ML20112J477

Audit of Environ Qualification of Safety-Related Electrical Equipment at Palo Verde Nuclear Generating Station Unit 1

ML20112J477
Person / Time
Site:	Palo Verde, 05000000
Issue date:	09/30/1983
From:	Borgen R, Holloway E EG&G, INC.
To:	NRC
Shared Package
ML20105B661	List: ML20112J145 ML20112J349 ML20112J410 ML20112J431 ML20112J439 ML20112J453 ML20112J477 ML20112J489
References
CON-FIN-A-6415, FOIA-84-262 EGG-EA-6416, NUDOCS 8501180295
Download: ML20112J477 (41)

Text

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EGG-EA-6416 SEPTEMBER 1983 9

. AUDIT OF THE ENVIRONMENTAL QUALIFICATION OF SAFETY-

, RELATED ELECTRICAL EQUIPMENT AT PALO VERDE NUCLEAR

.4 GENERATING STATION, UNIT 1 R. A. Borgen E. R. Holloway Idaho National Engineering Laboratory Operated by the U.S. Department of Energy

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. .s 8501180295 840524 l hth4-262 PDR Prepared for the U.S. NUCLEAR REGULATORY COMMISSION g Under DOE Contract No. DE-AC07-761001570 S G g G idanc l FIN No. A6415 N

EGG-EA-6414 4

AUDIT OF THE ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT AT PALO VERDE NUCLEAR GENERATING STATION, UNIT 1 Docket Number STN-50-528 R. A. Borgen E. R. Holloway Published September 1983

. Reliability and Statistics Branch Engineering Analysis Division EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 under 00E Contract No. DE-AC07-76IO01570 FIN No. A6415

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ABSTRACT

Palo Verde Nuclear Generating Station, Unit 1, was audited to determine the environmental qualification of safety related electrical equipment. Results of the audit are summarized in this report.

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SUMMARY

An audit of the environmental qualification of safety-related electrical equipment at Palo Verde Nuclear Generating Station, Unit 1, was conducted by a team composed of representatives of the Reliability and Statistics Branch of EG&G Idaho, Inc., and the Nuclear Regulatory Commission (NRC) staff. - Qualification deficiencies for individual equipment items are provided in Appendix A. Summaries of the central file reviews are provided in Appendix 8. It was concluded from the audit that the applicant must supply additional information to the staff before a determination of acceptability can be made.

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FOREWORD This report is supplied as part of the " Equipment Qualification Case Reviews" being conducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Engineering, Equipment Qualification Branch by EG&G Idaho, Inc., Reliability and Statistics Branch.

The U.S. Nuclear Regulatory Commission funded the work under the authorization, B&R 20-19-40-41-2.

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r-CONTENTS ABSTRACT ...... ....................................................... 11

SUMMARY

............................................................... iii FOREWORD ............................................................... iv :

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1. INTRODUCTION ..................................................... 1 l

2. BACKGROUND ....................................................... 2 1

3. PURPOSE .......................................................... 4 ,

4. SCOPE .........................'................................... s

5. EVALUATION ....................................................... 6

6. CONCLUSIONS ...................................................... 15 l 1

7. REFERENCES ....................................................... 16 i

i APPENDIX A--EQUIPMENT QUALIFICATION STATUS ............................ 17 APPENDIX B--SUMMARIES OF CENTRAL FILE REVIEWS ......................... 27 l l

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AUDIT OF THE ENVIRONMENTAL QUALIFICATION OF

, SAFETY-RELATED ELECTRICAL EQUIPMENT AT PALO VERDE NUCLEAR GENERATING STATION, UNIT 1

1. INTRODUCTION Equipment which is used to perform a necessary safety function must be demonstrated to be capable of maintaining functional operability under all service conditions postulated to occur during its installed life for the time it is required to operate. This requirement, which is embodied in

, General Design Criteria 1 and 4 of Appendix A and Sections III, XI, and XVII of Appendix B to 10 CFR 50, 4 s applicable to equipment located inside as well as outside containment. More detailed requirements and guidance relating to the methods and procedures for demonstrating this capability has been set forth in 10 CFR 50.49, " Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants," and NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment." This NUREG supplements IEEE Standard 323-1974, and various NRC Regulatory Guides and industry standards.

On June 28-30, 1983 a team comprised of representatives of the Reliability and Statistics Branch of EG&G Idaho, Inc., and the NRC staff conducted an audit of the environmental qualification of safety-related electrical equipment for Palo Verde Nuclear Generating Station Unit 1. The work effort consisted of: (a) a pre-audit review of the licensee's submittal, (b) an audit of the licensee's central files for selected equipment items, and (c) an onsite visual inspection of the equipment items. Qualification deficiencies for individual equipment items are provided in Appendix A. Summaries of the central file reviews are provided in Appendix B.

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2. BACKGROUND NUREG-0588 was issued in December 1979 to promote a more orderly and systematic implementation of equipment qualification programs by industry E

and to provide guidance to the NRC staff for its use in ongoing licensing reviews. The positions contained in this report provide guidance on (a) how to establish environmental service conditions, (b) how to select methods which are considered appropriate for qualifying equipment in different areas of the plant, and (c) other specific topics such as margin, aging, and documentation.

In February 1980 the NRC requested certain near term Operating License (OL) applicants to review and evaluate the environmental qualification documentation for each item of safety related electric equipment and to identify the degree to which their qualification programs comply with the i staff positions discussed in NUREG-0588. IE Bulletin 79-018 " Environmental Qualification of Class 1E Equipment," issued January 14, 1980, and its

! supplements. dated February 29, September 30, and October 24, 1980 '

established environmental qualification requirements for operating

{ reactors. This bulletin and its supplements were provided to OL applicants for consideration in their review.

j A final rule on environmental qualification of electric equipment important to safety for nuclear power plants became effective on February 22, 1983. This rule, Section 50.49 of 10 CFR Part 50, specifies the requirements to be met for demonstrating the environmental qualification of electrical equipment important to safety located in a j harsh environment. In accordance with 10 CFR 50.49, the eiectrical equipment in Palo Verde Nuclear Generating Station Unit 1 may be qualified in accordance with the acceptance criteria specified in Category I of NUREG-0588.

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The qualification requirements for mechanical equipment are -

principally contained in Appendices A and B of 10 CFR 50. The

• qualification methods defined in NUREG-0588 can also be applied to mechanical equipment, i

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In response to the above requirements, the applicant has provided

. equipment qualification information in letters dated February 7,1983 anc June 24, 1983 to supplement the information contained in Section 3.11 of the FSAR.

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3. PURPOSE The purpose of this audit is to evaluate the adequacy of the Palo Verde Nuclear Generating Station, Unit 1, environmental qualification '

program for electric equipment important to safety as defined in '

10 CFR 50.49, and.for safety-related mechanical equipment. A discussion of open items, as well as any unresolved issues, is provided in this report.

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4. SCOPE The scope of this report includes an evaluatio'nof the completeness of the list of equipment to be qualified, the criteria which they must meet, the environments in which they must function, and an assessment of the qualification documentation for the equipment. The principal area of review was the qualification of safety-related equipment which must function in order to prevent or mitigate the consequences of a loss-of-coolant accident (LOCA) or high energy line break (HELB) inside or outside of containment, while subjected to the harsh environments associated with these accidents. It is limited to electric equipment important to safety within the scope of 10 CFR 50.49, and safety related mechanical equipment.

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5. EVALUATION The evaluation of the applicant's environmental qualification program included an onsite examination of electrical equipment, audits of qualification documentation, and a review of the applicant's submittals for '

completeness and acceptability of systems and components, qualification methods, and accident environments. The criteria described in NUREG-0800, Rev. 2, Section 3.11 and NUREG-0588, Category I form the basis for the evaluation of the adequacy of the applicant's qualification program.

Revision 1 of NUREG-0588 was utilized to clarify staff positions as required.

The staff' performed an audit of the applicant's qualification documentation and installed electrical equipment on June 28-30, 1983. The audit consisted of a review of ten files containing equipment qualification documentation. The staff's findings during the audit are discussed in detail in Section 5.4 of this Technical Evaluation Report (TER).

5.1 Completeness of Eouioment Important to Safety The applicant was directed to (a) establish a list of systems ard components that are required to prevent or mitigate a LOCA or an HELS and (b) identify components needed to perform the function of safety related display. instrumentation, post-accident sampling and monitoring, and radiation monitoring.

Based upon information in the applicant's submittal, the staff has verified and determined that the systems included in the applicant's submittal are those required to achieve or support: (a) emergency reactor shutdown, (b) containment isolation, (c) reactor core cooling, (d) containment heat removal, (e) core residual heat removal, and .

(f) prevention of significant release of radioactive material to the environment.

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In order to comply with 10 CFR 50.49, the following information,must

'be submitted by the applicant prior to granting of an operating license for

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Palo Verce Unit 1:

1. A list of all nonsafety-related electrical equipment, located in a harsh environment, whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions by the safety-related equipment. A description of the methods used to identify this equipment must also be included.

The nonsafety-related equipment identified must be included in the environmental qualification program.

2. A statement that all safety-related electrical equipment in a harsh environment, as defined in the scope of 10 CFR 50.49, is included in the list of equipment identified in the Equipment Qualification (EQ) submittal.

3. A list of all_ post-accident monitoring equipment currently installed, or that will be installed prior to plan operation, that is specified as Category 1 and 2 in Revision 2 of Regulatory Guide 1.97 and is located in a harsh environment. The equipment identified must be included in the environmental qualification program.

5.2 Qualification Methods 5.2.1 Electrical Equipment in a Harsh Environment Detailed procedures for qualifying safety-related electrical equipment fn a harsh environment are defined in NUREG-0588. The criteria in this NUREG are also applicable to other equipment important to safety as defined in 10 CFR 50.49. Type testing of equipment in a sequence consisting of pre-aging (thermal, radiation, and mechanical), seismic and dynamic loading, and exposure to LOCA/HELB conditions (where applicable) is the preferred method of qualification. However, in a number of cases the 7

applicant has extrapolated partial test data to estaolish equipment

, qualification. A review of this analysis finds the approach to be adequate '

except as noted in this report.

5.3 Service Conditions NUREG-0588 defines the methods to be utilized for determining the environmental conditions associated with loss-of-coolant accidents or high energy line breaks, inside o'r outside of containment. The review and evaluation of the adequacy of these environmental conditions are described.

below. The qualification documentation has been reviewed to ensure that the qualification conditions envelop the conditions established by the applicant.

5.3.1 Temperature, Pressure, and Humidity Conditions Inside the primary Containment The applicant provided the LOCA/MSLB profiles used for equipment qualification. The peak values resulting from these profiles are as follows:

Maximum Maximum Temperature Pressure, Humidity F psig  %

LOCA 300 49 100 MSLB 370 41 100 5.3.2 Temperature, Pressure, and Humidity Conditions Outside the Primary Containment The applicant has provided the, temperature, pressure, and humidity conditions associated with high energy line breaks outside containment. A

• screening criterion of saturation temperature at the calculated pressure was used to verify that the parameters identified by the applicant are "'

acceptable. The NRC has previously accepted the HELB analysis in SER Section 3.6.

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5.3.3 Suor.ergence The maximum submergence level established by the applicant is

, 10-1/2 f t. above the containment floor. All safety related equipment subjected to submergence is or will be qualified for the time duration necessary for it to complete it's safety function. The applicant should provide a list of all equipment important to safety subjected to submergence and provide justification that failure of this equipment will not affect the safety function of any other equipment and will not mislead the operator.

The effects of flooding on equipment located outside containment have been evaluated to assure that safety shutdown can be achieved.

5.3.4 Chemical Spray Chemical spray is used for containment heat removal following a design basis accident. The specified composition of the spray is 4400 ppm Boron buffered with 50 ppm hydrazine with a pH of 7.0 to 8.5 using trisodium phosphate. Equipment inside containment was reviewed for qualification under the above condition.

5.3.5 Aging The aging program requirements for Palo Verde Nuclear Generating Station (pVNGS) electrical equipment are defined in Section 4, Category I of NUREG-0588. The degrading influences of temperature, radiation, vibration and mechanical stresses should be considered and included in the aging program. This requires the establishment of a qualified life with maintenance and replacement schedules based on the findings.

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5.3.6 Radiation (Inside and Outside Containment)

The applicant has provided values for the radiation levels postulated to ex,ist following a LOCA. The application and methodology employed to determine these values were presented to the applicant in NUREG-0588 and 9

NUREG-0737, " Clarification of TMI Action Plan Requirements." The review determined that the values to which equipment was qualified enveloped the

• requirements identified by the applicant.

The value specified for use in equipment qualification in the containment is 3.3 x 107 rads gamma and 1 x 10 8 to 2 x 108 rads 0

beta. In the auxiliary building, a value of 6.4 x 10 rads gamma has been utilized in areas with recirculatory fluid lines.

5.4 Environmental Qualification Audit An audit was conducted of the applicant's qualification documentation and installed equipment on June 28-30, 1983. Ten equipment items were reviewed to determine if the test data and analyses in the' files supported the qualification status determined by the applicant. The following comments, excluding the open items already covered previously in this TER, were made during the. audit:

1. The files were poorly arranged and pertinent information was neither included nor referenced.

2. During the plant walkdown the applicant was unable to locate two items (Rockbestos cables inside containment and the pre-amplifier for the Excore Detector). One item (ASCO Solenoid valve) didn't have a. nameplate attached.

3. The applicant is using 30 days post accident operability time for most equipment items. Only one equipment item (HPCS pump motor) has been identified to have an operability time of 4 months. The

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applicant's justification for using 30 days post-accident operability time is under review.

4. Ten files were reviewed during the audit. Several comments were .

made on these files. The applicant was informed during the exit interview that these comments should also be applied to other files, if applicable.

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l The comments on specific files are as follows: .

1. Rockbestos Cable (13EM-058) o According to Test Report E058-13-4, Radiation Aging was finished by July 1976 while the DBA test was not done until February 1978. What happened in between?

o Temperature and pressure profiles do not envelop the required profiles. No explanation is provided in the files regarding the acceptability of these profiles.

o Margin is not included for the temperature and pressure parameters. Justification is not provided in the file.

o Operability time required is for 30 days while the test duration was only 16 days. How was the 30 days operability time with margin achieved?

o The status of the cable in the submittal and in the file is defined as qualified, however, Arizona Public Service (APS) has not approved the file.

o The checksheet should identify the applicability of different test reports, e.g., test report E058-40-3 is not applicable to Palo Verde but is kept in the files only for reference and E058-39-3 is applicable only for outside containment.

o Although an outstanding deviation report was issued by the APS for the subject cable, the status of the report was still defined as qualified.

o The file was not arranged in a logical manner and much correspondence related to qualification was not -included in the file.

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2. ASCO Solenoid Valve o The aging analysis did not take into account the temperature rise for the solenoid valve because of the energized '

condition. During the plant walkdown a thermometer was placed on the surface of the solenoid ana a temperature of 141.4*F was noted while the ambient temperature was only 98 F. It should be noted that 141.4 F was only the surface temperature. The temperature inside the valve (coil and seat) will be higher than this value. The aging analysis should take into account the temperature rise because of the energized condition.

o There are three test reports included for the solenoid valve; however the information relating one test report to 2

another could not be easily followed. The checksheet should refer to a specific report and paragraph where the utilized information is located. Also if report AQS 2768 is used for outside containment only, then it should be so stated.

3. Feedwater Isolation Valve Solenoid Valve o The qualification testing does not envelop the required pressure profile. The steam test was done at atmospheric pressure while the required pressure was 21 psig. The file didn't justify the deficiency in the test report.

4 Rotork A C Motor Operators o The applicant used the data identified as the specified test environment as the qualification data and not the actual .

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o The number of test cycles used in the applicants analysis was not the actual number of cycles performed on the valve.

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l o The 10 C rule was used for aging analysis without any justification.

5. RTO's and Thermocouples (Weed Inst.)

o 005 #41 states that the review of the file is complete and the only outstanding item is related to submergence.

However, the checklist in the file identified many open items which are not resolved. It is recommended to change the status to not qualified unless all the items are identified on the Qualification Data Sheet (QDS). The review of the report was not conducted because of many open items.

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6. Pneumatic Low Leakage Damper o Why was the test sequence of radiation aging before thermal aging used especially since radiation is the only harsh environment (excluding seismic)?

7. Barton Transmitter o The correspondence between the applicant and Barton about the possible problem with lead corrosion and applicant's resolution of the problem should be included in the qualification file.

The resolution of the concerns identified above must be provided 60 cays prior to issuance of an operating license.

5.5 Outstanding Equipment For safety-related items not having complete qualification documentation, the applicant has provided commitments for corrective action and schedules for completion. For items not expected to have full qualification, analyses will be performed in accordance with paragraph (i) 13

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of 10 CFR 50.49 to ensure that the plant can be operated safely cending completion of environmental qualification. These analyses must be *

• submitted prior.to granting of an operating license. Section 5.1 discusses

{ additional equipment which may require qualification based on JU 10 CFR 50.49. These items must be identified and included in the

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. qualification program.

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6. CONCLUSIONS The Palo Verde Nuclear Generating Station programs for the

, environmental qualification of electrical and mechanical equipment has been reviewed. This review has included the systems selected for qualification, the environmental conditions resulting from design basis accidents, the methods used for qualification and the documentation for specific items of equipment. The following items are outstanding and must be resolved prior to issuance of an operating license.

. 1. Information demonstrating qualification of all electrical equipment located in a harsh environment, including certain nonsafety-related equipment, equipment required by the TMI action plan and installed Regulatory Guide 1.97 equipment, should be provided, or justification for interim operation in accordance with 10 CFR 50.49 must be supplied. The qualification information or justifications should be submitted with sufficient time for review and approval prior to issuance of an operating license.

2. Additional information concerning surveillance techniques to be utilized for detecting age-related degradation of equipment, should be provided.

3. Equipment in the containment potentially exposed to submergence must be identified and qualified, as discussed in Section 5.3.3.

4. Tne outstanding items identified during the site audit by the staff must be resolved as described in Section 5.4 Based on these considerations, it is concluded that satisfactory completion of the corrective actions identified herein will' ensure conformance with the requirements of 10 CFR 50.49 and relevant parts of General Design Criteria 1 and 4 of Appendix A; Sections III, XI, and XVII of Appendix B, 10 CFR 50; and the criteria specified in NUREG-0588.

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7. REFERENCES

1. Interim Staff Position on Environmental Qualification of Safety Related Electrical Equipment, NUREG-0588.

2. IEEE Standard for Qualifying Class IE Equipment for Nuclear Power .

Generating. Stations, IEEE Std. 323-1974. - *

3. Environmental Qualification Repo'rt 'per Requirements of NUREG-0588, Palo Verde Nuclear Generating Station, Arizona Public Service Company.

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APPENDIX A EQUIPMENT QUALIFICATION STATUS 4

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APPENDIX A EQUIPMENT OVALIFICATION STATUS Tables A-1 and A-2 list qualification status for Balance of Plant (80P) equipment and Nuclear Steam Supply System (NSSS) equipment respectively. The following code is used for status designation.

Legend A -

material-aging evaluation; replacement schedule; ongoing

. equipment surveillance CS -

chemical spray EXN -

exempted equipment justification inadeouate H -

humidity I -

HELB evaluation outside containment not completed M -

margin P -

pressure Q

qualified pending implementation of aging program QI -

qualification information being developed QM -

qualification method QT qualification time R radiation RPS -

equipment relocation or replacement schedule provided RTS -

retest, schedule provided S -

submergence SEN separate effects qualification justification inadequate T -

tempdrature AC -

accuracy 19

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These deficiencies do not necessarily mean that the equipment is unqualified. However, the deficiencies are cause for concern and require -

further case-by-case evaluation. The applicant should resolve these deficiencies and document the resolutions in an auditable form. -

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A I Aht i A-1 PAtO VERDf NuCifAR GIN!RAllNG SIAllON ItALANCI Of PLANI QUAL 15ICAllON SIAIUS Reviewer

_ _ _ . _ . . . (quip *f n t , __ _ Manufacts_tret__ Model [ valuation _

1. 8s80V motor control centers General Electric Prototype 1C7700 'QI

?. Auxiliary relay cabinets Harlo Corp. Prototype Qi

3. SkV power cables Anaconda FR-[P insulation Q

4. Low voltage electrical penet ra t ion Conax Corp. Prototype R. M. CS, A

5. Medium voltage electrical penetration Conax Corp. Prototype R,A

6. Instrumentation electrical pene t ra t ion Conax Corp. Prototype Q

l. 600V control cable Brand Rex Xt PE insulation QI

8. 600V powe r cable Rockbestos XLPE Insulation P, T, QT, A

9. 600V power cable Brand Rex XLPE Insulation Q

10. Coaxial cables Brand Rex XLPE Q

11. Shielded instrument thermocouple Anaconda IR-EP insulation Q extension cables [riccson

12. Shielded instrumentation cable Anaconda FR-EP insulation Q

13. Outside containment prefab cables Bendix/ Anaconda fR-EP insulation Q CPf Jacket Ira . Inside containment pref'ab cables Blw Silicon Q1 Insislation and

Jacket

15. Inside containment prefab cables BlW IhfP insulation M, R, A

, CPE Jacket

16. Ring tongue terminals AMP kynar Insulation M, R, A IF. High voltage terminations 5 and 15kV Raychem NHVI-ilVCM Q 18 Heat shrinkable motor connection kits Raychem NHCK-NCHk-NESk- RA WCSF-N

19. Ilea t shrinkable motor connection ki ts Raychem NHCk8 QI 20, 4.16kV transfer swi tcle III If Pl -C M, A

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TABl[ A-1 (Continued)

[.gujpment _ Reviewer

_,,Nept![a g(gre r Mode 1 [yajua Rqn_

21. stydrogen recombiner hand swi tch Consip Custom PTSB-C2-12-C Q1 Line Pump panel hand switch CE SM8 10 CC 211

22. Pressure transmitter Rosemount 1153CB7 Q 1153C86

23. Level transmitter Rosemount 1853DB4 Q 24 F low transmi tter Rosemount 1153085 Q

25. Pressure transmitter Rosemotent 1153GB9 Q

26. 9/P converter Masonellan 8005A Q

27. Ifydrogen moni tor Consip Delphi kill QI t 28. Acoustic monitors Technology for 88N4241SO QI Energy Corp.

29. Acoustic monitors

- Technology for 424-C2 Q1 E ne rgy Co rp.

30. Acoustic monitors Technology for 2273AMI QI Inergy Corp.

31. Preamp l i fie r Technology for T[C-504A QI f nergy Corp.

32. Transient shield technology for TEC-160-2 QI Inergy Corp.

33. Radiation resistant cable Technology for 2273-C2 QI Energy Corp.

3 fa . Level t ransmi t te r T ra nsa me rica ---

QI DeLaval Cens Sensors

35. Level switch Magnetrol F L S-X-M PX-S I MD Q 4DC

36. HID's and thermocouples Need Instrument HID's 611-18-C4- QI Company C-2-A2-0 IC's K282506 -

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Heviewer

_ _ _ _ _ _ . _ _ , _ ~ [quipment . , _ _ _ _ _ _ Manufacture r _,_ Model [va l ua ti on _

31. DC motor operated valve actuator Control $8-00-10 Q Components Inc.

38. - AC motor operated valve actuator Control S8-00-10 Q Components Inc.

39. Position transmitters Control [XG9-X3-X6P6- IP Components Inc. X6P6-13MH31 40 Position transmi tters Control HPA-10000 Qi Components Inc.

41 Limit switch Control NAMCO EA 180 Q Components

42. Solenoid valve Control ASCO c06-381-44 Q Components

43. Solenoid valve la rge t Rock inc. 76HH-Family QI

44. AC motor opera tor L imi to rque SMB-00-2 Q

45. AC, motor opera tor Limitorque SMB-00-40 Qi SMB-005 SM8-00-2

46. Aux i l ia ry feedwa te r pump moto r Westinghouse 4x6x 10-1/2 8 Q MSD-8 SIG

47. Aux i l i a ry feedwa te r pump mo to r con t ro l Bingham- Custom Built QI system Willamette

48. Auxiliary feedwater pump system valve Limi to rque SMB-000 Q actuator

49. Cool ing wa ter system motor Westinghouse HSWF F rame 6808L Q

$0. MSIV pressure t ransmi tter Ancho r/Da rl i ng Rosemount Qi 11$2GP-0-A-92-PB

$1. MSIV. TWlV limit switch Ancho r/Da r l i ng NAMCO E A180 Q

$2 IWlV pressure transmitter Ancho r/Da r l ing Rosemount QB 1 152G P A Pil 23 2

TABLE A-1 (Continued)

Equipment Reviewer Ma nu[aglu re r Model [ valuation

53. fWlV, MSIV solenoid valve '.

Ancho r/Da rl i ng Skinner Qi VSH65590-125VDC VSH65600-125VDC

54. Damper actuator Wa l d i nge r Co rp. ITT. Nit 95 QI

55. Damper actuator Wa ldinger Corp. Fisher 656 Q1

56. Solenoid valve Wa ldinger Corp. ASCO NP8320 QI

51. Limit switches Wa ld i nge r Co rp. NAMCO EA-170 QI

$8. . Ai r handling uni t motors Reliance ---

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59. Ai r handling uni t motors Reliance ---

Q

60. Air handling unit heaters Nutherm ---

QI 61 Ilydrogen recombiner fan motors Westinghouse ---

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62. flydrogen recombiner blower motor Rockwell Reliance QI

63. AC motor opera tors Dre s se r 7NA81-43 Qi industries

64. AC motor operators Limitorque SB-00-7.5 QI

65. Limit switch Na mco EA-180 Q

66. AC and DC motor operators Limi to rque SB-0-40 i SMP-00-25 67 Limit switch Na mco EA-180 Q

68. Solenoid valve ASCO NP831654E Q

69. AC motor operators Limitorque SB-00-7.5 Q

70. Limit switch Namco EA-180 Q

71. Solenoid valve Valcor V70900-37-4 QI 24 S

e IABLE A-2. PALO VERDE NUCLEAR CINEHATINC STATION STEAM SYSILM [QUIPMENT QUAliflCATION STATUS Heviewer

[W8 3 pmerlt ____ Ma nufactu rem r __.!9odel [ya lunj on

-1. Motor operated valve actuator L imi to rque SB-0-25 QI Sil-3-150 SH-000-5 SB-00-10 Sft-0-10 SB-0-25 S0-1-10 4 Sil-1-60 S8-00-15

2. Motor operated valve actuator Limi to rque SH-0-25 ( DC) QI

3. Motor operated va lve actuator Limi to rque SCM-Ois-5 QI SCM-Ola-7-1/2

4. Motor operated valve actuator L imi to rque SMB-00-10 Q1 SMB-1-40 SMB-1-60 SMB-3-100

5. t.imit switch Namco EA180-11302 S, A

6. Solenoid valvo Valcor VS26-5683-45 S. A -

V526-5683-864

7. Solenoid valve Target Rock 71L-001 QI 77t-003

8. Solenoid vasie Target Rock 71L-002 QI

9. Solenoid valve ASCO NP8320A187E S, A 10 Solenoid valve ASCO NP8320A187E S, A

11. Solenoid valve ASCO NP8321A10 Q

12. IIPIS pump motor Westinghouse 5810ll-WP2 Q

13. IPIS pump motor Westinghouse 5010P39-WP2 Q 184 Spray pump motor Westinghouse 5808P39-WP2 Q

15. Cha rging pump motor Westinghouse 1:084 IS-DP Q

16. Pump motor Heliance 18fICilfC-X1 Q 25

TABl0 A-2. ( Con t. i nued )

Equjpme!!L Reviewer Manufitetyst ,

Ngde_1 Lya [ua t f or! ,

17. Level and pressure transmitter I T T Ba rton 7 68: Q1

18. L evel and pressure transmitter ITT Barton 763 QI

19. Pressure transai tter Rosemount 1153AD8 Q1 20 Safety channel detector assembly Westinghouse WL-24036 Q

21. Safety channel preamp ri f ter assembly Westinghouse WL-28037 4 Q

22. Transmitter rack C-E ---

QI

23. Pressure transmitter Rosemount 1153-8 Q

24. RTD RDF Corp. 21245 .

QI k

4 1

26 4 .

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APPENDIX 8 SUMMARIES OF CENTRAL FILE REVIEWS 9

6 27

APPENDIX B SUMMARIES OF CENTRAL FILE REVIEWS ROSEMOUNT PRESSURE TRANSMITTER, MODEL NO.1153GB9 PLANT I.D. NO. J-AFB-PT-17 This transmitter is located in the main steam support structure and provides auxiliary feedwater pump discharge pressure indications to the control room.

The specified accident environmental parameters are: temperature, 5

120*F; relative humidity, 90%; radiation,1 x 10 rads; and an operacility time of 30 days.

Testing of an identical model is recorded in Rosemount Reports No. 108026, No. 108025 Rev. A, and other supplemental reports. Thermal aging was done at 203*F for 69 days, yielding a qualified life of 6.38 years utilizing Arrhenius methodology. Radiation aging was conducted at varying dose rates yielding a total integrated cose of 2.44 x 107 rads T. HELB testing was conducted at a peak temperature of 318*F and a peak pressure of 72'psig.

It is concluded that the Rosemount Model 1153GB9 is qualified for the environmental parameters specified, and that the applicants file is sufficient to support this conclusion.

29

SKINNER V5H65590-125V,DC SOLEN 0ID VALVE PLANT I.D. NO. J-SGB-UY-1328 This solenoid valve, which is an accessory to an anchor darling feedwater isolation valve, is located on the main steam support structure.

The feedwater isolation valves and the main steam isolation valves close upon a high energy line break in the main steam or feedwater lines.

The specified accident environmental parameters are: temperature, 300 F for 15 minutes; pressure, 21 psig for 15 minutes; radiation, 1 x 6

10 rads I; and an operability time of 30 days. No submergence parameter is identified.

Testing and analysis on a similar valve are detailed in Anchor Darling Report No. E9023-QR-2, Rev. A Wyle Test Report 43847-2, and various supplements.

Thermal aging at 250*F for 317.6 hrs was conducted on the O.C.

solenoid, and utilizing arrhenius methodology at an ambient temperature of 120*F with 20*F added for conductive and convective heat addition from the process fluid, a qualified life of 2.72 years was calculated. A check of this qualified life in the qualification maintenance program revealed a changeout period of 2.5 years. Radiation aging of the D.C. solenoid was 5

conducced at 2.7 x 10 rads T per hour for a total integrated dose of >

2.5 x 106 rads T.

After seismic excitation, the assembly was subjected to a steam test at 328*F peak temperature at atmospheric pressure. Pressure testing was not done, although an operability test on another valve showed operability at 21 psig in an air atmosphere. The lack of any pressure testing under a .

ste.ra atmosphere remains an open item.

An operability time of 15 minutes to valve closure was based on the -

maximum time to blowdown one steam generator. No documented investigation 30

of possible valve failure to remain closed after closure and the required

, time was' in the file, however this was addressed satisf actorily during the j audit.

j.

1 It is concluded that the D.C. solenoid valve (and its associated feedwater isolation valve) is not qualified due to lack of pressure testing to the required profile. This remains an open item. In addition, the applicant has several. listed open items.

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DIFFERENTIAL PRESSURE TRANSMITTERS BARTON MODEL 764 These transmitters are located in the containment building and are used to measure steam generator differential pressure and pressurizer level. Three of the steam generator DP transmitters are located below flood level, but are not required during a DBE which would cause submergence.

The environmental conditions to which the transmitters could be subjected are: temperature, 370 F; pressure, 60 psig; relative humidity, 1,00*4; radiation, 3.3 x 107 rads T TID and 1.0 x 108 rads S TID; chemical spray, 4400 ppm H3 80 3 , 50 ppm hydrazine, and pH 7.0-8.5; and ,

operability, 30 min for steam generator DP and 30 days for pressurizer level. ole to the short operability requirements of the steam generator DP transmitters submergence was not addressed.

Environmental testing has been performed on this model transmitter and is reported in Barton Report R3-764-9. The maximum environmental conditions to which the tested transmitters were subjected are: temperature, 420*F; pressure, 75 psig; relative humidity satu ated 8

steam / air; radiation, 2 x 10 rads i TID; chemical spray, 2700 ppm H380 3 and pH 10.5; and operability, 100 days. The transmitters were 5

aged for 1830 hours0.0212 days <br />0.508 hours <br />0.00303 weeks <br />6.96315e-4 months <br /> at 257 F and 5 x 10 pressure variation prior to testing. Aging by the Arrhenius methodology is equivalent to greater than 40 years.

During thermal testing, the transmitters experienced a non-repeatability problem at 320*F which affected the accuracy. Combustion Engineering (CE) is reviewing, (a) the Barton tests and documentation which is used to determine the cause of the problem and (b) the corrective action

, ,. proposed oy Barton to correct the problem. This information is contained in the EQ file.

A leakage current path through the shafts of the zero and span

_ potentiometer mounts to the mounting bracket was discovered by Barton.

This resulted in non-repeatability at 320*F. The solution is to 33

install a fiberglass insulator between the potentiometer shafts and the mounting bracket. The fiberglass material is the same as that used in the -

circuit board so its performance under specified conditions is known and previous tests results were not affected.

After DBE testing one transmitter was discovered to have a broken lead. The broken lead, which was thought to have broken during transit from one test facility to another, was sent to another lab for analysis.

It was determined that corrosion, due to chlorine activated flux, sodium trisulfate, and high temperature caused a decrease in the tensile strength of the upper wire. The test lab suggested that the copper wire external to the transmitter core be replaced with a standard stainless steel wire.

This wire should be welded to the iron base through pins of the gland.

There is no information contained in the file as to what the applicant intends to do.

During the audit the applicant stated a letter was sent to Barton addressing the possible lead wire problem, but this letter was not in the file. Also, the applicant stated the transmitters are securely mounted and lead wires which may become brittle or weakened by corrosion will not

, vibrate or move during operational testing. Therefore the stainless steel solder corrections are unnecessary.

In conclusion the transmitters are qualified for their specified environment pending the outcome of the thermal non-repeatability and the corroded lead problems.

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r WESTINGH0bSE WL-24037 EXCORE SAFETY CHANNEL PREAMPLIFIER / FILTER ASSEMBLY AND WL-24036 EXCORE SAFETY CHANNEL DETECTOR ASSEMBLY

. This equipment is part of the reactor protection system and is used to initiate a reactor shutdown and is located in the containment butiding.

The detector assembly is located at 93 feet 3 inches and the preamplifiers

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and filter assembly is located at 140 feet.

The specified accident environment for the detector assembly is: temperature, 370 F; pressure, 60 psig; relative humidity, steam / air; radiation, 2.5 x 10 6rads TIO (1/4 of 40 year dose); and operability time, 2 minutes. Chemical spray is not applicable.

The test environment was: temperature, 385 F; pressure, 65 psig; relative humidity, saturated steam; radiation, 4.8 x 10 6rads TIO on transition joints, and operability time, 10 minutes. The applicant provided analysis which extended the operability time margin to one hour.

The specified accident environment for the preamplifier is: temperature,120*F; relative humidity, 90*.; radiation, 3 x 10# rads TID. Chemical spray and submergence are not applicable. The operability time is N/A because the preamplifier is not required to function during LOCA, MSLB, or FWLC.

, The test environment was: temperature, 385 F; pressure, 76 psig; relative humidity, saturated steam; radiation, 6 x 10# rads T10 7, 7 x 3

10 rads TIO neutrons, and operability time, 2 minutes. Analysis was used to extend the time margin to I hour.

The specified environment for the filter is: temperature, 370 F; pressure, 60 psig; relative humidity, steam / air mixture; radiation, 3 x

• \ 4 10 rads TID; and operabilty time, 2 minutes.

35

The test environment was: temperature, 376 F; pressure, 62 psig,;

relative humidity, saturated steam; radiation, 6 x 104 rads TIO r, 7 x -

3 10 rads TID neutrons, and operability time 2 minutes. Analysis was used to extend time margin to one hour. -

In conclusion the excore detector and preamplifier / filter assemblies are qualified for their specified environment and documentation is contained in the applicants file.

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ROTORK 7 NAB 1-43 AC MOTOR OPERATORS These motor operators are located in the containment building in the

- auxiliary building. They are used to provide containment isolation in the event of an accident.

The specified environment is: temperature, 370 F; pressure, 60 psig; relative humidity, steam / air mixture; radiation 3.3 x 107 rads TID I, 2 x 108rads TID S; chemical spray, borated water at pH 8.5; and operability time, 30 days. Submergence is not applicable because once the valves operate following an accident, they are not required to operath again.

The~ applicant's review of this file was inadequate, e g., the wrong curve was used for comparison of specified and test environment, and the 10 C rule was used for aging analysis, without justification.

In conclusion, the file does not support qualification of the Rotork AC motor operator.

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m e,. ' 0 " U.S. NUCLEAR REGULATORY COMMISSION BIBLIOGRAPHIC DATA SHEET EGG-EA-6416 4 TITLF ANU SV8 TITLE 2 Ites e oim*/

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. Audit of the Environmental Qualification of Safety-Related Electrical Equipment :t Palo Verde Nuclear 3 RECIPf EN T S ACCESSION NC Generatino Station. Unit 1

. 7 AUTHOHiSi 5 O ATE T'EPORT COVPLE TE D

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R. A. Borgen, E. R. Holloway 3eotember 983 9 Pk H6 ORMING ORGANi2ATION N AME AND M AILING ADDRESS (tacivow 2,0 Cones DATE REPORT ISSUED woNTw lv&&R SeOtember 1983 EG&G Idaho, Inc, s ,c ,,,, ,,,,, , ,

Idaho Falls, ID 83415 8 fatave (Nwal 12 SPONSOHiNG ORGANilAT ON N AME ANf) M A6UNG ADORESS isar/ var 2 0 Coaal 10 PROJE CT T ASK. WOR % UNIT NO Engineering Division of Office of Nuclear Hegulatory desearcn ,, nN No U.S. Nuclear Regulatory Commission A-6415 Washington, DC 20555 IJ T V PE OF RE PO R T u d.,res #

• E m'00 Cov t af D (face ove IS StapPLEVEN TARY NOTES 14 Itse - tw eal I6 AHST H Ai' f (200 orns or 1,*ist Palo Verde Nuclear Generating Station, Unit I was audited to determine the environmental qualification of safety-related electrical equipment. Pesults of the audit are summarized in this report.

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