IR 05000302/1993025

Insp Rept 50-302/93-25 on 931018-22.No Violations Noted. Major Areas Inspected:Design Changes & Plant Mods,Technical Support Activities & Engineering & Followup on Previously Identified Insp Findings

ML20058J732
Person / Time
Site:	Crystal River
Issue date:	11/18/1993
From:	Casto C, Matt Thomas NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20058J718	List: IR 05000302/1993025 ML20058J716
References
50-302-93-25, NUDOCS 9312140218
Download: ML20058J732 (11)

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l S o UNITED STATES f# e atc ,% NUCLEAR REGULATORY COMMISSION r

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REGION 11 I

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5 E 101 MAR ETTA STREET, N.W.. SUITE 29:X)

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ATLANTA, GEORGIA 30323 0199

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l Report No.: 50-302/93-25 j

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Licensee: Florida Power Corporation 3201 -34th Street, South l

, St. Petersburg, FL 33733 l Docket No.: 50-302 License No.: DPR-72 l r

Facility Name: Crystal River 3  ;

Inspection Conducted: October 18-22, 1993 l Inspector: Y #6tolm M. Thomas W //-/7-98 Date Signed

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Accompanying Inspectors: H. Whitener  !

n G. Wiseman l Approved by: ( ...e e < i CP Ca'sto, Chief ~'

h Mk / /- /fd 7 Date Signed l Test Programs Section j Engineering Branch  !

Division of Reactor Safety

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SUMMARY Scope:

f This routine, announced inspection was conducted in the areas of design changes and plant modific7.tions, engineering and technical support activitie and followup on previously identified inspection findings.

. Results:

In the areas inspected, violations or deviations were not identifie In general, the quality and technical content of the modification approval record (MAR) packages reviewed was adequate. Some of the 50.59 safety evaluations lacked detail. Both Site Nuclesr Engineering Services (SNES) and Nuclear

. Plant Technical Support (NPTS) provided timely support to the plant. The licensee had taken actions to address the weaknesses identified in the System Engineering group during the previous SALP period. The licensee's self assessment activities were a positive example of management's ongoing efforts to identify areas that need improvement.

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One inspector followup item was identified relating to the licensee's recalculation of torque switch setpoint PDR ADOCK 05000302 G PDR

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.i REPORT DETAILS I Persons Contacted f l Licensee Employees  !

i S. Balliet, Supervisor, Site Nuclear Engineering Services (SNES) ;

G. Boldt, Vice President, Nuclear Production  !

• W. Brewer, Supervisor, Nuclear Plant Technical Support (NPTS) {

D. Czufin, Supervisor, NPTS  !

• M. Donovan, Supervisor, SNES _ !

l C. Doyel, Nuclear Engineering Supervisor, Nuclear Operations Engineering ;

I *C. Dutcher, Superintendent, Nuclear Projects i

• J. Frijouf, Nuclear Regulatory Specialist, Nuclear Compliance

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• E. Froats, Manager, Nuclear Compliance i R. Knoll, Nuclear Engineering Supervisor, Configuration Management l

• P. McKee, Director, Quality Programs i l *R. McLaughlin, Nuclear Regulatory Specialist, Nuclear Compliance _i

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• L. Moffatt, Manager, NPTS l B. Moore, Manager, Nuclear Integrated Scheduling I i *S. Robinson, Manager, Nuclear Quality Assessments )

• P. Tanguay, Director, Nuclear Operations Engineering and Projects J. Terry, Manager, Nuclear Plant Systems Engineering, NPTS l *A. Washburn, System Engineer, NPTS

• R. Widell, Director, Nuclear Operations Site Support 1

• K. Wilson, Manager, Nuclear Licensing

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Other licensee employees contacted included engineers, QA/QC personnel, l

craftsmen, operators, and administrative personnel.

j NRC Personnel

• R. Butcher, Senior Resident Inspector l *P. Holmes-Ray, Senior Resident Inspector

*T. Cooper, Resident Inspector

• Attended exit meeting l

l Acrsnyms and initialisms used throughout this report are listed in the l last paragraph.

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I Design Changes and Plant Modifications (37700) Plant Modifications to Improve Reactor Safety The inspectors reviewed the licensee's initiatives to identify and implement plant modifications to improve reactor safet Documentation reviewed included Revision 3 of the Master Schedule which covered fuel cycles 9 through 11 and included three mid-cycle maintenance outages and three refueling outages.

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I The primary purpose of the Master Schedule was to provide a means ;

of defining and controlling major work for both operating and l outage periods. The scope of the Master Schedule includes all !

major modifications (greater than $50,000), significant corrective '

or preventive maintenance, inspections, and tests. The Master Schedule also controls the scheduling of regulatory requirement The Master Scheduling Group (MSG), which consists of management l representatives from engineering, nuclear plant operations, ;

nuclear maintenance, site support, and nuclear materials, is i responsible for prioritizing major MARS. The Change Review Team l (CRT), which consists of representatives from the same groups as !

the MSG, is responsible for reviewing and prioritizing minor MAR l The inspectors reviewed selected CRT meeting minutes and the j results of a MAR cutting meeting. Both meetings involved '

reviewing and prioritizing MARS for inclusion in either a i refueling outage or mid-cycle outage . schedule. The inspectors !

determined that the MARS were appropriately prioritize {

The inspectors reviewed the listed documentation and concluded that the licensee had demonstrated the use of an adequate l prioritization process for identifying and implementing plant i modification t b. Planning, Development, and Implementation of Plant Modifications

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The inspectors reviewed the Modification Approval Record (MAR) :

packages listed below to determine the adequacy of evaluations to meet 10 CFR 50.59 requirements; verify that the MARS were reviewed :

and approved in accordance with TS and applicable administrative - !

controls; verify the subject modifications were installed (for l those that could be physically inspected) in accordance with the t MAR package and other applicable design documents; verify that _!

applicable plant operating and design documents (drawings, plant ;

procedures, FSAR, TS, etc.) were revised to reflect the subject modifications; verify that the modifications were reviewed and incorporated into the operations training program as applicable; ;

and verify that post modification test requirements were specified .

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and adequate testing was performe !

MARS 86-12-02-03,-04 and -05, Nuclear Service Closed Cycle Cooling ;

Water (SW) Temperature Control with Recirculating Emergency

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Nuclear Service and Decay Heat Seawater (RW)

This modification was accomplished during the recent mid-cycle :

maintenance outage 9M to provide temperature control'of the SW !

system. During the colder months of the year, the temperature i control is designed to maintain the SW temperatures at 80 degrees F or above to minimize thermal stresses on SW cooled equipment and reduce in-plant piping condensation which has been a problem in the past. The MARS were reviewed to verify that proper evaluations had been conducted from the engineering stage through the installation of modification hardware. The inspectors walked down !

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the modification with the system engineer and verified that the as-built configurations and material condition of installed <

l piping, pipe supports, valvrs, and instruments were in accordance i with the MAR package The system engineer demonstrated excellent l knowledge of the plant systems and was well versed in the !

l modification details.

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l MAR 91-06-10-01, Filtration of the Domestic Water (D0) to the RW !

l Pump Bearings

This modification was installed in February 1993, to filter the .

normal domestic water supply to the RW pump bearings. The purpose l

, of the filters was to remove lime and calcium carbonate suspended i l deposits from the bearing cooling water flow stream. Deposits of i these compounds have been contributing factors to previous pump bearing cooling low flow problems. The installation package was reviewed to verify that the changes had been review and approved in accordance with 10 CFR 50.59. A modification walk down was i

conducted by the NRC inspectors to verify the installation was l performed in a technically adequate manner. The new filtration

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system has been very effective at removing the suspended solids from the bearing cooling water system. No bearing cooling water clogging problems have been experienced since the filtration system has been operational. However, the operating filter has required frequent cleaning (almost daily) by plant maintenance personne l MAR 90-05-01-01, Fire Damper Operability This modification involved the addition of double negator springs (two springs on each side of a fire damper) to safety related fire dampers AHFD 223, 224, 225, 226, 266, 271, 273, and 239 to address j fire damper closing problems described in NRC Information Notice (IN) 89-52 and LER 90-02. The MAR had been partially implemented but modifications to damper AHFD 239 were deferred pending further engineering study described in Request For Engineering Assistance (REA) Number 921744. The status of the licensee's corrective j actions for this issue is discussed in HRC Inspection Report 50- l 302/93-2 MAR 93-06-07-01, FWV-28, 29, and 30 NRC GL 89-10 Resolution This modification involves changing the motor from 3600 RPM to I 1800 RPM for feedwater valve FWV-28 in order to improve the valve's reliability by reducing the speed of the motor which in turn reduces the operator inertia. This MAR will also increase the size of the motor power cables for FWV-28, 29, and 30 in order to improve motor voltage and performance. This MAR resulted from a problem identified during the licensee's Electrical Calculation Enhancement Program. Problem report PR 92-0178 was written to document the original issue. The inspectors reviewed the MAR package and associated 50.59 safety evaluation to verify that l

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design inputs included, but were not limited to, valve stroke l' times, motor weight differences, EQ, fire protection, voltage requirements, etc. This MAR had not been implemented at the time .

of this inspectio '

MAR 93-07-01-01, MOV Brake Removal  ;

This modification disabled the electric motor brakes on the makeup ,

system motor operated valves (MOV) MUV-58 and MVV-73. This was !

accomplished by disconnecting the electrical leads of the brake in l the Limitorque operator limit switch termination box and '

decoupling the motor from the brake assembly by removing the brake internal stationary and rotating disc. This problem was reported to the NRC in Licensee Event Report (LER) 93-08, dated August 4, 1993. The inspectors reviewed the 50.59 safety evaluation and !

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verified that applicable design inputs were considered. This included, but was not limited to, inputs such as fire protection, EQ, diesel generator loading, GL 89-10 MOV Program, etc. The inspectors also reviewed Limitorque Maintenance Update 92-2, which discusses the motor brake concern. The inspectors noted that the I safety evaluation for the MAR did not discuss the Maintenance

! Update concern for M0Vs with non-locking gearing and worm gear l

ratios less than 30:1. This concern was discussed with engineering personnel who provided the inspectors with additional

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valve and actuator design input data. The inspectors reviewed the l additional design data and determined that the gearing ratios were greater than 30:1 and removing the motor brakes was acceptabl The inspectors noted that, although the appropriate design inputs were considered, additional detail in the safety evaluation

, addressing the acceptability of removing the motor brakes based on l the gear ratios would have enhanced the safety evaluation.

l The inspectors concluded from reviewing the above MAR packages that the overall quality and technical content were adequate. However, the level of detail provided for some of the 50.59 safety evaluations needed l

improvement. There were no violations or deviations identified.

l 3. Engineering Support Activities Organization and Staffing Engineering and technical support were provided by both on-site and corporate organizations. On-site technical support was provided mainly by NPTS (which included system engineering) and SNES. This support included equipment performance trending, responding to REAs, preparing MARS, performing safety evaluations, failure analysis, etc. On-site engineering was realigned in l November 1992, with SNES continuing to report to the Director of l Nuclear Operations Engineering and Projects. NPTS now reports to l the Vice President Nuclear Production. Licensee personnel l indicated that this change enables NPTS to be more timely in providing support to the plant. The duties and responsibilities

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of the various engineering departments are described in procedure NEP-102, Organization of Nuclear Operations Engineering and Projects.

The insnectors interviewed licensee personnel and reviewed station records to evaluate the engineering involvement in support of day to day plant operations. The type of records reviewed included but was not limited to engineering reports of system status, problems, and performance; problem reports and associated actions; failure analysis and trends; ar.d MAR packages. Procedures reviewed included CP-111, Initiation and Processing of Precursor Cards and Problem Reports and N0D-14, Evaluating Operability. A sample of Problem Reports was reviewed to determine that the description of the problem, operability determination, root cause evaluation, and corrective action plan were adequately performed.

The inspectors concluded that the problem reports were adequately processed and engineering support in the resolutien of day to day plant problems was evident. Problem Reports reviewed included the following:

PR N Subject 92-0168 lhru Wall Leakage on SW Header 92-0207 Inadequate feedwater Flow 93-0043 Main Steam Safety Valve Incorrect Lift Pressure 93-0062 Foreign Material in M11V-22 93-0125 Torque Switch failure on EFV-33 93-0198 DHV-12 Torque Switch Setting 93-0209 RHW-124 Hanger Support Inadequate In regard to PR 93-0198, the licensee identified a condition where the torque switch trip (TST) set point was based on a linear extrapolation from test data obtained at 52 percent of design basis differential pressure (DBDP) to 100 percent DBDP. Current industry information indicates that a linear extrapolation from less than 80 percent of DBDP may under predict the required thrust at design basis differential pressure and flow conditions. based on a reevaluation of thrust required at full DBDP and flow conditions the licensee concluded that the TST setpoint for MOV DHV-12 is set below the minimum required per the thrust calculation. A valve operability evaluation was performed and the valve was considered operable based on the following:

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DHV-12 safety function is to open.

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The opening TST is bypassed for 21 percent of the valve strok The high thrust values occur at the initial opening (disk r pull out) of the valve while the TST is bypasse Recalculation, using valve test data hnd current industry information, indicate that the motor operator is capable of opening the valve at the higher thrust values with a 14 ;

percent margi The licensee has initiated a corrective action item to review all MOVs for a similar condition and issue wu k orders to reset torque switches to the new valves at the next outage. This matter is identified for followup inspection during the phase II inspection of the Generic letter 89-10 progra '

IFI 50-302/93-25-01, Review licensee recalculation of MOV thrust requirements and resetting of torque switch trip setpoin The inspectors reviewed SNES and NPTS engineering reports to evaluate engineering activities to support system reliability and safe plant operation. These reports serve to provide management '

the oversight of overall plant system status, problems, and problem resolution. The review included portions of the !

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NPSE Monthly Reports - October, November, December 1992

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NPSE Quarterly Reports - 1st and 2nd Quarters,1993

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SNES Quarterly Reports - 4th Quarter 1992,1st, 2nd and 3rd Quarters, 1993 7.e reports provided objective evidente that engineering has ,

or active in providing plant modifications, monitoring system pm /ormance, identifying system problems, establishing root cause d resolving problems. In the previous SALP evaluation a weakness was identified concerning system engineering monitoring system performance and performing system walkdowns. Based on this '

review system engineering performance in these areas has improve An additional responsibility has been assigned to system engineers which involves review and priority ranking of all work orders on ,

the enginearv assigned systems. This responsibility enhances the system manager concept which the licensee has recently introduced. !

The system engineer is traditionally responsible for the performance and reliability of assigned systems. This concept involves giving the system engineer the authority commensurate with this responsibilit ,

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The inspectors concluded that SNES and NPTS were adequately l staffed with knowledgeable personnel. Timely and effective )

engineering and technical support were being provided to the j plan '

An example of th n imely support was demonstrated during the l preparation and im;. . ementation of MAR 93-07-01-01, which is j discussed in paragraph 2.b. of this inspection report. Problem I report PR 93-0173 was initiated on June 28, 1993, for the motor !

brake issue. Engineering received the PR on July 2,1993. SNES !

?repared, issued, and supported implementation of the MAR package j between July 3-7, 199 '

Another example of timely engineering support was noted when the common train SW suction pipe header was discovered to be leaking due to deterioration from external corrosion. An emergency temporary MAR was irsued to encapsulate the section of pipe that was in the poorest condition. In addition, extensive calculations were performed to show that the remainder of the header, although in a degraded condition, still met all code requirements and is considered to be in operable condition. These prompt actions by engineering were instrumental in keeping the plant out of a TS action statement that would have required immsdiate plant ,aut dow A proactive action by engineering was demonstrated when MAR 92-04-11-01 was issued to provide for replacement of the original main turbine generator hydrogen gas regulator. The original ..gulator had degraded to the point where the pressure was falling below l minimum. The new type regulator will assure that hydrogen cooling ;

will be maintained at a pressure that does not degrade the main i generator's design efficiency. This is considered to be a l proactive design change that eliminates a problem prior to having an adverse effect on plant power productivit . Quality Assurance (QA) Audits and Self Assessment Activities l The inspectors reviewed QA audits, assessments, and Quality Program system functional audits of Nuclear Plant Systems Engineering safety related activities conducted by the Nuclear Quality Programs i organization. The assessments and audits were part of the overall !

Quality Assurance Program at Crystal River. The inspectors reviewed results of the following Site Quality activities that were either completed or in progress:

93-09-N0E0, September 1993, Assessment of Nuclear Operations ,

Engineering Organization  !

93-07-HVAC, July 1993, System Functional Audit of Heating, j Ventilation, Air Conditioning (HVAC) {

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92-10-N0E0, September 1992, Assessment of Nuclear Operations Engineering Organization

92-11-RWSW, November 1992, System functional Audit of Service Water (RWSW)

In addition to reviewing results of the above activities, the inspectors l reviewed several engineering response memorandums to the audit -

i observations and recommendations.

The inspectors found that the Quality Programs organization has been ,

heasily involved in assessing engineering activities. This included detailed reviews of the plant MOV Program and specific Safety System Functional Inspections (SSFIs). Quality Programs has been aggressive in identifying safety system functional problems and other areas of weakness. A number of Problem Reports were initiated to document corrective actions to the audit findings. These efforts have identified

to site senior management several program areas where improvement is neede In addition to the QA audits, the licensee also perfo,med a self assestment of Nuclear Engineering using the INPO document 90-009, Guidelines for the Conduct of Design Engineering. Approximately 400 questions were developed based on a review of the INP0 document. The self assessment began in 1990 and was conducted over a three year period. During the assessment, 152 recommendations were generated. The recommendations were prioritized and an action plan was developed. The inspectors reviewed selected recommendations and the summary of actions taken. It was determined that the licensee had either taken actions or already had controls in place to address the recommendations.

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Other self assessment functions performed by the licensee included trending various activities assigned to SNES and NPTS. These trends provided engineering with information on areas such as REAs that needed additional effort to reduce the backlo The inspectors considered the QA audits and self assessment activities to be positive examples of the licensee's efforts to identify areas that need enhancement in order to improve the engineering support being provided to the plan Violations or deviations were not identified in the areas inspecte . Followup on Previously Identified Inspection Findings (92701, 92702) (Closed) Violation 50-302/92-11-01, concerning prompt corrective actions not being taken relative to the failure of emergency feedwater valve EFV-14 to fully close during differential pressure (d/p) testing on October 13, 199 .. . ~

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l In their response to the violation dated July 9,1992, the licensee specified corrective actions which included replacing valve EFV-14 and its associated motor operator with a different design. Similar valves EFV-11, 32, and 33 were also modifie EFV-11 and its associated motor operator were replaced with a I different design; EFV-32 and 33 motor operators were modified to increase thrus The modifications for the above EFVs were reviewed and are discussed in NRC inspection report 50-302/92-19. The modifications and the post modification testing performed were i determined to be adequate. In addition to the modifications, the i licensee's response to the violation also stated that the lessons L learned from this finding would be incorporated into the MOV program by September 30, 1992. The inspectors reviewed training records which indicated that training was provided on September 17, 1992, to personnel involved in the preparation of d/p calculations for GL 89-10 HOVs. The training addressed i lessons learned from this finding. The inspectors considered the !

actions taken by the licensee adequately addressed the violatio ;

This violation is close l l (0 pen) IFI 50-302/92-19-01, concerning the licensee's resolution l of problem reports POPR-90-0058 and CMPR-91-000 Problem report l P0PR-90-0058 was related to reactor building spray valves BSV-16 )

l and BSV-17 not meeting stroke time requirements, and problem !

report CMPR-91-0008 was related to feedwater discharge isolation !

valves FWV-31 and FWV-32 not meeting closure time requirement During review of this item, the inspectors noted that problem report PR 93-218 concerning building spray start time was interrelated with POPR-90-0058 and PR 93-0022 concerning reactor l building temperatures following a main steam line break was

! interrelated with CMPR-91-0008. These new prs had not been.

l resolved at the time of this inspection. The inspectors will i

continue to follow the resolution of- these related prs during future inspections. This IFI will remain ope . Exit Interview j l

l The inspection scope and results were summarized on October 22, 1993, I l with those persons indicated in paragraph 1. The inspectors described the areas inspected and discussed in detail the inspection finding One IFI (50-302/93-25-01) was identified concerning the recalculation and setting certain TST setpoints. Proprietary information is not contained in this report. Dissenting comments were not received from the license I J

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7. Acronyms and Initialisms BSV Building Spray Valve CFR Code of Federal Regulations CRT Change Review Team DBDP Design Basis Differential Pressura DHV Decay Heat Valve d/p Differential Pressure EFV Emergency Feedwater Valve EQ Environmental Qualification F Fahrenheit FSAR Final Safety Analysis Repo*T FWV Feedwater Valve GL Generic Letter HVAC Heating Ventilation and Air Conditioning IFI Inspector Followup Item IN Information Notice INP0 Institute of Nuclear Power Operations LER Licensee Event Report MAR Modification Approval Record MOV Motor Operated Valve MSG Master Scheduling Group MUV Makeup Valve

.EP Nuclear Engineering Procedure NPTS Nuclear Plant Technical Support PR Problem Report QA Quality Assurance QC Quality Control REA Request for Engineering Assistance RPM Revolutions Per Minute RW Nuclear Service and Decay Heat Seawater SALP Systematic Assessment of Licensee Performance SNES Site Nuclear Engineering Services SSFI Safety System Functional Inspection SW Nuclear Service Closed Cycle Cooling Water TS Technical Specifications TST Torque Switch Trip