ML20004C188

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Environ Qualification of Safety-Related Electrical Equipment,Ie Bulletin 79-01B
ML20004C188
Person / Time
Site: Point Beach NextEra Energy icon.png
Issue date: 12/04/1980
From: Jablonski F
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20004C182 List:
References
IEB-79-01B, IEB-79-1B, NUDOCS 8106010715
Download: ML20004C188 (17)
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ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT IEB 79-01B TECHNICAL EVALUATION REPORT DOCKET NO. 50-266 DATED: December 4, 1980 Licensee: Wisconsin Electric Power Co.

Type Reactor: W PWR Plant: Point Beach Unit 1 Prepared by F. J. Jablonski Engineering Support Section Reactor Construction and Engineering Support Branch, RIII i

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CONTENTS Pag Introduction 1

Background and Discussion 1

Summary of Licensee Actions / Statements 1

System Comparison 2

Equipment Evaluation 2 8 Caveat 2

Conclusion 2-3 Attachments:

1 1.

Referenced Test Reports 2.

Onsite Inspection Report t

3a.

Generic Issues 3b.

Site Specific Issues 4.

Licensee System List 5.

NRR's System List 6.

Category Criteria 7.

LER's 8.

Unresolved Generic - Specific Issues 9.

Concurrence Code 1

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e Introduction 1/ or use as input to the This report is submitted in accordance with TI 2515/41 f

Safety Evaluation Report on qualification of Class IE electrical equipment in-stalled in potentially " harsh" er.iironmental areas at this facility.

Background and Discussion IE Bulletin No. 79-012/ required the licensee to perform a detailed review of the environmental qualification of Class 1E equipment to ensure that the equip-ment would function under (i.e. during and following) postulated accident con-ditions.

The Technical Evaluation Report (TER) is based on IE's review of the licensee's submittal for ccaformance with the DOR guildelines or NUREG-0588, a site inspec-tion of selected system components, to EQB'sreviewofcomponenttestreports.37rifyaccuracyofthesubmittal,and Licensee submittals were received on April 18, 1980, September 12, 1980, October 30, 1980.

The site inspection was completed on April 24, 1980. bI G specificguidancewasrequestedfromIE/NRRheadquarters.ggericandsite

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Summary of Licensee Actions / Statements i

The environmental qualification of a number of components could not be completely documented because of the unavailability of detailed equipment qualification records. However, licensee believes the components would perform their safety-related functions under postulated accident conditions. The components include solenoid valves, limit switches, level switches, diaphrams, o-rings, electrical conductor seal assemblies, and a limited amount of electrical cable. These components will be replaced with environmentally-qualified, equivalent equipment. Qualification of pressure and differential pressure transmitters will be accomplished by replacing the original transmitters with environmentally-qualified transmitters.

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1/ Technical Evaluation Report (TER) On Results Of Staff Actions Taken l

To Verify Reactor Licensee Response To IEB 79-01B And Supplemental l

Info rmation.

2/ Environmental Qualification of Class IE Equipment.

'3/ Attachment 1.

{/ Attachment 2.

5/ Attachements 3a and 3b.

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9 System Comparison A comparison was made between the system list provided by the licenseeb/

I and a similar list provided to IE by NRR during a meeting in Bethesda, MD on September 30, 1980. The following systems were not included in the li-censee's submittal.

Safeguards Actuation Main and Auxiliary Steam Isolation Main and Auxiliary Feedwater Isolation Containment Air Purification / Cleanup Containment Combustible Gas Control Accumulator Pressurizer Spray Power Operated Relief Valves Steam Dump Containment Radiation Monitoring Containment Radiation / Sampling Service Water Emergency Power Control Room Habitability Safety Equipment Ventilation Equipment Evaluation ClassIEeggpmentwasevaluated,thatis,placedintofiveseparate categories.- Result of the evaluation follows:

(See pages following)

Caveg Test reports and other documentation which licensees referenced as estab-lishing environmental qualification were reviewed for acceptability by NRR, Environmental Qualification Branch.

(Reference Attachment 3a, memorandum dated June 20, 1980 Hayes to Jordan.)

This TER does not include information about seismic of fire withstand capability.

It should therefore not be inferred that Category I equipment meets all necessary qualification requirements.

Conclusion Based on IE's review of the licensee's submittal, the nite inspection, and licensee's proposed actions, it cannot be concluded that there is reasonable 6/ Attachment 4, 7/ Attachment 5.

8/ Attachment 6. -

assurance all components installed at the Point Beach Unit 1 Nuclear Power Plant are environmentally qualified and installation methods of environmentally qualified coLponents would not contribute to the l'ailure of such components during a potential accident.

A positive conclusion cannot be made until:

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All matters referred to IEHQS/NRR have been satisfied.El 2.

The 15 systems missing from the licensee's submittal have been evaluated by NRR.

(Page 2) 3.

The negative equipment evaluations have been reviewed by NRR.

(Pages 4 thru 8.)

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9/ Attachment 8.

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QUALIFICATION REFERENCES '1 WCAP 7410-L (Volume I & II), Topical Report Environmental Testing of Engineered Safety Features Related Equipn:ent, Wet ti'nghouse Electric Corp., Pittsburgh, PA., Dec., 1970. 2 WCAP 7829, Fan Cooler Motor Unit Test, Westinghouse Electric Corp., Pittsburgh, PA., April,1972. 3 WCAP 7343-L, Topical Report Irradiation Testing of Reactor Containment Fan Cooler Motor Insulation, Westinghouse Electric Corporation, Pittsburgh, PA., June,1969. 4 Qualification of NAMC0 Crotrols Limit Switch Model EA-180 to IEEE Standards 344 ('75), 323 ('74), and 382 ('72), Revision 1, ACME - Cleveland Development Co., Highland Hiights, OH., March 3, 1978. Estimation of Qualified Life of EA180 Series Nuclear Switch, Revision Dated Feb. 27, 1980, NAMC0 Controls, Cleveland, OH. Test Plan For the Qualification of Series EA180 and EA740 Switches For Use In Nuclear Power Plants In Compliance with IEEE Standards 323-74, 382-72, and 344-75, Revision 1, July 26, 1979, NAMC0 Controls, Cleveland 0H. Bechtel Letter From D. H. Clark to D. K. Porter, dated June, 1980, NAMC0 Position Switches 5 Westinghouse Letter From R. L. Korner to W. F. Geisheker with the { following Attachments, dated May 22, 1978, Qualification Data for the Point Beach Nuclear Power Plants Units #1 and #2; 1. PEN-RLK-3-16-01, Accident Environment Test Report 2. PEN-ACD-4-72-03, Accident Environment Test Report 3. ETL Report 5261 Reports of Seismic Tests on Electrical 4. ETL Report 5275 Penetrations for Wes!.inghouse 5. Test Report on Incident Testing of Triax Penetration WEPC0 letter from R. L. Cantrell to T. J. Rodgers, dated March 1,1974, Electrical Penetrations Point Beach Nuclear Plant. WMPCo letter from R. L. Cantrell to Roger Newton, dated April 15, 1968, Electrical Penetrations. WMPCo letter from R. L. Cantrell to A. A. Simmons, Project Manager - Westinghouse, dated September 9,1968, Point Beach Nuclear Plant Electrical Penetrations. Westinghouse letter from A. A. Simmons to Glenn A. Reed, dated October 8,1968, Point Beach Nuclear Plant Electrical Penetrations. Westinghouse Tube Division, Electrical Penetrations Quality Control Production Record Sheet, and attachments. Crouse-Hinds Company Drawing Nos. 0100349, 0100382, 0100411, 0100334, 0100044. Test Report s ATTACHMENT 1

e QUALIFICATION REFERENCES f 6 Westinghouse Letter From R. L. Korner to W. F. Geisheker with Attachments, dated July 28, 1978, Point Beach Nuclear Plant , ualification of Containment Electrical Penetration Safeguards I Q Splices. 1. PEN-TR-78-45, Boric Acid Effect on Medium Voltage Ceramic Seal-Bushing. 2. PEN-TR-78-ll, Statement on Effect of Borated Water on Westinghouse Penetrations for the Angra Nuclear Plant. 3. Brunswick Nuclear Plant Drawing Hos. E-2457, E-2453, E-2452. 4. WEP, WIS Drawing Nos. 31402, 31396, 31400, 150-31393, 150-31394, 150-31396. 7 IE Bulletin 79-01B, Enclosure 4, Appendix C, Table C-1, Nuclear Regulatory Commission, Region III, Glen Ellyn, IL., Jan. 16, 1980. 8 Instruction Manual and Parts List, Fisher Controls Type 546 Electro-Pneumatic Transducer, Fisher Controls Co., Marshalltown, IA., Nov.,1968. Fisher Controls Letter from Bill R. Flowers of W. D. Ehrke Co., Inc., to R. K. Hanneman, dated Sept. 29, 1980, Point Beach Nuclear Plant Environmental Qualification of Fisher Components. 9 WCAP-7354-L, Topical Report Supolier Post Accident Testing of Process Instrumentation, Westinghouse Electric Corp., Pittsburgh, PA., July,1969. ( 10 Westinghouse Letter from C. A. Lins to R. K. Hanneman, dated June 2,1980, Point Beach Nuclear Plant Bulletin 79-018 Motor Qualification. WCAPF 5., Environmental Qualification of Class IE Motors For Nuclear Out-0f-Containment Use, Westinghouse Electric Corp., Pittsburgh, FA., June, 1976. 11 Westinghouse Letter WFP78-531, From R. L. Kelly to W. F. Geisheker with Attachments, dated June 28, 1978, Qualification of Containment Electrical Penetration Safeguards Splices. Westin' house Teletype PBW-B-3070 From N. E. Bush to J. K. Leslie of Bechtel, g l dated February 5,1970, Splicing Information. Bechtel Letter From H. E. Morris to W. F. Geisheker with Attachments, dated April 27, 1978, Point Beach Nuclear Plant Contatinrent Electrical Penetration Splices: 1. Bechtel Drawing SK-E-165, Splicing Requirements for Penetration f Lead Wires. l 2. Bechtel Chronological List of Correspondence. l 3. Bechtel letter PBW-W 2789C From J. K. Leslie to W. B. Henderson l of Westinghouse with Attachments, dated February 10, 1970, Penetration Splices. 4. Bechtel Letter From J. K. Leslie to W. B. Henderson of Westinghouse l with Attachments, dated March 3,1970, Penetration Splices. l l ATTACHMENT 1 L

QUALIFICATION REFERENCES 5. Westinghouse Teletype PBW-B-3179 From N. E. Bush to J. K. Leslie of Bechtel, dated March 5,1970, Safeguard Cable Splices in the Containment. 6. Westinghouse Teletype PBW-B-3211, From N. E. Bush to J. K. Leslie of Bechtel, dated March 13, 1970, Containntent Safectuards Splices. 7. Bechtel Letter PBB-W-2905 From J. F. Leslie to W. B. Henderson of Westinghouse, dated March 17, 1970, Splices for Safeguards Cables Inside Containment 1r Deleted 13 Deleted 14-Boston Insulated Wire & Cable Co. Letter dated April 23, 1980, from L. S. Lisker to R. K. Hanneman. Report B901, BIW Bostrad7 and Bostrad75 - Flame and Radiation Resistant Cables for Nuclear Power Plants, Soston Insulated Wire & Cable Company, Boston, MA., September,1969. 15 Report IPS-348, Test Report - Steam Line Break /LOCA Exposure of Field Cables and Terminal Bl_ocks For American Electric Power Conax Corporation, Buffalo, N.Y., May, 1978. 16 Qualification Type Test Report, limitoroue Valve Actuators For Class IE Service Outside Primary Containment _, Limitorque Corporation Test 8.aboratory, Lynchburg, Virginia, June 7,1976. 17 Robert 0. Bolt and James G. Carroll, California Research Corporation, Radiation Effects on Organic Materials, Richmond, California, Academic Press, New York, 1963. 18 Westinghouse Letter from C. A. Lins to R. K. Hanneman, dated June 2, 1980, P_oint Beach Nuclear Plant, Bulletin 79-01B, Motor Qualification. Westinghouse Letter from C. A. Lins to F.. K. Hanneman, dated August 29, 1980, Point Beach Nuclear Power Plant Ecuipment Qualification NRC Bulletin 79-01B Containment Spray Pump Motors Containment Fan Cooler Motors. WEPCo Letter from R. K. Hanneman to C. A. Lins, dated September 8, 1980, Environmental Qualification of Containment Spray Pumo and Component Cooling Motors at Point Beach Nuclear Plant. Westinghouse Letter from C. A. Lins to R. K. Hanneman, dated October 7,1980, Point Beach Nuclear Power Plant Environmental Qualification of Containment Spray Pump and Component Cooling Pump Motors at Point Beach _ Nuclear Power Plant, with attachment: Westinghouse Research Report 71-lC2-RADMC-R1, Proprietary Class 2, dated December 31,1970 (Revised April 10,1971), The Effect of Radiation on Insulating Materials Used in Westinghouse Medium Motors, by John Bartks, Westinghouse Research Laboratories. ATTACHMENT 1

QUALIFICATION REFERENCES 19 Foxboro Letter from G. Tennesen to R. K. Hanneman, dated August 5,1980, Resistance Temperature Detectors. Installation Instructions and Parts List, "Dynatherm Resistance Bulbs with Aluminum Cap-Type Head, Model DB-1 Series", The Foxboro Company, January 1964 20 Amoco Oil Company Letter from T. M. Warne, dated September 15, 1980, Radiation Resistance of Amoco Oil Lubricants; Project 4210, with attachment: The Effects of Radiation or Lubricants in Nuclear Generating-Stations, by James S. Ferrie, Paul Leinonen, Dr. B. Neil, and E. Wharton (Ontario Hydro Research Division), for presentation at ASLE 35th Annual Meeting, Anaheim, California, May 1980. 21 Mobil Oil Letter from J. Kestly to R. K. Hannemen dated October 13, 1980, Radiation Information Wisconsin Electric Power, with attachments for radiation test data of Mobilgrease 28. 22 Kerite Letter from R. A. Olson to R. K. 'Hanneman, dated October 22, 1980, with attachment: P_oint Beach Nuclear Plant, Wisconsin Electric Power Company, LOCA Qualification of Kerite 600 Volt HTK Insulated, FR Jacketed Power Cable. 2S Rome Cable Corporation Letter from D. D. Sand to R. K. Hanneman, dated April 9,1980. Rome Cable Corporation letter from D. D. Sand to P. R. Belhumeur dated March 19,1971. 24 Okonite Company Letter from J. S. Lasky to R K. Hanneman, dated May 9,1980, With Attachment. Blodgett, R. B. & Fisher, R.G., " Insulations and Jackets for Control and Power Cables in Thermal Reactor Nuclear Generating Stations, IEEE Transactions on Power Apparatus and Systems, Volume PAS-88, No. 5, May 1369. 25 " Type Test Cable Qualification Program and Data for Nuclear Plant Designed Life Simulation Through Simultaneous Exposure", Franklin Institute Research Laboratories, Final Report F-C3694, Philadelphia, Pennsylvania, January 1974. 26 Lancaster, Ron, " Qualification of Safety-Related Equipment Used in Nuclear Power Generating Stations including the Effects of Aging", Reliability Conference for the Electric Power Industry,1980. Carfagno, S. P. & Gibson, R. J., '_'A R_eview of Equipment Aging Theory and Technolog", Electric Power Research Institute, Final Report NP-ISS8, Palo Alto, California, September 1980. ATTACHMENT 1

La.nLD D W.LL I.vCLEM. REGUL A10F.y cort.t.'tLLION i REGION lli b/ / ?. '.. C .. r / 799 ROOSEVELT ROAD "g '... 4 d' GLEN ELLYN. ILLINols 80137 ~ /', October 21, 1980 l l MEMORANDUM FOR: E. L. Jordan, Assistant Director, Division of Reactor Operations Inspection, IE:HQ TERU: G. Fiore111, Chief, Reactor Construction and Engineering Support Branch FROM: D. W. Hayes, Chief, hgineering Support Section 1

SUBJECT:

SCREENING REVIElf 0F LICENSEE RESPONSE TO IEB 79-01B AND

SUMMARY

OF INSPECTION OF INSTALLED SYSTEMS AT POINT BEACH UNIT 1 - DOCKET 50-266 Frank Jablonski has completed the inspection phase at Point Beach Unit 1 in response to IEB 79-01B. A valkdown was conducted on October 10, 1980 to inspect installed components associated with the systems listed on the attachment; all compenents located outside containment. Observations: Motors 8 Motors for residual heat removal and component cooling water, the Auxiliary Coolant System, did not have nameplate data identical to the submittal. Both vere stamped :'.nsulation class "B", neither was stamped "Thermalastic Epoxy" as stated in the submittal. Special insulation qualities were defined in correspondence between licensee and manufacturer, however, the correspondence was not reviewed during the inspection; the submittal vill be corrected. Respective Westinghouse model numbers TBDP and ABDP vere accurate. Motor Operated Valve The operator for the valve supplying component cooling water to the residual heat removal heat exchangers was a Limitorque type SMB-00 with a Reliance motor, insulation class "B"; installation complementary with submittal. Limit Switches. Transmitters NAMCO Snaplock D2400X limit switches installed on the RHR heat exchanger outlet and by-pass valves were scheduled for replacement, because no qualification documentation was available. Onsite Inspection ATTACHMENT 2

l i j Foxboro transmitter Model 611GM, used for measuredent of RHR pump discharge pressure, was scheduled for replacement. Foxboro transmitters Model 613DM, used for measurement of outlet flow from the RER and CCW heat exchangers were scheduled for replacement because j manufacturer specified modifications (MCA) had not been completed. j Resistance temperature devices nameplate data complemented the submittal data. Transducers (Converters) Fisher Controls converters type 546, tsed for controlling the RER heat exchanger outlet and by-pass valves, were being reviewed for qualification. The converters were contained in a NEC Class 1, Group!D enclosure. Miscellaneous All equipment locations were verified to be complete and accurate. as stated in the submittal. Plant equipment identification numbers reported in the submittal were determined to be either as stated, different than stated, or non-existent. For evample, transmitter and pump motor identification were as stated; motor and air operated valves vere different; transducers, limit switches, and resistance temperature devices non-existent. Physical location of the latter components, that is those without plant identification, provided reasonable assurance of correctness. (i.e., equipment was that listed in the submittal.) Several typographical errors were also identified on those submittal sheets thus far redeved. in Qnclusion Except as reported above, the equipment descriptions provided by the licensee on the system component evaluation worksheets for the systems identified were complete and accurate. The licensee was made aware of the apparent discrepancies. A detailed review vill be made by the licensee and the response amended, o&WD. W. Hayes, ylef Engineering Support Section 1

Attachment:

As stated cc: J. G. Keppler G. Fiorelli V.~D. Thomas, IE:HQ Resident Inspector ATTACIH 6 T 2 s,_,___

A d= All Components outside Contain=ent Plant Identification Generic Name-1-P10 1-P11 Pump Motor (RHR) 1-ACT3 Pump Motor (CCW) 1-AC62h Valve Motor operator (CCk Electro-Pneumatic 1-AC626 Transducer (RER) Electro-Pneumatic 1-PT628 Transducer (RHR) 1-PI619 Pressure Transmitter (RHF 1-FT619 Flow Transmitter (RHR) 1-TE627 Flow Transmitter (CCW) 1-TE621 Temperature Element (PJIR) Temperature Element (CCW)- T. TTACEDCTI 2

[g + c UN!1ED ST ATES ~ f* '

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NUCLEAR REGULATORY COMMISSION 2d. E REGION lli g s., [f 4 f 7ee nooseveLT noan GLEN ELLYN. ILLINols 60137 July 23,1980 MEMORANDUM FOR: E. l.. Jordan, Assistant Director, Division of Reactor Operations inspection, IE:HQ THRU: G. Florelli, Chief, Reactor Construction and Engineering Support Branch FROM: D. W. Hayes, Chief, Engineering Support Section 2

SUBJECT:

IEB 79-01B (A/I F03067180) Attached is a copy of a memorandum dated July 17, 1980 received from Frank Jablonski relative to IEB 79-01B. It is being forwarded for your information and solicited guidance. The question of Identification of safety related systems and components (paragraph No. I of the metro) is an old one. I disagree with Frank in j that I feel that this identification is a responsibilltv of the licensee, not the NRC. He must know his plant. I do agree, hwever, that more guidance is needed for our inspectors in this area. This is especially important for those inspectors that have not had reactor operating experience. The significant differences in master IIsts that Frank discusses in paragraph two does raise questions. We can only compare these lists against the SAR. Review and evaluation beyond this is assumed to be an NRR function. In regard to Frank's question - should we assume the licensee's response to IEB 79-01B to be complete and correct - I have told him yes. Further, that if he identifies significant incompleteness in the response, or incorrect Information during his reviews, to bring these to my attention so appropriate action can be recommended. Coments and further guidance is requested concerning matters discussed in paragraphs 3 and 4 of Frank's memo.

  1. 7" D. W. Hayes, Chief Engineering Support Section 2 Generic Issues ATTACHMENT 3a

[Dl?J l O D O

E. L. Jordan 2 July 23, 1980 I i i

Attachment:

F. J. Jablonski Memo to D.W. Hayes dtd 7/17/80 cc w/ attachment: J. G. Keppler, Rl l i V. D. Thomas, IE:HQ A. Finkel, RI R. Hardwick, Ril D. Mcdonald, RIV J. Elin, RV R. F. Hel shman, Ri l l -> F. J. Jab lonski, Rl l i i l t ATTACHMENT 3a

![g= =eq 'e, UNITED STATES 7, NUCLEAR REGULATORY COMMJSSION { E ~ REGION lli o 8 799 ROOSEVELT ROAD E 'U j GLEN ELLYN. ILLINols 60137 July 17, 1980 9 MEMORANDUM FOR: D. W. Hayes, Chief Enginee-ing Support Section 1 FROM: F. J. Jablonski, Reactor Inspector

SUBJECT:

FORMULATING TECHNICAL EVALUATION REPORTS (TER) - REVIEW OF lEB 79-01B RE: MEMO TO YOU DATED JUNE 16, 1980 - SAME SUBJECT Since the review of IEB 79-01B is continual, new discrepancies continue to show up; discrepancies are not necessarily the licensees'. As you know, there is no specific nuclear power plant design required by NRC. Further, the designation of safety related systems is somewhat arbitrary and inconsistent. In fact, the NRC places responsibil'.ty for classifying safety related systems on the licensee. Action item No. I of 79-018 requested each IIcensee to provide a " master list" of all ESF systems in their respective plant required to function { during a postulated accident. Appendix A to 79-01B lists " typical" } equipment / functions needed for mitigation of an accident. A comparison of master ilsts was made of four licensees with simliar Westinghouse PWRs (see Attachment 1). Arbitrary selection and non-standard nomenclature ( of systems makes evaluation of the master lists extremely difficult. NRCt requested each IIcensee to submit the Information under oath. Shouldthek information therefore be assumed complete and correct? j lt is extremely frustrating to review responses which vary so much in attention to detail, depth of review, etc. As stated previously in the draft TER for D.C. Cook, because I as a principal reviewer lack detailed systems / operations experience, further guidance is requested. Another TER related matter is motor Ized velves equipped with Limitorque operators (see Attachment 2). As can be seen, each test report is for a scecific unit type including motor type and insulation class. Almost all licensees refer to the various test reports as qualification documentation for all series of operator types; never is name plate data e provided. For example, test report No. 600456 (SMP.-0-40, Reliance Motor with Class RH insulation) may be listed for all operators from series SMB-000 to SMB-5; motor nam plate data not provided. Without the name plate data and the basis for extrapolation, a meaningful evaluation cannot be made. ATTACHMENT 3a hAN 1.2 &Sb \\DO y i

1 D.W. Hayes July 17, 1980 it is requested that this memorandum he forwarded to IE:HQS as an addition to A/l F03067180 with the same copy distribution. j' 5 F. J. Jablonski Reactor inspector Attachments: 1. Comparison of Master Lists 2. Motor Operated Valve Tests cc: J. G. Keppler G. Flore111 i 1 ATTACHMENT 3a

ATTACHVZNT 1 SYSTEMS P.I. G,QQL E PT. BCH. Aux. F.W. X X X Chem. & Vol. Cont. X 2 X X Cntmt. Air Hndtg. X X X Cntmt. H Cont. X X 7 Cntet. S5. X X 1 Main Stm. X X X Aux. Stm. X Stm. Dump X Rx CLnt. X X X X 3 Res. Ht. X 2 X 3 Saf. Inj.gm. X 2 X X CLg. Water X Esnt'L. Serv. Wat. X Comp. CLg. Vat. X 3 Aux. CLnt.g CLg.2 Emerg. Cor 1 X 1 X Cntmt. Purge X Rx. Bldg. Vent X Inst. & Prot. X Rx. Trip. Act. X Rx. Cont. & Prot. X Rad. Monit. X Rx. Hot Samp. X Stn. & Inst. Air X Stm. Gen.BD X Post Acc. Monit. X Rem. Sht. dn. Monit. X Cntmt. Isol. X X Mn. Stm. Isol. X Mn. FW Isol. X l l l l ) ATTACHMENT 3a

~ ATTACHMENT 2 MOTOR OPERATED VALVES NOV's 1. There are basically two type series of Limitorque operators: SMB and SB. The operators are sized from 000 (smallest) to 5 (Largest) as follows: SMB-000, Sm-00 SM/SB-O' ] This series may SMB/SB-1 This series may also also include WB y SM/SB-2 include SB SM8/SB-3 ~\\Thisseriesmay SMB/SB-4 SMB-5 s f be suffixed "T" 2. Test Reports include: Report No. Date Unit Type Environment Motor Type Insulation

a. 600198 1-2-69 SMB-0-15*

PWR Reliance SpeciaL Hi No Radiation Temp

b. 600426 4-30-76 SM-0-25*

BWR Peerless H 7 (B-0009) 1x10 R DC 0 340

c. 600376A 5-15-76 SMB-0-25*

BWR Reliance RH 8 FIRL F-C 2x10 3441

d. 600456 12-9-75 SMB-0-40*

PWR Relian ce RH g 2x10

e. 600461 6-7-76 SMB-0-25*

Outside Reliance B Cntmt7 2x10 l

f. WCAP7410L 12-70 SMB-00 B

7744 8-71 l

  • denotes foot pounds of torque only SMB-0 has been tested seismically Re: a,b,c ATTACHMENT 3a

I !)pareg f jo UNITED STATES g NUCLEAR REGULATORY COMMISSION y., o E f wAssincTON. D. C. 20555 s a %,,,. / SSINS #6820 JUL 3 1980 . MEMORANDUM FOR: I. R. Rosztoczy, Branch Chief, Equipment Qualification Branch, Division of Engineerir.g. NRR M THRU: E f . L. Jordan, Assistant Director for Technical Programs, Division of Reactor Operations Inspection, LE FROM: V. D. Thomas, Task Manager, Review Group, IEB 79-01B. Divi '..: of Reactor Operations Inspection. IE

SUBJECT:

REQUEST FOR NRC POSITIONS ON REVIEW QUESTIONS OF IEB-79-OlB LICENSEE RESP 0NSES In accordance to our verbal agreement, we would be happy if you would provide positions on the questions noted in the enclosed memoranda. Since it is essential to establish a unifenn approach to the review effort to obviate the questions being generated in the on-going review of licensee responses, we will be happy to meet with your staff to discuss these concerns to expedite resolution of the issues. l fY VincentD. Thomas]TaskManager Review Group IEB 79-01B

Enclosures:

1. Memo D. W. Hayes to G. Fiorelli, RIII dated June 20, 1980. 2. Memo F. Jablonski to D. Hayes, RIII dated Jun 16, 1980. 3. Memo F. Jablonski to D. Hayes, RIII DATED June 10, 1980. cc: w/ enclosures E. L. Jordan, IE V. S. Noonan, NRR G. Fiorelli, HIII g D. W. Hayes, RIII A. Finkel, RI R. Hardwick, RII

f. Jablonski, RIII D. Mcdonald, RIV J. Elin, RV dul 71980 ATTACHMENT 3a dnP 9001030 Lzq

i,... c UNITED ST ATES ^,

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7

( NUCLEAR REGULATORY COMMISSION j REGION lli O, g g 799 ROOSEVELT RoAo [ GLEN ELLYN. ILLINots 60137 June 20, 1980 MEMORANDUM FOR: E. L. Jordan, Assistant Director, Division of i Reactor Operations inspection, IE:HQ hA. Flore111, Chief, Reactor Construction and THR. O Engineering Support Branch FROM: D. W. Hayes, Chief Engineering Support Section 1

SUBJECT:

IEB 79-01B (A/I F03067180) Attached are two memorandums from one of my inspectors, Frank Jablonski. The first is dated June 10, 1980 and the second June 16, 1980. Both memos raise basic que<,tions for which we require guidance to complete our review of responsas to IEB 79-018. By this mer.o I also would like to confirm our understanding that NRR (Environmental qualification Branch) will revi'w for acceptability all test reports and other documentation which licensees reference as establishing environmental qualification of Instrument / electrical equipment. In connection with this, we are sending under separate cover test reports, etc. In our possession to be forwarded to the Environmental qualification Branch. (We further understand that the IEB 79-01B task group, on a volunteer basis, may agree to review some of these documents). l The status or schedule for site inspections and review / evaluation of the final reports is also attached. Please note that every IIcensee has asked for some sort of time extension to submit their first report. We understand that the other regions have had similar reporting problems. Assuming that all our IIcensees meet their extended submittal dates, we should complete our site inspections, reviews, and technical evaluation l l ATTACHMENT 3a A 900 fML5 L

E. L. Jordan 2 June 20, 1980 reports by the end of December 1980. Further delays in the submittals or any unforeseen events will hamper our ability to rneet the new February 1,1981 deadline. .s & s D. W. Haye, Chief Engineering Support Section 1 Attachments: 1. Memo F. Jablonski to D. Hayes 6/10/80 2. Memo F. Jablonski to D. Hayes 6/16/80 3 Inspection Status / Schedule 4. " Separate Cover" List (Test Reports Sent to IE:HQ) - Separate' Cover: See Attachrent 4 cc w/ attachments 1, 3, & 4 only: J. G. Keppler G. Fiorelli l V. D. Thomas, IE:HQ l A. Finkel, R1 R. Hardwick, Ril D. Mcdonald, RIV J. Elin, RV R. F. Heishe.an I i s i ATTACHMENT 3a

./ g- 'i UNITED STATES -f,.k;g,'s.,g,g [Y NUCLEAR REGULATORY COMMISSION gy '(,,., c REGION ill g v7f 799 ROOSEVELT ROAD

  • ...f GLEN ELLYN,ILLINots 60137 June 10, 1980 MEMORANDUM F0R:

D. W. Hayes, Chief, Engineering Support Section 1 FROM: F. J. Jablonski, Reactor Inspector

SUBJECT:

EFFECT OF PREVIOUS NRR REVIEW ON MATTERS RELATING TO IEB 79-018 In almost every licensee response to IEB 79-018 there is a subtle or direct reference to matters apparently reviewed by NRR. Because of the referenced dates it is assumed by me that NRR has given either tacit or direct approval to the references; examples follow: 1. ALL licensees refer to their FSARs for establishing the list of engineered safety feature systems and environmental data such as temperature, pressure, radiation, etc. 2. One licensee, Wisconsin Public Service Corporation, states that "The AEC, in their " Safety Evaluation of the Kewaunee Plant", Section 7.5, issued July 24, 1972, concluded that our criteria and testing program for environmental qualification were adequate". It is further stated that "Our FSAR, which was approved by the AEC, discusses at tenjth the post accident conditions and required qualifi-cations for applicable equipment. (See Section 7.5 of the Kewaunee FSAR.)" 3. Two licensees, American Electric Power and Wisconsin Public Service Corporation, have discussed the effect of-components below flood level simply by referencing letters previously 8 cubmitted to the NRC, or FSAR questions / answers as follows:

  • AEP

-- Letter dated 9-29-75 from Tillinghast (AEP) to Kniel (NRC); FSAR question 40.10 Appendix Q.

  • WPSC Letter dated 2-2-76 from James (WPSC) to Purple (NRC).

y toow,in-s} ATTACHMENT 3a

t June 10, 1980 2 D. W. hayes My specific concerns are: Is it to be assumed that the referenced FSAR parameters, No' 1 above, are correct, i.e. reviewed by NRR7 If the answer is yes, then should it also be assumed that No. 2 above is likewise adequate? (If the answer is no, then none of the Licensee responses which reference the FSAR can be assumed to be correct.) Reference No. 3, even though a component may not be required to - operate subsequent to flooding, what effect will short circuits have on containment electrical penetrations? Was this considered by NRR? I am requesting that these questions / concerns be forwarded to the Assistant Director, Division of Reactor Operations Inspection for resolution. dd / F. J. Jablonski Reactor Inspector cc: J. G. Keppler G. Fiorelli l l l I ATTACHMENT 3a

p, UNITED ST ATES 5, NUCLEAR REGULATORY COMMISSION 3- ~ REGION lli o, h [ 799 roosevelt ROAD gv g CLEN ELLYN. ILUNols 60m June 16, 1980 4 MEMORANDUM FOR: D. W. Hayes, Chief. Engineering Support Section 1 FROM: F. J. Jablonski, Reactor Inspector

SUBJECT:

FORMULATING TECHNICAL EVALUATION REPORTS (TER) - REVIEW OF lEB 79-01B in accordance with IEB 79-01B, an overall conclusion relative to the qualification of Instrument electrical equipment is to be made for each operating plant based on a screening review of all plant systems, and by a detailed review and ob.servation of specific system coments. Unresolved concerns previously

  • Identified by Rill Inspectors during reviews of IEC 78-08 and IES 75-01 along with subsequently identi?ied concerns make it difficult for us to formulate meaningful TERs for certain plants.

The previous unresolved concerns are documented in the memorandums IIsted below (1,2,3) and are relterated in Attachment A to this memo. Subsequently identifled concerns are IIsted in Attachments B, C, and D. To assure uniform evaluation, guidance is needed for these items. Please forward these concerns to IE:HQ. 1. Tl 2515/13 - Qualification of Safety Kelated Electrical Equipment Flore111 to Snlezek, 10/13/78 2. Same title es I., Flore111 to Klinger, 12/78 l 3 Review Status of Responses to IEB 79-01, Hayes to Jordan, 9/5/79 f~ G$$wh(s' 1 F. J. Jablonski Reactor inspector l l

Enclosures:

As Stated cc: J. G. Keppler G. Floreill l V. D. Thomas, IE:HQ A. Finkel, Rt R. Hardwick, Ril D. Mcdonald, RIV J. Elin, RV ATTACHMENT 3a n + w u m S N'ss 7s M t 9 3 y 5

~T,;~ii2M. P ~ ~ UtHTED STATES V pr l$," *ff I f * ;,,,47j NUCLEAR t'EGULATORY COMMISSION ofo C.~' l .g pg/ p W ASNINGTON. D. C. 20555 A/D IIPAo $,1 c " g \\, i / FFct'.S II5LU SEP 1 1 1930 - / nCCES l} ~ ROLM5 pggnLE A1 a Dacket No. 50-266 -301 MEMORANDUM FOR: D. W. Hayes, Chief Engineeri.ng Support Section 1, Region III FROM: E. L. Jordan, Assistant Director for Technical Programs, Division of Reactor Operations Inspection, IE

SUBJECT:

LACK OF SEPARATION CRITERIA AT POINT BEACH (AITS Your request for a review of the Point Bea,ch discussion of GDC-4 was located 'on pages 4.1-4 thru 4.1-5 of the FSAR and mislabeled as GDC-40. A copy ' is enclosed. The response to your request for a definition of a high energy line is included in the enclosed memorandum to E. L. Jordan from D. G. Eisenhut, " Lack of Separation Criteria at Point Beach" September 3,1980. This memo-randum, also, outlines the NRR program on the review of pipe break criteria. Action Item F03059680 is closed. j . Jordan, Assistant Director ar fo echnical Programs d q. p.. Division of Reactor Operations Inspection

Enclosures:

1. Point Beach FSAR pages 4.1-4&5 2. Memo D. G. Eisenhut to E. L. Jordan dated September 3, 1980 cc: J. Fair, NRR RONS Regional Branch Chiefs CONTACT: H. A. Wilber, IE 49-28180 .r Site Specific Issues NI2ACIcmiT 3b 81DlO 200 Eb

"icsile Protection Criterion: Adequate protection for those engineered safety features, the-failures of which could cause an undue risk to the health and safety of the public, shall be provided against dynamic effects -/ ~ and missiles that might result from plant equipment failures. (GDC 40) y ~ The dynanic effects during LIowdown following a loss-of-coolant accident are evaluated in the detailed layout and design of the high pressure equipment and barriers which afford missile protection. Fluid and mechanical driving forces are calculated, and consideration is given to possible damage due to fluid jets and secondary missiles which might be produced. The steam generators are supported, guided and restrained in a manner which prevents rupture of the steam side of a generator, the ' steam lines and the feedwater piping as a result of forces created by a Reactor Coolant System pipe rupture. These supportrf guides and restraints also prevent rupture s s 5 of the primary side of a steam generator as a result of forces created by a steam or feedwater line rupture, s, .s.. e The mechanical consequences of a pipe rupture are' restricted by design such that the functional capability of the engineered safety features is not impaired. o ATTACICErr 3b

f 'l DISTRIBUTION: Central Files-SEP, 3 1930 ORAB RDG N5W Lem,ac,um ~~ MEMORAliotM FOR: Edward L. Jordan, Assistant Director for Technical j Programs, Division of Reactor Operation Inspection FROM: Darrell G. Eisenhut, Director Division of Licensing 4 S!!BJECT: LACK OF SEPARATION CRITERIA AT POINT' BEACH 2 l l

REFERENCE:

Mernorandist E.L. Jordan to D.G. Eisenhut, dated August 1,1980. In response to your request in the above referenced memorandum, we have reviewed the pipe break criteria that was presented and have investigated the current licensing activities in the area of pipe break criteria. You 4 specifically requested that we review the interpretation of the criteria-for a high energy system and provide a schedule for any proposed actions that may be necessary. p We have determined from our review of the pipe break criteria that the j cor: ect interpretation of a high energy line is a system where either y the fluid temperature is greater than 200*F or the fluid pressure is 4 greater than 275 psig. However, Appendix A of APCSB 3-1 defines a high energy fluid system as a fluid system that during normal plant conditions il meets the temperature ort pressure limits. Therefore, those portions t of systems such as ECCS systems that do not exceed the temperature or j pressure limits during normal operation 1d not be classified as high energy. This should resolve the Regio neerns for the safety ll injection lines in the pump room. n. N=J c.] As part of the NRC's Systematic Evaluation Program, the pipe break I criteria for the SEP plants was reviewed. The results of this reYieW revealed. inconsistent application of the pipe break criteria for both .[. inside and outside containment applications (see enclosed memo). As a result of this study a gcneric letter to the SEP licensees has been - - r prepared to address the application of pipe break criteria inside the containment. A generic letter to all other licensees addmssing pipe break criteria inside the containment is currently, planned. Resolution of the pipe break criteria outside the containment is pending the results of the inside the containment reviews. h original stEned W parrell c. nsed .{ i Darrell G. Eisenhut, Director gl00300fz) Division-of Licensing contact: J. Fair, X27357 0,,9C. W ,y_ g ycromp.y m maem d.......... r....m...

t... -.~ "

Amcan 3d , ~ _ _ _ _ _ _ - -, - - - - - m.

e s L"... E D f.14.1 E !., NUCLE AR RE GL'L A T OR Y CO*.'. MISSION { z( j lc nL GioN it g, g, g 799 RooSE VELT ROAD e GLEN ELLYN,ILLINots 60137 May 15, 1980 MEMORANDUM FOR: E. L. Jordan, Assistant Di rector,-Division of Reactor Operations inspection, IE:HQ THRU: . Florelli, Chief, Reactor Construction and Engineering Support Branch FROM: D. W. Hayes, Chief, Engineering Support Section 1

SUBJECT:

LACK OF SEPARATION CRITERIA AT POINT BEACH (A/l F03059680) Ref: April 24,1980 memorandum to D. W. Hayes f rom F. J. Jablonski, same subject (copy attached) Per our discussion, please advise us of the NRC position relative to l the matter discussed in Mr. Jablonski's memorandum. Our cursory review of the FSAR for Point Beach did not locate where overall plant requirements per GDC-4, " Environmental and Missile Design i Basis", and GDC-5, " Sharing of Structures, Systems and Components" were l discussed. However, in regard to the containment spray and safety injection pumps Figure 1.2-5 shows that these pumos for both Ur.its 1 and 2 have a common location without separation. In connection with the safety injection (SI) pumps, we would also like a clarification of what constitutes a high energy line. Our interpretatier. f rom Regulatory Guide 1.46 is that both a pressure above 275 psig and a temperature above 200 F must exist. Specifically, are the Si pump,F discharge lines which operate at about 1500 psig and less than 200 l considered high energy lines? l l No further review or action on our part is planned pending receipt of l your response. J O. W. Hayes Chief, Engineering Support Section 1 cc: l J.G. Keppler G. Fiorelli R.F. Heishman ATTACIDENT 3b R.F. Varnick F.J. Jabicr. ski -g dAA NO 82.2 o E N

U N n i v : : A I E:, NUCLE AR REGUL AT")RY CO?.'.P.'. LESION .**b,..)l I REGION lli

e.,,,

p'~ 799 ROOSEVE LT RO AD GLEN E LLYN,it LINOts 60137 4 f s a.... May 15, 1980 --3 MEMORANDUM FOR: D. W. Hayes, Chief. Engineering Support Section 1 FROM: F. J. Jablonski, Reactor inspector SUBJECT : POTENTIAL DISCREPANCY WITH CRITERIA USED IN THE SER OF HIGH ENERGY LINE FAILURE AT POINT BEACH RE: MEMO HAYES-JORDAN MAY 2,1980 On May 14, 1980 I had a discussion wi th Mr. C.J. DeBevec, IE:HQ, regarding the above. Mr. DeBevec explained that Regulatory Guide 1.46 states only what a high energy piping system is not.

Further, Mr. DeBevec's understanding of AEC meetings held several years ago about the same subject confirms no discrepancy exists with criteria used in the SER et the Point Beach Nuclear Power Plant.

(HI energy - where the temperature and pressure conditions of the fluid exceed l 200 F and 275 psig). 'T. O -47' G J-Gb l,.w j (, F. J. Jablonski Reactor inspector Engineering Support Section 1 cc: J.G. Kepple r G. Fiorelli j J. Smlth V.D. Thomas, IE:HQ C.J. DeBevec, IE:HQ l ATTACIDErf 3b { tbio 200.L4

or.s1 L u c ; t. t L i, g WUCL.E AR R EGUL4 T ORY COM.'.ilSSION - \\ g. p REGION lli

  • y. c,j (g 799 ROOSEVELT ROAD

.....f oleo' E LLYN. ILLINols 60137 April 24, 1980 /\\, \\ I '\\ MEMORANDUM FOR: b., W Hayes, Chief, Engineering Support Section 1 a FROM: F. J. Jablonski

SUBJECT:

LACK OF SEPARATION CRITERIA AT POINT BEACH During my trip to Point Beach on April 21, 1980 relative to IEB 79-01B, I ebserved what appeared to be a total lack of separation criteria. NOTE: Only the SI system was observed. For example the SI and CS pumps for both Units 1 and 2 share a common room without any separation between redundant pumps of Unit 1 or 2, or between Unit 1 and 2. ~ The same condition exists at a different elevation for the Spray Additive Tanks. Another example is the use of a single penetration for the passage of i redundant cables used for indication of containment sump level. Separation is beyond the scope of IEB 79-01; therefore there is a need l ( for a separate memo. I realize separation criteria may have been different 15 years ago; however, these observations should be documented. h6 F. J. Jablonski Reactor Inspector Engineering Support Section 1 cc: J. G. Keppler E. L. Jcedon, IE:HQ ATTACIDENT 3b b g gC w w-w wc. w m "=" -ma}}

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