ML18085A210

Forwards Suppl 3 to IE Bulletin 79-01B, Environ Qualification of Class IE Equipment. No Written Response Required

ML18085A210
Person / Time
Site:	Salem
Issue date:	10/28/1980
From:	Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	Schneider F Public Service Enterprise Group
References
NUDOCS 8011130008
Download: ML18085A210 (67)

Text

- ---- ----

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 631 PARK AVENUE KING OF-PRUSSIA, PENNSYLVANIA 19406 Docket No. 50-311 Public Service Electric and Gas Company ATTN:

Mr. F. W. Schneider Vice President - Production 80 Park Plaza - 15A

_ Newark, New Jersey 07101 Gentlemen:

October 28, 1980 C-"".::::C~

.-.

c*.~

-:-IC::

The enclosed Supplement No. 3 to IE Bulletin 79-0lB, 11Environmental Qualification of Class lE Equipment," is forwarded to you for information.

It clarifies two issues raised by Supplement No. 2.

These are (1) the submittal of qualification information of equipment from TMI Action Plan requirements and (2) the qualification of equipment which is required to achieve a cold shutdown condition. This action is a result of industry feedback to NRC regarding interpretation of Supplement No. 2.

No written response is required.

IE Bulletin 79-0lB dated January 14, 1980, Supplemental Information to IEB 79-0lB dated February 29, 1980, and Supplement No. 2 to IEB 79-018 dated September 30, 1980 were previously transmitted to you for action at Salem Unit 1.

They should also have been provided for information at Unit 2.

As as matter of record to ensure that the docket file is complete, a copy of each is enclosed.

If you desire additional information regarding this matter, please contact this office.

CONTACT:

S. D. Ebneter (215-337-5283)

.so111ao 008 Sincerely,

~e:.~~

Director

I,

<!(.

Public Service Electric and Gas 2

• Company _

Enclosures:

A.

Supplement No. 3 to IE Bulletin No.

79-0lB~ dated October 24, 1980 B.

IEB No. 79-0lB, dated January 14, 1980 with four attachments 1

• SEP Pl ants *

2. ~Master List (Typical)

. 3.*

• System Component Evaluation Work Sheet

4. Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment C.

Supplemental Information to IEB No. 79-0lB, dated February 29, 1980 with one attachment

1. Generic Questions and Answers to IE Bulletin No. 79-0lB D.

Supplement No. 2 t~ IE Bulletin No. 79-0lB, dated September 30, 1980 with one attachment

l. Generic Questions and Answers to IEB No. 79-0lB and Memorandum and Order (CLI-80-21) dated May 23, 1980 E.

Li st of Recently Is sued IE Bu 11 eti.ns cc w/encls:

F. P. Librizzi, General Manag~r - Electric Production E. N. Schwalje, Manager - Quality Assurance R. L. Mi ttl, General Manager - Licensing and Environment H. J. Midura, Manager - Salem Generatfog Station

.R. A. Uderitz, General Manager - Fuel Supply Department

• e UNITED STATES

• A SSINS No.:

6820 Accession No.:

8008220248

!EB Sup#3 to 79-018 NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.

20555 October 24, 1980 IE *Bulletin N.o.79-018 Supplement No. 3:

ENVIRONMENTAL QUALIFICATION OF °CLASS

• IE EQUIPMENT Description of Circumstances:

Two. issues were raised by Supplement No. 2 which require -clarification.

These are:

(1) the dates required for submittal of qualification.information for TM! related equipment, and (2) whether the equipment required to achieve a Cold Shutdown condition must be environmentally qualified -if the licensing

• basis for the plant was a Hot Safe Shutdown condition.

(1) Supplement No. 2 (Q. l, Q.S) addressed the minimum cold shutdown require-ments.

The staff position on thi~ issue is that the licen~ee must identify and environmentally qualify the equipment needed to complete one method (path) of achieving and maintaining a cold shutdown condition.

The equipment of other paths must be reviewed to assure that its failure will not aggravate or contribute to the accident (ref.

Q.5 Supp. No. 2).

Due to an inconsistency between Supplement No. 1 and Supplement No. 2, the staff_ position on this issue was unc.lear.

Therefore, the following will apply:

a~

The qualification information for equipment needed to achieve and maintain a Hot Safe Shutdown condition must be submitted not later than November 1, 1980.

b.

The qualification information for equipment required to achieve.and maintain a Cold Shutdown condition (ref. Q.1 and Q.5 of Supplement No. 2) must be submitted not later than February 1, 1981.

(2)

IEB 79-018 required a 90 day response which was due in mid-April 1980.

• Supplement 1(Feb._1980) informed licensees that equipment which was 11planned11 to be installed as a result of lessons learned need not be addressed in that response.

Some of this equipment has since been installed.

Supplement No. 2 (Q.S, Q.21). identified that the staff posi-tion was that equipment which is* installed should be treated in a manner similar to all other safety-related electrical equipment and be addressed in the Nove~ber 1, 1980 submittal.

This position represents no change in staff position r~garding the scope of the review.

However, since the staff position on this issue was unclear the following will apply:

a.

b.

Qualification information for installed TMI Action Plan equipment must be submitted by February 1, 1981.

Qualification information for future TMI Action Plan equipment (ref.

NUREG-0737, when issued), which requires NRC pre-implementation review, must be submitted with the pre-implementation review data.

.e

!EB 79-018 Sup #3 October.24, 1980 Page 2 of.2

c.

Qualification information for TM! Action Plan* equipment :currently under NRC review should be submitted as soon as possible~

d.

Quaiifi'cation information *for TMI J\\ction Plan equipment not yet

. installed which does not require pre*implementation review should be submitted to NRC for review by the implementation date.

The above items 1 and 2 represent no change in staff position regarding the

• scope of *the 79-,0lB Supplement 2 review.

IE *Bulletin No.79-018 was issued under *a blanket GAO.cle.arance (8180225

• (R0072), clearance expired July 31, 1980) specifically for identified generic problems.

Suppl~ment.No. 3 to Bulletin 79-018 is for information, hence no GAO clearance is required.

I

• ENCLOSURE 1.

.1 UNITED STATES NUCLEAR REGULATORY COMMISSION.

SSINS No.: 6820 Accessions No.:

7910250528 OFFICE OF.INSPECTION ANO ENFORCEMENT WASHINGTON, O.C. *20555 lE Bulletin No. 79-0lB Date:. January.14, 1980 P.age l. of 4.

• ENVIRONMENTAL QUALIFICATION OF CLASS IE EQUIPMENT Description of Circumstances;*

IE Bullet.in* No~* 79~_01 *required the licensee* to perform a detailed*. review* of the. environmental qualification of Class IE electrical equipment to ensure that the equipment will function under- (Le. duri'ng and following) postulated

• accident condi ti ans.

The NRC staff has completed the initial review of licensees' responses to*

Bulletin No. 79...,;0l *. *Based on this review, additional information is needed to facilitate completion of.the NRC evaluation* of the adequacy of environmental qualification of Class IE electrical equipinent in the operating facilities~

In addition to requesting more *deta i.l ed information, the scope of this Bu 11 eti.n ia: expanded to. resolve* safety concerns relating to design basis environments.

and current qualification criteria not addressed in the facilities' FSARS.

These include high energy line breaks (HELB) inside and outsi.depriinary contain-ment,.. a.gi ng, and submergence.

• , "GUIDELINES FOR EVAWATING ENVIRONMENTAL QUALIFICATION OF CLASS IE ELECTRICAL EQUIPMENT IN OPERATING REACTORS 11, provides the guidelines and criteria the staff will use* in evaluating the adequacy of the licensee's Class IE equipment evaluation in response to thi.s Bulletin.

In general, the reporting* prob.lems encountered in the original responses and the-additional information needed can be grouped into the following areas:

1.

All Class IE electrical equipment required to function under the postulated accident conditions, both inside and outside primary containment, was not.

included in the.- responses.

Z..

In many cases, the specifi'c information requested by the Bulletin for each component of Class IE equipment was not reported.

3. *Different methods and/or formats were used in providing. the written evidence of Class IE electrical equipment qualifications.

Some licensees used the System Analysis Method.which proved to be the most effective approach.

This method includes the following information:

a.

Identification of the protective plant systems required to function under postulated accident conditions.

The postulated accident conditions are defined as those environmental conditions result.ing

. from botn LOCA and/or HELB inside primary containment and HELB outside the primary contairiment.

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Enclosure l

• 1 e**.

IE Bulleti"n No. 79*018 Date:

January 14, 1980 Page 2 of 4 b *.. Identification-.of the Class IE electrical equipment items.within

. each of the systems identified in Item a, that are r.equired to

function under the postulated ac.cident conditio.ns.

~

. c-. *. *The* correl'ation.b.etween the environmenta1 data requirements spec:ified

.. * ****in the FSAR and the environmental qualificatjon test* data fo-r each

.. Cl~s IE electrical e.quipment item identified in.Item b above.

4.. -

• Additional data not previously addressed in IE Bulleti~* No.: 79-0l are ne*eded to determine the adequacy of the environmental *qualification of Class. IE.. e 1 ectri cal equipment.*. *These data address *component. aging and operability in a submerged condition.

Action To Be Taken By Licensees Of All Power Reactor Facili"ties With An -Operating License (Except those ll SEP Pl ants. Listed on Attachment 1). *

• 1;. *

• Provide-a*.. master list11 of all Engineered Safety Feature Systems (Plant Protection Systems) required to function under postulated accident conditions.

Accident conditions.are defined as the LOCA/HELB inside containment, and HELR outside containment.

For each system within (including cables, EPA's. terminal blocks, etc.) the mas.ter list identify each Class IE

• electrical equipment item that is required to function under accident conditions. Pages 1 and 2 of Attachment 2 are standard formats to be used foY.. the 11master l is~* with typical information ; ncl uded.

Electrical equipment items, which* are components of systems listed in Appendix A of Attachment 4, which are assumed to operate in the FSAR safety analysis and are relied an to mitigate design basis events are considered within the scope of this Bulletin, regardless whether or not they were-classified as part of the engineered safety features when the plant was originally licensed to operate.. The necessity for further up grading of nonsafety-related plant systems will be* dependent on the-outccme of the licensees and the NRC reviews subsequent to TMI/2.

l.

For each class IE electrical equipment item identified in Item 1, provide

• written evidence of its environmental qualification to support the capa-bility of the item to function under postulated accident conditions. For those class IE electrical equipment items not having adequate qualifica-tion data* available, identify your plans for determining qualifications of* these items and your schedule for completing this action.

Provide this in the format of Attachment 3.

3.

For equipment identifed in Items l and 2 provide service condition profiles (i.e., temperature, pressure, etc., as a function of time). These data should be provided for design basis accident conditions and qualification tests performed.

This data may be provided in profile or tabular form.

.lJ e

Enclosure I

• IE Bulletin No.79-018 Date*:-

January 14 > 1980

.Page 3 of 4

. 4.

• 5.

6.:

7.

'--~.

. Evaluate the q*ualification :~f your c1:ass* IE electrical equipment against the guidelines provided in Attachment 4... Attachment 5, uinte*rim Staff

..

• Position on Envi:ronmental Qualification of Safety-Re1ated Electrical

• Equipment," provides supplemental information to be used.with these

. guidelines.. For the equipment identified as having 110utstanding Items"

• *by Attachment 3, provide. *a detailed "Equipment Qualification Plan.

11 lnclude..in this plan specific actions which will be takerito determine equipment qualification and the schedule for compl et.i ng the actions.

Identify.the maximum *expected flood level inside the primary* containment resulting from postulated accidents. Specify this flood level by elevation such* as the 620 foot elevation *. Provide this infonnation in the fonnat.

o.f Attachment. 3 *.

. SUbmit a.

11 Licensee Event Report 11 (LERJ for any Class IE.electrical equipment.

• 1tein. which has been determi!1ed as not being capable of meeting environmental qualification requirements for service intended.

Send the L£R to the appropriate NRCRegional Office within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of identification.* !f plant operation is to cont.inue following. identification, provide justifi-cation for such operation in the LER.

Provide a detailed written report within 14-days* of i denti fi cation to the appropriate NRC. Regi ona 1 Office.

Those: itens which were previously reportad to the NRC as not being qualified per-IEB-79-0.1. do. not require an LER.

Complete the actions spec.ified. by this *bulletin in accordance with th~

following schedule:

.Ca) Submit a writ.ten report required by Items 1, Z, and 3 within 45 days from* receipt of this Bulletin.

(b) Submit a written report required by !tems 4 and 5 wi.thi n 90 days from

. receipt of this Bulletin.

This information is requested under the provisions of 10 CF.R 50.54(f).. Accordingly, you are requested to pr.a vi de within the ti me. periods specified in I terns 7. a and 7.b above, written statements. of the above infonnation, signed under oath or affinnation.

• submit the reports to the Director of the appropriate NRC Regional Office.

Send a copy of your report to the U.S. Nuclear Regulatory Commission, Office*

of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, O.t.

20555.

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IE Bulletin No. 79-018 Date: January 14, 1980 Page 4 of 4 Approved by.GAO, 8180225 (R0072); clearance expires 7/31/80. Approval was given under a blanket clearance specifically for identified generic problems.

Attachments:

1.

List of SEP Plants

2.

Master List Standard Fonnat, Typical

3.

• System Component Evaluation Work Sheet

4.

Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors 5..

Interim Staff Position on Environmental Qualification of Safety-Related Equipment (To Addressees Only)

I*

Plant Dresden 1

..

• Yankee Rowe Big Rock. Point San Onofre 1 Haddam Neck Lacrosse

. Oyster Creek R. E. Ginna Oresden Z Mil 1 s.tohe 1 Palisades

'-~

• .SEP Plants ta *IE Bulletin 79-018 Region

• III I III V'

I III

.r r III*

I III

--*-~-... _

~_.... __... _

-_.... ______ --....... ----+-------------+-----*.-**----**----*-**-*-* --***--..--i* -*

• ..;;*---;.;..-....;.;;.._;.,;;...;.~.-;;.;.;..r- --*---*-

lLT 594 *:** -

• LEVEL*TRANSMITTER x

-1LS 210 LIMIT SWITCH

• x II.. SYSTEM=

AUTOMATIC DEPRESSURIZATION SYSTEM~ (ADS)

COMPONENTS*

~..

+-------------,------------..,....------~~~~~----..;.-.;..__,J;-* *-*--* -

1..oca-c.1on Plant Identifieation Inside-Primary Outside Primary.

I Number-Generic Name Containment Canta i nment

~-~_._.........,...._,,___. _____ ~_,_,__;,~----~----~~-+-~~--------~-+-~---------------~~-

B2.1-R001 VALYE MOTOR OPERATOR x

B2.1-F003

• 621-FOlO.

• .SOLENOID VALVE PRESSUR°E SWITCH. *.

. =* **.-

.. :--:: ~:~.

x

• x

-~tacnmerrf No~ z. to. IE Bulletin *19-01s III.

sYSTEM~--- RHR EQUIPMENT/GOMPONENTS~*(ifyp1cal)

- o---*

n-F.,

a.

. * ***~--- ~-*,:_--~-- --~*;,*.. ~ :_- e.. **COMPONENTS~:

• i

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Location Plant Identification Nu;:Jber*

___... --- -. - -... ~ -

16xP455.*

Generic Name 0-RING GASKET Inside Primary -

Containment x :.

Outside Primary

• Canta i nment

~~C~~E=P=~~;~~c=t~as~s~~~E~,~,~-~.*-:--~*-~.r--_--~---.------.-----+----.;.._~-~-~--~-~-.::.:;;;,;--**=.~--~-~---:.:;,:_-~---~**~*=*~-~-:=;..:~--~

• . -.Westinghouse, lOOC.

ELECTRICAL PENETRATION ASSEMB~Y-:.. ~ -~: __ ~- ~~

KULKA No.*. ET35. * *. -

TERMINAL BOARD

• ONKONITE,-looov, 3t ~ *.

Black POW.ER -CABLE

~i..-:-

X BRAND lOW-40.. -

LUBRICATE OIL

---,-

• To I<B59 *('Boston --

Wire & Cable)

~:-. --------*-

Cutler Hanr.ler TB No. 6 RAY CH EM XYZ Scotch Ho INSTRUMENTATION CABLE TERMINAL BOX

• CABLE -.SPLICE.

INSULATING TAPE T&& Na. 10 INSULATE)

TERMINAL LUG Y Brand Epa;cy* No~.

SEALA.NT

x.

_x x

---* --.. --**.. *-:.. -- t--------

x x

x x

x

x.

x x

X.

x

~*-~-~-~--=-=-=-=-=-=-=....::..::=l=T=l:::t.~-------:---~~~-+~--__;;;-=~-~--~---~--~-~--~--~--~--~*~-~--~-~--_.;;;;;:.:.;::~;:::;,~~~--

• When a. component is not identified by plant identification number, use the manufacturer, made 1 number, seri a 1 n.umber-, etc..

• • Like components may be referenced.

1 Fnc111 ty i

. Unit;.. 1 1

. I 1'

i j * '

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SYSTEM COMPON~NT ~V/\\LUATION WQRK SllEET

. (Ty~1 cal ) *.. :

Attachment \\No. 3.to IE Bulletin. No. 79-01

.. Page l of 3.

' 0 ocket;

~

I EQUIPMENT DESCRIPTION

• ENV I RONME~T DOCUMENTATION.REF*

\\

• ~

1.

QUALIFICAJ oN oµTsTANDING *

~p~:~.L~t~~Sp-~~c1MfT1--. ~Q~1~a~l~1~f'1~-~-~sp~e~c,~r~1-~~4~~~1T1t~-,--~.. : M~THOO*.

\\

IJEMS

. i I

i ar~mp er*

cation cat1on catjon.

catjoo *

• Sys lem :.

RHR Op~r.at1ng l5 m1fl*

JOO 1i11n, l

• 5..

• S1mu1taneou~.1 *None Plant, ID No. lPT466

T 1 me 1,

1 i

• Test i. ' :,.*,. *!

,. 1,. 1 '.*

Component:

PRESSURE TRANSMITTER*

Manufacture:

Fischer-Porter Co *.

Model Number:

50-EN-1071~DCXN-NS

. 1 t

Funct oh:.

. *: i 1.!'.* i. *,.

. Accident Mon1torhlg*;l.. !t Accuracy: Spec:

5%

Demo.n:

4%

Servtce:

RHR Pump* lA..

Discharge *Pressure *

.:. S/Nl07 Locat1on:

Conta1nmf;!nt

• I Tempera'ture

.. '(Of).

I Pressure (PSIA)

. Re lathe HumidHy(%)

Chemical Spray SEE ACCIDENT ANO TEST PROFILES PROVIDED 100%

100%

I l

,~ 1 I

l l

Rad1at1on 4xl0 6ra~~

1.2~10 8rad~ *

2.

ii ~

I f I I

i i

Aging 40 yrs 40 yrs 3

I; flood 'Level Elev:

620' I *,

• Not Not,,

Above Flood Leve 1 :

• Yes Submergence ; Requ1 red Requ1 red No X*

'II i

I

. I 5

5 5

6

• f**.

7, 8

.. I S1multaneou*

Test. i None,**

1 S1mul taneou*

None Test Simul taneow None rest See Note 1 Sequential Test'***

.,

• None 1

• I

1. Sequent1i l Test None

2. Eng* Ana ysis

• None.:

. See Note 2 I

I

(' *. '*

tDOc~menta t1on Referen~~~ d :_.. I j.

• Notes: *

• * *

• I *

1.

XYZ Lette:r ~o, 237-1,* d~ted November 2, 1979, i 1.

FSAR Chapter J, -Para~raph 3.11 J ~.

FSAR Chapter 14, Par~~raph 14.~.JJl

3.

Technical Spec1f1cat19n 3.4."1, Pa~agraph A has been sent to MFG.j requesting the qua11f1cation !

4. Technica1.Specfffcat1Qri 4.6.5, P~~agraph D 5,, FIRL T~~t Report No. i i~oo dated Nqvember 2, 1972 *

6.

Fischer and Porter c9t Test,Repor~ No. 2500-1.*

7. ;A. D. DOD Eng~n~er1n~1Evaluat1on 4ata.Report No. 6932

. a,,, Hylle Laboratory."eprr.R~'. :~6~ ',

. 1nformat1on.*

• If quall1f,cation not detennined acceptable by Decembe,r 15, 197~ ~ ~omponent.

w1l l be replaced durfog refue 11 ng: outage March 1980,

• !I, I

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In the FSAR submergen'ce was. not *considered an environmental parapiete~. ABC ~aboratory.

15* to perform submergence test in April 1980.

\\*

I* to IE Bulletin 79-018

.*Page 2 of 3 SYSTEM COMPONENT EVALUATION WORK SHEET

.INSTRUCTIONS

1. * *. Equipment

Description:

• . *

• Provide the speci fie i nformat.ion requested for

.: each Class: IE electrical component.

• P.rovide component location,. specific information such as.the building, access. floor elevations, and whether*

• _ : the component is above the fiood level elevation.

In addition, provide

• . the _specified and demonstrated accuracies of a1l instruments for their

. tr.ip functions and/or post actident monitoring requirements.. Cables, EPA's, terminal blocks, and other items shall be identified as part of the_.engi_neered safety features syste_ms.

2..

Environment:

L. i st. va 1 ues. for each en vi ronmenta 1 parameter indicated.

• .List the 11specification -values 11 obtained from postulated accident analysis

• in the 11SPEC11 column.

List the "qualification va.lues 11 obtained from test reports,. engineering analysis data, etc. in the "Qual" column.

Tempera-ture*,.pressure, etc., as a function of time shall be* provided in profile or tabular form *. Specify the--time period that the component or equipment is required ta function and identify the document which provides the basis for this time intervaT.

It is expected that some listed parameters were not requested of the licensee at the time of their license issuance.

Address each parameter condition during this review. If it is determined that a parameter such as submergence or a service condition such as aging was not previously considered, identify it as an 110utstanding Item. 11

3.

Oocumentation

Reference:

Reference the documents from which information

• was obtained in the nspec 11 column.

Identify the document, paragraph, etc.,. that contafns the postulated accident environmental specification data-. In the* "Qual" column identify the* document, paragraph, etc., that contains the environmental qualification data.

4.

Qualification Method:

Identify the method of qualification.

To* describe the qualification method use words such as simultaneous test, comparison test, sequential test, and/or engineering/mathematical analysis.

Wortis such as utestu and/or "analysis 11 when used alone do not adequately identify the qualification method.

5.

Outstanding Items:

Identify parameters for which no qualification data is presently available. Also, identify parameters,-service conditions, or environments not previously address_ed during FSAR environmental quali-fication analysis such as submergence, qualified life (aging), or HELB.

Identify in the "Notes 11 section on page i of this attachment the actions planned for determining qualification and the schedule for completing these actions.

I 1.'

EQUIPMENT DESCRIPTION NOTE 1 I'

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POSTUl-AT~O ACCIDENT ENVIRONMENT NOTE 2 I

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I.,*/

• .. 1 ii

,, 1 of I~

1 ~ullqt,n 1 79~0~~.

page 3 of 3 I

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..,.2-S~RVlt~ CONDlTION PROFILES QUAl-lfJCATlOti TEST

~NVIRONM~NT NOT~ 3 ACCURACY ACCURACY REQUIREMENTS DEMONSTRATEP

. NOTE 4 NOTE 5 *

.EXCEPTIONS OR

• REMARKS NOTE 6

~!!,:,

NOTES:

1.

2.

3.

4.

5.

6.

Refer to "Equipment Descr1pt1Pn" on Page l of th1~ ~oclos~re.

• Provide suffic1ent*values of temperature an~ pr~~sure as a function of time in tabular f~r~'to draw a characteristic profil~.

Provide sufficient values of temp~rature and pressure &s & fuoctipn pf time fpr, whicn. eq~ipm~ntwas qualifie.

• to draw a characteristic profile. Present th1s informati~Hl in tabulilr form.

Provide the accuracy requirements for sensor~ and transmitters for trip funct1ons and/or post ac<;1dent monitoring.

as used in the plant safety analysis.

Provide the accuracy demonstrated l)y sensors and transp11tters during the qualification test regarding the trip functions and/or post accident monitoring as appl1cabl~.

, J t I

I*

I I Identify any exception or deviation between specified s~rv1ce conct1tlon and qualification service condition and justificatiol) tp e~p.l~j,-i,iaq:eptance Qf deviaijQn.

• i *

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~:*

Attachment No~ 4 -tF"""rE***sulTeTiii""" 7g:.u1s- ~

.. 'siJroar;AR. **~~L~~;~~~~;* :~~~O~M:T~~

LIFICATION 1

I*

OF CLASS IE ELECTRICAL EQUIPMENT IN OPERATING REACTORS l.o Int?"Oduction 2.0 Discussion.

3.0 Identification of Class IE Eguiprnent 4.0

• Sel""Vice Conditions

---

----- --- -* r -

4.1 Service Conditions Inside Containment for a Loss of Coolant Acdden( (LOCA)

T.. Temperature and Pressure Steam Conditions

2.

Radiation

3.

Submergence

~. Chemical Scrays 4.Z Service Conditions for a PWR Main Stear.i Line Break (MSLS)

Inside Containment

l. Temoerature* and* Pressure Steam Concitior..s Z.

Radiation Subme~enc:e

4. Chemical Scrays 4 *. 3 Service Conditions Outside Containment 4.3.l Areas Sub*ect to a Severe Envil"'Onment as a Result of a Hioh Enero Line Break HELB 4.3.Z Areas Where Fluids are Recirculate<:! From Inside Containment to Accomclish Lono-ien;: Erne!"'aencv Core*Cooling Following a LOCA 1..

Temoerature, Pressure and Re1 ative Hurni di ty

2.

Radiation -

3. Subr.iercence

4.

Chemical Sorays

---

~-

J..':;

t\\.. 1.a1:;11111e11 i;.

• ~u*. '+ i;.u..L co. OU I I t= i;. lrl 1;;1-U

. I 4 Paga 2 of 33. 9*.* -

- 2.;..

4.3.3 Areas Nonnal1y M*a-f:~.tained at Room Conditions 5.0 Qual i.fication Methods 5.T :selection of Qualification* Method

, ~-2 : otJalifi eation by TYJ?e Testing *

-

• 1. Si.mulated Service.Conditions and Test Duration

2.

• Test Soecimen *

3. - Test Seguence* * *

4. Test Specimen Aging S_ -Functfonal Testing and Failure Criteria 6-Insta 11 ati on Interfaces

_.. ------*-*-----S~l a Combination of Methods. (Test, EvaJuati on,*

5.0. Margin T_a Aaina S.cr Documentation Appendix A - Typical Equi')lnent/Functions Needed for Mitigation of a LOCA or MSLB Accident Appendix S - Guidelines for Evaluating Radiation Service Conditions Inside. Canta i nment for a LOCA and MSLB Accident Appendix C - Thermal and Radiation Aging Degradation of Selected Materials

Page 3 of 33

. _: 6UIDELINES9R EVALUATING ENVIRONMENTAL ot9IFICATION ~-- ---

OF CLASS IE ELECTR°ICAt EQUIPMENT IN* OPERATING. REACTORS

1.0 INTRODUCTION

• . On February 8, 1979,. the NRC Office of Inspection and Enforcement issued

• IE 841letin 79~01, entitled, "Environmental Qualification of Class IE.

Equipment.n This bulletin requested that licensees for operating power

_. reactors ~omplete within 120 days their reviews of equipment qualification begun ear-Tfer in connection with IE Circular 78-08.

The objective of IE Circular 78-0S was td initiate a. review by the* 1 icensees to detennine whether proper documentation existed to verify that all Class IE electrical equipment would function ~s required in the host.ile environment which could result from design basis events.

The licensees' reviews are now essentially complete and the NRC staff has:

begun to evaluate the results. This docu~ent sets forth guidelines for the NRC s.taff to use. in its evaluations of the licensees 1 responses to IE Bulletin 79-01 and selected associated qualification documentation.

The ob*jective of the evaluations using these guidelines is to identify Class IE equipment whose documentation does not provide reasonable assurance-of environ-mental qualification. All such equipment identified will then be subjected to a plant applicatioFi.-specific evaluation to detennine whether it should be_

requalified or replaced with a component whose qualification has been* adequately verified:

These guide l i nes a re intended to be used by the NRC staff ~o eva 1 ua te the qualification methods used for existing equipment in a particular cl ass of plants, i.e., currently oper_ating reactors including SEP plants.

Attacnmem: NO. '+-"to r1:. tsUI 1et.1rr 1::1'."'uus Page 4 of 33.

-'2 -

e.

Equipm"'1tin other classes of plants *not yet licensed to pperate; or replaceme":t ~quiprnent.for. operating reactors, may be subject to uifferent

.. requi renentS such as those set forth in NUREG-0588, Inte~irn Staff Position

. *~_*on Enviromiental Qu~lification of Safety-Related Electrical Equipment *

'. ~.

_ In addition to its reviews in connection with IE Bull et in 79-01 the staff*

is e~gaged in o~er gen.eric* !"eviews that include aspects of the equipment qualification issue.: TMI-2 lessons learned and.the effects of failures of non.;.tlass IE control and indication equipment are examples of these generic reviews *.

In some cases these. guidelines may be applicable, however, this detennination. will be made as part of that related generic review *.

2. 0 DISCUSSION IEEE Std. 323-197~

1 is the current. industry standard for environmental qualification of safety-related e*lectrical equipment.

This standard was first issued as a trail use standard, IEEE Std. 323-1971, in 1971 and 1 ater after substantial revision, the cu.rrent* version was issued in 1974.

Both versions of the standard set forth generic requirements for equipment quali-fication but the: 1974 standard. includes speci*fic requirements for aging, margins, and maintaining documentation records that were not inc1uded *in the 197.1 trial use standard.

• The*:tntent of this document is not to provide guidelines for implementing eittier version of IEEE Std. 323 for operating reactors.

In fact most of-;*_

the operating reactors are not *c0tm1itt.ed to comply with. any particular industry standard for e1ectrical equipment qualification.

However, all of the operating reactors are required to comply with the General Design Criteria 1IEEE Std. 323-1974, aIEEE Standard for Quali.fying Class IE Equipment for

.Nuc1ear Power Generating Stati_ons.

11

~.

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Page 5 of 33

..e e

• . specified in Appendix.A of 10 CFR 50~ General Desi"gn Criterion 4 states fn part that yst?"Uctures. systems and components *important to safet~ sha11

• be designed to accomodate the affects of and to oe compatfole with the.*..

environmental conditions associated with nonnal operation, maintenance,.

testing and postulat~d accidents, including loss..of-coo.lant acc:idents.M The intent.of these guidelines is to provide a basis for judgements required to c:>nfirm that operating reactors are in compliance with. General Design Crit:rion 4.

3.0. IDENTIFICATION OF. CLASS TE EQUIPMENT Cl ass IE equiprne.rJt inc.1 udes a 11 el ectri c:a l equipment needed to a chi eve emer;enc-1 reactor shutdown, containment iso1 ation, reactor cor~ ecol i ng, containment and reactor heat removal, and prevention of s.igrri_ficant release

_of radioactive material to the environment...

Typical systems included* in pressurized and boiling water reactor designs to perfonn.these functions for-tne east severe pi;Jstulated loss* of coolant accident (LOCA) and main*

st_eat:li.ne creak acc:*id\\:nt (MSLB) are listed in Appendix A.

More detatl ed descri. pti ens of the Cl ass IE equ i prnent insta l1 ed at specific plan-.s can* be obtained from FSARs, Technical specifications, and emergency procadures...

Although _variation i_n nomenclatlJre may exist at the various plants, environmental qua 1 if-tea ti on of toose systens which perform* the functions

- identified in Appendix A should be evaluated against the-appropriate service conditions (Section 4.. 0),

The guidelines in this document are applicable to all components necessary for operation of the ~ystems 1 isted in Appendix A including but not 1 imited to valves. meters, cables, connectors, relays, switches, transmitters and valve position indicator$~

1

_,~---- -~tacnmem:.*No. 4 to IE Bui 1et1n 1~"."0IB Page 6 of 33

• * ~-

4.0 SERVICE CONDITIONS In order to determine the ade.quac:y *of the qual tfi cation of equi*pment it i's necessary to specify the environment the* equl-pment ts exposed to during nonral and accident conditions with* a requirement 'to. remain functional, These environments are r~ferred to as the.. service conditions*.... *

• The approved servi.ce conditions specifi:ed in the FSAR or other licensee*

su_bmittals are acceptable. un1 ess otherwi:se noted in the guidelines discussued bel~~

4, 1 Service Conditions Insi'de Containment for a Loss of Cool ant. Accident (LO_CA) l, iemoerature and Pressure "Stearn Conditions ~ In general,._.the.. c.o.ntainritent te::perature.and pressure conditi:ons as *a. function of time should be

~sed. on the analyses i.n the FSAR.,; In the specific case of pressure

.suppressi'on. ty-pe containments, the. follo"Hing minimum high tempeature

onditicns should oe used:* {.11-BWR Drywe11s

• 34QOF for-6 hours; and CZI FWR tee* condenser Lower Compartments ~ 3400f for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />..

• Z..

?..adi"ati"on - When specifying r-adiation service conditions.for equipment

~posad to* radi a ti-on dur-i ng-norma 1 oper~ting and accident condtti ons, the r.ormal operattng dose should. be added to the dose received during the course of an accident~ Gutdelines for eva1uating beta and gamma radiat1~on ser1ii:e condi'ttons for general areas insi"de containment are

~rovi~ed below~ Radiation service conditt~ns for equipment located cirectly above the containment sump; in the vicini*ty of filters, or.

su~erged tn contaminated li.quids must ce evaluated on a case by case tasis~ Gui_-delines for these evaluations are not pro~id~d in this

-**--*-**------~~~*-

Page 7 of 33 e**

Gaitma* Radiation Doses - A total ga1T1Tia dose radtation servtce condtt1on of 2 ~ 107 RADS is acceptable for Class IE equipir. ** 1t located in* ge.neral areas inside containment for PWRs wi'th dry type containments,. Where a dose _less than this value has been spec1fi_ed 1 an application specific *

. evaluation must be performed to detenntne if the dose specifi"ed is

- acceptable, Procedures for evaluating radiati*on service conditions*. -

in such* ~ses are provtaed in Appendix B, The procedures in Appendi*x B are based on* the cal cul ati on for a typi ca 1 ?WR reported in Appendi :x 0 of NUP£G.0588 T,

Ga!rtr.a dose ra.diati*on service conditions for BWRs and PWRs with ice

• condenser containments must be evaluated on a case by case oasis,

• Since the procedures i.n Appendi-x B are based on a ca1 cu1ation for a typical PK°R. wtth a dry type contai.ment 1 they are not directly applicable to S"..W.S and othe!! containment _types:

However, doses "for these other plant confi"gurat'tons may be evaluated using similar procedures with conser1ati"ve dose assumpti-ons and adjustment factors developed on a case by case bas ts..

Beta Radi-ation Dos-es

• Bet~ radi.a.tion doses genera 11y _.are ___ less __ sign'ffjcant than. gantna radiation doses for equi.~ent qua1ifi*cation" This is due-to the low penetrating powe.r of beta part1*c1 es in comparison to gamma rays of equiva1ent energy"'

Of the general* classes of electrical equipment.

in a plant (e"g"' cables.~ tnstrume.nt transmitters, va 1 ve operators 1 containment penetrati'Cnsl, electri.cal caole i"S considered the most

• - 1NUREG-058S, Interim Staff Position on Environmental Qualification of Safety~Related Electr-ical Equipment~

Ml.~cu;nmem; l'iO. '+

~o J.t. i:su 1 1 e10 in / :::i-u.Lo

. Page *8 of 33 --

. vulnerable to damage from beta radiation~ Assuming a TIO 14844 *

. source* term,. the average maximum beta energy and isotopic abundance

. wi.ll vary as a function of time following an accident. If these*

. ' -pa~~ters are c:Cnsidered in a detailed calculation, the conse?"Vative ' _*

.~b~~:~ur.ace dose of 1~40.xx 108 PADS rep.orted in Appendix *Def NUREG

. 0588 would be :reduced by approximately a facto~ or ten within 30 mils

• • of the surface of electrical.cable insulation of unit density.

An

• additional 40 mils.of i.nsulation (total of 70 mils) results in an.other

• factir of l O reduction in dose..

Any structures or other equipment in the vicinity of the equipment of interest would act as shielding to *.

furt.'1er reduce-. beta doses.

If it ca!t be-shown, by assuming a conserva-tive unshiel.ded surface. beta do~e of 2.0 x 108 RADS and considering the shielding, factors discussed. here**, that the beta dose to radiation sensitive* e~ui-pment interna 1 s would be less than or equa 1 to l O~ of the total garrma dose. to which an item of equipment has been qualified, then that equipment may be considered qualified fol'"' the total radiation environment (gamma. plus beta). If this.criterion is not satisfied the-radiation service condition* should ce determined by the sum of the ;ar.ma and beta doses.

3. Submeroence.. The. preferred method of protection again~..:L.tne._effects cf sufrcergency is to locate equipment above the water flooding level.

Specifying saturated steam as a service condition during type testing of-equii::cent that will.become flooded in service is not an acceptable alternative for actually flooding the equipment during the test.

Pa~e 9 of 33

4. * 'CrJnta.inment Sara'- Equipment exposed to cli.emiJP sp~ sh.ould De qualified for the mst severe chemi eel envtronment (a*ctdi'c or basic) woich could exist~'- Deminera1ized water spr~ should not be exempt from consideration as a potentially adverse service condition..

• 4.-2 Service Conditions for a PWR Main Steam Line BreaK (MSLB) Inside Containment Equi;:znent required to functfon in a steam line break environment must

• be. qt:al ified for the* high temperature and pressure that could result.

In sane cases the environmental stress on exposed equipment may be

. higher than that resulting from a LOCA, in *others it may be no more severe-than for ! LOCA due to the automatic operation of a containment spray system.

1.

Te!i:!'Oera:ure and Pressure Steam Conditions - Equi pnent __ q~~] iJied_ fo-r:_ __ _

a LOCA envi ronmant is considered qua 1 i fi ed fo i a MSLB ac.ci dent environ-i:er.t irrplants with automatic: spray systems not subject to disabling single ccrnponent failures.

T.his position is based on the* nsest

~°':ir.'iate 11

• ca1cul ation of a typi c:a1 plant peak temperature and pressure anc= a therr:ia: analysis of typical components inside containment.ll The fina I acc:ptabi 1 ity of this approach, i.e., use of the nBest Estimate 11 is pending the completion of Task Action Plan A-21 '*Main Steamline 5reak Inside Containment.

Class IE equipment installed in plants without automatic spray systems or plants wit~ ~pray systems subject to disabling single failures or delayed in.itiation should be qualified for a MSLB accident environment detennined by a plant specific analysis. Acceptable methods ls~ HU~E6 0455., Short Tenn Safety Assess111erit on tne Eri~i-~~m;~-tai*-~

Qua1ific:ation of Safety-Related E1e~trica1 Equipment cf ~EP Operating.

React.ors, for a more detailed discussion of the best estimate calculation.

. l"\\... COacrnm::rr~ l~O *. 'f-~

!:SU I I ei:.1 n I 9-0IB

'. Page 10 of 33

• .e 9* -

for performing. such an analysis for operating reactors are provided in Section 1.2 for Category II pl ants in 'NUREG-0588, Interim Staff*

Position on Environmental Qualification of Safety-Related E1 1?Ctrical Equipc:Jent,.

z.

• Radiation - S~ as* Section 4. 1 above except that a conservative.

gaima dose of 2 :x 106 RADS is acceptable.

3 *. Sub~erasnce* ~,Same as Section 4~ 1 above.

4.. Che::ical Sor"avs - Same as Section 4.1 above..

4.3 Ser"lice Conditions Outside of Containment 4.J.l Areas S:ibject tc a.Severe Environmentas*a Result ofa.Hiah Eneray L.. S

~ 'H-L~)

ine r:air;. ;, :. ~

Ser-Vice conditions for areas outside containment exposed to a HELB were evaluat;d on a ~lant. by plant basis. as part of a program initiated. by".

• the staff i:1 !Jecanber,.. 197Z to evaluate the effects of a HELB.

The

!!quip;nent r:quired to mitigata the event was also identified.

This.

equipme!It should be qualified for the servi ca conditions reviewed and approve= :n ~"le ri~~=- Sa:::y Evaluation Report for each specific plant.

4.3.2 Areas ~iere Fluids are Recirculated from Inside Containment to Accomclish Lonc-irm Core Coolina Followino a LOCA T.

ieµ:)e!"ature and R~lative Humiditv - One* hundred perce11.t _

_r_tj_~_t.iv~_ hum.idity shouTd be estab1 ished as a service condit_ion in confined spaces.

The teir;>erature and pressure as a function of time should be based on the*

p1_a.'1t unique ana1ys7s reported in the FSAR.

~.*

---

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-9 e..

2. - Radiation - Due to diffi:aren~es in equipment arrangement within these areas and the significant effect of this factor on doses, radiation service conditions must be evaluated on a case* by case basis.

In general, a dose of at least 4 x 106 RADS would be expected.

3..

Su.bmeroence - Not applicable.

4.

Chemical Sorays - Not app1icab1e.

4..* 3. 3 Ar=!as Norma 11 y Maintained at Room Condit i ans

• C1 as:. IE equipi.ler.t 1 ocated. in these areas does not e.xperi en*ce si grtifi cant stress due to a change in service conditions.during a design basis event.

This equipment was designed and installed using standard engine~ring prac!icss and incustry codes and standards (e.g.,_ ANSI, NEMA, National

Electric Code).

Based on these factors, failures of equipment in these areas during a desi sn basis event ~re expected to be random except to t.lie ex'tent that they may be due to aging Ol'"" fa i1 ures of air conditioning or ventilation systems.

Therefore., no special consideration need be given to the. environiitental qual i fi caticin of C1 ass rE equipment in these areas provided the aging re~uirements discussed in Section 7.0 below are satisfied and the areas are maintained at room conditions by redundant air conditioning or

• ventilation syster.is served by the ens i te emergency e 1 ectri cal powe?"" sys tern.

Equip:nent located in-areas not served by redundant systems powered from onsite era~r9ency sources should be quali.fied for the environmental extremes which could result from a failu~e of the systems as detennined from a plant specific analysis.

5.0 QUAL!F!CATION MITriOOS

Page 12. of 33

• e

5. l Selection of Qualification Method*

-, The choice of qualification method employed for a particular application of equipment is largely a matter of technical judgement based on such..

factors as: (1) the:severity of the s~r.vice.co~di_tioris; (2) the stractura1 and*tii!terial 'canplexity of the equipment; and (3) 'the degree of certainty *.

required in the cjual ifi cation. procedure "c Le., the safety importance

• . of the equipment<function).

Bas.ed on these considerations". type testing is tlie preferred method of q~alification for electricai" eq~ipment 1oca.ted

• 1nside* containmen! required to.mitigate: the consequences of design basis

.events> i.e., C1ass. IE equipment (see Section. 3.0 above) *. As. a minimum~.

the qualification for severe-temperature,: pressure, and steam service conditions for Class IE. equipment should be based on type testing.

QuaTifica!i on fol'" other serviCe conditions such as radiation and chemical.

sprays may be by ana*1ysis (evaluati.cn) supported by test data (s~ Section S-.3 below).

E:cc:e;:itio11.s to these general guidelines must be justified an a case by case basis.

5.2 Ouali~ica~1or. bv iyPe Testino The eva*iuation of test plans and results* should include consideration of the fo1 lowing factors:

l.: Sirr.ulated Service Conditions and Test Duration - The environment. in-the.-------.----*

test chamber s~ould be established and.maintained so that;-; envelopes the service conditions defined in accordance with. Section 4.0 above.*

Tne time duration of the test should be at 1 east as long a~ the period from the initiation of the accident unti 1 the tempe?'"a.ture and* pressure service condi:ions return to.essenti.ally the same levels that existed before the postulated accident. *A shorter test duration may be acceptable

J\\W:QQ\\llleJHUC:il Cl 110

• 1 illliW WU I I WW.. IT{ --- Page 13 of 33 e**

if specific analyses are provided to demonstrate that the materials involved t 11 not experien~d*e significa.nt atcel erated thermal aging during th*~ per_"iod not te~ted.

z. Test. Scecimen - The test specimen should be the same m.94..eL_a.?_th.~,...

equipment being qualified. The type test should only be considered valid for equipment identical in design and material construction to the test specimen.

Any deviation~ should be evaluated as part of the qualifica-tion documentation (see also Section 8.0 below).

3. Test Seauence - The component being tested should be exposed to a st:ar:i/air environm~nt at elevated temperature, and pressure in the sequence defined for its service conditions. *Where radiation is a se!'Vica condition which is to be* considered as part of a type test, it may**be app1ied at any time during the test sequence-provided the component does: not contain any materials which are known to be susceptible tc significant radiation damage at the service condition levels or ma~erials whose susceptibility to radiation damage is not known (see "p-e~d.:x ""'

f"I !" '*. I i,_ J

• If the component contains any such materi a 1 s, the radi ati an dose should be applied prior to or concurrent wi:h exposure to the elevated

_teperature and pressure steam/air environment.

The same test specimen shculd be. used throughout the test sequence far all service conditions.

the equipment is ~o be qualified for by type testing. *The type test shcu1d only be considered valid. for the service conditions applied to the same test specimen in the appropriate sequence.

4.

Te.st Sceci~en Aaing - Tests which were successful using test spe.timens which had not been preaged may be considered acceptable provided the c~i:;icnen~ does not contain materials which are known to be susceptible

~* '!"..,.*

Attachment No. 4 to IE Bulletin 79-ors* *

'I

.* Page 14 ow3 _

1.

to significant degradatiOn due*to thermal and radiation agir.; (see Section

• _ 7.0} *. If ~e component contains such materials a qualified life for the

.. COfilP9nent must be established on a case by case basis.* Arrhenius techniques

. are generall;t considered acceptable for thermal 'aging..

s* *. -Functional Testing and Failure Criteria - Operational mode~--~~~~~-9._-... - ----------

... sneuld be "representative 'of the actual application' requirements (e.g., ccmpcinents which opet~te normally energized in the plant should be nonnally energized during the tests, motor and electrical cable loading during the test should be representative o~ actual o~rating conditions).* Failure criteria should include instrument accuracy requirenen ts bas*ed on the maximum error ass urned in the plant safecy analyses *. If a compon.ent fails at any time during tha test,.* even in a. so *called. "fail safe" mode,. the *test should be cons1dered inconclusive with regard to demonstrating the ability of the. component. to function for-the entire period prior to the failure.

6.

!nsta11ation Intsrfaces - The equipment" mounting and electrical or me:hanical seals used during the type test should be representative of the actual insta 11 ati on for the test to be considered conclusive.

The equipment ~ual ification program shou1 d include an as-bui1 t ir.spection in the field to verify that* equipment was installed as it was tested. Particular emph~sis.should be placed on comncn

-p~blems such as protective enclosures installed upside down with drain holes at the top and penetrations in equipment housings for electrical connections being left unsealed or susceptible* to moisture incursion through stranded con_ductors.

~----=

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Page 15 of 33 e

- 13 5.3 Qualification by a Ccrnbination of Methods (Test, Evaluation, Analysis As discussed in Section 5.1 above~ an item of Class IE.. equipment may be shown to be qualified for ~ complete spectrum cf service conditions even though it was only type tested for high temperature, pressure and steam. The qualification for service conditions such as radiation and chemical *sprays may be demonstrated by ana 1ysis (evaluation).

In.

such cases the overall qualification is said to be by a combination of methods.

Fo11owins are two specific examples of ~r=cedures that are considered ac:ceptab1 e. Other simi 1 ar procedures :nay al so be reviewed and fc~n: acceptable on a case by case basis.

l. Raciatfor! Ouaiif'ication - Some of the earlier tvcP. tris_~~perfo_nned ___. _

2.

for ope?9ati"rTg reactors did not include radiation as a service condition.. ln these cases the equipment may be shown to be radiation qu.a1ified by perfonning a* calcu1ation of the dose expected, i:a:king into account the time the eq~i;:ment is required to re:11ain functional and its location using the methods described in Appendix B, and analyzing the effect of the ca1cu1ated dose on the waterials used in the equipment (see Appendix C).

As a general rule, the time required to remain functional assumed for dose calculati~ns should be at least l hour.

Chemical Scray Ouali"fication -: Components enc1osed en~~-.!"~Jy i_I}_

corro~ion resistant cases (e.g... stainless sts~1) may be shown.

to be c;ua l ifi ed for a chemical environment by an ana 1ysi s of*

• h

... h rt.,

h 1

'"e er.ects Oi.. e pa icu.ar c ern1ca s on 1.:-: ;i!!":..1cu ar enc o-sure materials..

The effects of chemical spra/s on the pressure integrtty of any gaskets or sea1s'present shcuid be considered in the analysis.

t

-14.

Attachment NO. 4 to Page *16 of 33 e;

• 6.0 Marain

-*:*::[-

IEEE Std. 323~ 1974 cl*>' ines margin as the difference between the most sev~re specified se~vice *candition-s of the plant and the conditions used_

.. *in type te~ting *to account for nonnal variation~ in cQ:mtercial production of* equipment and reas~nable. errors in 'defining satisfactory pe~fonnance.

Section 6.3.1.5 of the standard provides suggest_ed.factors to be applied to the s~rvice c~nditions to assure adequate margins *. The factor app1 ied to the time equipment is required to remain functional. is the most significant in terms of the additional* confidence in qual ificat.ion that is achieved by adding margins to service conditions when estab1ishihg tes-= environments.

For this reason, spec.ial consideration was given to t.ie time required to remain functional when the guidelines* for Fun~tional -

Tes-ting and Failure-Criteria in Section 5.2 above were estab.lished *.

In addit"fon, a11 of the guide1 ines. in Section. 4-.o for estab1 ishing service CtJncitions include ccnservatiSi.is which assure ~~rgins between the service condi!ions specified and the actual conditions which could realistically b: expectsd in a design basis event.

Therefore; if the guidelines in Section 4.0 and 5.2 are satisfiec,no separate margin factors are required to be added to the service conditions when specifying test conditions.

7.0 Aaino Irnp1 icit in the-staff position in Regulatory Guide l.89-with regard to bacKfitting IEEE Std. 323-1974 is the staff's conclusion that the incremental improvement in safety from arbitrarily requiring that a*

specific qualified life. be demor.strated for a11 C1ass IE equipment is not sufficient tc justify the expense for plants already constructed ar.doperating.. ihis position does not, however, exclude equipment

-~...

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- 15

• • us1ng materi a Ts that have-been identified as being susceptible to significant-degradation due to thermal and radiation aging.

Component ca1ntenance or replacement schedules should include considerations of the specific aging characteristics of the compon.ent materials. Ongoing

-Programs should exist at the plant to review surveillance and maintenance records to assure that equipment which is exhib_i ting age* related degrada-tion will be identified and rep1 aced as necessary. Appendix C contains a listing of materials which may be found in nuclear power plants along with an indication of the material susceptability to thenna1 and radiation aging.

8. 0 Ooc!Jr:!enta ti on Comp1e~e* and auditable records must be available for qualification by any of the methods described in Section 5.0 above to be. considered valid.

These records.should describe the qualification method in sufficient ceta.il to veri.fy that all of the guidelines have been sa:tisfi ed.

A simp1 e vendor cert.ifi cation of compliance with a design specification should not be considered adequate.

Attachment No. 4 to IE Bulletin 7~"".0lB.

Page 18 of 33

• APPENDIX A

. JYPICAL EQUIPMENT /FUNCTIONS N.EEOED FOR

.MITIGATION OF A LOCA OR MSLB ACCIDENT Engin~red Safe.guards Actu~tion. * *

• Reactor Protection..

• Containnent Isolation

'Stearnline *Isolation Main F:edwater Shutdown arid Isolation Emergency Power.

Emergency Core Coo1ing1 Contair.ment Heat Removal Canta i1T.1ent Fission. Product. Renova 1

~ontair:nent Coiti:lustib1e Gas Control

. Auxiliary Feedwater

~ntai1.nent Ventilation

~ntain.::e11t Radiation Monitoring Control Roo:n Habitability Systems (e.g., HVAC, Radiation Filters)

Venti1ation for Areas Containing Safety Equipnent Component Cooling Service Water Emergency ShutdownZ --

Post Ac=ident Sampling and Monitoring3 Radiaticn Monito~ing 3 Safety ~elated Display !nstrumentation3

. \\...

M"WQ:iri..lilll~ii"" ilU*

f c;Q

..,y11,;;;;a -lll-----,-.ii1-U-.U Page 19 of 33 e 1These,systels will differ fo.r FWRs and BWRs; and for old.. r and newer pla.;,ts.

In each case the system features which allow fo

• transfer to recirculation cooling mode and establishment of long tenn cooling wit:J boron ;:Jrecipitation control are to be. considered as part of the system to be eva l ua ted.

2Emergency shut:dcwn systems include those systems used to bring the plant to a coid shutdown condition following accidents which do not

. result in a breach of the reactor coolant pressure boundary tog~ther with a rapid depressurization of the reactor coo1ant system.

Examples of such syst~s and equipment ars the RHR system, PORVs, RCIC, pressurizer sprays, checical and volume control-system, and steam dump systems.

3Mor; specific id:ntification of these types of equipment can be found in :he pl ant emerg:nc:y procedures *

• Attachment No. 4. to IE Bulletin 79-018 APPENDIX B Page. 20 of-3.

PRClcr!!tl?.£5 AJR 87ALUATING t:...."f".A RADIATION SERVI-CE CONDrT!OflS

~-

• Introduc::ticm and DisC"JSsion

• The adequacy Df ~rma. radiation. servtte condi:tic:ins specified for inside

• . contaiane.nt durin;.a lOCA or M.Sl.3 accident canoe.verified by assuming a eonseNative dose *at the corttail'lElent centeri ine and adjusting the dose according the. plant sped ~ic panmeters~ The purpose of this appendix

. '{s to.identi*fy -toosa paraineters :i.-hose effect on the total ganma dose is easy *tQ quantify wi"tll a high degres of confidence and describe procedures which r.ay be tis;d to ~ke these :ffects into consideration.

The bases for t.ie proc:eC!.Jres a:id restrictions for their use are as follows:

(11 A ccnsar1a:ive dose at the containment c:nterl ine cf 2 ~ 1 o7 RADS for a !..OCA and 2...x 1aE YaS for a MSLB acc:.ident has been assumed.

This assmi;ltion and. all t.~e *dose ratas used in the procedure out-li.ned ~e1o'W ar~aased on the.methods a.fld sample* calculation descM~&d m Aj:pe.'ldi-:c 0 of* h'TJ?~S... QSE.3., "Interim Staff Position on Environ~ental Qua 1 ifi cation of Safety-Re 1 atad E1 ectri cal Equip-ment.,** Therefore.. a11 the 1ir.:itaticns listed in Appendix 0 of MJRS... 0588 apply to t.~ese procedures.

(2)

The s~p1e calcut.ation in Appendix D cf HUREG-0588 is for a 4,000 w..:tn pressur'fzed -Wate:- reactor Housed tn a 2.. 52 x 106 ft3 contain-cient wi"-"~ an icdine scr.xbbir.g spray system, A similar. calculation withotr-iodine sCT"U:ibin; sprey's would increase* t11e dose to equipment ap~roxiuctely 15~-. The c::inser-tative dose* of 2 x 107 RADS assumed J

• ~.**

-~-----------,--.n,..-,t,;cit,;~arc::\\OO:n11m111~errr1 i::-t;--railt:O-.- t 1.U.1.t. CU I I e.l.*I n I ~-UIC

-2 Page 21 of 33.

e in the procedure below includes sufficient conservatism to

. *. aceount for this factd:r *. Therefore, the pro*~::.dure *is. also applicable to plants without an iodine scrubbing spray system.

(3)

Shielding* calculations are based on an average garmia energy of.

. 1 :HEV. derived from TID 14844, (4-). These -procedures are not applicable to equipm~nt located directly above the containment sump, submerged in contaminated liquids,

• or-nea~ filters. *Doses specified for equipment located in these

• areas must be evaluated on a case* by case basis.

(5} *Since the dose adjustient factors use~ iri these procedures are based on a calculation for a typical pressurized water reactor with a dry type containment, they are not directly applicable to.

boiling water reactors or other containment types.

However, doses for these other plant configurations may be evaluated using similar procedures with conservative dose assumptions and adjustment.factors developed on a case by case basis.

Procedure Figures. l _ throuah 4 provide factors* to be* app1 ied to the conservative f

dose to correct the dose for the following plant specific parameters:

(1} *reactor power level; (2) containme~t volume; (3) shielding; (4) compartment volume;* and (5) time equipment is required to remain functiona 1.

. It Attachment NO.- 4 to IE BU 11 etin 79:-018 Page-22 of 33

• e t'.*

The procedure for using the figures is best i11ustrated by an example.*

Consider the following case.

The radiation service condition for a particular itan of equipment has been specified as 2 x 106 RADS.

The application specifi-c parameters are:

Reactor power level -* 3,000 MWth Containment volume - 2. 5 x 106 ft3.

Compar-..me~t V~1ume - 8,000 ft3 Thickness of compartment shield wa11 (concrete) ~ 24" Time* equipment is required to remain functional - l hr.

The probl~~ is to make a reasonable estimate of the dose that the.equipment could be expected to receive in order to eyaiuate the adequacy of the radiation se?"Vice condition specification.

• Stec 1 Enter the nomogram in Figure 1 at 3,000 MWth reactor power level and 2.5 x 105 ft3 containment volume and read a. 30-day integrated dose of 7

l. 5 x. 10 AADS.

Steo 2 Enter Figure 2 at a dose of 1.. 5 x 107 RADS and 24" of concrete shielding for the compartment the equipment js located in and read 4.5 x. 104 RADS.

This is the dose the equipment receives from sources outside the compart-ment.

To this must be added the dose from sources inside the compartment

. {Step 3).

SteD 3 Ente~ Figure 3 *at 8,000 ft3 and read a correction factor of 0.13.

The dose due to sources inside the compartment would then be 0.13(1.5x107) 5

= 1.95 x 10 RADS.

The sums of the doses from steps 2 and 3 equa 1 s:

4.5 x 1a4 RADS + 0.13 (1.5 x 107)- RADS = 2.0 x 106 RADS -

---

:----~l"\\rtr.ca:rrc".Tnrmm~err:mr:i-w-.--~o lt. JjU 1 1 e"trn t !:1:-Uljj I

I

• Page

• 2*3 of 33 4- -

Steo 4 Enter Figul".'e _4 at 1 hcur and read a correction factor of O. 15.

Apply this factor to the sum of the doses determined from steps 2 and 3 to correct the 30 day total dose to the equipment inside the compartment to 1. hour.

o.1s <~.a x* 1as1 = 3 x 105 RAos*

In th"is particular example the service condition of 2 x 106 RADS specified is conservative with respect to the estimated dose of 3 x 105 RADS *calculated in steps l through 4 and is, therefore, acceptable.

• ~., It I e

CONTAINMENT VOLUME {ft3}

~

FOR'CONTA1NMENT VOLUME AND REACTO:t POWER LOCA DOSE CORRECTIONS* e

. Attachment No. 4 to *rE Bulletin 79-018 Page_ 2~ of 33 3x106 2x10S MWrH 1x106 4000 3000 2000.

5 x10S 1000 4x.105 500

. lx ios zx ios 200 1 x ios

-~* -.

30 DAY INTEGRATED yDOSE 4 X. 10 7 3 x 107

--z x 107 1 x*107 s-x ios 4x 106 3x106

.. 2.S x 106 2.0 x 106

, x 106

'*'r.SLB ACCIDEIIT DOSES SHOULD BE READ AS A FACTOR OF l 0 LESS

~~~*

.;: :oosE CORRECTION FACTOR FOR CONCRETE SHIEt.DlNG.

~. *... : :..,

l Y"O~LYJ

• Attac t No. 4-to IE Bulletin 79-018 10S Tx107 c:J*

• Z' -

Q

..J..

w* 1 xioa co -...

.,_ -~

0 Q.

. c:: -

tU, x 10S 7 **

rn:

0

~--

Q

, x104

• • , ;(' 103 I I I I

I I

J I I I

I 108 107 106 10S

• r COSE {RADS) WITHOUT SHIELDING (FROM FlGURE il

F1GURE 3 DOSE CoR.,-ioN FACTOR FOR CO.RTMENTVOLUM=*

• Attachment No. 4 to IE Bulletin 79-018

• Page 26 of 33 106

. 105 104

,_~L.---J~--"'----~--"'--~~--"'----"'----"----~--.._ ____________ --.t o*

.2

.4

.6

.8 i.o CORRECTION FACTOR

a:.

0 t-'

u 2 1.0 =

(Y) 0 M_

lf-t-

• Ou.

.,.... lU N 0:

~ a:

• ~*o.

• U

• w

(/).

0 a

<(

i DOSE CORRECTION FOR TIME hEOUIRED TO REMAIN FUNCTIONAL

,01L--..l~L-.l-LLU..LL.~-'---L.....L-..L.J...L.LJLI-~...._-'-..J.-J.-1-L..._._.~__..~~-'-._._.J.a-~-----------

• 1 1.0 10 100 1000 TIME REOUIREiJ TO REMAl.N FUNCTIONAL (HnsJ

Attachment No. 4 to I~ Bulletin 79~018.

• *. Page 28 of 33 A APPENDIX C W

ltrER?W. AND RADIATION AGING DEGRADATION OF SELECTED MATERIALS Table C-1 is a partial list of materials*which may be found in a nuclear power plant along with an indication of the material susceptibility to radiation and thermal aging.

Susceptibility to significaf!t thennal aging in a 4S°C environment and normal atmosphere* for 10 or 40 years is indicated by an (*) in the appro-priate cillumn.

Significant aging degra.dation is defined as that amount of degradation that would pl ace in subs tan ti a 1 doubt the ability of typical equipment using these materia 1 s to function in a hos ti 1 e en vi ronrnen t.

. Susceptibility to radiation damage is indicated by the dose level and the obser1ed effect identified in the column headed.BASIS.

The meaning of the tern:s used. to characterize the dose effect is as* fo 11 ows:

1° Threshold. - Refers to damage threshold, which is the radiation exposure requ.ired *to change at 1 east one phys i ca 1 property of the material.

Percent Change of Property - Refers to the radiation exposure required to* change the physical property noted by the percent.

1 Allowable -* Refers to the radiation which can be absorbed before serious degradation occurs.*

Tne infor.nation in this appendix is based on a literature search of sources including t.1le National Technical Infonnation Service (NTIS), the National Aeronautics and Space Administration's Scientific and Technical Aerospace Report (STAR), NTIS Government Report Announcements and Index (GRA), and

Attacnment' NO. 4 to IE *1:1u 11 ei:1IT1!:t-U.1jj -

Page 29 of 33 e various manufacturers data ryports.

The materials list is* not to be r=:lnsidered a11 inciusive neither 1s it to be used *as a basis for '

specifying materials to. be used for specific applications within a nuc1ear plant. The list is *sol.ely *.intended for use by the NRC staff in making judgements as to the possibility of a particular material J

in a particular application being susceptible to significant degradation due* tO radiation or thermal *aging.

• The data base for therinal and radiation agi_ng in engineering materials is rapidly expanding at this time.

As additional information becomes available Table C-l will be updated accordingly.

TABLE C-l THERMAL AND RADiATION AGING DEGRADATION OF SELECTED MATERIALS

'l'Yl'ES OP l':QIHl'IU-:tl'I' I

l'O'J'~'rJllf, t'OR fi I mu t' J CAN1' llllDU.'1' lON SUSCEll' 111 l I. l'l'Y Al.llO AGlllG llffOWH ----

llAllS ftA1'1*:n I Al,.

AS 10 Yllfi 40 YRS

(;J\\HHA llASIS lnte<JrAted Circuits CIC)

10)

'l'hceehold x

x x

rf-HOS J11te'.)ratod Ciccuita CIC) 104

~

ic x

x C-HOS Tra1111!11toca 104 x

x x

x x

x Pio.Jue 104

. x x

x x

x X*.

Sil lcon-Controllod 104 x

x x

x

x.

x llectl f leru lnte'.)rated Circuits CIC) 104 x

x x

x x

AnalO<J Vulcanized Fiber t

105 x

x x

x x

rish Paper 105 x

x

~

x x

x x

x Polyestea*

(unfilled)

~

105 x

x x

x x

x x

x x

Nylon olya111ide "

105 x

x x

x x

x x

x

x.

x x

x x

Polycaroo11ate 106 x

x x

x x,*

x x

Polyl*lde 106 x

x x

x x

Chlorouulfonated Poly-lypalo11 10' Allowable x x

x x

x ethyle110 Duno-rf lllR/tU-l06 Threshold x

x x

x die lubller ll1teqratod Clrcuitu (IC) 106 x

ll x

x x

x

• . *1*1*1.

lJiallyl Phthalate

>AP 106 SJ Uconu Ruhbe1 106 II x

x

• Indicates that there is data available which shows a potential for sign1f1cant thermal aging of the materials when exposed to normal operat1ng conditions for either 10 or 40 years as indicated.

x x

x

.X x

x x

x x

x K

x wz WO ll

.J:>.

X*

rt-0 m

x OJ c:

ro

'f;'i-....

~

ta

. I 8

OJ

r TYrEs Ot' t::QUI~*tum*1*

ron:m JAL ftAUIA'flON fOR SUSCEl"l'IBla.I'l'V S Jr.tU f I CANT Al.SO l\\0100 ICNOWH RADS HA't1-:AIAL AS 10 YllS 40 Yll9 GMflh BASIS SIJR llubbor 106 x

x CA1>AoltDc* - Tantalum 106 Allowable x

x x

x x

x x

Del lr:ln

~

io5

• rhco11hold x

x x

X*

x x

Tohel 106 16\\ Loll&

x Ji x

of Klonga tlon G.l'. l'hanollo li'-*oso 106 rhre11hold x

x x

x x

ll x

x x x x

x RTV Bille~

io7 Allowable x

x x

x*

x Cycolao (IW'#)

101

'l'hceuhold x

K

• -0

'11

. 101 c.o lntogcatod flrcult* (let x

x x

I fD '11 n ECL w

107

....,3, t"O~VAC x

x x

x x

x x

fD 107

. 0

I NEHJ\\ l'olyo11toc GlAH x

-ti Ltaalnato11, Grade Gl'O-J x

w WO NEHA Polyo11t&C Glas*

107 Laalnatu11, Grado Gl'O-]

x x

x l'olyotbylonca 107

!lovable x x

x x

x rt 107 II 0

Huopreno x

x x

x x

x x

x x

x x x x

El'R

'thylene-101 1-i II x

x JT1

• i-opylont1 m

tullbec c:

Polytheraalozo l08 rhr11shold x x

x x

x CD Ci-01111-J.lnkod 107 llowablo rt-....

rolyethylentii

I C4paoltoc11 - Hylac 108

'-I x

x x

X*

x

\\0 I

0.....

OJ

\\

11/14/79 TVfHS Uf' ~01..11rH£tfr (~l'l'IUH WIUCll ttA'l't;&UAI. *~..- ISH t'OUNU) l'OmNTIAL RADIA'flOH FOR

&USCEPTIDILITV

&IGNlflCAHT Al.SO AGIOO IKNOWM.

llAPS HA'ft:RIAL AS 10 VRS 40 vns GAlflA PMIS i>oll(a11Uona io1 llowal>le x

"'*°*

c-ld41 108 24\\ Loua x

x x

x I

f Elonga tion Roaiatoca - Nlre-llollnd to9 rhro11hold x

I I

I I

I x

Raalatora - C41rbon io9 JI JI x

I x

x JI COtot.l()jll tlon Capaolton -

~r-io 109 UowAl>la I

I I

I I

I capaoltor* - Ola**

109 I

x I

I x

capaoitac* -.Hica 109 x

x x

• x I

"'O ):

Pl C"1 lo9 NEHP. Tha..oaattlng I

I

'° C"1 CD Ill a-Jnataa, Gcado XXXl'C n

w ::i Kt:HA ThunM>aottln<,1 to9 I"

x N3 CD r.a.inatea, Gl'ado XXXP 0

s 109

-ti rt HEMA TlwinM>aettlng

~

I

• -lnatoa, Gl'adot Xl'X w ::2 WO HEHA Thol'llOaottlng io9 I

I Laminatoa, Gl'ado life

..f:>

NEHA Thonaoeettlng io9 I

I I

x x

I rl 0

.1..i*lR4tea, Gl'ado XX tlEHA Tbel'll011ettlng 109 x

I I

x I

x

~

.Laainatoe, Gl'ado XXP NEHA 'l'hor.oaettln<J 109 I

I I

I x

x La~inatou, Gl'adu XXX 109 rf HEMA Thoraouotting x

Lalllnal:.au, Grlldci CB

s NEHA Thor110uettinrJ io9 x

lC IAainatau, GJrado C I c

I; tu'ITt:ll I AL HHHJ\\ Tho.raoaettlng LA*lnateu, Grade L HEMA Thur-.osett!ng LA*lnatea, Grade La NEHA Tho.-.011ottlng l.aa!nateM, a.-ade FR-l HEHA TholWIODottlng lA*looto11, Grade l'R-l tlEHA Tho.-aooot tlnCJ Lll*lnatoa, G.-ado fR-..

Ha\\A ThoD110oettl9 1.a.. 1nat110, Gracia rR-10 llCHA Tho.-mouottl119 1.a*ln*te11, Grado A lll!HI\\ Thon*ouottln')

IA*lnatoa, Grado AA HEHi\\ 1'hon110oettl11<J Laa~nateu, Grade G-J lll!:HA The1111011ott1119 l.aalnatoa, Grade 0-11 tlEHA ThonaosottlQCJ La*lnatou, a.-ade rn-5 lto~htou - ru.

l'O'l'l::trr JAL t'OR 81 Gii U' ICAHT AGlN(J 10 us 40 YRS llAl>IATIOH 6USCEl"l'lDJ LlTV ltMS GAlflJ\\

DA!il8 109 Throuhold 109 109 109

.109 109 109 to9 109 9

10 J09 lOlO 11/14/79 i....

1'Vl'BU OF ~OUlrHHNT (Hl'l'lflH !tlllJCll Hll'l'ERJl\\1..HAY am FOUHUI x

I I

x x

x x

x x

x I

x x

x x

x x

x x

x x

1

• I I

I I

rt' 0

• Bul let*i n

.No.

• .. 79-13 (Rev. 2) 79-17 (Rev.* 1) 79-25 79-02 (Rev.. 2) 79-26 79-27 79-28 a

ENCLOSURE 2 RECENTLY ISSUED.IE BULLETINS Subject *

• - *

• Date Issued Cracking i.n Feedwater

• 10/17 /79 System Piping Pipe Cracks in Stagnant

• 10/29/79 Borated Water Systems Failu_res of Westinghouse W2/79 BFO Relays in Safety-

. Related Systems Pipe Base Plate Designs W8/79 Using Concrete Expansion Bolts Boron Loss From BWR 11/20/79 Contra 1 B 1 ades

• Loss.of Non-Class-1-E 11/30/79 Instrumentation and Con-trol Power System Bus During Operation Possible Malfunction 12/7/79 of NAMCO Model EA180 Limit Switches at Elevated Temperatures IE Bulletin No. 79-0lB

• Date: January

• 14; 1980.

Page 1 of 1

.. Issued To All PWRs with.an OL and Designated Ap-plicants (for Action):

All Other Power

• Reactor Faci 1 it i es with an Operating License (OL) or Con*

.structian Permit (CP)

(far.Information)

A 11 PWRs w.i th an OL (for Action). All other Power Reactor

. F aci l i ti es with an OL ot CP (for In~

formation)

All* Power Reactor F aci l it i es wi th an OL or, CP (for Action)

A 11 Power Reactor.

Faci 1 iti es with an OL or CP All BWR Power Reactor Facilities with an OL All Power Reactor Facilities with an OL and those nearing Licensing (for Action)

All Power Reactor Facilities with a CP (for Information).

All Power Reactor Facilities with an OL or.CP

(.

T ENCLOSURE 1 UNITED STATES.

NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION.AND ENFORCEMENT WASHINGTON, D.C.

20555 SSINS No: 6820 Accessions No:

7912190696 ENVIRONMENTAL QUALIFICATION OF CLASS IE EQUIPMENT IE Bulletin No. 79-018 Supplemental Information Date:

February 29, 1980 Page l of I Enclosed are the generic questions and answers which resulted from NRC Task Group/Licensee 11workshop 11 meetings held recently in NRC Regional Offices regarding IE Bu11etin No. 79-0IB.

This information is intended to further the understanding *of the qualification review process and reporting requirements o.f the Bulletin.

The further intent of this information is to assist the licensees in providing a method of approach acceptable to the.assigned NRC Task Review Group. in determining adequacy of the environmental qualification of Class IE Electrical Equipment installed at their respective facilities.

It.should be recognized that the review of the licensee 1s responses may generate additional need for guidance of finalized resolution of the environmental qualification issue.

Attachment:

Generic Questions and Answers

. to IE Bulletin No. 79-0IB

Attachment to Supplemental Information to IE Bulletin No. 79-018

. GENERIC QUESTIONS AND ANSWERS TO IE BULLETIN NO...79-018 Question 1-

!EB 79-018 indicates the scope of the task.ii only that equipment exposed to a harsh environ~ent. Enclosure 4, Section 4.3.3 identified areas outside of containment not exposed to harsh environmental conditions as the results of an accident. Should these areas be included in our evaluations?

Answer 1 No. Although the guidelines encompass all safety-related electrical equipment and components, the scope of IEB 79-018 fs: limited to only that electrical equipment which is exp*osed

• to the harsh environments identified in action item 1,

• including where fluids are recirculated from inside containment to accamp 1 i sh long-term cooling fa 11 owing a LOCA. A 11 equipment and components identified in action item 1 shall be included in the subsequent actions required by IEB 79-018.

Question 2 IEB 79-0lBaction 1tem* 1 and Enclosure 4 indicate that emergency procedures be used to identify equipment to be included in the master list. Should all the equipment identified in the emergency procedures be i nc.1 uded in the master 1 i st?

Answer 2 All the equipment the licensee relies upon in the emergency procedures to mitigate design basis events that may be exposed to a harsh environment must be identified in response to Question 1.

It is not the intent of this task to change

. the existing procedures by removing references to equipment or components that are considered nonessential and not environmentally qualified. This master list identifies all equipment and components that must be evaluated in response to action item 4. A determination should be made that sufficient equipment is environmentally qualified to permit accident mitigation. A tabulation of other equipment or components which are referenced in the Emergency Procedures but are not relied upon should be available for NRC review. Justification should also be available so that this non-qualified equipment wi 11 not be mis 1 eadi ng to the operator..

Question 3 Is note 2 of Appendix A to Enclosure 4 within the scope of this task?

Answer 3 Only those emergency shutdown systems that could be used for mitigation of a LOCA or HELB and are exposed to a harsh environment identified in response to Question 1. Licensee review should:

(1) identify equipment that could be used to achieve cold shutdown following LOCA or HELB; and (2) determine if environmental qualification exists. For equipment that is not environmentally qualified the licensee should either provide plans to qualify this equipment or provide justification that qualification is not needed to achieve safe shutdown to meet licensing requirements applicable to your facility.

. I

Question 4 Answer 4 Question 5 Answer 5 Question 6 Answer 6 Question 7 Answer 7 Question 8 Attachment to Supplemental Information

.to IE Bulletin No. 79-018 2 -

What is the basis for the 340 Degrees F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> req~irement identified in Enclosure 4 and NUREG 0588, Figure C l?

for minimum high temperature ~onditions in pressure suppression tyPe containments, we do not require that 340 Degrees F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. be used for BWR drywells or that 340 Degrees F for 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />sbe used for PWR ice condenser lower compartments. These values are from a bounding high temperature-profile (see Section 1.1 and 1. 2 of NUREG 0588) that can be used in. lieu of a plant specific profile,. provided that expected pressure and humidity conditions as a function of time are accounted for;

.. In general, the containment temperature and pressure ~onditions as a function of time should be based on analyses in the FSAR.

However, these conditions should bound thatexpected for coolant and steam line breaks inside the containment. The steam line break conditions should include superheated conditions: the peak temperature, and subsequent temperature/pressure profile as a function of time.. If containment spray is to be used, the impact of the spray on required equipment sho.uld be accounted for.

Should equipment or systems which are presently planned to be. modified as a result of actions, such as lessons learned, to be. included in the response to IEB 79-018?

No.

Will there be any other actions required when the NRC completes its evaluation of the responses to IEB 79-:018 and any related corrective-actions deemed necessary are complete?

The NRC* staff does not foresee any additional actions for the electrical equipment and components which are included in the scope of IEB 79-018; however, if new problems or concerns are iden_ti fi ed, *appropriate action wi 11 be tak.en.

Are Spare Parts required to meet 79-018?

Yes. The Spare *Parts are required to meet the same criteria as the installed electrical_ equipment or component resulting from the evaluation of IEB 79-018.

The instruction sheet for Enclosure 3 "System Component Evaluation 11 indicates that outstanding items be identified.

What is the definition of outstanding it.ems?

Answer 8 Question 9 Answer 9 Ques.ti on 10 Answer 10 Question 11 Answer 11 Question 12 Answer 12 Question 13 Answer 13

. *~

~

Attachment to Supplemental Information to IE Bulletin No. 79-018 An outstanding item is defined as that item that does not

• meet.the environmental qua l i fi cat ion guide l in.es and requirements of IEB 79-01.

Are the requirements and positions in NUREG 0588 the same

.as those in NUREG 0578 in relation to environmental qualification of electrical equipment and components?

Yes.

NUREG 0588 is out for comment. Does the staff expect any significant changes which may impact this effort?

No.

When it is determin~d, as a result of the efforts required by IEB 79-018, that specific equipment be upgraded, are the guidelines in Enclosure 4 td be used?

As a minimum the same requirements that were used to determine the acceptability of the electrical equipment and components within the scope of IEB 79-0lB may be used; however, if equipment is available which meets the requi.rements of IEEE 323-1974 it should.be used.

Does the Licensee Event Report CLER) requirements of IEB 79-018 supercede or change the reporting requirements already defined?

No. The requirement for reporting in IEB 70-018 does not change the reporting requirements defined in the license*

conditions.

Are only those items known to be unqualified immediately reportable whereas items for which there is no data or insufficient data are open items to be resolved, but are not immediately reportable?

When a determination has been made that the existing data is inadequate or no data-exists to have reasonable assurance that. the Cl ass IE e 1 ectri ca 1 equi pni"ent components can perform their safety-related function required in the specified FSAR environments, that is reportable per IEB 79-018. The time and technical judgments required to make the determination should be based on the significance of the specific equipment, components and the discrepancies.

Question i4 Answer 14 Question 15 Answer 15 Question* 16 Answer 16

• Question 17 Answer 17

,\\

• e Attachment to Supplemental Information

• to IE Bulletin No. 79-018.

Are th~ results of an evaluation using the materials identified in Enclosure 4, Appendix C, of an acceptable ~ethod of.

• addressing the effects of aging within the scope of Bulletin 79-018?

Yes, for those materials on the list, however, Appendix C indicates this is a partial list. Your evalu~tion in response to IEB 79-018 may identify other materials that are susceptible

• to significant degradation.

What are the sources Appendix C-1 used to identify the materials in Table C-1 and establ'ish the failure levels?

Typical sources for the information are given in Appendix C.

Your information of materials not on this table should identify the source for your evaluation.

Is additional effort or calculations required for.radiation servi~e conditions if previous efforts did not utilize the methodology or assumptions identified in NUREG 0588?

Yes, the extent of the effort required will be dependent on the significance of the difference in methodology and assumptions.

Will extension of time be granted for schedules if identified in IEB 79-018 action item 7.

The schedule was based on the significance of the safety concerns relating to the adequacy of environmental qualification of electrical equipment or components. Any projected deviations from these schedules should b~ identified to the Regional Office by a written request. The NRC staff will make a determination on a case-by-case basis.

UNITED STATES NUC~;AR REGULATORY COMMISSION OFFICE OF* INSPECTION AND ENFORCEMENT

_WASHINGTON, D.C.

20555 September 30, 1980 SSINS:* 6820 A~cess,ion No. :

I~ Supplement No. 2 to Bulletin 79-olB*: ENVIRONMENTAL QUALIFICATION OF CLASS lE EQUIPMENT

• Enclosed are the generic questions and answers which resulted from NRC/Licensee meetings in NRC Regional Offices during the week of July 14, 1980 regarding environmental qualification.of Class.lE equipment in use at power reactor facilities. These answers address specific questions asked during the meetings.

Due to the generic nature of some of these questions, the staff is issuing'*

them as a bulletin supplement.

The regional meetings highlighted the fact that in some cases, the scope and depth of the 79-0lB. review was not clear to licensees. Therefore, these answers may affect your 79-018 submittal. These submittals are required by a separate order to be completed by November 1, 1980.

Some answers given in Supplement No. 1 to IEB-79-018 are superseded by these answers.

For example, in Bulletin Supplement No. 1, issued on February 29, 1980, the answer to question No., 5 specified that TMI lessons learned equipment was not included in the review.

However, due to the extension of the response date from April 14, 1980 to November 1, 1980, this equipment is now being addressed since its installation is either complete or required before the issuance of the February 1, 1981 SER.* (See Question No. 21 of this Supplement.)

No specific response is requested by this Supplement; however, all answers contained in the enclosure to this Supplement should be carefully reviewed and considered for applicability in your response to IEB 79-0lB.

IE Bulletin No. 79-018 was issued under a blanket GAO clearance (B180225 (R0072); clearance expired July 31, 1980) specifically for identified generic problems.

Supplement No. 2 to Bulletin 79-018 is for information, hence no GAO clearance* is required.

Attachment *.

1. Generic Questions and Answers to IEB-79-0lB and Memorandum and Order (CLI-80-21) dated May 23, 1980

GENERIC QUESTIONS AND ANSWERS TO !EB 79-018 AND MEMORANDUM AND ORDER (CLI-80-21) DATED MAY 23, 1980 Q.1 Define the scope of review with respect to the June 1982 deadline.

What is required beyond the June 1982 date for qualification?

A.1 By June 30, 1982, all safety-related electrical equipment potentially exposed to a harsh environment in nuclear generating stations, licensed to operate on or before June 30, 1982, shall be qualified to either the DOR guidelines or NUREG-0588 (as applicable).

Safety-related electrical equipment are those required in bringing the plant to a cold shutdown condition and to mitigate the consequences of the accident.

The qualification of safety-related electrical equipment to function in environmental extremes, not associated with accident conditions, is the responsibility of the licensee to evaluate and document in a form that will be available for the NRC to audit.

Qualification to assure functioning in mild environments must be completed by June 30, 1982.

The qualification schedules for consideration of the dynamic loading of safety-related equipment (electrical and mechanical) and the environmental qualification review of mechanical equipment are being developed.

It is the intention of the staff to initiate this effort as soon as possible.

Q.2 Clarify the required submittal dates for ORs, NTOLs, and CPs.

What about Ols whose 100% license is not expected by June 1982?

A.2 The required schedule for submitting information in response to the Commission Order and Memorandum (CLI-80-21) is provided b.elow.

• Plants who have received an operating license, either for full or limited power operation, are required to meet the schedule for operating reactors.

Plants who have committed, to the NRC, to meet schedules in advance of those provided below are required to meet that commitment.

In all cases, plants are required to have their equipment fully qualified to the applicable standards either by June 30, 1982, or by the time the operating license is granted, whichever comes later.

Operating Reactors and NTOL (operating license expected by February 1, 1981)

Submittal to be received no later than November 1, 1980 Ols (operating license expected by June 30, 1982)

Submittal to be r~ceived no later than 4 months prior to issuance of operating license.

Ols and CPs (operating license expected after June 30, 1982)

Submittal to be received no later than 6 months prior to issuance of operating license.

• Q.3 Define the requirements and applicable criteria for ORs, NTOLs,- and Ols.

Specifically address the NTOLs whose CP SER is prior to July 1974 and after July 1974.

Can a CP whose SER is prior to 1974 use the DOR guidelines?

A.3 Table 1 describes the application of each document.

All operating reactors as of,May 23, 1980, will be evaluated against the DOR guidelines.

In cases where the DOR guidelines do not provide sufficient detai_l, but NUREG-0588 Category II does~ NUREG~0588 will be used.

ORs TABLE 1 REQUIREMENTS Ols CPs DOR GUIDELINES USE NUREG-0588 AS NECESSARY REPLACEMENT COMPONENTS USE NUREG-0588 (CAL I}

CP SER CP SER Before 7/1/74 After 7/1/74 NUREG-0588(CAT.II)

NUREG-0588(CAT.I)

NUREG-0588(CAT.I) or

.NEW RULE WHEN IN EFFECT All plants licensed after May 23, 1980, shall conform to NUREG-0588.

In accordance with Regulatory Gui de l. 89, a 11 such operating 1 i censes for facilities whose construction permit SER is dated July 1, 1974 or later, are to be reviewed against IEEE Std. 323-1974.

Thus, for these licensees, the operating license applicant is to qualify equipment to the Category I. column in NUREG-0588.

For op~rating licenses issued after May 23, 1980, whose construction permit SER is dated before July 1, 1974, the operating license applfcant is to qualify equipment to.at least Category II column of NUR.EG-0588; unless the licensee made commitment in the construction permit record to use the 1974 standard, or unless the operating licensee applica-tion record indicates -that the 1974 standard is to be used, in such cases Column I of NUREG-0588 is to be used.

While there are differences between the Category II column of NUREG-0588 and the DOR guidelines, the differences are in details and in the optional part of the documents.

The minimum requirements set forth by these documents are general and compatible.

Thus, the minimum standards set by either o*f the two documents are equally applicable to ORs and NTOLs.

Q.4 Clarify the reporting requirements for LERs with respect to Part 50.55e vs79-018.

Are only those items, known to be unqualified, immediately reportable?

Are items, for.which there are no data or for which there are insuf-ficient data, open items to be resolved, but are not immediately reportable?

A.4 The requirement for reporting in !EB 79-018 does not change the reporting requirements defined in the license conditions.

In general, CPs should report via 50.SS(e).

Operating plants should use the LER.

When a determination has been made that reasonable assurance does not exist to ensu~ that the Class IE electrical equipment component(s) can perform their safety-related function, that is reportable.

Inadequate or no data are factors in this determination.

The time and technical judgements required to make the determination should be based on the si.gnificance of this specific equipment, components, and the discrepancies.

Q.5 How does the 11Q 11 list review interface with the EQB effort? Can the NRC provide more specific guidance on how to pick out the required safety-related equipment?

A.5 Tiie 11Q11 lfst provides a source from which the required equipment may be selected.

The information required to be submitted by November 1, 1980, is for safety-related electrical equipment potsntially exposed to a harsh environment resulting from an accident.

Safety-related equipment are those required to help bring the plant to cold shu.tdown and to mitigate the accident (LOCA, HELB inside or outside containment).

11Mitigate11 includes safety-related functions such as containment isolation, and prevention of significant release of radioactive material.

In order to "pick out 11 the safety-related equipment, the licensee should generate a list of safety functions typically performed by plant safety systems.

Examples are listed in Table II.

For each safety function identified in Table II, list the systems, subsystems, or components assumed available in the plant FSAR or emergency procedures to perform that function during a LOCA or any HELB inside or outsid~ containment.

If a plant specific safety function not listed in Table II is identified, that function and the corresponding systems or equipment to perform the function should be added to the 1icensee 1 s 1 i st.

The systems and equipment identified above should be included regardless of the original classification when the plant received its operating license; i.e., some control grade equipment will probably be named in emergency procedures.

However, if plant emergency procedures specify a preferred mode of accident mitigation involving equipment recognized by the licensee as unlikely to meet environmental qualification criteria, an alternate mode of performing the safety function and qualifiable equipment may be identified.

In such cases, the emergency procedures must clearly indicate how the operator is to use environmentally qualified safety-related display instrumentation to diagnose failure to perform such safety functions.

Plant emergency procedures typically include provisions for the operator to sample or monitor radioactivity levels or combustible gas levels, to confirm that valves are in the correct position, to monitor flow or temperature, etc.

Some of these functions are essential for correct operator action, to mitigate accidents, and prevent radioactive releases.

When this is the case, the radiation sensors, valve position indicators, pressure transmitters, thermo-couples, etc., should be qualified to function in the relevant accident environment.

Licensees should, therefore, review their emergency procedures to determine the electri.cal components needed to. perform the functions of Safety-Related Display Information, Post Accident Sampling and Monitoring, and Radiation Monitoring.

When equipment impli.ed by the emergency procedures is not listed, justificiation must be provided

• that failure of such equipment would not prevent accident mitigation

. or release of radioactivity.

Equipment now indicated in emergency procedures in response to TMI-2 Lessons* Learned shou'd be listed.

Equipment which is or will be installed due to TMI Lessons Learned should be addressed similar to other existing safety-related equipment (e.g., saturation meter, sump level indicators, torus water volume, etc.).

The licensee should document anticipated service conditions in every portion of the plant where the environment could be influenced by the accident or its consequences.

These service conditions should also be correlated with the safety-related systems and subsystems identified above.

Whenever an item of safety-related equipment may*

be located in an environment outside the range of normal conditions, due to the harsh environment resulting from the accident, and the equipment is needed to mitigate the consequences of the accident, place it on the Jist of equipment in a potentially hostile environ-ment.

Conclusions which show that equipment is unqualified should include a basis for continued plant operation.

TABLE II TYPICAL EQUIPMENT/FUNCTIONS NEEDED FOR MITIGATION OF A LOCA OR MSLB ACCIDENT Engineered Safeguards Actuation Reactor Protection Containment Isolation Steamline Isolation Main Feedwater Shutdown and Isolation Emergency Power

.. Emergency Core Cooling Containment Heat Removal Containment Fission Product Removal Containment.Combustible Gas Control Auxiliary Feedwater Containment Ventilation Containment Radiation Monitoring Control Room Habitability Systems (e.g., HVAC, Radiation Filters)

Ventilation for Areas Containing Safety Equipment Component Cooling Service Water Emergency Shutdown Post Accident Sampling and Monitoring Radiation Monitoring Safety Related Display Instrumentation (l) These systems will differ for PWRs and BWRs and for older and.newer plants.

In each case, the system features which allow for transfer to recirculation cooling mode and establishment of long-term cooling with boron precipitation control are to be considered as part of the system to be evaluated.

(2) Emergency shutdown systems include those systems used to bring the plant to a cold shutdown condition following accidents which do not result in a breach of the reactor coolant pressure boundary *fogether with a rapid depressurization of the reactor coolant system.

Examples of such systems and equipment are the RHR system, PORVs, RCIC, pressurizer sprays, chemical and volume control systam, and steam dump systems.

(3) More specific identification of these types of equipment can be found in the plant emergency proc~dures.

Q.6 NUREG-0588 was issued for comment.

Will any changes impact the requirements established by the Commission memorandum and order?

Will the daughter standards referenced be corrected/changed?

A.6 The requirement established by the Commission memorandum and order will not change as a result of comments on NUREG-0588.

No substan-tive changes are anticipated in NUREG-0588 or in referenced daughter standards.

A revision is anticipated, making corrections.

Q.7 Can IEEE Std. 650(Standards for Qualification of Class IE static battery chargers and inverters for nuclear power generating stations) be used for qualifying the balance of plant components which are not exposed to harsh environments?

A.7 The methods and procedures relating to design stress analysis, aging of electrical/electronic components and the stress test identified in this standard are acceptable for qualifying the balance of plant components which are not exposed to harsh environments.

-.5 -

Q.8 Provide the staff 1s definition of 11central lacation 11 for qualifica-tion documentation.

What documentation is expected to be maintained?

Will it be acceptable to maintain summary test *reports at the utility central file and provide a reference to the NSSS Vendor 1s file for the actual test reports? Does NRC require test reports to be sub-mitted to support qualification?

A.8 The central location should be at the utilities corporate head-quarters or plant site. Both the DOR guidelines and NUREG-0588 specify that sufficient information must be available to verify that the safety~related electrical equipment has been qualified in Q.9 A.9 Q *. 10 A.10 Q.11

• accordance with the guidance and requirements.

Details for the information and documentation required fo~ type tests, operating experience, analysis, and extrapolation of test data from operating experience are provided in Section 5 of NUREG~0588 and Section 8 of IEEE Std. 323-74.

The staff will accept summary test reports maintained at the utility 1s central file which reference the actual test reports and data available in. a single location at the NSSS.vendor* s faci 1 ity.

• The Licensee/ Applicant must make the determination that necessary information and documentation, to support qualification of equipment, i*s in conformance with DOR guidelines and NUREG-0588.

This vendor

• information file must be maintained current, auditable and available throughout the life of the referencing plant.

Test reports are not required to be submitted.

Test report references.

must be included in the plant submittals and these reports must be available for staff review on demand.

The staff was directed* to codify, by Technical Specification, some of the requirements of the Order.

Can you give some of the details of this requirement, how the staff expects to meet this directive and when?

The staff has proposed to *the Commission changes to the Technical Specifications (e.g., Appendix A Section 6.10 of the license) which require the establishment and maintenance of a centrally located file which will contain the information necessary to verify the qualification adequacy of all safety-related electrical equipment.

With respect to the NRC data base, how will utilities address and

• obtain information from it?

The industry access method for the data base. will be addressed in the final stages of system development.

This information should be available by mid-1981.

Licensees will be informed at that time.

How should submittals containing data and qualification information be submitted? What format* should we use if we have several facili-ties at different stages (OR, NTOL, CP)?

A.11 The qualification information and data shou.ld be submitted with the appropriate officer's notarized sworn statements.

The format for the data should be in accordance with the format provided in I&E Bulletin 79-0lB or the letters provided to the plants in the SEP

• program.

Either format is acceptable.

Q.12 Is testing required of equipment which completes its safety-related function within the first minute(s) of a LOCA or HELB?

(E.g.,

nuclear instrumentation or other instruments providing RPS inputs, isolation valves, etc.)

A.12

• The staff does not require that the nuclear instrurrn:!ntation and its associated components be environmentally qualified for a LOCA or HELB.

The nuclear instrumentation system is used for transient conditions but is not required for a LOCA or HELB.

The staff does requi~ that equipment designed to perform its safety-

~lated function within a short time into an event be qualified for a period of at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in excess of the time assumed in the accident analysis.

The staff has indicated that time is the most significant factor in terms of the margins required to provide an acceptable confidence level that a safety-related function will be completed.

Our judgment of at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is based on the acceptance of a type test for a single unit ana the spectrum of accidents (small and large breaks) bounded by the single test.

Also see answer to question 21.

Q.13 Testing is currently being performed a.n some equipment, and contracts have been issued for testing additional equipment specifying confor-mance to IEEE Std 323-1971.

For sequential testing, how do we factor in aging? If earl.Y test failure occurs due to 11 non E-Q" mechanisms, can the test be extrapolated using analytical methods?

A.13 Sequential testing requirements are specified in NUREG-0588 and the DOR guidelines.

Licensees must follow the test requirements of the applicable document.

1.

If the test has been completed without aging in sequence, justiftcation for such a deviation must be submitted.

2.

If testing of a given component has been scheduled but not initiated, the test sequence/program should be modified to include aging.

3.

Test programs in progress should be evaluated regarding the ability to comply by incorporating aging in the proper sequence.

These would then fall in the first or second category.

When a failure occurs due to a non-EQ related mechanism, acceptability of analysis to extrapolate the test data would be dependent on several considerations (e.g., the specific function being demonstrated, the

• fai 1 ure mechanism, when the *tai 1 ure occurred, etc.), may be very difficult to achieve.

If such a failure occurs it may be more prudent to correct the failure and continue with the test.

Q.14 What is the definition of harsh environment?

How are the environ-mental profiles* defined outside containment?

A.14

.Harsh environment is defined by the limiting conditions, as specified in IE Bulletin 79-018, resulting from the entire spectrum of LOCAs HELBs.

Specifically, the harsh environment from a LOCA considers the worst parameters resulting over the spectrum of postulated break sizes, break locations and single failures.

Similarly, the HELBs inside and outside of containment consider the spectrum of breaks including main steam and feedwater line breaks.

The parameters to be considered are:

temperature, pressure, humidity, caustic spray, radiation, duration of exposure, aging and submergence. *Mechanical and flow-induced vibrations and seismic effects will be considered separ_ately.

Environmental profiles for HELB outside of containment have not been generically established due to the uniqueness of each facility.

Service conditions for areas outside containment exposed to a HELB must be evaluated on a plant-by-plant basis.

Each of the parameters li.sted above must be considered.

Acceptable engineering methods should be used for this calculation.

Temperature and pressure history may be available from earlier HELB evalations.

The radiation source terms are discussed under Question 18 below.

Further guidance for selecting the piping systems and conducting the review are delineated in Regulatory Guide 1.46 and Standard Review Plans 3.6.l and 3.6.2.

Q.15 The DOR Guidelines and NUREG-0588 give time and temperature parameters.

Can we use different values of these parameters? Will plant-specific profiles still be with the guidance provided?

Q.15 For minimum high temperature conditions in pressure-suppression-type containments, we do not require that 340 F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> be used for BWR drywells or that 340 F for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> be used for PWR ice condenser lower compartments.

These values are a screening device, per the Guidelines, and can be used in lieu of a plant-specific profile, provided that expected pressure and humidity conditions as a function of time are accounted for.

In general; the containment temperature and pressure conditions as a function of time should be based on analyses in the FSAR.

However, these conditions should bound thoseexpected for coolant and steam line breaks inside the containment with due consideration of analytical uncertainties.

The steam line break condition should include superheated conditions:

the peak temperature, and subsequent temperature/ pressure profile as a function of time.

If containment spray is to be used, the impact of the spray on required equipment should be accounted for.

• The adequacy of a plant-specific profile is dependent on the assump-tion and design consi~erations at the time the profiles were developed.

The DOR guidelines and NUREG-0588 provide guidance and c~nsiderations required to determine if the plant-specific profiles encompass the LOCA and HELB inside containment.

Q~16 Could you elaborate on what the staff expects with regard to quality assurance?

If parts or subcomponents are purchased from a vendor who does not have a quality assurance program, can it be qualified to meet IEEE Std. 323-74 requirements?

A.16 The QA programs should accommodate any increased scope due to the new environmental qualification documentation requirements.

Proce-dures incorporated by the licensee for data acquisition should be documented and available for staff review upon request.

Requirements for QA programs*are provided in Part 50, Appendix 8, of the Code of Federal Regulations.

Part 50, *Appendix B of the Code of Federal Regulations states that the applicant/licensee shall be responsible for the establishment and execution of quality assurance programs.

Specifically in purchasing parts or components, it is the responsibility of the licensee/applicant to ensure that the applicable quality assurance procedures for their plant are met.

In determining the qualification status of existing equipment purchased from a vendor, where a QA program did not exist, the utility should consider the following:

1.

The complexity of design, complexity of manufacturing process, and end use.

2.

Past performance of vendor.

3.

Past operating history of products, especially similar products, made by_ vendor.

4.

Procedures, equipment, and results of environmental qualifica-tion testing relative to those for other equipment for which a QA program was applied.

Q.17 Define the requirements for "replacement parts." Are they the same for "spare" parts? Clearly discuss the alternatives for existing inventories of parts/components.

If equipment is ordered to meet IEEE Std. 323-1974 standard but lead time exceeds June 1982, can we use IEEE Std. 323-1971 qualified components in the interim?

A.17 The requirements for 11replacement 11 and 11 spare 11 parts are the same for the purposes of complying with the Commission or.der and

10 -

memorandum.. After May 1980> all parts used to replace presently i nsta 11 ed parts sha 11 be qualified to Category I of NUREG-0588 11unless there are sound reasons to the contrary. 11 Nonavailability and/or the fact that the part to be used as a replacement is a spare part purchased prior to May 23, 1980> and is in stock are among the factors to be considered in weighing whether there are 11sound reasons to the contrary.

11 All replacement parts shall as a minimum conform to the requirements described in the answer to question 3.

Justifica-tion for dev.iation from Category I or NUREG~0588 shall be documented by the licensee and records shall be available for audit,. upon request by the NRC.

Q.18 DOR Guid~lines, NUREG-0588 and NURE~-0578, define or give guidance for calculating radiation source terms.

However, since one is more restrictive than the other, which do we use?

A.18 Both the DOR guidelines and NUREG-0588 are similar in that they provide the methods for determining the radiation source term when considering LOCA events inside containment (100%.noble gases/50%

iodine/1% particulates). These methods consider the radiation source term resulting from an event which completely depressurizes the primary sys~m and releases the source term inventory to the containment.

NUREG-0578 provides the radiation source term to be used for deter-mining the qualification doses for equipment in close proximity tD recirculting fluid systems inside and outside of containment as a result of LOCA.

This method considers a LOCA event in which the primary system may not depressurize and the source term inventory remains in the coolant.

NUREG-0588 also provides the radiation source term to be used for qualifying equipment following non-LOCA *events both inside and outside containment (10% noble gases/10% iodine/0% particulates).

When developing radiation source terms for equipment qualification, the licensee must ensure consideration is given to those events which provide the most bounding conditions.

The following table**

summarizes these considerations:

Outside Containment LOCA NUREG-0578 (100/50/1 in RCS)

NON-LOCA HELB NUREG-0588 (10/10/0 in RCS)

Inside Containment Larger of NUREG-0588 (100/50/l in containment) or NUREG-0578 (100/50/1 in RCS)

NUREG-0588 (10/10/0 in RCS)

Q.19 Can gamma equivalents be used rather than beta exposure for radiation qualification?

A.19 Yes.

Gamma equivalents may be used when consideration of the contri-butions of beta exposure have been included in accordance with the guidance given in the DOR guidelines and NUREG-0588.

Cobalt 60 is one acceptable gamma radiation source for environmental qualification of safety-related equipment.

Cesium 137 may also be used.

Q.20 If a piece of equipment will become submerged after completing its required action, must it be qualified for submergence?

A.20 If the equipment (1) meets the guiadance and requirements of the DOR guidelines or NUREG-0588 for the LOCA and HELB (small and large breaks) accidents and (2) licensees demonstrate that its failure will not adversely affect any safety-related function or mislead the operator after submergence, the equipment could be considered exempt from that portion (submergence) of qualification.

Q.21 What qualification is required of Reactor Pressure Vessel internal*

instrumentation (e.g., thermocouples) and new instruments required as the result of TMI Lessons Learned?

A.21 TMI Lessons Learned instrumentation will be considered in the February l, 1981 SER.

This equipment is subject to the same require-ments as other safety-related electrical equipment.

The guidance and requirements of NUREG-0588 referenced daughter standards, and Reg Guides will be used by the staff in assessing the adequacy of the qualification information.

The in-core environment should consider the worst source term for radiation effects, the worst humidity for the corresponding temperature, and high temperatures consistent with that of a damaged core.

Q.22 Is qualification 11by use 11 an acceptable method (e.g., CRDM's in BWRs)?

A.22 Qualification by use has limited application.

Often the equipment has never seen the harsh environment and no conclusions can be drawn as to its operability in a harsh environment.

Some -qualification

- 12. -

based on operating experience is a recognized method subject to the requirements of NUREG-0588 and the Guidelines.

Credit can be taken for the natural aging of the equipment and far the location of the

• equipment or other portions of the overall qualification information.

Q.23 How long should "long term" equipment be qualified for environmental qualification?

A.23 "Long term" for the purpose of qualifying equipment for a harsh environment is variable.. A determination of 11 long term 11 for qualifi-cation of equipment should be based on the considerations listed below far each postulated accident scenario. Justification for the value used should be provided with the equipment qualification documentation.

1.

The time period over which the equipment fs required to bring the plant to cold shutdown and to mitigate the consequences of the accident.

2.

The ability to change, modify or add equipment during the course of the accident or in mitigating its effects which will provide the same safety-related function.

Q.24 Why do we want component surface temperature rather than the bulk environment temperature?

A.24 Temperature measurements are raquired during the qualification testing to establish that the component was subjected to the most severe temperature environment postulated to occur.

These temperature measurements are required to be made as close to the component surface as practicable to ensure that they are representative of the environment in which the component is tested.

The surface temperature of the component, although not specifically required, is considered to be a conservative measurement of the test temperature environment.

IE Bulletin No. 79-018 Supplement No. 3 October 24, 1980 Bulletin No.

Subject RECENTLY ISSUED IE BULLETINS Date Issued Supplement 2

• Environmental Qualifica-9/30/80 to 79-018 tion of Class lE 80-22 80-21 Revision 1 to 79-26 Revi siori 1 to 80-19 80-20 80-19 80-18 Supplement 3 to 80-17 Supplement 2 to 80-17

_Equipment Automation Industries, Model 200-520-008 (Not Used)*

9/12/80 Boron Loss from 8/29/80 BWR Control 6lades Failures of Mercury-8/15/80 Wetted Matrix Relays in Reactor Protective Systems of Operating Nuclear Power Plants Designed by Combustion Engineering Failures of Westinghouse 7/31/80 Type W-2 Spring Return to Neutral Control Switches Fai.1 ures of Mercury-7 /31/80 Wetted Mat r.i ~ Re 1 ays i n Reactor Protective Systems of Operating Nuclear Power Plants Designed by Combustion Engineering Maintenance of Adequate 7/24/80 Minimum Flow Thru Centrifugal Charging Pumps Following Secondary Side High Energy Line Rupture Failure of Control Rods 8/22/80 to Insert During a

• scram at a BWR failure o~ Control R6ds 7/22/BO to Insert During a Scram at a BWR Issued To All holders of a power reactor OL or CP All holders of a radiography license All holders of a BWR power reactor OL All holders.of a power reactor OL or CP All holders of a power reactor OL or CP All holders of a power reactor OL or CP All holders of a PWR power reactor OL or CP All holders of a BWR power reactor OL or CP All holders of a BWR power reactor OL