ML20004C191

Environ Qualification of Safety-Related Electrical Equipment,Ie Bulletin 79-01B

ML20004C191
Person / Time
Site:	Point Beach
Issue date:	12/04/1980
From:	Jablonski F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20004C182	List: ML20004C181 ML20004C186 ML20004C188 ML20004C191
References
IEB-79-01B, IEB-79-1B, NUDOCS 8106010719
Download: ML20004C191 (20)

Text

O O

ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT IEB 79-01B TECHNICAL EVALUATION REPORT DOCKET NO. 50-301 DATED: December 4, 1980 Licensee: Wisconsin Electric Power Co.

Type Reactor: W PWR Plant: Point Beach Unit 2 Prepared by F. J. Jablonski Engineering Support Section Reactor Construction and Engineering Support Branch, RIII i

n 82 es02 o 7 q

CONTENTS Page 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.

Referenced Test Reports 2.

Onsite Inspection Report i

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 i

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Introduction This report is submitted in accordance with TI 2515/41 / for use as input to the I

Safety Evaluation Report on qualification of Class IE electrical equipment in-stalled in potentially " harsh" environmental 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 IE 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 conformance with the D0R guildelines or NUREG-0588, a site insper-tion of selected system components, to rify accuracy of the submittal, and EQB's review of component test reports Licensee submittals were received on April 18, 1980, Septemi;er 12, 1980, October 30, 1980.

The site inspection was completed on April 24, 1980. S specificguidancewasrequestedfromIE/NRRheadquarters.gericandsite G

Summary of Licensee Actions / Statements The environmental qualification of a number of components could not be completely documented because of the unavailability of detailed quipment 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 ar. ant of electrical cable. These components will be replaced with enviror. mentally-qualified, equivalent equipment. Qualification of pressure and differential pressure transmitters will be accomplished by replacing the original transmitten with environmentally qualified transmitters.

If Technical Evaluation Report (TER) On Results Of Staff Actions Taken To Verify Reactor Licensee Response To IEB 79-01B And Supplemental Information.

2/ Environmental Qualification of Class IE Equipment.

3/ Attachment 1.

4_/ Attachment 2.

l S/ Attachements 3a and 3b.

I

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

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

Safeguards Actuation Main and Auxiliary Steam Isolation Main and Auxiliary Feedwater Isolation Containment Air Purification / Cleanup Co ainment 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 Class IE enyipment was evaluated, that is, placed into five seperate categories.2' Result of the evaluation follows:

(See pages following)

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

(Reference Attac> ment 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 site inspection, and licensee's proposed actions, it cannot be concluded that there is reasonable 6/ Attachment 4, 7/ Attachment 5.

8/ Attachment 6.

F assurance all components installed at the Point Beach Unit 2 Nuclear Power Plant are environmentally qualified and installation methods of environmentally qualified components would not contribute to the failure of such components during a potential accident.

A positive conclusion cannot be made until:

I 1.

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

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

(Page 2) 3.

The negative equipmeat evaluations have been reviewed by NRR.

(Pages 4 thru 8.)

9/ Attachment 8.

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QUALIFICATl_0f( REFEREf1CES 1 WCAP 7410-L (Volume I & II), Topical Report Environmental Testing of Engineered Safety Features Related Equipment, Westinghouse 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, Top _i_ cal Report Irradiation Testing of Reactor Containment Fan Cooler Motor Insu1ation, Westinghouse Electric Corporation, ~ ~ Pittsburgh, PA., June, 1969. 4 qualification of ftAMC0 Controls Limit Switch Model EA-180 to IEEE Standards 3T4 TiSR23 ( 747, and 3'82Tr2]7. Revision 1, ACME - Cleveland Development Co., Highland Heights, OH., March 3, 1978. Estimation of Qualified Life of EA180 Series f;uclear Switch, Revision Dated Feb. 27, 1980,fiAMC0 Controls, Cleveland, OH. Test Plan For the Qualification of Series EA180 and EA740 Switches For Use In fiuclear Power Plants In Compliance with IEEE Standards 323-74, 38_2-72, and 344-75, Revision 1, July 26, 1979, fiAMCO Controls, Cleveland, OH. Bechtel Letter From D. H. Clark to D. K. Porter, dated June,1980, !!AMC0 Position Switches 5 Westinghouse Letter From R. L. Korner to H. F. Geisheker with the f following Attachments, dated May 22, 1978, Qualification Data for \\ the Point Beach fluclear Power Plants Units #1 and #2; 1. PEll-RLK-3-16-01, Accident Environment Test Report 2. PEft-ACD-4-72-03, Accident Environment Test Report 3. ETL Report 5261 Reports of Seismic Tests on Electrical 4. ETL Report 5275 Penetrations for Westinghouse 5. Test Report on Incident Testing of Triax Penetration WEPC0 letter from R. L. Cantrell to T. J. Rodgers, dated March 1,1974, l Electrical Penetrations Point Beach ?!uclear Plant. l WMPCo letter from R. L. Cantrell to Roger flewton, dated April 15, 1968, Electrical Penetrations. WMPCo letter from R. L. Cantrell to A. A. Simmons, Project Manager - t Westinghouse, dated September 9,1968, Point Beach fluclear Plant Electrical l Penetrations. Westinghouse letter from A. A. Simmons to Glenn A.. Reed, dated October 8,1968, Point Beach !!uclear Plant Electrical Penetrations. Ucstinghouse Tube Division, Electrical Penetrations Quality Control Production Record Sheet, and attachments. Crouse-Ilinds Company Drawing fios. 0100349, 0100382, 0100411, 0100334, 0100044. 9 9 T Reports P001 BRIGINAL est

j;.AllFICATION REFERLflCES / 6 Westinghouse letter From R. L. Korner to W. F. Geishcher with Attachnents, dated July 28, 1978, Point Beach fluclear Plant Qualification of Containment Electrical Penetration Safeguards ' Splices. 1. PEft-TR-78-45, Boric Acid Effect on Medium Voltage Ceramic Scal-Bushing. 2. PEN-TR-78-11. Statement on Effect of Borated Water on Westinghouse Penetrations for the Angra fluclear Plant. 3. Brunswick fluclear Plant Drawing flos. E-2457, E-2453. E-2452. 4. WEP, WIS Drawing fios. 31402, 31396, 31400, 150-31393, 150-31394, 150-31396. 7 IE Bulletin 79-018, Enclosure 4, Appendix C. Table C-1, Nuclear Regulatory Commission, Region 111, 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 ?!uclear Plant Environmental Qualification of Fisher Components. 9 WCAP-7354-l, Topical Report Supplier 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. WCAP875., Environmental Qualification of Class IE Motors For Nuclear _ Out-0f-Containment Use, Westinghouse Electric Corp., Pittsburgh, PA., 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'ghouse Teletype PBW-B-3070 From N. E. Bush to J. K. Leslie of Bechtel, 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 Contatinment Electrical Penetration Splices: 1. Bechtel Drawing SK-E-165, Splicing Required.ents for Penetration lead Wires. 2. Bechtel Chronological List of Correspondence. 3. Bechtel letter PBW-W 2789C From J. K. Leslie to W. B. Henderson of Westinghouse with Attachcents, dated February 10, 1970, _ Penetration Splices. 4. Bechtel Letter From J. K. Leslie to W. B. Henderson of Westinghouse i with Attachments, dated March 3, 1970, Penetration Splices. i j P00R ORIGINAL

QUALIFICAT10flREFEREf1CES ~ 5 Westinghouse Teletype PBW-B-3179 From it. E. Bush to J. K. Lesite of Bechtel, dated fiarch 5,1970, Safeguard Cable Splices in the _ Containment. 6. Westinghouse Teletype PBW-B-3211, From N. E. Bush te J. K. Leslie of Bechtel, dated March 13, 1970, Containn.ent Safenuards Splices. 7. Bechtel Letter PBB-W-2905 From J. K. Leslie to W. B. lienderson-of Westinghouse, dated March 17, 1970, Splices for Safeguards Cables Inside Containment 12 Deleted 13 Deleted 14 Boston Insulated Wire & Cable Co. Letter dat,ed April 23, 1980, from L. S. Lisker to R. K. Ilanneman. Report B901, BIW Bostrad7 and Bostrad7S - Flame and Radiation Resistant, Cables for Nuclear Power Plants, Boston Insulated Wire & Cable Comp ay, Boston, MA., September, 1969. 15 Report IPS-348, Test Report - S_ team Line Break /LOCA Exposure of Field Cables and Terminal Blocks For American Electric Power, Conax Corporation, Buf falo, ft.Y., May,1978. ~ 16 Qualification Type Test Report, limitorque Valve Actuators For Class IE ' Service Outside P'rimary Containment,137nftorque Corporation Test Laboratory, Lynchburg, Virginia, June 7,1976. 17 Robert O. Bolt and James G. Carroll, California Research Corporation, Radiation Effects o_n Organic Materials, Richmond, California, Academic Press, New York, 1963. 18 Westinghouse letter from C. A. Lins to R. K. Ilanneman, dated June 2, 1980, P_oint Beach fluclear Plant, Bulletin 79-018, Motor Qualification _. Westinghouse letter from C. A. Lins to R. K. Hanneman, dated August 29, 1980, Point Beach fluclear Power Plant Eouipmentlualification NRC Bulletin _79-01B Conta_i_nment Spray Pump Motors Containment Fan thoTeFilotors_. WEPCo letter from R. K. Itanneman to C. A. Lins, dated September 8, 1980, Envi_ronmental Qualification of Containment Spray Pump and_ ~~ Component Cooling IEtors at Point Beach fluclear Plant. Westinghouse letter from C. A. Lins to R. K. Hanneman, dated October 7,1980, Point Beach fluclear Power Plant Environmental _ ualificatic_n _of Conta_inment Spray Pump and Component Cooling Pump _ QMotors at Point _ Beach fluclear Power Plant, with attachment: Westinghouse Research Report 71-lC2-RAD'iC-R1, Proprietary Class 2, dated December 31, 1970 (Revised April 10,1971), The Effect of . Radiation on Insul_ating Materials __Used in Westinghouse Ifedium t Motors, by John Bartks, Westinghouse Research Laboratories. ATTACHMENT 1 POOR ORIGINAL

O QUAllFICAT10fl REFEREllCES 19 foxboro Letter from G. Tennesen to R. K. Ilanneman, dated August 5, 1980, Resistance Tqmperature __ Detectors. Installation Instructions and Parts List, "Dynathenn Resistance Bulbs with Aluminum Cap-Type Ifead, Model 08-1 Series", The Foxboro Company, January 1964. 20 Amoco Oil Company Letter from T. M. Warne, d:ted September 15, 1980, Radiation Resistance of Amoco Oil Lubricants: Project 4210, with attachment: The Effacts of Radiation or Lubricants in fluclear Generating Stations, by James S. Ferrie, Paul Leinonen. Dr. B. Tieil, and E. Wharton T0iitario ifydro Research Division), for presentation at ASLE 35th Annual l'2eting, Anaheim, California, May 1980. 21 Mobil Oil Letter from J. Kestly to R. K. Hannemen dated October 13, 1980, Radiation Information Wisconsin Electric Pcwer, with attachments for radiation test data. of ifo~biTg~rease 28. 22 Kerite Letter from R. A. Olson to R. K. 'Hanneman, dated October 22, 1980, with attachment: Point Beach f!uclear Plant, Wisconsin Electric Power Company, LOCA_ Qual'iTicaTion 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 9,1930. Rome Cable Corporation letter from D. D. Sand to P. R. balbumeur dated March 19, 1971. 24 Okonite Company Letter from J. S. Lasky to R. K. Hanneman, daed May 9, 1980, With Attachment. Blodgett, R. B. & F sher, R.G., " Insulations and Jackets for Control and Power Cables in Thermal Reactor fluclear Generating Stations _, IEEE Transacti_ons on Power Apparatus and Systems, Volume PAS-88, tio. 5, May 1969. 25 ." Type Test Cable _gualification Proaram and Data for fluclear Plant Designed Life Simulation ThrouSh Simultaneous E go_s_ure", Franklin Institute Research Laboratories, Final Re ort F-C3694, Philadelphia, Pennsylvania, January 1974. 26 Lancaster, Ron, "Qualifi_ cation of Safety-Related E.quipment Used in !!uclear Power Generating Stations Including the Effects of_ Aoing". Reliability Conference for the Electric Power Industry,1980. Carfagno, S. P. & Gibson, R. J., "A Review of Equipment Aging _ Theory and Technolocy", Electric Power Research Institute, Final Report tiP-ISS8, Palo Alto, California, September 1980. ATTACHMENT 1 P00R ORIGINAL

'c, UNITED ST AT ES 8 NUCLEAR REGULATORY COMMISSION {g 7 .. E REGION lli o, p' v [ 799 ROOSEVELT ROAD g j cLEN ELLYN.1LLINois 60137 April 24,1980 MEMORANDUM FOR: V. D. Thomas, Technical Programs, 91 vision of Rea,ctor Operations Inspection, IE:HJ S THRU: d(, [,, D. W. Hayes, Chief, Engineering Support Section 1 FROM: F. J. Jablonski, Reactor Inspector, Engineering Support Section 1

SUBJECT:

SCREENING REVIEW OF LICENSEE RESPONSE TO IEB 79-01B AND SUP. MARY OF INSPECTION OF INSTALLED SYSTEM 0 '.T POINT BEACH UNITS 1 AND 2 - DOCKET NOS. 50-266 AND 50-301 We have completed our initial screening review of the Point Beach facility response to IES 79-01B, and the inspection phase of the system audit. A walkdown was conducted on April 21, 1980 to inspect installed components associated with the SI system. Pri' - to the walkdown P&ID, elementary control, and wiring diagrams were. .wed. Components listed en the attachment were observed. NOTE: G.../ Unit 2 containment was inspected visually. Observations: Motors Pump and valve motors of the SI and CS systems for both Units 1 and 2 were located in the same room of the auxiliary building at i ( elevation {8 feet. Radiation is the limiting environmental f actor. (1.11 x '10 rads by analysis). 8 SI Pump - 4.16 KV, Model 688.5-H, Thermatastic Insulation, 2 x 10 rads, H 6 CS Pump - 480V, Model TBDP, 1 x 10 rads, M Motor Operated Valves Operators were either type SMB-0 or SMB-00 and had Class B motor insulation. Motors were either Reliance or Peerless. One operator switch compart.nent cover in containment was removed. Switch materials were red and black plastic. Previously, on some valves, the licensee had replaced the red / black plastic switches with a gray metamine material. RIII has experienced this same change at Onsite Inspection ATTACHMENT 2 l

V. D. Thomas April 25,1980 other reactor sites. It is believed the change was made to enhance resistance to damage from radiation. RIII wiLL followup with limitorque. Limit Switches, Transducers. Transnitters ALL Located in the auxiliary building. The only potential harsh environment is radiation; wiLL be included in review of 90 day response. Levet Switches Containment sump level switches (indication only) Magnetrol A-153 will be replaced. Rome cable associated with the switches may be replaced if not ultimately proven to be qualified for anticipated radiation Levels. Cablis and Solices None of the 600 volt power and control cables nor splices had any manufacturer's identification markings on the jacket. There was no physical way of verifying that installed components complement installation records or test data. Penetrations No model numbers or other identification was apparent. Radiation tests-have not been performed; analysis is yet to be performed. Miscellaneous ALL components inside containment were above flood level, i.e., greater than 15.1 feet; aLL components were oriented properly. NOTE: Location of critical components may not necessarily be identical in Unit 1. There may be a serious discrepancy with the criteria used in the SER of high energy line failure outside of containment at Point Beach. Th rough-out the SER (enclosure to NRC letter dated May 7,1976, Lear to Burste'.nl the pagameters constituting a high energy piping system were a temperature of 200 F and pressure of 275 psig. Regulatory Guide 1.46 clearly states the 200, regulatory position, i.e., to be low energy a piping system must be F or less and 275 psi or less. Statedpositively,apjpingsystem becomes high energy if the temperature is greater than 200 F or pressure l ATTACHMENT 2 l

V. D. Thomas -3 April 25, 1980 is greater than 275 psig. Please evaluate this apparent discrepancy. 9 9-aLLsL-I F. J. Jablonski Reactor Inspector Engineering Support Section 1 I.

Attachment:

Class IE Elec. Comp. List ~ cc w/ attachment J.G. Keppler G. Fiorelli R. Warnick, Ops J. Smith, Ops 1 i ATTACHMENT 2

+

.u Y: I : :. r.t beach i:.:citar Plant, Units 1 and 2

. CKLT !.0.: 50- 26 6, and 5 0-3 01 MASTER LIST (CLASS IE ELECTRICAL EQUIPMENT RECUIRED TO FUN UNDER POSTULATED ACCIDENT CONDITICNS) I. SYSTEM: Safety Injection COMPONENTS LOCATION PLANT IDENTIFICATION NUMBER INSIDE PRIMARY OUTSIDE PRIMARY GENERIC NAME CONTAINMENT CONTAINMENT l-P14A&B 2-P14A&B Pump Motors l-P15 A &B-X 2-P15A&B Pump Motors X l-SI866A&B Valve Motor 2-ST866AtB onerators l-SIB 51A&B Valve Motor X 2-SIO51A&B Operators X .-SIB 71A&B Valve Motor 2-SIB 71A&B Operators X l-SIB 60A,B,C,&D Valve Motor 2-SIB 60A,B,C,&D Operators l-SI852A&B Valve Motor X 2-SIB 52A&B Operators X l-SIB 78B&D Valve Motor 2-SI87BB&D Opcrators X Namco " Snap-Lock" D2400X Limit Switches X l-SI836A&B Electro-Pneumatic ? c:T 9 w a rn meancancorm l-PT922&923 Pressure X 2-PT922&923 Transmitters l-FT924&925 X 2-FT924&925 Flow Transmitters .1-FT9 2 8 - X 2-FT928 Flow Transmitters .-g_rI;9;; X o '2 -FI T 9 3 0 Flow Transmitters 1-LT931 X 2-LT931 Level Transmitters 1-LC942A&B X 2-LC942A&B Level Switches X e ATTACHMENT 2 1 e s

3, Safety *njection (Contino.>d)

• COMPONENTS

'"""n" FLANT IDENTIFICATION INSIDE PRIMARY CUTSIDE PRIMARY NUMBER GENERIC NAME CONTAINMENT CONT AIN..ENT i ' 600 Volt AC Power Kerite Cables X 5,000 volt AC Okonite Power Cables X 600 Volt AC Power Kerite Cables X X 600 Volt Control Rome Cables X X 600 Volt Control Kerite Cables X x Instrumentation Tite Cables X bw. ton Insulatior Instrumentation wire r. rahle cables X X em_ Crouse Hinds / West - - Penetrarions X X Bechtel/Raychem Cable Splices X '~ O WD en ene m 4 6 e eumsmm e e amme e se e g my 9 1 l e l l t e_ N ATTACHMENT 2 i l

,( o' - - *!; t-Jf $ 6 Q'~)/* % '4UC L E A R h L u a i T C5 y cut.t'.*i,5 t Cf. n M G T *.* l t

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,, s cLEN tLLyu stuhois sotar 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. 54. Hayes, Chief. Engineering Support Section 2

SUBJECT:

IEB 79-01B (A/l F03067180) Attached is a copy of a memorandum dated July 17, 1980 received from Frank Jablonski relative to IEB 79-018. It is being forwarded for your information and solicited guidance. The question of Identification of safety related systems and comonents (paragraph No.1 of the memo) is an old one. I disagree with Frank in that I feel that this identification is a responsibility 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 lists 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 questler. - should we assume the licensee's response I 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 recorxnended.

Coninents and further guidance is requested concerning matters discussed in paragraphs 3 and 4 of Frank's memo. 4 . J-7 ** D. W. Hayes, Chief Engineering Support Section 2 i ? Generic Issues ATTACHf'ENT 3a fo i2310o 8 3

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

Attachment:

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

e

1. EIC

/g&

1. "e UNITED STATES

'n NUCLEAR REGULATORY COMMJSSION 2 - - E REGION lli i 7eo roosevelt ROAD g.*....[ GLEN ELLYN,ILUNols 80137 July 17, 1980 - :b. MEMORANDUM FOR: D. W. Hayes, Chief. Enginee-Ing Support Section 1 FROM: F. J. Jablonski, Reactor inspector

SUBJECT:

FORMULATING TECHNICAL EVALUATION REPORTS (TER) - REVIEW 0F IEB 79-018 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 IIcensees'. 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 responsibility for classi fying safety related systems on the licensee. I Action item No. 1 of 79-01B requested each licensee to provide a " master list" of all ESF systems in their respective plant required to function during a postulated accident. Appendix A to 79-018 lists " typical" equipment / functions needed for raitigation of an accident. A comparison of master ilsts was made of four licensees with similar Westinghouse PWRs-1 (see Attachment 1). Arbitrary selection and non-standard nomencisture \\ of systems makes evaluation of the master lists extremely difficult. NRCl requested each IIcensee to submit the information under oath. Shouldthel information therefore be assumed complete and correct? l it is extremely frustrating to review responses which vary so much in attention to detail, depth of review, etc. As stated previously in the i draft TER for D.C. Cook, because I as a principal reviewer lack detailed systems / operations experience, further guidance is requested. l Another TER related matter is motorized valves equipped with Limitorque j l operators (see Attachment 2). As can be seen, each test report is for a specific 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 l s provided. For example, test report No. 600456 (SMB-0-40, Reliance Motor i with Class RH insulation) may be listed for all operators from series l SMB-000 to SMB-5; motor name plate data not provided. Without the name plate data and the basis for extrapolation, a meaningful evaluation cannot be made. l l l ATTACHMENT 3a W Q(hD k

1 D.W. Hayes 2-July 17, 1980 i it is requested that this memorandum be forwarded to IE:HQS as an addition to A/l F03067180 with the same copy distribution. e

• [

F. J. Jablonski Reactor inspector Attachments: 1. Comparison of Master Lists 2. Motor Operated Valve Tests cc: J. G. Keppler i G. Floreill ATTACHMENT 3a ---,,+ r-- ..,c- .,w.,.. .a, ,,-.,--_,-__.-.,m,,- r,

ATTACHMENT 1 SYSTEMS ,P de fdQK f1W PT. BCH, a Aux. F.W. X X X Chem. & Vol. Cont. X 2 X X Cntmt. Air Hndtg. X X X Cntet. H Cont. X X p Cntmt. Sp. X X 1 Main Stm. X X X Aux. Stm. X Stm. Dump X Rx Clnt. X X X X Res. 4t. X 2 X 3 Saf. Inj.(m. X 2 X X Clg. Water X Esnt't. Serv. Wat. X Comp. Clg. Wat. X 3 Aux. CLnt.g CLg.2 i Emerg. Cur 1 X 1 X Cntmt. Purge X Rx. Bldg. Vent X Inst. & Prot. X Rx. Trip. Act. X Rx. Cont. 8 Prot. X Rad. Monit. X Rx. Hot Samp. X Stn. & Inst. Air X Stm. Gen.8D X Post Acc. Monit. X Rem. Sht. dn. Monit. X Cntmt. Isol. X X Mn. Stm. Isol. X Mn. FW Isol. X i l ATTACHMENT 3a

ATTACHMENT 2 MOTOR OPERATED VALVES MOV'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 SMB-00}0 SMB/SB- } This series may SMB/SB-1 > This series may also also include WB SPE/SB-2 include SB SMB/SB-3 SMB/SB-4 This series may ~ ~ SMB-5 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 Temo

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

BWR Peerless H 7 (B-0009) 1x10 R DC l 3400 l

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

BWR Reliance RH 0 FIRL F-C 2x10 3441

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

PWR Reliance RH g 2x10 i

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

Outside Reliance B Cntmt 7 2x10

f. WCAP7410L 12-70 SMB-00 8

7744 8-71 l denotes foot pounds of torque only SMB-0 has been tested seismically Re: a, b, c ATTACHMENT 3a

1. ps Ettoq'o, e

UNITED STATES l' ~j NUCLEAR REGULATORY COMMISSION ^ E WASHING TON. D. C. 20555 o, 8 ~%.,.# SSINS #6820 JUL 3 1980 . MEMORANDUM FOR:

2. R. Rosztoczy, Branch Chief. Equipment Qualification Branch, Division of Engineering, NRR M

THRU: f . L. Jordan, Assistant Director for Technical Programs, E Division of Reactor Operations Inspection IE FROM: V. D. Thomas, Task Manager, Review Group, IEB 79-01B, Division of Reactor Operations Inspection, IE

SUBJECT:

REQUEST FOR NRC POSITIONS ON REVIEW QUESTIONS OF IEB-79-01B LICENSEE RESPONSES In accordance to our verbal positions on the questions nagreement, we would be happy if you would provide oted in the enclosed memoranda. Since it is essential to establish a uniform approach to the review effort' to obviate the questions being generated in the on-goi.ng review of licensee responses, we will be happy to meet with your staff to discuss these concerns to expedite resolution of the issues. ~ b Vincent D. Thomas, Task Manager 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 i G. Fiorelli, RIII g-D. W. Hayes, RIII A. Finkel, RI R. Hardwick, RII

f. Jablonski, RIII D. Mcdonald, RIV J. Elin, RV dUL 71980 pop :-otM A M MEM 3a

i s'* e*'4 e , c UNITED STATES ! }*., g,( NUCLEAR REGULATORY COMMISSION a"pr/ r REGION ill

g(q[3bbJ O, g

[ 799 roosevelt ROAD %*v y' CLEN ELLYN. ILLINots 50137 June 20, 1980 MEMORANDUM FOR: E. L. Jordan, Assistant Director, Divisios, of Reactor Operations inspection, IE:HQ THRU: -G. Flore111, Chief, Reactor Construction and Engineering Support Branch FROM: D. W. Hayes, Chief, Engineering Suppoi t Section 1

SUBJECT:

IEB 79-018 (A/l 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 questions for which we require guidance to complete our review of responses to IEB 79-018. By this memo I also would like to confirm our understanding that NRR (Environmental qualification Branch) will review for acceptability all test reports and other documentation which IIcensees 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 lEB 79-01B task group, on a volunteer basir, 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 licensee 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 licensees meet their extended submittal dates, we should complete our site inspections, reviews, and technical evaluation l l f ATTACHMENT 3a h 3. TD0$9 D $

June 20,1980 E. L. Jordan 2 reports by the end of December 1980. Further delays in the submittals or any unforeseen events will hamper our ability to meet the new February 1,1981 deadline. a& D. W. Haye, Chief Engineering Support Section 1 Attathments: 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: Ses Attachtent 4 cc w/ attachments 1, 3, & 4 enly: J. G. Keppler G. Florelli V. D. Thomas, IE:HQ l A. Finkel, R1 R. Hardwick, Ril D. Mcdonald, RIV 1 J. Elin, RV R. F. Heishcan s t l ATTACHMENT 3a

i

• \\*

/, UNITED STATES ah,s.,fj NUCLEAR REGULATORY COMMISSION e REGION lli g,i ,8 o 799 ROOSEVELT RO Ao ' d,. j' CLEN ELLYN,ILLINols 60137 June 10, 1980 MEMORANDUM FOR: 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 79-01B 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 approvat 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 length 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 submitted 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.
• UPSC Letter dated 2-2-76 from James (WPSC) to Purple (NRC).

ya 10010-02gf ATTACHMENT 3a

9 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 NRR? 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.) Ref erence 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. i l Ch Z t F. J. Jablonski Reactor Inspector cc: J. G. Keppler G. Fiorelli l l l ATTACHMENT 3a l l

c !\\ /,,' UtJITED STATES NUCLEAR REGULATORY CCMMISSION .? y,, {,4 ' REGION lli i o,. ) 799 ROOSEVELT ROAD g-.....f GLEN ELLYN. ILLINols 60137 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 IE3 79-01B in accordance with IEB 79-OlB, 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 components. Unresolved concerns previously' identified by Rill Inspectors during reviews of IEC 78-08 and IEB 79-01 along with subsequently ident! fled concerns make it difficult for us to formulate meaningful TERs for certain plants. The previous unresolved concerns are documented in the memorandums listed below (1,2,3) and are reiterated in Attachment A to this memo. Subsequently identified concerns are listed in Attachments B, C, and D. To assure uniform evaluation, gt.idance is needed for these items. Please forward these concerns to IE:HQ. 1. Ti 2515/13 - Qualification of Safety Related Electrical Equipment Flore111 to Sniezek, 10/13/78 2. Same title as 1., Flore111 to Klinger, 12/78 3 Review Status of Responses to IEB 79-01, Hayes to Jordan, 9/5/79 -('~ ( O w b ( [ F. J. Jablonski Reactor inspector

Enclosures:

As Stated cc: J. G. Keppler G. Flore111 V. D. Thomas, IE:HQ A. Finkel, Rt R. Hardwick, RIi D. Mcdonald, RIV J. Elin, RV D 9 ATTACHMENT 3a f $D ff

Pn : lit. iiI.U '-[.. _ ' 5 [" 'e, UNITED STAT Es / nin l ,5.* C.*_! _ _.' {. h ( 'n NUCLEAR REGULATORY COMMIS.SION ofo P.a t:1 W ASHINGTON. D. C. 20555 A/o l{PA0 ,f \\ / FFr.liS il5LU SEP 11 1980 - / nCCES P.0t,HS ];(JgjlFILE,A ( ~ Dacket No. 50-266 -301 MEMORANDUM FOR: D. W. Hayes, Chief. Engineering Support Section 1, Region III FROM: E. L. Jordan, Assistant Director foe Technical Programs, Division of Reactor Operations Inspection, IE '

SUBJECT:

LACK OF SEPARATION CRITERIA AT POINT BEACH (AITS F0305968 Your request for a review of the Point Beach discussion of GDC-4 was located

• on pages 4.1-4 thru 4.1-5 of the FSAR and, mislabeled as GDC-40.

' is enclosed. ~ A copy 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. war . Jordan, Assistant Director fo echnical Programs 8 i 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 Site Specific Issues ATTACIU4ENT 3b pp #6 d

Missile Protection Criterion: Adequate protection for those engineered safety features, the-failures of which could cause an undue risk to the health and T. safety of the public, shall be provided against dynamic ef fects J and missiles that might result from plant equipment failures. ~ (GDC 40) b ~ The dynanic ef fects duning bIowdown 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 mecha'nical 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 supportsf guides and restraints also prevent rupture s s N of the primary side of a steam generator as a result of forces created by a steam or feedwater line rupture. s The mechanical consequences of a pipe rupture are' restricted by design such that the functional capability of the engineered safety features is not impaired. e I l N1"ACIDEIT 3b i I h t

/' DISTRIBUTION: Central Files-SEP 3 1930 ORAB RDG s TuTc E,acht FOR: %acar unm ~- MDt0RAliD Edward L. Jordan, Assistant Director for Technical Programs, Divisicn of Reactor Operation Inspection FROM: Darrell G. Eisenhut, Director, Division of Licensing l 4

SUBJECT:

LACK OF SEPARATION CRITERIA AT POINT' BEACH I

REFERENCE:

Hmorandu::, E.L. Jordan to D.G. Eisenhut, f g dated August 1,1980. i In response to your request in the above referenced xxsnorandum, 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 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. li e We have detemined froci our rtview of the pipe break criteria that the .I correct interpretation of a high energy line is a sysusn where either the fluid temperature is greater than 200*F or the fluid pressure is I greater than 275 psig. However, Appendix A of APCSB 3-1 defines a high energy fluid system as a fluid systen that during normal plant conditions il meets the temperature ort. pressure limits. Therefore, those portions -}y i of systems such as ECCS system.; that do not exceed the temperature or pressure limits during normal operation ld not be classified as high energy. This should resolve the Regio neerns for the safety _. 3 injection lines in the pu:np room. 88L*J T.l s.N ~ As part of the NRC's Systematic Evaluation Piogram, the pipe break - b criteria for the SEP plants was reviesed. The results of this review revealed. inconsistent application of the pipe break criteria for both iniide and outside containmant applications (see enclosed memo). As . ij. a result of this study a generic letter to the SEP licensees has been y', prepared to address the application of pipe break criteria inside the 9 contain:aent. A generic letter to all other Itcensees addressing pipe break criteria inside the containment is currently, planned. Resolution ?, of the pipe break criteria outside the containment is pending the results l p, of the inside the contairnent reviews. Original signed hY G [ (' parre110. EisenhuI t Ihrrq11 G. Eisenhut, Director fC Diviston of Licensing e

• fff P00R ORIGINAL J. Fair, X27357 orrec. >

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,/ 'e' UNilI D s1 AT LS M NUCLE AR REGULAlORY COraMISSION ~ 5 - h h' [I HE GION lli 799 RoostVELT noAo ? GLEN ELLY N, ILLINots LO137 a...+ 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. V. Hayes, Chief Engineering Support Section 1

SUBJECT:

LACK OF SEPARATION CRITERI A AT POINT BEACH (A/l F03059680) Ref: April 24, 1980 memorandum to D. V. Hayes f rom F. J. Jablonski, same subject (copy. attached) Per our discussion, please advise us of the NRC position relative to 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, "Environn. ntal and Missile Design Basis", and GDC-5, " Sharing of Structures, systems and Components" were discussed. However, in regard to the containment spray and safety injection pumps Figure 1.2-5 shows that these pumps for both Units 1 and 2 liave a corxnon location wi thout separation. In connection with the safety injection (SI) pumps, we would also like a clari fication of what constitutes a high energy line. Our Interpretation from 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 discharge lines which operate at about 1500 psig and less than 200 F considered high energy lines? No further review or action on our part is planned pending receipt of your response. D. W. Hayes gggf210b Chief Engineering Support Section I cc: J.G. Keppler G. Fiorelli R.F. Heishman ATTACH!ENT 3b R.F. Varnick F.J. Jablonski P00R ORIGINAL

4 s. .s / -3 UNilLD SI A105 NUCLEAR REGULATORY COMMIF.SION / ' k/ j' $ 2 REGION lli S, F; 8 799 ROOSEVELT RoAO GLEN ELLYN,lL LINoiS 60137 Hay 15, 1980 4 MEMORANDUM FOR: D. W. Hayes, Chief, Engineering Support Section 1 FROM: F. J. Jablonski, Reactor Inspector

SUBJECT:

POTENTI AL DISCREPANCY WITH CRITERI A USED IN THE SER OF HIGH ENERGY LINE FAILURE AT POINT BEACH RE: HEMO 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 at the Point Beach Nuclear Power Plant.

(HI energy - whege the temperature and pressure conditions of the fluid exceed 200 F and 275 psig). T.Q O. .,. ; - G v G, s K, F. J. Jablonski Reactor inspector Engineering Support Section I cc: J.G. Keppler / G. Fiorelli J. Smlth V.D. Thomas, IE:HQ C.J. DeBevec, IE:HQ l ATTACHMENT 3b O flat 08003 y -e ,v-e 4

e

• ),

Ufelli O !.T Al E5 I,s. f- % NUCLEAR REGULATORY COMMISSION y'--* ' I ~ y,gd[y,g,r. f r REGION lli n 799 ROOSEVELT ROAD vf CLEN ELLYN,ILUNots 60137 April 24, 1980 o \\ ayes, Chief,EnginaeringSupportSection1 \\ MEMORANDUM FOR: W s H 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 observed 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 sade condition exists at a different elevation for the Spray Additive Tanks. Another example is the use of a single penetration for the passage of redundant cables used for indication of containment sump level. Separation is beyond the scope of IEB 79-01; therefore there is a need ( for a separate memo. I realize separation criteria may have been different 15 years ago; however, these observations should be documented. [ q,_ h6

_nc_;/; n F. J. Jablonski Reactor Inspector i

Engineering Support Section 1 cc: J. G. Keppler E. L. Jordon, IE:HQ l ATTACHMEIT 3b D}}