ML20094N414
| ML20094N414 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 02/19/1992 |
| From: | BECHTEL CORP. |
| To: | |
| References | |
| CIVP-S-043, CIVP-S-43, NUDOCS 9204060344 | |
| Download: ML20094N414 (14) | |
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. . x Th'E NOTES CONTAINED HEREIN ARE TO BE USED FOR UNIT FI AND UNI T
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TABLE OF COMTENTS FOR TRAY & CONDUIT
)
< DETAILS & NOTES Sheets.
Subieet
- i ..
lA ;
"1) , Current revisions of individual sheets. .
2), Cable. Trays, Supports and their included cables. 2, 2A, 3, 3A', 3B, 3C, 3I
- 3) Unistrut cable support' systems. *!
- 4) Conduits, - Rigid galvanized ~ steel,' alum'inum, flexible 4, 4A, 5 steel, associated fittings.. .A
- 5) class "IE" conduit support. . SA, SB
- 6) . Terminal Boxes (Painting & Monitoring). . 6 7
- 7) Non-Metallic conduit.
7A ,
- 8) Cabh channel. 3
- 9) Raceway marki.ng standards (paint, etc.). 7B, 7C, 7D -
- 10) Synbols & their definiticns. 8,, 9, 10
- 11) Detail for conduit entering tray, 10A
! 12) Details for fire-s* ops & definition. 15, 16, 17, 18, 20, 21 l J 12A) Details for fire seals / air seals and watertight seals 18A, 18T, 180, 18D
& definitions. g 16E, 18F 12B) Details for fire-seals & air seals for control room.
- 13) Watertight seals & internal conduit sealing. 19, 10A, 19B l 5, '
22 sO 14) Flexible conduit termination' details. t$ 23 g "O 15) 208V motor - tyr Acal raceway installation. , l 2 gg Local control station - typical raceway installation 23A gw 16) ($ for 4160 & 600 V motors.
- 17) Typical. raceway instal'1ation for cooler fan motor. 23B g ,w i N 18) Typical raceway installation for motor operated valves. 23C 005604
- 2. pc
- 19) Typical raceway installation for sump pumps & control _ 23D 9 P*
g stations. SOUTHERN SERVICES UIC. BECHTEL CORP. JOB 7597 - 0 3 /2 0 ,, b..k.W...... ALABAMA POWER COMPANY O y". W d'bw"W. g ,",,gg,.g - i ?.ss REV. AS NOTED . ,y,3c,
,,y m JOSDH M. FARLEY NUCLEAR PLANT TRAY & CONDUlT DETAILS & NOTES
.m 3-!T-7 ~). ..u. . ....... .. .ex. . IA A . P a .... . A-177 541
i - , Sheets
; subject . ,
- 20) Typical raceway installation to electri:a1 devices 23E on scienoid operated pneumatic control valves.
- 21) Typical raceway installation at instrument racks or 23r where groups of instruments are lo'cated together at
. equipment, 24 j ).22) Power receptable installati'on on handrail.
I 25 23)
'Sgcondary shield wall penetrations f
inside containment. .
'd 26A 2;4), " Void .
26B
- 25) Typical installation cable tray along contairiment valt.
' 27 ,
- 26) Void. . .
27A
- 27) Void. .
- 28) Separationandrequiredbarriers-cableIspreadingroom.
27B 27C
- 29) Void..
- 30) Separation where redundant trays or channels crossover. 27D I '
27E
- 31) Void.
*
- 3 2') Minimum separation between redundant trays & channels 27F -
& conduits. .
- 33) . Typical conduit supports to equipment. 28
- 34) instrumentation cable tray connection box. 29
- 35) "C" train conduit installation. SC
$ +) Electrical equipment location criteria.
- 23G
{ 23J-1, g , ;, ll; ') Modifications to NAMCO Limit S moisgre_ incursion during LOCA,(for ._ es to prevent ,
-E.Q. equipment 9 23J-2, 23K, 23L, 23M, 23N 23P, 235, DM --
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- 38) Specification Data - Application Guide 180, 18H , 18J , 18K, E
Kaowool blanket wrap f or fire protection 18L, 18M, 18N, 18P, yp 18R, ISS, 18T, 18U, 2o ' . 18V , e . E
@ .g 39) Void 23005605 SOUTHERN SERVICES INC.
fE BECHTEL CORP. JOB 7597 03/20 ALABAMA POWER COMPANY Sa - MW-h u " c8j
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. 0 wt M 5005 TRAY ACONDUIT DETAIL S a' NOTES n y.o-1s, w ath u .7 an,.
lA B .,2 9 . ..,.' A-l 77541 q 1 l . n h..E-7 .u ....... .m.
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- 41) Typical Raceway Termination to Penetration Cabinets 2-23V c
l 005606 i I ORIG 5UAL DRAWING, Rev.__ _j TRANSFERRED TG SCS BECHTEL CORP. JOB 7597 03/20 SOUTHERN SERVICES INC.
" ALABAMA POWER COMPANY h h.)zissut [ kc"oast.
- 2. P/a.M REV.PEAKN'S B Etr JOSEPH M. FARLEY NUCLEAR PLANT "suwscr 1-%E5REV.O.) 4 4 s.sezsoos m oit JP,,uTRAY & CONDUlT- DETAll R 4 NOTES
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l RACEWAY NO'IES - POWER. C0ffrROL AND INSTRLHENTATION 1.0 CAB 1.E TRAYS 1.1 Cable trays for use in the auxiliary building and the turbine building shall be aluminum or hot dip galvanized per ND1/. Pub. No. VE-1. Cable trays for use in tha containment and outdoors chall be hot dip galvani:cd after fabrication g per NEMA Pub. No. VE 1. ' All trays have an ef fective depth of 3". Lov level ) instrumentation trays shall have solid corrugated bottom unless otherwise .
% %pecified.
s All other trays shall be laddet type with 9" rung spacing. 1.2 Tray width shall be a maximum of 24 inches'and a minimum of 6 inche). 1.3 The venical spacing of trays shall be sufficient to pertuit unobstructed pulling of all cables unless otherwise spccified. . 1.4 With stacked trays, rnedium voltage power cab 1(s shall be in the upper trays, low voltage power and control cables shell be '.n the intermediate trays, and instrumentation cables shall be in the icwpr +. cays. 1.5 Fittings , accessories, and h'ardware shall be of the'same materials and finish as the straight t.r a y s . Minimum radii of side' rails on horizontal elbows, . vertical risers, tees, and crosses shall'.be.24 inches except' as otherwise-indicated on drawings. S 1.6 Tray support spans shall not exceed 10 feet. NEMA support locations for y fittings shall be followed where practicable. . 1.7 Tray elevations shown on drawings shall be the bottom of the tray rail.
>s N *0 1.8 Wherever a tray crosses a building expansion joint, an expansion cont.ector fh shall be installed. Ground cable installed on the tray shall be looped to Eo allow for expansion.
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< st 1.9 Where changes in elevation will not permit standard. fittings to be used, adjust-hd
.a y able splice plates shall be insta11ec in the field as required. Where standard fittings are used, the degree of incline will be noted on the drawings.
2 4 3% 1,10 Trays running vertically through floors or grating shall have a removabic , g cover to 7 feet above the floor or grating to provide protection against' o damago to. cables. 1.11 At points where trays run vertically through floors, a concrete curb of 3 inch minimum height or a 3" x 3/8" steel toe plate (min.) ehall be installed to 005G07
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protect the trays from damage. 1.12 Where 20' separation criterion cannot be maintained per the Farley Fire Protectio t Re-evaluation Report, all. redundant trays / channels that contain safe shutdown cables shall be wrapped with KA0 WOOL as specified on D-180500 thru D-180560 and A-180561, Sheets 1 thru 49, for Unit '1. For Unit 2 see D-204556, 204565-570, - 587-590, 597, 598, 600-603, 630-635, 641-644, 648-653, 672, 686, 687, 689-691, , 698, 699, 706, 707, 716', 719, 733-740; 745, 746, 749;-750, 756-760, 764-768, 778-780, 732-794, 797-796, 204800. - BECHTEL CORP. J03 7597 03/ 20 SOUTHERN SERVICES INC. onM . "- , W ,n"hs
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n mlm lk,o,,mpp,{ ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT e . M [ [.I NS L REV. G WJiED sus.trer _8 ""' TRAY ( CONDUIT DETAILS ( NOTES
,., M,_ 9 2-N/ REV AS MOTE D m
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4 gm 1.12a Where 3 foot horizontal or 5 foot vertical separation of redundant traysf
+
channels cannot be ma'intained per Reg. Guide 1.75. barriers of " thick marinite shall be installed between th6 trays /channelo_grpoie_cti,ng one _.f ogt, b used Jgygn,d in,the,. ed,ge,p_f,_the most installations hQav%ite tray ,onI or Types both M aresidesg acceptMarinite _a_bley' Type XL w
- V bu)Eitutes.{ Rigid steel conduit may be .used in all future cases to afford separati 3 where 3 f t. horizontal & $ f t. vertical cannot be maintained. ,
1.13 , When. cabte trays run under grating or wherever they are exposed to the deposit of foreign materials, solid removable covers shall be installed
- in the topmost tray.
1.13A To avoid ~ damage to cable insulation when " spilling out" of a tray, cables
. will be protected by using tray sof teners as manufactured by C. C. Spralding & Company "Jif fy Trim" or equal, ,
1.13B Under no circumstances are cables to be supported by Ty-Raps; i.e., do not support cables from botton: of tray, tray supports or unistrut. 1.13C Cable tray dr,op-out fittings shall be used
- . 1) Where a cable or group of cables leaves a tray between the rungs
. when, in the judgment of the field engineers, prob.,ble vibration or differe'etal movement between the cable and the tray could cause caule damage, and
- 2) Where bundled groups of cables leave a tray between the rungs.
1.13D Where Ty-Raps are ustd to train and/or secure cables in trays or . b equipment inside containment; only. radiation resistant ties, such as
- b. T&B Type TYZ-25M,or stainless steel shall be used.
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5A a a w , 003608 BECHTEL CORP. JOB 7597 03 SOUTHERN SERVICES INC.
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1.14 A minimum spacing of one cable diameter shall be maintained in cable trays for medium voltage and large low voltagr power cables. 1.15 khere penetrationf are required in valls or slabs of water right compartments, cable transit assemblies as manufactured by Nelson Electric Company shall be installed to maintain the watertight integrity. In order to expedite Unit 2 completion, Field may use silicone foam and/or' concrete to maintain watertight integrity. 1.16 Dimensions shown to cable trays are to the outside edge of the tray rail. 1.17 All field joints shall be made with devices which will not diminish the strength of electrical ground continuity of .he tray assembly. 1.18 Ca.les in vertical riser trays shall be properly supported. Cable grips shall be used for supporting cables on vertical risers of more than 20' where the cable is not secured to the raceway system. 1.19 Any opening into the main control room and in floors of ceilings for vertical ventillated tray installation shall have fire stops in accordance with the details shown herein. 1.20 For all cases where deviations of greater than 3" occur from the tray installatia dw;;s. the field shall submit on FCR 1:o engineerin g for approval. For deviations less than 3"; it will be the fields responsibility to ensure th.st their is no interference with.pioing etc. since there could be an unfav-orable tolerance build-up,ductvork, between equipment, tiie discip lines. 1.21 Contents Deleted. 1.22 The unsupported horizontal distance for cables shall not exceed: For all I. A.C. cables - 48" (Provided cable is secured any distance within, but not to exceed 2'-0" from any box, equipment, device, etc.) , For cables 1/0 and 1.cger (not I.A.C.)-36" 9 For cables smaller than 1/O-24" 1.23 The unsupported vertical distance for cables shall not exceed:
\
Fower . cables with IAC - 6'-0" l $m 0 Control cables 4'-0" c: & instrumentation cables er 1.24 Instrumentation tray, by the definition of " totally enclosed" shall have a' Eo 1 cast 85% of the total surface area of the tray (computed circumferentially) l 3: y 4 cc completely covered. The intent here is that at no location should the cables
; It be complet g ex2osyduexc gt_where__they must pass claough a watertight pene-g $ y' 'a -
g erationgcpp ng out of coEits to ecble tmy)orwhere they run exposed from troy to
< p. penetration terminct boxes as rir detail 3, D-177698. At these points. the distence o(
,'j g$ exposed cable shallnot exceed two feet. Locations not covered by the B5% criteric cre:
'iE 1.24A Instrument cables running between control room e,gu,ip,m nt and cable spreading
'o- rp.pn_.tu ys as Sheet 3E cf A-1775
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{ , ! 1. 24B DKitrument cables for RPI system in the containment building' l '-* 1.24C Control and Instrumentatiors Cables of the some train, supported Qp%IAs in the free air shcli I sncil be grouped in sepcrate bundIes.Controlcnd Instrumentati
" be separated es for as possible.For Uryit 2 refer to-design dreni jsNuUS l -
- 12. 24D 1nstrurnentation trcys used for Neutron Flux Monitorino System
- SOUTHERN SERVICES INC.
, BECHJEL CORP JOB 7597 -o3/20
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'OSEPH M. FARLEY NUCLEAR PLANT TRAY f CONDUlT DETAILS XNOTES g/g is yps ,a a y 29 . cos A- l7'541 16 3
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O 97B ? ISSUED 7CONST ALABAMA POWER COMPANY ' h;"[,Q"7/ ~ suancy JOSEPH M. FARLEY NUCLEAR PLANT mm TRAY 4 CONDUlT DETAILS 4 NOTES APP. ., nr. //;u . u,u . . n . .uu . 3^ ., 2 9 .on. A-17754I O
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BECHTEL CORP. JOB 7597-03 . SOUTHERil SERVICES INC.
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. .'.I. .... . ~o 97 3 issu[o o consT
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, O 97-73 ISSUED FOR CCH5T
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MOTE'. THESE ARE MON-SEISMIC INTERMEDIATE 5 - SUPPORTS ONLY, . t BC':HTEL CORP. JOB 7597-03 SOUTHERN SERVICES'INC. e a .ar. .svi. - ALABAMA POWER - COMPANY-N$.gg'.3 ~$.]. $ O 9 7*73 ISSUED FOR CONST b( / . suesser JOSEPH-M. FARLEY NUCLEAR PLANT. m TRAY &CONDUlT DETAll8. NOTES hg .- A-177541 O a.m .. ff. . 9)AA. 2D _ ., 29
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gte ' sy ; 003614 i1,- w 3 E u i FLR.El WO" MOTES , I. TYPICAL UMISTRUT ARRAMGMEMT HUMG FROM TRAY SUPPORTS TO ALLOW CABLE SUPPORTIMG FROM CORTROL ROOM CABIMET5 TO CABLE SPREADIMG ROOM TR AYS AMD TERMIMATIOM CABIMET5 Z. SEE SUPPORT PLAM DW4 D-lT] )30 4 TRAY PLAU DW6. D-lTIM4. dTHIS DETAIL TO BE. USE.D FOR CouTROL C.ABLEco ONLY \M UHtT" U\lSTRUT CABLESUPPORT SYSTEMS (TYP) BECHTEL CORP. JOB 7597-03 SOUTHERN SERVICES INC. ALABAMA POWER COMPANY O /t fce STR Tiou RW& [,f. 5 7/w$ W. PEA Pt>J 5-810~ suanc7 JOSEPH M. FARLEY NUCLEAR PLANT m
/ iose w o-so#l2 TRAY & CONDUlT DETAILS 4 NOTES
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' 2 [0 CONDUITS 1
e 2 > 1' Rigid conduits shall be aluminum or galvanized steel ot standard wall thickness (U L approved). Embedded condt.its shall be galvanized steel and exposed con-duits shall be. aluminum. All coaduits inside the containment shall be gal-vanized steel. ' 2.2 Flexibic steel conduits shall be galvanized with an outer liquid-tight i jacket. Fittings shall be of the water-tight compression type. 4 2.3 Fittings sha'11 be of the same naterial and finish as their r '
- conduits, however, malleable iron couplings may be used to couple aluminum conduit to rigid galvanized steel conduit in the following areas or ene Aux.
- i Bldg.:
- 1. Control Room
- 2. Cable Spreading Roor
- 3. Switchgest Rooms
- 4. CRDM Room
- 5. Computer Room
- 6. Elect. Penetratien Room
. 7. Cable Chases
- 8. Laboratories.
- 9. Balance of Plant except: -
- a. . Circulating Water Structures
- b. Water Treatment Plant 005615 .
. c. Cooling Towers *
- d. Cooling Tower Switchgear B1dgs.
- e. Pump Roost, of River Water. Structure.
- f. Fump Rooms of Service Water Structure
- g. Chlorine Structures . .
ma3*e"aNe Connectkons toen2*f'ikmotor tetk' h'al*Ycxes,ta'11Ed." iYs" 2.4 instatuments, or other equipment sbbject'to vi'.catinti or removal shall be made with liquid-tight . , flexible steel conduit, unless otherwise specified. 2.5 Whcie applicable, weep holes shall be- provided at the low point of embedded conduit runs to avoid accumulation of moisture. 0-2.6 A minimum of 4" spacing is required between embedded conduitshh minimum size of embedded conduits shall be 1"(unless noted otherwise. ll 4g . i$ 2.7 All conduits to be embedded in coacrete shall be inspected to assure , i O >- j continuity and correct position before concrete is pouted, and shall be ter-
, minated with couplings flush with the finished floor line, except as noted.
<Ey 2.8 All cabcdded conduit endo shall be plugged or capped during construction. ~
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*9 .
Ends of a, pare conduits chall. be kept. capped or plugged. When a conduit crosses a building expansion joint, a conduit expansion coupling or a liquidtig,ht flexible metal conduit shall be installed to allow [ l% g for expansion. Flexible conduits are.not.to exceed 24"'ii. length.=A flexible < g bonding juupar shall be used with each expansion joint or flexible conduit. ' D 2.9A Redundant duJrd. conduit shall be installed to maintain a minimum of 1" y , separation.
-BECHTEL CORP. -JOB 7597-03/ 2 O SOUTHERN SERVICES INC.
- n
.. //A. ALABAMA POWER COMPANY n . .... .. ..p. .
.3.'SLt s Rev. per PCN B-8A
- 2 2156 nev o-25 l, d JOSEPH M. FARLEY NUCLEAR PLANT- ,
cc . ..e L-p.- n 6.s Rev per comments u.susd;cr TRAY ( CONDUIT ~ DETA!LS 4 NOTES
/ DETAlt.
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l i !.+ . 1 4 2.10 Standard conduits for power circuits shall be connected to ground at one
, point in a continuous run by means of a grounding bushing.
2.11' A'11 threads on exposed and embedded conduit shall be painted with zine ; i rich paint before fittings are attached. All conduits cut in the field l shall be reamed and free of burrs before installation. i ' 2.12 Condui:s shall be supported in accordance with the National Electrical j Code Table 346-12 unless otherwise specified. , ;. 2.13 When conduit penetrations are made through secondary walls inside the
. containment, the conduit shall be installed at such an angle that the j
bottom of the conduit I.D. on one side of the shield shall be at least j 3 inches above the top of the conduit I.D. on the opposite side of the shield. a 2.13A The type of flexible metal conduit to be used in the Containment Building is Servicair Company "Serviflex" Hose
- Catalog No. SS63 **-S or equal:
** - Indicates size as follows:
06 - 3/4" i 08 - 1" 10 1/4" 12 1/2" 16 - 2" , l If the' stainless steel "Serviflex" is used, malleable fittings are ' l
. permitted, provided they a're zine coated. ;
Also acceptable for use in the Containment is Greenfield Electro-Plated Galvanized Flexible Conduit and Fittings. t i I I 005G1G*
~~ -
ORIGIN AL. DRAWING. M*V - TRANSFERRED TO SCS , BECHTEL CORP. JOB 7597 03 SOUTHERN SERVICES INC.- w o, oat. ..vistow if
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o Wh4 Fas2 1 ste m ted ALABAMA POWER COMPANY
,RQp,. I d4h5 N PER BM64-1 susacy JOSEPH M. FARLEY NUCLEAR PLANT
~ TRAY / CONDUlT DETAILTNOTES u, omit
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2,14 Elevation of conduits in concrete slaba sball be determined by field forces to avoid drains, pipes, recesses, and large reinforcing bars. Conduit elbows sball be located so that the entire bend is below the face of the
, floor slab.
I 2.-15 Where a change of direction is required for conduits in concrete slabs, long radius bends shall be used where possible. 2.16 Steel conduits installed in underground runs shall be encased in concrete with 4 inch minimum coverege. 2.17 Terminate steel conduits in concrete pull boxes with busbings 2 inches ' beyond the inside face of the wall unless otherwise specified. 2.16 Routing of exposed conduits are shown diagramatically. Field to determine e, . exact route in the event of conflict with piping, pipo hangerai, ductwork, etc. Also, field to maintain a minimum of 1/2" between outside edges of instrument-ation 6 :entrol conduit. For conduit containing 4KV-Po,;er ciacuits a minimum of 1 conduit,disaeter spacing:f:mst be maintained. For conduit containing 600V-power circuits, spacing must be no less than 1/4 of largest conduit diameter. I 2.19 The elevation of exposed tonduits should be as close to building steel or concrete as is possible to maintain a straight run and/or a minimum of bends. 2.20 Conduit bangers and supports shall be a field detail. Use continudus slot n- framing (Power-strut, Unistrut, etc.) and conduit clamps as fu shed. 2.21 Conduit runs of aot less than 100 feet but having more than two 90 degree bends sna11 be provided with pull boxes after every second bend. Pull l boxes shall be provided ewry 150 feet + in straight conduit runs, l 2.22 For outdoor or below grade equipment not having threaded hubs, Myers f Electric Company "Scru-tite" fittings shall be~ installed. D-l 2.23 All conduits shall be sloped towcrd pull boxes, or manholes for drainage. % 2.24 sbedded conduits shall be shown diagrammatic 1y. Final Routing to bg - d.ccrmined in field, however stub eps as shown on dwgs, will be tirm. j 2.25 When supporting exposed corduit from walls or ceiling, .all conduit larger than 4" require welded studs. Conduit 4" and less mr.y bc supported with cinch anchors. All cinch anchors should be installed l per manufacturers' specifications. 005617 l 2.25 Field to add reducers oushings, etc. to make final connections - ! to equipment as required. b -
'1 ORIGIN A1. DRAWING, Rev-TRANSFERRED TO SCS __
SOUTHERN SERVICES INC. BECH_TEL CORP. JOB 7597 - 0 3
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*^* TRAY 4' CONDUIT DETAILS ( NOTES i
App. O I .. . . . . 7 9/f'//.1 AtEV PER BE-Sb ntM. . .N SUPER $EDES SCALE s% - or d SHEETS
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2.27 Unless otheivise sper:ified, when installing conduit $4PPort haagers
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005618 2.28 Cable Tray Seismic Supports may be utilize to support conduit subject to the following limitations ORIGINM DRAWING Rev. 3 TRANSFERRED TO SCS
~
BECHTEL CORP. JOB 7597-03/ 2 0 SOUTHERN SERVICES INC. 4 3 FOit A P AL ALABAMA POWER COMPANY REV #0 MOIED sus.rse r JOSEPH M. FARLE? NUCLEAR PLANT e dWA [ S 434I@ % moJwkM3 ?LuTRAY & CONDUlT DETAILS a. NOTES Ap
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Auxiliary Building ,'
. .. . l w Conduit Max. Number of Conduits to.be_
Jupported-Adjacent to'Each Cable Tray
. 3", 3 'I, 4" 2 2" & 2\" '3 1\" 4 1" 6 3/4" 12 o.
Containment o 2\" to 4" . 1
'l " & 2" 2 .
1" 5 3/4" 9 _ ,_ _ . -~- ~.-~......~..,..._ . . _...... -. 2*29 d Tray supported conduits require intermediate supports if the maximum ._ h +
/'qcnduit span, as specified on B-203554 -is exceeded. ,,/)
~~
I ,* 2.30 PrYv~iIe~slipp o'rI'sa~t"Tolh7 id e s'oTIf1'~cEiiddi't" bend sT *a s'Ei~o's e to bend as practicable.
;0TEJ
. 1. Welding of brackets, plates, supports, etc. to the containment liner shall be done only at embedded plates ("A , P2, P3), hanger connection plates (HPI), beam support brackets (F1, 72), and scaffold lugs.
Welding directly to 1/4" liner plate is prohibited, uni ss specifically authorized in writing by the engineer.
- 2. Drilled-in anchor bolts shall not be used in the containment exterior wall without prior authorization of the engineer. When authorized.
drilling shall be by air percussion drill only; core drilling is prohibited. , -
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ORIGINAL DRAWING, Rev- 3 005619 TRANSFERRCJ TO SCS
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BECHTEL CORP. JOB 7597-03/ 2 0 SOUTHERN SERVICES INC.
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"..g u ~~~~~ o 3/Wu roe. co m ra x ,og ALABAMA POWER COMPANY es.& w &. [. 5 1 %7a3 REV.- AS uoTE.D ,u,,,c, JOSEPH M. FARLEY NUCLEAR PLANT a,
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- 1. "C" train circuits are chose which are capabic of performing a "B function or. an "A" function at any given time. i Toensuretheintegrityofthese"C".thaincircuits,noothercc.blesother than those presently reflect'ed in the EE553 progr. shall be pulled'thru "C" tra'.n condttits . .
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g% # I 4 l 005620 E ORIG 8h.'t 1RAWING, Rw. I W .'SFERRED TO SCS "C" TRAIN CONDUF \1S A _LATION BCCHTEL CORo JOB 7597 -0 3 / 2 0 SOUTHERN SERVICES INC. ALABAMA ' POWER COMPANY
, aaDDW-fh' 9,f3}7FOR 5bbST R.
( - 2//7Q[ l M 8/ REV'D AS MOTED - sumacr JOSEPH M. FARLEY NUCLEAR PLANT o,.ra _T RAY & CONDUlT DETAILS & NOTES
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V l Apr.
. 5 C ,,2 9 ... . A-177 54i l .. .fw. 22.. .u........ .m.
3.0 CABLE TRAY AND CONOU1T SUPPOP.T'AND FITTINGS , 3.1 Channel type macbers and fitting,s for the support of cable trays and conduits shall be green enamel steel, aluminum, or hot die galvanized steel, except for those used ins M e the containments which shall be hot dip galveaized after fabrication or plain steel coated per SS-1102-31. Plain steel coated fittings are also approved for use in the. Auxiliary Building. 3.2 Screws, nuts and other parts shall be of the same material and finish as support . 4.0 TERMINAL BOXES a 4.1 In dry locations, other than the containment, pull boxes and outlet boxes
'shall be aluminum with screw fastened covers. Inside the concainment they shall be code gauge galvanized sheet steel. Terminal Aoxes sball
.be of. Qe_same.m.atetia.Ls and U nish,withJ_Lnged co 3fSTAINLE55 --
3 TEEP ( boxes _MAY._BE. USED IN A1.1. INDOOR LOCATIOpy 4.2 ,In wet or damp locations indoors, other than the containment, g boxes shall be code gauge aluminum JIC or NEMA 12 type. Inside the containment thqy shall be code gauge galvanized sheet steel. 4.3 Box exteriors inside the containment shall be finished as described in Specification SS-1102-31. 4.4 Outdoor boxes used in the intake area shall be stainless steel unless
- otherwise noted. *
- 4.5 Embedded boxes shall be of galvanized steel or epoxy erameled inside and outside to prevent corrosion.
, 4.6 Boxes mounted on concrete surfaces shall be placed a minimum of
~
k inch off the surface. 4.7 Power receptacle boxes shall be mounted 3 feet above the finished
. floor or grade.
4- 8 '
. Pushbuttons, magnetic switches, etc. shall be mounted 5 ' -0" above -
the floor unless otherwise shown. 4.3 50E5 TRAT ME. SCREWS FOR COVER FASTENING MA'/605621 NELE55 AR'l>WHEN LOVER CLANW5 AP2 STkUPEDat - F,EPLACED WDR bbl.T AND NUT"6F ST\mE SilE OR WITH h 1.h%ER SCREW. 1 a ORIGINAL. ORAWING, Rev _ TRANSFERRED TO SCS -
~
BECHTEL CORP. JOB 7597-03 SOUTHERN SERVICES INC.
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on . 47-.. . . "Q*% s ML Al.ABAMA POWER COMPANY 5 h ea gz.y n num- esy .r. c)/M.h . 3 %s. {yo-[H2%)v.?o? acy JOSEPH M. FARLEY NUCLEAR PLANT W% TRAY 4 CONDUlT DETAILS 4 NOTES t- s.,n .o L % u.o APP. g'. [. 7! o .u ..
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6 o,_23_ . ,c,. A- 17754 i S l 1
c 5.0 NON-METALLIC CONDUIT 5.1 Non-metallic conduit runs placed in concrete encasements shall be as specified on specific drawings and associated bills o'f materials. 5.2 All conduit to be embedded in concrete shall be inspet.ted to assure continuity and correct position before conc, rete is poured. 5.3 All underground condtiit shall slo e r a minimum of 2" per 100 feet j toward ma or between g h
- 5.4 TermiAWe non-metallic conduits with female adapter coupling flush f with finished floor. .
5.5 When duct banks are dead-ended for the purpose of being e ended later, the duct tiers shall be staggered. Each tier shall extend 18 inches farther out than the tier above it. 5.6 Conduit ends shall be plugged or capped during construction. Ends of spare conduits shall be kept capped or plugged. 5.7 Minimum conduit size shall be it int hes. 5.8 PVC conduits shall be terminated in conecete pull boxes or manholes l with PVC end bells unless otherwise noted. I 5.9 PVC conduits may be' installed underground without concrete covering; refer to specific drawings for instructions. 005622 ORIGINAL DRAWING, Rev 1
- TRANSFERRED TO SCS BECHTEL CORP. JOB 7597-03 SOUTHERN SERVICES INC.
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ALABAMA POWER COMPANY 1 5 5'7/ REV.PER S.S.LTR ["g "M ' 1 M.M ' R Ev. A S NOT E.D
. suanc7 JOSEPH M. FARLEY NUCLEAR PLANT o rut. TRAY ( CONDUlT DETAILS $ NOTES c, .fth e, _29_ . m. A- 177341 ? . w.. / /vr/i/ ........ .. ,,. m . . 7
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- l 6.0 Cable Channel 6.1 Cable ch:,cnel for use in the auxiliary building and the turbine building shall b aluminum. Cable channel for use in the containment shall be hot
. dip galvanized steel af ter fabrication.
6.2 Channel width shall be 4", the side rails shall be 1 1/2" for steel and 1 3/4" for aluminum. 6.3 Fittkngs, accessories, and hardware shall be of the same materials and finish as the straight channel. Minit:um Radii of side tails on horizontal
& vertical elbows shall be 24" except as otherwise indicated on drawings.
6.4 In general, channel systems shall be continuous and where char.ges in elevation will not permit standard fittings to be used, adjustable splice plates will be installed in the field as required. However; if achieving continuity is ,
- impractical due to physical limitations, the requirements for unsupported cable distances as outlined in sections 1.22 and 1.23 shall apply.
6.5 han ers and supports shall be 5nacco7dinhwith Cable F Civil Dwg. B- channel,20355T.) Use continuous slot inecal framing (power-strut,. 7 n'isirut., etc.) and channel clamps as required. 6.6 P.auting of cable channel is shown disgrammaticly. Field to determine exset route in case of conflict with piping, pipe hangers , ductwork, etc. Also, field to maintain proper separation of redundant systems as outlined on Shs. 276,270 and'27F. . 005623 I'> 3 ORIGINAt. ORAWING, Rev. TRANSFERRED TO SCS BECHTEL CORP. JOB 7597-03 SOUTHERtl SERVICES INC.
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ALABAMA POWER COMPANY I, ]/g] 6 %% CORaccxED it u . sussect JOSEPH M. FARLEY NUCLEAR PLANT ' 1 '* O AG #" ocra TR AY( CONDUIT DETAILS 4 NOTES gs ,,,jjk/ g,5 Mih5 Rev. G.co omIMh2 .v,s 7 . og. , en, None y 7A 29 at<ms A-177541 G
4 i 7.0 RACEWAY MARKItc STANDARDS . 7.1 ceneral To facilitate cable pulling and inspection, all raceways shall be permanently I marked with their assigned raceway numbers and safeguard train or channel color by means cf prefabricated raceway markers or by stencil painting. All receway g markers.inside the, containment shall.be stencil painted..All others shall be 7 .,1,1 e s N I The field is to provide identification for the raceways that nt wrapped with *
~
kaowool. This can bu accomplished by using a stainless steel tag. sad embossing the raceway numbers on the tag. The tag may then be wrapped around the kaowool bundle witti an appropriate tie wire. Thes e tags should be attached at intervals .I
- 7. '
unsk - % p- -- ~ Raceway markers for application on all tray, and unistrut, or equivalent, shall consist of a marker which presents a color diamond of the proper color scheme, as well as the ,last six digits of the raceway number. These markers shall be applied to the outside of the rail of the raceway, at the si'e most accessible to view, at the following points along the raceway section: i (a) At. each end of the section designated with a triangle, "A", l on the tray and conduit layout drawing. , (b) At twenty (20) foot, maximum, intervals along the section. 4 (c) At any other point or points required for clarification. 005624' i 7;3 conduit & cfiannel . Marking ] l '
')
l 7 , 3'.1 The conduits and 4 inci6 channel sha.11 be marked at ooth ends with a race- $ way marker which presents a color diamond of the proper color scheme as j
- well as the last six digits of the raceway number. In addition, the con-duits and channel shall have a color diamond of the proper color scheme
. applied at 20 foot intervals along their length, where visible, and on each side of wall and floor penetrations. These- diamond markers shall be k applied.at a position most accessible to view. Conduit and channel shall be marked with a complete raceway marker (last 6 characters plus a diamond 2 marker) on each sida of a Nelson penetration where the conduit or channel ag$
c': . are no.t run continuous. through.the penetration. Where 4 inch channel are not continuous along their length, each discontinuous section shall be marked with , g C a complete raceway marker at one' point most accessible to view. 3o 7.3.2 Raceway, markers for application of all embedded conduit or sleeves shall 3y be applied to a metal plate Phich shall be permanehtly attached to the con-4m crete surface.adjace-t to each end of the. conduit, or metal tags, of compat-
$Ey ible m'at'erial for the. area, stamped with the raceway number ' attached with S.S. wire to the conduit.
E '.h 7.3.3 Each embedded pull box shall have conduit markers permanently affixed ad-g 6- jacent to conduit entry into pullbox. As an alternate method of labeling m ', ., embedded conduits lamacoid tags of the proper train color will be glued O directly to the finished concrete wall using Good Year Pliobond or equal. 91 BECHfEL CORP. JOB 7597-03/?.o SOUTHERN SERVICES INC. , FoR APP oV A\. ALABAMA POWER COMPANY N f O i 1 7 fof72 1NCORR 1 thE-NR wm : JOSEPH E FARLEY NUCLEAR PLANT [ .h,,,.f 4,1
^ 3 /472 21.u
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TRAY 4 CONDUlT DETAll5 4 NOTES NONE~ ,,,. 7 6 ,, x ,,,,1 ,
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7.4 Marke r s f or a pplica tior).. on . raceways..ou t_sid.g__th_e i ,c on ta,in,m,an t will be furnished by(Alabama Pcwer Compc.ny. 5 { f[
~ ~ . . . . . . _ ....~. . ._.
7.5 Stencil painting raceway markers inside the containment shall be done using one of the paints listed in section 7.6, accurding to 'the preparation procedurea outlined therein. 7.5.1 Each stencil painted raceway numbers shall present the six (6) character raceway number as __it, agpea,rs on the tray a,nL _ conduit layout,,drgAt the right end of each rdceway number I (/shallraceways. be painted a safeguard train color .. mark, for all..._ safeguard s 7.5.2 All. painted raceway numbers shall be 1/2,1 rc 2 inch depending on available surface area, red, block characters with no background color. 7.5.3 All safeguard color marks shall consist of a 1 inch by 1 inch, minimum, square painted in accordance witn the safe-guard colors presented in section '.5.4. 7.5.4 Co?. ors used in raceway number safeguard color marks shall be as tabulated below. SAFEGU4RD CHANNELS AND COLOR CODES , , Train , Train Or Safeguard Color' i, .ha rac t er
- Channel N** Non-Safeguard or Normal None Required A Train A Red 005625 E Train B Blue b >
C*** Train AB Red and Blue f S8" "'""' ** " ! 0 2 Safeguard Channel 2 Green 3 Safeguard Channel 3 Orange
- Safeguard Channel 4 y$
'A 4 Silver Outlined in Black I
gQ
- Note 1: The train character is that cbaracter appearing in the third ng
~
character position of the spaces marked: on the EE-553-2 l L 3m Cc - rr.ceway installation cards. RACEWAY NO. 4 cc
*g g ** Note 2: Markers for raceways of this train require no safcguard . .J gva color mark.
h *** Note 3: Markers for raceways of this train require two (2) safe-
' O. b guard color marks - one red and one blue.
;n: '
,o _
! BECHTEL CORP. JOB 7597-03 SOUTHERN SERVICES INC
- Mo? DATE mzyt. nom on _D O. .Gt,.
o N,42 FoR APPROVAL ALABAMA POWER COMPANY
~
Rev. Per FCN 5- ' JOSEPH M. FARLEY NUCLEAR PLANT'
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[EW*" m TRAY 6 CONDUIT DETAILS 4 NOTES oin # .d..... lf .u........ .c.u NOWE . l c_ o ,_ a _ .o m . A- 17754i 1
7.6 Paints used in stencil painting raceway numbers shall be . applied as detailed below. 7.6.1 Surfaces without breaks in the ;;alvani::cd steel zine coating and without field welds and cuts, shall be cleaned and prepared in accordance with the p int manufacturer's recommendations for painting galvanized surfaces and the procedure outlined in paragraph 7.6.3. 7.6.2 Surfaces with breaks in the zine coating, and all y field welds and cuts shall be thoroughly cleaned in accordance with the paint manufacturer's recommendations. If stricter instructions do not appear. elsewhere, the f ollowing shall be considered th2 minimum acceptable standard: weld spatter an( slag shall be removed in accordance with Steel Structures , ,, Painting Council (SSPC) - SP3.**The areas shall then be rnopped i w!th a recommended solvent, to remove welding flux and other t impurities, and washed with fresh water. After air drying, one [ coat of zine rich primer, recommended by the paint manufacturer l for galvanized surfaces, shall be applied. j 7.6.3 Af ter surface preparction in accordance with sect!.ons v 7.,6.1 or 7.6.2, and immediately prior to paine.ina, cha e- %. j shall he vashed or viced wi r} Mineral Spirits in accordance with j SSPC-Pl-63, fc11 owed by light sanding to get " tooth".. i
'.).4 Acceptable paints for use with this procedure are '
as lis.ed below. - Phe,noline No. 305 finish - 1 coat @4 to 5 mils - dry film thickness (DFT)
* [
j , OK a Mobil, t @ 4 mils (DFTi l Pittsburgh Paint Quick Dry Spray Ensmel in g AerosgCys ,
%v i
- Colors shall be a' visual match with the following:
PIENOLINE (CARBOLINE) MOBIL . Red #502 Accent Red 12- R-3 Blue #119 Navy Blue 12-B-7 I Yellow #627 Yellow 20-Y-5 QQ%g I Orange #409 Orange 12-Y-15 Grey #720 Silver Grey 12-F-18 Green #351 Green 20-G-ll
** REFERENCE Steel Structures Painting Council '
'2-.
ORIGINAL DRAWING. Ilev.
*4400 Fifth Avenu Pittsburgh, Pennsylvania 15213 TRANSFERRED TO SCS BECHTEL CORP. JOB 7597-03 SOUTHERN SERVICES INC.
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navi.eou ALABAMA POWER COMPANY ~ o n.6 72 FoR APPROYA L vn .- 1 . 9/t3/x P*r BE-573 Afid sumaci JOSEPH M. FARLEY NUCLEAR PLANT cah3f.[. 1 1 dW74 PER BE-G*>5 omn. TRAY 4 CONDUlT DETAILS 4 NOTES q f a,
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Symbol, Definition Conduit Turning Towards observer C f i 0 f Conduit Tursiing Away From Observer ( - f f Exposed Conduit f-- - - -f Embedded Conduit
. Conduit Tee Branch
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~ pr -_ M Flexible Conduit
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o Pushbutton 005627 a' Power or Control Panel p g TRANSFERRED TO SCS BECHTEL CORP. 0B 7597-03 SOUTHERN SERVICES INC.
~~~~
7/ REV.PER S S.LTR ALABAMA POWER COMPANY
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+ SEE SPECIFIC ORAWINGS FOR MARK NUMBERD BECHTEL CORP. JOB 7597-03 SOUTHERN SERVICES INC.
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- OVER E" UTILIIE DETAIL OM 5H. I*> OF A-177541 WITH CND. ENTERING SIDE.oF BOX 'IMSTEAD OF TRAY. /
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u CONDUlT TO TRAY DROP bUT~ ' BECHTEL CORP. JOB 7397-03 SOUTHERN SERV!CES INC. om ALABAMA POWER COMPANY Fon
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FULL STORY POST FOR CLASS IE CABLE TRAYS s
.I NOTES 005631
[ @, 1. ARRANGEMENT FOR ~.ONNECTioH OF B RA CKETS 'THAT SUPPORT _
.o TRAYS IS THE SAME AS THE HAMatR TYPE. SE18 SHEET,12.
2.Pc57. DIAMETER 15 VARIABLE fro:A L'/C To6", EAC.M POST I j -@F JHALL BE. Sl?.E3 BY INDIVIDUAL CALCULATIO - S TAKINCa gE
< INTO ACCOUNT THE LOAD, LLNSTH OF POET, r4D DESIGN
=
o n.E . EARTH QUAKE C RITERI A. --
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-a BECHTEL COf#. JOB 7$97 03 SOUTHERN SERVICES INC.
, I sn REV.P S . MR.
ALABAMA POWER COMPANY ., [gf 2' H at veinta ., JOSEPH M. FARLEY NUCLEAR PLANT f m. TRAY ( CONDUlT DETAILS & NOTES I
,CE!/vz/j . . . NOME = _u. _ - A-177541
~ ~ - _ _ _ _ _ _ _ . _ _ _ . _ . _ _ . - . - - . - - . . _ _ - - . . - - _ - - - - - _ . _
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- WELO TO EMBEDDED N P $TELL INSERTS 1 A A h
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h~ ema oossaa yg y vestire.euxte NOTES 2a ci h i h ) L DIAMETER "D" IS VARIAbLi FROM zo , 2N2"To G" AMD in DETERMlHED BY IMD)VIDUAL CALCULATIONS TAXING 3:<E m INTo Account LENcTH "L" AND DE%lON E.ARTHQUAKE, LOAD,
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k N 523 I h< h. TRA'r SIZE LENGTH D M AX l.OAD TRAY SIZE LENGTH D MAX LOAD s' 14%* 700 LBS S' NOT RECOMMEN010 12' 17%" 525'25 12* 17%' 1000LBS ! 18* 25 %~ 350LBS 18' 25Y 725 LBS ! 24' NOT RECCMMENDED i 1/ 2S%" 500LBS i
. 0056 h*
i NOTES BILL OF MATE. RIALS i No QTY MFG, No. OC SC R IPTION ! l. ALL MATERIAL UNISTRUT OR i p. lO75 BRACKLT l EQUlYALENT. g g p.loto NUT l E.To SE USED IN CLASS lt. 3 3 .... 1/2"nt" CAP SCRE.W R ACEWAY SYSTLMS ONLY 4 gg gooo Sige:tg egAygtt , WHERE VERTICAL SPAcikh 5 i P-IS93 BRACK.LT ! IS CRITICAk ' G & P' lool DOUBLE CHAMMEL.
- 3. MATER 1ALS SHOWN INTABLE y y p.32 53 CONCALTE INSERTS AT RIC=HTTo BE INCLuoED IN THRU-ALL APPLICABLE. SILLS OF P 3170 l MATERIALt.A r
A LLNCaTHS AS RLQutRLD i YolCA'_ TRAY OR CONJUIT BRACXET BECHTEL CORP, JOB 7597 03 SOUTHERN SERVICES INC.
^^ ALABAMA POWEN COMPANY
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.m.! 2! .-. w w.n. NONE . _ ti. ., _1;_ ... A- 17 7 541 -
i ORIGINAL DRAWING, Rev. 2_ TRANSFERRED TO SCS
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- 1. ALL MATERIAL UNIS f RUT ; g p.lO2.6 ANGLE. FITTlHG cm. EqulVALENT. 1 1 P 2452 b R A C E.
1.5EISMIC CRITERIA To BL S + P* lO l O nut- - < E CHECKLD BY STRUCTURAt. 4 4 - y2'13xt CAP SCRE.W SECTION FOR- USE IM class 5 m F1OOI DOUBLE. C HANNE4.- ( lE. TRAY SYSTEMS. 6
- P-looo SINQ LE, CH AN N E.L
= 1. 3.SEE SHEET IS, NOTE. 5. Mie LLHQiTHE A% Rt.QutRED TYPICAL . , TRAY OR CONDUlT BRACKET SUPPORTED FROM WALL BECHTEL CORP. JOB 7597 03 SOUTHERN SERVICES INC._
~-
8.LTR. ALABAMA POWER COMPANY 1 s 7/ REV.PE 7)$ A[.~.; r. 5-tu voict.D .. JOSEPH M. FARLEY NUCLEAR PLANT __ TRAY 4 CON 0utT DETAILS-4 NOTES
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NONE- 14 .,_?,B. m A 17 7541 . ORIGINAL DRAWING, Rev 1 -
' TRANSFERRED TO SCS
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- SCRE.WS (TYPIC AQ SEE. NOTE. 3 1/2"MARINITE XL(CUTINTWO .,
SEE icTE1
' PIECES- ANCHORED WITH
- MACH, SCRE.WS TO WALL OR B
TLoOR.) SEE < SH.i6 _OPENINC1 IN 5EEMOIE \' CABLE TfMY gfLCOP.OR. WALL 23 y
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L______________________J MINMdM A" -i- l l j mA Ki3 sEE SH.16 SECTION C-C sS (FROM SHEET IG $18) 005635
*a 3m NOTES ,
l,
- l. FILL LARGE VoiOS VJITH \HORGAMiG THERMAL IMSOLMiMG r
\ 0$ WOOL. ' Y.$ 2. SEAL OPENtWG SETVJEEN THE TVJo PnECES OF MARiNlTE BOARD VJiTH 8/4."
$ $- "FLAMEMASTIC" ll A MASTIC.THE MAST lC "oHOULD BE APPL)ED TO THE g
o FIRE STOP IN THINMER LA'fERS OF ABOUT 1/8" TO AVotD CRACKING A%
- THE MATERI AL DRIE*a, T'1P. VJHERE "FLAMEMASTlC" IS TO BE APPL \ED, S. COVER SCRE.W HEADS WITH 1/8'l T. HICK " FLAMEMASTIC" llA MASTIC.
- 4. FIRE STOPS ARE PROVIDED TO % TOP THE PROPAGATION OF FnRE FPOM ORE ROOM 1RTO AMOTHER OR WALLS FLODRS OR CEILING % THAT HAVE 3
~NO CLA%SIFICAT\0R OR RC)MG, FIRE STOP FOR CABLE TRAYS THROUGH FLOOR OF.
WAL_ OPEN NGS - BECHTEL CORP. JOB 7597-03/2 0 _ SOUTHERN SERVICES INC. 5- @ voto E EE-%1 ALABAMA POWER COMPANY [f/[" ,1 ;~[ 2 7-2 74, R E v. As MOT E D sunarcT JOSEPH M._FARLEY NUCLEAR PLANT a b29: REv. As NOTED T RAY < CONOUlT DE TAILS e NOTES 4
) D ET^"-
j/ #g//Ul....__ g 4 2 WI VO)D o , _._ / . .. ... NONE . / s e, x ..m A- 17 75 415 l
.VE RTIC AL. CABLE - TRAY TO HAVE TRA'f COVER A MINIMUM TOC 7:FT. ABOVC FLOCR LEVEL TO PROVIDE PROTEC.Tio N FOR : ' CAB LES.. .
M i l I I I I l l/2 MARINITE XL ANCHORED WITH ] (SCE iCTE.S)
. MACHINE SCRtWS l l +
(TYPICAL') ; ! I OPENING l I I C6 ' l 6C h -l l l l SHIS s . [' l 11 , h I gi ef4
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g (g* 7 "^* ee Appty THICK'penaox. iA~FLAMEMASTIC[ SECTION A-A Eo . 71A (SEE NOTE 2.) (FROM SH 15) , 3m + . . g: . . a*
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NOTES '
.00560s ~
Ee ?.'" l~. INSTALL MtRlHtTE XL ' AF BOTTOM BEFORE. APPLYING y -
FLAMEMASTIC" INSIDE PENE rRATIONS o 2.. .CcAT CAE>LES' "AND MARINITE XL AT PENETRATION WIT H"FL AME MASTIC 7l& #
- 3. JNSTALL MARINITE XL AT TOP SWHEN FLAMEMASTIC '
APPLICATION 15 COMPLETED. NO SEALINC, REQU IRE O. FIRE STOP FOR CABLE TRAYS THROUGH FLOOR OPENINGS BECHTEL CORP. JOB 7597-03 SOUTHERN SERVICES INC.
' ao. u o*T navi. ion
" " M.A '"""~~ ~ ALABAMA POWEE COMPANY
/g 3 5 5 7/ REv.PER s.s.LTa.
sj ', $ $" 2 AN4 REV. AS HOMD susarer JOSEPH M. FARLEY NUCLEAR PLANT
$j ,,, ! . . . . . . .
h vo - py.3c.%I DET^n T' RAY 4 CO N D U IT D E TAILSJ NOTES NONE .s _t E _o ,._ n ... A- 177 541 4 J. oare .R .u ..com. ____ ]
'l/2" MARINITE- )(L U SE. INORGANIC THERMAL INSULATINS WOOL AS REQUIRED TO FILL LARGE VotDS.
NOT REQUIRED IF SLEEVE. I S 1/4" T H IC K " F L AM EM ASTIG" INSIDE A PANEL. ll A APPLIED FROM ABOVE -- Ah10' BE. LOW (=oEE NOT E '2, - SH. l5\ '
\ : .-- ,. n CONCRETE.
~
ELECTRICAL FLOOR SLAB 1, _. CABLE.S
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{ SLE E.VE. THROUC,H k SLAB._._ I i
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t
' AROUND CABLE.S. SEAL l/4'-20 MA HIN._ SCRE.WS GAP BETWECN THE. TWO
, d0 SPREA D ,C.FLAM E., MASTIC'. PIECE.S WITH" FLAME MASTIC" l f! so 71 A MASTIC t/8 '. THICK.. .
.o
- 7lA MASTIC . ,
o >- ' OVER J SCP.E.W .. HEA.DS . . '
!$ @ ") 005637 55 0 'USE INORC, ANIC THERMAL iMSUL ATINC, WoDL AS RE.CtUIR.E.D To
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FILL - LARCd. VotDS ARoOND . CABLE.S THROUGH THE MARINITE 25
- F O RM. SM AL L Vot DS U P TO l/4" CAN BE FILLED WITH 9 *~ 4'FLAMEMASTIC." COMPOUND. THE THERMAL IN5ULATINCh _ Wool o ; .- , HALL BE BgW'S KAO Wook BO LK OR E.Q U AL . .
FIRE STOP FOR TA3LES THRUUGH FLOOPs SLEEVES SOUTHERN SERVICES INC. BECHTEL CORP. JOB 7597 03/ 2 0
,' o na,_.. " -
*y, "Q "g,28E-% ALABAMA' POWER COMPANY l [."jg. ..
2 7194 REV As NOTE.0- suurer _ - JOSEPH M. FARLEY NUCLEAR PLANT , 5 3-2 M REV A*o HoTEo J 4 M'.3/ VOlO "l_ RAY 6 CONDUIT DETAIL 5 d NOTES t. NONE . 17 or 23.....,. kU7Mi 5 fjob..'.......![.u...ces. .cu
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e .. . , . sEE sw. is 3.a SEC. TION BB
>u EM -
O (FROM SMEET 15) . Ee NOTES - kw E t tM57M.). MARtH1TE'XL FORM BEF ORE APPLYtt% , i a: 6 "Ft AM5 MASTIC"lMSiOE, PENETR ATtON . C "- '2.. COAT CAELES AMD MAR \M\TE* %L. tMStDE AMD
'y E 007S10E. fEMETRATLOM WITM"PLAMEMA% TIC." 71 A,.
z4 EE 5- 005638
. P! 5. STOP FOR CABLE TRA% THROMM WALL OPEMiMGS BECHTEL CORP. JOB 7597-03/ 2 0 SOUTHERN SERVICES INC.
" ~k ,\N. N O. - D ATE REW'380N Al ABAMA POWER COMPANY D- ~"
5 5ff2 VotD PER 2BE%I ' 1@ r..f..f..lg.
,, 2 A2-7/ REv. As NOT E.D w : JOSEPH M. FARLEY NUCLEAR PLANT N$ voto = ='m TRAY 4 CONDUlT DETAIL 5 d NOTES g '3 ///[
,,,. g/vrb/ -j A-l 77541 5' o.1. ...t .u,..... . .co. NOME : .m is o,_2s....,.
~
%. WALL,
/[.
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QPENiblQ FILLED WITH SILICONE FOAM'
/
j 4 // [ , CA.BLE \ r i . t t , i Oc9AOo OPSb / ht a 3,
'b CABLE TRA l % %
1 ' l J; _,
- j $
- _ ELEVATION _ NO SCALE SECTION 1-I
- ,o .
+
e >- NOTES.- - i f - lif. FIRE SEALS ARE PROVIDED ou FIRE RATED WALLS' VJHICH $AY , . Te, d . 6E COINCIDENT WITH AIRTIGHT- WALLS.
$ 2. FIRE -RATED FLOORS, VJALLS AMO; CELL lNGS ARE DELINEATED
{y$ z ON ARCHITECTURAL DWG'S, D-l%OS6 THRU 06S AMD z$ D-?.06 05 8 THRU OG3, o 5. AIRTIGHT AWD WAT ERT'IGHT~ FLOORS WALLS AND CEIL.lNGS 3 ' ARE DELINE ATED ON ARCHITEC TURAL- DvJG'S.D-l%oG5 THRO O Yt. AN D D 'I OG O G 5 T HRu 10 l '2_.
~
- 4. FI RE SEALS AND air SE. ALS ARE. TO E>E INSTALLED fN -
l ACCORDAMCE WIT H SPEC. %%-Ilo'L - 13'2 AND AS OUTL\ME'O . HE R E. l W. i
-0056C9 i
BECHTEL CORP. JOB 7597-03/ 2 0 SOUTHERN SERVICES INC.
"*- *^" "'
- o. 4..tB ALABAMA POWER COMPANY o- shk Fem e(.s" '"
r u e,% c5Nhb 5 547/ A00E0 HOTE) suancT JOSEPH M. FARLEY NUCLEAR PLANT 2 727/, RE.V. - A5 MOTE.D DN* TRNT 4 CO!JDulT DETAILS 4 MOTES oan. 47$ susa.. son. .ma a l8d or _D_ . ,, A-177541 Y
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l - l,/ /i, ,/; , I " i ' , CABLES ff 1 l l l h/ /
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i OPENINQ FILLED
! I WIT'H SILICONE l h 'e I,
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I j i I L , 6 i 8 {' Floor St,AS } ,,,g i I 4--- 1 . ! ELEVATION NO SCALE SECTION I-I i I q ~/ / ,/ / / / / / I'/7 G/'7 7~/ 7/ /9 i 7 , 'i // // / /// // / / // / ///
/
,. y 2 / / / / / / / / / / / / ' / /_,,,/.
o S o 90'$'$ WALL WHCRE APPLICABLE N 0 C oo s o o r
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g8 .- - j s y. cAsLe aista OPEbjlNG FILLED WITH cAatts silicone P'OAM
<g us 0056CO Z<m NO'? E N 9* PLAN g 1. I:OR GEhlE.Rh L NOTEca SE E SH. ) 6 A, i
BECHTEL CORP. JOB 7597-03/ 2 0 SOUTHERN SERVICES INC.
"" " AL.ABAMA POWER COMPANY 3 4 ron c nucTiou c,&WfL. ( [ 3 V-N-7; ADDED NOTE sur.n cy JOSEPH M. FARLEY NUCLEAR PLANT 3
dTl;2.s REv. A5 14OTE.0
^'
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. - e,v- -2 = ., , w+.y=-,e-,-t se,,--.+n--..- -,f, , ,g- -vy--,. 3.y-+v w r -t- w++-w ,- ..gr,+>. .ra
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'L 'L 'L l t I e
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\ I 4-.
! SILICONE FOAM CIRCULAR SLEEVE RECTANGULAR SLEEVE I ELEVATION S i !- g. NO SCALE v.
/ WALL -
j g . OPENIN4 FILLED WITH
-o -
SILICONE FOAM l j { 9 -v A l 5g C fh i C 'w --- 7 [ i
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--- /
l h 0 . l g coAlouir l NOTE.s /j/ /, j
- 1. FOR GENERA,L NOTES -
! \ SEE SM.16 A. [,- '
'W SECTION I-I 005641 '
NO SCALE i . ! BECHTEL CORP. JOB 7597-03/2O SOUTHERN SERVICES INC.
,~ O 3 r/74 FOR T RUCTick' 3 4N.7'1 ADDED NOTE .a JOSEPH M. FARLEY NUCLEAR PLANT
[ en.8WA._g/ DENL TRAY 4. CONDUIT DETAILS 1 NOTES 2 7-1-74 REV AS NonD gq A sM. b or d smarts A-177541 ' 3I
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fig P.O. BOX 338 ** 2 WEBSTER, NEW YORK 14580 Y (716) 265-1600
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, . PLAN OF CIRCULAR SLEEVE _ PLE.N OF RECTANGULAR SLEEVE NO SCALE O conjouir
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Q SECTION I-I 005642 NO SCALE FIGURE 8 BECHTEL CORP. JOB 7597-03/ 2 O SOUTHERN SERVICES INC. is aa #4 -- o '," g (({" ALABAMA POWER COMPANY h "gsg. g 3 N+p D O E D W T E. ,g JOSEPH M. FARLEY NUCLEAR PLANT D N"- e,2 './-2-M REV. A S N OTED [ .g7f/
. . . .u........ ._ .*lBRo,2EL.om. A-17754( Y u - ~~ j
,. i SILICORE FOAM AS ' '
SPECIFIED Oil -- SHEET
- ISA MIN.11 WE%tSTING --
CABLES TO TERMINAL BLOCK n - n, o SEAL COWDulT WITH i MINERAL WOOL tr ei vt gitico? f SPARE CONDUlT CAPPED ON SOTH 1 SILICONE FOAM PRIOR TO Fos=-- ENDS FOR FUTURE l POURlWG OF SILICONE FOAM '
- USE !
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BECHTEL CORP. JOB 7597-03/20 SOUTHERN SERVICES INC.
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. ALABAMA POWER COMPANY O !/a/u FoK (oNSTRUCTiod
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.n KOTE ,
- m (l.RACEWAvs THAT ORIGIMATE ouTSIDE THE COMTROL ROOMN j_$ MUST f6E' SEALED WITH A MINIMUM OF 11" 0F c o1LICONE
<m FOAM AT THE COWOUlT TERMINAT\0N IM THE COMTRO '
E{ ROOM AFTER ALL C.ABLE IS ' POLLED, Y$ w -
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,~ Y 005644 BECHTEL CORP. JOB 7597-03/2 O SOUTHERN SERVICES INC.
ALABAMA POWER COMPANY
." O I D0R C JST CTIOM cxM.R[.25 3 391$[ '[ suancy JOSEPH M. FARLEY NUCLEAR PLANT REV.k5HOTED D
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.. m . .u . LefE.o,21. m. A-I77541 Y_
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SPECIFICATION DATA _ APPLICATION CUIDE KAWOOL BLANKET WRAP _FOR FIRE _ PROTECTION t
, 1.0- Scope ..
1.1
'Ihis specification guide covers the application of Kaowool Blanket
. wrap around electrical raceways for fire protection.
2.0 Ma terial_e_ and,_Ecul oment 2.1 The insulation material shall be manufactured by and supplied by the Babcock & Silcox, Refractories Division, Augusta, Georgia 30903, or its
;. authorized distributora. The materials shall include the following: , '.' 2.1.1 Kaovool Blanket, needled, 8# per cubic foot, 48" wide,1" thick, standard length.and \" thick for inner wraps only.
2.1.2 High by 3Mperformance. Company, St. filament Paul tape,1" vide, such as that manufactured Minneso M a g Number g B,or ,its 4\ 2.1.3 (Anchor 3[tinent_a17apetypeJ8happroved -
~ ' " ~ ~ equisalent. One s Regular waxed. carbon stWhanding material, 3/4" vide, 0.015" thick, 2.1'.3a Regular carbon steel handing - not waxed - 3/4" wide, .015" thick may be used in place of the waxed banding. However; installers should periodically check to ensure banding installation is not~ degrading the Knowool .or Zetex wrap.
2.1.4
. . Tensioning too1~ for banding materisi such as that manuf actured by Signode Company, Chicago, Illinois, Model Number C-38-34, seals such as that manufactured by Signode Company Model Number 34SPC, and crimping tool such as that manufactured by Signede Corporation bdel Number SRC 3423, or its equivalent.
3*.0 Deliverv and Stora[e , 3.1 . Materials should be delivered in the original, tightly sealed containers or unopened packages, all clearly labeled with the manufacturer's name, product identification and lot numbers where appropriate. 3.2 The Kaowool products shall be from weather. stored within shelter to prevent damage
~
n n r t a e-ORIGINAL DRAWING, Rev7 TRANSFERRED TO SCS i SECHTEL CORP. JOB 7597-03/20 SOUTHERN SERVICES INC.
~
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d ISSlb5bNST. ALABAMA POWER COMPANY [g]li[ , ,_ l J 3@ INC. BE-4515 3 JOSEPH M. FARLEY NUCLEAR PLANT 2 492 INC PCN Pr81-10&5 susact
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.n k INC.PCN B 821509 4 ocra TRAY 8 CONDUlT DETAILS i NOTES e i+ o-3,n . u ,, . . . e . . - - .c. .
. 18G y
\ GT- tcV.n A-17754 l 13 L ,
g.. t 4.0 contrae m ,, 4 '.1 Kaowool Blanket wrap shall be applied by the field.
,. 4 .2 The field shall be responsible for selecting and supplying all labor . as well as furnishing other materials required to complete the job satisfactorily in accordance with the written specifications.,
5.0 Kaovool_B_1_anket Installation 5.1 For cable trays with no cover and not completely full, Kaowool blanket , shall be cut to the same width as the cable tray and laid on top of the cables to bring the level of the Kaowool to the same hsight as the
, top of the tray. For cable trays with covers 3r trays that are -
sufficiently full, no Kaowool filler blanket is required. 5.2 For the interior wrap, the Kaovool blanket shall be cut to a length sufficient to wrap completely around the perimeter of the cable tray, plus provide an overlap of blanket of not less than F . (ly: (2) wraps " in thickness may be substituted for 1" wrap far inner wraps only) a 5.3 The interior wrap of blanket shall be held in place using 1" vide
,. filament tape. The tape shall be spaced not less than 4" from the ends of each blanket and not more than 14" apart from each other.
5.4 The next adjacent int rior wrap of bisaket shall be cut to completely wrap around the perimeter of the cable tray with enough excess to-provide an overlap of not less th.r.h 3" The adjacest blankets shall
. be placed tightly together to', prevent any gaps in the butt joints between each blanket. ,
5.5 For the exterior blanket vrap, cut blanket to sufficient length to I wrap completely around the perimeter of the interior wrap with O sufficient excess to provide an overlap joint of not less than 3".
,e The blankets to be installed on the exterior layer shall be cut if K
%g needed, and properly located to ensure that the butt joints on the erteri r layer f blanket are not within 12" of any butt joints on the O
eH interior blanket wrap. EQ
- 3E 5.6 The exterior blanket shall be held in place using regular carbon
[gu. steel banding material as specified. The bands shall be placed not c more than 4" from any butt joint on the exterior blanket, nor mora J than 14" apart from each other. The tension on the banding material j <$t shall be sufff.c19 nt to hold the blanket snugly in place, but not to 3% cause any significant cutting or damage
~
the blanket material. E o 5.7
'm 00564G Along the length of the cable tray in areas where support members ',
come in contact with the protected tray, such supports shall be , wrapped using'the same techniques to a distance not less than 6' from the protected cable tray. / % -- , 5',/ BECHTEL CORP. JOB 7397 03/20 . SOUTHERN SERVICES INC.
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ALABAMA POWER COMPANY l .ggg i J3S IMC. BE-4515-3 suasse, JOSEPH M. FARLEY NUCLEAR PLANT
- z. o.g w.etu n s m o9o-s k NOTFS l
arra TRAY & CONDUlT DETAILS
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5.8 .In-areas such as floors, walls, ceilings where the blanket vrap ends, fire protection seals shall 'be. installed using mastic coatings
.as specified. (One type of sealaniwhich may be used is a silicone rubber' sealant - SF 60 - as made by BISCO). The scalant shall be '
j trowel or sprayed es pletely around the wrapped cable tray at the floor, walls,'or ceiling, The seals shall be not less than 4." ' thick and extend not less than 8" onto the Kaowool wrap,. nor less than 8" onto the floor, wall or ceiling. Another type of sealant which may 1 may be used is Carboline Intermastic #285. In addition; at all l
.. knowool terminations char are not at a 3 hour fire-rated wall, l Flamemastic #71A(or Flamemastic 77' Mist r s manufactured by _
Flamemaster Corporation may be use-. . . 6 5.8a At locations where a curb -fs used 'to protE the conduit stbb-ups from moisture, only the top. surface area of the curb needs
- to be protected with-the mastic. coating, s.
5.9 For Kaowool installation of conduits and channels, follow the same procedure used'for cable trays.with the dimensions shown on Sheets 18M, ISN, ISP,18R end 18S of thic drawing. 5.10 ForMechanic51pretectionofknowoolWrap,whererequiredto (A) prevent exposure to heavy wear or mechanical damage, the field
'is-directed' to insta11' Nevtex Co. product #10801-2440,* sytle
~ '
- B00 no'n-aluminized .zetex" cloth.OThis type nf protection shall be.used on power, control and inst'rumentation level' raceways. ,
.2 . . . .
(B) If additonal power raceways are required to be wrapped other rthan those shown on.the dwgs./ change notice, Bechtel Engineering must be given written notification before the wrappin'g is installed. This is required so that the " Watts /Ft." can be calculated and compared with the Prototype Test to ensure that no unusual heat build-up will cause the conductor temperature to exceed 900 C. (C) If additonal instrumentation or control raceways are required to be wrapped other than those shown on the drawings / change notices, Bechtel Engineering must be given written notification of the . . identity of all such raceways so.they may be properly documented.' 005647 ORIGINAL DRAWING, Rev. 8 TRANSFERRED TO SCS l SOUTHERN SERVICES INC. ( SECHTEL CORP. JOB 7597-03/ 20
" N- "' "
ALABAMA POWER COMPANY O E Fh bNST. 4 $Woz INC. BE-4544 JOSEPH M. FARLEY NUCLEAR PLANT
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- 3 CARBOM STEEL BAUD e OR'EQUIVALEMT
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- BECHTEL- CORP. JOB-7597-03/20 -
SOUTHERN SERVICES INC.
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ALABAMA POWER COMPANY O W/e ISS.FOR CONST. sumacy -JOSEPH M. FARLEY NUCLEAR PLANT
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~
005649 W PARTIALLY FILLED TRAY REQUIRES FILLER. PER SECTIOM 5.1. ORIGINAL DRAWING, Rev _ O TRANSFERRED TO SCS BECHTEL CORP. JOB 7597-03 / 20 SOUTHERN SERVICES'INC.
~~~
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i BECHTEL CORP. JOB 7597 03 /20 SOUTHERN SERVICES INC.
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9 SOUTHERN SERVICES INC. BECHTEL CORP. JOB 7597 03 /20
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BECHTEL CORP. JOB 7597 03/20 SOUTHEFfN~ SERVICES INC. l
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4
'Q BECHTEL CORP. JOB 7097 03/20 SOUTHERN SERYlCES INC. ,
**= a.m'o a
.. 0jb a o-co ALABAMA POWER COMPANY e gs
'eg susner JOSEPH M. FARLEY NUCLEAR PLANT
,,///.[g. :
om,cTRAY & CONDUIT DETAILS & NOTES
/
..,6 nit 4-
... S+ P . m ...... .. . N ,, ie+ cf 2 9 A-177541 W in 7
c _ NELSON FRAMES
./ _/ RGS-6 (TYPICAL) r r
~ MARK NO. BSA + MARK NO. BSC FOR 12"x 12" FOR la"x 24"
,~
PENETRATION PENETR ATION wMARK NO. BSB l + MARK NO. BSD FOR 12" 18" FOR I 2 "x 30" PENE.TR ATION PENETRATION
-~
MARK NO. BSE - FOR 12"x 3 G" PENETRATION M ARK NO.
. =
B5F FOR I 2"x 4 E" PE_NETRATION l --- M ARK NO. B5G FOR 12148" PENETR ATION
--- M AR K N O. '6H D. _
FOR I E"x 5* PENETR ATION f, j FOR OULKHE.AO AT SOUTH "g -M ARK NO. B31 TUNNEL ONLY e M' S E E. O W G, 0 - 17678G Eo . 3E eEg -MARK No. B57 FOR BULKHEAO AT SOUTH TUNNEL ONLY
-s-U 2-M R ", '+ wore sus.icer JOSEPH M. FARLEY NUCLEAR PLANT g mm TRAY 6 CONDUlT DETAILS 6 NOTES A99 . ..
D ATE ., .7 gyPER3gggg SCALE S H. OF 79 gMggyg
_~ , h 8 ' f
- WATER TIGHT om AIR .
TIGHT ff Fim.E PR.OOF AREA *. e
--> C '
go u
' > P_l" (WHERE FOSSIBLE)
- j FOR PL ATE DETAIL
~
APPLY..FLAMEMASTK.. ' REQ'D SEE A 177541
... f 8/4" THICK (SEE NOTE 1) ..e a FIELD SHEET 19.
6
- WELD p e .
> i
- l CABLE TRAY CONDulT,
-- - ANDC '
(ABLE TRAY, CONDUlT, g- - EXP 051ON AND CHANNEL. . proof FITTING / EE=EEEEE E E -+ E -~ - - - H.x ' EE ( .- :. L n[ [. - P _ .- . -" - QNDUlT 60lN6 THRU g40NDUlT GOING TURU.j/
, NEL50N .RENETRATION,
-- , ($ NEL50N PENETRATION (WHERE P0551BLE)
RGS-4 CABLE TRAMSIT 1 ,
! y
- e 4
l 5 HOP + WELDED TO PLATE (,
,, *, To FIT Blockout SY e i. 4. MELSoM ELECTRIC Co.
i0 EXTEMDIMG TR. A.Y, CON D. ' 4
-SEE SHEE7 lb E NCTES 3 5, G , 14 8 my ~ AND CHANNELS INTO PENET# , ^
ri F . RATION $ OPTtON AL .
- E. LECT R ic A L
^
ZQ . PE NET RAT IO M , . {g r: m
--. C.
SEE SHT.15
'h.. ---
oa
'z NOTES Ig
~
- 1. COAT ' CABLES IWSIDE AMD OUTSIDE PEMETRATIO N @TH
) F FLAMEMASTIC. 11 A ~ w -m, n ._ n ..
W _ 2.FOR ALTERNATE METHOD OF SEAUNC,- SEE NOTE 2, SH.198. ) o WATE R TIGHT PENE TRA~'O \ S_ ~ SOUTHERN SERVICES INC. y, BECHTEL CORP. JOB 7597 - 03 / 2 0' b,,4 o [$ ro R. kPOVAL ALABAMA POWER COMPANY 6%, d [MdA 2-Y1/ RU ^s worEo 4 S >D; REV' AS ROTED sus;r.cr JOSEPH M. FARLEY NUCLEAR PLANT
-/ 5 D"^TR AY & CONDU!T DETAILS E NOTES g 3 N 74 REV A5 N OTED
[I[./6hM2 .u w.cu. .cu uouE . 19A ;, _a_ .mm, A-17754115
1 - a To expsdite Unit 2 completien, ths Field may use silicone foam elastomer
- 2. 'and/or concrete and/or non-shrink grout per Inquiry No. SS-1102-6 tc seal penetrations for fire, water o'r air tight rated walls. N /
Silicone foam and elastomer shall be installed in accordane_e wiJh m Inquiry No. SS-1102-132 and(APCo Production Deparlmedroc.1}P-0-MP-35.1.) The Field shall be responsible for maintaining proper train and service level separati_on for all ctbles !.nstalled. When physical limitations make it impractical to maintain the 3 ft, horizontal and 5 ft, vertical separation
-of redundant circuits, these circuits shall be routed in rigid steel conduit thru the silicone foam. .
- 3. When silicone foam and/or concretesand/or non-shrink grout is used, spare conduit sleeves are to be installed and capped. The field is to notify engineering of the number, size and location of these sleeves that are inct:alled in each penetration by "as-built sketches."
When concrete and/or non-shrink grout is used, all cables shall be routed 4 thru the penetration in galvanized steel conduits. 2
- 5. Wherever practical, all aluminum conduit that penetrate fire wall or pressure tight boundaries shall be fitted with explosion-proof or condulet fittings and be filled with silicone foam at the boundary. These boundary f: s rs, walls, 6 ceilings are delineated on architectural dwgs. D-176058 thru D-176063 for Unit #1 & D-206058 thru D-206063 for Unit #2. When it is not practical to seal the conduit in this manner, aluminum conduits shall be sealed at both ends wi:h silicone foam. In this case, typical points of application would be at condulet
. fittings, pu11 boxes, & tray entrances.
ro . All rigid steel condu[te. that penetrate fire walls or pressure tight ' -
. boundriesaretobe-sealedasstatedgelow:
- a. If both ends of the conduit terminate further than 5 f t, from the wall, then only one end of the conduit needs to be scaled ,
anc the seal can be placed at either end.
- b. If one end of the conduit terminates further than 5 f t. from the wall, end the other end terminates less than 5 ft. from the wall, !
a seal should be placed in the conduit end which terminates less than 5 ft, from the wall. , c, if both ends of the conduit terminate less than 5 ft. from the wall, a I seal should be placed in both ends of the conduit unless the seal can be applied within :he portion of the conduit that penetrates :he wall. 7, Airtight and watertight floors, walls and ceilings are delineated on archi-tectural drawings D-176065 thru D-176072.
- 6. For alternate watertight seals see sheets 18C and 18D.
ORIGINA1. DRAWING, Rev. f - TRANSFERRED TO SCS SOUTHERN SERVICES INC. 7 BECHTEL CORP . JOB 7597 03/20
~~
m ALABAMA POWER CCMPANY 3 # R9. [g" g, 5 *l.s/n hv. Per PCN S- , , , , JOSEPH M. FARLEY NUCLEAR PLANT f , fje N"" ,,,y m TRAY &CONDUlT DETAILS & NOTES
}
m.0fM. ....
.mi ..US . ........
.cm - . ~
19 8 ,,2 ) .,m . A-17754I Q l
.g. _ ,. -C APPLY 'FLAMEMASTIC" 71 A TO FRONT AND BACK OF i TRAY AND CABLE ON TOP '
AND BOTTOM OF MARtutTE %L
~
CABLES SHALL BE COATED , 12 ABCVE $ BELOW THE MARINITE ._ " I"x '/8 FLAT IRON, SOTH SIDES
-S : TRAY OF TRAY, CINCH ANCHOR WITH f/4"-ao MAC H. SCR EWS . .
THERWlAL IN SULATING ' M WOOL iM VotDS-
- I " % , I$ 'n h
'- ' , r~t t
- c. ~.
< - - ', .p a e ,
,, % s.
+ ' ( l ll a
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g;;. m 4 "' .1
'.'" /2"MARINITE- XL : /H ,g khz,hl%-
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, % g
~ I f y}
lp/,, h- ' ' ~ ~g. '-$,
'~
9 h . h / " ~ ' - '
' / ,9 ;Y
, n- .
n ' ' FILL CRACKS WITH 5
- FLAMEM ASTIC'
- p .
[ ' an r% b 4 ^ Eg
" ~ ~-
- L.)"x I"X I/B" USIN6
[g Tl/T-20 MACH. SCREWS WITH C : ====
~ !
u- h NUT (WASHER AS REQUIRED Ner
!E NOTES
- 1. THis DETA\L 1:oR FIRE STOPS n
COVER SCREW HEAD, iA QR VERILCAL TRANS IGJ NO P_.Ost,.ER %EG.UIRED FOR WASHER p'FLAMEMASTlwN 1/8' THICH
/
J RPPEMDU R. 71 A MASTIC (TYPIC AL') _, 6 j 0056E1 x - _ -
" ALTERel6TE FIRE STOP rOR VERTICAL CAB _!E TRAYS --
BECHTEL CORP. JOB' 7597-03/20 SOUTHERN SERVICES INC.
== M A - - ALABAMA POWER COMPANY 5 5 h REIEfvED V010
, [g., PER'A', Bl%T'INf6k excT JOSEPH M. FARLEY NUCLEAR PLANT y 9/zMn REV.PER,?(.N B _ "" ' TRAY d CONDUlT DETAtLS ( NOTES 'ls 8%-f.037 F EV.0 )RE ,
A99..fY21 om .up uco , - .co, g zo o, n_,,,,,, A-I 7 7 5 41 6
~ . . _.m.. .
c~ \ l
.s-. , -lf'a-
. Je*-3 6 l-(. WALL ' '? c TRA'Y k. ' ~'. '
. q:.g. -
' . 1 J
L J /- ~ FLOOR i i, CLEAR MIN. f , , , s . 3 MAX. ,,
^ A,, .
A. ^ i . CONCRETE CURS
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- s. .
N
!!I gj 'JTo BE sh N agsf M NE 7
ei PROTECTIO fy - 1F REQ'D E i SECTION 'A-A'
- oosss/E\
VERTICAL FLOOR PENETRATION CURB BECHTEL CORP. JOB 7597-03 SOUTHERN SERVICES INC. [f .
,, gg,, n o.
1 our arvi ion 5-5 7/ REV.PER S.S.LTR. ALABAMA POWER COMPANY 2 2+1/ RE.V. AS MOTED sua m JOSEPH M. FARLEY NUCLEAR PLANT
. ]- '"fgGf,',[ TRAY elCONDUIT DETAILS 6 NOTES m ,c APP ..
. .gL_o,_22 .ncer. , A- l 7 7 5 41 2
' on....! . '. .uren.com. .eu. NONE _
4 2 6 (PROVIDED WITl4 ITEM i4 5)
.u.= W k
rk ELECTRICAL EQulPMENT [ MIM.FLEXtBLE CoRDulT LEM6TH5 - CouDulT 51Z.E HARDWARE MARK UuMBER lE
~
SIZE 0F FLEX ho_Tggy,M6N .Q, @ @ @ @ @
'4 4" 11" 2/4"'- H AL) HAld KBS KAQ HAB HSB I" 13" I" - HAP HBO KBC KAR HAC HBC IV4 " 14" IV #4 - H AG. HBP KBD KA5 HAD HBD 1V2 ' n" _
IVz "- H APs HBQ KBE KAT HAE HBE 2" ?_ l " 2" - H AS HBR KBF KAU HAF -HBF 2vz" 28" ZVt ". H AT ' HB5 KB6 KAV HA.6 HBG S" 34' 3" - H AU HBT KBH KAW HAH HBH l 4" 47_" 4" - H AW HBV KBI KAX HAJ HBJ TYPICAL FLEXIBLE CONNECTION TO TER9fP.esa HOUSING OF EQUIR SUBJECT TO VIBRATION . ORIGINAL DRAWING, Rev. I TRANSFERRED TO SCS BECHTEL CORP. JOB 7597-03 SOUTHERtl SERVICES INC. ALABAMA POWER COMPANY
- 4. "{9yg3 g Cy "c,[M3 " " - .,1 ,y ,,rer JOSEPH M. FARLEY NUCLEAR FLANT
/*
g ~ DETAM. TRAY 4 CONDUIT DETAILS 4 NOTES NWE A-17754 I
.-,.JMh3l.o..m... .m
.g at o, 29 . m. I
N y C.00 TROL COMDulT '
- AS REQ'D.
-POWER couDulT FOR .
208 volt MOTORS (FOR COODulT Sl'LE ( MARK TERMIMATE COMDulT - OuMBER SEE PLA9 DW6.) 05196 LocKuuT ( 1950 LATED BuSRIMG g a M SUPPORT coMDutT To UW15TRUT(P-ioco) 3 0 M I R O L_ j; 05106 P-to08 THRU P 'l0ZO ouE HOLE STATIOM couDulT OR EQUAL, li I t, am 1 FIELD WELD . uul5TRUT To' ll o,0* FACTORY BEVD STRAt6HT FLEX.CouRECTOR
'o
.. O!AUMELSTAMD-i FLEX C0900lT 1 i
- j. I 1i e e sc seE sa2o X 'I X ,
. Y f"'- motor i '
f-FIMI5HED FL0oR
" Ii toulP FouMDATI'09
-t
~
. ': * . *~.*. .... .
e . LOCAL ' CONTROL STATION 4' CdNDUlT SUPPORT- L)ETA l TYP FOR 208 V MOTORS 005664
~
ORIGINAL DRAWING, Rev- /
- TRANSFERRED TO SCS BECHTEL CORP. JOB 7597-03_ SOUTHERtl SERVICES INC. l
.. ZufD. .
*^{,'y Al ABAMA POWER COMPANY q, q
$F7kl t susarcy JOSEPH M. FARLEY NUCLEAR PLANT oma, TRAY (CONDUIT DETAILS ( NOTES aee.M. lll .j .
A-17754 I i. !
.c.u.. Ooot - ss 2 a_. o, 9e ... .
omaf 10 7.i6 .......
I
?
3'-4.l CHAMMEL WELDED To BASE ~y,~yj- f TYP'
~
- PLATE BY 3 HOP FABRKATOR e ._ _ew
-coMTRot couDulT 'c"
- TERM 10 ATE USING t.0(tpuT 4 IDSULATED At REQ'D, ',w .
FOR =* B05HluG At RE@. d f 4'(ABLE CHAMMELCHAMMEL(Q6 00MBERI - -O r
~ SEE PLAM DW6,) -@-
m L "x FoR /I 8 Vg"x 8"[h. COMTtot STATioM---- Qy M@ S.38 Vs) (HOLE. RE.oHEAD / il sesscTE'l ~ , m ye Jig"e $1)*/ SECTION "A-A' ' il (TYP. 4. PLACES) j(BATE. PLATE DETAIL) l
'o ) WTEr um Amon CABLE
-f a fI I r-
- ASLE. TE.RMIMAToft.
c . Il '
~
w p' MOTOR
. !! "A'
".A_",
<PiI7, '
y", i l," EQulP.F0000ATIOM k /FtMI54Ep FLOOR -
\ ,i . C' ' C'.
k
' [s- (.' .
{i.t i
~
NOTES I
~
. PIOOO(MBC)UDISTRUT WELDED TO(.HAMUEL
. STAMD. SUPPORT CABLE CHAOMEL To 005665 . ;
dMISTRUT ustMG (7.)WCA,2 LKA CHAMMEL . , lockout l
. I LOCAL CONTROL STATION 4 CABL E CHANNEL SUPPOR DETAIL-TYP FOR 4.l6 KV 4 600V MOTORS
~
' ORIGINAL DRAWING, Rev- /
TRANSFERRED TO SCS , SOUTHERN SERVICES INC. BECHTEL CORP. JOB'7597 03
"*- *^" ALABAMA POWER COMPANY
**;'{"
= DL ..
l 2fg=d/k i *+15 b5
- M D suurcr - JOSEPH M. FARLEY NUCLEAR PLANT m,,_ TRAY ( CONDUlT DETAILS ( NOTES )
. MM N,
.uu Mo0E . n& o, p . m. A-177541 1 i
. dl > . m .. m .
- 4. ,
1 ( (00LER d COOLER FAO MOTOR
/
g i x /
+ du. . _
j i
, I- L x
~#
(ABLE TERM 10ATOR-IMTERLOCKED ARMOR , 4" CABLE CHAMDEL (FOR CHAMMEL MARK UOMBER SEE PLAM DW(a.)- TYPICAL CONNECTION FOR COOLER MOTOR _
. . ~
Oossco
' . .. . ? ,
,,?
8 ..
.; ,:i s
.~
-I ORIGINAL DRAWING, Rev._ o g_
j] TRANSFERRED TO SCS : - BECHTEL CORP. JOB 7597-03 . SOUTHERtl SERVICES INC. _ [- 0 [ 15 2FOR RoVAL
,,,h,}, ; ,w,y JOSEPH M. FARLEY NUCLEAR PLANT om TRAY 4 CONDUIT DETAILS 4 NOTES m ..M h o l MoME .i 3 A % ,,_ m . m . A-17 7541 -
ur. . . [M(.u...... .ca. l A _ _ _ _ _ _ _ _ _ .
, i- .
.- Jy
~ 4" CABLE CHAMMEL(FOR CRAUDEL MARK UUMBER
=
COMTROL COMDut_t SEE PL Ak) DWh) FLEX % k = IMTERLOCKED N TERMIMATE CouDulT AT M.O.V,05196 LOCKL)UT ( ( ARMOR CABLE IMSULATED Bu51-1196 ( f Q r D r, (y,- , CABLE TERM 10ATOR
, - (
' MOTOR OPERATED
< , VALVE l l .
n, n . JW rrr1 rm trm rm N $ M M
" " " 00:i6
~ 0%
j ORIGINAL DRAWING, Rev- O ' ', TRANSFERRED TO SCS
~
.$ TkPICAL CONN. DF. TAIL FOR MOTOR OPERATED VALVE BECHTEL CORP. JOB 7597-03 SOUTHERN SERVICES INC.
"~ ~~ ALABAMA POWER COMPANY Y2[71Fo VAL.
,","f/ f l E'o ,
,,,, JOSEPH M. FARLEY NUCLEAR PLANT orrm TRAY 4 CONDUlT DETAILS 4-NOTES aee.hM.
A-LI7754I o
. n 4A 1 ou... 7...
.c,m_MoME . u c ., n . m.
_4 f-4"(ABLE
/ DETAll ou C.HAMOEL SUPPORT STAOD (FOR IMTERLCCKED
SEE SHEET tJo. ) ARMOR CABLE , i ] (00 TROL k SUMP PUMP I C00Dul" ;- rw s
' h f= 2 CABLE TERM 19ATOR s -
/
j ALTERMATORd dI w.
-) /
u e.) e.> i -Flul5H FLOOR
;: -ll
- l . a ,' .
. 4 ',
.d.
l ' 4 .f,
*i l
@.3=3 EE3 l l.
?
] 1 1
i I
, _L ;
i l; I i Nd9 1 i.
, ,ws/'
i-i
. I COMBINATION SUMP PUMP AND CONTROL .- j
^ STATION ' INSTALLATION 00:3668
~
0 ORIGINAL DRAWING, Rev. O TRANSFERRED TO SCS SOUTHERN SERVICES INC. BECHTEL CORP. JOB 7597-03 .
"~ ALABAMA POWER COMPANY
. ,[" ,, 0 4b7zFon ovAt.
ausarer JOSEPH M. FARLEY NUCLEAR PLANT
.; . ' A-2/6..
~
cic .
otra TRAY ( CONDUlT DETAILS d NOTE !
6
~
4,, ..h
. . p o , o_ a _ .,.. n . A-177541 - o
.ugpouE un .I 71 .u........
l NOTES: *L ;< l
- 1. FIE'L D TO CONNECT FROM, J-BOX TO SOLENOID VALVE AND LIMIT '
SWITCHES A5 PER C ONNECTION DIAC.R AM. '
- 2. FLEX \BLE CONDU\T NOT EXCEEDING 5 FEET IN LENGTH fAAY BE UST.D FKDNi1%E ALDVE T'/ PES OF DEN \tES TD C.ONNECT TD THE EAPOSED
, CONDUIT SYSTE tA OR .5UNCT\0N LOX. ;
- 5. I.N CASES INNOLNiMG RTDS' OR EACEPTIONS TO THE AE0% CRTTER\A,
- a TBE i\ ELD 3 BALL CONTACT ENG\MEER\MG TOR SPECIF\C AN AL'15\5 hND AP PROVAL. -
.. .. 4;.~;ALL TNTERNAL W \ KING 5%CULO EE IiOME TN ACCORDANCE WITH TME f.. AV?KOTKmTE CONNECT \ON DRAWiMG.
- 5. 70h.1VNCTION BOX AND RALEWAY INSTALLAT\0N RI.GU)REfAY,Jilb.
> ALENN)RONinENTALLN QUAL \T\ED VALVES, SEE PRAwiNG5fA-lM9 (5,H $ D -\ b) 3OO S H. l B.' - - - "
A ~ i
, a CONDUlT SIZE A5 TE RM. STRI SHOWN ON PLAN,
= -
E 4-50LENotD VAL.vE j 50LENOID' VALVE ~+{ l' T JUNCTION BOX. . j 7 r-#-CLQ-n : INST. AIR
' d[-i =
l
*-:-3/4" AND/OR /?.* RICwlD AND/OR
- FLE X CONDUlT TO SUIT BY b ~ LIMIT 5 WITCH FIELD NO NUMBER ASSIGiNED.
, PIPE ,,d ':.'gy'g g Q r'
FLExlBLE CONDulT NOT - CONT ROL VALVE TO E XCEED 5'
.[
TYPICAL CONDUlT CONNECTIONS TO ELECTRICAL
- 005669 DEVICES ON SOLENOIDgERATED PNEUMATIC CONTROL VALVES TO bti INSTAL _L ED' BY FIELD _. .
ORIGINAL DRAWING, Rev._ l TRANSFERRED TO SCS BECHTEL CORP. JOB 7597 03 50UTHERtl SERVICES INC.
"- *^" " " * * "
ond M uf81 AL ABAMA POWER COMPANY c c.:S /4-3 8 REN.PER PLN suwser JOSEPH M. FARLEY NUCLEAR PLANT
~
E .0- O. [M D:7AM TRAY (CONDUIT DETAILS ( NOTES DA SUMm.gDES SCALE NONE ,,,. 2 3 E o, 2 9 . ...,. A454 3
m L 9 f SOLENOID % 8 __,
/###### ,'
-9 - [ CND SIZE A5 SHOWN 0N PLAN
-//-{>_G-/f%I N5T A I R s
"j ~ 5 01 VLV
/ JUNC BOX
/
LIMIT
. MLIMIT SV I T CH -)6 L
T :#/###/#/:.
., PIPE %c:jjjjjjjjjjjjj: >
i .. RIGID AND/OR FLEX
^..- f X f, CND TO SUIT BY FLD NO NUMBER ASSIGNED j
'!E
;,'s.~; CONTROLVALVE
-9/4" AND/OR I/2" RIGID AND/OR FLEX CONDUIT TO SUIT BY FLD NO NUMBER ASSIGNED -
NOTES: . 6 .- -
- 1. NOTES ONE,T.HREE AND FIVEi.ON SHEET 29E APPLY TO' THIS SOLENOID JUNCTION-BOX ARRANGEMENT.
* +:
2.$ 5EE' MANUFACTURERS CONNE.CTION INFORMATION
'FOR CONNECTOR COUPLING AND CABLE PURCHASED VITH NAMCO LIMIT SWITCH 9.FOR FLEX CONDUIT CONNECTION AT LIMIT SUITCH SEE DETAIL NO.5 ON SHEET 29E-2
. TYPICAL CONDUIT CONNEfJION TO ELECTRICA L
. DEVICES ON 50LEN01D ' VALVES LISTED BELOU SV9109-A,5V9104-A.5V9991-8,5V9992-R,.5V9999-B, SV9994-B,5V9765-A & SV9766-A 005670 A .
0 9-23-H5 Bqdm R APVD PER PCN 0-8%-t-2744 pgA dA4 Z4 JJR X REV 0-19 DESCRIPTION APP R.1 APP R. 2 APPR. 3 APPR 4 APPR.5 REV. DA* BY C H K'O Southern Company Services, Inc. Foa ALABAMA POWER COMPANY "5" TRAY-& C0ND0IT DETAIL 5 AND ' ,,"',' [ ' ,,,,,,, ,,,,, ,,y , NOTES A-inwi nei -o ' BGT 7765 cwx o Jfys R
' hoRaws.
"" = - _ - _ _
__ _ _ _r ~;_ _
~
~
p
.' }
NOTE: 6 3 IF SPACE LIMITATIONS ARE NAMCO LIMIT ENCOUNTERED DUE TO THE 3 LENGTH OF CONN ASSEMBLY, SVITCH THE INSTALLATION OF !" MNPT TO I" FNPT 90*ELBOV MAY BE REQUIRED BETWEEN NAMCO LIMIT SV AND RCPT. ELBOV THREADS SHOULD DE SEALED PER INSTRUCTIONS IN MANUAL U265060, THE SECTION FOR RCPT ASSEMBLY
- TO LIMIT SVITCH.
\
t=scM
^
- CONN COUPLER (PURCHASED VITH
.c , .. ,
* ;..* .'~
LIMIT SVITCH) f, , ,. COUPLING l i : -FERRULE LOCK RING I" CND CPLG 5.5. FLEX CND M"'"" I"
$$$5A5#I 1
b h K') , 005671 ' vC
. DETAIL- NO. 5 l i APVD PER PCN B-84-1-2744 #44 DAB DECMA X O M3 85 BGT JM REV 0-13 _ _ _ _
APPR.2 APP R. 3 APPR.4 APPR. 5 DESCRIPTION APP R.1
"- REV. DATE BY l CHK'O Southern Company Services, Inc. FoR ALABAMA POWER COMPANY TRAY & CONDUIT DETAILS AND '""""'.D. PROJECT l DWQ . NO. SHEET REV.
NOTES A - ! 775t4 l 23E-2 0-
- .-- .. .....s... i
, e TO TRAY COMTROL - I.W ST RUM ENTATIO N !
COMDUlT - NUMBE.R IF REQ'D) :P T CouDulT N U MB E.R (4 SILE. P $ SIZ E. P E R. PL AN
. PER PLAW , .
t- - J
..I b 4 7 4~ WlRE.WAY x .
g (INSTRUMEN TATICM ) i i ! H 'k 4X4 ~ WlREWAY
. I MsTRU M E N T & ( COMTR.O L )
RAiK TYPICAL RACEWAY IMSTALLATION AT luSTRUMEMT RACKS OR GROUP OF INSTRUMENTS AT THE EQUIPMEMT id OT ES
- 1 IMD1VIDUAL INSTRUMEMTS ARE. LISTED AN D SMOWN OM
,. TH'i _ TR AY $ COM DU lT DRAWIN GS .
' -
- 2. FIELD SH ALL R.UM 3/2' RIGID d/OR FLEX CORDUiT FROM
. [ WI,REWAYS To E ACH INSTRUMEMT. TH EC & CONDUlTS SH ALL *
,- '" NOT BE. SHOWN OM THE. DRAWlMGS AN D TH EY SHAL L MOT BE
.. .',:. NUMBERED , THE CONDOIT LENGTH SHALL NOT EXCEED 15 FEE.T.
. 3.AL'INSTRUMEMT $)COMTROL DE. VICES (PRESS . S ARMS , E.TC.
S A M E.~TR Alk! D ESIG M ATIO N W I L L B E. FE D F Ro M . D I FF E.R E.NT WI R EW AYS , CONN.ECTE D BY SE.PA R AT E. COM Du t T S To THE. RESPECTlYE TR AYS . .
~
- 4. WH E M-' IM STR UM E. NT' 2.E.R MilhALS WILL ' M OT' ACC E PT GCHE.DULED ~ CAB LE.S , A .TE.RMIM A L SO)C MAY BE, ADDED WITHIN SIG HT AND NOT MORE. TH AM 10 FEE.T OF THE. .t NSTR.OMEMT . (U SE .ONLY QUALI FI E.D BORES I N SIDE. THE. COW TAIN MENT.) No BOK MUMBER SMALL BE ASS t G N E.D'.~ o n , ,.
ORIGINAL. DRAWING, Rev. '6"
. TRANSFERRED TO SCS
- BECHTEL CORP. JOB 7597 03 SOUTHERN SERVICES INC.
COMST ALABAMA POWER COMPANY O (SSUE F 4 JOSEPH M. FARLEY NUCLEAR PLANT 4 A .f sunner otra. TRAY 4 COMDU IT DETAILS d NOTES
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ELECTRICAL EQUIPMENT LOCATION CRITERIA , 1. Electrical T1oor Mounted Equipment will be dim.nsioned on Equipment Location Drawings for Electrical Floor Mounted Equipment. A 6-inch deviation will be permitted. .
- 2. In the absence of any specific locating dimensis s, the Field is to locate
. all wall mounted Electrical Equipment within the room boundaries as designated on the Electrical Tray and Conduit Drawings.
- - - _ . . . _ _ _ _ - .- ~ . ..~....- -.. ~ , ,,,. ,
[i. OESIGN CHANGES INYOLVING NEW OR RELOCATED EQUIP DIMENSION TOLERANCES FOR ALL LOCATING DIMENSIONS INCLUDED WITHIN
\ THE CHANGE. IN CASES WHERE THE FIELD DESIRES TO DEVI ATE BEYOND THE
'N DIMENSION AL TOLER ANCE 3 ENGINEERING APPROVAL SHOULD BE OBTAINED. j
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005673 ORIGINAL DRAWING, Rev. O TAANSFERRED TO SCS BECHTEL CORP. JOB 7597 - 0 3 /2 0 SOUTHERN SERVICES INC.
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. . . .h t,W... . ALABAMA POWER COMPANY -
152 NM O bA'M41EORP.? v .... ... ./w.. . ... ./j.& , M.w M ,u ac, JOSEPH M. FARLEY NUCLEAR PLANT
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$6[*[ g/p TRAY & CONUIT DETAILS & NOTES
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- 1) 'In order to prevent moisture leakage through the stranded wire conductors, modifi- )
cations are to be made to the Namco Limit Switches listed below: I l Valve Location Valve Location ; Number System or PJMfFg,,,, 'ai HV 3103* Pressurizer Sample Li s (SEE NOTE 10 ON SH g 23)-2) HV 3104* P.essurirer Saufple Lin (SEE NOTE 10 ON SH.23>)-2) , HV 3184* - CCW from RCP Thermal Barr SEE NOTE 110N SH.23)(2) ' HV 3376* CIMT Sump Discharge (SEE NOTE N SH. 23)-2 ) HV 3443* CCW from Let own Heat Exchanger ( 40TE 110N 23)2) HV 3765
- Reactor Lo Sample Line(SEE NOTE 10 0 - H.y ,
J-2) HV 3766* Accumul or Sample Isolation (SEE NOTE 10 ON SH.23J-2) HV 3196* CTMT turge solation (SEE NOTE 110N SH.23) 2) f HV 3197
- CTITPurgeIso tion (SEE NOTE 110N SH.Z3J 2)
HV 3999A* hteactorCavityCo ing System (SEE NOTE 110N SH.23J-2) 3V 3999B* Reactor Cavity Cooli System (SEE[OTE 110N SH.23J-2) HV 3179A
- St. Gen.1A BLwdn. Sompte n. e (S[E NOTE 110N SH.23J-2)
HV 3179B
- 1A Blwdn. Sanple Isjl (f SEE NOTE 110N SH.23J 2)
HV 3180A* St. Gen.
' 41wdn Samppi lBit(SEE NOTE 110N SH.23)*2 av 31 sob
- f s ' St. Gen.18 BD . Sorqxe Isin.( SEE NOTE 110N SH 23) 2)
HV 3181A* -
'St. Gen.1C Blwd Sopple Isin.( SEE NOTE 110N SH.23J 2)
'~
nV 31 B* St. Gen.1C B!wdn. Sompte Isin.( SEE NOTE 110N SH.23J-2) l HV 3101 Reactor Coolant ample Valve ( SEE NOTE 11 ON SH.23J-2)
' ample Valve ( SEE NOTE 110N SH.23J 2)
HV 3102* Reactor Coolan d g\ % N
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m m %. s N. ela c l9
$gE l
'E g< 005674 a
51 O av 3179C
- Stecu Generator 1A Blowdown (SEE NOTE 110N SH.23J-2)
VOID BECHTEL CORP. JOB'7597- 03/20 SOUTHERN SERVICES INC.
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.. bE10. twdu U IE wm ALABAMA POWER COMPANY bTk suasser JOSEPH M. FARLEY NUCLEAR PLANT
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VOIDF.D TEB.fD' Mgi us-2. es FEv.o.m /gf ui.T_ RAY & CONDUIT DETAILS & NOTES jfg N ..g 3 H _ ,, 29 ,,,,n. A-177541 Y
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- alve Locat%
Valve Location System or Pipinn
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[ 6 % W' Steam Generator IB Blowdown .WO%gSH.23J 2) Y.R r- Steam cenerator IC Blo@tivn(SEE NOTE 110N'5 23J 2) B J , 8047
- Pressurizer Relief Tank to Weste cessinq l System ( SEE NOTE 11 SH.23J 2) 8149A* themical an olume Control System etdownHed F.xchan Line (SEE NOTE 110N SH. 2 8149B*, Chemic' nd Volume Control System Letdo b ) Heat ,
E tang Line (SEE NOTE 11 ON SH.23)-2) !
,. t- i 8149C* Chemical and.1 me Control System Letdown Heat l Exchanger Line EE NOTE 110N SH.23J 2) i l
8871* Accumulator Tank Tes ine(SEE NOT 110N SH.23J-2) 7126* - Reactor Coolant Drain Ta , Heat Exchanger Line(SEE NOT SH.23J-2) p ,7, 252034* ' Pressurt: PSV (SEE NO 11 ON SH.23)-2)
.- 2 S2035 *- Presaurizer Y(SEE NOTE 11 ON SH.23) 2) ,
252035 * . Pres'surizer P V (SEE NOTE 11 ON SH.23J- 2) HV 33 Main Steam . HV 3368 . ' Main Steam
-* HV 3368C ',-
-Main Steam (
ain Steam [... HV 3369A
' . . _Maiti-s te am
;. .' HV 3369B '
HV.3369C Main Steam _
. h7 370A
. Main Steam .
HV 337 .. Main Steam ' s HV 3370C Main Steam ' HV.3976A Main Steam 'h RIGINA1, DRAWING. Dev - 1 I Main Steam a HV 3976B TRANSFERRED TO SCS
.HV 3976C Main Steam
. h 005G75 HV 3227A Auxiliary Feedwater .
$V3227B Auxiliary Feedwater V0ID HV 3227C Auxiliary Feedwater BECHTEL CORP. JOB 7597 - 0 3 /2 0 SOUTHERN SERVICES INC.
.. b tTA1 . ""'*g". hLABAMA POWER COMPANY
= - O w ene tes-s47 JOSEPH M. FARLEY NUCLEAR PLANT
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'4 .'. ' j Valve Location [ . ,, Valve Location
- System or Piping Ntmber _,_
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PV' 3371A . . Main Steam Atmospheric Relief Hain Steam Atmospheric Relief EV 3371B 006670 PV 3371C Hain Steam Atmospheric Relief ,. BECHTEL CORP. JOB 7597 03/20 SOUTHERN SERVICES INC.
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g"lf'j'[c,est; ALABAMA POWER COMPANY 0 JOSEPH M. FARLEY NUCLEAR PLANT
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.,23J-l c,r 2 9 w. A-177541
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- 2) For Namco Limit Switches with loss then 4 enductocs see Note 1 on Sheet 23M.
- 3) For Namco Limit Switches with 4 conductors see Sheet 23R& 23X.
- 4) herNamcoLimitSwitcheswithmorethan4conductorsseeSheet23M.
4
- 6) "or the Bill of Materials requited to make these modificatiens see Sheet 23P.
~
- 7) For sealing the conductor on solenoid valva junction box end, see Sheet 23L.
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_..__m g _ ._ - 1 B)f For procedure on how to install Raychem breakout kit on Limit Switches having conductor other than J02. see Sheet 23U-1. m =~a". t?'"* u N
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. . 005677-ORIGINAL DRAWING, Rev- [ __
TRANSFERRED TO SCS e a ^ BECHTEL CORP. JOB 7597 03/20
- SOUTHERu SERVICES INC.
$Ic. oust. ALABAMA POWER COMPANY o [); issu[
JOSEPH M. FARLEY NUCLEAR PLANT l .
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*"*' TRAY & CONDUIT DETAILS & NOTES aMed
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j ( .. i METHOD OF SEALING COAX' CONNECT 6R5 FOR CABLE $CHEME N0'.5-2VAIS0llE,2VAI90llG,2VBIS009E & 2VBIS009G -AT THE ELECTRICAL ! PENETRATION INSIDE CTMT. C FOR 02D21RE0027A A & RE00278-B >
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h l gMOUr, G PANEL 3 WCSF 500 < i , WCSF-200 - SHRINK SLEEVE (6") C,t
/~ COAX JAC g ... ;;,..
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l L-WCSF '400 LWCSF-T0g L-COAX. ' CAB.L' % lvCSF-200 4 4 EAT SilRINK .-
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SLEEV,E C6") 3 ,, y
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005685 VOID
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JMR VOI.DED PER E5 BB It32. AAke - B-14-8i RTP vW ud M i
~0 4-/5-68 RWK -JMR APPV.,PER PCN B-87 2-4684 REV.O. DEKiJec.
CHK'D DESCRIPTION APPR.1 APPR,2 APP R. 3* k. PPM,4 l APPR.8 REV. DATE SY. Southern Company Services, inc. rDa- Alabama Power Company J.M. FARLEY- NUCLEAR PLANT " " " " ' ' * ~~ ' **'" " "' TRAY & CONDUIT DETAILS & NOTES ~ - oMAww . RWK - Tyrfo cHno JMR .
- - -n - . . . . - . . . --. ._ , - _ _ _
., OTY, DESCRIPhlON 80 , ,RAYCHEM INSTALLATION KITS
, NO. NCBK-04-04 WHICH INCLUDES CABLE BREAKOUT NO. 502AB23-52/144, CUAllFIED FOR CLASS lE USE.
1 i i 1 005686 ORIGIN AL DRAWING, Rev.- O TRANSFERRED TO SCS
~~
BECHTEL CORP. JOB 7597 03/20 SOUTHERN SERVICES INC.
" '@~ I o 5$z ALABAMA POWER COMPANY Pg. Si IEoast.k
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. ,,, JOSEPH M. FARLEY NUCLEAR PLANT D .m,u TRAY & CONDUIT DETAILS & NOTES 1 wit &. .,23S of.2958. A-177541
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*N , - ). PROCEDURE NR APPLYING CHICO-A4 COMP 0tND TO THE N1PPLE OF NAMCO's EA-180 L1 HIT SWITCH WHERE RAYCHEM BREAK-0UT IS ATTACHED ntering from A. For switches which are vertically mounted w '
m., the bottom:
- 1. Remove the six screws holding the~ .ront cover plate to th switch being careful not'to loose the 0 .~ngs and Bellville washe.. u'nder '
each screw. Carefully remove the cover plate and gaske.
)
2. The following eps describe one method for applying the r .i " to the switch seal. The field has t$ opti of using other methods if desired. However,
- is important that no more than 1\ oz. of C 0 is applied to
- each switch, and that no Ch 0 finds its way to the switch internals. ! ,
- 3. Attach a short length of tygon tubing to t i end f a veterinarian type syringe which has been calibrated in e t . ounces (oz.) or.
d cubic centi .ers . ,. 4 Mix a s .all quantity of CHIL A4 using m ing instructions printed . on th< centainer label. I S. Using syringe with tygon tubing a tached, withdraw 2-3 oz. (35-50cc)
;e syringe.
ofthQ11guidCHICOmixtureinto
- 6. Through he open side of the swice I, carefully 'nsert the free nd of th tygon tubing into the .pe nipple attached to the
. s itch, unt. it bottoms on the F ychem breakout seal.
Insure tt. t the CHIC mixture does no et on the switch internals. ,
- 7. Inji. t 1\ oz. into .
. c nipple, then carefully withdraw tubing.
- 8. Dispos
.h syringe nd tubing thoroughly.
of the remainder of the CHICO in the syringe, and rinse
- 9. te ;ce swi -h cover plate usini; gasket, screws, 0-rings ap05687 Bell'r 4 was rs previously removed.
ORIGINAL DRAWING, Rev O
' TRANSFERRED TO SCS
]
SOUTHERN SERVICES INC. BECHTEL CORP. JOB 7597 03/20
"@ ~' * ' "
isbuUo on ALABAMA POWER COMPANY
=- O const. twcon9.-
Jh- sta e, st-itB4 3 sumaser JOSEPH N. FARLEY NUCLEAR PLANT
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, co gy s-su-4sas aou M"' TRAY & CONDUIT DETAILS &
esmud,N'""L7N . 235-I or 29 A-177541y
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B. For switches which are noti vertically mou ed or have wirino entering irom ths' top.
' '. Remov'e switch h om its mounting, leav g flex conduit and se undisturbed. ,,
- 2. Support switch in a Verti, cal p ition with w ng entering froc.
the bottom.-
- 3. Perform steps A.1 throu 4 9 aboviW.' . .
-4. Allow a minimum of s hou for CHICO ot'r cure,before mo%g switch.
I
- 5. After CHICO has red reinstall itch in its' original po . tion. j
. j
- \
NOTE I The work Lt time for CHICO is,about . ; min -.. n. k required prior to a ,nving CHICO s. 'd be performed
,efore mixing.
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005688 ORIGINAL. DRAWING, Rev. O y h TRANSFERRED TO SCS BECHTEL CORP. JOB 7597 03/20 . SOUTHERN SERVICES INC.
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I
,. e' TERMINATION METHOD FOR DEVICES SUBJECT l TO .HIGH EMPERATURE 1 I i
l
.s l EXISTIN G -
l CONDUIT i l 5' OF 3/4" FLEX FIELD TO LOCATE
=
WITH TWO AND INSTALL JUNCTION CONDUCTORS OF BOX FROM FIELD CABLE DESCRIBED STOCK AND SPLICE BELON . 25
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= EXISTING DEVICE RECOMMENDED CABLE:ROCKTHERM TYPE SR-ML OR -
EQUAL RATED FOR 600V,1eo C. THE CONDUCTOR Size SHOULD MATCH THE EXISTING CABLE. 005689 ORIGINAL DRAWING, Rev. O TRANSFERRED TO SCG , BECHTEL CORP. JOB 7597 03/20 SOUTHERN SERVICES INC. ca BjD ' ALABAMA POWER COMPANY "g/d; .,.,,[,[ { g33 3 *['* ',,g 5
$9seTn.Pl!i!!ML JOSEPH M. FARLEY NUCLEAR PLANT
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om,uTRAY & CONDUli DETAILS & NOTFC; p.gJ.-......
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__NAMCO LIMIT SWITCH ' N6TE_: THIS DETAIL IS APPLt TO STEfA MOUN7EL Li SWITCHES ON M0/880 , IN. C. ONLY. ,
'3 r ~
3 ~ I" NLPPLE-(NIPRLE FI'LL ~~ l WITH APPR0x. ipr. 01..
- f. . 2HICo. A4_ SEALING MPOUND) .
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~ RAYCHEM (502AB 52/l44 Q*) twt IBREAMOULFlTTff6 OVE ScoNDUlT' NIPPLE. $ I/C'*lE (JDZ) l"_..,
CABLE 10 BE CUT 70 SunT . COMPRESS) (i EXISTING Cot 4DUCTOR SMLLP -sototn_,- OfGU CTONTG .- 1
~
%. m SWITCH I
w w:- ' z.: fij . 3b7,, f g y m y.m sueve _ . ccMPp.ESStoN LUGi i __ (TYPICAL) __ a cuir- _ . DkTAl_ A NOTE: A CONOULET IS TO BE
'lNSTALLED IN EACH RICalD ._
n %~N-CONDUlT RUN TO HOUSE
.THIS CONNE.CTIOM.
-j i - SEE DE IL*A" ._
1 (Typical on ,
' I "lL COM CTO R)
L_ _J ME" ; , HEN LE.SS THAN 4 CONOUCTORS ARE USED PER v-" CABC. PORT A $HoRT PIECE OF Jo2 CABLE. (LENCtTH SUI IS TO BE INSERTED IN THE SPARE LECab OF THE. BREAKOUT AND HEAT SHRIMK OVER TERMil4AL . ._ . LUC 3 BARREL AS PER DE' TAIL. 4 2 3 SHEET. 2340%S0 ' ORIGINAL DRAWING, Rey. O TRANSFERRED TO SCS BECHTEL CORP. JOB 7597 03/20 SOUTHERN SERVICES INC.
"*' *= **"*'*a J
- c. njb... ALABAMA POWER COMPANY .
,jg ,_ O h/. Ig. h$d$p g g.sg.!s43 susue, JOSEPH M. FARLEY NUCLEAR PLANT l 2 Wa VUI.DED YER PCN ,DJ1^ TRAY &CONEUIT DETAILS & NOTES
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PROCEDURE FOR INSTALLING RAYCHEM CABLE BREAKOUT F1TTlNGS TO STEM MOUNTED LIMIT SWITCHES ON MOV8806A.
.. MOV8 808 B & MOV8 808C _._._
2 1.) Remove the existing conductors from each li .c switch and lug each co 'uctor. *
*2.) Terminate small lengths of J02 . cable from each imit switch and install .
-' raychem CB4-2 cable breakout fitting as per Drawl ' A-177541/Sht. 23K& 23
. . . , ,' . , ; !!OTr.
;.JWhen less than 4 conductors are us ..per cable breakout, a hort piece of.
- 02
',. cable -(length to suit) is to be . tced in the. spare legs o. *he cabic breakout and heat shrink'over t mins lug barrel as shown on u Ning.
A-177541/Sht. 23L. - 3.) Install a ring tongua terminal luglon each .>s' conductor and solder at, shown ' on Drawing A-177541/Sht. 23U. .
. 4.) Bolt the existing conductor. and J02 cabit. ring tongus *erminal .ugs together and apply a WCSF.-N sleeve over the J02 conductor,and heat s.. 'nk
^
- a shown on Drawing A-177541/She . , .
5.) ApplyaWCSF-f'sleeveuertheentire onnection with r ,,rox. 1" of excess sleeve on eac ide of the connection at. heat shrink
, 6.)- A condulet is to be installed in each affec d rigid conduit run to house this con eptit . .
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.S1 VOID SOUTHERN SERVICES INC.
BECHTEL CORP. JOB 7597 05/20
** ALABAMA POWER COMPANY issut c Hsr4 L O JOSEPH M. FARLEY NUCLEAR PLANT surner ca OL- $.h"$.l'ek2-g,EmTRAY4CONDUlT DETAILSS NOTES tt)y 3 vivsi MDB M KN Y
cC.M W . . ......!* * "' N."N .23Ul A-17754I
,. ,. I INSULATION OF BOLTED CONNECTION -
SHIM SLEEYE - - - SPLICE SEALIN SbEVE BOL AD b m % = ..... : m m (, w ::. .:::::= = : "'-Q,_ &" ' . " ' * * * '
- 2'NIN 3' NORM .
/
NOTES: L USE RAYCHEM TY E WCSF-N TUBI G ONLY. /
- 2. INSTRUCTION APPLICATI0)[SUPPLIED A55H WITH L BE HE INKIT. ACCORD NCE WpTH M 3.ONLY A QUALIffED T SHALL BE U ED FOR CLA55 IE APPLICA-
+
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LOW VOLTAGE (1000V. OR LESS)IN-LINE SPLICE BOLTED CONNECTION FOR NUCLEAR (LOCA) QUALIFIED APPLICATIONS 005692 (CLASSlE) YOID I T D Ef-66 RWK JMR voiDEO PER PCN 6-8b 0-3b58 REV.0 /M,,e O 9 - 1 F2- 66 JLO JMR APPVD. DRAWN PER S-8613730REVO MC.A r46s 4WCd/d(L k REV. DATE BY CH K'D DESCRIPTION APPR.1 APP R. 2 APP R. 3 JAPP R. 4 APP R. $ Southern Company Services, Inc. roa Alabama Power Company J.M.FARLEY NUCLEAR PLANT l,",','",","*"' ,,, , ,o . ,,,,, , , y. TRAY & CONDUIT DETAILS & NOTES 23U2 DRAWNdLQ TYPED CHK'D JM R A-1775 Lt l l
RACEWAY TERMINATION DETAIL @ PENETRATION PROTECTION CABlNET (TYPlCAL OF ALL PLA ES) i ALL PHASES PER CABLE NOTES: 1. BLIND END PLATE (FOR CABLE TRAY) DRILLED 70 ACCEPT BO'A CONNECTOR END OF CABLE TE.RMillATOR.
- 2. FOR LARGE N2- OF CABLES, A LARGER GALV. PLATE MAY BE SUPPLIED BY THE FIELD AS REQulRED, 9
*-CONDu\T NOT CONNECTED h TO CABINET TRAY NOT CONNECTED TO CAB.
g . ARMORED CABLE i ,,,,, TERMINATOR
\ g g
e iE 4 ALL PvtASES BAMDED y.[ __E;;__ TOGETHER + TRA)HED
.m'To POWER CABLE C- .ELECTR) CAL PEuET.
e> qq uing PROTECTION CAB 1 NET c 7 1 .
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i j i OABLE TIE' i > g E ' I '. 005690 i y , E 1 1 2 1 5 - y BECHTEL CORP. JOB 7597 03/20 SOUTHERN SERVICES INC.
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,g, ALABAMA POWER COMPANY
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v m t PLUG INTERIOR WITH THIS TYPE OF
.- SILICON FOAM CONDUIT REQUIRES l CABLE DROP NO INTERNAL
,, SEAL
['(WHERE APPLICA5LE)
/2 SILICON FOAM i
^ '
BE AD OF R.T.V.
"",, OR CAULK I
[ I 2- 2 /2" SILICON F0AM \
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w Mygyg.
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.. I EQUIPME NT ..
. LxAOw00t ..
PACKING TO ., PROVIDE DAM IF VENT OPENINGS OR FANS PRECLUDE (MAY BE COATED SEALING TOP 0F WITH VIM ASCO -
+ OR EQUAL) +
EQUIP, A RAIN SHIELD MAY BE SUBSTITUTFD FOR EQUIPMENT C ABLE SEALING THE OPENING OPENINGS. 4 SEALING BEAD NOT REQUIRED IF WATERTIGHT CONDUIT HUBS ARE - , USED SPRINKLER PROTECTION OF ELECTRICAL EQUIPMENT 005694 NOTE: THIS. DETAIL MAY BE USED ANYWHERE ELECTRICAL EQUIPMENT IS SUBJECT TO DAM AGE FROM SPRIN';8.ER r>YSTEM OPEG4 TION, _ , _ , , . 0 5-2 88 RJP JMR APVD.D.R AWN PER PCN B-852- MN c/td OMD[#N%T 38 37 REV.0-12 l-REV. DATE BY CH K'D D ESCRIPTION l APPR.1 AP9R.2 APPR.3 APPR. 4 l MPR.i So, thern Company Services, inc. FDR _ qabama Powerfo.m paw J.M. FARLEY NUCLEAR TL ANT. 'Q'j,Q "'"]Y Dwo. so. sHtEr Rev. TR AY & CONDUIT. DETAILS & NOTES O ~
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POWER RECEPTACLE INSTALLATION ON %2 RAIL
- ORIGINA1. DRAWING, Rev-- /-
TRANSFERRED TO SCS
"' SOUTHERN SERVICES INC.
BECHTEL CORP. JOB 7597-03 557/ REV . P E.1 .S.LiR, 1 , ev.h.2S susarcr _ JOSEPH M. FARLEY NUCLEAR PLANT-xy,c TRAY & CONDUIT DETAIL $ NOTES APP .
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'AE.TER 'CASCE INSTALLATION x _
ELEVATION CouDulT To EXTEUD ' 4' BEYOMD WALL WITH TH READS. UNLESS - # OTHERWISE MOTED i SEC.ONDARY SHIELD WALL
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ORIGINAL DRAWING, Rev.d-TRANSFERRED TO SCS BECHTEL CORP. JOB 7597-03 SOUTHERN SERVICES INC.
* ~ '
557/ REV.P .S:LTR ALABAMA POWER COMPANY [",f/h,/t 7 2. MW2 RtV AS NoTED suanc7 JOSEPH M. FARLEY NUCLEAR PLANT orra TRAY d CONDUlT DETAILS $ NOTES APP. . .A. e d. ... /.?L.u,....... .eu. . .; n .. , s . ..,. A- 17754l 2
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J<g . CONDUlT & EQUIPMENT
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1 BECHTEL CORP. JOB 7597 "' 03
.. M.A .. ..
" * - *^ ' * " "
ALABAMA POWER COMPANY 1 5f7! REV. P ER S.S.LTR
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8' 2(THIS DETAIL) FIRE STOP BARRIER FOR VERTICAL REDUNDANT SYS. BECHTEL CORP. JOB 7597 03 SOUTHEPN SERVICES INC.
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tO' O' y g MOTE 'O A'20'-O' MIMIMUM VERTICAL 'b N AMD HORIZOMTAL SEPERATIOD 15 REQUIRED BETWEE.M TRAYS 7gAig A e OF REDUMDAUT SYSTEMS ~
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- to' O" SEPERATiou CANMOT '*; TRAluA BE' M AINTAINED. .;., ,;jd liii ;j '; + .
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., . _ . s-MINIMUM SEPERATION BETWEEN TRAYS ffd&h XCCHANNELS)IN A MISSLE ZONE ORIGINAL DRAWING, Roy. L TRANSFERRED TO SCS SOUTHERN SERVICES INC.
~
BECHTEL CORP. JOB 7597 03
"" N -- c. " m ALABAMA POWER COMPANY
*,% ",'.[ 'Z"~ ~ I MiB A5 NOTE.D "O 5 b"w $
- SUBJECT - -
l#'4 Ne*** o,7a TRAY & CONDUlT DETAILS ( NOTES Jf
.. 401.~.J....... .m. woes .a , * ...n . A- 177 54? 2 c .
l . \ t" OR TWKE. THE DIAMETER
.0F THE LAR6EST CABLE i
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~ -
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SAME TRAIM SYSTELA i l Q OMt_Y. ; o , i TWICE THE DIAMETER i 2 OF THE LARGEST CABLE
$ 1 MINIMUM DISTANCE OF TRAYS BELOW j l INTERFERENCES WITHOUT BARRIERS l
- i I
.! 005702 I - -
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TRANSFERRED TO SCS ._. BECHTEL CORP. JOB 7b97-03 SOUTHERN SERVICES INC. ALABAMA POWER COMPANY [ O ( .<. Fon VAw
.$ x Q.2 ,["~.J 2 Mi A5 NOTED suancT JOSEPH M. FARLEY NUCLEAR PLANT
$.$ "yo$ng"o ""* ocre. TRAY 4 CONDUlT DETAILS ( NOTES
\ .auf Ill.. .un...... .eu. MODE 6 V A . n _....,. A- 177 5 41 3
m . . _ - - _ _ -._ . _ . . _ _ . - _ - 4 3 b o" TO BE DETERMINED TO BE DETERMtNED G MIN.n i
- IN FIELD TO SU IT -
- IN FIELD TO SUIT ~ ~
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REQulRED WHEM SiO"RORIIobTAL
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SEPERATIOM CAlJMOT BE
. MAlldTAlfJE.D BETWEEM E
7 Y TRAYS OF REDUUDAMT, O, SYSTEMS WHEN 5-REDUMDAMT SYSTEMS SEPARATION . CANNOT b ' BE MAINTAINED. wwHEtLBARRtERS ARE REQ'D. [SEE I.l?a SHEET M 1 4lW:
'500D ALUMINUM COVER. -
.TRA10 A .
4 . MINIMUM SEPERATION BETWEEN REDUNDANT TRAYS AND CHANNELS WITH i REQUIRED BARRIERS IN CABLE SPREA h: . _ ROOM , 6 ORIGINAL. ORAWING, Roy - TRANSFERRED TO SCS , SOUTHERN SERVICES INC. BECHTEL CORP. JOB 7597 03/2 0
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~
INSTRUMENTATION TRAYS AND POWER AND , . CONTROL T. RAYS _ 005704 n ,.; ,, .a ..
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~dRIGINAL DRAWING, Rev. I '
/ f h',j TRANSFERRED TO SCS , ';
(i :i g g ..< ( . BECHTEL CORP. JOB 7597-03 SOUTHERN SERVICES INC.
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NOTE ' 4 4 ijg ggig,g A 15"MIM. CLEAR SEPARATiciJ BARRIER 15 REQUIRED BETWEEM o> p REDUNDAMT TRAYS AT A- 2C ' (Ross-0VER. A '/z'MARINITE . BARRIER 15 ALSO REQUIRED ~~ e
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4 005705
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MINIMUM SEPARATION BETWEEN TRAYS ' AND .) f6IANNELSlAT A CROSS-OVER !N AN g _- ELECTRICAL FIRE ZONE WITH REQUIREO-ORIGINAL. DRAWING, Rev- / BARRIERS. TRANSFERRED TO SCS BECHTEL CORP. JOB 7597-03. SOUTHERil SERVICES INC.
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~
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ORIGINAL DRAWING, Rev. i l . '1 " .. . TRANSFERRED TO SCS l !J
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O OVAL ALABAMA POWER COMPANY J FOR AP h - p7
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TO BE DETERMINED I - 0,, IN FIELD TO sulT ( . TO- BE DETERMINED : IN FIELD TO sulT L 12" 12" a = - - 1 TR AIN A TRAIN B N ~
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PR AcinCAL FROM ALL REcuNDAMT SAFEGUARD TRA'1 OR CHAMMEL 2.REDUNDAWT SAFEGUARD CONOulT SHALL DE INST ALLED
- TO MA)WTAIN A MINIMUM OF 1" SEPARATION, S.I.M UL11T "'2. ALL SUBSEQUENT SAFEGUARD COMDulT SHALL'6E
- GALVAN 11EO RIGID STEEL, ORIGINAL DRAWING, Rev..- ,/ .
TRANSFERRED TO SCS . V h V U VI S EPA 3AT Ol\ 3 ETW EEN~ R E JUN JA\'~ . TRAYS, CHANNELS AND CO\l]UITS BECHTEL CORP. JOB 7597-03/ 2 0 SOUTHERN SERVICES INC. ao. oa u . sa ass g ,,, , , ALABAMA POWER COMPANY
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, BECHTEL CORP. JOB 7597-03 SOUTHERN SERVICES INC.
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8Q RAYCHEM INSTALLATION KITS NO. NCBK-04-04 WHICH INCLUDES CABLE BREAKOUT NO. 502A823-52/144, OUALIFIED FOR CLASS lE USE, 005710 Q jg APPROVAL f -i f $. oars I j"h*n5ur.r.7jd ( i 1 1 ; 1
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1 O $,, gogg,;gg,%.unner JOSEPH M. FARLEY NUCLEAR PLANT 3 3/4#1 RE'i FER NN'S 6-g/jf B+2-4Sf6 EV.0-lot PFZ" TRAY & CONDUIT DETAILS & NO U.x.-nd.[ ......y o2E m o t 7 N , 2 3 K ,2 ') . m. A-177541 T _ . _ _ . . _ _ _ __~
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y. BILL OF MATERIAL QUANTITY DESCRIPTION "AYCHEM CAT # (502A823-52/144, NCBK-04-04) i 250 able breakout kits with installation instructiens, qualified for Class 1E use ) inside the containment. j 3 - l 1 l 110 kits are required for Unit 1. I Remaining 140 kits for Unit 2 including spares will be delivered , in'6 to 8 weeks after placement of j order. 200 RAYCHEM WCSF-115-6-N, cable sleeves, ' with installation instructions, qualified for Class 1E use inside the containment. _ j These kits a.re required.to replace field' stock which is being used for .
.the Unit 1 modification.- .
Determined RTV Primer, G.E. Cat. No. SS-4004. by Field . Determined G.E. Silicone , Cat. No RTV-60. by Field 3 RAYCHEM Heat Shrink Tubing WCSF-070-6-N. Determined C rt.c ' w .ds sealing compound CHICO A4s by Field 005682 l ORIGINAL DRAWING, Rev. L TRANSFERRED TO SCS BECHTEL CORP. JOB 7597 - 0 3 /20 , SOUTHERN SERVICES INC.
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AlabamaPower Or@: D ._ DOCAt iRG A M nViCl mesourhemsect caysem ORfNCH October 11, 1978
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Dockei: ho'.' 50-348
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l66 79418 PW/ .-. ' U.f S. t.'uclear Regulatory Co= mission ff - Region 11 Director. tif-N'
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101 Marietta Street, N.W. 7 Suite 3100 // g', ' Atlanta, Georgia ATTH: Mr. James. P. O'Reilly
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JOSEPH M. TARLEY NUCLEAR PLANT-1P.F NO.M QUALIFICATION OF ELECTRICAL TEPJIINAi, BLOCK'
. r],_. Gentlemen: . [~ ~ ' . ..
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~'
As requested in the NRC IE Circular 78-08, a study of environmental
~
qualifications of, electrical equipment in the Tarley-Unit I has been con-ducted. The study revealed that a full test report was not available for certain terminal blocks-inside the containment ; ssociated with containment isolation valves, vide-range and narrow-range temperature monitoring, and power-operated relief valves. The terminal blocks are Type ZWM, Series M-25000, manufactured by States Company, a division of Multi-Amp Corporation. They are twelve, fifteen and sixteen pole front connected terminal blocks with white marker strips. -The blocks are mounted inside terminal boxes, one terminal block per terminal box.
.The molded base is fabricated from general purpose Durey #791, a brand name of Fooker Chemical and Plastics Corporation for a two-stage asbestos-free phenolic. The barrier strip is fabricated from Allied Chemical Grade FP-10SO polypropylene. Both materials are highly reaistant to beat and moisture..
The following is a summary of available information pertaining to the environmental qualificatirn of the terminal blocks: -
- (a) States Company . terminal blocks were used on the motor control
' center which was tested for environmantal conditions simulating the effects
- of a high energy line break in E. I. Hatch Nuclear Plant, The terminal blocks were tested to a maximum of 2500F, 29 psis, and 100 percent humidity vith no detrimental effects.
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4 'O U. S. Nuclear Regulatory Co= mission 1J- page 2 October 11, 1978 0062723 (b) Test results compiled by Battelle Memorial Institute ir.ficate that the wood flour-filled phenol formaldchyde retained useful properties beyond , the exposure dose of 8.3 X 109 RAD. This material also showed good radia-tion staM11ty above 5000F. - * (c) polypropylene has been tested bv Epstein ok k'estinghouse and the data compiled by Battelle. It was reported that there is a slight increar.e in stiffness and some loss of impact strenEth up to 107 FAD. At that dose, the changes in electrical proper:ies are minimal. (d) The States Company teminal blocks with enclosure were exposed to 1 X 107 RADS (at Georgia Tech) with no significant. change to their resistance or dielectric strength, The above noted tests indicate that the States teminal blocks are very high quality products which should easily teit$ stand the Farley post-accident environments. However, Alabama Nser Compr..; has initiated a testing pro-gram to further verify the perfernance of the blocks in the prest-accident environment. This program vill be co=pleted by the end of the first refuel-ing outage. i L/ The following justification summarizes the detailed investigation of I the safety-related equipment associated with these terminal blocks (a) Fif ty-six terminal blocks are used to connect field viring to valves which are used for the purpose of containment isolatD n. The majority of these valves are. closed during normal operation and remain closed during pest-accident conditions. All containment isolation valves close on loss
. of power and are not required to reopen following containment isolation signal. "Ihe power interruption required to effect a post-accident valve closure vill occur even if the terminal block f ailed.
(b) Two terminal blocks serve the pressurizer power-operated relief l valves which are nomally closed. These valveu close' on loss of power and are not required to open after an accident. (c) Twelve terminal blocks are used in the RCS temperature vide-range monitoring and narrev-range trip circuits. The RCS temperature T c/Th wide-range monitoring is only required for monitoring emergency plant cooldown , rate during later stages of plant life. The RCS temperature Te/Th narrov- ' range instrumentation only provides a trip function for a narrow class of Jov probability steam line break accidents (0.3 sq. f t. to 0.7 sq. f t. break ' area). This trip function occurs within the first 100 seconds of the acci-
. dent.
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- U. S. Nuclear Regulatory Connission Page 3 -
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October 11.-1978 This subject was reviewed by the Plant Operations Review Comittee and the Nuclear Operations Review Board. Based on the above they determined that thit Iarley Nuclear Plant can continue operation without af fecting the health and safety of the public. ,, ; Yours truly, e h fD&[:h%f T. L. Clayte , Jr. ILCJr:bhj cc: Mr. R, A. Thott.aa - Mr. G. F. Trowbridge be: Mr. J. T. Young Mr. H. C Thrast. - Mr. O. *a. Kingsley Mr. W. C, Hairston Mr. T. N. Epps
- Mr. C. Biddinger Mr. Ozen Batum Mr. D. M. Crowe Mr. A. A. Vizzi Mr.-H. A. Sindt Mr. R. S. Longdon '
Mr. R. F. Mcdonald / e 9 O
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- : Alabama Power Company Atta: Mr. F. L. Clayton, Jr.
4f lh[ , i Senior Vice President Post Office Box 2641 -O[ h' Birmingham, Alabama 35291
,i [ f Centlement Enclosed is IE Eulletin 79-01. It requires action by you with regard to power reactor facilities with an operating license except for the 11 SEP plants which are listed in Enclosure 3.
This Bulletin is also being sent for information to the 11 SEP plants and
-jf] all power reactor facilities with a construction permit. No action or
( V
/ written respense is required is- construction permit facilities or the il SEP plants.
Should you have quest, ions regarding this Bulletin or the actions required of you, pleese contact this office.
' Sincerely, A
c__N 6 %.- James P. O'Reilly
^ 9 1 rector
Enclosures:
IE Bulletin No. 79-01
- 1. ,
- 2. ist of IE Bulletics
^
Irsued in the past i 12 months
- 3. List of SEP Plants (11) . :. 3. T. ' i V i ... S
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rit i 879 1 4 O Inclosed is IE Bulletin 79-01. It requires action by you with regard to
. Power reactor facilities with an operating license except for the 11 SEP plants which are listed in Enclosure 3.
This Bulletin is also being sent-for information to the 11 SEP plants and all power reactor facilities with a construction permit. No action or written response is required for construction permit facilities or the 11 SEP plants. ' Should you have questions regarding this Bulletin or the actions required of you, please contact this office. Sincerely. c >\is.c
^. James P. O'Reilly
< Director
Enclosures:
- 1. IE Bulletin No. 79-01
- 2. List of IE' Bulletins Issued in the past -
12 months
- 3. List of SEP Plants (11) 0
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irN (' a-FE8 8 579 1 J . . kl' Alabama Power Company- . 0051373 cc w/ encl: Mr. W G. Bairston, III m Plant Manager Drawer 470.
- 3 Ashford, Alabama 36312
% Mr. R. E. Rollands, Jr.
V .-QA Supervisor-P. O. Box.V Ashford, Alabama 36312~ Hr. A. R. Barton ExecutivelVice Presioent-Post Office Box 2641
. Birmingham, Alabama 35291
- r. d Thrash, Manager-
. Nuclear Generation Post Office-!3ex 2641 '
Birmingham, Alabama
. ,C Mr. C. Biddinger, Jr., Manager
( Corporate Quality Assurance
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_ qO. . . ' . 005;37, ; UNITED STATES NUCLEAR RRCLTLATORY COPMIS$10N OFFICE OF INSFECTION AND ENFORCBIENT 11ASRINGTON, D.C.- 20555
- February 8, 197) .
IE Du11stin No. 79 : ENVIRONMENTAL QUALIFICATION OF CIASS IE EQU1FMENT Description of Circumstances: The intent of IE Circular 78-08 was to highlight to all licensees' , - important lessons learned from environments 1 qualification deficiencies
. reported by individual licensees. In this regard . licensees were requested to ensaine installed safety-related electrical equipment .
and determine that proper documentation existed which provided assurance. that this equipment wou'i function under postulated accident conditions.
, The scope of.IE Circular 18-08 was such broader than other previously issued Bulletinsand Circulars (such as IES 78-06 and IER 78-02) which
% Y ' addressed specific component failtires. The intest of this Sm11stia is to raise the threshold of-IE Circular M-03 to the level of a Bulletin; e
~1.e., action-requiring a licensee response.
. Inspections conducted to date by the NRC of licensees' activities in -
^
. response to. IE Circular 78-08 have identified one component which licensees have found to be unqualified for service within the LOCA environment..
Specificially, unqualified stam sounted limit switches'(SMLS)...other than those identified in previously issued IE Bulletin 78-04, were found, to be installed ed safety-related valver inside contaitret st Ah Duane Arnold and Quad' Cities'I and 2 Nuclear Cenerating Staticas.- The e~salified switches - are idotified as NAM:0 Models SL2-C-11, $3CML, sal-31, SA1-32, D1200j,
' EA-700 and EA-770 switches. According to the. man'afacturer, these switches -
- are designed only for general purpose applications and.are mot considered suitable. devices- for service-in the LOCA envirossent. Consequently (
switches are being replaced at the above power plants with qualified compo ents.- Also, R C inspection of component qualification has identified equipment
. which does not have documentation indicating it is qualified for the LOCA-environment. The inspections have also identified that the licensees' O ,
re-review and resolution of probles areas are. mot receiving the level of ~ attention from all licensees which the NRC believes is warranted. Because of the- protracted schedule for completion of the re-review, we are now requesting the power reactor facilities with sperattag licenses to 8 , expedite c.ompletion of their re-review progran or131selly requested by IE Circular 78-06 dated May 31, 1978. Page 1 ef 3
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. ak . , . .- () O 0051370 IE Dulletin No. 79-01 February 8, 1979 Action to se Taken By Licensees of All Power Reactor Facilities (Except Those 11 SEP Plants Listed on Enclosure 3) With An Operating License: -
- 1. Cosplete the re-review program described in II Circular 78-08 within 120 days of receipt of this Sulletin.
- 2. Determine if the types of stam mounted limit switches described above are being used or plassed for use on safety-related ve?ves which are located inside containment at your facility. If oo provide a written report to the NRC within the time frame specified and to the address specified in item 4 below.
- 3. Provide written evidence of the qualification of electrical equipment required to fur.ction under accident conditions.* For those items nr t havirio complete qualification data available for review. identify your plans for determining qualification, either by testing or engineering analysis, or combination of these, or by replacement with qualified equipment. Inclu.e your achedule for completing these actions and your justification for contiar 4
- . ( operation, s
submit this information to the Director. Division of Reactor Opera-tions Inspection. Office of Inspection and Enforcement. Nuclear Regulatory Couseission. Washington, D.C. 20555 with a copy to the appropriate NRC Regional Of fice within 120 days of receipt of this Bulletin. 4 Report any items which are identified as not meeting qualification requirements fer service intended to the Director. Division of Operating Reactors. Office of Nuclear 2 actor Regulation. Nuclear
. Regulatory Comeission. Washington. D.C. 20555 with a copy to the appropriate.NRC Regional Office within 24 hours of identification.
If plant operation is to continue following identification, providt justification for such operation. Provide a detailed written report within la days of identification to NRR. with a copy to the appre-1 priate NRC Regional Office.
. 'l
- This written evidence should-include: 1) component description;
- 2) description of the accident environment; 3) the environment to which the component or equipment is qualified;.4) the manner of I qualification which should include test methods such as sequential, synergistic, etc., and 5) identification of the specific supporting I qualification documentation.
Prge 2 of 3 L e .
k ((., - - - (. O 0051369 12 Bulletin No. 79-01 February 8, 1979 No additional written response to this IE Bulletin is required other than those responses described above. WitC inspectors will testinue to monitor the licensees' progress in coeP lating the requested action described above. If additional information is required, contact the Director of the appropriate NRC Regional Office. Approved by GAO 3180225 Ot0072); clearance expires 7/31/80. Approval was given under a blanket clearsnee specifically for identified generic problems.
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- Plant Region ,
Dresden 1 'HI Yankee Rowe I nig Rock Point III 9' San Onofre 1 'I ' V I Baddam Nect III Lacrosse Oyster Creek I .. , R. I. Ginna 1 I Dresden 2 UI *I I Millstone
. Palisades III
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_ . . ~ _- . . __ _ _ _ . _ _ _ _ . _ _ _ . l I c . g O 0051367 ( .. . IE Bulletin No. 7 b 01 February 8, 1979 LISTING OF 1E BULLETINS >
!$50ED IN 1AST TWELVE MONTR$
Sulletin Subject Date Iseued teamed To noe 76-09 BWR Drywell Leakage 6/14/78 All WR Power Fatha Assacfated with Raatter Facilities ' 1madeguare Drywell veh an OL er CP Closures Y. 78-10 3ergen-Paterson 6/27/78 All WWit Power Rydraulic Shock Rat.etor Facilities Suppressor Accumulator with an OL or CP ' Spring Cof.ls 78-11 Examination of Mark 1 7/21/76 DWR Power Reactor Containment Torus Facilities for Welds action: Peach Bottoe 2 and 3 Quad Cities 1 and 2, Ratch 1, Monti-
. J(q-f sailo and Veruset Yaakse 78-12 Atypical Weld Material 9/29/78 All Powett Reactor '
in Reactor Pressure Facilities with an ! Vessel Welds OL or CP l 78-22A Atypical Weld Material 11/24/78 All Power Reactor in Reactor Pressure '
.icilities with an Vessel Welds L or CP 78-13 Failures 7a Source Meads 10/27/78 A11.gener31 and of Kay-Ray, 7ac. , Causes specific licen% -s Models 7050, 70508, 7051, with the subjet.
70513, 7MD, 7M03, 7M1 II,ay-Ray, Inc. and 7M1B gausea 74-14 Deteriorar. ion of Suna-N 12/19/78 All CE BWR facilities Components In ASCO . with aa OL or CP
. Solenoids Eaclosure 2 Page 2 of 2
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g' . O 0051368 : ( .. . IE Su11stin No. 75 01 February 8,1979 k I L!fTING OF IE SULLETINS
- ISSUED IN LAST TWELVE IGNTE$ ,
tulletin Subject Date Issued Isoved to no. 78-03 Potential taplosive 2/8/78 All SWR F r Ces Misture Accumula- Reactor Facilities . tions Associated with with an OL or CP SWR Ofigas System Operations 78- 04 Environmental Qualf- 2/21/78 All Power Raatter fication of-Certain Facilities with na ; Stes Mounted Limit OL or CP ;
$ witches !aside Reactor Costa 1 ament 78-05 Malfunctiostag of 4/14/78 All Power Reactor '
Circuit Braaker Facilities with sa > Auxiliary Contact OL or CP ' Mechanism-General Model CR105X 78-06 Defective Cutler- 5/31/78 All Power Reactor Hammer, Type H Relays Facilities with an With DC Ceils . 01. or CP 78-07 Protection afforded 6/12/78 All l'ower Reactor-by Air-Line Respirators Tacilities with an and Supplied-Air Moods OL, c11 class E and F Research Reactors with
, , en OL, all Fuel Cycle Facilities with an OL, !
and all Priority 1 Material Licumaees 76-06 Radiation Levels fron 6/12/78 All Power and Fuel Element Transfer Research Reactor Tube s Facilities with a Paol Element transfer tube and an OL. 8 Rectosure 2 PaSe 1 af 2 i
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f 2} St:ff hh. 24 So- Bye /.u f swp , *ffL < UNITEDSTATES 9Np%u.: useu phe, NUCLEAR REGULATORY COMMISSION Accessions No.:
- CFFICE OF INSPECTION AND ENFORCEMENT 7910250028
' WASHINGTON, D.C. 20555 '92 ftAR 13 P12 :00 January 14, 1980 d i te m si c,w IEBNtYEidoE79d1B -
ENVIRONMENTAL QUALIFICATION OF CLASS IE EQUIPMENT Description of Cire w tances: IE Bulletin 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 (i.e. during and following) postulated accident conditions. The NRC staff has completed the initial review of licensees' responses to Bulletin No. 79-01. Based on this review, additional information is needed to f acilitate completion of the NRC evaluation of the adequacy of environmental qualification of Class IE electrical equipment in the operating facilities. In addition to requesting more detailed information, the scope of this Bulletin is expanded to resolve safety concerns relating to design basis environments and current qualification criteria not addressed in the facilities' FSARS. Tnese include high energy line breaks (HELB) inside and outside primary containment,
- .ging, and submergence.
~= __
Enclosure 4, " GUIDELINES FOR EVALUATING ENVIRONMENTAL QUALIFICATION OF CLASS IE ELECTRICAL EQUIPMENT IN OPERATING REACTORS", provides the guidelines and criteria the staff will use in evaluating the adequacy of the licensee's Class IE equipment evaluation in response to this Bulletin. In general, the reporting problems 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 contain-ment, was not included in the responses.
- 2. In many cases, the specific information requested by the Bulletin for each component of Class IE equipment was not reported.
- 3. Diffeant methods and/or formats were used in providing the written evidence of Class IE electrical equipment qualifications. Some licensees used the System Analysis Nethod 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 resulting from both LOCA and/or HELB inside primary containment and HELB
": outside the primary containment.
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' January /S,1980 IE Bulletin No. 79-01B Page 2 of 3 f==
- b. Identification of the Class IE electrical equipment items within each of the systems identified in Item a, that are required to function under the postulated accident conditions.
- c. The correlation between the environmental data requirements specified in the FSAR and the environmental qualification test data for each Class IE electrical equipment item identified in Item b above.
- 4. Additional data not previously addressed in IE Bulletin No. 79-01 a,re needed to determine the adequacy of the environmentai qualification of Class IE electrical equipment. These data address component aging and operability in a submerged condition.
Action To Be Taken By Licensees Of All Power Reactor Facilities With An Operating License (Except those 11 SEP Plants Listed on Enclosure 1)
- 1. Provide a " master list" of all Engineered Safety Feature Systems (Plant Pro-tection Systems) required to function under postulated accident conditions.
Accident conditions are defined as the LOCA/HELB inside containment, and HELB outside containment. For each system within (including cables, EPA's termintl blocks, etc.) the master list identify each Class IE electrical equipment item that is required to function under accident conditions. Pages 1 and 2 of Enclosure 2 are standard formats to be used for-the " master
+ list" with typical information included.
Electrical equipment items, which are components of systems listed in Appendix A of Enclosure 4, which are assumed to operate in the FSAR safety analysis and are relied on to mitigate design basis events are considercd within the scope of this Bulletin, regardless whether or not they were classified as part of the engineered safaty features when the plant was orginally licensed to operate. The necessity for further up grading of nonsafety-related plant systems will be dependent on the outcome of the licensees and the NRC reviews subsequent to TMI/2.
- 2. For each class IE electrical equipment item identified in Item 1, provide written evidence of its environmental qualification to support the capability of the item to f unction 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. Drovide this in the format of Enclosure 3.
- 3. For equipment identifed in Items 1 and 2 provide service condition These data profiles (i.e., temperature, pressure, etc., as a function of time).
should be provided for design basis accident conditions and qualification tests performed. This data may be provided in profile or tabular form. l
- =
l
IE Bulletin No. 79-01B _ January 14, 1980 p _,; Page 3 of 3
- 4. Evaluate the qualification of your Class IE electrical equipment against the guidelines provided in Enclosure 4 Enclosure 5, .nterim Staf f Position on Environmental Qualification of Safety-Related Electrical Equipment,"
provides supplemental information to be used with these gulielines. Fct the equipment identified as having " Outstanding Items" by Enclosure 3, provide a detailed " Equipment Qualification Plan." Include in this plan specific actions which will be taken to determine equipment qualification and the schedule for completing the actions. .
- 5. 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 information in the format of Enclosure 3.
- 6. Submit a " Licensee Event Report" (LER) for any Class IE electrical equipment item which has been determined as not being capable of meeting environmental qualification requirements for service intended. Send the LER to the approp-riate NRC Regional Office within 24 hours of identification. If plant opera-tion is to continue following identification, provide justification for such operation in the LER. Provide a detailed written report within 14 days of icentification to the appropriate NRC Regional Office. Those items which were previously reported to the NRC ab not being qualified per IEB-79-01 do e6En not require an LER.
- 7. Complete the actions specified by this bulletin in accordance with the following schedule:
(a) Submit a written report recuired by Items 1, 2, and 3 within 45 days from receipt of this Bulletin. (b) Submit a written report required by Items 4 and 5 within 90 days from receipt of this Bulletin. This information is requested under the provisions of 10 CFR 50.54(f). Accordingly, you are requested to provide within the time periods specified in Items 7.a and 7.b above, written statements of the above information, signed under oath or affirmation. Submit the reports to the Director of the appropriate NRC Re0i onal Office. Send & copy of your report to the U.S. Nuclear Regulatory Commission, Office-of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D.C. 20555. Approved by GAO, B180225 (R0072); clearance expires 7/31/80. Approval was given under a blanket clearance specifically for identified generic problems.
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Enclosure !
*f~_" IE Bulletin No. 79-01B i
( Januaryfy,1980 ! RECENTLY ISSUED IE BULLETINS Subject Date issued itsued To Bulletin r No. l 12/7/79 All power reactor 79-28 Possible Malfunction of facilities with an . Namco Model EA 180 Limit OL or a CP Switches at Elevated
- Temperatures-11/30/79 All power reactor 79 27 Loss Of Non-Class-1-E facilities holding i Instrumentation and OLs and to those.
Control Power System Bus During Operation nearing licensing l 11/20/79 All BWR power reactor 79 Boron Loss From BWR Control Blades facilities:with an ' OL Failures of Westinghouse 11/2/79 All power reactor 79 25 facilities with an
- BF0 Relays In Safety-Related OL or CP
##" - Systems
,7 .
~
Pipe. Cracks In Stagnant 30/29/79 All PWR's with-an -+ 79-17 - OL and for information (Rev. 1) Borated Water System At to other po'.er reactors' . PWR Plants-
-r 9/27/79 All: power reactor - 79-24 ~ Frozen Lines facilities which have either OLs or cps and are in the late-stage [
of construction 9/12/79 A11 Power Reactor 79-23 Potential Failure of Facilities with an Emergency Diesel Operating License or Generator Field a construction permit: Exciter Transformer . Seismic Analyses For 9/7/79 All Power Reactor
'79-14' fac'lities with an (Supplement 2) As-Built Safety-Related OL or a'CP j
-Piping Systems '
Possible Leakage of Tubes 9/5/79 To Each Licensee
'79-22 who Receives Tubes of Tritium Gas in Time' of Tritium Gas- :
pieces for Luminosity Used in Timepieces for Luminosity ,
-e
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Enclosure 1 SEP Plants Y^ Peaton Plant
- III Dresden 1 I
- Yankee Rowe III Big Rock Point y San Oncfre 1 I Hsddate Heck III Lacrosse I Oyster Creek I, R. E. Ginna III Dresden 2 I Millstone III Palisades i
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- q.
Facility: XYZ Enclosure 2 Jocket No.. 50-XXX HASTER LIST (Typical)
~
(Class IE Electrical Equipment Required to Function (. Lnder Postulated Accident Conditions)
- 1. SYSTEM: RESIDUAL HEAT REMOVAL (RHR)
COMPONENTS l Location Plant Identification Inside Primary Outside Primary Numoer Generic Name Containment Containment , IPT 456 PRESSURE TRANSMITTER x ILT 594 LEVEL TRANSMITTER x ILS 210 LIMIT SWITCH x
=
f4?+ II. SYSTEM: AUTOMATIC DEPRESSUR12ATION SYSTEM (ADS) l l - t COMPONENTS loc 8Llon Outside Primary Plant Identification Inside Primary Generic Name Containment Containment Number B21-R001 VALVE MOTOR OPERATOR x SOLEN 0ID VALVE x i B21-F003 PRESSURE SWITCH x B21-F010 l
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. . _ . ~ - . . _ . . . _ - . . _ . _ . _ . . _ . _ . _ _ _ _ _ _ . _ ... _ _ ._ . .__ . ._.__
Rl:R EQUIPMENT /COMr0NENTS (Typical) i
. 411. SYSTEM: i " ** COMPONENTS 1 ,
y ' Location : 1 Inside' Primary Outside Primary PlintIdentification Containment
}
Number
- Genaric Name Containment 16xP455 0 RING GAtKET x >
~~
j EPA, Class E, l Westinghouse,100C ELECTRICAL PENETRATION ASSEMBLY X
' TERMINAL BOARD x KULKA No. ET35 ,
ONKONITE,1000V, JC , x x Black- - PO,WER CABLE-X BRAND 10W-40 LUBRICATE OIL x , 15 KB69 (Boston Wire & Cable) = INSTRUMENTATION CABLE x x 1 Cutler Hamer TB- x No. 6 TERMINAL BOX RAYCHEM XYZ CABLE SPLICE x x Scotch No. 54 INSULATING TAPE- x . p= . T&B No. 10 INSULATED TERMINAL LUG x x x I Y Brand Epoxy No.. - SEALANT ' 111
- When a component is not identified by plant identification number, use the manufacturer, model number, serial number, etc.
** Like compo".ents may be referenced.
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I,z<c. . .. Enclosure 3
~e SYSTEM COMPONENT EVALUATION WORK SHEET INSTRUCTIONS
- 1. Equipment
Description:
Provide the specific information requested for each Class TE electrical component. Provide component location, specific information such as the building, access ficor elevations, and whether the component is above the flood level elevation. In addition, provide the specified and demonstrated accuracies of all instruments for their trip functions and/or post accident monitoring requirements. Cables, EPA's, terminal blocks, and other items shall be identified as part of the engineered safety features systems.
- 2. Environment: List values for each environmental parameter indicated.
List the " specification values" obtained from postulated accident analysis in t"t " SPEC" column. List the " qualification values" obtained from test Tempera-rep; engineering analysis data, etc. in the " Qual" column. ture, p. .ssure, etc. , as a function of time shall be provided in profile or tabular form. Specify the time period that the componene or equipment is required to function and identify the document which provides the basis for this time interval. It is expected that some listed parameters were not requested cf the licensee at the time of their license issuance. Address each param:ter
- 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 "Dutstanding Item."
- 3. Documantation
Reference:
Reference the documents from which information was obtained in the " Spec" column. Identify the document, paragraph, etc., that contains 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. Words such as " test" and/or " analysis" 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 addressed during FSAR environmental qualification analysis such as submergence, qualified life (aging), or HELB. Identify in the " Notes" section on page 1 of this enclosure the actions planned for determining quaIcation and the schedule for completing these actions.
cility: PC9e 1 cr r . g. je y .
~
m!i SYSTEM COMPONENT EV" !10N WORK SHEET * *
- t:
1
- k;t: 7.i (Typ ,ji
- 3 '
ENVIRONMENT DOCMNTATION REP QUIPMENT DESCRIPTION QUALIFICATION OUTSTANDI NG Parameter Speci{i- QuaH{i- Specifi- Qualiti- MET 1100 ITEMS cannn ra ti nn cation cation '
- stem: RHR Operating 15 min. 300 min. 1 5 Simul taneous None icnt ID No. IPT456 Time Test mponent: ' '
Tempera tum I 5 Simultaneou
- ESSURE TRANSMITTER SEE ACCIDENT AND
( F) TEST PROFILES , Test None nufccture: PROVIDED scher-Pcrter Co. Pressure * (PSIA) 1 5 Simultaneou: None . del Number: Test ' -EN-1071-BCXN-NS
. Relative Humidity (%) 100% 100% 1 5 Simultaneou< lione ction:
ident Monitoring i .i Test Chemical N 0
- . urccy: Spec: 5% Spray 3 3 1 See Note 1 Demon: 4%
NA0H 'vics: RilR Pump 1A 6 0 Sequential charge Pressure Radiation 4x10 rads 1.2x10 rad i 2 6 Test None WO7 1. Sequentti I ction: Containment Aging 40 yrs 40 yrs 3 7, 8
- 2. En nal yst od Lcvel Elev: 620' ) Not Not None
.ve Flood Level: Yes l Submergence Required Required See Note 2 No x' cumentation
References:
Notes : FSAR Chapter 3 Paragraph 3.11 1. XYZ Letter No. 237-1, dated November 2, 1979, FSAR Chapter 14 Paragraph 14.2.3.1 has been sent to MFG. requesting the ocalification Technical Specification 3.4.1, Paragraph A informa tion. If qualification not detmnined T* chnical Specification 4.6.5, Paragraph B acceptable by December 15, 1979., component F2'4L Test Report No. 30G0 dated November 2,1972 - will be replaced during refueling outage March 1980. Fi:e-her and Porter Co.' Test Report No. 2500-1 A. b. 000 Engineering Evaluation Data Report No. 6932 2. In the FSAR submergence was not considered Wylie Laboratory Report No. 467 an environmental parameter. ABC Laboratory is to perform submergence test in April 1980.
Enclosure 4 GUIDELINE 5 FOR EVALUATING EfNIRONMENTAL 00ALIFICATION 0F CLASS IE ELECTRICAL EQUIPMENT _ lH OPERATING REACTORS 1.0 Introduction 2.0 Dis cussion 3.0 Identification of Class _ IE Ecuioment. 4.0 Service Conditions _ 4.1 service Conditions Itsi+r Containment for a loss of Coolen: Accioec- (LO T ~
- 1. Temeerature and Pressure Steam _ Conditions
- 2. Radiation
- 3. Suorergence 4 Che-ical Sorays 4.2 service enditions for a P'AR rain Steam Line Ereak (MSLB)
Insi:e C:ntainmett
- 1. Temeerature and Pressure Steam Conditions
- 2. Radiation
- 3. Subcergence 4 Cherical Sorays 4.3 Service Conditions Outside Containne,n_t.
- 4. 3.1 Areas Subf ect to a Severe Environment as a Result o' a nien Eneroy Line Break thELB) 4.3.2 A-eas Where Fluids are Recirculated Fron Inside C:ntain ent to Accomolish Lont-itre Emergency Core Cooline Following a LOCA
- 1. Tem:erature, Pressure and Relati',e Humidity
- 2. Radiation
- 3. Subrergence L. Che-ical Sorays
2 4.3.3 Areas Normally Mat'.tained at Room Condition,s 1 5.0 Ovalificati.on Methods 5.1 Selection of Oualification Method 5.2 Oualification by Type Testing
-1. Simulated Service Conditions and Test Duration
. 2. Test Scecimen
- 3. Tes t, Seouence
- 4. Tes; Soecimen Aging ,
- 5. Func}ional Testine and Failure Criteria
- 6. Installation Interfaces 5.3 Oualification by a Combination of Methods (Test, Evaluation, Ana lysis )
4 6.0 Maroin
.7.0 Acino 8.0- Documentation
~
Appendix. A - Typical Equipment / Functions Needed for Mitigation of a LOCA or MELS Accident Appendix B - Guidelines for Evaluating Radiation Service Conditions
- Inside Containment for a LOCA and MSLB Accident Appendix C --Thent.a1 and Radiation Aging Degradation of Selected Materials e
.....n
GUIDELINES FOR EVALUATING ENVIRONMENTAL 00ALIFILa 10N ! I OF CLASS IE ELECTRICAL E00lPMENT INOPERATINGREACTQRS, S 4
1.0 INTRODUCTION
~
On February 8,1979, the NRC Office of 1nspection and Enforcement issued IE Bulletin 79-01, entitled, " Environmental Qualification of Class IE Equipment." This bulletin requested that licensees for operating powar reactors complete within 120 days their reviews of equipment qualification begun earlier in connection with IE Circular 78-08. The objective of IE Circular 78-08 was to initiate a review by the licensees to deterrine i whether proper docu.entation existed to verify that all Class IE electrical equipment would function as required in the hostile environment which could , result fror design basis events. The licensees' reviews are now essentially complete and the NRC staff has . begun to evaluate the re:ults. This document sets forth guidelines for the NRC staff to use in its evaluations of the licensees' responses to IE Bulletin 79-01 'and selected associated Qualification documentation. The objective of the evaluations using these guideli..es is to identify Class IE equipment whose docu entation does not provide reasonable assurance of environ-mental qualification. All such equipment identified will then be subjected , to a plant applicatico specific evaluation to detemine whether it should be requalified or replaced with a component whose qualification has been adequately verified. These saidelines are intended to be used by tne NRC staff to evaluate the
~
qualification rtethods uses '*ar txisting equipment in a particular class of plants, i.e., currently operating reactors including SEP plants. a ... _ . __ _ _ _ _ . . _ _ _ , .
1 2 Equipmcat in other classes of plants not yet licensed to operate, or
, replacement equipment for operating reactors, may be subject'to aifferent requirenents such as those set forth in NUREG-0588, Interim Staff Position j on Environmental Qualification of Safety-Related Electrical Equipment.
In addition to its reviews in connection with IE Bulletin 79-01 the staff is e,ngaged in other generic reviews that include aspects of the equipment qualification issue. TMI-2 lessons learned and the effects of failures of non-Class IE control and indication equipment are examples of these generic reviews. In sore cases these guidelines may be applicable, however, this deterrination will be nade as part of that related generic review, 2.0 DISCUSSION IEEE Std. 323-1974 I is ..ie current industry standard for environmental qualification of safety-related electrical equipment. This standard was first issued as a trail use standard, IEEE Std. 323-1971, in 1971 and later
- after substantial revision, the current version was issued in 1974 Both
..rsions of the standard set forth generic requirements for equipment quali-fication Lut the 1974 standard includes specific requiraments for aging, l
margins and maintrining dccumentation records that were not included in the 1971 trial use standard. The intent of this document is not to provide guidelines for implementing either version of IEEE Std. 323 for operating reactors. In fact most of the operating reactors are not committed to comply with any particular industry standard for electrical equipment qualification. However, all of the operating reactors are required to comply with the General Design Criteria I IEEE Std. 323-1974, "IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations."
7-
- 3- ,
specified in Appendix A of 10 CFR 50. General Lesign Criterion 4 states is, part tnat ' structures, systems and components important to safet; shall be designed to accomodate the affects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing and postulated accidents, including loss-of-coolant accidents." The intent of these gu'delines is to provide a basis for judgements required to confirm that operating reactors I are in compliance with General Design Criterien 4. 3.0 IDENTIFICATION 0: CLAS$ IE E0'J1FMENT Class IE equipment includes all electrical equipment needed to achieve energency reactor shutdown, containment isolation, reactor core cooling, containment and reactor heat removal, and prevention of significant release of radioective retirial to the environment. Typical systems included in pressurized and builing water reactor designs to perform these functions for the nest severo postulated loss of coolant accident (LOCA) and main steanline break accident (MSLB) are listed in Appendix A. More detailed descrictions of the Class li equipment installed at specific plants can be obtained from FSARs, Technical specifications, and emergency procedures. Although variation in nomenclature may exist at the various plants, environmental qualification of those systems which perform the functions
'dentified in Apoendix A should be evalua:ed against the appropriate service conditions (Section 4.0). .
The guidelines in this document are applicable to all components necessary l for operation of the systems listad in Appendix A including but not limited to valves, 6stors, cables, connectors, relays, switches, transmitters and valve position indicators. L .
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. -. . = . . . . - . - . - _ - . . - - . - ._ .-_ -
4.0 SERVICE CONDITIONS In order to detemine the adequacy of the qualification of equipment it is necessary to specify the environnent the equipment is exposed to during norr.a1 and accident conditions with a requirement to remain functional, Thr.sa environments are referred to as the ' service conditions." The approved service conditions specified in the FSAR or other licensee submittals are acceptable, unless otherwise noted in the guidelines discussued beloo . 4.1 Service Conditions Inside Containment for a loss of Co7 ant 1 Accident (LOCA) 1, Temoerature and Pressure Steam Conditions . In general, the containment temperature and pressure conditions as a function r f time should be I based on the analyses in the FSAR, In the specifi: case of pressure l l . suppression type containments, the following minimum high tempeature " conditions should be used: (1) BWR Drywells - 3400F for 6 hours; and (2). ?WR Ice Condenser Lower Compartments . 3400F for 3 hours,
- 2. ?.adiation - When speci*ying radiation service conditions for equipment exposed to radiation during normal operating and accident conditions, the nomal operating dose should be added to the dose received during the course of an accident. - Guidelines for evaluating beta and game radiation service conditions for general areas inside containment are provided below. Radiatien service conditions for equipment located directly above the containment sump, in the vicinity of filters, or
[ l-_ submerged in contaminated liquids must be evaluat'ed on a case by case , \ l basis, Gui.delines for these evaluations are not provided in this document. L . . -_ . - - - . . _ -- -
-. =-
.L o5 Gama Radiatien Doses - A total gama dose radiatien service condition ' of 2 x 10 7PADS is acceptable for Class IE equipt..it located in general areas inside containment for PWRs with dry type containments, )fhere a dose less than this value has been specified, an application specific evaluation must be perforned to detemine if the dose specified is acceptable. Procedures for evaluating radiation service conditions <
in such cases are provided in Appendix B, The procedures in Appendix B are based on the calculation for a typical PWR reported in Appendix I D of NUREG-05BS . Gama dose radiation service conditions for BWRs and PWRs with ice condenser containments must be evaluated on a case by case basis, Since the procedures in Appendix B are based on a calculation for a typical PWR with a dry type contatnmet, they are not directly applicable to BWRs and other containment types, however, doses #9r these other
. elant configurations may be evaluated using similar procedures with conservative dose assumptions and adjustment factors developed on a case by case basis.
Beta Radiation Doses - Beta radiation doses generally are less significant than garma radiation doses for equipment qualification. This is due to the low penetrating power of beta particles in comparison to gamma rays of equivalent energy, Of the general classes of electrical equipment in a plant (e.g., cables, instrument transmitters, valve operators, containment penetrations), electrical cable is considered the most l tOREG-05SS, hnterim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment.
.. . -= - - . - - . . - - . . = _ - - . - . -. . .. --
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vulnerable to damage from beta radiation. Assuming a TID 14844 source tem, the average raaximum beta energy and isotepic abundan;e will vary as a function of time following an accident. If these parame*ers are cunsidered in a detailed calculation, the conservative beta surface dose of 1.40 x x 108 R/DS reported in Appendix D of NUREG 0588 would be reduced by approximately a factor of ten within 30 mils of the surface of electrical cable insulation of un;t density. 7.n additional 40 mils of insulation Itotal of 70 mils) results in another . factor of 10 reduction in dose. Any structures or other equipment in the vicinity of the equipment of interest would act as shielding to further reduce beta doses. If it can be shown, by assuming a conserva-tive unshielded surface beta dose of 2.0 x 108 PADS and considering the shielding factors discussed here, that the beta dose to radiation sensitive equipment internals would be less than or equal to 10% of the :otal gar:na dose to which an item of equipment has been qualified, then thu equi;nent may be considered qualified for the total radiation envirorr.ent (gamma plus beta). If this criterion is not satisfied the radiation service condition should be determined by the sum of the gama and beta doses.
- 3. Submercence - The preferred method of protection against the effects of subreergency is to locate equipment above the water flooding level.
Specifyir.g saturated steam as a service condition during type testing of ' equip ent that will- become flooded in service is not an acceptable alternative for actually flooding the equipment during the test. g $_, ~ _ .
7
- 4. Containment Soravs - Equipment exposed to chemical sprays should be qualified for the most severe chemical environment (acidic or basic) which could exist. Demineralized water sprays should not be exempt fmm consideration as a potentially adverse service condition.
4.2 Service Conditions for a PWR Main Steam Line Break (MSLB) Inside Containment Equipment required to function in a steam line break environment must be
- qualified for the high temperature and pressure that could result.
In same cases the environmental stress on exposed equipc.ent may be higher than that resulting from a LOCA, in others it may be no more severe than for a LOCA due to the automatic operation of a containment spray system.
- 1. Temoerature and Pressure Steam conditions - Ecuipment qualified for a LOCA envirenmer.t is considered qualified for a MSLB accident environ-ment in plants with automatic spray systems not subject to disabling single component failures. This position is based on the "Best Estimate" calculation of a typical plant peak temperature and pressure and a thermal analysis of typical components inside containment.M The final ac:eptability of this approach, i.e., use of the "Best Estimate",
is pending the completion of Task Action Plan A-21, Main Steamline Break Inside Containment. Class IE equipment installed in plants without automatic spray systems or plants with gpray systems subject to disabling single . failures or delayed initiation should be qualified for a MSLB accident environment determined by a plant specific analysis. Acceptable methods
-*I See NUREG- GaSE, Short Term Safety Assessment on the Environmental Qualification of Safety-Related Electrical Equipment of SEP Operating Reactors, for a more detailed discussion of the best estimate calculation.
, 8- -
l l ia-i i= for performing such an analysis for operating reactors are. provided . in Section 1.2 for Category II plants in NUREG-05BB Interim Staff Position on Environmental Qualification of Safety-Related Eb!ctrical Equipment.
- 2. Radiation - Sare as Section 4.1 above except that a conservative gpmma dose of 2 x 10I RADS is acceptable.
- 3. Submeroence - tame as Section 4.1 above.
- 4. Chemical Sorays - Same as Section 4.1 above.
4.3 Service Conditions Outside of Containment 4.3.1 Areas Subject to a Severe Environment as a Result of a Hich Enerey Line Break (HELB) Service conditions for areas outside containment expesed to a HELB were evaluated on a plant by plant basis as part of a program initiated by the staff in December,1972 to evaluate the effects of a HELB. The equipment required to mitigate the event was also identified. This equipment should-be qualified for the service conditions reviewed and approved :n tne if:.5 Se'ety Evaluaticn Report for ea:r. speci'ic plant. 4.3.2 Areas k'here Fluids are Recirculated from Inside Containment to Accomolish Lone-Term Core Cooline Followine a LOCA -
- 1. Termerature and Relative Humidity - One hundred cercent relative humidity t * ~ Us t established as a service condition in confined spaces. The
-J.' re and pressure as a function of time should be based on the plant nique analysis reported in the FSAR.
. =
, , ~ . _ ,
9
- 2. Radiation - Due to differences in equipment arrangement within these arets and the significant effect of this factor on doses, radiation se,'ce conditions must be evaluated on a case by case basis.- In general, a dose of at least 4 x 10 6 RADS would be expected.
- 3. Submergence - Not applicable.
- 4. ' Chemical Soreys - Not toplicable.
4.3.3 Arees Norna11y Maintained at Room Conditions Class IE equipment 1::ated in these areas does no* experience significant stress due to a change in service conditions during-a design basis event. This equipment was designed and installed using standard engineering practices and industry cedes and standards (e.g., ANSI, NEMA, National
. Electric Code). Based on these factors, failures of equipment in these areas during a design basis event are expected to be random except to the extent that they may be due to aging or failures of air conditioning or ventilation syster.s . Therefore, no special consideration need be given to the environmental cualification of Class IE equipment in these areas provided the aging recuirements discussed in Section 7.0 below are satisfied.and the areas are reaintained at roon conditions by redundant air conditioning or ventilation 'syster.s served by the onsite emergency electrical power system.
Equipment located in areas not served by redundant systems powered from onsite ecerpency sources should be qualified for the environnental extremes which could result from a failure of the systems as determined from a plant specific analysis. t 5.0 .0'JA'.I ICAT!0N METHO S m- -
5.1 Selection of 0931ification Method The cnoice of qualification method employed for a particular application of equipment is largely a tratter of technical judgement basid on such factors as: (1) the severity of the service conditions; (2) the structural and raterial complexity of the equipment; and (3) the degree of certainty required in the qualification procedure (i.e., the safety importance of the equiprent functiori). Based on these considerations, type testing is the preferred method of qualification for electrical equipment located insic'e contain: rent required to mitigate the consequences of design basis events, i.e., Class IE equipment (see Section 3.0 above). As a minimum, , the cualification for severe temperature, pressure, and steam service l l conditions fer Class IE equipment should be based on type testing. IQualificati'on for other service conditions such as radiation and chemica' sprays may be by analysis (evaluation) supported by test data (see Section
- 5. 3 below). Exceptions to these general guidelines must be justified on a fase by case basis.
5.2 Ouali'ica-ion by Tyre Testine The evaluation of test plans and results should include consideration of l ! the f:11owing factors:
- 1. Simulated Service Conditions and Test Duration - The environment in the test chamber should be established and maintained so that it envelopes the service conditions defined in accordance with Section 4.0 above.
The time duration of the test should be at least as long as the period from the-initiation of the accident until the temperature and pressure l - service c:nditions return to essentially the same levels that existed l before the postulated accident. A shorter test duration may be acceptable i
if specific' analyses are provided to demonstrate that the materials , invrlved i 11 not experience significent accelerated themal aging j dur ..ig the period not tested.
- 2. T_est Soecimen - The test specimen should be the same model as the equipn.ent being qualified. The type test should only be considered valid for equipment identical in design and material construction to the test specimen. Any deviations should be evaluated as part of the qualifica-tion documentation (see also Section 8.0 below).
3 Test Secuence - The component being tested should be exposed to a - steam / air environment at elevated temperature, and prtssure in the secuence defined fcr its service conditions. Where radiation is a service condition which is to be considered as part of a type test, it may be ac:'.iH at any time during the test sequence provided the component does not centain ar.y materials which are known to be susceptible to significant radiation damage at the service condition levels or caterials wnese suscestibility to radiation damage is not known (see Ap:endix C). :f tr,e component contains any such materials, the radiation ccse should be apolied prior to or concurrent with exposure to the elevated te perature anc pressure steam / air environment. The same test specimen shculd be used throuchout the test sequence for all service conditions the equipment is to be qualified for by type testing. The type test
.. should only be considered valid for the cervice conditions applied to the same test specimen in the appropriate sequence.
- a. Test Sce:iren Acine - Tests which were successful using test specimens wnich hao not been creaged may be considered acceptable orovided the
. :::.cenen;;;es not :cntain naterials which are known to be susceotible
12 - to significant degradation due to themal and radiation agin (see Section 7.0). If the component contains such materials a qualified life for the component must be established on a case by case basis. Arrhenius techniques are generally considered acceptable for thermal aging. ,
- 5. Functional Testine and railure criteria - Operational modes tested i
. should be representative of tbt actual application requirements (e.g., components which operate normally energized in the plant should be normally energized during the tests, motor and electrical ,
cable leading-during the test should be representative of actual operating conditions). Failure criteria should include instrument accuracy recuirements based on the maximum error assumed in the plant safety analyses. If a component fails at any time during the test, even in a so called " fail safe" mode, the test should be considered inconclusive with regard to demonstrating the ability of the cceponent to function for the enti i Seriod prior to the failure.
- 6. Installation Interfaces - The ecuipment mounting and electrical or mechanical seals used during the type test should be representative of the actual installation for the test to be considered conciusive.
The equipment qualification program should include an as-built inspection in the field to verify that equipment was installed as it was tested. Particular emphasis should be placed on comen problems 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 conductors.
,-$+ 4- -E -
1
- 13 .
I 5.3 Ovalification by a Combination of Methods (Test, Evaluatier, Analysis As discussed in Section 5.1 above, an item of Class IE equipment may be shown to be qualified fer a complete spectrum of service conditions even though it was only type tested for high temperature, pressure and steam, The quali,fication for service conditions such as radiation and -hemical sprays ray be demonstrated by analysis (evaluation). In such cases the overall cualification is said to be by a combination of meth:ds. Fo11csing are two specific examples of ;rocedures that are considered accescatie. Other similar procedures r.ay also be reviewed anc 'esn: ac:e;tacle or h case by case basis.
- 1. Raciatier Caelificeciog - Some of the earlier tvo* tests eerformed for operating reactors did not include radiation as a service condition. In these cases the equipment may be show 3 be F-radiation qualified by perfonning a calculation of t~ne dose ex;e:ted, taking into account the time the equipment is required to remain fun:tional and its location using tie metnods described in Appendix S, and analyzf ng the . 'fect of the calculated dose on the materials used in the equipment (see Appendix C). As a ger.eral rule, the time required to remain fun:tional assumed for dose calculations should be at least 1 hour.
- 2. Chemical Scray Oualification Components enclosed entirely in corrosion resistant cases (e.g., stainless steel) may be shown to be qualified for a chemical environment ty ae, analysis of the effects of the carticular chemicals on the : articular enclo-su're materialss The effects of chemical spra./s on the pressure integrity of any gaskets or seals present shcuid be considered in the analysis.
14 6.0 Karoin 3EEE Std. 323-1974 e ines margin as the difference between the most stvere specified service con /itions of the plant and the conditions used in type testing to account for nonnal variations in comercial production l of equipment and reasonable errors in defining satisfactory perfomance. Section 6.3.1.5 of the standard provides suggested .* actors to be applied to the servien conditio'sn ti assure adequate margins. The f actor applied to the time equipment is required to remain functional is the most significant in ter s of tne additional confidence in cualification that is achieved by adding margins to service conditions when establishing test environments. For this reason, special consideration was given to the time required te remain functional when the guidelines for Functional Testing and Failure Criteria in Section 5.2 above were established. In additio .11 of the guidelines in Section 4.C .or establishing service conditions include conservatisms which assure margins between the service conditions specified and the actual conditions which could realistically be expected in a design basis event. Therefore, if the guicelines in Section 4.0 and 5.2 are satisfied,no separate margin f actors are required to be added to the service conditions wnen specifying test conditions. 7.0 Acin:: , Itplici: in the staff position in Regulatory Guide; 1.89 with regard to backfitting IEEE Std. 323-1974 is the staff's conclusion that the increcental improverer.t in safety from arbitrarily requiring that a specific qualified life be demonstrated for all Class IE equipment is not sufficient to justify the expense for plants already constructed i
.- and o;eca, ting. This position does not, however, exclude etuipment
_ - , _ _ , _ _ , _ , _ . , _ , _ ._-__m..., _ ,. ._,,m. , . . . _ _ . . . . - - . _ , _
( , using materials that have been identified as being susceptible to , significant degradation due to thertal and radiation aging. Component maintenance or replacement schedules should include considerations of the specific aging characteristics of the component materials. Ongoing programs should exist at the plant to review surveillance and maintenance records to assure that equipment which is exhibiting age related degrada-tion will be identified and replaced as necessary. Appendix C contains a listing of materials whien may be found in nuclear power plants along with c .1 ind'. cation of the material susceptability to thermal and radiation aging.
- 8. 0 gcu entation Comple 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 retcrds should describe the qualification method in sufficient detail to verify that all of the guidelines h0ve been satisfied. A simple vencer certification of compliance with a design soecification snould not be considered adequate. 8 O 1 (
,n . - _ - _ _ _ _ _ _ _ _
l.. o APPENDIX A
. , TYPICAL EOUIPMEh*T/ FUNCTIONS-NEEDED FOR HITIEATIO 17F A LOCA OR MSLB ACCIDENT c
Engineered Safeguards Actuation Reactor Protection Containnent Isolation Steamline Isolation Main Feedwatee Shutdown and Isolation _ Emergency Power
. Emergency Core Coolingl Contair. ment Heat Renoval Containment Fission Product Removal Containment C'ombustible Gas Control Auxiliary Feedwater Con'a r :.:f' Ventilation C'a a i e n . Radiat'7n Monitoring Cor'-i: R.:< a Habitability Systems (e.g.. HVAC, Radiation Filters) .
Ventila.i.n for Arcas Containing Safety Equipment Component Cooling Service Water Emergen:y Shutdown 2 Post Ac:ident Sampling and Monitoring 3 Radiation Monitoring 3 Safety Related Display Instrumentation 3 l l _ _ _ - - _ - _ - _ _ _ - _ - _ - _ _ _ _ _ l
[ . These systems will differ for PWRs and BWRs and for old r and newer plants. In each case the system features which allow fo transfer to recirculation coolieg mode-and establishment of long term cooling with boren erecipitation control are to be considered as part of the system a be evaluated.
' 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 pressurr, boundary together 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 volums control system, and steam dump systems.
3 More specific identification of thesa types of equipment can be found in the plant emergency procedures. 9 j 1
. APPENDIX B - PROCEDURES FOR EVALUATING GAK4A RADIATION SERVICE CONDITIONS Introduction and Discussion The adequacy of gama radiation service conditions specified for inside containment during a LOCA or MSLB accident can be verified by assuming a conservative dose at the containment centerline and adjusting the dose according the plant specific parameters, The purpose of this appendix TLs to identify those pare neters whose effect on the total gamma dose is easy to quantify with a high degree of confidence and describe procedures which may be used to take these effects into consioeration. The bases for the procedures and restrictions for their use are as follows: (1) A conservacive dose at the containment centerline of 2 x 107 RACS for a LOCA and 2 x 106 RADS for a MSLB accident has been assumed. This assumption and all the dose rates used in the procedure out-lined below are based on the methods and sample calculation described in Appendix D of NUREG-0588, " Interim Staff Position on Environmen+al Qualification of Safety-Related Electrical Equip-ment." Therefore, all the limitations listed in Appendix D of NUREG 05SS apply to these procedures. (2) The sample calculation in Appendix D of NUREG-05SS is for a 4,000 2%'th pressurized water reactor housed in c 2.52 x 10 6 ft3 contain-ment with an iodine scrubbing spray system. A similar calculation-without iodine scrubbing sprays would increase the dose to equipment 7 approximately 15%. The conservative dose of 2 x 10 RADS assumed 5
2 in the Oroced;te below includes sufficient conservatism to < account for tr.is fact:r. Therefore, the pro idure is also applicable to plants without an iodine scrubbing spray system. (3) Shielding cal:ulations are based on an average gamma energy of 1 MEV derived fror TIC 14E44, (4) These pro:edures are not applicable to equipment located directly above the centainrer.t semp, submerged in contaminated liquids, or near filters. Doses specified for equipment located in these areas mus: be evalucted on a case by case basis. . ( 5 ', Since tr.e dese adjustrent factors used in these procedures are based or t ca.:ula:ior for a typical pressurized water reactor with a dry type cor.: air. ment, they are not directly applicable to boilir.g water reac: cts or other containment types. However, deses f:r these o:ner plan: configurations may be evaluated usir.g sim'lar procedures with conservative dose assumptions and adjus:nen factors developed on a case by case basis. Pr:cedure Fi;ures ! through a previde factors to be applied to the conservative dose to corre:: the dose fer the following plant specific parameters: (l' reactor power level; (2) containment volume; (3) shielding; (4) corsartment volune; and (5) time equipment is required to remain functicnal.
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.3. (
i g . The procedure for using. theLfigures.is best illustrated by an example, 4: Consider th'e following.. case.~ The radiation service condition for a I particular item of equipment 1has been specified as 2 x 106 RADS. The l application specific parameters are: 1 b Reactor power level - 3,000 MWth containment. volume - 2.5 x 106 ft 3 Compartmerit Volume - 8,000 ft3 3 Thickness of compartment shield wall (concrete) - 24" ,
, - Time ecuipeent is required to rema'in functional - 1 hr.
! The-prcblem-is to make a reasonable estimate of the dose that the equipment , could be- expected to receive in order to evaluate the adequacy of the y Lradiation service condition specification. Steo' 1
- Enter the nomocram in Figure 1 at 3,000 MWth reactor power level' end - -
? 2.5.x 10 6ft3 containment volume and' read a 30-day integrated dose of.
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- Enter Figure 2' at a dose of '1.5 x 107 RADS and 24"J of cencrete shielding L
4 for the compartment the equipment is located in and read 4.'S x".10 ' 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 lQ 4 M ' (Step 3). Sten 3 ? k k Enter Figure.3 at 8,000 ft3 and read a correction factor of 0.13. The
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dose due. :o sources 1inside the compartment would then be 0.13 (1.5 x 107 ) -
= 1.95 x'10 6 ,? ADS. The sums of the doses from steps 2 and 3 equals:
p h- 4.5 x 10 4RADS + 0.13 (1.5 x 10 7) RADS = 2.0 x 106 RADS E n .- . .,: .- ~, ,- . - ..,,,,,-- , . _ . - ,,- - . .; - ... .., ,,
4 Steo 4 Enter Figure 4 at 1 hour and read a correction factor of 0.15. Apply this factor to the sum of the ' doses detenmined from-steps 2 and 3 to-
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- correct _the 20 day total dose to the equipment inside the compartment to 1 hour.
0.15 (4 0 x 106 ) = 3 x 10 5 RADS 6 In th'is carticular exarple the service condition of 2 x 10 RADS specified is censervative with respect to the estimated dose of 3 x 5
- 10 RA:S calculated in steps 1 through 4 and is, therefore, acceptable.
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*MSL3 ACCIDENT DOSES SWO';Lu SE READ AS A FACTOR OF 10 LESS
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? ~~= OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 February 29, 1980 IE Bulletin No. 79-01B Supplement Information ENVIRONMENTAL QUALIFICATION OF CLASS IF EQUIPMENT Enclosed are the generic questions and answers which resulted from NRC Task Group / Licensee " workshop" meetings held recently in NRC Regional Offices regarding IE Bulletin No. 79-01B. This information is intended to further the understanding of the qualification review process and reporting requirements of 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's responses may generate additional need for guidance of finalized resolution of the environmental
. qualification issue.
..=.
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Enclosure:
Generic Questions and Answers . to IE Bulletin No. 79-01B 32= 5:.
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? ;.==.. GENERIC QUESTIONS AND ANSWERS TO IE BULLETIN No. 79-01B '
Question 1 IEB 79-01C indicates the scope cf the task is only that equipment exposed to a harsh environment. Enclosure 4, Sectica 4.3.3 '.dentified areas outside of cnntaintrent not exposed to harsh environmental conditions as tha results of an accident. Should these areas be included in our evaluations? Answer I No. Although the guidelines encompass all safety-related electrical equipme.t and cnmponents, the scope of IEB 79-01B is limited to only that electrict1 equipment which is exposed to the harih environments ictatified in action item 1, including where fluids are recirculated from inside containment to accomplish long-term cooling following a LOCA. All equipment and components idertiffed in activn item 1 shall be included in the subsequent actions :equired by IEB 73-01B. Quescion 2 IEB 79-01B action itm 1 and Enclosure 4 indicate that amergency procedures be uscC to identify equipment to be includec in the master list. Should all the equioment identified in the emergency procedures be included in the master list? Answer 2 All the equipment the licensee relies upon in the emergency procedures to mitigste design basis events that may be
..=G: exposed to a harsh environment must be identified in response
~
to Question 1. It is not the intent of this task tu change, the existing procedures by removing references to equipment or exponents 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 will not be misleading 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 enviionment 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 H:..... is not needed to achieve safe snutdown to meet licensing
; ~ requirements applicable to your facility.
. . _ _ -. . _ . -_ __ _ ~ . . . . _ _ _ _ _ _ __ _ _ _
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- Question 4 What is tiie basis for the 340 Degrees F for 6 hours requirement identified in En:losure 4 and NUREG 0588, Figure C 17 Answer 4 For minimum high temperature conditions in pressure suppression type containments, wo do not reouire that 340 Degrees F for 6 hours be used for SWR drywells or that 340 Degrees F for 3 hours be used for PWR ice condenser lower compartments. These values are from a bounding high temperature profile 0,ee Section 1.1 and 1.2 of NUREG.0588) that can be used in lieu of a plarit" specific humidity conditio, nsprofile, provided as a function of time that are expected pressure accounted for. and In general, the containment temperatt.re and pressure conditions as a function of time should be based on analfses in the FSAR.
However, these conditions should bound that expected 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 .he spray on required equipment should be accounted for. Question 5 Should equipment or systems which are presently planned to
'_ ..m4 bemodifiedasa-resultofactions,suchaslessonslearned, to be included in the response to IEB 79-01B?
\
c Answer 5 No. Question 6 Will there be any other actions required when the NRC completes its evaluation of the responses to IEB 79-01B and anf related corrective actions deemed necessary are complete? Answer 6 The NRC staff does not foresee.any additional actions for the electrical equipment and components which are included-in the scope of IEB 79-01B; however, if new problems or concerns are identified, appropriate-action will be taken. Question 7 Are Spare Parts required to meet 79-01B? Answer 7 Yes. The-Spare Parts are required to meet the same criteria
'as the installed electrical equipment or component er'sulting from the evaluation of IEB 79-018.-
Question 8 The instruction sheet for Enclosurc 3." System Compenent Evaluation" indicates that outstanding items be identified. What is the definition of outstending items? J::7.:1. r - * . - ~ o . - r -s -r-, , -
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qq= Answer B' An outstanding item is de'ined as that item that does not meet the envircraiental qualification guidelines and requirements of IEB 79-01. Question 9 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? Answer 9 Yes. Question 10 NUREG 0588 is out for comment. Does the staff expect any significant changes which may impact this effort? Answer 10 No. Question 11 When it is determined, as a result of the efforts required by IEB 79-01B, that specific equipment be upgraded, are the guidelines in Enclosure 4 to be used? Answer 11 As a minimum the same requirements that were used to determine the acceptability of the electrical equipment and components within the scope of IEB 79-01B may be used; however, if
-equipment is available which meets the requirements of .eE==E IEEF. 323-1974 it should be used.
E Question 12 Does the . Licensee Event Report (LER) requirements of IEB 79-01B supercede or change tne reporting requirements already defined? Answer 12 No. The requirement for reporting in IEB 70-01B does not change the reporting requirements defined in the license conditions. Question 13 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? Answer 13 When a determination has been made that the existing data is inadequate or no data exists to have reasonable assurance that the Class IE electrical equipment components can perform their safety-related function required in the specified FSAR environments, that is reportable per IEB 79-01B. The time and technical judgments required to make the determination should be based on the significance of the specific equipment, components and the discrepancies, e i I l
(: -4' p.=t.;c Question 14 Are the results of an evaluation using the materials identified in Enclosure 4, Appendix C, of an acceptable method of addressing the effects of aging within the scope of Bulletin 79-01B7 Answer 14 Yes, for those materials on the list, however, Appendix C indicates this is a partial list. Your evaluation in response to IEB 79-01B may identify other ma*.erials that are susceptible to significant degradation. Question 15 What are the sources Appendix C-3 used to identify the materials in Table C-1 and establish the failure levelt? Answer 15 Typical sources for the informatisn are given in Aopendix C. Your information of materials not in this table should identify the source for your evaluation. Question 16 Is additional effort or calculatiras required for radiation service conditions if previous efforts did not utilize the methodology or assumptions identified in NUREG 0586? Answer 16 Yes, the extent of the effort required will be dependent'on the significance of the difference in methodology G7D;; and assun.ptions. Qcestion 17 Will extension of time be granted for schedules if identified in IEB 79-01B action item 7. dnswer 17 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 thess schedules should be identified to the Regional Office by a written request. The NRC staff will rake a determination on & case-by-case bisis.
~ .4L:
IE Bulletin No. 79-01B Enclosure
- ((l; Supplement Information February 29, 1980 RECENTLY ISSUED IE BULLETINS Bulletin Subject Date Issued Issued To No.
80-04 Analysis of a PWR Main 2/8/80 All PWR reactor facilities
. Steam Line Break With holding OLs and to those Continued Feedwater nearing licensing Addition 80-03 Loss of Charccal From 2/6/80 All holders of Power Standard Type II, 2 Inch, Reactor Ols and cps Tray Adsorber Cells 80-02 Inadequate Quality 1/21/80 All BWR licenses with
- Assuran:e for Nuclear a CP or OL 80-01 Operability of ADS Valve 1/11/80 All BWR power reactor Pneumatic Supply facilities with and OL 79-01B Environmental Qualification 1/14/80 All power reactor iEEEE of Class IE Equipment facilities with an OL 79-28 Possible Malfunction of 12/7/79 All power reactor Namco Model EA 180 Limit facilities with an
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Switches at Elevated OL or a CP Temperatures 79-27 Loss Of Non-Class-1-E 11/30/79 All power reactor Instrumentation and facilities holding Control Power System Bus OLs and to those During Operation nearing licensing 79-26 Boron Loss From BWR 11/20/79 All BWR power reactor Control Blades facilities with an 0L 79-25 Failures of Westinghocse 11/2/79 All power reactor BFD Relays In Safety-Related- facilities with an OL or CP Systems 79-17 Pipe Cracks In Stagnant 10/29/79 All PWR's with an (Rev. 1) Borated Water System At OL and for information PWR Plants to other power reactors I
~: 1
SSINS: d Acccssion No.: IEB 79-01B, Supple. 2 tlNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 September 29, 1980 IE Supplement No. 2 to Bulletin 79-018: ENVIRONMENTAL QUALIFICATION OF CLASS 1E EQUIPMENT En:iosed are the generic questions and answers which resulted from NRC/ Licensee meetings in NRC Regionel Offices during the week of July 14, 1980 regarding environmental qualificatien of Class 1E equipment in use at power reactor fscilities. Tnese 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 79-01B review was not clear to licensees. Therefore, these scswers may affect your 79-01B submittal. These submittals are required by t separate order to be completed by November 1, 1980. Some answers given in Supplement No. I to IEB-79-01B 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 learnea 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 Questien No. 21 of this Supplement.) No specific response is requested Ly 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 79-01B. IE Bulletin No. 79-01B was issued under a blanket GAD clearance (B180225 (RD072); clearance expired July 31, 1980) specifically for identified generic problems. Supplement No. 2 to Bulletin 79-01B is for information, hence no GAO clearance is required.
Enclosures:
- 1. Generic Questions and Answers to IEB-79-01B and Memorandum and Order (CLI-80-21) cated May 23, 1980
.. *b as g f. e. 0.
Q.1 Define the scope of review with respect to the June 1982 deadline. What is required beyond the June 1982 date for qualification? 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 DDR guidelines or NUREG-0585 (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 ass.ure functioning in mild environments must be completed by June 30, 1982. The qualification schedules for consideration of the dynamic loading cf safety-related equipment (electrical and mechanical) and the environmental qualification review of mechanical equipment are being rP ' loped. It is the intention of the staff to initiate this effort as sa n 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 requirec schedule for submitting infurmation in response to the Commission Order and Memorandum (CLI-80-21) is provided below. Plants who have received an operating license, either for full or liraited power operation, are required to mett the schedule for operating reactors. Plants who have committed, to the NRC, to meet schedules in advance nf those provided below are required to meet that commitment. In all cases, plants are required to hhve their equipment fully qualified to the applicable standards either by June 30, 1980, or by the time the operat' g license is granted, whichever comes later. 00eratinc 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 received no later than 4 months prior to issuance of operating license OLs and cps (operating license expected af ter June 30, 1982)
- Submittal to be receivM m later than 6 months prior to issuance of operating license.
l l
5 g ' r 2-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 af ter--July.1974. Can a CP whose SER is prior to 1974 use the DDR guidelines? A. 3.- Table-1-describes the-application of-each document. _All operating reactors 'as of liay 23, 1980, will be_ evaluated against the DDR guidelines. In cases where the 00R guidelines do not provide
'* sufficient detail, but_ NUREG-0588 Category _II_ coes, NUREG-0588 will '
be used. TABLE 1 t. REQUIREMENTS ) 3 Ols cps >= ors LOOR GUIDELINES CP SER CP SER-Before-7/1/74 After 7/1/74
-USE NUREG-0588 NUREG-058B(CAT.II) NUREG-0588(CAT.I) NUREL.s588(CAT.I) 4 -AS-NECESSARY or NEW RULE-WHEN REPLACEMENT COMPONEN'S USE NUREG-0588 (CAT.I) IN EFFECT.
All plants licensed after May 23. 1980, shall conform to NUREG-0588. In accordance with-Regulatory Guide 1.89, all- such opcrating' licenses s for facilities whose construction permit SER-is dated July 1, 1974 or later,Eare to'be reviewed against IEEElStd.- 323-1974. Thus, for-these licensees, the operating license applicant is to. qualify. equipment to^ the' Category I' column- ini NUREG-0588.- For operating: licenses issued after May 23,-1980, whose construction permit?SER is-
~
dated before July 1,1974,. the operating'. license- applicant is to qualify' equipment to at_ least- Category II column of NUREG-0588; unless the' licensee:made commitment J in the construction permit record to use-the 1974 standard, or unless the-operating-licensee-applica-
-tien 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 DDR 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 of-the two documents are equally
~
applicable to ors and NT0Ls.
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- Q.4 Clarify the reporting requirements for LERs with respect to Part 50.55e vs 79-01B.
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Are only those items, known to be unqualified, immediately reportable? Are items, for which there are no data or for w?.ich there are insuf-ficient data, open items to be resolved, but are not immediately reportable? A. 4 The requirement for reporting in IEB 79-01B does not change the reporting requirements defined in the license conditions. In general, cps should report via 50.55e. Operating plants should use the LER. When a determination has been made that reasonable assurance does not exist to ensure 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 sh:uld be based on the significance of this specific equipment, components, and the discrepancies.
~ ~
.Q. 5 How does the "Q" 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 The "Q" list 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 potentially exposed to a harsh environment rcsulting from an accident. Safety-related equipment are those required to help bring the plant to cold shutdown and to mitigate the accident (LOCA, HELB inside or outside containment).
" Mitigate" includes safety-related functions such as containment isolation, and prevention of significant release of r6dioactive material.
In order to " pick out" 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 outside 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 licensee's list. The systems and equipment identified above should be incluced 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 tF.e safety function and qualifiable equipment may be identified. In such cases, the emergency ptocedures must clearly indicate how the
9 operator is to use environmentally qualified safety-related display instrumentation to diagnose failure to perform such safety functions. Plent 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 When this is the case, the radiation prevent racioactive releases. sensors, valve position indicators, pressure transmitters, thermo-couples, etc. , should be cualified to function in tue relevant accident environment. Licensees should, therefore, review their emergency procedures to determine the electrical components needed to perform the functions of Safety-Related Display Information, Post Accident Sampling and Monitoring, and Radiation Monitoring. When equipment implied by th*e 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 Equipment which is or will be Lessons Learned rhould be listed. 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 bo 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 list of equipment in a potentially hostile environ-ment. Conclusions which show that equipment is unqualified should
;nclude a basis for continued plant operation.
TABLE II TYPICAL EQUIPMENT / FUNCTIONS NEEDED FOR MITIGATION OF A LOCA OR MSLS ACCIDENT Engineered Safeguards Actuation Reactor Protection Containment Isolation Steamline Isolation M'ain Feedwater Shutdown and Isolation Emergency Power
" Emergency Core Cooling Containment Heat Removal Containment Fission Product Removal Containment Combustible Gas Control Auxiliary Feedsater Containment Ventilation Containment Radiation Monituring 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 (1) 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 shetdown condition following accidents which do not result in a breach of the reactor coolant pressure boundcry together with a rapid depressurization of the reactor coolant system. Examples of such systems and equipment'are the RHR system, PORVs, RCIC, pressuri?ir sprays, chemical and volume control system, and steam dump systems. (3) More specific identification of these types of equipment can be found in
-the plant emergency-procedures.
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 anc invertors 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.
3 .
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9 Q.8 Provide the staff's definition of " central location" 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's 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 DDR guidelines and NUREG-0588 specify that sufficient information must be available to verify that the safety-related electrical equiprent has been qualified in accordance with the guidance and requirements. Details for the information and documentation required for type' tests, operating experience, analysis, and extrapolttien 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's central file which reference the actual' test reports and data available in a single location at the NSSS vendor's facility. The Licensee / Applicant must make the determination that necessary
-information and documentation, to support qualification of equipment, is in conformance with 00R 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. y
.Q.9 The staff was directed to codify, by Tech,1 cal 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?
i A.9- 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. Q.10 With respect to the NRC data base, how will utilities address and
- obtain information from it?
A.10 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 *2981. Licensees will be informed at that time. How should submittals containing data and qualification information Q.11 be submitted? What format should we use if we have several facili-ties at different stages (OR, NTOL, CP)? U ' _ _ . _ . . ,
' 11 The qualification information and data should 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 79-01B 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 (E.g., function within the first minute (s) of a LOCA or HELB? nuclear instrumentation or other instruments providing RPS inputs, isolation valves, etc.) A.12 The staff does not require that the nuclear instrumentation 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 require that equipment designed to perform its safety-related function within a short time into an event be qualified for a period of at least 1 hour in excess of the time assumed in the accident analysis. The staff has indicated that time is the most significant f actor 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 I hour is based on the acceptance of a type test for a single unit and the spectrum of Also accidents (small and large breaks) bounded by the single test. see answer to question 21. Q.13 Testing is currently being performed on some equipment, and contracts hsve been issued for testing additional equipment specifying confor-mance to IEEE Std 323-1971. For sequential testing, how do we factor in aging? If early test failure occurs due to "non E-Q" mechanisms, can the test be extrapolatao using analytical methods? A.13 Sequential testing requirements are specified in NUREG-0588 and the DDR guide 11..es. Licensees must follow the test requirements of the applicable document.
- 1. If the test has been completed without aging in sequence, justification for such a deviation must be submitted.
- 2. If testing of a given component has been scheduled but not initiated, the test coquence/ 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 f all 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
.g.
failure mechanism, when the f ailure 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 contair. ment? A.14 Harsh environment is defined by the limiting conditions, as specified in IE Bulletin 79-01B, resulting from the entire spectrum of LOCAs HELBs. Specifically, the harsh environment from a LOCA coasiders the worst parameters resulting over the spectrum of postulated break rizes, break locations and single failures. Similarly, the HELSs inside and outside of containment consider the spectrum cf breaks including main steam and feedwater line breaks. The parameters to
~
be considered are: temperature, pressure, humidity, caustic sprsy, Mecha cal radiation, duration of exposure, aging and submergence. and flow-induced vibrations and seismic effects will be consic red separately. 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 listed above must be considered. Acceptable engineering methods Temperature and pressure should be used for this calculation. The radiation history may be available from earlier HELB evalations.Further guidance source terms are discussed under Question 18 below. for selecting the piping systems and conducting the review are delineated in Regulatory Guide 1.46 and Standard Review Plans 3.6.1 and 3.6.2. The DDR Guidelines and NUREG-0588 give time and temperature Will Q.15 parameters. Can we use different values of these parameters? plart-specific profiles still be with the guidance provided? For minimum high temperature conditions in pressure-suppression-type Q.15 containments, we do not require that 340 F for 6 hours de used for SWR drywells or that 340 F for 3 hours 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 However, function of time should be based on analyses in the FSAR. these conditions should bound those espected for coolant and steam line breaks inside the containment with-due consideration of analytical uncertainties. The steam line break condition should the peak temperature, and subsequent include superheated conditions: 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.
4 The adecuacy of a plant-specific profile is dependent on the assump-tion and design considerations at the time the profiles were developed. The DDR guidelines and NUREG-0588 provide guidance end considerations required to determine if the plant-specific profiles en:ompass 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 3EEE Std. 323-74 requirements? A.16 The QA programs should accommodate any increased scope due to the new cnvironmental qualification documentation requirements. Proce-dures incorporated by the licensee for data acquisition should be documented and available for sta'f review upon request. Requirements for QA programs are provided in Part 50, Appendix B, of the Code of Federal Regulations. Part 50, Appendix 8 of the Code cf 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 o" 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 " replacement" and " spare" parts are the same for the purposes of complying with the Commission order and
\
'i memorandum. After May 1980, all parts used to replace presently .
installed parts shall be qualified to Category I of NUREG-0588 ,
"unless there are sound reasons to the contrary." Nonavail2bility 1 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 " sound reasons -! to the contrary." All replacement parts shall as t minimum conform to the requirements described in the answer to question 3. Justifica-tion for deviation from Category i or NUREG-0588 shall be documented by the licensee and. records shall be available for udit,_upon ' request by the NRC. Q.18- 00R Guidelines, NUREG-0588 and NUREG-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 00R 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 roethods consider the radiation source term resulting from an event which completely depressurizes the primary system 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 equiement in close proximity to recircuiting 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 thosa avents which provide the most-bounding conditions. The-following table summarizes these considerations: . LOCA NON-LOCA HELB NUREG-0578 NUREG-0588
. Outside Containment (100/50/1 (10/10/0 in RCS) in RCS) 4
-,. ~ __
._,-7, . , .m+ ,,. , ,
.. e c 11 e
Inside Containment larger of NUREG-05BB NUREG-05BB (100/50/1 (10/10/0 in containment) in RCS) or NUREG-0578 (100/50/1 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 (J seta exposure have been included in accordance with the guidance given in the DDR guidelines and NURcG-0588. Cobalt 60 is one acceptable gamma radiation source for environmental qualification of safet; 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 DDR 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 equiprint could be considered exempt from that portion (submerger 1) of qualification, Q.21 Wr.at qualification is required of Reactor Pressure Vessel internal instrumentation (e.g., thermocouples) and new instruments required as the result of THI Lessons Learned? A.21 THI Lessons Learned instrumentation will be consido ed in the February 1,1981 SER. This equipment is subject to the same require-ments as other safety-related electrical equipment. The guidancs ( and requirements of NUREG-0588 referenced daughter standards, and Reg Guides will be useo 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 "by use" an acceptable method (e.g., CRDH's in BWRs)? Qualification by use has limited application. Often the equipment A.22 has never seen the harsh environment and no conclusions can be drawn as to its operability in a harsh environment. Some qualification l I
12 - i based en operating experience is a recognized method subject to the l requirements of NUREG 0588 and the Guidelines. Credit can be taken l for the natural aging of the equipment and for the location of the j equipment or other portions of the overall qualification information. l Q.23 dow 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 "long term" for qualifi-cation of equipment should be based on the considerations listed below for each postulated accident scenario. Justification for the value used should be provided with the equipment qualification documentation.
- 1. The tira period over which the equipment is 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.?4 Why do we want component surface temperature rather than the bulk environment temperature? A.24 Temperature measurements are required during the qualification testing to establish that the component was subjectedThese to thetemperature most severe temperature environment postulated to occur. measurements are required to be made as close to the component surface as practicable to ensure tnat 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. 9 4 4
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~ IEB 79-01B, Supple. 2 ,
September 25, 1980 RECENTLY ISSUED IE BULLETINS
, i 4
Bulletin Date Issued Issued To No. Subject Automation Industries, 9/11/80 All radiography 80-22 licensees Model 200-520-008 sealed-source connectnrs 8/29/80 All BWR power 79-26 Boron loss f rom BWR Revision 1 control blades facilities with an OL Failures of Westinghouse 7/31/80 To each nuclear 80-20 power facility in Type V-2 fpring Return to Neutral Control Switches your region having an OL or a CP 7/31/80 All nuclear power 80-19 Failures of Mercury-Wette6 Matrix Relays in facilities having Reactor Protective Systems eithe: an OL or a CP of Operating Nuclear Power Plants Designed by Combus-tion ingineering Maintenance of Adequate 7/24/80 All PWR power reactor 80-18 f acilities holding OLs Minimum Flow Thru Centrifugal Charging Pumps Following 'and to those PWRs Secondary Side High Energy nearing licensing Line Rupture Failures Revaaled by 7/22/80 All BWR power reactor Supplement 2 facilities holding OLs to 80-17 Testing Subsequent to Failure of Control Rods-to Insert During a Scram at a BWR 7/18/80 All BWR power reactor Supplement 1 Failure of Control Rods facilities holding OLs to 80-17 to Insert During a Scram at a BWR 7/3/80 All SWR power reactor 80-17 Failure of Control Rods facilities holding OLs to Insert During a Scram at a BWR Potential Misapplication of 6/27/80 All Power Reactor 80-15 Facilities with an Rosemount Inc., Models 1151 and 1152 Pressure Transmitters OL or 0 fP with Either " A" or "D" Output - Codes 0
f. ~t S$1Nb No.: betu Accession No.:
- 8008:20248 IEB Sup #3 to 73-01B !
- UNITED STATES NUCLEAR REGULATORY COMM15510N OFFICE OF INSPECTION AND ENFORCEMENT '
W SHINGTON, D.C. 20555 October 24, 1980 IE Bulletin No. 79 01B Supplement No. 3: ENVIRONMENTAL QUALIFICATION OF CLASS IE EQUIPMENT Description of Circumstances: Two issues were raised by Supplement No. 2 which require clarification. The' e are: (1) the dates required for submittal of qualification information for TMI 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 Pot Safe Shutdown condition. . (1) Supplement No. 2 (Q.1, Q.5) address'ed the minimum cold shutdown require-ments. The staff oosition on this issue is that the licensee must identify and envir'onmentally qualify the equipment needed to complete one method (path) of achieving and maintaining a cold shutdown condition. The equipment of o' er 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. I and Supplement No. 2, the staff position on this issue was uncitar. Therefore, the following will apply: .
- a. The qualif b ation information for equipment needed to achieve and maintain a at 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 79-01B required a 90 day response which was due in mid-M ril 1980. Supplement 1 (Feb. 1980) informed licensees that equipment which was
" planned" 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.5, Q.21) identified that W stdf posi-tion was that equipment which is installed should be treatec in a manner similar to all other safety-related electrical equipment and De addressed in the November 1, 1980 submittal. This position represents no change in staff position regarding the scope of the review. However, since the staff position on this issue was unclear the following etil apply:
- a. Qualification information for installed THI Action Plan equipment must be submitted by February 1, 1981.
- b. 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.
"4 IEB 79-01B Sup #3
. October 24. 1980 i'
. Page 2 of 2
) c. Qualification information for TMI Action Plan equipment currently under NRC review should be submitted as soon as possible. .
- d. Qualification information for TMI Action Plan equipment not yet installed which does not requite pre-implementation review should be submitted to NRC for review by the implementation date.
.The above itemi 1 and 2 represent no change in staff position regarding the scope of the 79-01B Supplement 2 review.
IE Bulletin No. 79-01B was issued under a blanket GAD clearance (B180225 (R0072), clearance expired July 31,1980) specifically for identified generic
. problems. Supplement No. 3 to Bulletin 79-01B is for information, hence no GAD clearance is required.
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- s. . IEB 79-01 Sup #3
. . Oc?.ober 24, 1950
- RECENTLY ISSUED
, IE BULLETINS Bulletin No. Subject Date Issued Issued To ,,
Supplement 2 Environmental Qualification 9/30/80 All power reactor to 79-01B of Class IE Equipment facilities with an OL 80-22 t.utomation Industries, 9/11/80 All radiography Model 200-520-008 Sealed- licensees source Connectors , 79-26 Boron loss from BWR 8/29/80 All BWR power Revision 1 Control Blades facilities with an OL 80-20 Failures of Westinghouse 7/31/80 To each nuclear Type W-2 Spring Return , power facility in to Neutral Control SwP m your region having an OL or a CP 80-19 itilures of Me,cury- 't n/et All nucleae power Wetted Matrix Relays in facilitien having Reactor Protective Systems either an OL or a CP of Operating Nuclear Power Plants Designed by Combus-
. tion Engineering 80-18 Maintenance of Adequate 7/24/80 All PWR power reactor Minimum Flow Thru Centrifugal facilities holding OLs Charging Pumps Following and to those PWRs Secondary Side High Energy nearing licensing Line Rupture Supplement 2 Failures Reveeled by 7/22/80 All BWR power reactor to 80-17 Testing Subsequent to facilities holding OLs f ailure of Control Rods to Insert During a Scram at a BWR Supplement 1 Failure of Control huds 7/18/80 All BWR power reactor to 80-17 to Insert During a Scram facilities holding OLs at a BWR 80-17 Failure of Control Rods 7/3/80 All BWR pover reactor to Insert During a Scram f acilities holding Ots at a BWR
,80-16 Potential Misapplication of 6/27/80 All Power Reactor Rosemount Inc., Models 1151 Facilities with an and 1152 Pressure Transmitters OL or a CP with Either "A" or "D" Output Codes c ,
e- ,1 $ - 53 d'3Y8M/'b Start Ihh,53 Z t yl72 - qsy;y
, . 92 tiAR 13 P12':06 y i n i vi ii D :t it." f r,0cstIon t.':! wirl bH A hl 6*
NUCLEAR POWER STATION bW QUALIFICATION TYPE TEST REPORT LIMITORQUE VALVE ACTUATORS XM* TOR PWR SERVICE I' PROJECT # 600456 4 Tested per IEEE Standard 382-1972 , Test Perfez:ned 7 June 1974 to 22 November 1974 Prepared by.Limitorque Corporation Test Laboratory >, Prepared by tL , l Walter L. Syksy Test n ineer Date 'lZa 9 ff Approved C W /J . Denkowski Mief Engineer Date [
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i ..- + PWR Qualification Project #600456 T TABLE or CONTENTS W [ i i 1.0 Introduction .................................page 1 f 2.0 Identification of sample valve Actuator ......page 1 3.0 Type Test Procedure .................. 2.....page 2 3.1 Aging simulation 1 3.1.1 Thermal Aging 3.1.2 Mechanical Aging i 3.1.3 Radiation Aging l 3.2 Seismic Qualification 3.3 Radiation Exposure . 3;4 Accident Simulation 3.4.1 Test Description 3.4.2 Test Procedure for LOCA Test - Tigure 5 - Specified Accident Profile 3.5 Post Test Inspection 3.6 Post-LOCA Load Cycling Test 3.7 Final Inspection a 4.0 Test Results .................................page 11 4.1 - Mechanical Aging ,
- 4. 2 - Seismic Qualification 4.3 Radiation Aging & Exposure 4
I 1
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i . I l PWR Qualification Project #600456 .j 070F18 TABLE OF CONTENTS (cont. ) 4.4 Accident Environmental Simulation Test Results 4.4.1 Temperature and Pressure Profile Figure 6 - Actual Accident Profile Figure 7 - 1st Transient PWR Profile Figure 8 - 2nd Tran=ient PRR Profile 4.4.2 Chemical Spray Delivery 4.4.3 Chamber Humidity
'4.4.'4 Insuldtfon Re,sistance Measurements
. Table II - Insulition Resistance of Power & Control Leads 4.4.5 Actuator Iycling Date Table III - Valve Actuator Cycling Data 4.5 Post LOCA Inspection Figure 9 - Post LOCA Conditions (photograph) 4.6 Post LOCA Load Cycling 4.7 Final Inspection Results 4.7.1 Motor Inspection and Dismantling 4.7.2 Valve Actuator Inspection and Dismantling Figure 10 - T*st Unit Final Inspection (photograph) 5.0 conclusion ........................................page 29 i
II
~
PWR Qualification Project #600456 4 APPENDICES APPENDIX A - Reliance Electric Company's Certificate of Ccmpliance APPENDIX B - Mechanical Aging, Load Cycling and Thrust Measurements Appendix C - Certificate of Compliance from Isomedix Corporation APPENDIX D - Lockheed Environmental Labs' Seismic Report No. 3521-4911
'APPERDIX E - Figure 1 - Test Chamber Figur's ? - Steam Generator Figure 3 - Control & Instrumentation Panel APPENDIX F - Figure 4 - Schematic - Instrumentation and Power System Table I Summary of Data Acquisition system III
.. . . . .- -. . - - . = - - - . . .- . .. . _ - . . . - - . .
. ... f e
PWR Qualification 1.0 090550 ; Introduction l, A typical Limitorque valve Actuator, SKB-0 with a 40 ft-lb motor (SMB-0-40) was submitted for qualification to the type test specified by IEEE std. 382 '72' for service in a gres- ' i surized Water geactor (PWR) containment chamber in Nuclear Power Station application.
. The SMB-0-40 Valve Actuator was subjected to mechanical aging simulatio.n- to approximate 40 years service life, radiation ex-posure (Both Normal Life levels plus accumulative doses) cr.d other environmental conditions all as indicated in IEEE Std. .
382 '72'.. Addition $1 load cycling wa's performed after LOCA environmental conditions to determine the post ac:ident abilities of the valve actuator.
. +:
2.0 Identification of Sample valve Actuator TEST UNIT A limitorque SMB-0 Nuclear valve Actuator with a 40 ft-lb-nuclear containment motor (RH Insulation Class) was con-
- l. structed per standard nuclear bill of mat;;ials and stan-L
! dard' nuclear meter' specifications. The following inform-ation'was taken'from the identification tags: 4 1.
. 4 e< ,
=PWR Qualification
- t valve Ac tuator Type / Size . . . . . . . . . . . . . . . SMB-0 070551 ;
Manufacturer ......................Limiterque Corporation order Number ...................... 600456-A i Serial Number ..................... 189835 Electric Motor Information ; Size . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 0 f t - Lb s t a l l , 8 ft-lb run M anu f a c tur e r . . . . . . . . . . . . . . . . . . . . . . Re li anc e E le c tri: Company Identification number ............ 2Y267074AlIZ Full Load Speed .................. 1735 PPM-Frequency ................'........ 60 Hz ,
. . y voltage....................,....... 4'60 Volts ,
Insulation Class ..................RH Type ..............................P f 3.0 Type Test Procedure The type test. plan as described in IEEE Std. 382 '72', para- _ graph 4, consists of-three basi: parts:
- 1. Aging simulation i
- 2. Seismic Qualification
- 3. Accident Environmental Simulation
-2.
PWR Qualification b2 s As an added test margin, the test actuator was submitted to additional load cycling after completion of all the required environmental conditions and prior to final inspection. This additional load cycling is not a re-
~
quirement of IEEE Std. 382 '72'. A base test motor was processed with the Test Unit for additional engineering information. 3.1 Aging Simulation- (IIEE Std. 382 para. 4.2) 3.1.1 Thermal Aging i Thermal aging was performed on the motor stators , by the motor manufacturer (Reliance Electri: .C,om-pany) in cooperation with Limitorque ' Corporation. t The' motor stator was heat aged for 100 hours at , o 180 C. A certificate of compliance was supplied by Reliance Electric Company verifying the thermal aging of the stator (see Appendix A.)L 3.1.2 Mechanical Aoing-Mechanical Aging was performed on the Test Unit' by the Limitorque Test Laboratory. Data on the Aging & Post test Cycling is presented in Appen-dix B. Although IEEE Std. 382 '72' requires 500 cycles, the unit was eyeled thru 1208 cycles. 3.
PRR Qualification 070553 i each cycle consisting of one close stroke and one open stroke at room ambient conditions. The Limit-orque valve Actuator was seated at the end of the close stroke and the seating thrust monitored. The thrust applied was equivalent to the thrust & torque ratings of the SHS-0 actuator. A .ypical stroko time of 40 sec. was chosen for the actuator operat-ing time. 3.1.3 Radiation Aging (IEEE Std. 382 '72' Part II Section 1)
. The A'ging dose of 4 Megarads was combined with the accident dose (200 Megarad) perfEEEStd. 382 '72' part III and is diz:ussed in the following section 3.3 of this report.
3.2 Seismi
Qualification (IEEE Std. 382 Para. 4.3)
- l The seismi Qualification was performed by Lockheed Ele:-
l tronics, Inc. Environmental Laboratory on a Reaction vibration machine. The Test Unit with motor, was scanned in each of ths three major axis over a frequency range of 5 to 35 Hz to search for resonance. No resonance was found. 4.
_~ . 1 PWR Qualification 090534 l: The valve Actuator was mounted on a test fixture.to provide simulated valve seating loads, during the i dwell portions of the seismic qualifications. The lead imposed was equal to the rating of the test l actuator. ! ! The vibration machine was adgusted to a displacement (0.050 D. A.) equivalent to 3 g's acceleration at a frequency of 35 Hz. The test sample was then vibrated for a period of ten (10)' seconds at each even integer of frequency from 6 Hz to 34 Hz. The unit was operated during the dwell through one cycle from open limit-to-torque switch seated position and back to original point. Thw vibration machine was adjusted to a dis-placement (0.100" D. A.) equivalent to 6 g's ac-celeration at a frequency of 35 Hz. The test sample l was then vibrated for a period of ten (10) seconds at 35 Hz and operated during the dwell. l The dwell tests above were per$ormed in each of the three major axis. A report on the seismic Qualifi-cation was prepared by Lockheed Electronics Corpor- l l ations Environmental Laboratory '(Report No.3521-4811 and ! j r L is presented in Appendix D.) The duration of each i stroke was 40 seconds. / l- . s.
"< ^ . .
PwR Qualification =( 070535 l 3.3, ; Radiation Exposure (IEEE Std. 382 Part II Section 1) ' i The Limitorque Actuator and motor were subsected to a Gamma Ray Irradiation of 204 Megarads per IEEE Std. ; 382 '72' requirements. The. Test Unit was placed in a Cobalt-60 and Cesium-13' field of 1 Mrad / hour at an air equivalent _ dose. A total radiation dose of 204- Megarads was applied af ter thermal aging, ' mechanical aging and seismic qualification. The radiation expecure_was performed by Isomedix Corpor-
. . ation. A Test Certification was supplied by Isomedix . i corporation and is presented in Appendix C. ,
3.4 Accident-Simulation (IEEE Std. 382 ' 72 ') , 3 '. 4 .1 - - Test Description The test was performed' at Limitorques' Environmental
-Test Facility, see figures 1, 2 and 3 in Appendix E.
A schematic of the instrumentation system and a
,;(
summary'of instruments used during the test are . presented in Figure 4 and Table I located in Appen-dix F. The limitorque Actuator was mounted on a thrust tube attached to the side of the test chamber with the stem thrusting against. the load cell mounted'
- 6. ,
4 4
- y
. l
'PWR Qualification .
070536 4 ext,ernally to the test chamber. (see Figure 1, l Appendix E) ;
. 1 I
Control and power lead connections were made l l through flexible pressure tight conduit con-nections run between the units and the access l ports of the test chamber. The external wire ; harness was run to a junction box, where ter-minal strips provided access to each lead for monitoring insulation resistance. The terminal strips were wired to a control system (see
. Figure 4, Appendix F).
, The control panel illus-trated in Figure 3,(Appendix E) contains a power
, monitoring system to monitor line voltage, :urrent in each of the three (3) motor legs and the power
- onsumption of the motor.
Pressure and temperature were monitored on the multipoint temperature recorder and: strip chart recorder mounted on the test console (Figure 3, Appendix E). In addition to the automati: moni-toring system, the temperature and pressure was monitored by a pressure gauge and two thermometers mounted in the side wall of the test chamber (see Figures 1 and 4.) 7
PWR Qualification t During the rapid te tperaturt and prersure tran-sients, the chamber ambient and limit switch com-partment internal temperature and pressure were monitored continuously on the strip chart re- , corder. , i cooling coils mounted inside the chamber provided cooling capacity to reduce the temperature in the . chamber to the various temperature plateaus. A double spray system provided a reliable source of chemical spray during the test. profile. Flow i 4 meters mounted on the panel near the test chamber (see Figure 1, Appendix E) monitored the chemical fluid flow. Spray nozzles mounted on two sets of manifold s (3 nozzles per manifold) with the ability to switch manifold provided the proper spray pat- , tern. The pressure in each active manifold =et was monitored to indicate any restriction of the spray nozzle orifice. A back flush system was provided to back-flush the spray manifold. . 3 . 4 '. 2 Test Procedure for LOCA Test [ The Limitorque Valve Actuator was exposed to steam and chemical spray in accordance with the I i h 8.
. l . . - . . . , . . - - - , - - - . - - . - . _ . - _ . . - . - . - - -.- - . . . . - . ..
070fiipg PWR Qualification I- criteria listed in Table 1 in the "IEEE Guide for Type Test of Class 1 Electric valve operators for Nuclear Puwer Generating Stations" IEEE Std. 382 '72'. The temperature / pressure profile is illustrated in Figure 5, which also shows the sche-dule for measuring the insulation resistance of the power and control leads and eyeling of the Limitor-que valve Actuators. Duripg the first four days of the test, the specified temperature and pressures were maintained by the con-trolled injection of steam into the test chamber.
. 0 .
During the remainder of the test, the 200 F/10 psig state was maintained by filling the test :hamber with air controlled to the proper pressure and using elect-rical heaters. The atmosphere was kept saturated with water vapor by maintaining condensate in the bottom of the tank and by daily injections of steam. 3.5 Post Test Inspection A visual inspection of the limit switch compartment and the limit and torque switches was performed at the conclusion of the accident simulation. 9.
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PWR Qualification 070F(30' i, , 3.6 Post LOCA Load Cycling Test Similar to that performed at pre-test mechanical aging, the unit was cycled for a total of 794 cycles (one close
& one open stroke percycle) at room ambient conditions (data supplied in Appendix B).
3.7 Final Inspection A complete physical inspection of the test samples was msde after the completion of the Post LOCA Load Cycling to observe the conditions of the actuator. 4.0 Test Results - 4.1 MechaniealAgihg The unit was initally tested on 7 June 1974 and a thrust output of 20,162 lbs. was obtained at a torque rwitch setting of 1-3/4. (This value is an average. of 24 readings.) The unit remained on the test stand and was automatically cycled at room ambient conditions. The cycling test was performed from 7 June 1974 to 10 June 1974 for a total of 1200 cycles consisting of one torque rwitch closure in each cycle. The load was measured after the completien of the mechan-
, ical aging and an average of 10 readings produced a thrust output of 19,920 lbs.
.:l FWR Qualifi cat ion II O!791 f i4 The' test data obtained is presented in Appendix C.
4.2 seismic Qualification The seismic Qualification was performed at Lockheed ! Electronics Environmental Laboratory on 12 June 1974. ; The data recorded is presented in Lockheed test Re-port No. 3521-4811. (Appendix D.) The thrust load was not monitored during seismic testing however, . thrust readings taken after seismic and radiation, 19,350, ave' rage of three readings, was within three (3)- percent. of- the post mechanical aging value. . The output characteristics did not change during seismi: , testing or irradiation. The valve actuator and its
. limit and torque switch functioned during seismi:
testing. , 4.3 Radiation Aging & Accident Exposure 1 The exposttre to radiation of the Test Unit- was per-formed e.n 18 July 1974 at Isomedix Corporation. A total dose'of 204 Megarads was used. A Test certi-fi=ation may be - found in Appendix - C. ,
/4.4 Accident' Environmental Simulation Test Results ,
ThetLOCA Test was performed at Limitorques' Enviro-mental Test Facility.- T.b - environmental test was , started 22 August '1974 and completed 21 September 1974. 1 N
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PWR Qualification
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R4.4.1 Temperature and Pressure Profile The profile specified in paragraph 3.2.5 of this report was closely followed as evidenced in Figure
- 6. The transient data was obtained by means of the strip chart recorder. At the transient time of ten (10) seconds, the temperature was a temperature with-o in 91% of_the specified temperature (300 F.) A o
temperature of 300 F was reached in 15.2 seconds. The second transient closely approximated the first o reaching a, temperature of 300 F in,13.8 seconds.
. Copies of*th'e actual strip chart data are presented in Figures 7 & 8.
After the' transient and a dwell of 30 minutes at ' o 300 F, the test ambient was brought to a stable o condition of 250 F and 30 psig. The actual temp-n.ature' conditions were within minue 2% and plus 6% of specified temperature and the pressure con-ditions-were within plus or minus 3.5%. These conditions were maintained for the balance of four (4) days. At a test time of 96.1 hours (approx. 4 days) the o test ambient was lowered to 100 F and 10 psig. The chamber was maintained at these conditions by means 13. {. ..: .
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PWR Qualification O O*W)b ( of strip heaters and air injection through a , pressure controlled solenoid valve. Onces stability was reached, the ambient was maintained within plus 0% and minus 4% of specified temperature and plus 10% and minus 0% of specified pressure. 4.4.2 Chemical Spray Delivery The chemical mixture (per Table 1 of IEEE Std. 382-page 12) was prepared prior to start of the LOCA test and pH values measured. Tank No. 1 had a pH
'of 10.9 a'fter initial mixing. . Tank No. 2 had a pH of 10.5 after initisi m'ixing.
- The pH was moni-tored en a sa. pie taken from Tonk No. 1 at a test time of 0.1 hours (pH=11.1) aad after 4.4 hours (pH=11.1) . A sample of Tank No. 2 taken at 24 hours had a pH reading of 10.5.
The chemical flew was maintained at 0.6 gal / min in each sprav J t:id or an overall flow rate of 1.2 gal / min. A check was made of the average flow rate -by recording the total amour.t of chemical solution used in a given period of time. These average flow rates agreed with the recorded in-stantaneous flew rates.
l . PWR Qualification C90fS7 ( 4. 4. 3 - Chamber Hamidity The relative humidity of the chamber was main-tained at 100 ' percent by the periodic injection of steam and by maintaining the vapor condensate at the bottom.' of the thsmber at the same tem-perature as the air / vapor mixtare. The content _ of air in the air vapor mixture was minimized 4 by venting the chamber during the thermal tran-sients. ,4 5 4.4.4 Insulation' Resistance Measurements
. Insuletion resistance m'easurements to freund were made periodically on the power and control leads of'the-l Test Unit prior to operating the valve actuator (see Table II.) -
4.4.5 operator Cycling Data The test unit functioned without problems through-out the entire test. It is worthy to mention that during the final operational cycle (719.1 hours) n - tno close indicating-light exhibited a very dim glow when it should have been extinquished. This phenomena was noticed only on the "close" light circuit and no other indicating lights or cir:uits 1
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TABLE II
'g Insulation Resist.ance of Power and Contro1D Leads -
(All resistances are in Megohms except where a .K. indicates ' Kilo-ohms) (all measurements made to ground). . Time ' A E CONTROL CIRCUIT LEADS -' ' ' , g LEA DS ' Tcst - 1 (hr.) di W2 T-3 &1 61 71 41 43A 43B. 43C 45A. 45B 51 53A 53B 53c 55A 55B' .
*O 400 ~460,: .4D0 180 180 180' 2000 180 180; 180 180 180 180 190 180' 100 ISO 180
,. 0.15 160K 160K 160K 300K 400K 400K 4Q.0 400K 400K 400K 400K 400K 400K 400K 400K 400K 400K 400K 0.5 120K 120K 120K 280K 280K 280K 5.0 200K 200K 200K 280K 200K 200K 200K 280K 280K 280K 200K 3.9 100K 100K 100K 2.0 2.0 2.0 6.0 2.0 2.0 2.0 2.0 2.0 2.0 2.0 2.0 2.0 2.0 2.0 j 4.2 50K 50K 50K 400K 400K 400K 2.4 400K 400K 400K'400K 400K 400K 400K 400K 400K 400K 400K 95.5 80K 80K 00K 2.0K 2.0K 3.0K 40K 2'.0K 2.0K 2.0K 2.0K 2.OK 2.OK 2.0K 1.3K 2.0K 2.0K 2.0K 334.9 60K 60K ' 60K 1.5K 1.5K 2.0K 5.0K 1.5K }.5K 1.5K 1.5K 1.5K 1.5K 1.5K 1.5K 1.5K 1.5K 1.5K 719.1 60K 60K 60K 2.0K 2.0K 3.OK 5.0K 1.7K 1.7K 1.7K 1.7K 1.7K 1.7K 1.7K 1.7K 1.7K 1.7K 1.7K o .
- Check prior to starc of tpst. 4 3
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PWR Qualification- . 07059{} Il exhibited thsse characteristics. Subsequent in-vestigation concluded that the current flow through this light to duplicate the dim glow was insignifi-cant and coupled with its occurrance in the final hour of a 30 day test did not constitute a mal-function. The megger rekis*gs diminished during tae environ-mental test and the current & power requirements did increase slightly as the test in the environ-
~
mental chamber continued; however, this had no , effe,ct on the actuator performance. The stroke time remained constant throughout the test. . Also a slight variation in the measured output , thrust was noted and was a**.ributed to a change - in stem efficiency rather than actuator output torque change. It was noted that during periods of non-operation, the thrust tended to become lower, whereas during periods of frequent operat-ions, the thrust increased. Probably, the ambient temperature & moisture effected the lubricity of the lubricant used on the stem. A summary of the cycling data is presented in Table III. 20.
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7 1 TABLE VALVE ACTIRTOR CYCLING IRTA ~ - Time Potential O' PEN STROKE -CLOSE STROKE Afher (volts) Run Current #: Run Current
- Start (Amps) (Amps)
M Power of gg _ M Es g.
. L" (Watt i ME8g tn -
.J 48g Test T--1 T-1 'F-2 T-3 T-2 T-3 T-l' P2 T-3 $4 p' 'S g, -
$m 'F-1 F2 %3 O $$
,0 m M3 (hr.) -
py . Run
- Peak pv to 30 n. **
**O 490 495 490 4.6 4.8 4.6 620 42 4.6 4.8 l4.4 620 1350 .'42 5.0 19,375 4
.3 475 480 480 4.5' 4.6 4.5 620 42 4.3 4.3 4.2 600 1300 5.1 43 19,425
.6 485 490 485 4.7 4.8 4.6 680 42 4.8 4.9 .4.5 5.1 680 1320 43 20,825 4.1 485 490 485 4.8 4.9 4.6 620 42 4.8 5.0 4.6 5.1 610 1300 43 21,600 4.3 490 490 400 4.0 5.0 4.7 650 47 4.8 5.0 640 1350 4.6 5.1 43 22,150-
.95.5 495 500 495 5.1 5.2 5.0 725 42, 5'2 5.3 4.9 5.3 750 1500 43 22,650 335.4 485 490 485 4.8 4.9 4.6 650 41 49 5.0 3 4.6 5.0 650 1400 42 21,600 ,
719.1 495 500 490 4.9 5.2 4.9 675 '42 5.0 5.2 4.7 5.0 675 1500 42 18,550 719.5 495 500.. 490 5.0 5.2 4.9 700 42
, 5.0,3.2 4.7 5.6 675 19b0 42 21,350
- Due to rapid rise of current and power, considering the slow meter response
- times, these values to be considered as approximation of actual. peak.
** Check prior to start of test.
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FWR Qualification 090571 4 4.5 Post LOCA Inspection The post LOCA Inspection was performed 21 September 1974 after ? opening the test chamber. Photographs were taken of the test unit with the limit rwit=h compartment =over in place (see Figure 9). Externally, the Test Unit was clean looking with no unusual de-posits. The-limit rwitch compartment had approxi-mately one-eighth (1/8) of an inch of condensate in the bottom of the compartment. Both ,the limi,t and torque swit=hes were clean and functioned without mechanical di'ffi=ulties. The motor lead protective sleeving was split in several areas; hdwever, no damage was noted to the motor lead insulation. 4.6 Post LOCA Load Cycling The post LOCA nad Cycling was performed by the Limitorque Test Laboratory from 30 September 1974 to 4 October 1974. The thrust output of the Test Unit was measured prior to the start of the load cycling. The thrust output.was found to be 16,392 (an average of 6 readings). This was accomplished at the same torque
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07057.1
'PWR= Qualification:
30 *
' switch setting (1-3/4) as that used throughout the test.
The . torque output of the actuator as controlled by the torque switch -remains constant with the same , torque switch setting, however, the thrust monitor , in the _ test stem depends upon several factors in-eluding the efficiency of the acme threads. The lower thrust output monitored after the unit was . brought to room . temperature was _ attributed . to a degredation of stem-efficiency as a result of corrosion of the stee,1 etem and' deposition of'for- ' eign materials- from the exposure to _the steam and hemical spray and not, attributable to changes in the torque switch operating _ train or reduction in the torque output-of the actuator. The effect of the corrosion was most noticeable after the stem was eAposed to room Embient con-ditions for several hours. After the completion of.the 794 cycles,;the thrust monitored returned
.to-its original value indicating the repeated cy -
cling had removed the corrosive deposits in the stem thread ar'ea. The cycling data and thrust reading are presented in Appendix B.
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- PWR Qualification'-
070571 3? , A Base Test motor which experienced- all the en- :
. vironmental conditions was installed on the test actuator . af ter the planned post IccA cycling to obtain cyclic information on the base motor and -
provide additional-load cycling on the test unit. , The -SMB-0 Actuator (with the base: test motor) was-cycled forlan additional 2184 cycles. , The SMB-0 Actuator functioned without difficulty _
/throughout-the additional 2184 cycles.
TA-_ summary of'the load cycles. accumulated on the test unit is'as follows: s
. Pretest Mechanical Aging . . . . . . . . . . 120,8 cycle s LOCA Testing cycles ................... 9 cycles .
Post LOCA Load cycling .............. 794 cycles Base Test *ctor cycling ........... 2184-cycles
- - TOTAL 4195 cycler-i.
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PWR' Qualification 070$73 Ii 4.7 Final Inspection and Dismantling 4.7.1 Motor Inspection- and Dismantling The motor (used with the test unit during LOCA test) mounted on the Test Unit was removed from the SMB valve actuator and , dismantled-for inspection. The inspect- - ion was performed on 21 November 1974 with representatives from Reliance Electric Com-pany in
- attendance.
-The rotor turned freely p ior to dismant-
. ling the motor. The stator and rotor showed little evidence of corrosive build-up and no evidence of physical damage. Thu and
~
bell was particularly clean with little evi-dance of water. The bearing lubricant. was moist and the bearing. turned freely. 4.7.2. Valve Control Inspection and Dismantling The SMB-0-40 valve Actuator was completely dismantled for inspection on 22 November 1974. Photographs of the valve actuator components are presented in Figure 10. The torque switch and limit switch were re-moved from the SMB-0 Valve Actuator and the oc
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4 w-PWR Qualification
.e
6+ following observations were made: a.) The torque switch and limit switches worked freely. b.) The torque switch and limit switch
= pinions both showed signs of lubrication.
The grer e in the main housing and the handwheel clutch comparti.snt was dark in color but maintained its - lubricity . A slight amount of separation of the grease was noted. The 0-Ring and bearings seemed
~
in good condition with no wear noted. 5.0 Conclusion- ,
. The Limitorque valya Actuator SMB-0-40 was*subje=ted to a qualification test. consist. ng of a 30-day exposure to a steam chemical environment, including two temperature &
pressure _ transients from 120 degrees F to 300 degrees F in;approximately-lO seconds. . Prior to environmental testing, the motor was heat aged, the unit was mechani-cally' tested .and subjected -to gamma ray irradiation.
~
The unit was cycled with simulated valve seating loads - during environmental testing at. elevated temperatures and. pres-sures and after environmental test was additionally cy-L cled with a simulated valve seating load. L
- 28.
L l . -
' PWR Qualification C^0578
(, since the~ unit performed satisfactorily, throughout the test, it is concluded _-this test qualifies similar-Limitorque
. valve Actuators for use in a PWR containment c5 =1nbar where environmental conditions depicted by Table I in IEEE Std.'.382 '72' are encountered.
e 9 4 6 9 0 4 e b 29.
- - - - - - ~ - - - - - - - - - - - - - - - - - - - - - - - - - ~ _
PWR Qualification gqf)gyg . APPENDIX A Relianca Elactric Company - Cartificate , of Complianca
- 0
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. ~ ~ ^ _ _ _ _ _ _ _ _ _ - - - - - - - - - ' - ' - ' - ' ' - - ' - - - - ~
070590 ( MELIANChr si.scrmc commy ] CERTIFICATE OF COMPLIANCE Limitorque Corporation N 5114 Woodall Road **- Lynchburg, Virginia 24502 L J *
+?,b'ab IQ 4
.N %<e 5
- EQUIPMENT: Electric Motor
REFERENCE:
Purchase Order No. 600426-C FILE: Sales Order No.
- 2T- 267074A1 We certify that the equipment identified above has been desigt,0d, manufac-t u re rt , inspected, and/or tested in accordance with the recuirements estae-lished by the following speci ficat ions : RCP-242, L1=1 torque D/S 2 001-1 We further certify that the stator was heat aged 100 hours at 180}C.
g Quality Control Departtrent 5 I
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o i E 3 2. . 4:~.0 ilE N Eri?
- CAD / a5mTABwa.A. *m:0 44004 iTELE8*CNE 216 992 2175
s.; :..
- PWR Qualifi::ation' 07()5531 h -
APPDOIX 3
.i
. *j
; INITIAL TORQUE SWITCH SETTING
..;< _ MECHANICAL. AGING'
+ __ .
}.' . POST MECHANICAL- AGING THRUST MEASUREMENT
- POST - -SEISMIC QLTALIFICATICN AND RADIATION THRUST MEASUREMENT
- POST ENVIRONMENTAL THRUST MEASURIMENT . .. -
POST' ENVIRONMENTAL LOAD CYCLING'
-POST' LOAD CYCLING THRUST MEASUREMENT
-i l. 32.
.; e
SUMMARY
OF LOAD CYCLING DATA ( _ O ON.3 Specimen -TEST UNIT. Limtorque Valve Actuator Type SMB Size 0-Serial No. 189435 Motor si*e 40-ft-lb I. D. #2Y267074AlEZ Instrumentation. Load Cell: Capacity 30, 000 pounds Manufacture'r BLH Serial No. 2512 Strain Indicator: Manufacturer BLH Type.N Serial No. 443604 INITI AL -TORQUE SWhTCH SEITING Date 6/7/74 No.-of Readings Torque Switch Setting Thrust Ou r '.'It * (pounr! 5 al' 11,070 S "1 " 16,010 24 "1 3/4" 20,162
- Average of all readings MECRANICAL AGING Dates 6/7/74 to 6/10/74 Definition: One (1) cycle open Limit actuation to close torque
. Switch actuation to open limit actuation. Two (7) strokes per cycle.
Stroke Time: 54 sec
- Cycle Time: 1 min. 53 see Duty Cycle: ' RUN ' 7 cycles - 'OFF' 10 min.
Load (Thru st) : 00,16' pound, i Total No. of Cycles: 1208 h 33.
. . = .-~ .
s.: 1 090593-A. fN,
SUMMARY
OF LOAD CYCLING DATA (continued) A.l -
'* ;Thetunit was: cycled =for mechanical aging on a different load stand:than was used-in-the test and since the stroke was rionger in this-stand,-a longer stroke time was obtained.
POST MECHANICAL AGING THRUST MEASUREMENT Date: 6/10/74 No.'of Readings-
. Torque Switch Satting
~
Thrust Output * (pounds) 10 "1 3/4" 19,970 POST SEISMIC OUALIFICATION AND' RADIATION THRUST MEASUREMENT Date: 8/19/74
- No. of Readings Torque Switch'Satting Thrust Output * *
(pound s) 3 "I 3/4" 19,250 ,
- Average of all' readings
-POST ENVIRONMENTAL TEST THRUST MEASURDfENT Date: 9/30/74 No. of Readings- Torque Switch-Setting ~ Thrust Output *
(pound s) , 6 "I 3/4" -16,397 - Note: The' low output thrust: readings _are.a result of poor.stam efficie cy-es a result _of accumulated deposits on the acms threads of the stem. The-thrust measured-during the last test point of the --environmental test was ~>1,350 pound.a. T
) -i y..
e 34.
C'405SA
- SUM.O.RY OF LOAD CYCLING DATA (continued) . i; U .
POST ENVIRONMENTAL ICAD CYCLING - Dates */30/*<4 - 10/4/74 Definitions. One (1) cycle Open111mit to close torque switch actuation to open limit. Two (?) strokes per cycle. Stroke Time: 40 see Duty Cycle: 'RUN' 7 cylens - 'OFF' 10 minutes Load (Thrust) : 16,39? at start 19,667 at finish Total No.- of Cycles: 794-Note: The increase in thrust . output is due to improved stem efficiency. The. repeated cycling removed-the corrosion in-the threaded area of the stem.
. The load cycling.was discontinued during the night.and ran during the first shift.
POSTLOADCYCLINGTHRUSTMEhSUREMENT
- Date: 10/4/74 No. of Readings- Torque Switch Setting Thrust Cutput (pou.pd s) 3 "1 3/4" 19,667 -
Note: The output thrust' returned to the value recorded , after the pre-test mechanical aging.
- s.
E, 35.
PWR Qualification 070593 t. APPENDIX C_ Radiation Exposure - Isomedix Certi*icate of Performance 9 4 36.
1 f O
- 1.
- isomamix o;0596 July 19, 1974 Mr. W. J. Denkowksi Chief Engineer Limitorque Corporation 181 South Gulph Road King of Prussia, Pa. 19406
Dear Mr. Denkowski:
This will summarise the perimeters pertinent to the irradiation of one valve operator and meter assembly. Identification on the valve operator and motor ascembly was: SMS O Valve Control s/n 189835 . Reliahce 40 lb-ft motor I'.D. '2Y26707 4A1EI Units were placed in a co-60 field of 1x106 rad per hour, at an air equivalent dose. They were rotated several times during the exposure to achieve a more uniform ' dose distribution. Total dose received to the centerline of the unit was 204 = rad ' (air equivalent) with an overdose factor on the edges of the uni ts o f 1. 2. Irradiation was in' air and ambient temperature in a slight negative pressure. The temperature of the samples during irradiation did not exceed 1000F. Dosimetry was performed using a Victoreen Model 555 Integrating Dose Race Meter and Probe. The unit was calibrated on January 15, 1974 by the Victoreen Instrt.unent Company, using Cobalt-60 and Cesium-137 sources whose calibrations are trace-able to the U.S. National Bureau of Standards. A copy of the calibration certificate is available. Confirming dosimetry utilizing a Red Perspex system was also completed. , Isorriedix Inc. 25 Eastmans Road, Particoany, New Jersey (201) 887-4700 Meinne Aooresst post Office som 177, Persioneny, New Jersev 07o54 CHICAGO DIVISION
- 7s2a Neve ave.. Morton cro . nono. soots (312) 9461140
+
c ..- l
-4 Mr. W. J. Denkowski 2-- July 19, 1974 070397 Irradiation was completed July 18, 1974 and the units returned to you under separate cover.
Very truly yours,
@ M Ge Dietz Manager, Radiation Services
- GRD:k:n ~ -
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4 4 PWR Qualification L 0905c39 APPENDIX D Seismic Qualification - Lockheed Test Report O O 4 . e e t G e e I 39.
p_ , t C70599 4 5 TEST REPORT NO. 3521-4811 REPORT OF TEST ON LIMITOROUE CORPORATION SMSO OPERATOR W/ MOTOR -(40 FT. LB.) AND-MOTOR (25 FT. LB.) REPORT WRITER: - - - 3
- . M !I15Iltia f($P^#
TEST ENGINEER: I [ W. A. Black LO CK H E E D ELECTRONICS COMP ANY, INC. PLAIN FIELD, N EW JE RS EY DATE: Jure 17. 1974-APPROVED - BY: + N. Johny'p, Manager Environ % ntal Laceratory E --- me
, n o ---
9 Ape i e
. O'10590 TEST R DORT NO, _3521-(011 PURPOSE OF TEST:
To subject the test specimens to the Seismic Test referanced in Limitorque
.( Corporation Purchase Order Number 600456 dated June 11, 1974
- Limitorque Corporation MANUFACTURER:
iS114 Woodall Road Lynchburg, Virginia 24502 SPECD1 ENS TESTED: (e) SMSO Operator with 40 ft. Ib. motor.S/N 1SS_S_J5 j (b) Relianc,e 25 ft. Ib . . motor-10 NO. 2Y287074 AlEZ - APPLICABLE DOCUMENTS : LimitiirqCe7 erporation Purchase Order Number 600456 dated June 11, 1974 PROJECT NUMBER: 24-8041,3811 w OUANTITY OF , SPECIMENS TESTEO: One (1) each SECURITY OLASSIFICATION OF SPECIMENS TESTEOs UncimesiFied DATE TEST COMPLETEO June 12, 1974 LOCKHEEC ELECTRONICS CCMPANY, INC. TEST CONCUCTED BY: ENVIRCNMENTAL LABORATORY DISPOSITION OF Returned to Limitorque Corporation per SPECIMENS TESTEO: Lockheed Electronics Company, Incorpc- ' rated Peckirg Slip Number 97448 datec June 12, 1974. ASSTRACT: The test specimens were subjected to the Seismic Test referenced in Limitorque Corporation Purchase Order Number 600456 dated June 11, 1974 This test was completed with no visible evidence of external comage or resonances. Reaction-Type Vibration Machine, LAS TEST APPARATUS: Company Model RVH-72-SCOO, S/N S1401. 2 3 a e PAGE i OF 7
090591 I TEST REPORT NO. 3521-4911 Vibration Pickups, MB Company Type 124, j< TEST-APPARATUS - S/N 14074 and Type 126, S/N 14006.-
-(Continued)
Vibration Meter, MB Company Model M-G,
- S/N 539.
Dis 1-A-Gain Amplifiers, Unholtz-Dickie Model 610M, E. L. Number 463 and Model 510RM-3G, E. L. Number 464. Accelerometers, Endevec Model 22210, S/N NAS4 and FC55. , TEST PROCEOURE: The test tipecimens were secured to the test machine, as shown in Figures 1 through 3, and subjected to the follow-ing Seismic. Test in accordance with Limitorque Corporation Purchase Orcer Number 600456 dated June 11, 1974.
- 1. To determine resonant frequenciesg an exploratory scan was perf ormed in each of the three (3) major axes over the frequency range of 5 to 35 Hz with a maximum input -
' acceler.ation of 1.0 g's.
Ea. With no resenent frecuencies present, the test specimen was sue- , jacted to 10 second dwells at the frequencies specifice by the Limitorque Corporation representa-tive (see data sheets) in eacn axis. The vibration amplituce was maintained at the. maximum controll-able displacement from 5 hz to the frequency at which 3 g's was attained. The input was then maintained at 3 g's from that fee-quency up to 34 Hz. 2b. The test specimen was vibrated at 35 Hz at an input level of 6 + % g's for a ten (10) second dweII. The test specimens were actuated during part 2, and all performance monitoring wee performed by and the data retained by Philadelphia Gear Corporation per-sonnel. A U a X
'8 6 PAGE 2 OF 7
.7 ,
090592-
% TEST REPOR* NO. 3521-4811 The Vibtetion Test wee completed with
. TEST MESU',TS:
(, no visible evidence of external damage
% noted to either test specimen.
There were no resonances detected in. the three (3) exes of vincetion. RECOMMENDATIONS: None. Date merely submitted. Test Eneineers #IM. W. A. Black c
.E.
U a w e PAGE ' OF 7
090593 Test Report No. 3521-4811 (, a' FIGURE 1 I TEST SETUP - X AXIS X
~& l B
I i /) 3 'Y
- i
- . 5, a k.________..- n T- -
ma , FIGURE 2 , i' .; T8!:ST SETUP - Y AXIS .t~ - l- A. SMBO OPERATOR W/40' LES, MOTOR a i B. MOTOR 25' LES.
; . 4-A..
~ x.& ..s -
l~-- ; j- ,
~. -
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" , . }L.y,-- g 3 -
FIGURE 3 i ' '* M i , 7. A
. TEST SETUP - Z AXIS -
---igN .
3
~ ~.T
- a
.. l l ,
- + -'
I
-.-...-~.'..- -.1. ~ ._.. _ .._..
, . __m __ _ . _ - _ _ __ -. _ _ _ _ . _ _ _ - - . _ . . _ _ . - _
l VI3 RAT 20N TEST DATA SHIET
' AXIS: 2 Axis REPORT NO. 3521 - 4811 I
~ DATE: 6/12/74 I i r.uLORATORY (Fara.1) ii VARL PREQUINCT(Para.2.2)
Ez i INPUT CH. 1 ICE. 2 i f PUT l CH. 1 i CH. 2 NOTE: RECOROED DATA IS 4 11 i e DOUBLE ANFLITUDE 5 il .014 .016 .013 (INCHES) 6 16 .016 .016 .014 .054 .049 .047___ 7 il .017 .016 # .016 ENDURANCE TEST (PARA.2.23)i 8 11- .018 .017 l .017 .054 .C50 i
.052 Bs INPUT I DURATION 9 I .018 .017 i .018 35 .100 10 see.
10 t .018 .017 I .0125 .052 .052 .056 11 16 .018 .017 i .019 I I i 12 14 .018 .017 I .019 .051 .054 .058 13 ti .017 .018 i .020 TEST SPECIMIN , 14 i' .017 1 .018 1 .020 .050 .054 .059 NOMENCLATURE 15 14 .017 I .018 i .021 !
~~2t 1 017 i .018 i .021 .050 I . 0 5 '. l .060 1 ctr. With fixture 17 : .017 i .018 i .021 i 1 SM30 Operator 18 i, .017 I .019 I .022 .050 I .054 6 .062 - with motor
! 19 !! .017 I .019 I .022 I i 20 :' .017 I .019 I .022 .049 .055 i .063 21 64 .017 I .019 i .022 ' i SERIAL NO. 22 62 .017 i .019 .023 .048 l .056 I .064 23 il .017 6 .019 .023 e i 24 16 .017 1 .019 i .02; .048 i .056 I .067 25 li- .017 I .019 I .023 I I . 26 ii .017 'l .019 I .023 ' .048 i .056 t .066 MANUFACTURER 27 II .017 1 .019 I .023 i I 28 *t .017 i .019 i .024 __.048 i .056 *t .068 1 29 1. .017 1 .019 i .024 I i j LIMITORQUE CORPORA!!ON l 30 it .017 I 019 i .024 .048 i .056 ( .069 i 31 .i- .017 I .319 6 .024 i i i ACCELEROMETER LOCATION
. 32 =i .0165 I .0195 I .0245 .048 i .056 i .072 ,
33 t .0165 i .0195 t .0245 i i iCH. 1 l side of =cter i 34 i .0165 I .0195 .0245 .048 i .056 i .072 ' CE. 2 1 top of uni : 35 ;i .0165 I .0195 I .0245 I I 36 i i f l I REMARKS: 37 i l I I I 38 't i i t i i ' 39 il I I I i 40 +t i I i i i 41 !! I I i 1
- 42 18 I i i I
_,4 3 'l i i Operator actuated duringl 44 .i i this portion of test l 45 *i - 1 I 46 u . I ! 47 i l i l 48 1: i I I I 49 't i i ' i i 50 il I i I i l i i RES. nene H: l TEST ENGINEER FORM LEC 9223 I
'Jm. C. Black curr- 5 nr 7
0 % 893 V2BRATt0N TEST DATA SEEET AIIS E Aria REPORT NO. 3531 - 4811 DATE ~ 6/12/74 f U7" CRAIORY (Para, 1) l VUQ . FREQULiCY (Fa ra.2.27 ( N2
]i CH. 1 iCH. 2 _
_ i ,1NFUT- l CH. 1 CH. T' NOTE: RECORDED DATA IS DOUBl.E AMPLITUDE 3 g , q acu, -*-
,j .014 1.015 _ 1918, (INCH 13)
! a 014 t
.044 _ .035
_6 1 .017 .015 I .016 1 I DiDURANCE TEST (PARA.2.2 3)i
~
8 I 9 5 17 !
~
8 D 10" 9 .019 .017 .018 1 1
~ ~ ~
" : Jff l 10 see 10 I
.019 .01'7 .019 I .052 I .050 .050
.0js5 I .020 l 1.'. i
.0iS i !
12 i .0105 I ,018 l .020 _1 i 13 I .)18
.018 ! .021 i i UST SPEC W Ommm la IL i
1
.018
.018
.018
.018 1.021 i.021 i
i .0$0 j i .052 I .058 r 16 .0175 .019 i.022 y ' i 1 motor with fixture 17 4
.0175 .019 i .022 ,
, i 1 SMBO Operator
.017 with meter
~
~
18 .019 i.022 * ' j_,1 - .017 I .019 i.b23 i ! 20 '
.017 1 .019 I .023 i .048 I
.054 I .062 21 i .017 i .019 i .023 ; i i SERIAL NO.
22 i .017 I .019. t .023 i l i i -
~
23 i .017 i .019 I .0235 t I ! ! 24l l
.017 I .019 I .024 i t i 25 i .017 i .019 i .024 i .047 i .055 i .066
~
26 Ii .0165 i .019 i .024 I . I . EDOD. 27 it .0165 I .019 8 .025 ! i i < 28 ii .0165 i .019 I .025 i i t LIMITORCUE CORPORATION ! 29 )i .0165 i .0195 i .025 i L i i 30 ti .016 i .0195 i .026 i .047 4 .056 I .012
.016 ACCELEROMETER LOCATION li i i 95 I .026 i i i i 32 ll .016 i 5 1 .027 '
I I i
^
I 33 it .016 a 5 I .027 i i i CH. I rear of =ct:r t s' 34 ii .016 i .t #5 I .029 i i . I CH. 2 frent et nancwnee. i 35 1 .016 i .0195 1 .028 I i .047 i .056 I .086 36 i I 1i i i EE 37 i , I t 38 i . i 4 i e i 39 i i i I 40 i ! ' t i Operator actuated durit.it 41 l' 1 i 1 L tlis portion ei test
. 42 i i 43 i i
- 74
~
i i i 45 j i i 46 L$ I I i 47 ~ ; *. i ' 48 . _r . l , 49 I i s 50 l RES. none Fz IIST ENGINEER FORM LEC 922B
. Wm. C. Black I
SHIET 6 of 7
______ _ . _ _ _ _ _ _ _ . _ . _ _ _ _ ~ . . _ _ _ . _ . _ . _ l VIBRATION TIST DATA SHIET AXIS; Y Axis
kIPORT NO. 3521 - 4811 l DATE 6/12/74 i j l IXFLORATORY (Para. 1) i VARI . TRIQUDiCY ( Fa ra . 2 . 2 )
{ Hz INPUT CH. 1 "H. 2 IN7UT' CH. 1 CH. 2 ~i NOTE: RICORDED DATA IS y, yg g 5 .009 .016 .010 (INCHES) 6 .009 .014 .012 .052 .049 .032 i 7 .013 .017 .013
~
l OfDURANG TIST (PARA.!' .2 3)! , 8 .014 .017 .014 Hz IN PU". f DURATION i 9 .015 .017 .015 1 - .100 1 10 see 10 .0155 .017 .016 .051 .049 .042 ! i I 11 .016 .017 G16 1i 12 .0165 .017 .017
.017 7TST SPECDfEN 13 .017 .03 7 !
NOMENCLATURE ! 14 .017 .018 .017
- i 15 .017 .018 .0175 .049 .052 .047 i 16 .017 .018 .018 '
17 i .017 .018 .018 1 motor with fixture
, 1 SM50 Operator 18 iI .017 .018 .018 1 . i ~^
with :otor 19 Ii .017 .018 .0185 l ? i 20 I I_ .017 .0185 .019 .048 .052 1~.050 1 21 i .017 .0185 .019 , SERIAL NO. Y
.017 .0185 .019 i 23 i .017 .019 .019 I I 24 1 .017 .019 .020 .048 .054 I .052 i 25 i .017 .019 .020 .
l .. I 26 8' .017 .019 .020 i (
'. i MANUTACTURIR 27 ii .017 .019 .0205 _,
i i i 28 II .017 .0195 .. 021 - 6 i ' LIMITORQUE CORPORATION l 29 i! .017 .0195 _ 021 1 I . 30 i l .017 I .020 I .022 ,048 i .056 i .058 ! 31 ' .i i .017 '
.020 .0225 i i : ACCELEROMETIA LOCATION 32 i! .017 .020 .023 ! _
i i 33 16 .017 .020_ .023 F i i CH . 1! Top of noter i 34 il .017 .020 '
.024 ! i ! CH. 2i Tep front of i 35 II .017 I .0205 l .024 .049 I 658 i .062 l Handwheel 36 II l 4 j i ! RD'. ARKS :
37 11 I i ! l 38 I i i i i 39 l 1 i i l 40 i operator actuated durinit i el l this portion of test l
, 42 I i ! '
43 I I I 3 45 , ! 46 ,, 1 6
'7 I 48 1 49 i i 1 50 i i t i RES. none Hz TEST ENGINEER TORM LEC 9223 i
W. C. Black y h
PWR Qualification 090597 l. 4 APPENDIA E Figure 1 Te=t chamber Figure '/ Steam Generator Figure 3 Control and instruttentation Panel . O e e a 4
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8 i PWR Qualificatier. , i APPDTDIX F Figure 4 - Schemat.c4 - Instrumentation Table I Summary of Instruments used for Data Acquisition t 4 e 4A
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v I r ; ! c _ ', 'I' d., 1 ll u..,....'....'.m....,. .. g 3 g 1 - il dI 3 Al !
.i -
fili! M!t l T 54 ? M IT l,,;l{'I'j , m ,l'I'i'i!.lli 4* - H ff j '--> i t q !!2!!!!!!(iiill iIy ,- v 1 d..l. l
..i.- ...l. ,l.jv! 7,,l5.
. ... . .m 4
[Hj 4 g li,llI l l l _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - - _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ ~
070603 TABLE I
SUMMARY
OF DATA ACQUISITION SYSTEM t, ; 5m t *. ow ~ EU - 2 c "2 x 2 E 6 g MEASURING SYSTEM F < u$ k o' k
=
gn 5 = 2 a 'e d 4 < b U
$$ $ a Indicator Signal X-Ducer Monitoring 6 jy Conditioner Point p N qw . u) w 6w < m Em A 1* Feb. Feb.
iultipoint 1 1B T. C . -J ( *> ) Chan.ber Amb. 1974 1975 Temperature 2 2A T. C .-J ( 2) Test Unit
$ Recorder '3 3B T . C.. -J ( ~> ) L.S. Comp 5 Type J T. C. 4 4B T. C . -J-( 2 ) Chamber Amb. . @ 5 --
Room Amb. E Honeywell 6 -- Shorted Model No. 7,.10 5 T. C.-J Condensate K153x80-c- 9,11 6 T. C.-J Press x-Dw? t II-W6-65 9,12 7 T. C.-J Press x-De3 E Serial No. 13,15,17 8 T. C.-J Press x-D#4
$ TilP06-83004 14,16,18 9 T. C.-J Press x Del E 19,21,23 10 T. C.-J H 2 O input
[ 20,22,24 11 T. C.-J Hp0 output G
, 2. Strip Chart 1 Amplif. Press x-D#1 Chamber Amb. Mar. Sept .
- w Recorder 2 "
Press x-D4 Test Unit 1974 1974 5 8 Channel 3 Press x-D43 L.S. Comp.
$ 4 "
Press x-D*4 Chamber Amb. E Gulton 5 T.C. Mod, lA . T. C . -J ( 2 ) Chamber Amb,
" TR888 6 "
2B T. C . -J ( ~> ) Test Unit d 7 " 3A T.C.-J(2) L.S. Comp. E S/N 8 " 4A T.C.-J(2) Chamber Amb. p 3042802 s l
$ Note: Amp. TSC 801 b T.C. Mod. TSC 500J b
e 9
. 4 I
1 TABLE I (continued) O Ob94
SUMMARY
OF DATA ACQUISITION SYSTD4 t w w e 5
- 5 5
*9' . e 1: i:
o" *
$$ .D I 8N h ! deasuring System 55 si S U ". < % $ E EU3M '
Signal X-Ducer Monitoring u o Indicator conditioner Point 3 Mercury bulb Thermometer Chamber New 2/75
, Ambient 2/74 g Wexler 50-400 F
& Bi-metal Dial Thermometer
, w 4 Chamber New Ambient 2/74 2/75 h Wexler 50-400 F
- Dia PEessure Gage Chamber 5/74 C
h m 5 Pressure 8 $ Ashcroft 30 in Hg to 200 psig 5 $ e .. - Time of Day - Ti"* Of 6 Day - - Wall-Clock o
" Total - -
7 Running Time Clock b Test 1000 hours 0.1 hr. Time Resolution 8 Strain Indicator 20,000 lb Test Unit 12/73 1p eo - Bridge Load Thrust /74 w .'. BL &H Cell output
$ Typen BL&H f.
s3 3 S/N 443604 U~1 Q S/N 2512 mo
4 9
, TABLE I (continucd)
SUMMARY
OF DATA ACQUISITION SYSTEM N t - E w
-2 h Measuring System e09 e 0
e 0 b si e m 2 . l
$th '
nom' D W C
~
, O e
Z e l l
$E% 1>
{< $ Indic- Signal X-Ducer Monitorint b$ j3 \ 6 % .e* goa m m ator Condit. Point beU e U vJ i 9* ' 3 phase Recording Power New l Watimeter Conswnption 7/74 2/75 ! g of W Enterline Angus operator , Model A 601C ' S/N 192358
- 10. Panel Meters Test Unit New .
Current 3/74 3/75
. 3 meters one in each -
w p phase - 5 5 - -
* $ Low Range 0-10
$ $ amp meters W Triplett Type 430 0
4 Ri Range 0-50 g amp meters 3 Triolett Tvee 430 h
- 11. Panel Meters Test Unit New w 3 meters one acros- Voltage 3/74 3/75 g each phase h
l 5 0-500VAC Triplett 3 Type 430 1 l w
TABLE I (continued) 07()pgg
SUMMARY
OF DATA ACQUISITION SYSTEM ce z b b a eo . g* o Z b$ b N$ b u$ g[ h $ Measuring System $$ $* zu Signal X-Ducer Monitoring O E O Indicator Conditioner Point h
~ ~
g 12 Stopwatch . Stroke Time g $uog p Rever S/N 512406 m 13 Flow Meters (2) Chemical New 2/74 2/75 Fisher & Porter '
. . flow
$ k .
3 E 'Model S/N - m 10A1735Y 7309A0574A1 j 10A1735 7407A0403Al , t 14 2 Dial Pressure Gages Manifold New ( (2) Wesler Model BA14P Pressure Pump 2/74 Indicator 2/75
,e g g g Pressure only g g 1 Acco Halicoid U s 0-200 psig-15 PH Meter Comparison
& PH of Chem. Acainst-Solution Standard Solution ie 2e .,wwvkwaces
.3 y Motor & New yj James G. Biddle Control 2/74 2/75 g,a Leads gg Model 21159 m
# 732521 S/N i
j l j ! SWR Qualification OIO8 f)7 APPENDIX C Reliance Electric Company . Certificate of Complicance for Head Aging f
R A N 2 5 r t t r~ r * ' * ** e e v pa m. O@()R
.fD'ICAT4 of COMPLINICE Limitorque Corp.
$114 Woodall licsd 1.yneheurg Ya. 24502 March 3, 1972 EQUIP M : Qty.1, R56 Frame, 25 ft.ibs. Start Torque RETGENCE Parchase Order No. 355696-A FILE 8 Order No. 601962-P We certify that the equipnent ident fled above has been desicnod, nanufacturM, incpceted, and/or tested in accordance with the requirenents estaoli:. sac ey the folleving specificationst Insulation 0.F.I. 226 59 Tinish C.F.I. 336.163 Class k Bodistion Insulation llent e6ed 100 hrs. at 160*C Rotor treatment U-475 varnish Winding Epoxy coated 4824-3A Drain in B.E. Shi*1d - U.L. opproved
. [ @ *c R.O. Lunsford, P.1nnger 4]p Control Ecpartment d son, Indiana Mdg 0g $tidD BGL: st i4 i , , , . .. i i i . .. .e n i . T . . . . . . . . , m. , ,, ,, ,, , . , , ,, ,. .w n , r ,y , , , , , , ; ., ,, , ,, , , , . , , .
a-e 2.+ :.a+ m. 2 m.-_ . , __a_m. a - J4ma e. ,_A,# .A .a#- - - 4 -J- ---. - _ . - - - .La 4- Jeab.e as ,;..m.w.esw_..a- L.; --a. 4sm4. .w w e 6 0 F 6 070699 9 I o e 4 4 8 9 4 e ) } i ) i l
+
t
. t l
BWR Qualification .
.. . .i t
0"OGto .
- s. .
t APPENDIX D , h Radiation International, Inc. 4
. Certification' ~of Test - Unit Irradiation 4
't I
t 9 1
.]
e .
*= ". - . . . . . . . . . . . - . , . . - _,._.-...,,-.-.:.,._--..-,
- ,. . . . - - . , , , - - - - , , - , - , . . . . - - - , - ~ , , , , , , - - ,
. _ . _ . . . _ _ . . _ _ _ _ . - . _ . . . = . _ _ .._..__-_._..._-___..m.___.____.__. ,
. 1
.a . .
- t, 1
Onnet1 May 23, 1972 l IMr. W. H. St.tigelmann, Manager Nuclear Sys cens Laboratory ; Franklin Institute Research ' Laboratory . 20th and- Race streets Philadelphia, Pennsylvania- 19103 Dear-Mr. steigelmanni-
- This will summarize the parameters pertinent to the Arradiation of Limitorque Corporation's materials, described belov, per your Case C-3327._ i i .:*
SN601962P ) Motor used in FIRL test Reliance Electric Company ) - report #F-C3441 Madnetic Brake and Motor,
. Motor: SNC31960P-Ralf_anceElectricCompany Brake: Magnetic Brake Reliance Electric-Company 1 Model X7-62008-26
~
S N-2 -53 276 . 6 Ft.-Lb. Torque, 460V' valve and Motor Limitorque SMB Size O SN 144068-
' Motor.- Limitorque Reliance Electric Company SN601961-P x
v.<-y-. -,se, ,e y w ,,,, g-,. <e+i.,,,,,,,.,,.v-- , , 79=w ye-, f
I . . O s a 0906t2 4 Mr. W. H. Steigelmann May 23, .* 12 - l Components were placed in a cobalt-60 gamma field of 1.0 Mrad f per hour intensity for a period of 4 hours, and received a total # minimum dose of 4 Mrad. They were then shipped via RII truck to Ogden Technical Labs for testing. The units were then received for further irradiation, and were exposed in a 1.0 Mrad field for an additional 200 hours or 200 Mrad. Portions of the outer sides of the components received a 254 higher dose to assure that inner portions received the designated 204 Mrad. Components were turned several times during exposure to assure a more uniform dose distribution. Irradiation was in air at amL' int temperature and pressure. Radiant heat from the source heated the samples somewhat, but did not exceed 1000F, as con-firmed by measurement of an oil solution in the same relative position. Dosimetry was performed using a Victoreen Model 555 Integrating Dose Rate Meter and Probe. The unit was calibrated on January 15, 1971 by the Victoreen Instrument Company, using cobalt-60 and cesium-137 sources whosa calibrations are traceable to the U.S. National Bureau of Standards. A copy of the calibration certificate is availablo. , - Following irradiation, components were visually inspected'for obvious irradiation damage. None was noted. Irradiation was completed on May 8, 1972 and picked up by FIRL ' May 9, 1972. Very truly yours ,
~
(~
~
Ge Di t Manager, Radiation Cervices GRD:km 0 9
g f)~f i e i.
- q go-3ff D/~OIVO APCo Exhibit 08 i b' '
,,f , *s JN D UNITED STATES
.8 - Uh 'I l' 2 = ,,# g ,I :
fiUCLEAR REGULATORY COMMIS$10N MGION ll
%, ,/,
!- 101 MAmtst ta sT., N w., surra 310o Art. Aura.csoncia nesas *92. W j3 p g
- JAN 3 7 III) , ,
In Reply Refer To: wd,r;,g nev le $l evir.I, ; kil:JPO 50-348
@ q }l Arsl b . ( LIf p!'NW MWt ..g f i
f OCif / f 00 3 % V' ? O l lgp h v/a/n _ t Alabama Power Company U.F07 , C)- Attn: T. L. Clayton, Jr. ' Senior Vice Presidest }gC . Post Office Box 2641 21:1:ingham, Alabama 35291 ,
~Gentiament Please disregard TE Eulletin No. 79-015 and anelosures dated January 14, . -
1950. l Enclosed is II Lulletin No. 79-01B which requires action by you vith regard . l to your power reactor faci.11ty(tes) with an operating license. ' Should you have questions regarding this Bulletin or the actions required of you, please contact this office. [ i Sincerely, (N N . James P. 0'Reilly l Director-i
Enclosures:
- 1. II Eulletin No. 013 -
with- Et closures . 2. List of gecently Issued II Bulletins s i b Q \ l 1
..u,~.-...-_- ..,_.___u..,... . _ - - . . _ _ . _ . _ - _ _ . - . _ . . _ . _ . ~ . _ _ . . . . l
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- T--is ! 6 t
- e. ~ e . e Ada<nv/_ Ruok. , .
pag.- y sm ,mm so3a.v ,.SAA a Yt ? '. _d _ _ Of__,_______ __ u un u, _ . . . . d!/9,g h l.l l ' C'.m _ , _ _ _ . _ , _ d'MI I 0?I 'l .. ., _ -. ._ __-
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5
,6 e, s ns + t i JAN 171333
-Alabama Power Company -
2-ce v/ enc 1: 0051547 A. R. Barton Executive Vice President Post office Box 2641 ^ Birmingham, Alabama 35291 J. T. Young Vice President-Production Post office Box 2641 birmingham, Alabama 35291 C. Biddinger, Jr. . Manager-Corporate Quality Assurance Post Office Box 2641 . Birmingham, Alabama 35291 E. O. Thrash-Hanager-Nuclear Generation Post Office Box 2641 Birmingham, Alabama 35291 H. G. Hairston, III-Plant Manager
. Drawer 470 Ashford, Alabama 36312 R. E. Hollands, Jr.
QA Supervisor Post Office Box U Ashford, Alabama 36312
~~
{ e l-l l 1 . q *'%'W w I'r-v+'-FTPW rf- rr---1FD F 'rWe y- w ge j=yp br=& y WTF wpm y-g 9'MTMPWVPWs
o v 8 ' ., f f 0051548 UNITID STATES SSINS No.1 68"O NUCLEAR REGULATORY CO.tISSION Accessions Mo.t OTTICE OT INSpICTION AND EhTORCL".E.57 491025057u WASHINGTON, D.C. 20555 January 14, 1980 II Bulletin No. 79-01B ENVIRONMESTAL QUALITICATION OF CEASS IE EQUIPMEh7 Description of Circumstances: IE Bulletin 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 conditions. vill function under (i.e. during and following) postulated accident The NRC staff has completed the initial review of licensees' responses to Eulletin No. 79-01. 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 equipment in the operating facilitiss. In addition to requesting more detailed information, the scope of this Bulletin is expanded to resolve safety concerns relating to design basis environments and current qualification criteria not addressed in the facilities' TSARS. These include high aging, and energy line breaks (EELB) 1. side and outside primary containment, submergence. Enclosure 4, "GUIDELINIS FOR EVALUATING ENVIRONMLh7AL QUA11TICATION OT CLASS IE ELECTRICAL EQUIPMEh7 IN OPERATING REACTORS", provides the guidelines and criteria the staff villinuse evaluation in evaluating response to thisthe adequacy of the licensee's Class IE equipment Bidletin. In general, the reporting problems 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 contain-ment, was not included in the responses. 2. In many cases, the specific information requested by the Bulletin for each component of Class IE equipment was net 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 resulting from both LOCA and/or HELB inside primary containment and HELB outside the primary containment.
+ , ) 0051549 IE Bulletin Fo. 79-013 Janua ry 14, 1980 Page 2 of 3 j
i
- b. Identification of the Class II electrical equipment items within each of the systems identified in Item a, that are required to function under the postulated accident conditions, c.
The correlation between the environmental data requirements specified in the TSAR and the environmental qualification test data for each Class IE electrical equipment item identified in Item b above. 4. Additional data not previously addressed in TE Bulletin No. 79-01 are needed I to determine the adequacy of the environmental qualification of Class IE
- electrical a submerged equipment. condition.These data address component aging and operability in 1 l
Action To Be Taken By Licensees Of All Power Reactor Tacilities With An Operating license (Except those 11 SEP Plants Listed on Enclosure 1) ,
- 1. Provide a " master list" of all Engineered Safety Testure Systems (Plant Pro-tection Systems) required to function under postulated accident conditions.
Accident conditions are defined as the LOCA/EEEB inside containment, and HELB outside containment. For each system within (including cables, EPA's terminal blocks, etc.) the master list identify each Class IE electrical equipment item that is required to function under accident conditions. Pages 1 and 2 of Enclosure 2 are standard ' formats to be used for the "msster list" with typical information included. Electrical equipment items, which are components of systems listed in Appendix A of Enclosure 4, which are assumed to operate in the ySAR safety analysis and are relied on 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 orginally licensed to operate. The necessity for further up grading of nonsafety-related plant systems will be dependent on the outcome of the licensees and the NRC reviews subsequent to TMI/2. ' 2. for each class IE electrical equipment item identified in Item 1, prov;h written evidence of its environmental qualification to support the capability of the item to function. under postulated accident conditions. For those class II electrical equipment items not having adequate qualific; tion data available, identify your plant for determining qualifications of these items and your schedule for completing this action. Provide this in the format of Enclosure 3. l 3. For equipment identifed in Items 1 and 2 provide service condition profiles L (i.e., temperature, pressure, etc., as a function of time). These data l should be provided for design basis accident conditions and qualification t tests perforned. This data may be provided in profile or tabular form.
, ,, ~ , ,,
l e ; 0051550 i
't . 01B Janua ry 14, 1980 II f . Page 3 of 3 I
p e qualifiedt N f e. . % .E electrical equipment against 4.1 plines provideo in ; .losure 4. Cnclosure 5, " Interim Staff Position i
' paental Qualificatice of Safet, Related Electrical Equipment,"
' pupplemental information to be esed with these guidelines. For the f t )FIquipment Qualifica: ton Plan." Incluoe in this plan specificidenti I
phich will be taken to deterni,- equipment qualification and the pior completing the actions. S- the maximum expected flood level inside the primary containment from postulated accidents. Specify this flood level by elevation g 620 foot elevation. Provide this information in the format of p 3.
~
- 6. "I.icensee Event Report" (II.R) for say Class II electrical equipn'ent has been deterinimed as not being capable of meeting environmental on requirements for service inteeded. Send the 1.ER to the approp-Regional Office within 24 hours of identification. If pint opera-
. gas centinue following identification, provide justification for such in the I.ER. Provide a detailed written report withir.14 deys of p tion to the appropriate NRC Regional Offit. . N se items which geciously reported to the NRC as not being qua 111ed per IEE-79-01 do g pa an I.ER.
7* pth, actions specified by this bulletin in accordance with the g schedule:
'f g Wt a written report required by Itens 3, 2, and 3 within 45 days pp receipt of this Bulletin.
j 3. bait a written report required by Itens 4 and 5 within 90 days from gettipt of this Bulletin, 18 g en is requested under the provisions of 10 CFR 50.54(f).
- CC hyou 7.bare requested above, writtento provide within statements of thethe timeinfor1 above periods specified cation, signed in j ge affirr.ation, Sulast p worts to the Director of the appropriate NRC Regional Office.
of your report to the U.S. Nuclear Regulatory Connission. Office
'I I'8 M ,3.C. 20555. Send a toPf a and Enforcement, Division of Reactor Oper Washim;tesi Approval was P[ g S0 B180225 (RC07?.); clearance expires 7/31/80.g a blanket cle clos planta 2 Nsst 3' ,,, onent Evaluation Work Sheet Instructions 4* W
,11aps for Evaluating Environmental Qualification g glass IE Electrical Equipment in Operating Reactors i
3* ygferis 5taff Position on Invironmental Qualification of g,g,gy-Related Electrical Equipment (N1' REG-0588) Notu Dr A 'I ave enclosures are t be sent to the corporate offices only.
005155t s IE Bulletin No. 79-Olb Enclosure Janua ry 24, 1980 g RECENTIT ISSI.*ED IE BULLITINS Bulletin Subject Date Issued Issued To No. 80-01 Operability of Ads Valve 1/11/80 All BWR power reactor Pneumatic Supply facilities with an operating license 79-01B Environmental Qualifica- 1/14/80 All power reactor tion of Class 1E Equipment facilities with an operating license _ 79 Possible Malfunction of 12/7/79 All power reactor . Namco Model EA 180 Limit facilities with an Switches at Elevated OL or a CP Temperatures , 79-27 Loss of Non-Class-1-E 11/30/79 All power reactor Instrumentation and facilities holding Control Power System Bus OLs and to those ' During Operation nearing licensing 79-26 Boron Loss From BWR 11/2?/79 All BWR power- reactor control Blades facilities with an OL 79-25 Tailures of Vestinghouse 11/2/79 All power reactor BTD Relays-In Safety-Related facilities with an Systems OL or CP 79-17 Pipe Cracks In Stagnant 10/29/79 All PVR's with an 4 (Rev. 1) Berated Water System At OL and for information PWR Plants to other power reactors
, _ . 79-24 Trozen Lines 9/27/79 All power reactor facilities which have -
either OLs or cps an
- are in the late stage of construction 79-23 Potential Tailure of 9/12/79 All Power Reactor Emergency Diesel Tacilities with an Generator Tield Operating License or Exciter Transformer- a construction permit 79-14 Seismic Analyses Tor 9/7/79 All Power Reactor
.(Supplement 2) As-Built Safety-Reisted Taeilities with an Piping Systems OL or a CP 79-22 Possible Leakage of Tubes 9/5/79 To Each Licensee of Tritium Gas in Time- who Receives Tubes pieces for Luminosity of Tritium Gas Used in Timepieces for Luminosity
a 0051552
, Enclosure I SEP Plants Plant Region
- Dresden 1 ~~
311 Yankee Rowe I Big Rock Poi.nt III San Onofre 3 - y Haddam Neck I lacrosse III Oyster Creek I R. I. Ginna I Dresden 2 III , Millstone I Palisades III b u M f a
_ . . . . _ .- -- __ . . = - . .. -. .. _ _-_ . . t r e s e s s u . . u t-- y,~, ,;;. ..- : . . -- . - n.,..- :- 4 . Docket No.: 50-XXX
- r .ig .-- .
t.nclosure 2 ~
... . . . ,.T X...*J.;.i ~f.....MASTER LIST. f:
H."
!-?
. a ..a. .,-r:z .. t..3:-
.. >..=
-.:,is:f.s..f.w m(T.Ypica1). , .-. , -
. ' ; , q.-
,.- , y _'
.g.,.-
~ :;.. .,;. ":.;QV."X=rs-~Q a+..==L. . 9, . L's . . . . .
Y ' . ..
- ...$.'Af.-jQ 'FD.'f ,(CTitss i5'E1Ntfical Equip ent Ngu'tred to Function... Q
.. r , < . ' . . : .- .p.;T.
( *
- '.. . 6r ' '. . . Under Postulated L '-Accident Conditions)
~'
~; .- '**~*--: - - . . '
.1. '
SYSTEM: RESIDUAL HEAT REMOVAL' (P.HR) y'- l -
. ." :*:.5 . . ... -~.:~ - x . . . . .. * .
. . . . .: 4- ..
' . .: . . ,,, . . . COMPONENTS * ' .
.. .s * -
, . - ~
, , . e
. .. LOC a t.10 n Plant Identification .
Number
- Inside Primary Outside Primary Generic Name Centainment Containment 1.PT 456 .
, PRESSURE TRANSMITTER x ,
1LT 594 LEVEL TRARSHITTER x 1LS 210 LIMIT SWITCH x II. SYSTEM: AUTOMATIC DEPRESSURIZATION SYSTEM (ADS) . COMPONENTS Locauon
~ .
Plant Identification Inside Primary Nu:.ber Generic Name Outside Primary Containment Containment B21-R001 VALVE MOTOR OPERATOR x B21-F003 SOLENOID VALVE g B21-F010 PRESSURE SWITCH x i
- p
- c. - r ,- .
un.iv>ue , t . III.' SYSTEM: RMR
,JIPMENT/ COMPONENTS (Typical) 005l554
~ ~ '
** COMPONENTS'.
i Location
- Plant Identification Inside Primary -
Nt; .ber* Outside Primary Generic Name Contaiment Containment 16xP455 0-RING GASKET x
- EPA, Class E. .
~ ~
Wertinghouse. 100C
, ELECTRICAL PEhETRATION ASSEy. SLY X KULXA No. ET35 TERMINAL POARD x ONKONITE. 1000V, 3C ,
Black
- _ POWER CAS,LE x x -
X ERAND 10'4 40 LUBRICATE OIL x 15 KS69 (Boston Wire & Cable) INSTRUMENTATION CABLk , x x
' Cutler Hanmer TB
,; .g TERMINAL BOX x
RAYCHEM XYZ CABLE SPLICE x x Scotch No. 54 . INSULATING TAFE x T&E No.10 IliSULATD TEP.MINAL LU3 x Y Brand Epoxy No. . SEALANT x x 111
'inen a component is not identified by plant identification number, use the
** manufacturer, model number, serial ntr.ber, etc.
Like cc. ponents may be referenced. 4
. a
- 9 9
4 6
- c_ , ,
i a
; .=, .
- w ,
0051555 Enclosure 3 SYSTEM COMPOND7 EVALUATION WORK SEEIT IhSTRUCTIO,fS, '
- 1. 7.quipment
Description:
Provide the specific information requested for each Class II electrical component. Provide component location, specific information such as the building, access floor elevations, and whether '
' the component is above the flood level elevstion. In addition, provide the specified and Jemonstrated accuracies of all instruments for their trip functions and/o. t est accident monitoring requirements. Cables, EPA's, ternisal bloch , :M #ther iter chall be idsnified as part of ; the engineered safety f 2 cures systems.
- 2. Environment: List values for each environmental parnaritz 1. ated.
List the " specification values" obtained from postulater it analysis in the " SPEC" column. List the " qualification values" ootaineo trem test reports, engineering analysis data, etc. in the " Qual" column. Terpera-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 to function and identify the document which provides the basis fer this time interval. 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 submergecce or a service condition such as aging was not previously considered, identify it as an " Outstanding Item." 3. Documentation
Reference:
Reterence the docanents from which information was obtained in the " Spec" column. Identify the document, paragraph, etc., that contains the postulated accident environmental specification data. In the -" Qual" column identify the document, paragraph, etc., that contains the environacgtal qualification data.
- 4. Qualification Method: Identify the we hod of qualification. To describe the qualification method use words such as simultaneous test, comparison test, sequential' test, and/or engineering / mathematical analysis. Words such as " test" and/or " analysis" when used alone do not adequately identify the qualification method.
- 5. Outstandint Items: Identify parameters for which no qualification data is presently available. Also, identify parameters, service conditions, or environments not previously addressed during 7SAR environmental qualification analysis such as submergence, qualified life (aging), or ELB. Identify in the "liotes" section on page 1 of this enclosure the actions planned for determining gaalification and the schedule for
. completing these actions.
actifty:' ' SYST[HC0llPONENTEVALUi WORK SiiEET C " itjet. .: . (Typicai, ENV NMEN CUMENTAUDN REP EQUIPMENT DESCRIPTION QUALIFICAT!0N OUTST'ANDI!G Parameter SPecift- Qualifi- Speceit- lIlualian- ; HETil0D ITEMS . catinn ra tinn ra tion cation
;ystem: RilR Operating 15 min. 300 mic. 1 E Simultaneou:
None
'lant ID No. IPT456 Time Test ,
- omponent: - '
I
'RESSURE TRANSMITTER.
"'*** 1 5 Simul tcneou-SEE ACCIDENT AtiD
#@F)
( TEST PROFil_ES , Test, nm tanufacture: PROVIDED _ lscher-Porter Co.
- Pressure lodel Number:
(PSIA) i1 5 Simultaneou- None
;0-EN-1071-DCXN-NS Test .
Relative & 1 inctt:n:. thantolty(%) 100%' 100% ! 5 Simultaneou:
.- None
- cid;nt Mont toring l ..I Test Chemical N3003 / ' '
curacy: Spec: 5% Spray 1 ilA0ll See Note 1 Demon: 4% , rvic2: RllR Pump 1A 6 8 Sequential scharge Pre 3sure Radiation 4x10 rads 1.2x10 rad , 2 6 Test None S/Hl07 -
~
, I. Sequentf.' l -
cati:n: Containment Agfr:g 40 yrs
- 40 yrs 3 7, 8 "
( 2. E cn ood Level Eley: 620' ? Not Not ~ Hone cn ova Flood Level: Yes Submergence Required Required See Fote 2 . cn No x' l C Jcumentation
References:
Notes: ' . FSAR Chapter 3. Paragraph 3.11 1. XYZ Letter No. 237-1, dated November 2,1979, . FSAR Chapter 14. Paragraph 14.2.3.1 nas been sent to MFG. requesting the quaitfication- . T:chnical Specification 3.4.1. Paragraph A information. If qualification rot detennined . Technical Specification 4.6.5. Paragraph D acceptable by December 15, 1979, corgontnt FIRL Test Report No. 3 will be replaced during refueling cutage e'trrrh 1960. Fischer and Porter 00.'500 Testdated November Report No. 2500-1 2,1972 *
.A. B. 000 Engineering Evaluation Data Report No. 6932 2. In the FSAR submergence was not' considered Wylla Laboratory Report.No. 467 an environmental parameter. ABC Laboratory '
h to perfor.n submergence ' test in Aprl) 1980.
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Enclosure 4 l* .
- curett1NES Fon EvrtuaTiss EnvinosMEsTAt ovat!FICATier 0051557 0F CLASS IE ELECTRICAL Etv)fPMENT IW CPEUTING REACTORS 1.0 Introduction 2.0 Discussien 3.0 Identifiestien of Class TE Eeuipment 4.0 Service Cenditions
~
1 Service Conditisns Inside Containment fer a Less of - Coolant Aceteer.* (LO;A)
- 1. Te?.eerature and Pressure Steam Conditiens ,
- 2. Radiation
- 3. Submereence
- 4. Chemical Sertys 4.2 Service tenditiers fer s P'~R Main Steam Line Break (MSLB)
. jns1:e Centainmen:
- 1. T
_emeerature and Pressure Steam Cenditiens
- 2. Radiation
- 3. Subme*eenee 4 Chemical Scravs 4.3 Service Conditiens Outside Centeineent
- 4. 3.1 Areas Subi eet to a Severe Environment as a Result of a H1cn Enercy L'he Breer (MELB) 4.3.2 Areas Where Fluids are necirculated Fre, Inside
~ Containment to AccomeTisn tene-Terr Eme-cenev ~
fore Coo)1no Fellowine a LOCA
- 1. Temeeratore. Pressure and Relative Humidity
- 2. Radiation
- 3. Submereence
- 4. Cha-feel Sorays 4 * ,
. 0051558- -
4.3.3 Areas Nvtr 11v Mat,.tained at Reem Cenditiens 5.0 Oualificatien Met &d2 5.1 Selection of Oualification Methed - 5.2 Oualifiestion by Tvoe Testina
- 1. Simulated Service Certditions and Test Duratien
. 2. Test Soecimen
- 3. Test Seeuence
- 4. Test Seecinan Aoine
- 5. Functional Testine and Failure Criteria
- 6. Insta11atien InteMaces 5.3 Oualification bv a Cembination of Metheds (Test, Evaluatien.
A.alysisJ 6.0 ,Ma rcin 7.0 Acine E.0 Documentation Appendix A - Typical E:;uipment/ Functions Needed fer Mitigati:n of a LOCA or MSLS Accident
' Appendix S - Guidelines for Evaluating Radiation Service Cenditions Inside Containment for a LOCA and MSLB Accident Appendix C - Thermal and Radiation Aging Degradation of Selected Materials l
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- 0051559 GUIDELINES FDR EVALUATING ENVIRONMENTAL QUALIFICATION
_OT CLASS IE ELECTRICAL ECUIPMENT ' IN OPERATING REACTORS 1.0: INTROOUCTION On February 8,1979, the NRC Office of Inspection and Enforcement issued " 1E Bulletin 79-01, entitled, " Environmental Qualification of Class IE Equipment.' This bulleti- ',*;;uested that licensees for operating power - reactors complete within 120 days their reviews of equipment qualification s begun earlier in connection with IE C, cular 78-08. The objective of ' C. IE' Circular 78-03 was to initiate a review by the licensees to detemine - whether proper docc.en:ation existed to verify that all Class IE electrical - equipment would function as_ required in the hostile environment which could t
, result fro . design _ basis events.
The licensses' reviews _are now essentially complete and the NRC staff has begun to evaluate the results. This docu ent sets forth guidelines for the NRC staff to use_in its evaluations of the licensees' responses to IE
. Bulletin 73-01 and selected .1ssociated qualification documentation. The-objective of the. evaluations using these_ guidelines is to identify Class IE equipment whose documentation does not provide reasonable assurance of environ-i mental qualification. All-such ' equipment identified will then be subjected to a plar.t application specific evaluation to detemine whether it should be requalified or replaced with a-component whose qualification has been adequately
. yerified.
These guidelines are intended to be used by the NRC staff to evaluate the - l qualification methods used for existing equipment in a particular class of plants, i.e., currently operating reactors including SEP plants. e e = ,r . . - , ,w-o y ,--y-- ----- - -- -e -
x.- 0051580 Equiprvmt in other classes of plants not yet licensed to operate, or replacement equipment for operating reactors, may be subject to different -. . requirments such as those set forth in NUREG-05SS, Interim Staff Position on Environmental Qaalification of Safety-Related Electrical Ecuipment. In addition to its reviews in connection with II Bulletin 79-01 the staff is engaged in other generico reviews that include aspects of the equipment qualification issue. THI-2 lessons learned and the effects of failures of , non-Class IE control and indication equipment are examples of these generic r evi ews. In some cases these guidelines may be applicable, however, this detemination will be made as part of that related generic review. 2.0 DISCUSSION ' IEEE Std. 323-1974 I is the current industry standard for environmental qualification of safety-related electrical equipment. This standard was first issued as a trail use standard, IEEE Std. 323-1971, in 1971 and later . after substantial revision, the current version was issued in 1974. Both versions of the standard set forth generic requirements for equipment uali-fication but the 1974 standard includes specific requirements for aging, _ margins, and naintaining documentation records that were not included in
-the 1971 trial use standard.
The intent of this d$cument is not to provide guidelines for implementing either version of IEEE Std. 323 for operatin's reactors. In fact most of ;- the operating reactors are not comitted to comply with any particular industry standard for electrical equipment qualification. However, all of , the operating reactors are-required to comply with the General Design Criteria I IEEE Std.- 323-1974, "IEEE Standard for Quali.fying Class IE Ecuipment for Nuclear Power Generating Stations."
, 005156i
~
specified in Appendix A of 10 CFR 50. General Design Criterion 4 states in part that "structuras, systems and components important to safet; shall be designed to acc==odate the affects of and to be co=patib'e with the environmental conditions asso:iated with normal operation, maintenance, testing and postulated a::idents, including less-of-coolant accidents." The intent cf these guidelines is to provide a basis fnr judgements required f _ to confirm that operating renctors are in ce=pliance with General Cesign Critarion 4 3.0 IDENTL ICATION 02 CLtS$ IE E0'JIPMENT Class II equipment includes all electrical equipment needed to achieve er.ergen:y reactor shutdown, containment isolation, reactor core cooling, containment and rea: tor heat removal, and prevention of significant release
. of radioactive material to the environment, Typical systems included in pressuri:ed and boiling water reactor designs to perform these fun:tions for the cost severe postulated less of coolant ac:ident (LOCA) and main t .
steinli.ne break accident (MSLE) are listed in Appendix A. F. ore detailed descri;tiens of the Class II equipment installed at specific plants can be obtained from FSARs, Technical specifications, and emergency procedures. Although variation in nemenclature may exist at the various plants, environmental qualification of those systems which perform the functions
~
! identified in Appendix A should be evaluated against the appropriate service conditions (Section 4.0). l l The guidelines in this document are applicable to all components necesri y i for operation of the systers listed in Appendix A including but not limited to valves, m= tors, cables, connectors, relays, switches, transmitters and valve position indicators, ' l
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- 0051562 4.3 SEFNICE CocITIONS In order to determine the adequacy of the qualification of equipment it is necessary to specify the envirerpent the equipment is exposed te during nomal and accident conditions with a requirement to remain functional, These environments are re'ferred to as the ' service ccnditions.' .
The approved service conditions specified fn the FSAR or other licensee submittals are acceptable, unless othendse noted in the guidelines discussued below. - 4.1 - Service Conditions Inside Centai7ent for a 'Less ef Cecient Accident (LOCA) 14 Teneerature_ and Pressure Steam Cenditters In general, the containnient terperature and pressure conditions as a function of time should be based on the analyses in the FSAR, In the specific case of pressure.
, suppression- type containments, the followir.g minimum high tempeature conditions should be used: (1) BWR Drywells - 3400F for 6 hours; and (21 FWR Ice Condenser Lower Compartments - 3t.00F for 3 hours, 2..
Radittien --Vhen speci'ying radiation service conditions for equipment exposed to radiation during nomal operating and accident conditions, the nort.a1 operating-dose sheuld be added to the d:se received during the course of an accident. Guidelines for evaluating b ta and gama radittien service conditions for general areas inside containment are envided belows - Radiation service conditions for equipment-located dit eitly above the containment sump, in the vicinity of filters, or subnerged in contaminated. liquids must be evaluat'ed en a case by case basis, Guidelines for these evaluations are not provided in this document,
~
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0051563 Gt ca Radiation D=ses - A total gacma dose radiation servire condition of 2 I 107 RCS is acceptable for Class IE.equipm.at located in general L areas inside containment for~ Pkts with dry type containmentsb Where a d:se'1ess than'this value has'been specified', an application specific evaluation must be perfomeo to detemine if the dose specified is acceptable. Procedures for evaluating radiation service conditions
~
in such cases are provided in Appendix 8( The procedures in Appendix ',
- B arit based on the calculation for a typical M reported in Appendix D of WREG-0553I , .
glr:a dose radiati:n servi:e c:nditions for B'As and Ph%s with ice
- cencenser containments mus: be evaluated on a case by esse basis, Since the prccedures in Ap:endtx 3 are based on a calculation for a typical Phi with a dry type contatment,: they are a t. directly a:Plicable to E'@.s and other conttiment types [ F.owever, doses for these other plant configurattens cay be evaluated using similar proceduras with '
censervative cose assu=tiens and adjustment- fa:: ors develeped on a case - by --case bas is. Seta Radhttien Deses - Bett radiation doses generally are less significant than cacca radiation d:ses for equipment qualification. This is due to the_1cv penetrating power of beta particles in comparison to gama rays of equivalent energy, Of the seneral classes M electrical equipment in a plant (e g. ., cables, instr =ent transmitters, valve operators,
. containment penetratiens)., elec:rical cable is considered the most l
WREG-05S3,: Interim 5:aff Position on Envirernental Qualification of
' Safety-Related Electrical Ecuipcent. -
' ~
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.- -- = --
=-
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- 005156a vulnerable to da age frtr: beta radiation. Assuming a TID 14E44 source ter=, the average auxt=um beta energy and iset:pic abundance will vary as a function cf tion folloring an accident. If thr>e para eters art considered in a detailed calculat' ion, the conservative beta st.T face dose of 1.40 x x 108 RADS reported in Appendix D of NUREG OTES would be reduced by appr:ximately a factor of ten within 30 mils of the surface of ele:$ ical cable insulation of unit density. An additional 40 cils of insulation (t tal of 70 mils) results in another .
facter of 10 redu: tion in dese. Any stru::ures or other e uipment in ' the vicir.ity of the equip:er.t of interest would a:t as shielding to furt.ier redu:e beta d:ses. If it can be shown, by assuming a conserva-tive uns.'.ielted surface beta dose of 2.0 x 106 RADS and considering the shieldin; fa:t: s dis:ussed here, that the beta d:se to radiation sensitive equi;= tat internals would be less than or equal to lot of the total gar a desa to which an item of equipment has been qualified, then tha: eqt.i;:=ent may be c:tsidered qualified for the total radiation envirer. int (gc a plus bata). if this criterion is not satisfied the radiation service conti:icn should be determined by the sum of the gacna r.ni bets deses.
- 3. Sutnereer:e -Tne preferred method of protection against the effects cf sutraergen:y is ta locate equipeent above the water flooding level.
Specifying sat:: rated staan as a servica condition during type testing of'equipr.ent that will be:oce flooded in service is not an acceptable alternative for a:::: ally flooding the equipment during the test. 4
005t5G5
- 4. Cer.tainment Seravs - Eqvipment exposed to themical sprays should be quaiified for the ecst severe chemical enviro. ment (actdic or basic) which could exist. Deminerali:ed water sprays should not be execept from consideration es a potentially adverse service condition.
4.2 Se vice C nditiens for a WR Main Steam Line Bre!E (MSLB) inside Centain+ent Equipeent required to function in a steam line break environment must _ be ' qualified for the high te perature and pressure that could result. In sc.e cases the environmental stress on exp: sed equipent may be
- higher than that resulting from a LOCA, in others it may be no more severe than for a LOCA due to the automatic operation of a containment spray system.
,1. Te- eratu e and Pressure Steam Onditiens - Ecuipent qualified for a LO;A environment is considered qualified for a MSLB accident enviren-
. ment in plants with automatic spray systems n:t subject to disabling sing'e c:m; nent failures. This position is based on the "Best Estimate" calculation of a typical plant peak temperature and pressure and a ther a' analysis of typical components inside centainment.M The final ac:eptability of this approach, i.e., use of the "Best Estimate",
is pending the c mpletion of Task Action Plan A-21, Mein Steamline Break Inside Containment. Class IE equipment installed in plants without automatic spray systems o.r plants with . spray systems subject to disabling single , failures or delayed initiation should be qualified for a MSLB accident environment detennined by a plant specific . analysis. Acceptable methods l See .v; REG 0455, Short Term Safety Assessment on the Ervironmental Qualification of Safesy-Related Electrical Equipment of SEP Operating Rea tors, for a more detailed discussion of the best estimate calculation. a
\
0051166 for performing such an analysis for operating reacters are provided in Section 1.2 for Category II plants in NUREG-OSBB, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipeent.
- 2. Radiatien - Same as Section 4.1 above except that a conservative garma dose of 2 x 105RADS is acceptable.
- 3. Submereente - Same as Section 4.1 above.
4 Chemical Seravs - Same es Section 4.1 above. - 4.3 Service Conditions Oe: side of Centain-ent 4.3.1 Arees sub!e:t te a severe Enviren ent as a Pesult c' a Fich Enerev Line Ereek (HELE1 Service conditiens for areas outside containment exp: sed to a HELB were evaluated on a plant by plant basis as part of a program initiated by the staff in Decamter,1972 to evaluate the effects of a HELS. The equipment required to r.itigate the event was also identified. This equiprent should be qualified for the service conditions reviewed and aperovec in the H:' S Sa'ety Evaluat :en Report for ea:s. specific plant.
"'4.3.2 Areas 'there Fluids are Recireviated # rem Inside Centainment to Accereitsh Leno-Term Core Coeline Followine a LOCA
- 1. Te Ser!ture and Relative Hamidity - One hundred percent relative humidity shouTd be established as a service conditien in confined spaces. The temperature and pressure as a function of time should be based on the plant unique analysis reported in the FSAR.
1 i
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i 4 1
m . _. __ ._ _ _ _ _ _ _ . - . _ _ . . _
.,. : 005t567-
- 2. Radiation .Ne' to differences in equipment arrangement within these areas and the significant effect of this factor on dosts, radiation service conditions must be evaluated on a case by case basis. In general, a dose c' at least 4 x 10 'RA05 6 would be .
epected.
- 3. - Sukereenee - Not applicable.
~
- 4.
- Che-icui SY!vs - Not thplicable. '
4.3.3 stets Ner c11v Maintained at Roer Condittens Class IE e:;uipmert le:ated in these areas does not experien:e significant stress due t: a chance in service conditiens'during a dnsign basis event. This equipment was designed and installed vsing standard engineering practices and industry codes and standards (e.g., ANSI, NEMA, National Electric Coda). Based on these factors, failures of equipment in these areas during a design basis event are. expected to be randem except to the extent that they may be due to aging or failures of air conditioning or ventilation - syste .s. Therefore, r:o special consideration need be given to the environ,. ental qualification of Class IE equipment in these areas-provided the aging re:uirements discussed in Section 7.0 below are satisfied and the
- e.reas are maintained at room conditions by redundant air conditioning or ventilation systa s served by the onsite emergency electrical power system.
Equip ant located in areas not served by redundant syst-'s powered from onsite omreeney sources should be qualified for t! s envir aental-extremes which could result from a failure of the systems as determined from a plant specift: analysis. 5.0 . 0'J A1.1~IC ION METdOOS l i l
., - - .. , - -- , , - . . ~. ,-. .
' t 0051568 5.1 Selectien of tualificatien Methed The choice of qualificatien tiethod employed for a parti:ular application of equipment is largely a matter of technical judgement based on such fact:rs as: (1) the severity of the service conditions; (2) the structural and :aterial cec.plexity of the eqrfpment; and (3) the degree of certainty required in the qualification procedure (i.e., the safety importance . of the equiprent function). Based on these considerations, type testing is the preferred method cf qualification f:r electrical equipment located -
insice centain. en: re:uired to mitigate the conset;uences of design basis events, i.e., Class II equipr ent (see Section 3.0 above). As a minimu=, the 0,ualification for severe temperature, pressure, and steam service conditiens fer Class IE equipment should be based en type testing. Qualification for ethat servi:e conditions such as radiation and chemical s; rays may be by analysis (evaluation) supported by test data (see Se: tion 5.3 below). Ex:t::icns to these general guidelines must be justified on a case by case basis. 5.2 Ovali*i ca-t en bv Tv- e Testin: The evaluation of test plans and results should include consideration of the f:llowing fact:rs: 1. Simulated Semice Conditiens and Test Duration - The environment in the test chamber should be established a id maintained so that it envelopes the service conditions defined in accordance with Section 4.0 above. Tne time duration of the test should be at least as long as the period frc= the initiation of the accident until the temperature and pressure
~
service c:nditions return to essentially the same levels that existed hsfore the postulated accident. A shorter test duration may be acceptable
00515d9 if specific analyses art provided to demonstrate that the materials ' involved t 11 not experience significant accelerated themal aging during the period not tested. .
- 2. Test Soecimen - The test specimen should be the same inodel as the equipment being qualified. Th'e type test should only be considered valid for equipment identical in design and material construction to the *;:st-specimen. Any deviation,s should be evaluatee et part of the ( ,
; fica-
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' tien documentation (see also 'Section 8.0 below). . ,
3 Test Se:uence - The component being tested should be exposed to a - stet:./ air environment at elevated temperature, and pressere in the sequence defined for its service conditions. Where radiation is a service condition which is to be considered as ptrt of a type test, it r may be toolied at any time during the test scouence provided the component does not contain-any materials which are known to be susceptible to significant radiation damage at the service condition levels or materials whese susceptibility to radiation damage is not known (see A;:endix C). If the' component contains any such materials, the radiation dose shculd be applied prior to or concurrent with exposure to the elevated tes;erature and pressure steam / air environment. The same test specimen should be used throughout the test sequence for all. service conditions the equipment is to be qualified for by type testing. The type test should only be considered valid for the service conditions applied to the same test specimen in the appropriate sequence.
-4 Test Sce:fren
- cine - Tests which were successful using test specimens which had not been creaged may be considered acceptable provided the I
corpenen: does not contain materials which are known to be susceptible
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_a.,. --... 005157l' ' 4 5.3 Ovalification by a tambination ef Metheds (test. Evaluation. Analysis As discussed in Secticn 5.1 cbove, an item of Class IE equipment may be shown to be qualified for a complete spectrum of service conditions even though it was only type tested for high tem;erature, pressure
, and steam. The quali,fication for service conditions such as radiation and chemical- sprays may be demonstrated by analysis (evaluation). In such cases the ovtrail qualification is said to be by a combination of
~
me: hods. ollo.in; are t.c specific examples of crocedures that are '
' l cor.sidered acceptable. Other similar procedures may also be reviewed and fc.nc nece;tacie on a case by-case basis.
(
- 1. Radiation oualifientiea - Some of the-earlier tvo* tests performed i 2
'for operating reactors did not include radiation as a service condition. In these cases the equipment may be shown to be radiation qualified by performing a calculation of the dose extected, taking into account the time tne equipment is required te remain functional and its location using the methods described in Appendix B and analy:ing the effect of the-calculated dose on the materials used in the equipment-(see Appendix C). As a general rule, the time required to remain functional assumed for dose calculationr. should be at least 1 hour.
- 2. Cheefeal Serav Dualification - Components enclosed entirtly in corregion: resistant _ cases (e.g. stainless steel) may be shown to'be qualified for a chemical environment by ar. analysis of the- effects of t'; particular chemicals on tie ; articular enclo-sure materials. The effects of chemical spra/s on the pressure <
integrity of any gask' ets or seals'present shculd be considered , in the analysis.
0051570 to significant degradation due to.ther .a1 and radiation agir. . (see Section 7.0). If t.he w,,enent centains such materials a qualified life for the cocponent r.st be established on a case by case bas 15. Arrhenius techniques are generally considered acceptable for thermal aging.
- 5. Functional Testine and Failure Criteria - Operational modes tested
. should be representative of the actual application requirements (e.g., c:=penents which operate ner= ally energized in the plant should be nemally energized during the tests, meter and electrical ,
cable leacing during the test should be representative of actual operating cenditiens). Failure criteria should include instrument accuracy recuire ents based en the maxim.:m error assu ed in the plant safety analyses. If a component fails at any time during ( the test, even in a so called " fail safe" mode, the test should be censidered inconclusive with regard to demonstrating the ability of the co penent to function for the entire period prior to the failure.
- 6. Installation Inte da:es - The ecuipment neunting and electrical or mechanical seals used during the type test should be representative of the actual ins *allatien for the test to be considered conclusive.
The equipment qualification program should include an as-built ir.spection in the field to verify that equipment was installed as it was tested. Particular er:phasis should be placed on co:::non pmblems such as pmtective 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 conductors.
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0051572 ; 6.0 Marcin IEEE' Std. 323-1970 6 fnes margin as the difference between the most severe specified se vice conditions of the plant and the conditions used in type testing to account for nonnal ' variations in coernercial production of equipctent and reasonable errers in defining satisfactory peNomance.
- Section 6.3.1.5 of the standard provides suggested, factors to be applied
'* to'the service conditio*ns to assure adequate margins. The factor applied to the time equipment is required to remain functional, is the most ,
significant in terms of the adcitional confidence in qualification that-
- is achieved by adding margins to service _ ccnditions when establishing test environtents. For this reason, special c:nsiceration was given to the tir.e required to rec.ain functional when the guidelines for Functional
, Testing and Failure Criteria in-Section 5.2 above were established. in addition, all of the guidelines in Section 4.0 for establishing service conditions. include conservatisms which assure :rargins between the service
. conditions specified and the actual conditions which co',.1d realistically be expected Lin a design basis event. Therefore, if the guidelines'in
, Section 4.0 and 5.2 are satisfied ano separate margin factors are required '
to be added to the service conditions when specifying test conditions. 7.0 - Acin:
-Implicit in the staff position in Regulatory Guide 1.89 with regard to backfitting IEEE Std. 323-1974 is the staff's conclusion that the-Lincremental improvement in safety from arbitrarily requiring that'a -
specific qualified life be demnstrated for all Class IE equipment is not sufficient to justify the expense for plants already constructed and operating. This position does not, however, exclude equipment
4_ - m m. _, 4 ,, 15 - ' 0051573 Using r.ateritls that have been identified as being susceptible to significant degradation due to therr.41 and radiation aging. Component ' maintenance or replacement schedules should include considerations of the specific aging characteristics of the component r.aterials. -Ongoing programs should exist at the plant to review surveillance and maintenance records to assure that equipeent which is exhibiting age related degrada-tion will be identified and replaced as necessary. Appendix C contains a listing of nterials which may be found in nuclear power plants along with ' an indication of the material susceptability to themal and radiation aging. E.0 Docu entatier.
' C=plete 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 detail to verify that all of the guidelines have been satisfied. A simple vener certification of compliance with a design specification should not be considered adequate. M 4 4 0 e 9
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1 M LPPENDIX A , JYP1 CAL E0'JIPMENT/ FUNCTIONS NEEDED FOR HITIGATION OF A LOCA OR MstB ACCIDENT
;Er31ceered Safeguards Actuation Reactor Pmtection containment 1 solation Stesmline Isolation
'~
Main Feedwater Shutdem and Isolation'- Emergen:y foner. Eser;ency Core-Coolingl Containment Heat Reeval Containment Fission Product- Removal
-Containment Cembustible Gas Control Auxiliary Feed ater Containment Ventilation Containment Radiation Monitoring
- Control P.com Habitability Systees (e.g., HVAC, Radiation Filters)
Ventilation for Areas Containing Safety Equi; ment Component Cooling
- Service Water Emergency Shutdown 2 Pos: Accident: Sampling and Monitoring 3
- Rsdiation Monitoring 3
' Safety Related Display Instrumentation, l
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0051575'- 1 These plants.systems wil1 ~ diff'er for PWRs and SWRs, and for eld r and newer In each case the ' system features which allow fo transfer to
. recirculation cooling mede and establishment of long term cooling with boren precipitation control are to be considered as part of the system to e evaluated.
2 Emergency shutdown systems include those systen:s 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 together ~
- with a rapid depressurization of the reactor coolant syste. Examples of such syster.s and equipment are the RHR system. PORVs,- RCIC, pressurizer.
sprays, chcical_ and volume control systam, and steam dump systems.
^~-
kore specific' identification of these types of equipment can be found ', in the plant emergency procedures. , a m . 4 e e P S
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AppExoIx s . WW6 PR30D'Sls FOR EVALUAi!N*, MM p.ADIATION SERVICE COND!TIO55 Introduction and Distussion The adequacy of gama radiation servire conditions specified for inside contairment during a 1.0*A or MS!.B accident can be verified by assuming a conse:vative dose at the containment centerline and adjusting the dose according the plant specific parameters. The purpose of this appendix is to identi.fy those parameters whose.effe.ct on the total garr.a dose is easy to quantify with a high degree of confidence and describe procedures which may be used to tale these effects into consideration. ' The bases for the procedures and restrictions fe- their use are es follows:
. (1) A conserva:ive dose at the containment centerline of 2 x 107 RADS for a LO~A and 2 x 106 RADS for a P.51.3 accident has been assumed.
His assum;: tion and all the 'de'se rates used in the procedure out-
~
lir.ed below are based en the rnethods and sample calculation desertted in Appendix D of NUREG 0553, " Interim Staff Position on Envirencental Qualification of Safety-Related Electrical Eoyip-cier.t, " Therefore, all the limitations listed in Appendix D of h'JRIs 05SS apply to these procedures. (2) The sample calculation in Appendix D of NUREG-OSES is for a 4,000 MWth pressurized water reactor housed im a 2.52 x 105 fg3 contain-c:ent with an iodine scrubbing spray system, A similar calculation without iodine scrubbing sprays would increase the dose to equipment appmximately 15t. The conservative dose of 2 x 107 RADS assumed
a
. y.... .,
.in the procedure below includes sufficient conservatism to account for this factor. Therefore, the prendure is also applicable to plants without an iodine scrubbing spray system.
(3) .. Shielding calculations are based on en average gamna energy of 1 MDt derived'from TID 14B44
. (4) ' These'procedurts are not applicable to equipment located directly '
above the containment sump, submerged in contaminated liquids, ~ or near filters. Doses specified for equipment located in thes.e areas must be evaluated on a case by case basis. ' (5) Since the dose adjustment factors used in these procedures :.re
' based on a calculation for a typical pressuriced water reactor with a dry type containment, they are 'nct directly applicable to
, boiling' water reactors or other containment types. However, doses for these other pla '. configurations may be evaluated using similar procedures with conservative dose assumptions and adjustment factors developed'en a case by case basis, i precedure ~
Figures ~ 1. through 4 previde factors to be applied to the conservative
~
dose to correct the dose for the following plant specific parameters:
;(1) reactor power level; -(2) c:ntainment volumet (3) shielding; (4) co=pe,rtment volume; and -(5) time equipment is required _to remain functional.
l
. _= .
3, 0051577 The procedure for using the figures is best illustrated by an example. Consider the following case. The radiation service condition for a particular item of equipment has been specified as 2 x 105 RADS. The application specific parameters are: * ' Reactor power level - 3,000 MWth Containment volume - 2.5 x 106 ft 3 Compartment VIlume - 8,000 ft 3 Thickness of cor:partment shield wall (concre.e ) - 24 " Time equipment is required to recain func:ianal - 1 r.r. The pt:blem is to make a reasonable estimate of the dose that the eK;uipment could be expected to receive in order to evaluate the ade:;uacy of the radiation service condition specification.
- Stes 1 Enter the nomegram in Figure 1 at 3,000 MWth reactor power level and 2.5 x 106 f3 containment volume and read a 30-diy integrated dose of 1.5 x 107 PADS.
Stes 2 Enter Figure 2 at a dose of 1.5 x 107 MDS and 24" of concrete shielding for the compartment the equipment is located in and read 4.5 x 10 4 RADS. l ( This is the dose the equipment receives from sources outside the compart-ment. l To this must be added the dose from sources inside the ec=partment (Step 3). Steg 3, En' tar 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 'O.13 (1.5 7 x 10 )
= 1.95 x 10 6 RADS. The sums of the doses from steps 2 and 3 equal s:
4.5 x 104 RADS + 0.13 (1.5 x 107) RADS = 2.0 x 106 MDS
- b . ...
. . - 4-005t578 Stec-4 -
Enter figure 4 at 1: hour and read a correction factor of 0.15. . Apply this factor to the sum of the doses detemined from steps 2 and 3 to
- correct the.30 day-total dose' to-the equipment inside the compart:nent to 1' hour.
0.15 (4 0 x 10 6) = -3 x 105 RADS In this particular example the service condition of 2 x 105 ggg3 specified is conservative with respect to the estimated dese of 3 x . 10 3 RA 5- calculated in steps l'through 4 and is, therefore, acceptable. - ' e s.
.-l 4'
9 7. r e
- 7 - , y 1m . _ _ . - . . - - .
+ .: -
FIGURE 1
, , . t. t . . - - E MRAM FOR CONTAIF.ENT VOLW.E A!. AEACTOR POWER Lo:A cosE c0mcT20ns. 0051579
~
7 0NTAINMENT
' VOLUME (ft2) 3 x 10s -
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