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'I. T. Yin F Date

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NRC Consultants: C. J. Crane, Main Line Engineering Associates

M. Shlyamberg, Parameter, Inc.

Approved By: OMp /h f//4!M

'W. D. Shafef, Chief Date '

Maintenance and Outages Section

Insoection Summary

Inspection conducted July 6 throuch Auoust 24. 1993 (Report No. 50-

461/93010(DRS))

Areas Inspected: Special announced team inspection of engineering and

technical support and related management activities. The inspection was

conducted utilizing portions of inspection procedures 37700, 92701, and 92720

to ascertain whether engineering and technical support was effectively

accomplished and assessed by the licensee.

Results: Based on the items inspected, overall performance in engineering and

technical support was considered good. One violation with two examples of

inadequate corrective action and two inspection followup items were

identified.

9309240037 930917

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Inspection Summary 2  ;

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The most significant strengths relating to engineering support were management  !

commitment'to good engineering support, dedicated and experienced engineering -!

and plant management personnel, and good communication and coordination within  !

engineering and with other plant organizations.  ;

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The most significant weaknesses were the failure to properly identify and

correct causes of significant equipment failures, the failure to determine or .

provide objective evidence of system operability determination, lack of l

attention to detail in performing calculations and 10 CFR 50.59 safety  :

evaluations, and the misapplication of the alternate component procedure i

rather than using the design change process.

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DETAILS

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1.0 Principal Persons Contacted

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Illinois Power Company

• L. Haab, Chief Executive Officer

• C. Wells, Executive Vice President l

• J. Perry, Senior Vice President

o S. Clary, Supervisor, Procurement / Materials Engineering i

J. Cook, Vice President ,

• K. Graf, Director, Engineering Projects

• D. Korneman, Director, System and Reliability Engineering l

• oJ. Langley, Director, Design and Analysis Engineering

• J. Miller, Manager, Nuclear Station Engineering Department 1

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• oR. Phares, Director - Licensing l

0 J. Sipek, Supervisor, Regional Regulatory Interface i

• D. Waddell, Director, Programs and Administration l

• P. Yocum, Director, Operations  !

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U. S. Nuclear Reculatory Commission (U.S. NRC) i

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• P. Brochman, Senior Resident Inspector 1

F. Brush, Resident inspector  !

o B. Burgess, Chief, Operational Programs Section  :

o R. Hague, Chief, Projects Section 3B i

• F. Jablonski, Chief, Maintenance and Outage j

o G. Wright, Chief, Engineering Branch

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• Denotes those present at the exit meeting on August 3, 1993.  ;

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ODenotes those present at the supplemental meeting at the Region III office

in Glen Ellyn, Illinois, on August 24, 1993.

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Other persons were contacted as a matter of course during the inspection.

2.0 Licensee Action on Previous inspection Findinas

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A number of problems or concerns identified in past NRC inspections were

reviewed for appropriate licensee corrective actions. The items reviewed and i

the inspector's evaluations of the actions to address these issues are  !

discussed in this section,

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2.1 (Closed) Unresolved item (461/90024-Ol(DRS)1: Concerns regarding

the acceptability of engineering analysis of improper Raychem splices. The  ;

inspector reviewed the licensee's analysis and evaluation of the improper use i

of Raychem splices inside containment. The inspector also reviewed an '

evaluation of this issue by the NRC Office of Nuclear Reactor Regulation, I

which concurred in the licensee evaluation. The inspector established,

through discussions with licensee personnel and review of internal memorandum

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-Y-215619, dated July 23, 1993, that the continuity of the circuits, through l

the questionable Raychem splices, would be periodically verified by ,

surveillance or operability tests unless continued circuit continuity could be  !

verified by the use of annunciators or other means. The inspector has no '

further concerns on this issue and this item is closed.  !

2.2 (Closed) Unresolved Item (461/90028-02(DRS)): Concerns regarding

the acceptability of Raychem splices located inside containment in containment

penetrations and junction boxes, the basis for only inspecting Raychem splices ,

installed during P0-3 and RF-2 and the acceptability of splices involving the  :

improper use of AMP 34318 barrel splices. The licensee's analysis and

evaluation of the improper use of Raychem splices inside containment included

the use of oversized AMP 34318 splices and Raychem splices located at j

containment penetrations and inside junction boxes. .The inspector reviewed

licensee documents describing the action taken to resolve the issue as well as j

an evaluation of this issue by the NRC Office of Nuclear Reactor Regulation, j

which concurred in the licensee evaluation. In addition, the inspector ,

reviewed internai memorandum Y-215620, dated July 23, 1993, which described  !

the basis for only inspecting Raychem splices installed during P0-3 and RF-2.  !

The inspector has no further concerns on this issue and this item is closed. l

3.0 Inspection Ob.iectives

The objectives of the inspection were to determine if engineering activities '

that supported the Clinton Power Station (CPS) were properly coordinated and

effectively controlled and implemented. The inspectors focused on design j

changes and modifications, internal assessments of engineering, and actions

taken for the identification and resolution of technical issues and problems.

This was accomplished by observations of work activities, interviews with  !

selected personnel (including engineers and engineering management), and

reviews of records, procedures, and associated documentation.  !

3.1 Performance Data and System Selection  ;

The selection of systems and components for emphasis during this inspection

was based on a review of data from licensee event reports (LERs), latest SALP

information, and discussions with cognizant NRC personnel. The systems  ;

selected were specific electrical, mechanical, and instrumentation components j

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of the reactor core isolation cooling system (RCIC) and the control room *

heating, ventilation, and air conditioning (CVS) system. In addition,

activities and documentation involving other systems and components were

reviewed during the inspection, with consideration given to the systems  !

considered most safety significant. Substantial attention was given to the

residual heat removal (RHR) system, after possible problems were noted in this

area. )

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3.2 Observations of Plant Conditions

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On July 8,1993, the inspection team performed a walkdown of the RCIC and the

CVS systems to observe the material condition, adequacy of design for

maintenance and other supporting activities, indications of equipment

problems, and unusual conditions. A walkdown was also performed on the

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accessible portions of the RHR system to review pipe support configuration. .

The material condition nf other areas of the plant were routinely observed '

during tha performance of the inspection.  !

Other than conditions noted in Section 3.2.3, the inspectors concluded that

the mtterial condition was good with no unusual conditions noted. I

Accessibility for maintenance and other required activities was considered e

excellent for the CVS area and acceptable for the other areas. Details on the i

specific systems walkdowns follow. >

3.2.1 Reactor Core Isolation Coolina System

The inspectors walked down accessible portions of the RCIC system to assess

the material condition of the equipment and of the environment. The material

condition of the RCIC system was good and the accessibility of major system

components for maintenance was acceptable with limited access for some

components. There were no significant detrimental conditions noted and  ;

components that were in need of repair were properly identified.

3.2.2 Control Room Ventilation System

The Control Room Ventilation system (CVS) design exhibited excellent design in

the areas of separation and layout. The system was exclusively dedicated to

the control room complex and the system layout was excellent with ample room

for maintenance of all major components. The mechanical room area was well

lit and clean. There were no significant detrimental conditions noted and

components that were in need of repair were properly identified.

3.2.3 Residual Heat Removal System

The RHR system layout was not reviewed for maintenance accessibility. During

the walkdown, the team observed and investigated several instances where rigid

start type pipe supports were located so close to snubbers that the function -

of the snubbers would be affected. The team's discussions with licensee

personnel and a review of records in this area are documented in Paragraph

3.4.2 of this report.

3.3 Enaineerina and Technical Support

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Engineering and technical support at the CPS was provided by the Nuclear

Station Engineering Department (NSED). NSED was divided into four separate

organizations; design changes and modification responsibilities were assigned

to the Design and Analysis Engineering Group. Most of the other engineering

support to plant organizations was provided by the Systems Engineering Group.

Special projects and support functions were provided by the Engineering

Projects and Reliability Engineering Group and the Programs and Administration

Group.

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3.3.1 Systems Enaineerina Support

The Systems Engineering Group consisted of three system engineering sections. -

Systems engineers provided oversight of assigned' systems and focused on daily

operations and maintenance activities on the systems. These engineers aided  ;

plant operations and maintenance personnel in resolving technical issues and l

problems and were involved in complex maintenance evolutions involving the i

assigned systems. They also coordinated potential design changes with other  !

engineering organizations.

Based on the results of interviews and other inspection-activities, the

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inspectors determined that, in most cases, the system engineering staff was

competent and qualified. They exhibited a good sense of system ownership and

. communication and coordination with plant management, operations, maintenance,

and C&I were effective.

3.3.2 Enaineerina Support for Preventive and Predictive Maintenance

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The involvement and performance of the engineering staff in support of the

predictive and preventive maintenance (PM) program was good. Improvements  !

continued to be made in this area. A liaison had been recently established to  !

facilitate communications between engineering, operations, and maintenance in

this area. These changes appeared to have improved the quality of

communications and support of maintenance.

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3.3.2.1 Support for Preventive Maintenance

The team reviewed approximately 30 PM task deferral requests from 1993. All

deferrals were properly justified by technical staff raembers cognizant of the

activity and equipment affected. Only one of these PMs, inspection of the

dehydrator of a drywell chiller, had been delayed previously. The PM was

delayed because the PM required entry into a condition that was restrained by

a maintenance work request (MWR), which was awaiting parts. The PM was to be

worked in conjunction with the maintenance.

3.3.2.2 Support for Predictive Maintenance

Extensive use was made of thermography, acoustic monitoring, oil sample

analysis, vibration analysis, and other trending / diagnostic methods. These

methods were aggressively and effectively used to trend adverse equipment

conditions. The data were analyzed by system engineers and reliability i

engineering, and changes were made to correct the adverse material condition  ;

or to correct the PM program, as necessary. The application of the results  !

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from this testing resulted in increased equipment reliability and more  !

effective use of limited resources.

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Examples of recently identified condition monitoring trends included the use

of vibration analysis to diagnose pump coupling degradation, use of check

valve monitoring to discover the RHR "C" pump discharge valve was not fully

closing, and use of thermography to identify elevated temperatures of one of

the battery jumper cables for the horizontal fire pump. The latter example

was discovered during performance of a PM task created as part of corrective

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action for a CR initiated after the batteries on "B" fire pump caught fire due ,

to a hot terminal on a battery connection. These examples were typical of the '

effective application of diagnostic tools and analysis of the data to the PM -

. program.

3.3.3 Review and Evaluation of NRC and Industry information

The inspectors evaluated the effectiveness of the licensee's method for review

and evaluation of NRC and industry information. This review included the l

methods used to assure that vendor, industry, and NRC generic information was  ;

controlled, distributed, and evaluated and that appropriate corrective actions

were taken.

The Licensing Department had the overall responsibility for coordination of .

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review and evaluation of NRC and industry information. Upon receipt of this-

information an initial screening for applicability was performed. According

to the information provided by the licensee, only 40 to 50 percent of the i

incoming documents qualified as applicable to this facility. All documents l

were logged and tracked in accordance with the procedure L.1, " Operating l

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Experience Program". i

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After the documents were logged, they were forwarded to various departments i

responsible for the actual evaluation and implementation. Approximately 80

percent of all applicable documents were handled by the NSED. All departments

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, were required to provide a response to the Licensing Department with a

determin: tion of impact within 30 days. All of the applicable NRC and

industry information was tracked by the Licensing Department until the issue

was closed. The Licensing Department was also responsible for assembly of the

response package and preparation of the cover letter. In addition to the i

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tracking performed by the Licensing Department, all of the documents handled '

by the NSED were assigned an engineering work request (EWR) number and were

tracked as EWRs by the NSED.

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i One incident was noted where inadequate action was taken on a Part 21 j

notification on diesel generator air start solenoid valves. CR 92-08-031 l

documented a case where an NSED engineer, responding to a 10 CFR 50, Part 21 l

' notification, wrote an analysis and closed the item, without proper action

being taken. This item is discussed in more detail in Section 4.3.2 of this '

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3.4 Desian Chances and Modifications

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Nuclear project engineering had the primary responsibility for coordination,

evaluation, development, and installation of design changes and modifications.

This group consisted of engineers with mechanical, electrical and other

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engineering specialties. The primary purpose of the group was to coordinate

plant modifications, including design, safety reviews, installation, and post

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modification testing. The modification work was handled by project engineers,

who were assigned specific modifications. These engineers were actively

involved in all phases of assigned modifications including . installation. The

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engineers completed walkdowns, as necessary, to ensure proper design and

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resolution of design installation problems.

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Based on the results of interviews and other inspection activities, the

inspectors determined that the project engineers were knowledgeable of the

assigned modifications, and were competent and qualified. Communication and

coordination were effective between the project engineers, the system

engineers, plant management, operations, maintenance, and I&C.

3.4.1 Review of Modification Packaoes and Records

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The inspectors reviewed both open and closed modification packages to verify

that the packages were complete and accurate. The review included

verification that the description of the modification, the 10 CFR 50 59 safety

evaluation, installation instructions, documentation of work performed, post-

modification testing requirements and test records were adequate. In some

cases, other supporting records associated with the modifications, such as

calculations, were selected and reviewed to verify the adequacy and accuracy

of the engineering process.

Based on the review of the selected records, the inspectors concluded that the

modification process was effective. The team had no adverse comments

concerning the review of most of the modification packages; however, problems

were noted with a few modifications and supporting records. Some supporting

calculations and 50.59 safety evaluations or screenings were considered weak

and showed a lack of attention to detail. A variety of problems were found

during the review of records for a few modifications. In addition, several

problems were noted with respect to electrical cable fill in cable trays.

Examples noted in these areas, however, did not appear to have a significant

safety or operational impact on the plant.

With the exception of the noted deficiencies, the modification records

reviewed were adequately controlled and were consistent with regulatory

requirements.

3.4.1.1 Review of Goen Modification Packaoes

The eleven open modification packages reviewed during the inspection were:

o DC F004 -- This modification required that the Division I, Class lE

battery be replaced with a larger capacity battery. The modification

had been installed and the system had been released to operations;

however, final package review had not been completed and the package had

not been closed.

In addition to reviewing the records in the modification package, the

inspectors also reviewed calculations associated with the modification

to verify the accuracy of the engineering process. The electrical

analyses for the modification consisted of evaluation of voltage drop,

short circuit current, under voltage relay settings and breaker

coordination. The team noted several problems with the associated

calculations. Details on the calculation review are discussed in

Section 3.4.1 6 of this report.

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o RH-040 -- Increase the overall gear ratio for RHR system containment

isolation injection valves. The design was complete for this i

modification, however, the installation had not been started.

The safety evaluation had not been researched thoroughly. A valve  !

opening stroke time of 40 seconds was quoted from a table in the USAR,

yet elsewhere in the USAR and in the Technical Specifications, a value  !

of 37 seconds was given. This limit was based on the required Low ,

Pressure Coolant injection System response time of less than or equal to

37 seconds. The impact of this weak review was minimal since the

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operations department documented the discrepancy in a Detailed Impact

Assessment Form, which was completed as part of the modification

process.

o RI F007 -- This modification required the installation of a vent line

between the RCIC condensing pot level switch and valve IE51-F324A to  :

allow venting for calibration of the RCIC steam inlet drain pot level

switch IE51-N010. The modification had been installed and the system +

had been released to operations; however. final package review had not  !

been completed and the package had not been closed. i

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O RI F010 -- This modification required that flange connected pipe on the i

RCIC turbine steam exhaust line piping leading to the suppression pool -i

downcomers be replaced with welded pipe. The modification had been

installed and the system had been released to operations; however, final

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package review had not been completed and the package had not been  ;

closed. i

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o RI F0ll -- This modification required the installation of

environmentally qualified in-line quick disconnect electrical connectors

in class lE circuits located inside the drywell. This modification had I

been installed and the system had been released to operations; however, {

final package review had not been completed and the package had not been I

clesed.

During the review, the following observations were noted:

(1) The modification required that the electrical connectors be

environmentally qualified. The modification also required that a

connector cap be installed on the electrical connector prior to

submerging the connector during refueling.

The environmental qualification (EQ) evaluation was not performed

for the acceptability of submerging the connector in the de-mated

condition during refueling. We also found that_ documentation was

not requested or. received from the vendor (EGS/Patel) relative. to

the water tightness capability of the connector cap. Our concern

was that the connector cap could leak when submerged, thereby

admitting water into the connector possibly resulting in

degradation of the circuit at a later date af ter startup.

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(2) The Certificate of Conformance, provided by the supplier. referred

to EQ Test Report EGS-TR e13602-01, Revision A; however, licensee

personnel evaluated Revision 0 of this report; Revision A was not

reviewed and was not available onsite.

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Licensee personnel stated that the cap used to protect the disconnected

male portion of the connector during refueling had leaked under

submergence. The connector was subsequently dried out in accordance  :

with manufacturer's instructions under MWR 023097, verified to be ,

operational, and returned to service. The inspector was also told that ,

future occurrences of this condition would be prevented because covers

to provide complete submergence seals would be obtained and used.

Additionally, the licensee provided the necessary submergence evaluation .

with respect to maintaining equipment qualification by revising the l

impact assessment and the EQ review checksheet. A complete EQ package

had not been prepared for the installed electrical connectors as of the ,

date of this inspection and the connectors had not been added to the EQ

master list. Licensee personnel stated that the EQ package was ,

scheduled for preparation on the EQ Impact Assessment computer log  ;

tracking system. This is an inspection followup item pending completion i

of the submergence connector cap procurement, completion of the full EQ

package, and addition of the connectors to the master EQ list i

(461/93010-Ol(DRS)). 'i

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O RT 030 -- This modification required that the size of the RWCU warm-up

line be increased from one to two inches. The design was complete for

this modification and the installation had been authorized. The  :

modification was scheduled to be installed during RF-4, which was

expected to start in October 1993.  !

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o RT 034 -- This modification required that a six inch pipe with blind

flanges on both ends be installed through spare containment penetration  !

IMC-74 for use during reactor water cleanup decontamination activities.

The design was complete for this modification and the installation had  ;

been authorized. The modification was scheduled to be installed during

RF-4, which was expected to start in October 1993.

- o VCF 013 -- This modification "su ir. that the control room HVAC system

chlorine detectors be discon .ted .o abandoned in place. The design

for the modification was com '

ts ,dt the modification had not been

installed.

o VCF 019 -- This modification required that a high temperature alarm be

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installed on VC control panels to annunciate in the control room in case

of a steam line break and subsequent temperature rise in the area. This

modification had been installed and the system had been released to

operations; however, final package review had not been completed and the

package had not been closed.

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VCF 021 -- This modification was administrative in nature and revised

the part number on design documents for a HVAC Westronics recorder used

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to replace the originally installed control room HVAC Westronics

recorder. The replacement recorder had been purchased non-safety

related and was dedicated for safety related use. This modification had

been installed and the system had been released to operations; however, J

final package review had not been completed and the package had not been

closed.  ;

o WS 020 -- This-modification required the replacement of four existing

air operated temperature regulating valves in the plant service water

system. Design was complete but the modification had not been ,

installed. No schedule for ' installation of the modification was '

provided.

3.4.1.2 Review of Closed Modification Packages

The inspectors reviewed seven closed modification packages during the  :

inspection. The results of this review are presented below.

o A 147 -- Increase the height of the existing splash / spray guard

surrounding the Hydraulic Power Unit (HPU) skids to better contain  ;

future leakage of the Fyrquel EHC fluid. i

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0 HP 029 -- Replace valve actuator spring pack. The impact on the opening

or closing times was not addressed.

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o MS 041 -- Replace the original spring packs and torque limiter plates. '

o RE F002 -- Change the setpoints for various sump level instruments.

This modification did not provide an adequate description, safety

significance or justification of the proposed changes.

o RH F015 - Lower the tailpipes for containment penetrations MC-87, 20,

and 18 further into the suppression pool.

O SC F003 -- Change the setpoints for SLC te.nk level instruments. This

modification did not address the safety significance of the setpoint

accuracy on the SLC tank level. The impact system function is not

addressed.

o SX F013 -- Reroute line ISX36AB (shutdown service water return line from

ECCS RHR heat exchanger Room IB coil cabinet IVY 055) downstream of

orifice ISX12MB.

3.4.1.3 Review of Minor Modification Packaaes

Licensee personnel stated that a new method was implemented in March of 1993

to expedite the- review and approval of minor modifications. The inspectors

reviewed Procedure 0.54, " Plant and Engineering Changes," Revision 0, to

ensure that the process provided adequate controls and that regulatory

requirements were met.

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The procedure appeared to provide a very good method-for control of minor

modifications, four minor modification-packages were reviewed to evaluate the

control and effective implementation of the minor modification process.

Although there was a possible misapplication of the procedure noted and a lack

of attention to details was evident in some packages, the control of minor

modifications was considered to be acceptable.

The results of this review are summarized below.

O ECN 27869 -- Add a 90 degree elbow and a 3 inch nipple between the air

filter and the diesel generator air start compressor.  ;

A bill of material was not provided and the size, material and-other

requirements for the elbow and nipple were not specified. In addition, i

installation details such as the method of installation (i.e. threaded,

welded, etc.) were not provided.

The lack of details on this modification was discussed with licensee t

personnel. Since the effects of the modification did not appear to be

significant, no further action was considered necessary,

o ECN 27929 -- Modify MOV MO-518.

O ECN 27975 -- Replace the pinion gear and shaft gear on the RHR loop C

suction valve.  ;

o ECN 27995 -- This ECN authorized the machining of a reactor water

cleanup pump shaft.

This modification appeared to a possible misapplication of the minor

modificatien procedure. The design basis for sealing the pump shaft t

against leakage was based upon the use of a gasket and mechanical seal. '

Addition of the 0-Rings might constitute a change to this basis.

There was no evidence that the impact of machining groves on the RWCU

pump shaft was evaluated. No calculations or other information was  :

provided to demonstrate that the machining of the grooves would not l

affect the integrity of the pump shaft.

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The addition of the 0-rings affected pump maintenance, yet the  :

corresponding procedure was not identified as "affected".

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During discussions with licensee personnel on this matter, the ,

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inspectors were told that the modification to the pump shaft was

discussed and coordinated with the pump supplier who provided assurance l

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that there was no problem with the modified pump shaft and operation of

the pump would improve.

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o ECN 28006 -- This ECN authorized modification of MOV M0-401.  !

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3.4.1.4 Technical Evaluation for Replacement Items .

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During the review of modifications the inspectors noted that, in some cases,

the Technical Evaluation for Replacement -Items (1ERI) program was used to i

replace, supplement or support the modification process. Procedure D.59, '

" Alternate Part Control," Revision 1, and CPS instruction DE-14, " Technical '

Evaluation for Replacement items (TERI)," Revision 0, describe this program. t

The procedures were reviewed and the inspectors selected 13 TERI packages to i

review for adequacy and proper application. The results of this review are  ;

summarized below.

Most of the packages reviewed were acceptable and all but one package, SX-

P002, adequately identified differences in physical attributes. The assigned

TERI numbers were not prominently located on the packages and one package, M-

P013, based tne determination of " safety" on the frequency of component usage. l

Most of the seven valve packages failed to identify " seizure" as a failure

made and most of these packages did not identify the three functions, pressure

boundary, allowed flow and isolate flow. In at least one case, important

physical attributes were not addressed.

The inspectors noted that, in three cases, the TERI program appeared to be 3

applied to items beyond the scope of the program as described in DE-14. This '

paragraph described the scope of the TERI program as ". . . when the original  ;

part has been discontinued or is no longer available, has had a change in

manufacturing processes / designs / materials or when an alternate sub-tier

supplier is requested."

o CM-P001 -- This TERI evaluation authorized the use of a flywheel, for i

the containment monitoring compressors, made of ASTM A-576, Gr.1026, E

rather than ASTM A-576, Gr.1018, which was specified on the design  ;

documents.

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The flywheel was inadvertently fabricated from the wrong material when  ;

parts were made "inhousa" from stock identified as ASTM A-576, Gr.1018.  ;

During this review, there was no evidence that the problem with material

control was addressed or corrected. This indicated a weakness in the .

corrective action process which is discussed further in Section 4.3.2 of I

this report.

O

FP-P005 -- This TERI evaluation authorized t.he use of a solenoid '

operated starter to replace the centrifugal starter on diesel driven

fire pump "A". The starter configuration was different; the solenoid i

starter required an electrical wiring change which increased the  ;

involved wiring from two to five wires. i

o- M-P012 -- This TERI evaluation authorized the use of compression brass

t

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fittings on the plant chillers instead of flared brass fittings. ,

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Based on the review of these packages, the inspectors were concerned that the i

three packages did not meet the described scope of the TERI program and ~

actually appeared to be modifications. Safety screenings were performed on '

these items and the evaluations required by the TERI program appeared to '

ensure adequate control to ensure acceptability of the components. NRC  :

concerns in this area were discussed with licensee personnel and a meeting was ,

held on August 24, 1993, to discuss the issue. Licensee personnel presented

information clarifying the intent and scope of the TERI program. Personnel  ;

also stated that, during preliminary TERI reviews, emphasis would be placed on

ensuring the scope of the TERI program would be properly applied and that TERI

procedures would be reviewed to determine if the description for the program

application needed to be clarified in the procedure. This additional

information resolved NRC concerns in this area.

,

I

The TERI program appeared to provide a good method for expediting the review

and approval of alternate parts and the evaluation questions included in the .

procedures were very good. As indicated above, however, additional attention

i to detail appeared to be lacking in the implementation.

,

4

3.4.1.5 Review of Safety Evaluations and Screeninas

During the review of various design change packages, the inspectors reviewed .

'

the 10CFR50.59 safety evaluations or screenings included in the packages to

assess the adequacy of these evaluations. The safety evaluations were

reviewed for completeness, accuracy, and compliance with regulatory I

requirements. l

l

The observations for the safety evaluation / screening reviews follow.

)

o All of the packages reviewed, contained the 10CFR50.59 safety i

evaluations or screening and some of the packages contained supporting '

l information. Packages reviewed included modification, minor

j modification, TERI and temporary modification packages.

o Justifications for some of the answers in some of the safety evaluations

were inappropriate. For example, justifications for some answers in

safety evaluations for modification of some safety related equipment

components were, "the function of this component is not discussed in the

USAR."

o Many safety evaluations demonstrated a lack of attention to detail.

Incomplete answers were provided or the answers did not demonstrate an

understanding of the safety significance of the issues. This was

especially true for minor modifications and TERis. For example, the

safety evaluation for modification SCF003 addressed only the localized

,

impact of the actual changes and did not address overall plant safety.

o in some cases, all aspects of the change were not addressed in the

safety evaluations. For example, some of the modification packages

reviewed indicated that electrical cable trays would be filled above the

levels specified in the USAR. A separate safety evaluation was not

performed for the cable tray fill and there were no indications that the

issue was addressed in the modification safety evaluation.

)

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3.4.1.6 Review of Calculations

In order to complete the assessment of the design change and modification

process, the inspectors reviewed calculations required to be performed or

revised by the selected modifications. Calculations were reviewed for

completeness, depth and accuracy with emphasis on how well the calculations

supported the modification. Since the majority of the reviewed modification

packages did not require supporting calculations, several additional selected

calculations, not related to the reviewed packages, were evaluated.

Observations noted during these reviews are provided below.

o Calculation 19-D-10 was not revised as required, by modification package

DC-F004, to incorporate new replacement battery data.

o Calculation 19-D-14 contained an unconfirmed design input. The value

for the battery internal resist & ace was based on a S&L telephone

memorandum.

O Calculation 19-D-14 contained several calculation errors.

(1) Vendor information dated May 30, 1985, was used rather than the

new replacement battery information dated November 14, 1991. This

resulted in approximately a 5 percent error in the final result;

however, the acceptability of the batteries was not affected.

(2) The resistance of the 3/C #4/0 cables between battery 1A and the

MCC was incorrect.

(3) The cable lengths used in the calculation for the cables between

battery 1A and the MCC were incorrect.

O Calculation 19-D-14 used a non-conservative value for determining short

circuit current. A voltage of 119.5 VDC was used rather than the

maximum voltage of 130.5 VDC.

o Calculation 19-D-28 used a non-conservative value for determining the i

voltage at electrical equipment. The nominal voltage of '25 VDC was  !

used for the calculation rather than the fully charged battery "minirum"

voltage of 119.5 VDC.

1

o Calculation 19-D-28 contained an unsupported assumption. The

calculation assumed that the 125VDC Division I diesel generator  ;

auxiliary pump motors would operate satisfactorily at 87.5 VDC; however, l

there was no supporting documentation for this assumption.

o Calculation 19-D-33 contained a calculation error. The calculation

determined the continuous load current for the battery charger to be 160

' amps. The inspectors noted that calculation 19-D-28 and the battery

discharge characteristic curves indicated the continuous load to be 145

amps. This difference was not identified by licensee personnel.

13

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o Calculation 19-D-33 contained a note, " superseded by calculation'19-D-

10." Calculation 19-D-10 had not been revised to incorporate the new *

replacement battery data.

o Calculation 19-G-01 contained a calculation error. The calculation

indicated that the ampacities were based on Table 3-16, and Section 2.3

of IPCEA P-54-440 and were determined to be 133 amps. Based on IPCEA P-

54-440, the team found that the calculated value for ampacity for these

cables should be 114 amps. There was no documented justification in the

calculation to support the ampacity values.

l

0 _ Calculation IP-C-0005, Revision 01, did not address the possible impact 3

of the instrument tolerance on the setpoint. Licensee personnel

obtained information from GE during the inspection indicating that the  ;

setpoint nominal values included instrument inaccuracy so this problem

was of minor significance.

O Calculation IP-M-0076, Revision 2, failed to address the limitations of l

the results. The test results were valid for "as tested" configuration .

only. Changes to hardware could impact relative initiation of valve  ;

opening vs. pump start and invalidate the results. Subsequent review of

the test results and the calculation demonstrated tha+ an adequate  ;

margin existed to compensate for this uncertainty.  ;

o Calculation IP-M-0076, Revision 2, and other motor operated valve

^

calculations did not appear to consider abnormal events to determine '

maximum differential pressures as required by f4RC Generic Letter 89-10.

In Paragraph 2.1.1 of CPS Nuclear Engineering Standard ME-03.00, ,

Revision 3, emergency operating procedures (EOPs) were explicitly cited i

i as a reference source in the determination of maximum operating '

pressures. None of the reviewed calculations provided documented

evidence that the E0Ps were used in the determination of maximum

operating pressures. This matter is considered an inspection followup

item and will be reviewed during a subsequent inspection (461\93010-

02(DRS)). j

Based on the review of calculations, the inspectors concluded that some i

'

calculations were inaccurate and lacked attention to detail. It was evident  !

that improvements were needed in the preparation, checking and control of -  !

calculations. In some cases, electrical calculations, especially those

related to modification DC-F004, were weak, and at times were inaccurate and

non-conservative. Several calculation errors were noted and, in some cases, t

non-conservative inputs, unconfirmed design inputs and unsupported assumptions  !

were used.  ;

'

.

Many calculations, however, were thorough and exhibited a good understanding

of the problem. Other calculations appeared to be extremely detailed and ,

accurate. The DG HX mechanical calculations were of high quality and i

exhibited good engineering practices. r

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Although most of the individual problems in this area were not considered to  ;

have a significant effect on equipment function, considerable improvement was l

needed in this area. This matter was thoroughly discussed with licensee  ;

personnel.

3.4.2 $quhber and Ricid Restraint Intera_qtion

.

During the walkdown of the RHR Loop A and loop B pump suction piping, the

inspectors noted that three rigid restraints were located in close proximity i

to snubbers and were oriented in the same restraining directions. The '

inspectors questioned as to whether or not these snubbers would actually  ;

lockup during a design basis seismic event. During the review of the piping t

stress report, the inspectors found that the three rigid restraints were

modeled as snubbers. The piping analysis simply assumed that, after changing i

the snubbers to rigid retti aints, the nearby snubbers would still lockup- as  !

required.

.

The inspectors discussed the matter with licensee engineering personnel, which

included some Sargent and Lundy Engineers (S&L) personnel. The inspectors

reviewed documents and records provided and found that there was neither

mathematical nor experimental data to substantiate the position that the i

affected snubbers would perform their intended function. The inspectors

concluded that this area did not present a safety problem since the calculated  :

RHR piping stresses in the areas of concern were very low and the design i

appeared to be overly conservative. Licensee personnel indicated that the

questionable areas would be reexamined to determine if some of the affected

snubbers should be removed, -

j

3.4.3 Temporary Modifications

The methods to control temporary modifications were described in CPS

administrative procedure No. 1014.03, " Temporary Modifications," Revision 16.

The scope of this procedure was well defined; the procedure was comprehensive

and provided clear directions for implementation. >

The inspectors reviewed a list of temporary modifications, provided during the ,

inspection and selected nine temporary modification packages for review to

verify proper control of temporary modifications. The packages reviewed are

listed below:

0 88-058 - Install temporary fans in turbine building

o 90-004 - Remove the spool piece for the RHR heat exchanger auxiliary '

steam connection and install a blind flange to isolate RI from RHR '

o 90-027 - Remove ISX173B " Loss Of Power" input to annunciator

P601-5065-2f

o 91-006 -- Install control room annunciator enable / disable switch

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o 91-016 -- Install mechanical jumper to bypass the off-gas system

hydrogen analyzer

o 91-037 -- Install condensate oxygen injection

o 92-029 -- Install a soft patch and canister around drain line 2WS69A

upstream of drain valve 2WS110 to stop the leakage of lake water into

the condensate booster basement

o 92-040 -- Install additional RTDs near a suspected hot spot in the

drywell to monitor temperature during plant operations

o 92-093 -- Replace fuel oil in the Division I diesel storage tank

The inspectors noted that the control of temporary modifications had been

noted as a problem area in the engineering and technical support inspection

conducted in 1992 (inspection report 461/92005(DRS)). The temporary

modifications, listed as problems in the inspection report, had been closed l

and the number of temporary modifications had been reduced substantially ,

during the past year. Quality Assurance (QA) surveillance records indicated i

that QA had monitored the temporary modification reduction activities several

times during the year.

l

No problems were noted in the review of temporary modification packages.

Based on the review and the positive licensee actions in this area, the  ?

-

inspectors considered the methods used by the licensee to control temporary

modifications to be thorough and provide adequate control. ,

4.0 Self-Assessment of Enaineerina Activities

Self-assessment of engineering activities at the CPS consisted of audits and

surveillances of plant modifications and engineering support as well as ,

several assessments of engineering and the modification process. Overall, the

various assessments covered the spectrum of the engineering modification

process. ,

4.1 Audits and Surveillances

The inspectors reviewed recent QA audit and surveillance records and

interviewed personnel to determine the effectiveness of the licensee's self ,

assessment of engineering activities. QA performed audits and surveillances  !

of plant engineering activities in order to evaluate engineering performance. ,

The scope of the QA audits and surveillances of engineering was considered

adequate. *

.

4.1.1 Quality Assurance Audits

Comprehensive audits of the engineering group were normally conducted yearly

with additio9cl audits of supplemental engineering activities conducted as

needed. Records of six QA audits of engineering or engineering related

activities were reviewed. These audits were:

16

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Q38-92-01 -- Audit of the Nuclear Station Engineering Department (NSED)

Q38-92-06 -- Audit of RF-3 Refueling outage activities. This audit included a [

review of post modification testing for fuel handling equipment modifications.

Q38-92-07 -- Audit of Maintenance. This audit included a review of post

modification testing for modifications and field alterations. 6

Q38-92-22 -- Audit of Post Modification Testing for Plant Modification and

Field Alterations. '

Q38-93-04 -- Audit of Maintenance. This audit included a review of post

modification testing for modifications and field alterations.

Q38-93-09 -- Audit of the Nuclear Station Engineering Department (NSED).

Observations made during the review were as follows:

o Audit Q38-92-01 -- This audit was conducted from January 20 to

January 31, 1992. A weakness in the calculation review process was

noted in the audit. This appeared to be consistent with the findings of

the NRC inspectors since calculation problems were noted during this

inspection. Calculations are discussed in Section 3.4.1.6 of this

report.

O Audit Q38-93-09 -- This audit was conducted from April 5 to

April 27,1993, and included a review of the entire engineering support

activities. This included a review of the implementation of the

recently issued minor change procedure 0.54, " Plant and Engineering

Changes". This procedure was developed to expedite minor design

changes. As a result of the D.54 review, CR 03-93-04-027 was issued

because the D.54 process was used to authorize design changes outside

the procedural limitations. This appeared to be consistent with the NRC

findings noted during the inspection. See Section 3.4.1.3 of this

report.

The QA audits reviewed were detailed and performance based and appeared to

adequately cover engineering activities.

4.1.2 Ouality Assurance Surveillances

QA surveillances were used to supplement audits in the assessment of

engineering performance. The inspectors reviet.ed eight QA surveillance

-reports conducted on engineering and engineering related activities. These

surveillance reports were:

o Q-15128 -- This surveillance was. conducted in October of 1991, to verify

that temporary modifications were controlled per plant procedures,

o Q-15129 -- This surveillance was conducted in October of 1991, to verify

the adequacy of installation of temporary modification 91-030.

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o Q-15158 -- This surveillance was conducted in November and December of

1991, to monitor the installation of modification MT029-A.

o Q-16028 -- This surveillance was conducted in June of 1992, to review

the status of the Temporary Modification Reduction Plan.

o Q-16032 -- This surveillance was conducted in June of 1992, to verify

compliance with the temporary modifications procedure,

o Q-16040 -- This surveillance was conducted in July of 1992, to review

CRs associated with improper partial release of modifications for

operation, i

.

o Q-16135 -- This surveillance was conducted in April of 1993, to verify

compliance with the temporary modifications procedure.

o Q-16154 -- This surveillance was conducted in June of 1993, to monitor ,

the installation of modification SAF009.

The inspectors noted that four of the eight QA surveillances reviewed '

(Q-15128, Q-16028, Q-16032 and Q-16135) were related to the control of

temporary modifications which had been identified as a problem in the ,

engineering and technical support inspection conducted in June of 1992 (report -

1. 461/92005 (DRS)). These surveillance reports indicated that QA was aware of i

the temporary modification problem and appeared to be providing good follow up ,

to ensure proper resolution of a known problem.  !

4.2 Special Assessments of Enaineerina -

The Engineering Assurance (EA) group was primarily responsible for special

assessments of engineering. The group was formed approximately two years ago ,

and was tasked to review the engineering and modification process. Feedback I

on general and specific observations concerning the process were provided to i

engineering management for resolution. During the past two years, the group  ;

performed three assessments of engineering, primarily in the area of

modi fications. In addition, two other recent assessments of engineering were  :

performed; one by an internal assessment team and the other by a contractor.  ;

The assessments by the three different organizations are discussed separately.

4.2.1 Enaineerina Assurance Assessments .

The three EA assessments performed by the EA organization were:

o Informal Assessment 92E -- Review of selected temporary modifications.

This assessment was completed in November of 1992.

,

o Assessment 92F -- Evaluation of the NSED design process. This

assessment was completed in December of 1992.

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o  !'

Informal Assessment 92G -- Review of selected uninstalled modifications.

This assessment was completed in January of 1993. t

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At the time of the inspection, EA was reviewing and evaluating selected i

modifications to be implemented during RF-4. Modifications selected for i

review, if improperly designed, could result in failure with consequences, up I

to and including, a reactor scram or forced shutdown.

4.2.2 Special Internal Assessment Group. l

An internal group was organized to assess the effectiveness of the engineering

department in' meeting expectations. This assessment was performed from 3

December 7 to December 18, 1992, by conducting extensive personnel interviews '

and reviewing engineering department organization and processes for  ;

accomplishing work. The assessment focused on evaluating management and plant  !

staff expectations of NSED, and.the performance of engineering personnel in  !

meeting those expectations.

Results of this assessment indicated that plant staff personnel had a high I

regard for the engineers they work with. They believed the engineers were

technically competent and were improving in meeting the expectations of the

plant. There were also indications that improvements were still needed and

,

that the design process needed to be simplified and streamlined.

4.2.3 Assessment bY Outside Contractor

An independent assessment of the engineering modification process was  !

completed, by an outside contractor, in January of 1993 for the period of  ;

January 1992 through November 1992. The assessment report indicated that, in  ;

some cases, engineering solutions to plant problems were of a temporary l

nature, rather than a permanent fix. The majority of engineering effort was l

also found to be spent on low impact projects.

The report also indicated that the modification program needed to be changed

so that minor changes to the plant would not undergo the rigors of the i

complete modification process. This recommendation apparently precipitated

the development of the minor modification procedure D.54, which was recently

incorporated into the CPS modification process.

4.3 Trendina and Corrective Action

The inspectors reviewed the methods used by engineering to trend equipment

problems, investigate equipment problems for cause and provide adequate

corrective action. Significant problems or failures were documented on

condition reports, which were used as a mechanism for investigation to

determine the root causes and initiate actions to prevent recurrence.

Management involvement in the tracking and resolution of deficiencies and

corrective actions was good. The cooperation exhibited by individuals and

groups involved in this process appeared to be effective in achieving timely

resolution of problems.

Trending and corrective action are discussed in the following sections  :

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4.3.1 Trendino

Reliability engineering performed MWR and PM failure trending through I

systematic reviews of corrective maintenance history using the Power Plant

Maintenance Planning Systems database. The trending and failure histories

were reported- quarterly in the Material Condition Management Program (MCMP)

trend reports by Reliability Engineering. In these reports, the equipment

condition trends and failure trends observed over the previous quarter were  :

reported and the status of corrective actions was provided.  ;

,

The material condition trending program was extensive and included vibration  !

analysis, oil analysis, thermography, check valve monitoring, EDG monitoring,

MOV trending, battery trending, and in-service inspection. The failure  ;

trending included equipment failure trending, C&I surveillance failure ~

trending and the Component Failure Analysis Report (CFAR). i

Specified equipment failure data for designated components were submitted to

the fluclear Plant Reliability Data System (NPRDS) on a regular basis. The

licensee's use of the NPRDS through CFARs resulted in the identification of  :

above average failure rates for some plant components. The failure rates were

addressed and appropriate corrective actions were taken.

!

During the review of the modification packages, the inspectors noted

repetitive failures for Westronics recorders used in the reactor recirculation '

system and the control room HVAC. Reliability Engineering was aware of these t

failures as well as other Westronics recorder failures and was tracking these

failures for the different model types.

The teams' assessment of the MCMP trend reports and the ongoing work between '

reliability engineering, systems engineering, other technical staff, and i

management determined that repetitive problems were effectively identified and  ;

the corrective actions were appropriate. The use of trending data provided

early detection of problem areas and allowed implementation of timely  !

corrective action. The inspectors concluded that the objectives of the

trending program were met.

1

4.3.2 Corrective Action I

1

Significant problems or failures were documented on CRs and evaluated by the i

Corrective Action Review Board (CARB) to determine if the problem was quality

related and if a root cause analysis was required, assigned responsibility for ,

resolution of the CR to the appropriate department, and determined if the

!

immediate actions taken corrected the identified condition or were tracked by l

an approved tracking mechanism (i.e., MWR). This process was described in CPS '

Procedure 1016.01, " CPS Condition Reports", Revision 24.

The inspectors reviewed numerous condition reports and concluded that, in most

cases, CRs were properly processed, evaluated for cause and actions taken were

appropriate and timely. In some cases, however, the cause was not determined

~

and the cause investigation and corrective action were inadequate. This

, problem is illustrated by the two violation examples as well as the other

examples that follow.

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The inspectors reviewed 33 CRs to evaluate cause investigation and corrective

action. Discussion of the concerns in this area follows:

o 1-91-01-029 and 1-91-01-069

CR l-91-01-029 documented seven rigid restraint and snubber

deficiencies found on RHR pump A suction including binding, clamp '

slippage, one bent rigid restraint, and a few loosened nuts. One rigid  :

restraint binding on RHR pump B suction was also documented. This CR

was issued on January 8,1991, and closed on February 22, 1991.

CR l-91-01-069 documented six rigid restraint and snubber deficiencies  !

and one damaged pipe penetration seal found on RHR pump A suction and

RHR shutdown supply mode pumps A and B suction from reactor 6

recirculation loop B. The restraint deficiencies included one case of

embedded plate and concrete damage, one case of concrete spalling, one

case of a bent rigid restraint, and a number of bound up conditions.

This CR was issued on January 17, 1991, and closed on September 21, i

1992.

The root cause investigations of these CRs were inadequate. There was

no detailed documentation of conditions found on the eight bound up

rigid restraints and the determination of force directions and

magnitudes was not made. Although licensee personnel concluded that the

restraint damages were probably caused by improper adjustment of

supports by craft personnel during RF-2, there were no calculations or

explanations provided to determine that forces sufficient to damage the

supports could be present during the outage. The possibility of this

damage occurring during plant operations was not addressed.

.

All the damaged restraints were repaired or replaced and one restraint

was redesigned, fabricated and installed. All the deficient conditions

were restored to their proper design condition. Based on a walkdown of

the affected areas, the inspectors concluded that the repair and rework

to restore the defective restraints was adequate. i

Actions taken to correct the problems documented on both CRs included:

(1) walkdown of the RHR shutdown cooling suction at line temperature of

307* F, (2) observation of three pump starts on both loop A and loop B,

(3) review of RHR Operating Procedures, (4) review of work practices to  ;

determine if any additional inspection of supports was needed. Based on t

the results of these actions, licensee personnel concluded that the  !

restraint damages were probably caused by misadjustment of supports by  !

craft personnel during RF-2. Training on pipe support fundamentals and '

support adjustments was provided to craft personnel.

.

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The inspectors reviewed records, mapped the damage and deficiency

locations, discussed the matter thoroughly with licensee personnel and

saw no objective evidence to support the conclusion. The inspectors

-

concluded that the majority of the adverse conditions could not have i

been caused by improper restraint adjustments. In addition, no similar

problems had been identified by licensee personnel in other systems.

All the deficient conditions were identified during refueling outage

(RF-2), which started on October 14, 1990, and ended on March 9, 1991.

The inspectors reviewed licensee records and found that, prior to RF-2,

the licensee conducted an inspection of all the snubbers, three rigid

restraints, and one spring support in the subsystem in October 1990, no

problems were noted. The unsatisfactory explanation of the restraint

damage otserved, and a reported water hammer occurrence in November

1989, indicated the possibility that hardware deficiencies could have

been caused by water namer and could have existed during plant

operations. This possibility was not addressed.

The team concluded that the root cause evaluation was inadequate. There

was insufficient documentation to determine what activities were '

conducted in the vicinity of and on the affected piping during RF-2 that

could have caused the hardware damage and binding. No determination of

force directions and magnitudes could be made since there was no

detailed documentation of as found conditions for the eight bound up

rigid restraints. The licensee also failed to perform an operability

analysis for the as-found conditions to ensure plant operating license

conditions were maintained, and to ensure safe system operation if a i

similar event was to recur in the future. More than a dozen

deficiencies, including damage to four rigid pipe restraints, were  !

identified with pipe supports in the RHR area. The failure to

adequately assess the root cause and the safety significance of a

significant deficient condition is an example of a violation of ,

Criterion XVI of 10 CFR 50, Appendix B (461/93010-03A(DRS)). '

o 1-91-02-008 -

This CR documented four cracked welds on the rigid i

pipe guide for the RHR B pump discharge piping. The CR was issued on -

February 5,1991, and closed on Fabruary 28, 1991.

Licensee personnel were unable to identify the cause of the pipe

restraint weld cracks. The defect was believed to be defective

construction welds and the cracking was thought to be the result of

cyclic loading propagating into these defective welds. Even though the .

i

cause was thought to be due to defective construction welds, the welds

on the 17 other similar pipe guides in the RHR system were not inspected

prior to plant operation. All of the similar pipe guides were submerged

in the suppression pool during plant power operations. Based on the

team inquiries, the licensee stated that they believed two of the

similar restraints were without weld cracks since work was performed on

the pipe supports during RF-2 and no weld cracks were documented. t

,

!

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There was no objective evidence that an operability evaluation was  ;

performed to determine.the significance of the failure. During the '

inspection, licensee personnel determined that the weld cracks would not  ;

af fect piping operability since the welded members were under

compression loading. A licensee report, (Monitor Report 91-06-011,  !

dated June 5, 1991) stated that the failed and repaired ,nipe guide and

another pipe guide having nearly identical design will be inspected  !

during RF-4. The failure to perform an inspection of welds on similar  ;

pipe guides and the failure to assess the safety significance of the I

weld cracks is an example of a violation of Criterion XVI of 10 CFR 50, i

Appendix B (461/93010-03B(DRS)). '

o 1-92-08-031 -

LThis CR was written to document the premature failure  ;

,

of a Division Ill diesel generator air start solenoid valve. Illinois .l

Power Company had been notified by a 10 CFR 21 notification that ,

solenoid valves of the type used in the EDG air start system were

subject to " leak by" due to a weak spring. A stronger spring was

required to prevent this problem.

.

The NSED engineer, who performed the Part 21 evaluation, stated in his

analysis that the solenoid valve springs for all three diesels had been

replaced with the stronger springs. The valve spring for the division i

Ill EDG had not been changed and the problem was noted after the valve i

failed.

t

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The cause of the problem was attributed to personnel error; however, the  :

CR analysis did not address a reason for the personnel error, such as a  :

, lack of training or ineffective verification requirements.

o 1-92-10-052 -

This CR was written on two failures of replacement  ;

Gieger-Mueller (GM) tubes (stock code SH7410) after maintenance work.

Fifteen GM tubes from stock code SH7410 were sent to the vendor for i

testing and of the 15 tested, 11 passed. The conclusion was that, with  !

the exception of the four tubes which did not meet the acceptance  !

criteria, the remainder will function normally.

The team felt that the corrective actions taken in response to these  :

failures were inadequate. Seven failures out of a sample of seventeen l

should warrant a more rigorous investigation of the conditions that l

allowed so many defective components to be acquired.  :

o 3-92-07-031 -

This CR documented inadequate corrective actions

concerning the Master Equipment List (MEL). During the CR review, the

inspector noted that the individual who performed the root cause

evaluation did not have all of the training required by CPS procedures i

to perform root cause analyses. The lack of training did not appear to i

have adversely affected the quality of the root cause analysis.  !

Licensee personnel were notified and, since no other examples of this  ;

type problem were found, this was considered to be an isolated case. l

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4.3.3 Operability Analysis Reauirements

During the review of the action taken on CR l-91-02-008, which documented the

cracked welds on the RHR pipe guide,- the inspectors asked to review the system

operability analysis. Licensee personnel stated that the defects were found

during plant shutdown and the system was already inoperable for installation l

of a modification and an operability evaluation was not required. Licensee l

personnel, however, had previously concluded that the cracked welds were

caused by improper welding during construction combined with cyclic loading i

during plant operations. Based on this logic the failure occurred sometime l

during plant operations before the refueling outage. *

Similar justification for not doing an operability evaluation was applied to

the RHR restraint deficiencies and damages documented on CRs 1-91-01-029 and 1

1-91-01-069. The licensee's threshold for requiring an operability evaluation

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appeared to be unreasonably high and unrealistic. Based on these findings, r

licensee management indicated that they planned to review and revise the '

internal guidance for operability analysis.

5.0 Inspection Followup Items

[

Inspection followup items are matters which have been discussed with the  !

licensee, which will be reviewed further by the inspector (s), and which .;

involve some action on the part of the NRC or licensee or both. . Inspection  ;

followup items disclosed during the inspection are discussed in Sections

3.4.1.1. and 3.4.1.6. of this report.

6.0 Exit Meetina

The inspectors met at the Clinton Power Station with licensee representatives

(denoted in Paragraph 1) on August 3, 1993, to summarize the purpose, scope,

and findings of the inspection. The inspectors discussed the likely ,

information content of the inspection report with regard to documents and *

processes reviewed by the inspectors during the inspection, noting that one ,

document reviewed during the inspection was identified as proprietary.

Licensee personnel were requested to identify any proprietary information or >

material discussed during the exit meeting. The licensee did not identify any

information presented during the discussions as proprietary.

Additional information pertaining to the TERI program was obtained from the  ;

licensee during a meeting on August 24, 1993. During that meeting, attended r

by the Engineering Branch Chief, Division of Reactor Safety, the licensee

presented information describing the development, scope, depth and controls >

related to the TERI program. The information provided by the licensee r

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answered all NRC questions in that area.

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