ML20004D852

Environ Qualification of Safety-Related Electrical Equipment, Technical Evaluation Rept

ML20004D852
Person / Time
Site:	Duane Arnold
Issue date:	11/19/1980
From:	Julie Hughes NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML112231562	List: ML20004D852
References
IEB-79-01B, IEB-79-1B, NUDOCS 8106100220
Download: ML20004D852 (44)

Text

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~ ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT JO IEB 79-01B' TECH 2 CAL EVALUATION REPORT DOCKET No. 50-331 DATED: Novemiier 19,'1980 j '

-Licensee:

Iowa Electric Light & Power Company.

Type Reactor: BWR j

Plant: Duane Arnold i

Prepared by J. Hughes Engineering Support Section Reactor Construction and Engineering j.

Support Branch, RIII

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[agg Introduction 1

Background and Discussion 1

. Summary of Licensee ~ Actions / Statements 1

. System Comparison 2

Equipment. Evaluation 2-4-8

~

Caveat.

2 Conclusion 2

Attachments:

1.

Referenced Test Reports 2.

Onsite Inspection Report 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

-v-

Introduction 1/ or use as input

-This report is submitted in accordance with TI 2515/41 f

to the Safety Evaluation Report on qualification of Class IE electrical equipment installed in potentially " harsh environmental areas at this facility.

P 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 equipment would function under (i.e. during and following) postulated accident conditions.

The Technical Evaluation Report.(TER) is based on IE's review of the li-censee's submittal for conformance with the D0R guidelines or NUREG-0588, a sitt inspection of selected system components, to ve submittal, and EQB's review of component test reports.gpfy accuracy of the t

Licensee submittals were received on March 14, 1980, May 5, 1980, and October 31, 1980.

The site inspection was completed on April 14,1980.b/ Ge specificguidancewasrequestedfromIE/NRRheadquarters.yericandsite i

Summary of Licensee Actions / Statements Investigations by the licensee indicate that almost all components are either not subjected to harsh environmental service conditions or have qualification documentation. Components determined to have incomplete qualification documentation at the present time will be tested, shielded, relocated, or replaced as soon as possible.

In most instances, these actions will be taken during the 1981 refueling outage, but in no event later than June 30, 1982.

In each case, justification for continued operation of DAEC has been provided.

In the equipment qualification charts, the specified radiation dose is the calculated bounding dose from gamma radiation as per the licensing basis of the Duane Arnold Plant.

(Not the guidelines)

L 1/

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/-.

4/- Attachment 2.

_5 /

Attachements 3a and 3b.

i a

a h

System Comparison A. comparison was made between the system 9jlist provided by the licenseek!

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

Engineered Safeguards Actuation Low Pressure Coolant Injection Containment Spray Radiation Sampling Combustible Gas Control Closed Cooling Water System Reactor Water Cleanup System Reactor Recirculation System Equipment Evaluation Class 1Eeggppmentwasevaluated,thatis,placedintofiveseparate Cdtegories.-

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 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 site inspection, and licensee's proposed actions, it cannot be concluded that there is reasonable assurance all components installed at the Duane Arnold Energy Center are environmentally qualified and installation methods of environmentally qualified components would not contrib'ae to the failure of such components during a potential accident.

Based on several components categorized as IVb, the information submitted by the licensee did not fully and completely respond to the Order for Modification of License DPR-49. However, the licensee did provide justifica-tion for continu.ed operation.

6/.

-7/

Attachment S.

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~A' positive conclusion cannot-be made until:

~ 1. - 'All matters referred'to'IEHQS/NRR'ha've been satisfied.EI 2.

The S: 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,-5, 6,'and 8.)-

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LIST OF TEST REPORTS 1.

GE Qualification Report F01 for Electrical Penetration Assem";1y dated April 30, 1971.

2.

Letter by Mr. G. G. Sherwood of GI to Mr. D. G. Eisenhut of NRC dated December 2, 1977 3.

ITT 3arten Report No. R3-288A-1 dated May,1980 and letter of Mr. L. L. Blake, Jr. of ITT 3arton to Mr. J. C. Hink of 3echtel dated June 10, 1980.

4.

Limitorque Qualification Test Repcrt No. 30003, dated May 28, 1976.

5 Linitorque Qualification Test Report No. 600376A dated May 13, 1976.

6.

Franklin Institute Research Laboratories Final Report No. F-C3h41 dated September, 1972.

7 NAMCO letter to Bechtel dated September 8,1980 and vendor print 788L-AFED-E57-1.

8.

Barksdale Bulletin 730701-E dated 1979 9

Letter No. 3LIEG-80-378 of Mr. J. L. Hurley of Bechtel Associates Professicnal Corp. to Mr. Philip D. 'Jard of Icwa Electric Light and ?cwer Co. dated August h, 1980.

10.

ASCO Test Report No. AQS 21678/TR Rev. A.

11.

Atkematic Valve Co. Inc. letter frcm Gary Spear to Jim Hurley dated May 28, 1980, and Report No. 21 by, " Radiation Effect Information Center" Of 3attelle Me=orial Institute.

12.

Plant Equipment Design Engineering Memo No.126-62, test No. h.

13.

Letter by Mr. D. K. Vater of Target Rock Corp. to Mr. Ren Garris of Icwa Electric dated August 29, 1980, Target Rock Corp. Test Report No. 2375 dated September 26, 1979, Appendix C, Appendix I, and Target Rcck Test Report No. 2302 dated May 9, 1979 14.

Letter frem Me. Dan Whalen o.' Rosemount Inc. to Mr. J. L. Hurley of 3echtel dated July 11, 1981 15 Franklin Institute Research Laboratories Technical Report No. F-C2737 dated April 30, 1970.

16. Kerite Report No. EM-173A and 3 dated May 23, 1977 17 Ok: nite Ccepany Engineering Report 30. 127, Revisica 1, dated Nove=ber 5, 1971.

18.

Franklin Institute Research Lateratcrie: Technical Report No. F-CLO33-1 dated January,19't 5 P'00R0Rl l

List of Test Repcrts A'"TACHYI'IT 1

t'd.s

.h:?,Y

19. 3r[ad, din Institute Research Laboratories Technical Report 63fr F-Ch033-1 dated January, 1975

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20.

Raychem Report en Aging Study ~4CSF Cc= pound Report Ifo. ERR 2001 dated August 10, 1978.

21.

Raychea Draft " Interim Report of Flantrol Ther=al Aging Study" I

laboratory Report !!o. 5058 dated February, 1976.

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P.EM'RANDL:s FOR:

V. D. Th: as, Techni cal Progra-s, Livision of M

Ra cr C:ersticns Int e:.t ion, I E : H'l Q

Tr!EU:

4.

. Hayes, Chief, Engireering Su: port Secticn 1 g

F F.CM :

J. Hus'.es, Reactor :ntpector, Engineering Su; port Section 1 Q

L S '. 3 J E C T :

SCREENING REVIEW OF LICENSEE RES-NSES TO IES-79-01B AND SU".".ARY OF INSFECTION OF If. STALLED SYSTEM AT DUANE ARNOLD - FACILITY DOCKET No. 50-331

• te have ::rpleted our initial screening review of the Duane Arnold f s:ility rete:..se to !E3-77-O'3, and have cc pleted the in:pecticn phase cf the system audit.

I condu:ted a walkd:-n en March 11-12, 1930 to ins ect instatted cer:c.ents 7

asse:isted with the Cere S ray and RHR syste.s.

Fricr to the wat.id: n, the fell:-ing cc penents were selected for review:

Tag N..'er Cen;cnent Dra ing 252142 Linit ! witch fi-121 Rev 9 VZ'42

".a ual Valve L.O.

252'4 3 Linit S itch

2* 43 t'anual Valve L.O.

C '.' 2 ' 18 Testable Check Valve

.e.....,

c-o. 1a n

52I'E3 Li.-i : !. itch CV2'33 "estable Check Valve S '. Z ' I S S:lencid

SZ'III.E3 Limit S itch I'". ' ? 3 3 Lini:creue C: era:ce Valve t'

20 Rev ~3

!"'9:3 Li-itercue C; era:cr Valve l'. 9 Rev ~5

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V2 02 Te :able check Va'.ve

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a V.D. Theras April 14, 1980 Results of the inspection established that the instattatien of the selected cor inents was in accordance with specifications and drawings.

Nameplate r' a was censistent with the records and included serial number, tag number, motor type insulaticn class and ratings.

Electrical cables will be reviewed later under generic cceponents.

The-inspector questiened the Licensee rete ive to solenoid and limit switches for testable valves. The licensea stated that these valves were enty operated (tested) during a refueling outage, and that during c eratien they were in the safe position; therefore, the solenoids and switches need not be qualified.

Limit switches for normally locked open valves (manual type) also fall in this same classification.

The inspector agreed with the licensee's position.

The inspector requested the licensee to determine if the field run junction boxes located inside the centainment were pull boxes or connection boxes.

The licensee stated that if the boxes contained solices/ terminal blocks, they wculd establish that they were preperly envirennentally qualified.

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Reac*g r Inspector Enqjneering Support Section 1 l

cc:

J.G. Keppler, RIII G. Ficrelli, RIII G.C. Wright, RIII W.S. Little, RIII RIII Files N!"!AC:9EC 2 i

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'g s......f cLes tuvs iushoes wur 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-OlB (A/l F03067180)

Attached is a copy of a memorandum dated July 17, 1980 received from Frank Jablonsk! relative to lEB 79-018.

It is being forwarded for your Information and solicited guidance.

The question of identificatlor: of safety related systems and components (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 Ilcensee, not the NRC. He must know his plant.

I do agree, however, 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.;his is assumed to be an NRR function.

In regard to Frank's question - should we assume the licensee's response to IEB 73-OlB to be complete and correct - I have told him yes.

Fu r the r,

tnat if he Identifies ignificant incompleteness in the response, or incorrect Information during his reviews, to bring these to my attention so appropriate action can be recommended.

Connents and further guidance is requested concerning matters discussed in paragrap!'s 3 and 4 of Frank's memo.

/

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

t E. L. Jordan 2

July 23, 1980

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Attachment:

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F..J. Jablonski Memo to l

D.W. Hayes dtd 7/17/80 J

cc w/ attachment:

J. G. Keppler, Rlll V. D. Thomas. IE:HQ A. Finkel, RI R. Hardwick, Ril D. Mcdonald, Alv J. Elin, RV r

R. F. Helshman, Ri t t

-> F. J. Jablonski, Rlli l-5 1

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i ATTACHMENT 3a 4

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uy'o UNITED STATES E '

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NUCLEAR REGULATORY COMMJSSION

• -*//

t REGION lil o,

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J 7M roosevelt ROAD

,g oLEN ELLYN,lLLINois 60137 July 17, 1980

---S. MEMORANDUM FOR:

D..W. Hayes, Chief Enginee-ing Support Section 1 FRON:

F. J. Jablonski, Reactor inspector

SUBJECT:

FORMULATING TECHNICAL EVALUATION REPORTS (TER) -

REVIEW OF lEB 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; discrepancles 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 Inconsis* tnt.

In fact, the NRC places responsibility for classifying safe',y related systems on the 1icensee.

Action item No. 1 of 79-018 requested each IIcensee to provide a " master' t

list" of all ESF systems in their respective plant required to 5

during a postulated accident. Appendix A to 79-018 lists " typical" equipment / functions needed for mitigation of an accident.

A comparison of master lists was made of four licensees with similar Westinghouse PWRs,-

(see Attachment 1).

Arbitrary selection and non-standard nomenclature

\\

of systems makes evaluation of the master lists extremely difficult.

NRC L requested each licensee to submit the information under oath.

Shouldthek information therefore be assumed complete and correct?

It is extremely frustrating to review responses which vary so much in attention to detall, depth of review, etc. As stated previously in the draf t TIR for D.C. Cook, because I as a principal reviewer lack detailed systems / operations experience, further guidance is requested.

i Another TER related matter is motorized valves equipped with Limitorque operators (see Attachment 2).

As can be seen, each test report is for a specific unit typt 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 nace plate data e

provi dec.

For example, tes t report No. 600456 (SMB-0-40, Reliance Motor with Class RH insulation) may be listed for all operators from series SMB-000 to SMB-5; motor nare plate data not provided. Without the name plate data and the basis for extrapolation, a neaningful evaluation cannot be mace.

ATTACHMENT 3a

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.@*6.W. Hayes "

July 17, 1980

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It !s requested that this memorandum be fomarded to IE:HQS as an I

addition to A/l F030,67180 with the same copy distribution.

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./u F. J. Jablonski Reactor inspector i

4*tachments:

1.

Comparison of Master Lists f

2.

Notor Operuted Valve Tests I

cc:

J. G. Keppler E

G. Florelli l

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ATTACHMENT 3a I

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ATTACHMENT 1 SYSTPS P.I.

M E

PT. OCH.

l Aux. F.W.

X X

X Chem. & Vol. Cont.

X 2

X X

Cntet. Air Hndtg.

X X

X Cnte.t. H Cont.

X X

2 Cntmt. Sp.

X X

1 Main Stm.

X X

X Aux. Stm.

X Stm. Dumo X

Rx CLnt.

X X

X X

Res. Ht. Sm.

X 2

X 3

Saf. Inj.'

X 2

X X

CLg. Water X

Esnt't. Serv. Wat.

X Comp. Clg. Wat.

X 3

Emerg. Corg CLg.2 1

X 1

Aux. CLnt.

X Cntet. Purge X

Rx. Bldg. Vent X

Inst. & Prot.

X Rx. Trip. Act.

X Rx. Cont. & Prot.

X i

Rad. Monit.

X i

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.

y Mn. FW Isol.

X l

1 i

ATTACHMENT 3a I'

ATTACHMENT 2 MOTOR OPERATED VALVES MOV's 1.

There are basic $lly two type series of Limitorcue operators:

SMa and SB.

The operators are sized from 000 (smallest) to 5 (largest) as follows:

SMB-000]

SMB-00 SPS/SS-0

} This series may SPS/SS-1 This series may also also 4.Lude WB SMB/SB-2 )

include SB spa /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 Insulatien

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

PWR Reliance Special Hi No Radiation Temp

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

BWR Peerless H

7 (B-0009) 1x10 R DC 0

340

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

BWR Reliance RH 0

FIRL F-C 2x10 3441

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

PWR Re t'. an ce RH 2x10

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

Outside Reliance B

Cntet 7 2x10

f. WCAP741CL 12-70 SMB-00 B

7744 8-71

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

@ e%q'o 4

UNITED STATES -

E'

[,j NUCLEAR REGULATORY COMMISSION e

J.

• E o ?L W ASHINGTON, D. C. 20555 k.,

SSUiS #6820 JUL 3 1930

. MEMORANDUM FOR:

Z. R. Rosztocry, Branch Chief, Equipment Qualificawn 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 G oup, IEB 79-018 Division of Reactor Operations Inspection, IE St!BJECT:

REQUEST FOR NRC POSITIONS ON REVIEW QUESTIONS OF IEB-79-018 LICENSEE RESPONSES 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 unifom 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.

Y W/

Vincent D. Thomas', Task Manager Review Group, IEB 79-OlB

Enclosures:

1.

Memo D. W. Hayes to G. Fiorelli, RIII dated June 20, 1980.

2.

Memo F. Jablonski to D. H.yes, 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,'RIII '

D. W. Hayes, RIII A. Finkel, RI R. Hardwick, RII

f. Jablonski, RIII D. Mcdonald, RIV J. Elin, RV JUL 71983 l

ATTACHMENT 3a

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d UNITED ST ATES E 'yg [e NUCLEAR REGULATORY COMMISSION

g ".,,,//

a REGION lil Q,46Q

[

799 roosevelt ROAD

%.*.v j'

GLEN ELLYN. ILLINols 60137 June 20, 1980 PEMORANDUM FOR:

E. L. Jordan, Assistant Director, Division of Reactor Operations inspection, IE:HQ h-T4. Florelli, Chief, Reactor Cons truction and THRU:

0 Engineering Support Branch FROM:

D. V. Hayes, Chlef, Engineering Support Section 1

SUBJECT:

lEB 79-01B (A/l F03067180)

Attached are two memorandums from one of my inspectors, Frank Jablonski.

The fi rst is dated June 10, 1980 and the second June 16, 1980. Both memos raise casic questions for which we require guidance to complete our review af 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 estabi t shing environmental quali fication of Instrument / electrical equipment.

In connection with this, we are sending under c,eparate 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).

The status m' schedule for site Inspections and review / evaluation of the final reports is also attached. Please note that every IIcensee I:as 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 IIcensecs meet their extended submittal dates, we should cceplete our site inspections, reviews, and technical evaluation ATTACHMENT 3a

t E. L. Jordan 2

June 20, 1980 t

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 deadilne.

l 7/

m~

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 Attachment 4 cc w/attachcents 1, 3, & 4 enly:

J. G. Keppler G. Fioreill V. D. Thomas, IE:HQ A. Finkel, R1 i

R. Hardwick, Ril D. Mcdonald, RIV J. Elin, RV R. F. Meish=an s

ATTACHMENT 3a i

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'!.g',.,.,g,g NUCLE AR REGULATORY COMMISSION UNITED ST ATES

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,.t REGION 111 o h}

j3,f f 799 ROOSEVELT RoAo nap f GLEN ELLYN lLLINOls 60137 g

= v54 1

June 10, 1980 MEMORANDUM FOR:

D. W. Hayes, Chief, Engineering Support Section 1 FROM:

F. J. Jablonski, Reactor Inspector SUBJ ECT:

EFFECT OF PREVIOUS NRR REVIEW ON MATTERS RELATING TO IEB 79-01B In almost every licensee response to IES79-018 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 envirennental quali fication 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-cacions for applicable equipment.

(See Section 7.5 of the Kewaunee FS AR.)"

3.

Two licensees, American Electric Pcwer and Wiscensin 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 fecm Tillinghast (AEP) to Kniel (NRC); FSAR question 40.10 Appendix 0.

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

ATTACHMENT 3a

June 10, 1980 i

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 ce assumed that No. 2 above is likewise adequate?

(If th? answer is no, then ncne 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.

C&

J t

F. J. Jablonski Recctor Inspector cc:

J. G. Keppler G. Fiorelli i

1 ATTACHMENT 3a

o UNITED ST ATES

}/1 NUCLEAR REGULATORY COMt.ilSSION

'{ jk, & f f!

REGION lli o, p @ 4 j' 799 ROOSEVELT RoAo g-cLEN ELLYN. ILLINolS 60137 June 16,1980

~+ MEMORANDUM FOR:

D. W. Hayes, Chief Engineering Support Section I FROM:

F. J. Jablonski, Reactor inspector

SUBJECT:

FORMULATING TECHNICAL EVALUATION REPORTS (TER) -

REVIEW OF l EB 79-01B in accordance with IES79-01B, 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 observation of specific system components.

Unresolved concerns previously

• identified by Rlli inspectors during reviews of IEC 78-08 and IEB 79-01 along with subsequently identified 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 rei terated in Attachment A to this memo.

Subsequently identified concerns are IIsted in Attachments B, C, and D.

y To assure uniform evaluation, guidance is needed for these items.

Please forward these concerns to IE:HQ.

1.

Tl 2515/13 - Qualification of Safety Related Electrical Equipment Florelli to Sniezek, 10/13/78 2.

Same title as I., Florelli to Klinger, 12/78 3

Review Status of Re;ponses to IEB 79-01,- Hayes to Jordan. 9/5/79

[ *f G$$wY' F. J. Jablonski Reactor inspector

Enclosures:

As Stated cc:

J. G. Keppler G. Floreill V. D. Thomas, IE:HQ A. Finkel, RI R. Hardwick, Ri l D. Mcdonald, RIV J. Elin, RV ATTACHMENT 3a

ATTACHMENT A 1.

Foxboro Models EllGM and 611/613 transmitters with MCA modificat are believed by Rif t to be under a generic review by NRR.

It is Rill's further belief that the "MCA" modification does not make the transmitters suitable for use in a radiation environment.

Region til's understanding correct?

Is 2.

Several licensees have declined replacement of limit switches which provide position Indication of valves used for primary containment i

isolation.

Are these switches required to be qualified?

3.

GE cable type 51-57275 is used on penetrations manufactured by GE.

Penetrations with this type cable are Installed.at Monticello, -

Dresden 1 and 2, quad C,ities 1 and 2, and Duane Arnold.

The cables withstood LOCA tests performed by Wyle Laboratories, Report No.

44114-2; however, the cable did not pass the IPCEA S-19-81 vertical flame test.

Further, failed at radiation levels in excess of 5x10in the same Wyle test,6GE rads. We recognize thit in regard to GE cable type SI-57275 flame tests are not part of the environmental qualifications addressed in IES 79-OlB, but it makes no sense to find these penetrations acceptable per IES79-018 knowing that they may not meet other requirements.

Concerning GE type S1-58136 sable, this item should be evaluated on a generic basis since many of the early GE plants use this cable.

l 4.

One licensee, American Electric _ Power, lists a letter No. NS-TMA-1950, W to NRR, as technical reference for qualification of ITT Barton di f ferential pressure transmitters.

disposition and status of the letter. Please supply us with the l

ATTACHMENT 3a

ATTACHMENT B The following questions are based on our review of some licensee submittals to IES 79-018:

1.

Licensees maintain that aging is not a required consideration for components that are included in a routine periodic inspection and calibration program.

Is this acceptable?

2.

Licensees maintain that aging is a generic industry issue whose resolution is not clear; therefore, evaluation has not been made or will continue to be made as relevant Inforration is made available.

3 Licensees are referencing manufacturers' letters as establishing the qualification of ancillary parts such as lubricants, tapes, etc.

Is this acceptable or are manufacturers' test reports required?

4.

Limit switches used for valve position indication only have been deleted from the submittal. Licensees maintain that a valve outside containment in series with one inside can have its position verified visually following an accident.

Is this 3

acceptable?

5 The licensees maintain that neither valve position limit switches, solenoid valves, nor control cables for air operated containment isolation valves need be replaced or protected from the adverse envi ronmen t, including flood, because all postulated failures will result in the isolation valve assuming its fail-safe position.

is this acceptable.

t 8

6.

Some fan cooler motors do not me rads.

Qualification test was to 1.4x10gt FSAR requirement of 1.5x10 rads.

Licensee states radiation ATTACHMENT 3a

_,_-.____---L-_

it

.4 ATTACHMFNT B

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level'Is "close enough" to expected accident radiation level to be acceptable.

Is this acceptable to the NRC7 4

7.

Attichment D is a summary of problems Incurred dering a one year operation test of a containment fan cooler unit. Would you consider the test to be a success?

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ATTACHMENT C 1.

In lieu of a test report, what constitutes an acceptable Certificate of Compliance?

2.

What if the test specimen and installed component di f fer, e.g.,

model, type, etc?

3 What, as a minimum, must be included for an analysis to be accept-able?

4.

The guidance provided in Enclosure 4 of 79-01B allows analysis (evaluation) for survice conditions such as radiation and chemical

aorays, is analysis (evaluation) and " engineering judgment" the same thing?

5 Since effects of radiation and chemical spray are " allowed" to be analyzed (evaluated) for important components se:h as containcent electrical penetrations, is it prudent to require a licensee to prepare a full blown analysis to qualify a 7/C 12 AVG cable when a similar 5/C 14 AWG cable was actually tested and shown to be qualified?

6.

Provide us with + limits for evaluation of test data such as pressure, temperature, radiation, duration, chemical spray, and aging.

7 Most tests include only single components and the reports do not include any acceptance criteria. Test conclusions are that usually, no matter what happens during the test, the component is accepted.

This is commonly referred to as a " dead bug" test.

Provide us w!th

. minimum acceptance criteria requirements for a test and its report to be acceptable.

ATTACHf1ENT 3a

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i D.s fs11:vir3 is a tri:f :.;;:ry of the pr.bt::s in:urred 4:21r 1 C.3 : *.3 ;':Tr c:t... ".J :l J rs t ien ;'.t s e o f 1:::

y 211.!:sti.a t:st.

T1 CI L'::TT'I C7 r.72 t--T r

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12-13-75 75.3 1-3/4 1::urs

!!aints.-ance pt:blos caused L

less of :eer.

12-17-75 103.2 2-1/4 t:urs Tr:nsfor sr coil turned out.

12-25-75 373.3 2-1/4 hours Loss of plant p:ver.

1-13 76 027.5 15 in.

E1:ctrical stors c:used loss of pl:nt power.

2 '.7-76 1043 9 hiurs Sprsy rin;s plu::ed. Ri ::4 bypsss.*ar eco11n3 vster.

.shutisvn requir:d to drill holes in test ch::ber.

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2-21-76 1741.5 15 min.

Pltnt : sinter.aace req. sired 1

cut-off of pcver.

I 3-3-76 2211.5 40 min.

Electrical stori c:used two j

short shutdevns bec:use of 3

t:-,4rsry pcwor int erruption.

3-13-76 2352.7 4 tcurs Triped en overlor.J. Faulty soler.oid caused cendensate i

to back up in ch::ber.

Jl 3-10-76 2:54 2 hcurs Unkn:vn.

4-12-75 2047.5 10 days 3:srin: probics; see A;pendix C.

4-22-75 2 - *A 5-1/2 hours 1:nia:rn.

i 4-03-76

,*. 3 9 h urs L'ainom - Installed rxording j

e,ut- ::st e. 7 :nd volt::s to j

d:tcet r.uis..ca tri;s.

4-!3-75 2022.6 2-1/2 d:ys

!.bism.ce trip h e to overicad i

hestcr failure; 1:r th of sNt-l

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1::s Fr:./ :r1.::s st re. t il.!s u :11

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ATTACHMENT 3a

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.r.r Af.as 5-5-76 3140 i

No Trip Probier. with du p valves tut corrected withou'. shutdown.

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'-15 76 4719.5 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Fower (citure due to cic:trical stor=.

7 30-76 5109.8 i

e l

2-1/2 days Terminal teard ru;>tured i

causin: loss of pressure in cLt.ter. h.is teard was a ser.1 regired for i

testin: and w s not a part of equircent being qualified.

i 8-14-76 5369.2 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Power failure.

t-19-76 54 C 7.3 No Trip S11 ht t

c:u:ir. protic: in the controls

(

r. slight cyclin; of tc,tr:ture. Irotic correctcJ wit!.out s1.utdoen.

8-1G-76 5502.6 6-1/2 hours Tculty solenoid resultin:

)

in ec..dcr.:ste b::kir. up in ch: /. tr tr.d c u:ir., totor to trip c. ovcric:.d.

S-2C-76 L? 1.4 2-1/2 hours Solenoid did not function prot:rly end override circuit did r.ot op: rate.

ca overler.d.

Motor trir;:J C-30-76 5725 Ka Trip Tc.nper:ture w:s 4:nta slightly.

Eslenoid 1.:d fciled to cperate but overr c!c'.:r. ido circuit w:s vent:n,;

I roti c.:

with:ut shutds.n. w.s ccrrc:tcJ C-31-76 5752

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i 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Mater tripped on overlocd.

Flott vc1ve stu:k etusing cor.fcasetc to tuild up in c!='. c r.

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ATTACHMENT 3a

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• 4e a h s. 4 y g= 3 N b bu MEW P Kil AC'

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!*'_* *e, "mLCat0 oe J.T..e:rler s A? g _ 8 ra l t. let?

TIME ON L.E.431 07.

DaTE Kxr: grytn ggy 7 j g 9-4-16 5954.*

l-1/2 days treakdown in electrical tape K')

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and other ansulation -spplied by JOY at the ends of the Ict.d o

caused a short which tr.irned off one lacd resulting in a single j

hased condition. Lead end u

• (red.

/y 11-7-76 7354.3 2 days Lead seperation at terrinal board caused by bret.kdom of insulation, lead c brittle:.ent and vibretion. Le:l was repairc {

11-30-76 7840.5 20 hcurs nsulttien f:ilure sicilar to thet of 9-E 76 caused unit to trip.

12-24-76 6390.9

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1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Impmperly asscb!d *6ler. olds c:.used condensc.c to beck up in chc;:ber resulting in an e -"

• 4% ri

• t P 12-30-76 E530 No Trt},

Lost Fhase f1; continued to opertte in a single phase condition, ji 1-11 77 E 814'. 2 27 he,urs Power surre caused unit to tri;*. I l

E4.ccus e of sir.-le p'.c se conditio:.!

unit could rst be restcrted.

1 Repircd the lect et the ter:. ira.1 bocrd cr.d resu cd tt :t i r.:.

14.i s 1 n h L. s si ilcr to thct of 11-P-76.

3-9-77 10145.9 xI A-trent short in r.stor caused 7 '-

t e.m.ir. t ion o f t e s t.

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l ATTACHMENT 3a

!ione t

t l

f Site Specific Issues ATTACIBEiT 3b

q i:

1.

' Control' Rod Drive System 5

.2.

Pri=ary Containment Isolation and Nuclear Steam Supply Shutoff System 3.

!!ain Steam Line Isolation Valve leakage Centrol h.-

High-Pressure Coolant Injection System

-ls 5

Automatic Depressuri:stien System 4

6.

Core Spray System 1

7 Residual Heat Removal System 8.

Standby' Gas Treatment System 9

Standby AC Power Supply 10.. DC Pcver Supply 11.

Residual Heat Renoval Service Water System 12.

E=ergency Service Water System 13.

Leacter Protection System 14.

Reacter Core Isolatien Cooling System (Alternative Use Only) 15 Engineered Safeguard Ecoms Heating and Ventilating System

16. Control Building Heating and Ventilating System 17 Standby Diesel Generator Rcom Ventilation System

18. Energency Service Water Pump Rocs Heating and Ventilating

-System 19 Intake Structure Heating and Ventilating System 20.

River Intake System t

f21. Electrical and Control Panels a

f 22.

Pumphouse Drain Sump 23 Leak Detection Systems i

f 2k. -Centainment' Atmosphere Centici

[

r

?25. Main Feedvater (Alternative Use Only) i t

i

.26.. Ancillary Cenpenents

27. Area Radiation Monitoring (Alternative.Use Only)

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28. Nuclear Boiler / Containment Systems 29 UUREc'0576 Modificatien3

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SYSTEMS LIST

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GE E'4h 3

.1.

' Engineered Safeguards Actuation f

2.

Reactor Protecticn System l

3.

. containment Isolation T

n h.

Main Steam Isolation 5

High Pressure Coolant Injection 6.

Lev Pressure Coolant Injection 7

Autc=atic Depressurisation System a

8.

Core Spray i

9 Centain=ent Spray e

10. Residual Heat Removal i

11.

Standby Gas Treatment l

12.

E=ergency Fever 13 Service Water I

ik.

Radiation Monitoring 15 Radiatien Sampling

16. Ccabustible Gas Control 17 CRD Hydraulic System

[

18. Cicsed Ccoling Water System 19 Condensate and Feed Water System 20.

Reacter Water Cleanup System

21. ' Standby Liquid Centrol 22.

Reacter Recirc stien System i

23. Reactor Core Isolation Cooling r

TRE Systems List AZACDIE:IT 5

,c T

6 s n..i ~. y s n.

I.

Equierent is 2:alified fer Phnt Life r

a.'

Equipment =eets all w -licable t

require =ents of DDR delines or :.v!20-0588.

i i

l b.

Qualification by judgement =sy be l

acceptable with sufficient basis.

i Il.

Equirrent is Qualified with Festrictices 1-Equi;=ent =eets all applicable require =ents of DOR Guidelines er 1.TFIG-C588 vith the follevin6 limitations:

l a.

Equip =ent Qualificatica for service life less than the plant life.

l i

b.

Equipment. requires modification to

=eet qualification require =ents, such as relecatien er shielding.

1 III. - Equirrent is Exeerted fren Qualificatien t

Equi;=ent where safety related function can te acce=plished by reduniant fully qualified equi;=ent which =eets single failcre criteria.

IV.

Cualifica'.icn of E ui; tent Unresc1ved I

l a.

Qualification ~ertinc scheduled, but not ec plete.

i b.

Qualificatic: ?.eccrds cearch still in pregress.

V.

Equi;=ent ::ct Oualified

+ n.

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0 LER's None LER'S ATTAC!DIENT 7 l

l

-... 3 UNRESOLVED GENERIC - SPECIFIC ISSUES

'1.

No answer was rer received to the Generic Issues, Attachment 3a, discussed in ' attachment 2 of memorandum Hayes to Jordan dated June 20, 1980.

2.

.there a're no unresloved specific issues.

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P Unresolved Generic - Specific Issues ATTAClefENT 8 e

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f.

A.

Subject ' to reviev and approval by URR/ZQB of test reports of other documentation ~.

3.-

Meets or exceeds specified parameters #.

C.

Engineer Analysis (Aging).-

D.

Engineer Analysis (Radiation).

~

/

E..

Justification for Continued Operation.

7.

-Cc=ponents which inadequate test data exist, vill be tested, shielded, relocated, or replaced with suitable ec=penents, no later than June 30, 1982.

'G.

Cc=ponents vhich vill be replaced during the March,1981 refueling j

outage.cr shielded.

H.

Cannot evaluate (?).

-I.

This note in regard to censideration of qualified operating time for cenpenents which =ust be. qualified for integratei radiation deses only following an accident.

Qualification for pressure, temperature, and hu=idity is not applicable for these ec=penents during a high-energy line break because the cc=ponent is net located within the same confined vicinity as a high-energy line.

Because.the effects of integrated deses are cu=alative and time' or

. rate independent, operating time for these ce=ponents is not applicable frc= an equipment qualification standpoint.

r

~

i J.-

These Limitorque motor operators located in the rerus roc = have been qualified for operation at 250F for 2h hours and 200F for another 15 days. The te=perature in the -torus area vill reach approxi=ately 1

280F T seconds after the postulated EPC1 steam line. break, but falls t

i==ediately back down to approximately 200F. Based en the rationale s

that this high te=perature lasts for only a very short duration and the fact that the prototype test sequence subjected the test specimen to an elevated te=perature for a much lenger duration (2h hours) than expected i

c in actual service, the referenced test progra= is deemed to be adequate.

K.

1hese Li=itorque =ctor cperators located in the stes tunnel have been qualified for operation at 250F for 2k hours and 200F. for another 15 days. The te=perature in the stes: tunnel vill reach

.approximately 300F i==ediately after the postulated main stees line

[

break, but falls back down to approximately 200F in approximately 2 seconds. Based on the rationale that this high te=peratura 'asts for a very short duration and the prototype tes+, sequence sub 9.. ed the J

test specimen to an elevated te=perature for a r.uch lenger durati0n t

(2k~ hours) than expected in actual service, the referenced test program is deemed to'be adequate.

e i

1. Eeyond reviev'r's expertise to determine specification adequacy.

Concurrence Code AZACE'TC 9

?

_4

i*eo.

9 L.

An analysis of the high-pressure coolant injection (EPCI) system operation under the postulated accidents defined by NRC IE Bulletin 79-013 has shown that EPCI system ce=ponents are not subjected to a harsh environment for those accidents requiring the EPCI system to functinn.

M.

These switches do contain Buna N diaphrag= = ate

However, the expected radiation dose of less than 1 e 10{ial.

rad is wel16 below the radiation susceptibility threshold level of 1 x 10 rad given by the NRC in Appendix C to Bulletin 79-013.

N.

GE has qualified these devices for operation at ik8F during accident conditions. GE =aintains that this te=perature is the highest average ec=partment te=perature expected to be seen where the device is located. This te=perature is for the first hour and does not take into account temperature rise by direct stes= i=pingement. Our calculations indicate that the te=perature in these areas will exceed ik8F during the first few seconds of the respective accident.

However, GE amintains that its test philosophy is justified based on the folleving: all class 1E safety-related instruments are redundant and physically separated; four devices are generally available to previde the same functien; and = cst sensing functions occur within the first fev seconds of the onset of an accident and then are sealed in by.

centrol rocm relay logic.

O.

These ec=ponent. were supplied by General Elactric and have been qualified for vperation at te=peratures up to 212F. Calculatiens of expected temperature.following a pipe break in these roc =3 indicate that the rect te=perature =ay rise to approximately 2257 during the first few seconds. Also note that the te=perature switches associated with these te=perawure elements have a setpoint of 130F. Therefore, actuation vill take place-before the te=perature exceeds 212F in *he room. Based on the reasoning that this temperature is a trantient that lasts for a short period and that the elements would not reach +ae te=perature of the surrounding a=bient within this short period, these devices are suffi-ciently qualified.

P.

See Attach =ent 3a, July 17,1980 =e=orandum Jabicnski to Hayes,

" Motorized Valves".*

Y Q.

It is intended to replace these gaskets en an interval consistent with f

their qualificatien (1 year).

R.

Qualified for less than LO years; no mentien of program to verify replacement.

S.

Procurement and installation is in accordance vith require =ents of NU3EG-0578.

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