ML20154A545
| ML20154A545 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 04/28/1988 |
| From: | Henderson J PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML20154A543 | List: |
| References | |
| EE-12-0013, EE-12-13, TAC-61601, NUDOCS 8805160045 | |
| Download: ML20154A545 (72) | |
Text
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/ Public FORT ST. VRAIN NUCLEAR GENERATING STATION QJ)/ Serviceo rusuC SERVICE COMPANY OF COLORADO I I EE-12-0013-REV. B INTEGRATED SYSTEMS STUDY OF THE CONTROL R0D DRIVE MECHANISM R00 POSITION INSTRUMENTATION [, h,,' Prepared by: b -ul ~ dw k #M (7 pt rewcoa 4%W
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FO;T ST. VR AIN NUCLEAR GENERATING STATION Pubil'" PUBLIC SE .VICE COMPANY OF COLORADO ett C*- C W **
- O Service' CHECK LIST OF DESIGN VERIFICATION BY u m gar PAGE QUESTIONS FOR DESIGN REVIEW METHOD YES NO N/A b 1. Were the inputs correctly selected and incorporated into design?
y[] 2. Are assumptions necessary to perform the design activity adequately described and reasonable? Where necessary, are the assumptions identified for subsequent re verifications when tht, detailed design activities are completed? EU 3. Are the appropriate quality and quality assurance requirements specified?
- 4. Are the applicable codes, standards and regulatory requirements including issue and addenda
%] ] ] properly identified and are their requirements for design met?
N U S. Have applicable construction and operating experience been considered? UUE 6. Have the design interface requirements been satisfied? N UU 7. Was an appropriate design method used?
\] U 8. Is the output reasonable compared to inputs?
UU 9. Are the specified parts, eqd c n\ ent, and processes suitable for he required application?
- 4. W 9 "W .
- 10. Are the$ pretogpa5 specified materials compatible with each other and the design environmental conditions
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to which the material will be exposed? EU 11. Have adequate maintenance features and requirements been specified? Are accessibility and other cesign provisions adequate for performance of needed maintenance
)) 12.
and repair?
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- 13. Has adequate accessibility been provided to perform the in-service inspection expected to be required during the plant hfe?
UU 14. Has the design property considered radiation exposure to the public and plant personnel? R[%
- 15. Are the acceptance criteria incorporated in the design documents sufficient to allow verification that design requirements have been satisf actority accomplished?
Have adequate pre-operational and subsequent periodic test requirements been appropriately
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specified? %% g g,3 g %g p, g 6 UU] 17. Are adequate handling, storage, cleaning and shipping requirements specified? 40T
*-tF*d. For p rot n b - 18. Are adequate identification r%945equirements specified?
4 U 19. Are requirements for record preparation review, approval, retention, etc., adequately specified? NOTE: If the answer to any question is no, provide additionalinformation and resolution below. RESOLUTION OF DESIGN DEFICIENCIES UNCOVERED OURING THE DESIGN VERIFICATION PROCESS M 86 9 8' ck CLA d 9" O A 9 6abmL'vs. h % Q d r* s 3 Q Q @
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G Public FORT ST. VRAIN NUCLEAR GENERATING STATION EE-12-0013 Rev. B l D Serviceo rueuc senwce couram or cotonAno TABLE OF CONTENTS Page No. l 1.0 PURP0SE...................................................... 1 i 2.0
SUMMARY
....................................................... I ; 3.0 SC0PE........................................................ 1 4.0 APPROACH .................................................... 3 5.0 EVALUATION .................................................. 5
6.0 CONCLUSION
.................................................. 11
7.0 REFERENCES
.................................................. 13 f Appendix A Sketches Referenced in Section 3.0 Appendix B Photographs . -
Appendix C Notes / Letter Regarding NRC Staff Visit of 12/04/87 l Appendix 0 Anomalies Chart . Appendix E FMEA of Position Transmitters ! Appendix F FMEA of In/Out Limit Switches Appendix G FMEA of Slack Cable Switches Appendix H Specifications for New Components ; Appendix I Maintenance Information i Appendix J Failure Study of RPI Components from Conynercial Operation to Date i Appendix K FMEA of New Designs Appendix L REF 1 (Preliminary Report Related to the Restart and Continued Operation of Fort St. Vrain Nuclear GeneratingStation(excerptfrom)) ! Appendix M REF 2 (Evaluation of Integrated Systems Study of Control Rod Drive Mechanism Position Indication A Instrumentation for the Fort St. Vrain Nuclear T GeneratingStation) ' i i t ronu M4. U 5317 ,
Public Fon7 ST, VRAIN NUCLEAR GENEAATING STATION E 2-0013 G$9fVICO' Pusuc SERVICE COMPANY OF COLORADO s - I t ! 3.1 The ins trumentation evaluated included (reference Appendix A, sketches 1, 2 and 3): 9 a 3.1.1 Position transmitter: potentiometer,
- dual gang, 1 10 turn,1000 ohm 1 each per rod pair i 3.1.2 Full-in/ full-out position: switch, limit, double throw, single pole, 28vde, 10A, 2 pairs of 2 each per rad pair ;
3 3.1.3 Slack cable condition: switch, limit, double .
- throw, single pole, 28vde, 10A, 2 each per rod pair ,
3.1.4 Full retract position: switch, limit, double , throw, single pole, 28vde, 10A, 1 each per rod ! cable. (2 cables per rod pair) !
*Two electrically independent potentiometers on a comon shaf t !
j 3.2 Ancillary equipment evaluated included (reference Appendix
- A, sketches 4, I and z)
3 3.2.1 Orive shaft (potentiometer shaft): drives [ l potentiometer directly and cam drum indirectly. 1 , 4 each per rod pair - 3.2.2 Cam drum: moves cams relative to the full-in/ full-out limit switches, 1 each per rod pair 3.2.3 Cams: actuate full-in/ full-out limit switches, 2 a pairs of 2 each per rod pair ; j !
- 3.2.4 Calibrated counter balancin springs
- actuate slack cable switches upon oss of weight of
- control rod on springs l
3.3 Indication methods evaluated included: ) US.1 Full-in, full-out and slack ce le limit lights; ; i n. color coded, powered thru contact on relays (wired i fail safe), 1 each per condition per rod pair 3 3.3.2 Position indicator; analog, voltage based, r indication per position transmitter current output ' (part 1), 1 eaci per rod pair l 1 l l ! j i J i i seaw in we.n wt j
t Public FORT ST. VRAIN NUCLEAR GENERATING STATION EE-12-0013 SerVIC9 0 PUBLIC SERVICE COMPANY OF COLORADO 3.3.3 Position indicator; digital, voltage based, indication per position transmitter current output (part 2),1 set (selectable by rod group) 3.3.4 Relays; energized thru limit switches, provide power thru contact to indicating lights, 1 each per condition per rod pair 3.4 Associated passive components evaluated included: 3.4.1 Power supply: voltage, for limit switch circuits, 24vde, 1 each 3.4.2 Power supply: current, for all posi tior, transmitters,1 each 3.4.3 Cables, connectors, terminal lugs, terminal blocks, solder connections; as required 3.4.4 Surge protectors: intended to short high inductances caused by relay de-energization to negat .3 side of power supply instead of allowing it thi switch contacts, 1 per relay 4.0 APPROACH The following is the methodology used in the evaluation of RPI. Each component is evaluated separately unless multiple components and their actuation mechanisms are sufficiently similar to allow typical analysis. In such cases, the basis for similarity will
- be established within the evaluation.
I The general evaluation included: 4.1 Review existing data 4.1.1 Review reference 1 ar.d 2 to determine NRC concerns 4.1.2 Review available design and maintenance Ji documentation
. ):
4.f3 Interview knowledgeable personnel l FORM (A) W 22 $317
i Public FORT ST. VRAIN NUCLEAR GENERATING STATION EE-12-0013 service *' PUBLIC SERVICE COMPANY OF COLORADO 4.2 Gather new data 4.2.1 Disassemble components and make observations regarding the nature of failures (if any are known), types of damage and other pertinent information for future evaluation 4.2.2 Phot 39raph components for future evaluation (Reference Appendix B) for excerpts of photos , The component specific evaluation included: 4.3 Develop each component's Failure Modes and Effects Analysis . (FMEA) 4.3.1 Based on data obtained in 4.2 develop a conceptual FMEA for each component 4.3.2 Consider all data obtained in 4.1 and modify conceptual FMEA as required 4.3.3 Review materials and methods of manufacture of each component and modify conceptual FMEA as required. 4.3.4 Consult appropriate manufacturer's engineering and analytical organizttions for evaluation of conceptual FMEA and concurrence (or rejection) of modes and credible causes of failure 4.3.5 Prepare final FMEA 4.4 Identify any component or system weaknesses 4.4.1 Based on FMEA, determine what impact any weakness has on component opera tion , reliability and mechanisms for failure 4.5 Evaluate operability of components ! 4?841 Based on FMEA, determine level of component
"> operability and reliability 4.6 Evaluate causes of failure 4.6.1 Based on FMEA, determine causes of failure l
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Public FORT ST. VRAIN NUCLEAR GENERATING STATION EE-12-0013 OSerVlCe* PUBLIC SERVICE COMPANY OF COLORADO 4.7 Evaluate operational adequacy of the existing system 4.8 Propose interim solutions 4.8.1 Based on causes of failure, propose interim solutions to limit failures by removing as many causes of failure as possible 5.0 EVALUATION This evaluation is based on the following inputs:
- a. The scope of the evaluation is presented in section 3.0 of this study. Exceptions within the specified scope are documented and justified as appropriate.
- b. The approach of the evaluation is presented in section 4.0 of this study. Each component or assembly is evaluated on a case by case basis. Failure mode and effect analyses will be presented within the body of the study with references to appendices which contain
- the technical analysis including the failure tree analysis. An occasional reference by the reader to the appendices containing sketches (A) and photographs (B) will provide additional insight into the assembly or component being discussed.
5.1 A review of existing NRC docuraentation provides many questiens, concerns and an occasional misconception. Table 5.1 will provide a source reference, a basic issue of concern to PSC (from NRC documentation) and a reference to a section of this study. This table should provide the reader immediate access to any specific area of interest. [ [ FORM I AI M4 D M11
a Public FORT ST. VRAIN NUCLEAR GENERATING STATION EE-12-0013 O $9fVlCe* PUBLIC SERVICE COMPANY OF COLORADO
- TABLE 5.1 Reference Page/ Para /Line Issue RPl*
G-84392-REF1 P3-3,p4,L12 failure mode of pot shaft C,E (Appendix L) same P3-3,p4,L14 cam over rotation C,E same P3-3,p4,L16 pot failure modes E same P3-3,p4,L24 single point failure 5.2.6 same P3-4,p3,L1 circuit redundancy 5.2.5 same P3-4,p3,L5 single failure undetected 5.2.5 same P3-6, sect. 3.4 failure modes defined 5.2 E,F,G same P3-7,p7,L1 preclude overtravel damage C,0 same P3-8,p2 integrated system study 1.0 same P3-9, table 3.1 anomalies D UTB7237-REF2 P3,p2 and 3 criteria not applicable 3.0p1 (Appendix M) same P4,p3,L1 failure mode of switches 5.2.1 F same P4,p3,L6 moisture causes pitting 5.2.1 F same P5,pl,L1 corrosion not studied F----
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scme "P3,p2 L5 design environment H same P5.p3,L3 important to safety 3.0p1 same P5,p3,L3 design criteria 1 and 13 3.0pl same , P5,p4,L1 cam damages pot shaft C same ~d; "EG,p4,L9 pot replacement necessary C same P6,p3,L2 slack cable anomaly 5.2.2 same P8,p4,L1 maintenance I same P9,p3,L1 inadequate eval of failure 5.2,A BOEFG FORM IA) )M D E317
i PubilC FORT ST. VRAIN NUCLEAR GENERATING STATION EE-12-0013 Serv lCO* PUBLIC SERVICE COMPANY OF COLORADO
- Numbers represent sections of this report.
Letters represent appendices. 5.2 Evaluation Existing RPI 5.2.1 Evaluation of Full-In/ Full-out Limit Switches The evaluation began by researching the application of the limit switches that provide the indicating signals for full-in or full-out position of each rod pair. These limit switches are evaluated simultaneously since the switches, actuation methods and circuitry are identical. The problem identified was binding of the switches. This results in the full-in or full-out indication remaining energized after the rod pair is moved away from the corresponding position. Disassembly of the switches revealed pitting and gouging of the switch shaft and the deposition of the gouged materials onto the switch shaft housing. Further investigation into the cause of the failure revealed the exertion of excessive lateral forces onto the switch shaft by the actuating cam. This was verified by inspection of the disassembled switch, as the degradation occurs primarily on the side of the switch shaft opposite the cam approach. The manufacturer's review of the application confi rmed this evaluation. Additionally, the switch mancfacturer indicated the balance of the switch to be "in quite satisfactory condition". The manufacturer took specific interest in the condition of the electrical contacts which were found to be in good condition. The technical presentation of the failure modes and effects analysis including the failure analysis tree is presented in Appendix F.
.i.-
FORM (M 344 2M317
Public FORT ST. VRAIN NUCLEAR GENERATING STATION EE-12-0013 SerVlCe# PUBLIC SERVICE COMPANY OF COLORADO 5.2.2 Slack Cable Limit Switches The slack cable limit switches provide the indicating signals for the control rod pair in the event of a broken cable or a damaged control rod. The principle of operation is very simple. When the rods are suspended by the drive mechanism the limit switch is actuated, resulting in no indication to the control room. In the event of a broken cable or other failure in which weight is removed from the drive, a precision calibrated spring provides a force on the actuating plate. This action results in the change of state in the switch and thus indication in the control room. Contrary to Table 3.1 of reference 1, there has never been a known failure of the slack cable limit switches. The case cited in reference 1 (Table 3.1) was a realistic case of slack cable indication as the cable had broken. The manufacturer's review indicated proper application of the switch. The technical presentation of the failure modes and effects analysis including the failure analysis tree is presented in Appendix G. 5.2.3 Rod Retracted Limit Switch This switch is not evaluated in this study as it is not a part of the normal control scheme. Instead, it is a maintenance feature to indicate when the rod is properly withdrawn for extraction of the drive assembly from the core. The failure modes for these components are comparable to the full-in/ full-out limit switches, however, they
- have no impact on reactor operation. Therefore, no specific analysis exists. These switches are typically replaced on the same schedule as the full-in/ full-out and slack cable switches.
b I l l l l FORM (Al 344 22 $317 i
Public FORT ST. VRAIN NUCLEAR GENERATING STATIONEE-12-0013 O SerVICO* PUBLIC SERVICE COMPANY OF COLORADO I 5.2.4 Position Transmitters The position potentiometers provide two independent and continuous signals for indication of control rod position. Each of the position transmitters completes a current loop from which a voltage drop provides indication on both digital and analog meters. Each section of the potentiometer is evaluated as a separate component as appropriate. Any single failure which presents a dual impact is identified as such. The most common failure mode is breakage of the potentiometer body, coupling or wiper due to overdriving. A secondary possible failure mode is galvanic action. While discoloration (darkening) of the resistive element has been observed, no confirmed account of copper oxide production exists. The technical presentation of the failure modes and effects analysis including the failure analysis tree is presented in Appendix E. 5.2.5 Indicating Methods The methods utilized to indicate rod position in the control room were reviewed. They include: (1) 37 sets of full-in, full-out and slack cable indicating lights (2) 37 analog gauges for position indication and (3) a series of selectable digital meters for recundant position indication. The indicating lights are low voltage incandescent units wired to energize when the condition is to be indicated. The analog and digital gauges are i voltmeters scaled to indicate position. All l position indication methods (continuous) use a constant current source to provide a varying voltage dependent on potentiometer resistance which varies with rod position. All limit switches, potentiometers and rod position indicators are redundant. However, one indicating _ light is provided per pair of limit switches and are therefore not redundant, but are tested at i regular interval s . No concerns were identified relative to indication. FORM I A) 344 - 22 5317 _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - J
Public FORT ST. VRAIN NUCLEAR GENERATING STATION EE-12-0013 ev. B SerVICO' PUBLIC SERVICE COMPANY OF COLORADO 5.2.6 Integrated Component / System Review The movement of the control rod pair (by its drive motor in the inward and outward direction normally or by gravity with capacitive motor speed control during scram) is by a cable connected to a rotating drum. As the drum rotates, the rod pair is moved. This drum, through gearing, is coupled to a shaft which drives a cam wheel for full-in/ full-out switch actuation and to a potentiometer for position indication. This common shaft is a single failure point. A review of fai' lures indicates that prior to administrative limits (see note 1) being established, this comon shaft in a few instances has sheared resulting in loss of both limit switch and continuous RPI indication. However, subsequent to the establishment of administrative limits, no known failure has occurred. Note 1: Administrative control has been establisned in P-85242, 7-10-85 (interim rod indication technical specification) and G-85294, 7-23-85 (supplemental safety evaluation); incorporated into plant procedure 50P-12-01. The full-in/ full-oJt limit switch failures prior to 1985 refurbishment (Raf. Appendix J) appear to be from excessive operations and wear (from the testing phase). Further, there exists some probability of operator error during June 1984 due to surveillance "testing" of the rods for "full-in" position. Several full-in switches were known to have failed completely, along with others failing intermittently. As a result, the operators (prior to improved training) were likely to attempt to "overdrive" the rods, as observed by the NRC during their June 1984 visit. This doubtlessly contributed to the number of potentiometer failures during this period. An overall failure rate of .005 (Ref. Appendix J) shows operability of the existing system. This does not influence PSCo's desire to upgrade limit switch and continuous RPI design for improved maintainability and reliability. FO A M ' Al 344 22. M17
.I Public FORT ST. VRAIN NUCLEAR GENERATING STATION EE-12-0013 OSOfVlCO' PUBLIC SERVICE COMPANY OF COLORADO *
6.0 CONCLUSION
S 6.1 General 6.1.1 Limit Switches A review of the failures occurring after FSV began commercial operation (Appendix J) indicates operability of the existing system. However, based on the FMEA and operational experience, mechanical operation of the components for an extended time period may result in their failure. In the case of the full-in/out switches, actuation of the limit switch plungers by . cams results in mechanical wear of the switch shafts. One solution to this problem would be to replace the existing cams with cams having a less severe slope. This would but leaves the root cause resultofinfailure improved unchanged reliability,(mechanical ope ra tion) . An improved design would use a sensor that has no moving parts. Such a design in its conceptual stage including FMEA is shown in Appendix K. 6.1.2 Continuous Positien Indication For the potentiometers, an increase in the number of turns -(mechanical) will eliminate the concerns associated with "yo-yoing". Furthermore, if an operator was to overdrive the rods in error, the new component will not fail at the point where the existing component does. Improvements in the areas of materials will eliminate all other possible failure modes. This design, including FMEA, it shown in Appendix K. Both the sensors and potentiometers have been prototyped and are available for testing. row w w.n.un
Public FORT ST. VRAIN NUCLEAR GENERATING STATION EE-12-0013 ev. B SerVlCO PUBLIC SERVICE COMPANY OF COLORADO 6.2 Specific 6.2.1 Potentiometers 6.2.1.1 Potentiometers can be broken by overdriving (operator error) or by yo-yoing (mechanical rebounding of the rods as they reach the end of travel ) . This has been observed prior to and after FSV went into comnercial c;;t2 tion. The existing administrative lim'ts (Ref. S0P 12-01) do not allow "overdriving". This has resulted in zero failures subsequent to the 1985 refurbishment. To upgrade the potentiometers to the proposed design can only improve their reliability. 6.2.2 Full-in/out limit Switches 6.2.2.1 The proposed replacement sensors will provide a virtually maintenance free RPI system. These non-contact sensors are capable of function in temperatures up to approximately 500 degrees Fahrenheit at which point the solder connections melt. This limitir.g factor will not be of concern as the maximum application temperature is approximately 300 degrees Fahrenheit. 6.2.3 Slack Cable Switches 6.2.3.1 No failure modes are known to exist. The manufacturer's review indicates this to be a proper application of the switch. 6.2.4 System 6.2.4.1 The existing RPI system is operational, within Technical Specifications limits, and has operated without failure subsequent to the 1985 refurbishment. During the 1985 refurbishment, all limit switches and potentiometers were replaced. After 4.5 million plus hours of availability (Ref. App. J), this system is concluded to be acceptable for continued operation. However, for improved maintainability, it is desirable to proceed with component upgrades consistent with the overall CRDM temperature requalification plan and the CRDM maintenance schedule. voeu a> >u n un
i Public FORT ST. VRAIN NUCLEAR GENERATING STATION EE-12-0013 OSefVICO#.PUBLIC SERVICE COMPANY OF COLORADO
- TABLE 6.1 Sumary of Component Disposition Full-In Limit Switches Recommend replacement with non-contact sensors Full-Out Limit Switches Recomend replacement with non-contact sensors Slack Cable Limit Switches Leave As-Is; No credible failure modes Full Retract Limit Switch Leave As-Is; Not part of control circuitry F5s1 tion Potentiometer Recomend replacement with a unit having more turns and improved design
7.0 REFERENCES
7.1 Reference 1 Letter Denton to Walker Dated October 16, 1984 and enclosure titled "Preliminary Report Related to the Restart and Continued Opuation of Fort St. Vrain Nuclear Generating Station" 7.2 Reference 2 Letter, Heitner to Williams Dated July 31, 1987 and enclosure titled "Evaluation of Integrated Systems Study of Control Rod Drive Mechan' sm Rod Position Indication Instrumentation for the Fort St. Vrain Nuclear Generating Station" l I i l i i l l FOAM (Al 344 22 5317 L
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E 2-0013 FORT ST. VRAIN NUCLEAR GENERATING STATION OPublic SerVICO* PUBUC SERVICE COMPANY OF COLORADO APPENDIX B PHOTOGRAPHS FOmM (An 344 ,22 5317
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- AGE UNKNOWN
- EE-12-0013 Public FORT ST. VRAIN NUCLEAR GENERATING STATION SerVlCe' PUBLIC SERVICE COMPANY OF COLORADO APPENDIX C NOTES / LETTER REGARDING NRC STAFF VISIT OF 12/04/87 4
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NCIES Prior to the meeting of Dec. 1, 1987 a tour of the CRDM drive assembly was crovided to the NRC staff. It should be noted that scecific areas of concern to the staff were reviewed while observing the aceration of the CRDM test set-uc. One scacific concern was the striking of the cotentiometer shaft by the cams which presently operate the switches or by the targets which are proposed for use with the new sensors. By visually observing the aceration of the cam drum adjacent to the potentiometer shaft on a partially disassembled unit, it was apcarent that the targets (which are as large as the cams) do not strike the cotenticmeter shaft. This point is reiterated by the staff letter of Dec. 15, 1987 Page.1, paragraph 9, last line. i 1
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', UNITED STATES
.- j NUCLEAR REGULATORY COMMISSION
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' t rv ASHING 7oN. O. C. 20684 t Cecember 15. 1987
\ ,,,,,- g g g -2 ^ $ 7 Docket No. 50-267 8 *
.9 e MEMORAN00M FOR: Jose A. Calvo, Director ,
Project Directorate - IV . - : Division of Reactor Projects - III, *
. 1.E ..
IV, V and Special Projects , .g ", ; FRCH: Kenneth L. Heitner, Project Manager . S' ' Project Directorate - IV .
...d Division of Reactor Projects - III, E .f IV, V and Special Projects .L~
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SUBJECT:
SUM 4ARY OF MEETING TO DISCUSS CONTROL R00 ORIVE 9 7 ~ .": , .. MECHANISM (CRDM) TEMPERATURE REQUALIFICATIOh "Jm . AMD POSITION INSTRUMENTATION, DECEMBER 4, 1987 (TAC NOS 61601 AND 62198) The purpose of this meeting was to discuss the licensee's latest proposals for these licensing actions. The licensee's previous proposals for these areas have been evaluated by the staff in letters dated Decevaber 24, 1986, and July 31, 1987. The attendees at this meeting are listed in Enclosure 1. Pocition Instrumentation The licensee presented a preliminary review of the current control rod positier instrumentation. The presentation material is in Enclosure 2. The presenta-tion included a failure mode analysis of the present instrumentation. This analysis showed that the current rod-in and rod out limit switches were strong candidates for replacneent. The problems with the current design of the rod position potentiometers were also highlighted. The licensee stated that improved instrumentation had been selected as follows:
- The rod-in and rod-out limit switches would be replaced by proximity sensors, and
- The rod position indication potenticoster would be replaced by an improved potentiometer.
Both items would be able to withstand the temperature, humidity, radiation, and pressure environment of the CRDM. The licensee stated that their intent was to submit a new failure mode analysis l report. This report would clarify that there was corrosion failure of the I rod-in and rod-out limit switches. It would also indicate that there were no credible failures (and need to replace) the slack cable limit switches. There was also not a problem with the proximity switch targets nochanically inter-faring with the position potentiometer shafts. The staff noted that the licensee had not addressed other important criteria for the new instrumentation. The staff stated that the new instrumentation should clearly have service life equal to or exceeding that of the current instrumentation. Otherwise, the reactor operation could be constrained by inoperable control rod instrumentation. (The staff also notes that other factors, such as accuracy and long ters drift, should be cons;dered.) c, n Q j i . " '- / L2-]Il 95, cp
l l CROM Temperature Recualification i'he licensee reviewed the key steps in his proposed program for CROM temperature requalification (see Enclosure 3). They are as follows: Evaluate tne projected maximum CROM service temcerature. The licensee has developed a methodology for correlating tne CROM motor temcerature with reactor temperatures and coolant flow parameters. These include the orifice position settirig for each refueling region. Projected maximum temperatures for Cycle 4 are about 300*F in Region 12 and 275' f in Region 30. However, the licensee noted that some reductions could be achieved by reducing overall core flow and deliberately mismatching region outlet temperatures througn orifice adjustments (within the limits of LCO 4.1.7). There was also some uncertainty about the validity of tne correlations at nign power, which could result in lower temperatures. Some possibility existed that CROM temperature requalification would not be needed. Evaluate the CRCH system for acceptability at the higher temperature. This would include an evaluation of the materials of construction at higher temperature. The methodologies would be similar to those used to "qualify" materials under 10 CFR 50.49. Specific accepted methodologies would be used for special cases, such as CROM drive motor insulation. Perform an integrated systems test of an entire CROM assembly. The purpose of this test would be to evaluate potential synergistic and system effects of increased temperature, and Evaluate the role of the preventive maintenance program in assuring system performance in tne event the "qualified" component life is reduced by the elevated temperatures. This could iriciude more frequent maintenance of CROMs exposed to higher temperatures Future Actics The licensee proposed to submit a proposal for new red position instrumentation in April 19%, and for CROM temperature requalification in March 1983. The staff commented that the licensee should give careful attention to applicable and related regulatory requirements and guidance. This included: 10 CFR 50.62 GL 83-28 and related licensee submittals and staff evaluations, SRP 3.11, 4.5.1, and 4.6, The October 16, 1984 Assessment Report, and Subsequent Safety Evaluations on the CRDMs and CROM position instrumentation.
&' M Kenneth L. Heitner, Project Manager Project Directorate - IV Division of Reactor Projects - !!!,
IV, V and Special Projects
Enclosure:
As stated cc: See next page
Fr. R. O. Williams, Jr. Public Service Company of Colorado Fort St. Vrain CC* Ma. D. W. Warembourg, Manager Al',ert
. J. Hazle Director Nuclear Engineering Division Radiation Control Division Public Service Company Oecartment of Health of Colorado 4210 East lith Avenue P. O. Box 840 Cenver, Colorado 80220 Derver, Colorado 80201-C840 Mr. David Alberstein, 14/159A Mr. R. O. Williams, Jr. , Acting Manager GA Technologies, Inc. Nuclear Production Division Post Office Bca o6608 Public Service Comcany of Colorado San Diego, California 92138 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Mr. H. L. Brey, Manager Nuclear Licensing and Fuel Division Mr. P. F. Tomlinson, Manager Public Service Company of Colorado Quality Assurance Division P. O. Box 840 Public Service Company of Colorado Denver, Colorado 80201-0840 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Senior Resident Inspector U.S. Nuclear Regulatory Commission Mr. R. F. Walker P. O. Box 640 Public Sr:rvice Company of Colorado Platteville, Colorado 80651 Post Office Box 840 Denver, Colorado 80201-0840 Kelley, Stansfield & 0'Donnell Public Service Company Building Comitment Control Program Room 900 Coordinator 550 15th Street Public Service Company of Colorado Denver, Colorado 80202 2420 W. 26th Ave. Suite 100-0 Denver, Colorado 80211 Regional Administrator, Region IV U.S. Nuclear Regulatory Consnission 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Chairman, Board of County Coaunissioners of Weld County, Colorado i
Greeley, Colorado 80631 l Regional Representative l Radiation Programs
- Environmental Protecticn Agency 1 Denver Place 999 18th Street, Suite 1300 Denver, Colorado 80202-2413 t
i
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Enclosure 1 ATTENDEES NRC/PSC Meeting December 4, 1987 NAME ORGANIZATION Rick Burrows PSC/ Plant Engineering Ed Pitchkolan PSC/Special Projects M. E. Niehoff PSC/NED Greg Bates PSC/NED Jim Henderson PSC/NED J. R. Reesy PSC/NED J. L. Mauck NRC/NRR/ICSB Ken Heitner NRC/NRR/PD-IV Sam Chesnutt PSC/ Licensing I M. H. Holmes PSC/ Licensing C. Bomberger PSC/ Licensing Don Warembourg PSC/NED Frank Novachek PSC/NPD : i I
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I E\C. T.:I : ' Normal Operating Parameters 1 - Cable length % 240" from Fully Retracted to Fuli-In Lim.it. 2 - Cable length $ 190" from Full-Out to Full-In Limit. 3 - From Fully Retracted to Full-In ~~ 10 turns on 2-gang potentiereter. 4 - At Full-In 2 each limit Switches are actuated - wirec in series. Actuated by Cam. 5 - At Full-Out 2 each Limit Switches are actuated - wired in series. Actuated by Cam. 6 - On loss of a Rod (Cable Breaks) 2 N/C Limit Switches are actuated ; by counter balancing spring. I h I i i l l l I
All CRDM Position Associated Other Circuitry Instrumentation Anomalies from NRC Inspection Indication Components Based on Described Component (s) failure CRDM No./ Region Anomaly I - __ 1 -____.__________________. i
- 1) Rod-In Limit Switches Indicating Light Cable Connectors, 13/19 Faulty Rod-in Limit Switch, incorrect (Control Room (CR)) lerminal Lugs Blocks CR analog & digital indication (See 5,6)
Relays Surge Protec- 39/23 Incorrect CR Analog position indication, tion (Diac). Power Faulty Rod-In Limit Switch (See 5) Supply
- 2) Rod-Out Limit Switches Indicating Light Same as Above 6/1 I of 2 rod-out limit switches inoperable (CR) 26 faulty Rod-out switch a
29/ESW5 Faulty Rod-out switch 37/3 Faulty Rod-out switches a j 3) Rod Retracted Limit Indicating Light Cable Connectors, q Swi tches (Test Box) lerminal Lugs, Blocks
- 4) Slack Control Rod Cable Indicating Light Cable Connectors, 25/7 Faulty Slack-Cable Switcha l Limit Switches (CR) Terminal Lugs, Blocks, i Relays Sury Protec-
{ tion (Diode), Power { Supply J l 5) Position Analog Meter Cable Connectors 12/15 Incorrect CR analog position indication . Potentiometer (CR) Terminal Lugs, Blocks,13/19 Faulty Rod-in Limit. Switch, Incorrect j (2 Gang) Part 1) Power Supply CR analog and digital indication (See 1,6) 33/16 faulty analog position indication ! 39/23 Incorrect CR analog position indication. { Faulty rod-in limit switch (See 1)
- 6) Position Potentiometer Digital Meter Same as Above 13/19 faulty Rod-In Limit Switch, incorrect (2 Gang) Part 2 (CR) CR analog and digital indication (See 1,$)
4 Note - A known cable failure occurred in j this position; therefore. the (on-clusion of a f aulty slac k-t abit- switih is unfounded. i
wt FRILURE ANRLYSIS TREE (% cye.cTC 1 (?)~ scRT; U " N aCT'. ATE:: i i s.::c :- SCM. EN (i) m('a SCA. CLOSE: SMT 14* GOLL b oTwa actLca contacis l l
-EcHw!ca. CLtcTRIca.
t ru.L-Iwru.L-out LIMIT s. ITCH i i _ NOTES rmIttpc meg trrCCTS ANR.YS!3 1 - NOT R rMILUE (INTENDED A - INDICATING LIGHT ILLt.NINATCO, R NCT!ON) RCP91NS So VTER t.0S$ OF ST!M.A.LS 2 - NOT M CMCDIR.C FAILLRC McCC 9 - PTCHNICR. FAILtJIC 7 AN (NQ CIMCUIT MILSELC) LMRCDICTAE.C PeTURC WILL OCCut 3 - NOT A CREDIBLC rAILutC MODC C - No INDICATING LIGHT ILLud! PATED l (BRSCD CN MRTCRIR.S 383&44OSS3 f.PON ACTl,RT!CN - PCT BY JUDCCTN7 l 4 - NOT, SY JUDGCTNT, A CRCDIR.C A CRCljlR.C FAILUSC - $CC '4 ' I rAILust McCC (SMP ACTION O - INOCTERMIPeTC - A CA 9 IPSCO SWITCH) SCC 'C' CN POCCISC Po!NT & JM 5 - NOT A C9CDIR.C rAILt.MC MODE (IPSCD CN $TNLCTutM. 50 MCCHNICM. AUGGC2(SS OF l SWITCH) \ - _ -- . _- _- _ _ _ _
Based on Failure Analysis Wnat Can Fall ? Exclain anat would cause 'ailure. Pg. 1 - A - Full-In/ Full-Out Switch fails mechanically by jaming tre shaft in the completely actuated position. Cause - Shaft Failure Investigate cause of shaft failure.
Conclusions:
Shaft fails due to hign lateral forces causec by Cam angle >15' ( recomended max per Mfr). Failure is by erosion, deposition of eroded metal, corrosion until space between snaf t and heuting is filled and will not allow free movemetit . To verify: (1) Contact manufacturer for concurrence; (2) Check for cor-os t on/e rosion/deposi tion on only 1 side of shaft (side opposite Ctm approach) Verify as required. . P g . 1 - B - S ame a s " A " above, except position at which final failure occurs.* Cam may fail.' ; Pg. 1 Same as "A" above, except position at which final failure occurs. Pg. 1 - C - Full-In/ Full Out Swi tch fails electrically by sealing the contacts in the closed position. Not credible. Cause - High inductance discharge from relay coil. I A connitment for testing of surge suppressors is in place. I
"AGC !
FRILURE RNALYSIS TREE M) :~ m SLICE SER CPCt m mm SE:6 0':SEC i 07-ER Se+r T ::c.imc 5 I i T C W IC.AL C EC?4ICAL _SLACX CONTROL RCC CABLE LIMIT SWITCH NOTES FAILURE MOCE EFFECTS ANALYSIS 1- NOT A FAILtAC (INTENDED A- NO INDICATING LIGHT !LL plNATCO FUNCTION) UPON ACTtFTION - NOT BY E DGCMCNT 2- NOT A CRCDIBLE FAILURC MODC A CRCDIBLE FAILtRC - SCC ' 4 ' (No CIRCUIT RVAILAR.C) B- INJICATING LICHT RCMINS 3- NOT A CRCDIBLE FAILLAC ILLLN!MTED. T4,qti.Y WY MODC (LPON LOSS OF RCD TO RC'CyC T4 STIPtJ.US IS CABLC, SPRING PPPLIES TO RC 6 f T4 ( IGHT OF CCfJR. FORCC 70 % MICA T4 CONTROL RCD, VI A T4
-1/4' e) - If A JfH OCCIARCD CABLC. 9<LLD THIS FAILLRC M LOSS Cr R00 CABJ. OCCA. IT MD BC RCSOLvCD IT WOLAD K RCPLMCCD BY SWIT04 4CetPCCt1CNT DLMING KFUt319fCNT 3. RING RCPJtBI9 toff TESTING.
TESTING. SEC *B' SCC '3' 4 - NOT, BY JUDCCITNT, A CRCDIBLE FAILLRC MODC (SW ACTICN SWITCH) SCC 'A' 5- NOT A CRCDIBLE FAILURE MODC (BASCD ON STRUCTLRAL %D TCP.ICR. RUGCCDPESS Cr SWI'J4)
~ _ . - . . - .
l-Basec on Failure Analysis What Can Fail? '
~
xplain wnat would cause f ailure. [ Pg. 2 - A - Slack cable Limit Switen fails electrically by sealing tne !
,y contacts in the closec :osition. Not judged a credible '
failure. Cause - High inductive discharge from relay coil. E A commitment for testing of surge su; pressors 15 in place. Pg. 2 - B - Slack cable Limit Switch fails mechanically by jamming the shaft in the completely actuated position. Not a credible failure, i Cause - None known, t
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i. w1 1 FRILURE ANALYSIS TREE 3CL i 4 !% C0vCR 3CLTS Tui OL49 V:CC PLaiC_ h sory OB 0^> 1CTCG BCTaCCN '#cC4 g g;ga St::CS C0tP.ING 1 SL!OC SLIOC
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/ I relPCR (SLIDC)
INCORRCCT {3 /\ 4 RCSIST44C SWT SEPIEINGS OPCN Sa* _4C % iCM- CLCCTRICR. l l l CONTROL 900 PCSI' ION ACTENTI W CR I I
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Notes 1 - Not a Failure (Intended Function) 2 - Not a Credible Failure (Failure of some otner type required to orovice stimuli for tnis f a ilu re ) . Failure of tnis type for coth analog and digital cortions of a single ootentiometer simultaneously is not considered credible. 3 - Not a Credible Failure (due to the Structural / Mechanical ruggecness of the Part). 4 - Not a Credible Failure (due to the materials used (non-correcing. Dresumed a type of magnetic stainless), lubrication, low speec, limited number of operations, and limitations of load applied). 5 - Based on operating experience, no known failures of these parts have occurred. 6 - No known failure modes (Strictly a mechanical support (bebbin) for winding the conductor coil of the potentiometer. 7 - Not a Credible Failure Mode (Neither mechanical (wear) or Electrical (opening or shorting) can occur due to the limited length of conductor and its installed configuration. The conductor is made up of approx. 40 strands which produce an extremely flexible AWG 26 (accrox. by measurement). Its length is such that in the installed configuration, this jacketed conductor does not come into contact with the conductor coil of the potentiometer and thus cannot be worn or shorted. l l l
;. 3.2 Failure Mode Effects Analysis A - When the uceer wiper (slice) jams (at ne : o or cot:0m of tre slice
- arriage) due to being driven oeyond its cesign range, system 'ailure is iminent. One potential failure is ne icer guice (a car: of
- ne acer) is broken, rendering tre slide incoeraole or at best uncrecic able. This failure is particularly likely to oc:ur anen the cotentiometer is driven very slowly beyond its cesign limits.
'n' hen the potentiometer is driven at scram soeed (3 turns in accrox.
150 se:encs, approximately 3.2 rom), the f ailure mode is far more likely to be Failure Mode "B". (See Note below.) 3 - When the potentiometer is driven beyond design limits, at or near scram soeed, the failure mode is most likely to be breakage of ne body (housing), the drive coupling, or the drive coupling pin. The result is acoroximately the same as in f ailure mode " A" in wnien the slide is inoperable or unpredictable. (See Note below.) NOTE: In tne event of failure modes "A" or "B", the output and thus th& indication in the Control Room becomes unpredictable. Since the 2-gang potentiometer is 2 independent potentiometers, the following may occur: 1 - Each portion may perform adequately. 2 - One portion may be accurate while the other is inaccurate (linearly or non-linearly). 3 - Both portions may perfom identically though inaccurately. 1
- . ]..]
Failure Mode Effects Analysis C - When moisture is introduced into a cotentiemeter, as in decressuri-
- stion follo ing a moisture ingress event, an oxice may result, any uninsulated coooer used in the succort coil may react electrically with the conductor coil and produce coocer oxide. This failure mode is considered unlikely in most potentiometers; yet it nas oeen observed on a limited scale. The result would be to increase tre resistance between the wiper and the conductor coil. This ould result in erroneous indication in the Control Room. This failure mode may be cacaole of correction by drying the nelium environment prior to depressurization. It is also possible that operation of the potentiometer may remove the oxide once it nas formed, correcting to some degree the erroneous indication.
Based on Failure analysis What Can Fajl? Explain wnat aculd cause failure. Pg. 3 - A - Control Rod Posi ticii Potentiometer fails mecnanically by jamming the upper wiper resulting in nicer camage. Cause - Control Roo Drive Mechanism (CROM) is overceiven (forced beyond the intended design limit) in an attemot tc actuat; non-working Limit switenes. An administrative procedure now controls this action to limit this type of damage. Pg. 3 - B - Control Rod Position Potentiometer fails mechanically by janming the upper wiper resulting in body or coupling damage. Cause - Same as "A" above. Corrent same as "A" above. Pg. 3 - C - Control Rod Position Potentiometer fails electrically by producing copper oxide and thus changing the resistance of the potentiometer. This results in erroneous indication t in the Control Room. I Cause - Electrolysis induced by bare copper and other metals in a moist environment. Apparently, the added current of the circuit aid:. in the electrolysis. 1 - Contact manufacturer for concurrence. 2 - Verify visually. Contacted and verified as required. I t I
, . , _ _ _ . __ _ . _ _~ . _ __ , _ . _ _ _ , -. - _ . . . . . _ _ . . . . _ . _ _ . _ _ . . _ . _ _ _ . _ . _ _ _ _ _ _ . . . _
Scoce of Review for Reclace ent Proc.:ts For Position Transmitters (samole) Encocers Cotical Bidirectional BCD Magnetic Position Transducers Motors Steppers Potentiometers Selsyns Off-the-Shelf Custom Design Compan'ies Contacted (sample): Microswitch Litton Autotech Bailey Controls General Electric Westinghouse Beckman Many others
For Full-In/ Full-Out Position sensiqq Switches Roller (Perpendicular to Contact Bar) Roller (Longitudinal with Contact Bar) Non-roller type with slide actuation Sensors Infrared Magnetic Proximity - various types including Eddy current killed oscillator for sensor
- integrated Full-In/ Full-Out/ Position Transmitters
Selected Reclacement Comeonents
- 1. For Red-In/ Rod-Out Proximity Sensor -
Eddy Current Killed Oscillator by Microsait:n Prototype eX82383-FW A remotely mounted transmitter receiver is used with this :ceconent. See Note below.
- 2. For Position Indication 2-g:ng Potentiometer by Beckman Prototype #7239-2966-0 See Note below, i
Note: All items designed to operate within the following parameters: Temp = 300'F min. Humidity = Approx. 1005 Radiation = 1 Rad per Hour l, Pressure
- 845 psi.
All parameters encountered simultaneously. ' f 1 l l
Ev- . . . PROPOSED PLAN TCR REQUALITICAT:0N CT 0 ROM's The f ollowing integrated plan has been developed for requalification of the CRDMs. The plan consists of the following phases: (1) evaluate temperature ' proflies obtained from operational data and programs in the area of the OROM assemblies to establish the temperature to be used for qualification basis: (2) document age-related qualifications of the CRDMs in accordance with the approved TSV EQ prograa through evaluation using EQ DOR aging criteria: (3) for components which may possess insufficient quallfled life at the elevated qualification temperature, perfora a failure-modes-and-effects analysis to determine the consequences of component failure during accident and post-accident operation; and (4) perform a type test of a complete CRDM assembly to resolve outstanding concerns, includit.g red position indication, synergistic effects, moisture ingress, and operation at elevated temperature. NRC Standard Review Plans 3.11 and 4.6 will be reviewed and applicable concerns identified in these SRP't-will be incorporated into the integrated CRDM requalification program. Although various phases of this approach have been performed pieceaeal to document qualification of the CRDMs, it is PSC's position that this integrated plan will provide a complete qualification approach for qualifying ! the CRDM's. TSV's IQ progran has been audited and approved by the NRC and I I utilizes a standardized methodology for qualification of these assemblies. The following paragraphs provide additional information related to each of the areas to be investigated. (1) DOCUMENTATION OF QUALIFICATION TEMPERATURE: Analyze and document available data on the maxinua expected actor temperature for which the CRDMs should be qualified. Document (1) recent actor temperatures while operating at power, (2) the program used to estimate maxinus temperature, and (3) assumptions used in this program which will require monitoring / review when additional operating data is l available. The purpose of the program and analysis is to provide the basis for the qualification temperature to be used in the determination of qualified life. (2) DETERMINATION OF QUALIFIED LIFE: The qualification will be based on application of the TSV EQ program, which has been reviewed and approved by the NRC, to the qualification of the CRDM's. Qualification will be tailored specifically for the CRDM environment and expected operational requirements Yhe program will have to address the following points as a sinimua:
- a. description of the applicability of Dot since this component is located inside the PCRY while other EQ items are located outside the l FCRY in the Reactor or Turbine Buildings.
I b. description of "mild" environment for the CRDMs: define mild and determine worst case accident and post-accident operating environment. l
- c. evaluate post-accident operability requirements for these components l
to ensure that, if required to operate, evaluation of the post-accident environment is performed as well as identification of any appropriate testing requirements.
- d. Identify the functional, operational, and design requirements for the CRDMs to ensure that these characteristics are used as the basis for type testing of the CRDN assembly.
e, qualified life vill be established using standard DOR sethodology through use of saterial analysis. Due to the age of the CRDMs, use of this methodology should be acceptable; full thermal aging by type test is not required since it is not required of stallar vintage EQ related equipment and the CRDMs are installed in a mild environment. (Type testing is recommended to be perf ormed but is intended to resolve non-aging, synergistic concerns identified by the NRC in their submittals to PSC.) (3) FAILURE MODES AND ETTECTS ANALYSIS (FMEA): In addition to determination of the qualified life for degradable components of the CRDM, a FMEA will be needed to Justify uJe of components which are determined to have insuffici.nt qualified life at the elevated temperatures expected. If any such components are identified, the TMEA vill determine the consequences of failure of these components or parts. If failure can be proven to cause no adverse effect on the ability of the CRDM to operate or safely shutdown, no action would be required. If, however, the part has inadequate qualified life and its failure has unacceptable consequences, then the recommendation should be made to limit the reactor power level such that an acceptable temperature can be achieved or replace the component prior to the end of its useful life. (4) TYPE TESTING: Several pre-1985 documents from the NRC to PSC identify the need to perform type-testing of the CRDM assembly. (See attached page for major comments.) Due to the uncertainty of the results obtained during the last CRDM type test, it is deemed to be prudent to retest the CRDMs to demonstrate suitable performance; the retest will include operability of the rod position indication, operation following noisture ingress, and operation at elevated temperatures. The intent of this test is to demonstrate the operability of the CRDM and to demonstrate that synergistic ef f ects will not adversely affect performance of the CRDMs; it is NOT the intent of this type test to establish the qualified life of the CRDM components.
l 1 SUXXARY CT SIGNITICANT NRC CORRISPONDENCE RELATED TO THE CRDM QUALITICATION ISSUE G-82J84:
"Synergistle effects such as tolerance accumulation, wear induced alsalignments, lubricant redeposition, differential thermal expansion etc..
aandate reliance on total assembly qualification tests rather than on analysis or tests of individual components." G-84392:
"The licensee should determine whether compensating design and/or operational modifications are needed to sintatze noisture ingress to the CRDM cavities and minimize temperatures in the vicinity of the red drives.
In the event that temperatures recorded during plant operation prove to be higher than those for which the assembly was initially qualified, perform requalification testing of a CRDM assembly." Critical comments received f rom the NRC in the SER for the CRDM Temperature Requalification (G-86664) include the following:
"The licensee did not develop functional, operational, and design specification based on the environment that the CRD0A's would be expected to operate in. The licensee's submittal did not provide acceptance criteria developed from the functional, operational and design specification against which to evaluate the test results. The tests were not performed under conditicas that are representative of conditions in TSV j (i.e., tests were performed with dry helium, while noisture is potentially present in the FSV reactor especially if there is a noisture ingress event). The licensee did not provide information on the neehanical and electrical properties of the asterials in the CRDOAs as a function of temperature, humidity, pressure, and radiation. The test was performed for a very limited time (i.e., 14 days at 300 F). The licensee's submittal did not explain how the data would be extrapolated to the length of time the CRDOA's would be required to operate at elevated temperatures, i
In conclusion, G-46664 stated:
"the revised submittal should include or consider such items as functional, operational and desita specifications, acceptance ersteria, material specifications, material properties, test specifications, test operatinc procedures, test data, and the necessary supporting analyses, i
1
EE-12-0013 Public FORT ST. VR AIN NUCLEAR GENER ATING STATION Rev. B SerVlCO' PUBLIC SERVICE COMPANY OF COLORADO APPENDIX 0 ANOMALIES CHART I i l l FORM iAs W 22-5317
All CRDM P sition Ass::ciated Other Circuitry Anomalies from NRC Inspection Instrumentation Indication Components Based on Described Component (s) Failure CRDM No./ Region Anomaly
- 1) Rod-In Limit Switches Indicating Light Cable, Connectors, 13/19 Faulty Rod-In Limit Switch, Incorrect (Control Room (CR)) Terminal Lugs, Blocks CR analog & digital indication (See 5,6)
Relays, Surge Protec- 39/23 Incorrect CR Analog position indication, tion (Diac), Power Faulty Rod-In Limit Sw3tch (See S) Supply
- 2) Rod-Out Limit Switches Indicating Light Same as Above 6/1 1 of 2 rod-out limit switches inoperable (CR) 26 Faulty Rod-out switch 29/ESWS Faulty Rod-out switch 37/3 Faulty Rod-out switches
- 3) Rod Retracted Limit Indicating Light Cable. Connectors, Switches (Test Box) Terminal Lugs, Blocks
- 4) Slack Centrol Rod Cable Indicating Light Cable Connectors, 25/7 Faulty Slack-Cable Switch
- Limit Switches (CR) Terminal Lugs Blocks, Relays, Surge Protec-tion (Diode), Power Supply
- 5) Position Analog Meter Cable. Connectors 12/15 Incorrect CR analog position indication Potentiometer (CR) Terminal Lugs. Blocks, 13/19 Faulty Rod-In Limit Switch, Incorrect (2 Gang) Part 1) Power Supply CR analog and digital indication (See 1,6) 33/16 Faulty analog position indication 39/23 Incorrect CR analog position indication.
Faulty rod-in limit switch (See 1)
- 6) Position Potentiometer Digital Meter Same as Above 13/19 Faulty Rod-In Limit Switch, Incorrect (2 Gang) Part 2 (CR) CR analog and digital indication (See 1,S)
Note - A known cable failure occurred in this position; therefore, the con-clusion of a faulty slack-cable switch is unfounded.
0013
~~ PubllC FORT ST. VRAIN NUCLEAR GENERATING STATION Eg1
' SerVICO ' PUBLIC SERVICE COMPANY OF COLORADO APPENDIX E FMEA 0F POSITION TRANSMITTERS l
l 1 l 1 roav ' A> w a un
PmGC 3 FRILURE RNRLYSIS TREE a BOLT RING COVER BOLTS MOUNT CLfW1P
' LACE PLRTE_
BC WE R BODY k LOWER SLIDES COUPLING _ SLIDE SLIDE JAM JAM SUPPORT CONDUCTOR COIL COIL J UPPER LGER q
- \J/
SLIDE CARRIAGE COIL
~
r l WIPER (SLIDE) I C SWT BCf4 TINGS OPEN SHORT MEC W ICR. ELECTRICk," CONTROL RCD POSITION POTENTICfTTER
1 Page 3.1 l Notes ! 1 - Not a Failure (Intended Function) i i . 2 - Not a Credible Failure (Failure of some other type required to provide stimuli for this failure). Failure of this type for both analog ( and digital portions of a single potentiometer simultaneously is : not considered credible. l 3 - Not a Credible Failure (due to the Structural / Mechanical ruggedness !
- of the Part), i 4 - Net a Credible Failure - (due to the materials used [non-corroding,
[ presumed a type of magnetic stainless], lubrication, low speed, ' limited number of operations, and limitations of load applied). i 5 - Based on operating experience, no known failures of these parts I have occurred. l l 6 - No known failure modes (Strictly a mechanical support (bobbin] for winding the conductor coil of the potentiometer). l . t 7 - No Credible Failure Mode (Neither mechanical [ wear] or electrical ! (openingorshorting]canoccurduetothelimitedlengthofconductor I ! and its installed cor. figuration). The conductor is made up of approx. ! i 40 strands which produce an extremely flexible AWG 26 (approx. by I measurement). Its length is such that in the installed configuration, ! this jacketed conductor does not come into contat.t with the conductor I coil of the potentiometer and thus cannot be worn or shorted. [ t 4
}
1 > l .i I , t b i ! i r ! i f
l l Pg. 3.2 - i i Failure Mode Effects Analysis A - When the upper wiper (slide) Jams (at the top or bottom of the slide i i carriage) due to being driven beyond its design range, system failure is imminent. One potential failure is the wiper guide (a part of ! the wiper) is broken, rendering the slide inoperable or at best unpredictable. This failure is particularly likely to occur when ; the potentiometer is driven very slowly beyond its design limits. : When the potentiometer is driven at scram speed (8 turns in approx. 150 seconds. approximately 3.2 rpm), the failure mode is far more ! likely to be Failure Mode "B". (See Note below.) i i B - When .the potentiometer is driven beyond design limits, at or near scram speed, the failure mode is most likely to be breakage of the
- body (housing), the drive coupling, or the drive coupling pin. The
) resul t is approximately the same as in failure mode "A" in which the slide is inoperable or unpredictable. (See Note below.) l NOTE: In the event of failure modes "A" or "B", the output and thus ! j- the indication in the Control Room becomes unpredictable. Since ' the 2-gang potentiometer is 2 independent potentiometers, ; the following may occur: i 1 - Each portion may perform adequately. 1 .
- 2 - One portion may be accurate while the other is inaccurate ,
(linearly or non-linearly). ) 3 - Both portions may perform identically though inaccurately. i j 1 } . 2 I i t l
O
?(
Pg. 3.3 Failure Mode Effects Analysis C - When moisture is introduced into a potentiometer, as in depressuri-zation following a moisture ingress event, an oxide may result. Any uninsulated copper used in the support coil may react electrically with the conductor coil and produce copper oxide. This failure mode is considereo unlikely in most potentiometers; yet it has been observed on a limited scale. The result would- be to increase the resistance between the wiper and the conductor coil. This would result in erroneous indication in the Control Room. This failure mode may be capable of correction by drying the helium environment prior to depressurization. It is also possible that operation of the potentiometer may remove the oxide once it has formed. correcting to some degree the erroneous indication.
- - 013 j;' 3 Pubil] FORT ST. VRAIN NUCLEAR GENERATING STATION
'SerVlCO'- PUBLIC SERVICE COMPANY OF COLORADO t
APPENDIX F FMEA 0F IN/0UT LIMIT SWITCHES i scau iai x4 :: un
i pact t FRILURE RNRLYSIS TREE ; CCePLETELY PRRTIRLLY l , NORMR. ACTURTED I l ! O SLIDE JAM h SER. OPEN SEAL CLOSED hh } SHW T JAM ROLL OTHER ROLLER CONTACTS l MECHANICR. ELECTRICR. t t + FLLL-IN/FLLL-0UT LIMIT SWITCH i , NOTES FAILURE MODE EFFECTS ANALYSIS 1 - NOT A FRILLRE (INTENDED R - INDICRTING LIGHT ILLUMINRTED, FLNCTION) REMRINS 50 WTER LOSS OF STIMJ.US 2 - NOT A CREDIBLE FRILt.RC MODC B - MCCmNICR. FRILtRC OF 4J (NO CIRCUIT fWRILABLE)- LFPREDICTfS.E NATURE WILL OCCUR 3 - NOT A CREDIBLE FRILLRC MODE C - NO INDICRTING LIGHT ILLUMINRTED I (BRSED ON MRTERIR.S 303&44055) LPON ACTLETION - NOT BY JUDGEMENT , j 4 - NOT, BY JUDGEMENT, R CREDIBLE R CREDIBLE FRILURC - SCC '4' FRILLRC MODE (SNRP RCTION D - INDETERMINRTE - R OR B BRSED SWITCH) SCE 'C' ON PRECISE POINT OF JAM :
' S - NO CREDIBLE FRILURE MJDE {
(BRSED ON STRUCTURAL R4D MECHANICR. RUGGEDNESS OF I SWITCH) !
. au mm xm.m ., im ex ;
l !
6 l r I Based on Failure Analysis What Can Fail? , Explain what would cause failure.
'Pg. 1 - A - Full-In/ Full-Out Switch fails mechanically by jaming the I
shaft in the completely actuated position, t Cause - Shaft Failure ; Investigate cause of shaft failure. i
Conclusions:
Shaft fails due to high -lateral forces caused i' by Cam angle >15' (recommended max per Mfr). Failure is by erosion, deposition of eroded metal, corrosion until ;
~
space between shaft and housing is filled and will not allow free movement. To verify: (1) Contact manufacturer for l (2) Check for corrosion / erosion / deposition concurrence; on only 1 side of shaft (side opposite Cam approach) Verify as required. y i Pg.1 - B - Same as "A" above, except position at which final failure occurs.* Cam may fail.* i > 1 Pg.1 - D - Same as "A" above, except position at which final failure f occurs. i i i Pg. 1 - C - Full-In/ Full-Out Switch fails electrically by sealing the : contacts in the closed position. Not credible. Cause - High inductance discharge from relay coil.
}
- A commitment for testing of surge suppressors is in place.
I l 4 e I h t i P 4 i 4 / l v
EE-12-0013 l Public FORT ST, VRAIN NUCLEAR GENERATING STATION
/ Se rVIC O ' PUBUC SERVICE COMPANY OF C', L'f ~~. ADO l
APPENDIX G FMEA 0F SLACK CABLE SWITCHES roav s,w n wr
P8CE I ; FRILURE RNRLYSIS TREE i 3a" m SLIDE m SEAL 0$N mm SEAL CLOSED ' h OTHER SMFT l CONTACTS I I i i l MECHANICAL ELECTRICAL i j SLRCK CCNTROL ROD CABLE LIMIT SWITCH , H NOTES FAILURE MODE EFFECTS ANRLYSIS f 1- PCT R FRILLRC (INTENDED R- PC INDICRTItG LIGHT ILLLMINRTED ! FLNCTION) UPON ROTURTION - PCT SY AIDGEMENT : 2- 14T R CREDIDLC FRILtRC MODE A CRCDIBLE FRILLRC - SEE '4' l (No CIRCUIT RVAILRBLC) 'B- INDICATING LIGHT DOES NOT
'3 - PCT R CREDIBLE FRILLRC ILLLMINRTE. - NOT A CREDIBLE MODC (NO STItt.LUS TO FRILLRC MODC (to STIMLLUS
.! PRODUCC THIS FRILLRC) TO PRODUCC THIS FRILLRC). '. 4- NOT, BY JUDGCMENT, A SWITCH IS HELD RCTURTED [ CREDIBLC FRILLRC MODE NORMALLY FY4D CHVJGCS STRTC ; (SNPP RCTION SWITCH) SEE 'R' MEN WEIGHT IS REMOVED. ;
'5 - NO CREDIBLE FRILURE MODC (
(BRSED ON STRLETURRL AND ! I MccavaICa. RuGGCDNESS OF SWITCH) ' i j i I e agN!ED MTDt ECDEM 4.1M7 ITITI$ I i i
Based on Failure Analysis What Can Fail? Explain what would cause failure. Pg. 2 - A - Slack cable Limit Switch fails electrically by sealing the contacts in the closed position. Not judged a credible failure. Cause - High inductive discharge from relay coil. A commitment for testing of surge suppressors is in place. Pg. 2 - B - Slack cable Limit Switch fails mechanically by jaming the shaft in the completely actuated position. Not a credible failure. Cause - None known. t
E C013 Pubils FORT ST. VR AIN NUCLEAR GENERATING STATION
%' $9fVIC9" - Pusuc ssRvece COMPANY OF COLORADO APPENDIX H SPECIFICATIONS F0i. AEW COMPONENTS Note- that these new components will be specified and procured in accordance with qualification criteria of new qualification program.
Material characteristics, temperature limits, expected service life, synergistic effects FMEA will be performed prior to modification in accordance with established qualification program. , scau m w.n un
Pubil:; FORT ST, VRAIN NUCLE AR GENr <AT.NG STATION
- 2-0013 7
ServlCO' PUBLIC SERVICE COMPANY 4.F COLORADO POSITION SENSORS SPECIFICATION AND CERTIFICATION FOR PROTOTYPES hhh (kl 1
QUALITY RELATED PURCHASE REQUISITION , . . . . . . . . . . . . , , , . . . . . , P>n9n TF99En Conf 1rming Y6148
" * - ~ - " . .
N- 6194 ~5'*"="~"== I 1dRQ En ARnADWAY Cmn1. 4.- t
.... . -, c Am_. ; j 8-6-85
... ,,,, n F ri.v.E R . cn Ans1n 686 E STOREKEEPER, FORT ST.VRAIN: PLATTEVILLE, COLORADO 80651 l
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- 2. C CO L LE CT wu s.... , . . , = . . _ . . . . . . . . . . . . . < , , , w . ... a , ,. . ..., _. ,
THE FOLLOWING CHANGES PERTAIN TO ITEMS ON THISi ; l Pn - 1 I I rHANGF CFRTIFTCATTON PARAGRAPM l FQOM. MANHFACT11RFA 'S PCRTIFICATE nF CONFOR- ' - I ' MANCE THAT MATERIALS FURNISHED MEET OR ' l' i FYrren TWC QCnlifRFMcNT9 0F 1CTTFQ
~
l I ADG-85-0277 (5TTACHED). b mulum. l l l l T0: MANUFACTURER'S CERTIFICATE OF CONFOR- l l l MACE TWAT f%IERIALS FURNISWEn MEET Op ' EXCEED THE REQUIREMENTS OF PAGE 2 0F j j ' T.W. T R. Dil D.r. W_A R. C_n. R. .n F R . I i EXCEPT FOR THE AB0VE CHANGE, ALL OTHER DOCUMENTATION, : I panic Ann pairrt RFYATN UNCHE'GED. I ' ' i I l l l
! I i -
I l 4
- I ! ! I i i i 1 i SEE ATTACHED SHEET FOR QUALITY QUALIFIED : MANUFACTUPER l AND DOCUMENTATION RECUIREMENTS. SOURCE : SUPPLIER i l REQUIRED: SERVICES i SPEC,NO. EQUIP. NO. OA REVIEW c.. . . ....m.: _ _ 32 .
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NOTES 1 - ELECTRICAL CONNECTIONS - THREE 20 GAGE WIRES PER MIL-W-22759/43 "RED" TO TERMINAL 3 0F 405FM504-B SWITCH CARD, ' BLUE" TO TERMINAL 6, AND "YELLOW" TO GROUND. WIRE LENGTH 72 INCHES MINIMJM. 2 - ACTUATION: WHEN OPERATED WITH THE 405FW504-B SWITCH CARD OPERATE POINT FOR A TARGET OF 15-SPH STEEL, .625 DIAMETER BY .062 THICK IS BETWEEN .060 AND .086 INCHES, RELEASE POINT IS BETWEEN .003 AND
.012 INCHES OVER THE OPERATE POINT DISTANCE.
3 - MINIMJM OPERATING TEMPERATURE OF THE SENSOR IS -77'C. 4 - SENSOR IS DESIGNED AND MANUFACTURED TO MEET OR EXCEED THE FOLLOWING ENVIRONMENTAL PARAMETERS: a) TEMPERATURE 9 +155"C b) PRESSURE 9 845 P.S.I. c) HUMIDITY 9 100% d) RADIATION 9 1 RAD / HOUR e) ALL THE AB0VE ENCOUNTERED SIMULTANEOUSLY y PART MARXING THIS SURFACE
- 2.4 "
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MICR@ SWITCH 8 A8 8802t ILtlNQil 6 803t A OMSION C# HONEYWELL INC. l ' CERTIFICAT1ON UIIAT GOIS 02iFOPM TO SECIFICATICNS This will certify that MICRO SfITG catalog listing XE383-FW (Custcce: Part Ntaber Reference Hone) furnished Public Service of Coloracb on its Confir::d.ng Purchase Order N6194 and Suppler.ent il to TTSSCO (MICRO Si1T3 A.D.) , Qr: tract (Ibne) confoms to the di::ensional, perfourance and envirorcental requirements of referenced purd.aae order. ) i By: ' Harlan K. Einkley, Supervisor Approvals Engineering 13 August 1985 [
)
EE-12-0013 Rev. 1 POTENTIOMETER SPECIFICATION AND CERTIFICATION FOR PROTOTYPES ) l I i f f l I \ \ . -
l l m u.e. . . O PublicService 2420 W. 26th Avenue Suite 100-0 Denver, CO 80211 Company of Colorado P.o. Box 84o Denver. CO 802o1 - 084o (303)571 7511 February 8, 1985
. Fort St. Vrain Unit No. 1 NDG-85-0081 Beckman Industrial Corp.
901 0xford Street Toronto, Ontario Canada M8Z ST2 Attn: Dan McBride
Dear Mr. McBride:
Below listed are our revised requirements for the quantity of potentiometers shown on the attached requisition. Please note this letter is specification for construction of required potentiometers. Reouirements: 1 - 10 Turn 3600' Elect. 5400' Mech.. Elec. Turns Centered in Mechanical Turns 2 - Mounting Same as Bechman 7603 3 - 2 Gang - Electrically Independent 4 - 1000 OHMS t 5% 5 - Linearity 9 1/10 of 1% Error 6 - High Reliability Seal - Per Item 10 7 - High Temperature Reliability - 300'F Min. 8 - Power Rating, Watts 5 0 70'F, Derating to 0 0 Maximum Temperature Rating 9 - Taps - Conductors or Terminals Penetrating "Can" (May Be Axial) 10 - Direct Water Ingress Resistance to Maximum level Reasonably Attainable Maintaining Torque < 10 In. Oz.
~
11 - Seismic Qualification - STD for Series 7600 12 - Shock - - - - - - - - - STD for Series 7600 13 - Standard Tests Not Named Above 14 - Hydrolizable Chlorides Subject to Item 16 15 - Copper may be used only if it is sealed completely away from other parts. 16 - Due to the atmosphere involved, all materials require approval prior to their application. 17 - Pressure = 845 P.S.I. 18 - Voltage Tracking Sect. 1 to Sect. 2 = .25%. Should you have any questions or desire further information, please do not hesitate to contact me at (303) 571-7134. ry truly y s, ames G. Henderson JGH/dh Attachment i
Beckman industrial ~ '"c" *5* " c",2 ru::::,= ? r"JM:nN.'"1,?,s** *'"" January 8, 1985 Public Service of Colorado, 2420 West 26th Avenue, Suite 100D, Denver, Colorado, U.S.A. 80211 Attention: Jim Henderson Re: C of C For Model 7239-2966-0 This is to certify that the components supplied on our P/S 08993P, your order #N6113 were designed to withstand 845 P.S.I. and manufactured, inspected, tested and con-form to your letter NDG-85-0081, February 8, 1985. Yours very truly, BECKMAN INDUSTRIAL CORPORATION ELECTRONIC TECHNOLOGIES DIVISION ES/cc f a Smit , quality Control Manacer NMN igON s CWAN IMOV$tmiAk CORPomatt04 A $44$i0i AR Y 08 (W88 SON $ LICTRIC CQ 6
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l ,, 5 . .11 .. . . . . . . ! ,, 6 .. . ...12................ Ph2stng Within . ..).f. . of . 9..M Frint Shaft Extn. ..G.I.o.9, .. . . . l Reor ,, ,, . . . . .A)/8. . . . .. . .. . .. . . Fica Run.Out . .. C.:.Q.0.3.. . . . . Diameter Run.OuC..Q O/.I... .. . . .. . Shaft Run.Out . . O.4 C.Q/... .. ... .. .. . . . Sh2ft End Play . . C: .QE. .. . . . .. .. . . . Radlal Play . . . . .C:f. 9. b. . ... .. . . . .. . St:rting Torque . ../.C.... MAX.. In. oss. 8urrs . . . . ........./................................ Flnisb . .. /. . . . .. . . . . . . . . . . . . . . Oth:r Features .. .. .. . . . . . . . . . . . . .. Mechanical.. Rotation ..dd.YM.. .. ..
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Interoffice Memo NDG-88-0208 O Pubh. c Service ~ March 10, 1988 Pubuc Service Cey of Cokwado Date To L. Marauez From J. Henderson A ttn. Subj. Lack of Radiation Soecification in letter #NDG-85-0081 While the radiation requirements are not specifically addressed, they were covered in discussions with the manufacturer. This does not appear to be a significant concern, as these components are prototypes only (not for permanent installation). The permanent plant components must have the 1 rad per hour specification included.
- 4. G. Tiende'rson JGH/dh l
l L FORM 184 8% 22 0054
- - . _ _ _ - . .- ..- . . - - . . - - . ~ .. - -. -.=_. . .-.. . . .. .._.
. L E-12-0013
. Public FORT ST. VRAIN NUCLEAR GENERATING STATION .
4
$9fVlC98' 'PUBUC SERVICE COMPANY OF COLORADO ~
4 + I h 1 0 V i- : Y
-APPENDIX I q
!. MAINTENANCE INFORMATION I i r t i t l l r E I r I w I t I I r FOAM (Al 364 , 22 5317 4 W-T-T'v us y 7 e e P p f-M WT_m#eM 9-t& ', _- W PN.4-=TwyT-_ ., N M yN M*yw'WWPMC_
FORT ST. VRAIN NUCLEAR G NERATING STATION Page 1 of 29 PubilC SOfVlCCO PUBLIC SERVICE COMPANY OF COLORADO TITLE: CRDOA PREVEFT MAINTANENCE DEPARTMENT: MAINTENANCE SPONSIBLE 7
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UTHORIZED NEW .
'00T,6 - J987 !F EE DVE
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DCCF NUMBER (S) bl-)O)1 W^ HQC Review J~ i~CC/(50 A) 8PfmV2 '3/ Do not start test before Week # and must be completed by Sch. Clerk . O . ja W T.1 7 4 -.- ON
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8 g)% Public FORT ST. VRAIN NUCLEAR GENERATING STATION of 29 Vd7 S0rvicoo PUBUC SERVICE COMPANY CF COLORADO 5.46 H inspection and/or test acceptance criteria (s) are not met, THEN
~
initiate SSR to refurbish components as required and indicate extent of refurbishment below. SSR # N/A Remarks _ _f 5.47 Replace orifice position l potentiometer per approved procedure. Procedure (s) used: l NOTE: Red potentiometer and limit switch I j replacement should be performed during drive train refurbishment, if / refurbishment is required. l l 5.48 Replace rod position potentiometer per approved procedure. k-Procedure (s) used: i
' /
/
I 5.49 Replace slack cable, red in, red ( I out, and rod retract limit switches t I (2 each) per approved procedure (s). Procedure (s)used: _ f ~~ _
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l Test Conouctor Signature Date MRMiel372 22 3k3
** TOTAL PA ++
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1 E-12-0013 Public FORT ST. VRAIN NUCLEAR . GENERATING STATION O Service *1PUBUC SERVICE COMPANY OF COLORADO o. APPENDIX J FAILURE STUDY OF RPI COMPONENTS FROM C0!HERCIAL OPERATION TO DATE 4-A M 4Ai W . M M17
EVALUATION OF FAILURES OF R00 POSITION INDICATION
-COMPONENTS SUBSEQUENT TO COMMERCIAL OPERATION By J. G. Henderson February 7,1988
l 1 l REVIEW 0F MAINTENANCE DATA
- OATE COMPONENT SER! DISPOSITION 821229 In/Out 34 Type 1 811117 In R4 Type 2 810325 Relay 16 Type 3A 791107 Retract 17 Type 3 811110 Pot 17 Type 3 791107 Retract 17 Type 3 791103 Analog Ind 25 Cleaned Meter Case 791114 Pot 27 Zero Pot Adjusted 810706 Pot 31 Type 1 821018 Pot 5 Type 3 800804 Pot R23 Type 3 791107 In 1 Type 3(Assume In) 791109 In 17 Type 3 840217 In 19 Type 3(AssumeIn) 801031 Pot 27 Cancelled 840719 Slack 7 Not a Failure 821008 In 8 Type 1 840516 In 3 Type 1 830625 Pot 35 Type 3 790801 Pot 35 Type 2 No Failures 8507 thru 8802 Disposition Types:
- 1. Work description indicates no work required
- 2. Adjusted Component
- 3. Replaced Component 3A. Replaced component (mounted outside core)
Note: Sources of data include: a) '79 thru '84 component tracking report #5597 i b) '8507 thru '8802 data from system engineer r L
EVALUATION OF PERIOD BETWEEN FAILURES DATE COMP SER# MONTHS HOURS (THOUSANDS)*** 791107 In 1 --- --- 791109 In 17 .066 8.9 800804 Pot R23 9 1,216.2 811110 Pot 17 15 2,029.1 821018 Pot 5 11 1,486.5 830625 Pot 35 8 1,081.1 840217 In 19 8 1,081.1 8406XX Misc Misc 4 540.5* 8802XX ---- ---- 33 4,459.7** Multiple failures assumed to occur at same time (as the purpose of this document is to show period between failures, no benefit is gained by showing specific data)
** Calculated from July, 1985 (end of refurbishment)
*** Hours of availability / year = 365.25 x 24 hours x 37 rods x 5 components (Note 1)
Note 1 Five components is conservative as (2) rod-in and (2) rod-out switches are considered while the potentiometer (dual-gang) is' considered as (1) component. The slack cable switches are not considered as no failure of this component has has ever occurred. l l l l l
CONCLUSIONS APPROX. AVERAGE NUMBER OF FAILURES PER CALENDAR YEAR:
.P0TS:<1 SWITCHES:<1
- 1) Some of the components which failed thru the years prior to and during the '84 incident were part of the original installation. ,
- 2) Multiple failures during June, 1984 point to the high probability of operator error. The operators (as observed and reported by the NRC) tended to overdrive the rods in an attempt at obtaining correct indication from noncperational limit ~ switches and thus damaged the potentiometers as well.
- 3) The zero failure rate from refurbishment of 1985 (approximately July) thru February 1988 shows that enhanced operator training in operation of the control rod drives has been effective.
Furthermore, the availability of the new equipment for 4.5 million hours indicates the replacement parts have served well.
- 4) The reliability rate of the existing RPI system (based on a failure rate of I component per year) is:
( 1 failure per year }=.005 (37 rods x 5 components) 8ased on conclusion 2, above, 1984 failures are not included in failure cale as the data is not considered valid for the reason ' indicated. l f.
' _3
- r >
E ' Public IFORT ST. VRAIN NUCLEAR-GENERATIN'G STATION-
-2{0013-ServlCO* PUBUC SERVICE COMPANY OF COLORADO .,
l d ( 4 i L I APPENDIX K i
- i. FMEA 0F NEW DESIGNS-I I
I Fomu iA3 94 22.U17
FMEA of Proposed Replacement Sensors (1) Non-Cor. tact Sensing for Full-In/Dut Position Failure Modes Analyzed Credible? Comments t Jamming (mechanical) No Non-Contact, Double Nutted into place. Sensor Failure Yes Per Mfr. Cale: MTBF 0 257'F = 2 million hrs /8766 hrs /yr = 228 yrs Life MTBF 0 300*F = 1.5 million hrs /8766 hrs /yr=171 yrs. l (2) Extended Range Potentiometer Jamming (Mechanical) Yes If rod is driven approx. 30% beyond Full-In or Full-Retract, the potentiometer will jam and r fail like the existing component. Galvanic Corrosion No All materials are isolated from the copper winding core. Disparity in Output No .1% linearity
.25% Tracking Phase 1 to Phase 2.
l l i l l l Y
EE-12-0013 PUDllC) FORT. ST. VRAIM NUCLEAR GENERATING STATION Rev. B-
$9fVICO* PUBUC SERVICE COMPANY OF COLORADO -
i l- APPENDIX L (REF. 1) i Preliminary report related to the restart and continued operation of - Fort St. Vrain Nuclear Generating Station (excerpt from) 4 i FORM IA) 344 22 5317
p * **e JY. n,It UNITED STATES
! 'ef G NUCLEAR REGULATORY COMMISSION ih6
?.,' .rf4 j/
! w^sMiNcToN o e 2csss OCT 161984 -bbE Mr. R. F. Walker, President Public Service Co;pany of Colorado h- !-b~
P. O. Box 840 Oenver Colorado 80201
Dear Mr. Walker:
In early July I dire:ted my staff, with assistance from Region IV, to pe'- fem an audit of Fort St. Vrain operations including problem areas associated with.the June 23, 1984 event regarding the failure of a number of control rocs to insert. A preliminary report (copy attached) has been developed by the staff which documents the results of our assessment. The report contains findings that the staff believes should be implemented before and after station restart. These findings are contained in the Executive Sunmary and at the end of each section in the body of the report. The report is preliminary in that various options to solve staff's findings are available to the licensee and need to be discussed prior to final resolution. Public Service Company of Colorado should review and evaluate the report and detemine what followup actions are appropriate. We intend to schedule a meeting to discuss your proposed actions-to resolve these findings and will be in contact with you to schedule such a meeting. Sincerely.
/ -
Harolo R. Denton, Director Office of Nuclear Reactor Regulation
Enclosure:
As stated cc: See next page ! LI 'r pil I3 ' y . - s- \ 1p l l n :
3 CC G CL RCD INSTRUMENTATION ANCMALIES 3.1 Introduction On July 30, 1984, the NRC was infonned of numerous and various control red instrumentation anomalies in several refueling regions in the reactor. The 11even ancmalies included: simultaneous red-in and red-out indications, out-limit switch lights remaining lighted, indications of partial red with-drawal, no position signals, disparity between analog and digital red position information, and a slack-cable indication. A team of NRC technical personnel and their consultants from Los Alamos National Laboratory, visited the plant - site on August 1 .3, 1984, to review the instrumentation problems. This Section reports the results of that plant visit. Tnis Section incluces a description of the Control Rod Drive instrumentation characteristi,cs and.the t anomalies observed, a preliminary-evaluation of the aremalies and their effect - en verifying control red position, and recemnendations. The results are included in the assessment report because they relate to the overall l perfom.ance of the CROMs. i l 3.2 C_ontrol Rod Instrumentation As described in Section 2 of this report, the principal mechanical cor.penents i of the CRDM are the shin actor and motor brake assesely, the reduction gearing to the cable drum, and the control rod pair susppnded by cables from the drum. Integrated into the CROM and used to determine control rod positions are the instrumentation-related components that include the red position l potentiometers, rod-in and rod-out limit switches, limit switch cams and gear reducer, and the slack-c.able indication device. The mlative locations of these cosconents are shown on Figure 3.1. The control rod potentiometers are intended to provide continuous monitoring of control rod position. As shown in Figure 3.2, both of the ten-turn, l potentiometers are directly coupled to the rotation of the cable drum. The l l 3-1 1
coupling is provided by connection of the potentiometer snaft to the cable crum hub. The potentiometer shaft passes tnrougn the drum support, and a oinion gear on the shaft drives the limit switch cam wneel. Beyond the ; inion, a multi-jaw ocuoling drives the two potentiemeters. Outout from the ;otentio-meters is provided to the operator through separate analog and cigital rencouts in the control room. As previously mentioned, "rod-in" and "rod-out" position indication is providec by cam-actuated limit switches (the potentiometer snaft shewn in Figure 3.2 dlso drives the cam wneel). At the full red-in position, the cable anchor on the cable dmm and the cam are at the 6 o' clock position, and both rod-in limit switches are actuated. Full rod travel (190 f,1 inches) to the rod-out position causes the cam wheel to rotate clockwise 3/4 turn, actuating the pair ~ of red-out limit. switches. Limit switch actuation is indicated by steadily-lighted lamps in the control room. The slack-cable switch assemoly shown in Figure 3.3 is provided for monitoring the tension in the caeles supporting the control roc pair. The spring plunger exerts an upward force on the underside oi' the cable dme soindle, wnich counteracts the downward force caused by the control red weight. In the event a control red cable becomes slack, the upward force overcomes the control red weight, and microswitches are a:tuated. Slack cable indication is signallec in the control room by both an alarm and a light. , 3.3 gn,t,rol Red Drive. Instrumentation Review In response to the information received about instrumentation anomalies the NRC staff ar.d Los Alamos personnel reviewed the overall control rod asseemly as it relates to control rod drive perfonnance and control rod position indication. The team requested and received extensive, detailed information on the CRCM anc interfacing instmmentation. The major itams discussed were:
- 1. the instrunentation anomalies currently being experiented;
- 2. the CADM mechanical interface to the red position instrwentation, which includes the potentiometer drive gear and shaft acting as the 3-2 .
n ~
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pinion for the cam wneel drive and the red-in anc red-out limit switches, a multi-jaw coucling, and the two potentiometers; and
- 2. One electrical cacabilities of the CTOM instrumentation to detemine control red position uncer nonnal and adverse concitions.
The instrumentation anomalies ecserved at the time of the plant visit are sumari:ed in Table 3.1. , From a mechanica'l perspective, the CRCM instrumentation is inoted directly coupled to the motion of the cable drum, which dictates the movement of the control rods. However, under certain conditions, the meenanical aspects can actually inhibit the perfor nance of the instrumentation. The c:mnon shaft controlling both the rod position potentiometers and the limit switches is susceptible to damage by overeriving the control red pair past the rod-in position. This problem has been encountered when, for some reason, a given l co'. trol red pair, when supposedly fully inserted, may not actuate the red-in I limit switch. In an attempt to actuate the switch, the operator would typically try to overdrive the control rods to attain full insertion (even thougn the operation manuals (Ref.1) strongly advise against such a maneuver). i e In this case, the in-limit cam can rotate past the in-limit switch, and the out-limit cam can rotate to interfere with other mechanical components
~ resulting in damage to potentianeter shaft. This' damage can occur because ,
neither a positive stop is provided to restrict cam wheel overrotation nor is s there sufficient clearance for the cam to overrotate. If the shaft is damaged by overtorquing or shearing, or the multi-jaw coupling is displaced, the two red position potentiometers at the end of the shaft can give: the same j erroneous signal about the actual position of the control rods on both l 7 readouts, different analog and digital outputs for the same rod positien, or no ( output signals at all. On ,the other hand, if the control rods are overdriven L by 1/4 turm of the cable drum (about 10 inches), the cable. anchor can become wedged in the cable groove of the drum, and then upon. withdrawal, the control yrod cables may not feed onto the dnJa correctly, or can become entangled. It l 1s clear that this type of integration of mechanical and instrumentation
~
l components can result in a single point failure, can potentially result in l u 1 33
damage to :ne instrumentation, and does not provide an 1.idecendent incicatiori of control od full-in position. Discussions were held with tne licensee regarding the electrical inter' acing and cont-ci of the CEM and instmmention. Electrical schematics fr:m tne "Rod Centrol System huipment I-9303, Operation and Maintenance Manual". Ref.1 and the "Installation, 0: era:1on aiic Maintenance Manual for tne Con: ol anc Orificing Assemoly for the For: St. Vrain Reacter' Ref. 2, were reviewed by tne staff. The licensee stated that none of the control and indication circuitry for the CEM is classified as safety-related, except for the bypass circuitry used riuring a scram to energize the shim motor to drive in the centrol rods. - A nunter of questions were raised as to the effective redundancy in circuits such as red-in anc red-out limit switch indications. Acceroing to the elec-trical schematic (Figure 3.4), the red position limit switches may'not be, in effect, . redundant, because,the less of one switch of the pair is not , cetectable. -In other words, a single circuit switch failure would go undetected. If this occurs the loss of the second switch will result in ~ complete loss of full-in or full-out indication. The rod-in and red-out limit switenes art integrated into the mechanical aspects of the CEM to the extent nat the same switches are used for both rod drive control and rod position _ indication. Purchase specificatien1 were reviewed to deterwine how the CROM instrumentation should be expected to perfom in normal and adverse environmental c nditions. That review indicated that all CROM instrumentation is of commercial grace, anc that no special quality or safety-related specifications are recuired. No housing or shielding is provided to protect CROM instrumentation from the axisting environment. Under relatively nomal operating conditions, such as just. prior to tne control rod insertion failure, control red position instrveentation had been generally operable. Within two days after the control rod insertion failurt, all control rods were exercised, and rod position instrumentation was operable. Some two weeks after the insertion failurt event the core was depressurized. It was O ___ n _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ . _ _ . . _
after the depressuMzation that the broad array of ins:mmentation snemalies were cbserved duHng subsequent control red exercising. The staff, therefore, believes.that the red position instrumentation is e1uite likely to become unreliable when subjected to depressud:ation follcwing execsure to a hot, moist environment. It is likely tnat concensation of moisture and we::ing cf electrical components occurred after reactor pressure was reduced.
*he staff focused its attentien en the specific creblems regarcing :ne CTCM located in Regien 19, where both the analog and digital readouts indicated that the cortrol red was withdrawn about 40 inches. When asked if the instrumentation should be believed, the PSC responded that they did not believe ,
the installed red pcsition reaceuts, but that they had veMfied to their satisfaction that the contrel rod was indeed fully inser:ed. This conclusien was based on signature traces free watt-meter testing of the shim motor. Accorcing to PSC, at some point after depressuMzatien, the Region 19 control red pair was being exercised, and wnen the rod pair was fully inserted, the rod-in limit switch incication did not come on. The operator drove the red i pair in further so as to make contact with the limit switch. Again, no contac: was indicated. At some point duMng this attempt to get full red-in , j indication, the 40-inch offset was noticed. Therefore, a watt-eater test was performed on the Region 19 shim motor. The wattage reading of the shim metar was recorded as the rod pair was "yo-yo-ed," 1.e., a Pepetitive wit *crawal and insertion sequencing. Based on past experience and the Region 19 traces, PSC , concluded that the control rod pair was fully inserted. However, when :ne ' staff examined the Region 19 traces, the evidence was inconclusive. The inconclusive watt-meter traces in conjunction with analog / digital indications prompted the staff to request that a more definitive veMfication of the . Region 19 control rod position be performed. The veMfication technique selected by the licensee consisted of manually retracting the cortrol rod pair from the region, as if in preparat1h for refueling removal-1.e., the control rods are completely drawn up into the refueling penetration housing. A completely separate limit switch, indicates full control red retraction to the refueling position. If the e.ontrol rod were completely inserted . then full retraction would require an additional 50 inches 3-5 8 e
l
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of travel over the nonnal 190 inches of red travel (for a total of 240 inches) Frem the red-in position, corresponcing to a total of 330 turns of the manual rewind t:el. On the night of August 2,1954, the Region 19 c:ntrol rod pair was withdrawn 330 turns when the full retract positien indica:1en was cocained, proving tnat indeed the red had been fully inserted. Hewaver, it aisc cemenstrated that the installed red position instrumentatioc alone is insufficient to determine control rod positions under acverse cencitions, such as wnen mechanical damage may have occurred to red pcsition instrumentation. It is evident that centrol rod analog and digital positien incications can be in agreement without reflecting the tnJe red pair position. Statistically, no particular instrumentation ancmaly to be prevalent
- under these adverse circumstances. A sumary of the current status of 'the control red dMyes, instnanentation and orificing,, as prepared by PSC, shows tnat the anomalies tend to be evenly distributed aang red limit switching prcblems, digital and/or analog readout prcblems fm red position pctentio-meters, with the exception of the single slack-cable related problem. Thi s-pattern is in contrast to the histoMcal distribution, where red pesition potentiometer problems had been domir. ant. In reviewing past instrumentation anomalies as recorded on PTRs (Plant Trcuble Report) since plant startue, aporeximately 47% of the anomalies are rod position potentiometer problems, about 17% are limit-switch related, less than 25 are related to slack-cable indication,17% are connected to oMfice valve problems that are not germaine to this review, and 19 are related to motor / brake and other anmalies. In a aujerity of the cases, the "faulty" component was simply replaced.
3.4 $UMAAY In general, the instrumentation anomalies are believed to b1 the result of mechanical damage or exposing the CADMs to a hot, moist atmosphere and a subsequent core depressuMzation. As stated in Ref.1, proper CROM and instrumentation performance requires maintaining purge flow into the CRDM cavity, and maintaining dMye mechanism temperatures below 250*F. Loss of purge flow certainly contributed to the ingress of moist helium into the CRCM 3-6 O 9
i l l .. . cavities. The effects of depressuri:ation on instrunentation per#crmance are - not well understocc, but there is significant evidence that depressuri:atien in j conjunction with a moist enviro. ment can tend to increase instrumentation preolems. These pr:blems are likely caused by concensatien of moisture en alectrical c:cenents. In c:ntrast, the Region 19 instrwnentation ancmalies were most likely caused by overdriving the centrol reds, thereby damaging ne
- etentiemeter shaft, and rtsulting in amoigueus and erroneous pesition instrumentation readings.
3.5 CONCLUSION
S The following conclusions are based en the possibility of unreliable perdernance of the CRDM red position instrumentation under adverse conditions , and/or mechanical damage from overdriving, questions concerning instraentation recundancy, and the lack of independent red full-in position verification. The staff has deternined that the following actions must be completed prior to , restart: 1
- 1. To prevent CROM damage and to protect red position potentiometers and limit switches, plant procedures should be changed to prevent overcriving the control rocs past the red-in limit (yo-yo-ing).
- 2. Periodic surveillance of rod position potentiometers and switches sheule -
be developed and- implemented in interim precedures aad be peccesed f:r inclusion in the plant Tecnnical Specifications. This surveillance should include verification of limit switch operability and cenfirmation *Jat redundancy has not been lost. The following actions should be taken in the long-ters:
- 1. Damage due to overtravel should be precluded either by the installation of a positive mechanical stop or by providing sufficient clearence to prevent damage.
3-7
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- 2. An appropriate, independent and definitive means of verificatien of control red full-in position should be provided because the installed red position instrumentation can be inadequate to verify control red position. In the present fem. Watt-meter testing of the shim motor is consicered inacequate to verify full insertion cf control rocs. It is therefore concluded that the Watt-cater method be refined or an alternate method be developed to achieve sufficient resolution of red position and then formali:ed into a plant precedure.
- 3. Conduct an integrated systems study to resolve red position ' indication maintenance and operability questions.
3.6 REFERENCES
- 1. "Installat15n, Operation, and Maintenance Manual for the Control and Orificing Assenely for the Fort St. Vrain Reactor', GA-9806, May 1977.
- 2. "Rod Control System Equipment I-9303. Operation and Maintenance Manual".
E-115-265 (REY. 3), August 1979. l . l \ - l l l l - l l l 1 3-8 l
Table 3.1 Current CRDM and Instrumentation Anamelies CRCH Current Instmmentation
& Recion Problem 6 I 1 of 2 red-out limit switches inoperacle.
12 15 Incorrect Control Room (CR) analeg position indication. 13 19 Faulty red-in limit switch, incorrect CR analog and digital indication. . 25 7 Faulty slack-cable switch. 25 Faulty red-out switch. 29 ESW5* Faulty rod-out switch.
- 33 16 -
Faulty analog position indication. - , 37 3 Faulty red-out switches. 33 23 Incorrect CR analog position indication, faulty red-in limit switch.
- Equipnent Storage Well d
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t FORT ST. VRAIN NUCLEAR GENERATING STATION )2-0013.
'('3 PubilC -
$9fVIC9*- Pusuc senvece COMPANY OF COLORADO APPENDIX M (REF. 2)
Evaluation of integrated systems study of control rod drive mechanism position indication instrumentation for the Fort St. Vrain Nculear Generating Station. Iow io x4.u.uu
a ase
/c w%,
UMito STATES
! w % f i i NUCLEAR REGULATORY COMMISSION 9 gK nasmotos o c. rosse
/ July 31, 1987
, g l0 -p I Occket No. 50-267
\
Mr. R. O. Williams, Jr. Vice Presicent. Nuclear Operations Puolic Service Company of Colorado P. O. Box S40 Denver, Colorado 80201-0840
Dear Mr. Williams,
SUBJECT:
EVALUATION OF INTEGRATED SYSTEMS STUDY OF CONTROL R00 ORIV POSITION INDICATION INSTRUMENTATION FOR THE FORT ST. VRAIN NUCl GENERATING STATION (TAC NO. 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 (CROM) position instrumentation at Fort St. Vrain (FSV). This review was per-formed by our contractor EG&G, Idaho, Inc. Their Technical Evaluation Report (TER) is enclosed. The staff has reviewed this TER and concurs with the EG4G, Idaho, Inc. conclu-sien that Public Service Company of Colorado has not provided an acceptaole 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 fonvarding this TER to you to allow for another proposal for the upgracing 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-l fore OM8 clearance is not required under P.L. 96-511. l Sincerely, I
...,..,..n l i
Kenneth L. Heitner, Project Manager . 6.Y ,b i Project Directorate - IV . t". % o . l Oivision of Reactor Projects - III, F.::.4, , IV, and V and Special Projects 'c7e Office of Nuclear Reactor Regulation ,
"[G'" .
- ;;/
Enclosures:
As stated i EIJ
.,...o g g , .
ec w/ enclosure: l See next page j' TMN 7p9
Mr. R. O. Williams Public Service Company of Colorado Fort St. Vrain CC: Mr. O. W. Warembourg, Manager Albert J. Hazle, Director Nuclear Engineering Division Radiation Control Division Public Service Company Department of Heaith of Colorado P. O. Box 840 4210 East lith Avenue Denver, Colorado 80220 Denver, Colorado 80201 Mr. David Alberstein, 14/159A Mr. R. O. Williams, Acting Manager GA Technologies Inc. Nuclear Production Division Post Office Box 85608 Public Service Company of Colorado San Diego, California 92138 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Mr. M. L. Brey, Manager Nuclear Licensing and Fuel Division Mr. P. F. Tomlinson, Manager Public Service Company of Colorado Quality Assurance Division P. O. Box 840 Public Service Company o.f Colorado Denver, Colorado 80201 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Senior Resident Inspector U.S. Nuclear Regulatory Commission Mr. R. F. Walker P. 0.8ox 840 Public Service Company of Colorado Platteville, Colorado 80651 Post Office Box 840 Denver, Colorado 80201-0840 Xelley, 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 100-0 Denver, Colorado 80211 Regional Administrator, Region IV U.S. Nuclear Regulator 611 Ryan Plaza Orive, Suite y Commission 1000 Arlington, Texas 76011 Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative l Radiation Programs l Environmental Protection Agency ! 1 Denver Place l 999 18th Street, Suite 1300 l Denver, Colorado 80202-2413 1 t I
EGG-NTA-7705 May 1987 INFORMAL REPORT EVALUATION OF INTEGRATED SYSTEMS
/daho SR DY OF CONTROL R00 ORIVE MECHANISM RCO Nat/ons/ POSITION INDICATION INSTRUMENTATION FOR THE FORT ST. VRAIN NUCLEAR GENERATING STATION Engineering Laboratory _
Managed by the u S Ceoanment D. E. Jackson 0/ EN9V C. L. Nalezny I l 1 l l hEGsm
- Prepared for the wwe m m I wco$le#qo c U.S. t01 EAR EGIATORY C&filSSION t
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r r OlSCLAIMER This report was prepared as an account of work sponsored by an agency of tne United States Government. Nettner the United States Government nor any agency thereof, or any of their employees, senes any marranty, espretted or inelied, or assumes any legal Itaollity or responsibility for any third party's use. l of any information, apparatus, prodwCt or process disclosed in this report or represents that its use by swCM third party would not infringe privateif owned rights. l I I t s t e m ,
EGS-NTA-170$ EVALUATION OF INTEGRATED SYSTEMS STUDY OF CONTROL R00 ORIVE MECHANISMS R00 POSITION INDICATION INSTRUMENTATION FOR THE FORT ST. VRAIN NUCL:AR GENERATING STATION Occket No. 50-267 TAC No. 62198 INEL Reviewer - D. E. Jackson
!NEL 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 i l l l i Prepared for the l U.S. Nuclear Regulatory Commission Washington 0.C. 20555 under 00E Contract No. OE-AC07-761001570 ( FIN No. 06023 l l
ASSTRACT This EG&G Idaho, Inc., recort presents tne 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 Cocnission (NRC) by the licensee, Public Service o'f Colorado (PSC). The evaluation by EGSG Idaho, Inc., concludes that PSC has not complied with the NRC directive to prepare an engineering evaluation of the problems experienced with the cantrol 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 proposed replacement instruments that are important to safety (full-in limis .d thch and red position potentiometer) do not comply with the quality standard and instrumentation requirements of General Design Critoria No. 1 and 13 of Appendix A to 10CFR50. Docket No. 50-267 TAC No. 62198 11
m FOREWORO This report is supplied as part of tne "Review of Plant Specific Licensing Actions for Operating Reactors," Task 1-14 being con ~ ducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of PWR Licensing-8, by EG&G Idaho, Inc., NRR and !&E Support Branch. The U.S. Nuclear Regulatory Comission funded the work under authorization B&R 20-19-10-11-2, FIN No. 06023. I Oceket No. 50-267
, TAC No. 62198 I
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CONTENTS ABSTRACT ...................... . . . . . - . . . . . . . . . . . . . . . . . . ........ ... i' FOREWORD ............................................................ . 'i' 1.0 SACKGROUND .. .................................................... '
- 2. DESIGN BASES CRITERIA ............................................. 3 3.0 EVALUATION ....................................................... 4 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 INTEGQATED $YSTEMS STUOY OF CONTROL ROD ORIVE MECHANISMS R00 postil0N IN0! CAT!CN INSTRUMENTATION FOR THE F(QT ST. VRAIN NUCLEAR GENERATING STATION 1.0 BACKGROUNO Following a scram on the morning of June 23, 1984, at the Fort $t. 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 ,,as to include problem areas associated with the June 23 scram. The Control Rod Drive Mechanism (CROM) is comprised of the shim motor and motor brake assembly, the gear reduction to the cable drum and the control red pairs suspended by cables from the drum. Instrumentation-related l components, integrated into the CROM, used to determine control rod positions include the red position potentiometers, red-in and red-out limit i switenes, 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 red instrumentation anomalies in several refueling regions in the reactor I (Refence 1, section 3.1). The eleven anomalies included: simultanseus 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 red position information, and slack cable j indication. l The results of the NRC audit were documented in a preliminary report which was issued on October 16, 1964, (Reference 1). The report contains findings to be addressed both before and after plant restart. The NRC staff noted that a number of deficiences need to be corrected on a long-term basis 1 l l l l
4 following resta-t. ime licensee was directed ta suomit senecules ott-+e 6; cays of restart for ce?cle'.ing these items. One of tne items listec se:er "Actions Required Following testart" is "Conduct an integratec systems st cy (engineering evaluation) to resolve rod cosition indication, maintenance anc coerability cuestions." On August 15, 1986, Public Service Company of Colorado (DSC) issue a report entitled "Integrated Systems Study (Engineering Evaluation EE-12-0013) of the Control Rod Drive Mechanism Rod Position Indication l, Instrumentation" (Reference 2). This Engineering Evaluation by DSC was in 1 response to the requirement listed in Section 4.e of the Executive Summary l of Reference 1, and is the subject of this staff evaluation, j l I i i i I i .l
- 2. DESIGN BAS!$ C4!TERIA Tht follCwing General Oesign Criteria (GCC) of Accencix A to 10 CFR 50 aere acolied to the evaluation of tne Fort St. Vrain Integrated Systems Stacy of Control 2.od Drive Meenanisms Roc Position Indication instramentation.
Criteria 1 - Quality Stancards and Records. Structures, systems, and comconents important to safety shall be designed, fabricated, erected and tested to quality standards conmenserate with the importance of the safety function to be performed. 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 accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fiss1on process, the integrity of the reactor core, 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. 3
- 3. EVaLUAitCN l The PSC response to the NRC creliminary recort, s:ecifically accresses the requirement for an integrated systems study (engineering evaluation) to resolve red position indication (RP!) maintenance and coeracility cuestions. It was issuec in the form of an Engineering Evaluation (EE-12-0013) and documents the evaluation of several acti.e and :assive electrical components in the RP! system to identify any potential design deficiencies. These components included the control red-cair full-in and full-out limit switches, slack cable limit switenes and red pair cosition 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 Centrol tod-pair Full-in and Full-Out limit Switches The most common problem observed with the full-in and full-out limit switches was pitting, e.rosion 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 shaf t 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 conoansation of moisture. It is very likely that moisture caused pitting of the plunger and contributed to the failure of the red position 4 i _ _m. . - _ _
instrumentation. The study by pSC coes act include an evalwation of corrosion of the switch C0mconents, anc the effect corrosion nould nave on the mechanical operation of the snitches. PSC proposes to replace these snitenes with croximity sensors nnich are scocially designed for their acolteation and environment, This is an acceptacle solution if the sensors are designed to satisfy amoropriate functional, and operational recuirements, and are cualified 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. The full-in limit indications can be classified "important to safety" because they are the primary means of verifying that the control rod drives (CR0s) 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 GCC 1, or the instrumentation requirement of GOC 13. 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 shaf t, 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 f.ecessary to address the fact that the stainless steel targets will damage the potentiometer shafts when the system is overdriven. 5
Since tre anal 0g Cosition indicators are "imoortant to safetj," ; e out* limit Cam must be designed so that overdriving it will rot ca* age Ine potentiometer shaft, or so that they can not te overceiven. Trerefore, it is concluded that PSC nas not perfor ec a satisfact:ry engineering evaluation of the problems associated with tne full-in/ full-out switenes, mas not proposed solutions which will correct the orcelems, and has not complied with the requirements of GOC 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 mar.af acturer 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 addretsed by the licensee. The statement that the manufacturer confirmed this to be an acceptable application of this switch is not acceptible because it does not resolve the question of why the switch failed. The engineering evalustion 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 i limit switches and has not proposed any solution to correct the problem. i 3.3 position potentiemeters i i The engineering evaluation of the position potentiometers by PSC revealed that resistivity changes were caused by moisture intrusion into the I 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 I (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. 5 l [
pSC proposes to reolace trese potentic eters with new ores scecifically designed and fabricated for tnis acclication. The replacement potentiometers are to be built with a 10 ture electrical section centeres :n a 15 turn mechanical section anc mounted tne same as the Bechman Model 7(03s
,, mien are currently used. These and otner s:ecifications are included in "Specification for prototype Dotentiometers," Appendix 0, of Ref. 2. The proposed replacement potentiometers are an improvement. However, tecause they provide the operator with continuous position information for all the rods during operation and following a scram, they are considered "immortant to safety." Therefore, they must be designed, fabricated and procurred to the requirements of the appropriate GOCs. The specifications for the potentiometers presented in Reference 2 are appropriate for a comercial grade component, but did not comply with the quality standards and reporting 4
requirements of GOC 1, which are necessary for a component that must ccmply with GOC 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 comercial 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 preposed 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 toperature limits of the FSV Control Rod Drive and Orifice Assemblies (CRDCAs) to 3000F. The findings and deficiencies identiffe6 in Reference 4 are genert11y applicable to the CRDM 7
rod position instumentation. The operating environ?t9t of the CROMs an: the rod position instrumentation is the same. Therefore, the important to safety instrumentation should be subject to the same reovirements. The deficiencies in the CROM submittal (Reference 4) that are applicable to the i
, safety related roc position instrumentation are listed below.
- 1. PSC did not provide acceptance criteria developed from the functional, operational and design specifications against wnich to evaluate the proposal.
- 2. PSC did not provide information on the mechanical and electrical properties of materials in the RP! comoonents as a function of temperature, humidity, pressure, and radiation.
l 3. PSC did not address maintenance of RPI instrumentation. i i i 8
4 CCNCLUSICNS The licensee for tne Fort St. Vrain Nuclear Dener Station, nuclic Service of Colorado, has not provi:ed an acceptable proposal to upgrace a selected number of Control Roc Position Instrumentation Systems. The rational for this conclusion is given below. The submittal was reviened for compliance with the NRC requirement tnat DSC conduct an integrated systems study to resolve rod position indication maintenance and operability cuestions, and the applicable requirements of the General Design Criteria (Appendix A to 10 CFR 50) for "important to safety" instrumentation. O The licensee did not perform a thorough evaluation of the failures that were identified in the NRC preliminary report.bihe contribution of corrosionofthefull-inlimitswitcheswasnotevaluated,%hedesignofthe 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 overdrivenkhe failure of the slack cable limit switches was not evaluated,andhespecificationforthereplacementrodposition 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. Inaddition%heproposedreplacementinstrumentsthat are important to safety (full-in limit switch and rod position l potentiometer) did not comply with the quality standard and instrumentation requirements of GOC 1 and 13 of Appendix A to 10 CFR 50. i 9 l
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4.0 REFERENCES
- 1. NRC Letter, Centon to Walker, Daelimiaary Re ert :elateo to testart and Continued Operatin of Fort St. Vrain Nuclear Generatino Station, (G 84392) dated October 16, 1984
- 2. PSC Letter, Warembourg to Berkow, Fort St. Vrain Rod Position Instrumentation intearated Systems Study, (P-86522) dated l August 15, 1986.
- 3. U. S. Nuclear Reculatory Commission, Standard Review Plan, NUREG-0800, Rev. 2, July 1981.
l 4 NRC letter, Heitner to Williams, Control Rod Drive and Orifice Assembly 3000F Temperature Limits Fort St. Vrain Nuclear Generatina Station, G-86664, December 24, 1986. I 4 .l l l 10 l
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This EG1G Idaho. Inc.. report presents the results of an evaluation of an integrated system study (engineering evaluation) of the control red drive mechanism rod position indication instrumentatien for the Fort St. Vrain Nuclear Generating Station which was sutenitted to the Nuclear Regulatory Comission (NRC) by the licensee. Public Srvice of Colorado (PSC). Inc.. concludes that PSC has not complied with the NRC directive to prepare a
, engineering evaluation of the problems experienced with the control rod drive roc l position indication because it does not adequately address the problems that were experienced at the Fort St. Vrain Nuclear Generating Station, and does not procese acceptable component replacements.
In 6ddition, the components of the red 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. I
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