ML20215L933

Evaluation of Integrated Sys Study of CRD Mechanism Rod Position Indication Instrumentation for Fort St Vrain Nuclear Generating Station, Informal Technical Evaluation Rept

ML20215L933
Person / Time
Site:	Fort Saint Vrain
Issue date:	05/31/1987
From:	Diane Jackson, Nalezny C EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML20215L911	List: ML20215L908 ML20215L933
References
CON-FIN-D-6023 EGG-NTA-7705, TAC-62198, NUDOCS 8706260218
Download: ML20215L933 (17)

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, EVALUATION OF INTEGRATED SYSTEMS i O ,

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DISCLAIMER ,

This report was prepared as an account of work sponsored by an .;

agency of the Uniteo 3tstas Government. N,either 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,

, of any information, apparatus, product er process disclosed in this report or repre'sents that its use by such third party would not infringe privately owned rights.

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EVALUATION OF INTEGRATED SYSTEMS STUDY OF CONTROL R0D DRIVE MECHANISMS R0D POSITION INDICATION INSTRUMENTATION FOR THE FORT ST. VRAIN NUCLEAR GENERATING STATION Docket No. 50-267 TAC No. 62198 1 1

INEL Reviewer - D. E. Jackson ]

INEL Program Mgr - C. L. Nalezny '

NRC Lead Reviewer - R. W. Lasky NRC FSV Project Mgr - K. Heitner i NRC Program Mgr - M. Carrington l

Published May 1987 Idaho National Engineering Laboratory EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 ,

t Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-761001570 FIN No. D6023

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 mechanism 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 control rod drive rod position indication 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. j i

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Docket No. 50-267 TAC No. 62198 ii

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FOREWORD 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, 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.

l Doc'ket No. 50-267 TAC No. 62198 iii

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CONTENTS ABSTRACT .............................................................. ii FOREWORD .............................................................. iii 3

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1.0 BACKGROUND

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2. DESIGN BASES CRITERIA ............................................. 3 1 3.0 EVALUATION ....................................................... 4 ,

1 3.1 Control Rod-Pair Full-In and Full-Out Limit Switches ............. 4 3.2 Slack-Cable Limit Switches ....................................... 6 3.3 Position Potentiometers .......................................... 6 3.4 Results of Evaluation of Proposed Control Rod Drive Temperature Limits ............................................... 7

4.0 CONCLUSION

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5.0 REFERENCES

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EVALUATION OF INTEGRATED SYSTEMS j 4

STUDY OF CONTROL ROD ORIVE MECHANISMS R00 i POSITION INDICATION INSTRUMENTATION FOR THE FORT ST. VRAIN NUCLEAR GENERATING STATION 1 l

1.0 BACKGROUND

. Following 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 j audit was to include problem areas. associated with_the June 23 scram, k

The Control Rod Drive Mechanism (CRDM) is comprised of the shim motor j 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 indication devices.

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 anomali~es included: simultaneous rod-in and rod-nut 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

., indication.

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. The results of the NRC audit were documented in a preliminary' report I which was issued on October 16, 1984, (Reference 1). The report contains-findings to be addressed both before and after p1' ant restart. The.NRC' staff noted that a number of deficiences need to be corrected on.a:long-term basis' 1 ,

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following restart. 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 (engineering evaluation) to resolve rod position indication, maintenance and 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 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 tha Fort St. Vrain Integrated Systems Study of Control Rod Drive Mechanisms Rod Position Indication instrumentation.

Criteria 1 - Quality Standards and Records. Structures, systems, and components important to safety shall be designed, fabricated, erected and tested to quality standards commenserate with the importance of the safety function to te performed.

l Criteria 13 - Instrumentation and Control. Instrumentation shall be l provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for 1

accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor co.re, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls )

shall be provided to maintain these variables and systems within prescribed operating ranges.

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3. EVALUATION ]

l 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 )

1 (EE-12-0013) and documents the evaluation.of several active and passive 1 electrical components in the RPI system to identify any potential design l>

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 j 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 j design changes were in compliance with General Design Criteria (GDC) 1 and

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

3.1 Control Rod-pair Full-In and Full-Out Limit Switches j 1

The most common problem observed with the full-in and full-out limit j switches was pitting, errosion and corrosion of the switch shafts which caused binding of the switchs which resulted in inaccurate indications in i the control room. PSC stated that the p1tting 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.

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- 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 I 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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instrumentation. The study by PSC does not include an evaluation of corrosion of the switch components, and the effect corrosion would have on l

the mechanical operation of the switches, l i

PSC proposes to replace these switches with proximity sensors which are j specially designed for their application and environment. This is an i acceptable solution if the sensors are des.igned to satisfy appropriate functional, and operational requirements, and are qualified for the environment in which they must operate. However, the statements " designed for FSV's application and environment" and " capable of operation at several hundred degrees Fahrenheit above the requirements" are vague and do not adequately specify the operating conditions and functional requirements for the design, fabricaton, and procurement of the sensors. l 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 (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 l

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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 the fact that the stainless steel targets will damage the potentiometer shafts when the system is overdriven.

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Since the analog position indicators are "important to safety," the f out-limit cam must be designed so that overdriving it will not damage the  !

potentiometer shaft, or so that they can not be overdriven.

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 slack-cable limit switches revealed no known failures and the switch j manufacturer confirmed this to be an acceptable application of the switch.

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 licensee. The statement that the manufacturer confirmed this to be an acceptable application of this switch is not acceptable because it does not resolve the question of why the switch failed. The engineering evaluation should have included a comparison of the design requirements for the switch with the known operating conditions to determine if the failure was design, fabrication, or maintenance related. Therefore, it is concluded that PSC has not performed a satisfactory engineering evaluation of the slack-cable j limit switches and has not proposed any solution to correct the problem.

3.3 Position Potentiometers l 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 the 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 potentiometers are to be built with a 10 turn electrical section centered on 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 0, of Ref. 2. The proposed r_eplacement potentiometers are an improvement. However, because they provide the operator with continuous position information for all the

• rods during operation and following a scram, they are considered "important to safety." Therefore, they must be designed, fabricated and procurred to 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 with GDC 13. For instance, the requirements for a quality assurance program were not called out. Environmental conditions such as minimum and maximum temperatures, and maximum moisture content of the Helium atmosphere were not specified. If PSC procures commercial grade potentiometers for this 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 NRC 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). Refarence 4 is.an evaluation of a PSC oroposal to increase the operating temperature limits of the FSV Control Rod Drive and Orifice Assemblies (CRD0As) to 3000F. The findings and deficiencies identified in Reference 4 are generally applicable to the CRDM

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rod position instumentation. 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 CROM 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.

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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. j

3. PSC did not address maintenance of RPI instrumentation, i

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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 relacement targets for the full-in/ full-out limit switches did not consider 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 rep ~1acement rod position )

potentiometers did not include all the environmental conditions that-the- l components could be exposed to and did not define how the potentiometers-would be qualified. In addition, the proposed replacement instrumonts that 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 Continued Operatin of Fort St. Vrain Nuclear Generating Station,  !

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

2. PSC Letter, Warembourg to Berkow, Fort St. Vrain Rod Position Instrumentation Integrated Systems Study, (P-86522) dated j 1

August 15, 1986. ]

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3. U. S. Ncclear Regulatory Commission, Standard Review Plan, NUREG-0800, Rev. 2, July 198)..

4. NRC letter, Heitner to Williams, Control Rod Drive and Orifice Assembly i

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3000F Temoerature Limits Fort St. Vrain Nuclear Generating Station, G-86664, December 24, 1986. ,

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3. TITLE AND SUSflTLS JLEAV88 BANE Evaluation of Integrated Systems Study of Control Rod Drive Mechanism Rod Position Indication Instrumentation "'""'" " ' " " ' ' ' " "

For the Fort St. Vrain Nuclear Generating Station o rN nAa m,,,o,,,,, ,

May l1987 8 oArg pt,oni assuto D. E.' Jackson McNrN nAa C. L. Nalezny .. May l1987 7 ERsoRMING oRGANIZAT. oft NAa.4 ANQ MAILING AooRtSS (Jsetswe ge Goed B. (*MoJ4CTIT A&E/ WORE UNs T Num.StR INEL-EG&G Idaho-Idaho Falls, ID 83415 . t eiN oa caA r Nv .a D6023 -l

. J ta spoNsoneNG ORGAN 12ATioN NAME ANO MAlpNQ AoomtS5 isassverle Camps 16 Tvetofmeroni Division of Reactor Projects IV ' Informal l Office of Nuclear Reactor Regulation . ,...oo cov.R.o ,, e,.- ,

! U.S. Nuclear Regulatory Commission Washington, D.C. 20555 12 SUPPLEMENT ARY Noit5

, a 11 ASST R ACT (Jnl0 e. ores or s.ess l

This EG&G Idaho, Inc.," report presen.ts the results of an evaluation of

., an integrated system study (engineering evaluation) of the control . rod drive .

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mechanism 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, j

-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 3 position indication because it d.oes not adequately. address the problems that were  ;

experienced at the Fort St. Vrain Nuclear Generating Station,_and does not propose j 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 1d CFR 50.

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