ML20236K894

Discusses Review of Util 860815 Submittal of Integrated Sys Study of CRD Mechanism Position Instrumentation.Util Has Not Provided Acceptable Proposal to Upgrade Control Rod Position Instrumentation Sys.New Schedule Requested within 30 Days

ML20236K894
Person / Time
Site:	Fort Saint Vrain
Issue date:	07/31/1987
From:	Heitner K Office of Nuclear Reactor Regulation
To:	Robert Williams PUBLIC SERVICE CO. OF COLORADO
References
TAC-62198, NUDOCS 8708100036
Download: ML20236K894 (2)

Text

1 July 31, 1987 -

l Docket No. 50-267 l Mr. R. O. Williams, Jr.

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Vice President, Nuclear Operations Public Service Company of Colorado L O)b 1 P. O. Box 840 l Denver, Colorado 80201-0840 j

SUBJECT:

EVALUATION OF INTEGRATED SYSlEMS STUDY OF CONTROL R0D DRIVE MECHANISM l POSITION INDICATION INSTRUMENTATION FOR THE FORT ST. VRAIN NUCLEAR l GENERATING STATION (TAC N0. 62198)

We have completed our review of your submittal dated August 15, 1986 (P-86522),

which contained an integrated systems study of the control rod drive mechanism (CRDM) position instrumentation at Fort St. Vrain (FSV). This review was per-formed by our contractor EG&G, Idaho, Inc. Their Technical Evaluation Report I (TER) is enclosed. l

) 1 The staff has reviewed this TER and concurs with the EG&G, Idaho, Inc. conclu- l sion that Public Service Company of Colorado has not provided an acceptable i proposal to upgrade the control rod position instrumentation system for Fort St. Vrain. Section 4 of the TER gives the reasoning for this conclusion. We are forwarding this TER to you to allow for another proposal for the upgrading '

of the control rod position instrumentation system.

We request that you provide a new schedule for completing this proposal within 30 days of the date of this letter. The new proposal should be provided within the following 120 days after your schedule is established.

The information request in this letter affects fewer than 10 respondents; there-fore OMB clearance is not required under P.L.96-511.

Sincerely, lSf k[l M g6 07073; P Kenneth L. Heitner, Project Manager Project Directorate - IV 4 05000267 PDa Division of Reactor Projects - III, IV, and V and Special Projects Office of Nuclear Reactor Regulation

Enclosures:

As stated cc w/ enclosure:

See next page DISTRIBUTION F' )

(Docket FS e NRC POR Local PDR PD4 Reading F. Schroeder J. Calvo P. Noonan K. Heitner 0GC-Bethesda E. Jordan J. Partlow ACRS (10)

PD4 Plant File po_1v [

PD-IV M PD-IV )(H PNoonan

• KHeitner/tys JCalvo 07/%V87 07/*E/87 07/3l/87

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l Mr. R'. O. Williams .

Public Service Company of Colorado Fort St. Vrain  !

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Mr. D. W. Warembourg, Manager Albert J. Hazle, Director l

Nuclear Engineering Division Radiation Control Division  :

Public Service Company Department of Health ]

of Colorado 4210 East lith Avenue  ;

P. J. Box 840 Denver, Colorado 80220 Denver, Colorado 80201 1 Mr. David Alberstein, 14/159A Mr. R. O. Williams, Acting Manager  ;

GA Technologies, Inc. Nuclear Production Division l l Post Office Box 85608 Public Service Company of Colorado j San Diego, California 92138 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 il Mr. H. L. Brey, Manager  !

Nuclear Licensing and Fuel Division Mr. P. F. Tomlinson, Manager j Public Service Company of Colorado Quality Assurance Division l P. O. Box 840 Public Service Company of Colorado Y Denver, Colorado 80201 16805 Weld County Road 19-1/2 l Platteville, Colorado 80651 i Senior Resident Inspector j U.S. Nuclear Regulatory Commission Mr. R. F. Walker P. 0. Box 840 Public Service Company of Colorado {;

Platteville, Colorado 80651 Post Office Box 840 Denver, Colorado 80201-0840 l Kelley, Stansfield & 0'Donnell Public Service Company Building Commitment Control Program Room 900 Coordinator  ;

550 15th Street Public Service Company of Colorado .

Denver, Colorado 80202 2420 W. 26th Ave. Suite 100-D i Denver, Colorado ~B0211 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission i 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Chairman, Board of County Commissioners j of Weld County, Colorado l Greeley, Colorado 80631  ;

Regional Representative 'l Radiation Programs Environmental Protection Agency 1 Denver Place l 999 18th Street, Suite 1300 Denver, Colorado 80202-2413 1

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Enclosure

.'. EGG-NTA-7705 May 1987 INFORMAL REPORT i

EVALUATION OF INTEGRATED SYSTEMS STUDY OF CONTROL R00 DRIVE MECHANISM ROD

/daho POSITION INDICATION INSTRUMENTATION FOR THE National FORT ST. VRAIN NUCLEAR GENERATING STATION Engineerilig Laboratory Managed by the U.S.

Department D. E. Jackson ofEnergy C. L. Naiezny

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Prepared for the net No. DE AC07 ?t9lD015?O r

O.S. NUCLEAR REGULATORY C014ISSION PA I

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l DISCLAIMER i

This report was prepared as an account of work sponsored by an j agency of the United States Government. N,etther the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, )

i of any information, apparatus, product or process disclosed in l this report or represents that its use by such third party would l not infringe privately owned rights.

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EGG-NTA-7705 l

t EVALUATION OF INTEGRATED SYSTEMS STUDY OF CONTROL ROD DRIVE MECHANISMS R0D l POSITION INDICATION INSTRUMENTATION FOR THE l FORT ST. VRAIN NUCLCAR GENERATING STATION l Docket No. 50-267 TAC No. 62198 INEL Reviewer - D. E. Jackson INEL Program Mgr - C. L. Nalezny NRC Lead Reviewer - R. H. Lasky NRC FSV Project Mgr - K. Heitner NRC Program Mgr - M. Carrington Published May 1987 EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 l Under 00E Contract No. DE-AC07-761001570 FIN No. D6023 l

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t ABSTRACT This EG&G Idaho, Inc., report presents the results of an evaluation of an integrated systems study (engineering evaluation) of the control rod drive nechanism rod position indication instrumentation for the Fort St.

Vrain Nuclear Generating Station which was submitted to the Nuclear Regulatory Commission (NRC) by the licensee, Public Service of Colorado (PSC). The evaluation by EG&G Idaho, Inc., concludes that PSC has not complied with the NRC directive to prepare an engineering evaluation of the problems experienced with the c^ntrol rod drive rod position indication ,

I because it does not adequately address the problems that were experienced at the Fort St. Vrain Nuclear Generating Station, and does not propose acceptable component replacements. In addition, the proposed replacement instruments that are important to safety (full-in limit swithch and rod

! position potentiometer) do not comply with the quality standard and instrumentation requirements of General Design Criteria No.1 and 13 o'f Appendix A to 10CFR50.

Docket No. 50-267 TAC No. 62198 ii

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l FOREWORD 1 1

This report is supplied as part of the " Review of Plant Specific Licensing Actions for Operating Reactors," Task 1-14 being conducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, l Division of PWR Licensing-B, by EG&G Idaho, Inc., NRR and I&E Support Branch.

The U.S. Nuclear Regulatory Commission funded the work under authorization B&R 20-19-10-11-2, FIN No. D6023.

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Docket No. 50-267 TAC No. 62198 iii f

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6 CONTENTS ii ABSTRACT ..............................................................

iii FOREh0R0 .... .........................................................

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1.0 BACKGROUND

2. DESIGN BASES CRITERIA ............................................. 3 4

i 3.0 EVALUATION .......................................................

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l 3.1 Control Rod-Pair Full-In and Full-Out Limit Switches ............. 6 3.2 Slack-Cable Limit Switches ....................................... l 6 '

l 3.3 Position Potentiometers ..........................................

l 3.4 Results of Evaluation of Proposed Control Rod Drive Temperature Limits ............................................... 7 G

4.0 CONCLUSION

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5.0 REFERENCES

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1 EVALUATION OF INTEGRATED SYSTEMS STUDY OF CONTROL ROD ORIVE MECHANISMS ROD l POSITION INDICATION INSTRUMENTATION FOR THE FORT ST. W.AIN NUCLEAR GENERATING STATION ,

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1.0 BACKGROUND

Fpilowing a scram on the morning of June 23, 1984, at the Fort St.

Vrain Nuclear Generating Station (FSV), 6 of 37 rod pairs failed to scram.

As a result of this event, the Director of Nuclear Reactor Regulation (NRR) ordered that an audit of the overall operation of FSV be performed. This audit was to include problem areas associated with the June 23 scram.

The Control Rod Drive Mechanism (CRDM) is comprised of the shim motor I and motor brake assembly, the gear reduction to the cable drum and the control rod pairs suspended by cables from the drum. Instrumentation-related components, integrated into the CRDM, used to determine control rod positions include the rod position potentiometers, rod-in and rod-out limit switches, limit switch cams and gear reducers, and the slack-cable j l

indication devices.

l On July 30, 1984, the NRC was informed of numerous and varied control rod instrumentation anomalies in several refueling regions in the reactor (Refence 1, section 3.1). The eleven anomalies included: simultaneous rod-in and rod-out indication, out-limit switch lights remaining lighted, indications of partial rod withdrawal, no position signals, disparity between analog and digital rod position information, and slack cable I indication.

The results of the NRC audit were documented in a preliminary report which was issued on October 16, 1984, (Reference 1). The report contains findings to be addressed both before and after plant restart. The NRC staff noted that a number of deficiencies need to be corrected on a long-term basis 1

1 foilowingrestart. The licensee was directed to submit schedules within 60 days of restart for completing these items. One of the items listed under

" Actions Required Following Restart" is " Conduct an integrated systems study 'j (engineering evaluation) to resolve rod position indication, maintenance and I operability questions."

On August.15, 1986, Public Service Company of Colorado (PSC) issued a report entitled " Integrated Systems Study (Engineering Evaluation (

EE-12-0013) of the Control Rod Drive Mechanism Rod Position Indication Instrumentation" (Reference 2). This Engineering Evaluation by PSC was in l

response to the requirement listed in Section 4.e of the Executive Summary of Reference 1, and is the subject of this staff evaluation.

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2. DESIGN BASIS CRITERIA The following General Design Criteria (GDC) of Appendix A to 10 CFR 50 were applied to the evaluation of the Fort St. Vrain Integrated Systems l Study of Control Rod Drive Mechanisms Rod Position Indication instrumentation.

Criteria 1 - Quality Standards and Records. Structures, systems, and c6mponents important to safety shall be designed, fabricated, erected and tested to quality standards commensurate with the importance of the ,

i safety function to be performed. i Criteria 13 - Instrumentation and Control. Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for i accident conditions as appropriate to assure adequate safety, including l those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls 4 shall be provided to maintain these variables and systems within I prescribed operating ranges.

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3. EVALUATION The PSC response to the NRC preliminary report, specifically addresses the requirement for an integrated systems study (engineering evaluation) to resolve rod position indication (RpI) maintenance and operability questions. It was issued in the form of an Engineering Evaluation (EE-12-0013) and documents the evaluation of several active and passive electrical components in the RPI system to identify any potential design deficiencies. These components included the control rod-pair full-in and full-out limit switches, slack cable limit switches and rod-pair position potentiometers. In addition the associated lights and meters in the control room were evaluated. The EG&G evaluated the integrated systems study by PSC to determine if the root causes of the anomolies had been identified, if PSC had proposed design changes that would correct them, and if the proposed design changes were in compliance with General Design Criteria (GDC) 1 and

13. The results of the EG&G evaluation follow.

3.1 Control Red-pair Full-In and Full-Out Limit Switches The most common problem observed with the full-in and full-out limit switches was pitting, errosion and corrosion of the switch shafts which caused binding of the switchs which resulted in inaccurate indications in the control room. PSC stated that the pitting and erosion of the shafts, and the deposition of shaft metal on switch housings was due to the high angle at which the actuating cams contact the switch shafts exerting excessive lateral force on the shaft thereby causing binding and or breakage of the switch.

The conclusion that the high angle of the cams is responsible for the high lateral loads which caused the switches to bind is not supported. The NRC preliminary report stated that the instrumentation anomalies are believed to be the result of mechanical damage or exposure of the CRDMs to a hot, moist atmosphere and a subsequent core depressurization which resulted in condensation of moisture. It is very likely that moisture caused pitting of the plunger and contributed to the failure of the rod position 4

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b inst'rumentation. The study by PSC does not include an evaluation qf corrosion of the switch components, and the effect corrosion would bave on j the mechanical operation of the switches. y ]

l PSC proposes to replace these switches with proximity sensors which are specially designed for their application and environment. This is an acceptable solution if the sensors are designed to satisfy appropriate functional, and o;,erational requirements, and are qualified for the environment in which they must operate. However, the statements " designed [

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! for FSV's application and environment" and " capable of operation at several j hundred degrees Fahrenheit above the requirements" are vague and do not adequately specify the operating conditions and functional requirements for > l the design, fabricaton, and procurement of the sensors.

l The full-in limit indications can be classified "important to safety" because they are the primary means of verifying that the control rod drives ]

I (CRDs) have fulfilled their reactor scram safety function. The proposed changes to the full-in limit switch does not comply with the design standards requirement of GDC 1, or the instrumentation requirement of GDC

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The concern that the out-limit cam being overdriven and damaging the potentiometer shaft is not addressed in the PSC engineering evaluation (Reference 2). It is stated that, " targets for the sensors will be provided by replacing the existing cams with stainless steel cubes. These targets are the same size as the existing cams and will not cause any new mechanical interference problems." They will therefore have the same potential to damage the potentiometer drive coupling and shaft, due to overdriving, as the original cams. It is stated that administrative limits have been imposed to prevent overdriving the system. However, if it is still necessary to provide replacement potentiometers with additional turns to avoid being damaged when overdriven, it should also be necessary to address f the fact that the stainless steel targets will damage the potentiometer I shafts when the system is overdriven. ,

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< Sinc'e the' analog position indicators are "important to safety," the out-limit cam must be designed so that overdriving it will not damage the potentiometer shaft, or so that they can not be overdriven.

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Therefore, it is concluded that PSC has not performed a satisfactory engineering evaluation of the problems associated with the full-in/ full-out switches, has not proposed solutions which will correct the problems, and has not complied with the requirements of GDC 1 and 13.

3.2 Slack-Cable Limit Switches In Reference 2, PSC states that the engineering evaluation of the I l

slack-cable limit switches revealed no known failures and the switch j mar,afacturer confirmed this to be an acceptable application of the switch. l However, the list of anomalies in section 3.1 and table 3.1 (Reference 1) indicate a slack-cable switch failure which was not addressed by the l licensee. The statement that the manufacturer confirmed this to be an l acceptable application of this switch is not acceptable because it does not resolve the question of why the switch failed. The engineering evaluation j should have included a comparison of the design requirements for the switch  !

with the known cperating conditions to determine if the failure was design, l fabrication, or maintenance related. Therefore, it is concluded that PSC has not performed a satisfactory engineering evaluation of the slack-cable I

limit switches and has not proposed any solution to correct the problem. l 1

3.3 Position potentiometers The engineering evaluation of the position potentiometers by PSC revealed that resistivity changes were caused by moisture intrusion into the case and that driving these potentiometers past their limits had caused broken bodies and drive gears. The changes in resistivity caused some measurement uncertainty. Section 3.3.3 of the NRC preliminary report (Reference 1) states that overdriving the potentiometers can result in the out-limit cam rotating around to the point that it can interfere with and cause damage to.tne potentiometer shaft coupling.

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pSC proposes to replace these potentiometers with new ones specifically designed and fabricated for this application. The replacement j potentiometers are to be built with a 10 turn electrical section centered on j a 15 turn mechanical section and mounted the same as the Bechman Model 7603s {'

which are currently used. These and other specifications are included in

" Specification for Prototype Potentiometers," Appendix D, of Ref. 2. The proposed replacement potentiometers are an improvement. However, because j they provide the operator with continuous position information for all the l rods du/ing operation and following a scram, they are considered "important  !

to safety." Therefore, they must be designed, fabricated and procurred to f the requirements of the appropriate GDCs. The specifications for the potentiometers presented in Reference 2 are appropriate for a commercial grade component, but did not comply with the quality standards and reporting requirements of GDC 1, which are necessary for a component that must comply I 1

with GDC 13. For instance, the requirements for a quality assurance program l were not called out. Environmental conditions such as minimum and maximum temperatures, and maximum moisture content of the Helium atmosphere were not i specified. If PSC procures commercial grade potentiometers for this l l application, they should develop a formal program to qualify the  ;

components. In developing the qualification program, PCS should apply the applicable guidance contained in Chapters 3.11 and 4.6 of the kRC Standard Review Plan (Reference 3) which deal with environmental qualification of mechanical and electrical equipment, and the functional design of control rod drive systems.

3.4 Results of Evaluation of Proposed Control Rod Drive Temperature Limits Since the RPI instrumentation is an integral part of the CRDM, it is subject to the same design and functional requirements as the CRDMs. In this context, reference is made to the NRC Safety Evaluation Report (SER),

dated December 24, 1987 (Reference 4). Reference 4 is an evaluation of a PSC proposal to increase the operating temperature limits of the FSV Control ,

Rod Drive and Orifice Assemblies (CRDOAs) to 3000F, The findings and deficiencies identified in Reference 4 are generally applicable to the CRDM 7

tod position instrumentation. The operating environment of the CRDMs and the rod position instrumentation is the same. Therefore, the important to safety instrumentation should be subject to the same requirements. The deficiencies in the CRDM submittal (Reference 4) that are applicable to the safety related rod position instrumentation are listed below.

1. PSC did not provide acceptance criteria developed from the functional, operational and design specifications against which to evaluate the proposal.

2. PSC did not provide information on the mechanical and electrical properties of materials in the RPI components as a function of temperature, humidity, pressure, and radiation.

3. PSC did not address maintenance of RPI instruruentation.

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4. CONCLUSIONS The licensee for the Fort St. Vrain Nuclear Power Station, Public Service of Colorado, has not provided an acceptable proposal to upgrade a selected number of Control Rod Position Instrumentation Systems. The rational for this conclusion is given below.

The submittal was reviewed for compliance with the NRC requirement that PSC conduct an integrated systems study to resolve rod position indication maintenance and operability questions, and the applicable requirements of the General Design Criteria (Appendix A to 10 CFR 50) for "important to safety" instrumentation.

The licensee did not perform a thorough evaluation of the failures that were identified in the NRC preliminary report. The contribution of corrosion of the full-in limit switches was not evaluated, the design of the i

relacement targets for the full-in/ full-out limit switches did not consider 1

the potential for damaging the rod position potentiometers when the control rods are overdriven, the failure of the slack cable limit switches was not evaluated, and the specification for the replacement red position potentiometers did not include all the environmental conditions that the components could be exposed to and did not define how the potentiometers would be qualified. In addition, the proposed replacement instruments that I are important to safety (full-in limit switch and rod position potentiometer) did not comply with the quality standard and instrumentation requirements of GDC 1 and 13 of Appendix A to 10 CFR 50.

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4.0 REFERENCES

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1. NRC Letter, Denton to Walker, Preliminary Report Related to Restart and j Continued Operatin of Fort St. Vrain Nuclear Generating Station, l 1

l (G-84392) dated October 16, 1984. )

2. PSC Letter, Warembourg to Berkow, Fort St. Vrain Rod Position i i

Instrumentation Integrated Systems Study, (P-86522) dated )

August 15, 1986.

3. U. S. Nuclear Regulatory Commission, Standard Review Plan, NUREG-0800, l

Rev. 2, July 1981. l

4. NRC letter, Heitner to Williams, Control Rod Drive and Orifice Assembly 3000 F Temperature Limits Fort St. Vrain Nuclear Generating Station, G-86664, December 24, 1986.

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io. s,omscaiuo ono. darion =.we *=o maiu o aoonass ,, ,s.c , n.. ryen o, maronr Division of Reactor Projects IV Informal l Office of Nuclear Reactor Regulation . ,..ioo Cov...o ,,

U.S. Nuclear Regulatory Commission 1 Washington, D.C. 20555 t, SvP'LiutN T ARY NOrts 1

IJ Aelf maCT GOD averse .r 'out This EG&G Idaho, Inc., report presents the results of an evaluation of an integrated system study (engineering evaluation) of the control rod drive mechanism rod position indication instrumentation for the Fort St. Vrain Nuclear .

Generating Station which was submitted to the Nuclear Regulatory Corm 11ssion (NRC) I by the licensee, Public Service of Colorado (PSC). The evaluation by EG&G' Idaho, '

Inc., concludes that PSC has not complied with the NRC directive to prepare an engineering evaluation of the problems experienced with the control rod drive rod position indicatiori because it does not adequately address the problems that were experienced at the Fort St. Vrain Nuclear Generating Station, and does not propose acceptable component replacements. In addition, the components of the rod position instrumentation that are "important to safety," are not in compliance with General Design Criteria 1 and 13 of Appendix A to 10 CFR 50.

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