ML20115G357
| ML20115G357 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 06/08/1996 |
| From: | Drucker M, Sayles C SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML20115G352 | List: |
| References | |
| N-0480-003, N-0480-003-R04, N-480-3, N-480-3-R4, NUDOCS 9607190088 | |
| Download: ML20115G357 (120) | |
Text
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bALCULATION TITLE PAGE C 'Cy NO. ,AoE _ o, _
CCN CONVERSION:
- Calc.No. N =aM3 DCP/FIDCN/FCNNo.&Rev. N/A CCN NO. CCN.
t Subject FUELTRANSFER TUBE SHIELDING Sheet L,of 117 j System Number /Pnmary Station System Designator 2103 / XC1 SONOS Units 2 and 3 Q-Class III i
} Tech. Spec. Affecting? E NO YES, Section No. Equipment Tag No. not annlicable i
OONTROllID FROGRAMDATABASE NAME(S) VERSION /RELIASE NO.(S)
COMPUTER E FROGRAM PROGRAM / O AlsO,umDBELOW DATABASE D DATABASE i ACCORDING TO f,0123. XXIV.S.1 SONE 2 M SHIELD-BG D2 M D210.Ma RECORDS OFISSUES
_BZV, L DESCRIPTION SHTS FREP M APPR M DISC. (Prins nama/ sign /date) (Signature /dste) a WT SH'I'.
0 to 3 SEE SHEETS 4,5,23 AND $2 FOR N/A ORIO. FIS h REVISIONS 0 3DESCRITDONS.
j Bechtal PAGETUTALS,SIONA1URES, AND N/A IRE Other Other a
DATES 4
s S E NOTE R BEIDW M 04Bo-016 Ro Total.117 0 RUCKER h/h _ g g.g gg Other NFM last:187 IRE C.W.SAYLESI dh y/K,
, Other Other ORIO. H3 h j IRE Other Other i
ORIO. FIS Other IRE Other Odur Space far RPE Stamp,idanufy use of no sharnate calc., and notas as appbcable.
NOTE 1: The purpose of Revision 4 is as follow:
- 1) Revision 4 evaluates as-built shiciding and SONOS-specfic source term data to detenmne the dose rates in and around the fuel transfer tube for various concrete thicknesses
- 2) Due to the historical nature of prior revisions, Revision 4 does not delete any text from Revisions 0 through 3. Revisions 0 through 3 were each individually pagmated. To facilitate future use of this calculation, Revision 4 repaginates the entire calculation such that the pages are sequential.
- 3) Pages added: currently numbered pages 1,2,3, and 67 through 117 Pages revised: all(as required by item 2)
Pages deleted: none
- 4) The Quality Class III (non-safety related) designation is consistent with previous revisions of this calculation
- 5) System Number 2103 and System Des gnator XCl represent the Penetration Area, which is the shielded area ofinterest.
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I This Calculation was typed using " Wordperfect 6.l' as an electrmic typewiter. " Wordperfect" was not used for any computational j portions of this calculation.
This cale. was prepared for the identifal DCPNCN. DCPSCN completion and tumover acceptance to be venfied by receipt of a memorandum directma DCN Conversion. Upon receipt, this cale. represents the as-built condition. Memo date by 9607190088 960716 PDR ADOCK 05000361 l P ppg i
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gNPUT TNE DOS 00,0,N.TE,E..., POINT F IN c ARTESI AN C00PDIN ATES w,. _, .. T C0040!NATER 76*25 F (. TERMINAL INPUT !$ COMPLETE * ,
- (
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c . Myffy ToL % /306 fSI f. <' 54//~e
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'00$E P0!NT8 X"C00Rc!N.Afg
(
$*50000 00 F Y".C
. ....O OR....
6*25000 00 F D..I N-AT E yy7 g i '
~
/ &C[ ~ hh0-3-0 DISTANCE INTO LAS7 $HIELO (NO. 2) a 1 00000 00 F
( ( ENER67 00SE RATE I su! LOUP (MR 8.ROUP
... ......s .aa./HR) 1 3*1640 02 1 1380 03
( 2 3 6740 01 4 8550 04 1*2490 04 3 1e9240 0) 4 1 1290 01 2s7520 05 I 5 8*3330 00 1*8340 04 6 6e7970 00 4 7780 01 7 5 8270 00 8 7250 02 I $ 4*9790 00 3e0060"02
'f0TAL 3*5670 05
( d<( ( ( ( ( ( ( ( (
. .. w.;~.................... ............. ...
M,Q f *d foi Hs./364- (31 / f. Tij/- 4
/d[;jf]
TYPE '1' IF YOU WISH 70 INPUT A NEW DOSE POINie ' 12e tp you W!$4 TO RERUN THE PROGRAM
- I ',3 3 IF YOU W!SH TO TERMIN ATE TNE RUNI>1 b/E(7 hM INPUT TNE DOSE POINT IN CARTESIAN COORDINATES 4
q 'X'C00RDINATEI >6 5 , N-YBo O Y C00RotNA TE8 >6*25 F l TERMINAL INPUT !$ COMPLETE * , ( ( ( ( ( ( d( ( ( ( ( ( ( . ( O-
t
- .' %wim
/4 73 Tab m.1304- 23/ )?.g. sL i> s foffg V*
...C.O O R D.I NATg la , b f
( "90$E POINTI ...... X*COORDI.N 4 ... ATE ... .... 4*50000 00 F 6 25000 00 F U O q O* DISTANCE INTO LAsi 8HIELO (N0* 2) a , 2e00000 00 F N/-d N" NO~ 3*N ( ( ENER8Y COSE RATE ( GROUP 8UILOUP (
. .... 4.... 4 6.MR/NR) 1 1 0010 03 3 7230 00
( 2 6 2300 01 444550 02 3 289530 01 te3720 02 4 1 5990 01 9e0080 03 ( 5 1 1310 01 848930 02 6 Se9480 00 340840 00 7 fe5090 00 6*9680 01 k 8 6 2890 00 2 9140"03 TOTAL 140660 04 ( l d<( ( ( ( ( ( ( ( A' V (
.-...n ..... ......... .... . .... ...
( y ,, D 50
- lSbY' $bl f f. Y-l,Y=
TYPE '1' IF YOU WISN TO INPUT A NEW DOSE PO!Nie
-(
'2' IF YOU W!$N To RERUN THE PROGRAM
- 63 9 IF YOU WISH 70 TERMINATE THE RUNi>1
/d/p/f g .
p '>dth 5.m.g i l rNPUT THE DOSE POINT IN CARTESI AN C0ORDINA TES g X C00R0!NATE8 >7 3 P
$'/fE/ M S M *(
Y C00RDINATEt >6*25 7 QL Nyg-3 g I TERMINAL INPUT !$ COMPLETee , ( l ( (, ! t i ( l ( l l j 4 l U((. l k
)
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, (
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( N*d Ww'y f6 7j h f. & ( 'sosE POINT: ..c0x- Y-c00RDIN ORDINATE ......... ATE.
/ 0 / 3 0/ 73 f.$0000 00 F 6 25000 00 F
b Shcd16* W6'I'-k,,
Q' OtsTANet INT 0 Last sNIELD (No. 2) = 3 00000 00 F . g Cm N.qgg 3 o 1 ASSUMPTIONt USED IN DETERMINING sgLF ATTENUATION 8
- ENER G Y 4ROUP 13 MEWS ( A+R A0108)s22 e 3$7 3 ' MEWS C A+R ADIUS) SET TO 20 I ENERGY GROUP 18 81827*8311 si'8ET'TO 15e
' ~ ' '
I ENERGY 00$E RATE GROUP gutLOUP (MR/NR) 4..... ...e.a. ......... 1 2*8690 03 1*4760*02 2 947700 01 5 0080 00 3 441870 01 $d5370 00 ( 4 281340 01 ti9620 02 5 1e'4630 01 4 0040 01 6 1 1330 01 1*7180-01 ( 7 9*3640 00 4d7050 00 8 746890 00 245260"04 TOTAL 345170 02 ( 1 ( ( l ( ( ( s ( (
IF YOU WISH TO INPUT A NEW OOSE Po!NTe* TYPE '1'
'RI IF Y UO WISH 70 RE UN R THE PROGRAM * /g!30N3 d b '38 IF YOU WISH TO TERMINATE THE RUNf>1 g 7 5fy%
O y INPUT THE 00$E POINT IN.4ARTE8 TAN COORDINATES X COORDINATEI T C00R0!NATEt
>8 5 f
>6 25 F AN MY[O' 3' O
( TERMINAL INPUT IS COMPLETEe , i ( ( i ( l i 4 (
- (
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. ~ ~ ~ : .' '" .
( kA%3 Nd ' +
- 7 c A
/s/
/W14/73
+ - ( *008E POINT! .a X .C00R0!NA.TE
....... . .Y".C O O R.D. I.N A.T E 7
i 8 50000 00 F
. a....
6 25000 00 F 3;7gy gfig 3r O '
~
act u4<so-3-0 l O!8TANCE INTO LAST $N!ELO (N0s 2) a , 4 00000 00 F I ( ) A SSUMPTIONS usED gN DETERMINING SELF ATTENUATION! ( ENERGY GROUP 11 NEWS C A*R A0!US)825 *338 4 MEW 8 C A*R ADIUS) SET To 20* ENERGf GROUP 11 81834e1278 31 SET'To 25e , i ! ( i ( (NE RGY 00$E RATE GROUP (MR/ tu.! LOUP aa...NR) l ... u . . s. s. .... } I 1 fe7100 03 6 8810 05 ! 2 1 4900 02 5e3570*02
- 3 547040 01 143860*01
' ( 4 2*7090 01 1e1950 01 i 5 118070 01 260980 00 6 1 3790 01 1 e'04 5 0
- 0 2 I f 1 1290 01 3 2180-01 3 941990 00 1 9290 05 .-
TOTAL' 1*4580 01 1 j ( i
- ( ,
l 1 ( f ( ( d . f
( ' 5 ',A, ^ ^
.* TYPE '1' I F YOU WISH TO INPUT A NEW DOSE POINie 82' 3p YOU WISH TO RERUN TMs PA0 TRAMS ' /ggg/p
. (. e5' IF YOU WISH 70 TERMINATE THE AUNIP1 m% u% MEE7&o , y i 4
' OX g INPUT TMg 0085 POINT IN,CARTgSI AN COORDINATES C00RDINATE8 Y C00RDINAfg8 76 25 F 79 0 p
, , [gg gjd,3-0
! ( TsRNINAL input 1: COMPLaTae i i ( i I ! ( ( ( ( ( ( (
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\
. . _ . . _. . . _ . . . . . _ . . . . _ . . . . . . . _ . . 1 E , __ . m ,._
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'f 20st POINTI X*&00R0.!NAf
....... .... g T*
..C.0..
0.R... 01 v.a..T
. g
/d[,3 73 >> <
9300000 00 F 6 25000 00 F SHEET l l
&cl#1'?6-3-O '
1 01 STANCE INTO LAgT $N!gLD (N0s, 2) a 4 3 50000 00 F -' ' ( ASSUMPTIONS USED IN OgTgRMgNINg $g(F ATT ENUATION3 ENER8Y G ROUP 18 MEW 8( A'R A01U$)s26*8284 WEW5( A+RADtWS) ~ SET
~ * " TO 20'
( ENERGY GROUP 18 61837 2754 s1 SET 'To 25 i ENERGY GROUP 28 31m2649238 81 SET TO 25', i ( l i ( fNERsY 00sE RATE GROU ( ( .....P. g
. u.g.LO.U
.. P ..MR/NR) 1 1+2520 04 4 7210*06 ,
2 148100 02 640880*03 l I 3 646040 01 te1780*02 4 3 0240 01 2 4170 00 5 1*9850 01 Sa'0150'01 l 6 145000 01 248110*03 7 162250 01 Si8740*02 8 9 e'9 360 00 546540*06 l
' TOTAL 3*,0370 00 y.
( ( ( ( ( (*~ .
- y. .
CALCULATION TITLE SHEET 33 e,
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PMOJECT b JOS NO. DISCIPLINE ( uS>ECT EUE C 'T/?^WS /~6 /2 "YUDE S///fil) /A/f7, IEEK 1 ,,L, n 6C4-A
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d D td,hAlcn DATE r- to - 7r ** 'ALITY CLASSIF, zit OU
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l LEVEL Or Review h @ @ CHECK AS REQUIRED P.C. ST AMP IF R EQ'D
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ORIGIN AL i: SUE ,
. NAME DATE SIGNATURE h GROUPLEAD N M d, e- A EGS
- X f* . .% l l Y o' =^^b E 5 h SPECIALIST h CHIEF .
OTHER l RECORD OF REVISIONS N O. REVISION DATE ENG. CKR EGL EGS SPEC. CHIEF a#A l A A A ' t berg, tact iNClUbE3 l~ A E *- t i
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DATE pRosEcr ? S o A/&S V2/3 ,o, , /00?9 suaaEcr FUEG TRANSF&c 72)D F S ///ELDodf sHE -l ? ~~ OF V SHEE S
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, /9 73 , p &-/%
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LAO MIS 673 l
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CALCULATION CHEET #' 0-3J CALC. NO- f SIGNATURE / U DATE CHECKED DATE W3 FROJECT S Oh'M me no. /06 7 ? SUSJECT U 5 'VA'0TSf/I EO C' INIEA);Will?' SHEET "~OF AS S EE S 3 3 WES U LTS 4 SHtELD WAcc T E q ate tf r7E ^!T S Arti PG0Wh 0^I PA W Fok. Aal Att2 64 P BEk/t-Ed 7/Il TLISN EN6&L A^'b 1///E S C. s Hall t? & QJitt 6/nf MT3 AWE: 10 i 31 l l l 2A D IA Tie ri ii' C Ne~ RE Q )ll?fb W4 iL t/m fr (mallt) THH k NES S (PT) 14 I
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CALCULATION CHEET fj. i cat.c.No 980-3-) SIGNATURE / DATE # CHECKED - DATE /I FlOJECT O d N $'~d JOB NO. ! SUBJECT [upC 77/W//~M 77/3r fg/g&'o/ M SHEE OF S E t
- 2
} F0K IM sd jsYoT Q GNGEa/ T//&' 7"OE E 3N&L s*4Alb J'Y/6LA' 4/ALL, /80 AI (trtAGNE 1/7[ } b En2iT/dJ oF 30 2, ~?ib , i4 N'b s .
/0 o ?u oF .3/ff23//7 LJ Ett E O2Eb. Conc n GTE 2 ScA> teEirisarr's M e-e 9
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" AN ,4LTE RNAT/WE 70 S///EO w& THE fEnm/c GA A3' m
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LAO M38673
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CALCULATION CHEET
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- a. . . . . . . . . . . . ' / - - -
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3 4 1.5490 01 1.0560 04 1.1040 01 -
- ' ~ ~ - - - - ~ ~ - - ' - - -
5 c.4140 02 6 R.759D 00 3.0120 00 7.362D 0 0 -- 6.3350 01 ~ ~ -'- -- - - - - 7 8 6.161D 00 2.561D.03 -
- - T OT A L- - --- ~ 1.2 4 7 D 0 4 ~~ ~ -- ~---
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DOSE POINT X. COORDINATE Y. COORDINATE LAST SHIELD t NO ; ~31
.. _ _ . _ . - - ' - ^ - - - ~ - ~ ~ ~ ~ -
... _ _ .. 2 ---' 8.50 0 0E *0 0 P 67250 0E 10 T - 4. 0 0 0 00 0 0 f
~ ~ ~ - ---
D O S E R A T E --~~~ " - ~ ' ~ ~ ~ ~ ENERGY GROUP BUILDUP (MR/HR) - -
.g ...... ....... .........
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1 2.542n 03 2.3660 02 ' - - - - - - --- 2 0.0100 Ol' 8.441D 00 - - " - - 3 3.914D 01 9.0800 00 - - ~ ~ - - ~ - - ~ ~ ~ l _.4 .__ _ _. p , 0 2 8 0 01 ' 4.8350 02 ~ '- j 5 1.4010 01 5.8990 01 ' - ~ ~ ~~ ~ ~ ~~-- 6 ~~ ~~1.0890 01 ? . 32 9 D.~01 -- 7 0.0370 00 5.686D 00 . .
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, .g ..7 - TOTAL 5.660D 02 l DISTANCE IN 1 A uCSE POINT ~ X.~COORD f N ATE ~Y.C00ph!N ATE- TA ST SHIELD ~~ ~ ~ ~~- ~~~~- ~~ -" - I
- (NO. 3) 9.5000E 00 F 6.2500F 00 F 3
5.0000D 00 F ENERGY DOSE RATE ' ~~~~~ ~"~ ~ 6ROUP~~- BUILDUP ~' " ~(MR/HR)~ ~ ^~"-'~
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1.099D.04
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1.803D.02 . ... .. ... . - ..- - i
. . . 8 _. . 8.8050 00 P.7890 05
~ ~ P.674D Ol'-
. . . - ~~ -~~ ~ ~ --
' TOTAL l
i
- . - _ _. .. ~ . .. . . .. _. . .... . .- _
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( CYLSO RUN OF 04/10/75 tT 10.49.39 , ENENGY ENERGY SRC STRENGTH OflS 4 GROUPS (MEV) (MEV/CC*SEC) hc/ % )c-1-l ( 5.000000 01 1 1 7.280000 11 2 1.000000 00 1.920000 12
- ( 3 1.50000D 00 8.670000 10 4 2.00000D 00 7.77000D 11
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.......... ...... ____. ............ ..........__ l
( WATEP 3.750000 00 7 6000D-01 F 1.25000 01 F SHIELD MATERIAL DENSTYtG/CC) THICKNESS ( 1 WATER 1.00000D 00 1.12000 00 F 2 AIo 1.300000-03 3.00000 00 F ( 3 C09CRFTE 2.300000 00 1.00000 02 F 4 DISTANCE IN
- ( DOSE POINT X. COORDINATE Y. COORDINATE LAST SHIELD ;
- (No. 3)
O .......... ............ ............ ............ (- 1 7.5000E 00 F 1 25G0F 01 F 3.00000 00 F ; ENERGY NOSE RATE ( GROUP BUILDUP (MR/HR) 1 9.0830 02 P.547D 00 ( 2 E.803D 01 2.9550 02 3 2.793D 01 1.539D 02 4 1.5510 01 5.3090 03 ( 5 1 1070 01 4.7480 02 4.786D 00 1.524D 00 6 7 7.390D 00 3.2150 01 l ( 8 6.1900 00 1 304D.03 j . TOTAL 6.2700 03 1 ,i ( DISTANCE IN
- ( DOSE POINT x. COORDINATE Y.COOPDINATE LAST SHIELD (No. 3) l ............
) ( 2 A.5000E 00 F 1 2500F 01 F 4.0000D 00 F 4 ENERGY DOSE RATE - j GROUP BUILDUP (MR/HR) 1 2.543D 03 1.1840 02 ( 2 9.016D 01 4.2260 00 3 3.918D 01 4.5550 00 ! 4 P.0310 01 2.4310 02 ( 5 1 404D 01 2.976D 01 6 1.0920 01 1 1700 01 , I
~r - ~ 9.oe90 00'- ~ P.seu'00' 'f '
( - - 8 7.5020 00 ).3600-06/g ., kg 4/ .g.7f fgg j[op , TOTAL 2.8c70 02/g ,y, gggy jpj79 $3-( eu 7 m s s viu y ire: mis pdAcl N-V 60- 3-/ DISTANCE IN 7/sA J O- DOSE POINT x-COORDINATE Y-COORDINATE LAST SHIELD (NO. 3) ( 3 9.50000 00 F 1.2500E 01 F 5.0000D 00 F ENERGY DOSE RATE GROUP OUILDUP (HR/HR) I 6.8740 03 5.498D-05
, 2 1.3530 02 6.0420-02 3 5.3130 01 1.3460 01 4 2.5720 01 1.1130 01 , 5 1.T240 01 1.8670 00 6 1.3160 03 9.137D-03 7 1.081D 01 2.599D-01 6 8.8400 00 1.4210-05 TOTAL. 1.3460 01
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- DATE FROJECT 0NY A JOB NU.
YbO ?Y - SUBJECT NE0 bA [M N/N SHEET ' OF 6' 'S T , s
- 74S' ce pescETE s/Act irr</s7 73 e~
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CYLSO RUN-OF 04/08/75 AT,14.52.16
. l.
I V-Cec . $V l ' L. ENERGY 'ENER8Y " $RC JSTRENGTH GROUPS (MEV) (MEV/CC*SEC) . ..
, 1, -
5.00000D 01 . -
~
2 1.00000D 00 1.920000.12 7.28000011[k ,, 3 1.500000 00 8.67000D 10 a. '
.. ,.~ ...
4 2.00000D 00 7.77000D 11 #[f ' j ~*- ~ n ,' 5 2.600000 00 2.570000 10
~ ~~
-6' 3.000000 00 ?"' " .' ~ ' ' . - .
4.220000 07 h-
'T _._3.50000D 00 l 5.68 0 0 00 04'08 '0" ~~' C' -
8 -4.000000 00- 1.47000D
' ' ~'
h" ! 1 r / .- SOURCE NTL DENSTY(6/CC)' DIAMETER LENGTH WATER 3.750000 00 7 6000D 01 F 1.2500D 01 F '. J SHIELD MATERIAL DENSTYtG/CC) ! THICKNESS 1 WATER 1.00000D 00' 1.1200D 00 F ' 2 IRON 5.046000 003.0000D 00 F ' '
}
3 CONCRETE 2.300000 00 '1.00000 02 F e' DISTANCE IN J DOSE POINT X-COORDINATE Y.C01RDINATE LAST SHIELD .
. . .(.........
No.3)
< (:) . ....... . . 6...0000E 1
..........00 F 1.50000 00 F 6 2500.E 00 F.
ENERGY DOSE RATE .: GROUP BUILDUP (MR/NR) l 1 7.6440 04 5.4820 11
- 2 2.762D 02 2.3210 06 3 9.2380 01 3.897D 05 t.
4 4.349D 01 8.8870 03 "
. 5 2.9770 01 2.583D 03 6 2.3000 01 1.8190 05 7 1.942D 01 6.1410 04 ,
~ '
3.9210 08
~
.- 1 6060 01-
^
8' *
' TOTAL . 1.214D.02.. ..
.- s - .,? . _ _ _
~
.- .mv. t . .
e s , w .& '
~ '
' ~
9 '
~ l DISTANCE IN
~
I' DOSE PDINT x.CDDRDINATE' Y. COORDINATE LAST' SHIELD (No. 3)
~f 'l
.- .e........ 52...........s"
.p -
2 7 0000E 00 F 6 2500E 00 F 2.5000D 00 F ENER6Y DOSE RATE
~4 ,
() GROUP BUILDUP (MR/HR) 1 2.320D 05 2.8350 13 2 4.2920 02 3.400D.08 a 3 '1.2270 02 1 1170 06 i 4 5.2150 01 3.8080 04 . f 5 3.420n 01 1.500D.04 2.5830 01 1.3000 06
)) _
6
- ^~ ~ - ~~
l..' 2 149D U1 b.059U-US ~ I j i 8 1.757D 01 3.7770-09 ff
, 94y3. je .
5.8410-04
~ ~ "TDTAL i:
SONGt U MJ7
~
l . _ p,p 7ksa%-x 7v8 r sHi& Lb/NC . ', . j
$ OntA h $0=3=/
DISTANCE IN $fyp OM { X-CDDRDINATE Y-CDDRDINATE LAST SH ELD ppy (}D'DSEPDINT ,q, l :i 3 8 0000E 00 F 6.2500E 00 F 3.5000D 00 F
. ~ . . .-. . . . . _ . -. ~- -
t i ENERGY . DDSE RATE .. l GRDUP BUILDUP (MR/HR)
~
i 1 6.259D 05 1.3390-15 . 2 6.415D 02 4.922D-10 l 3 1.602D 02 3.2310-08 4 6.188D 01 1.6610 05'
~
I , -l 5 3.9040 01 8.8870-06 4 6 2.884D 01 9.493D-08
- - . 7 2.366D 01 4.3070-06 i
8 1.9120 01 3.7160-10 TOTAL 2.9940-05 4 e g 1 W . 6 9 4 1b *
- 9:
e .- .. .
"J O
< i e .
----J--------- -- _.. _ _ __ _ _,
0l k Y*bO W $6N$$Sh /0079 ShEE$foi CYL 0 CUN OF 04/09/75 AT 08.13.09 %
~
$ W A M & &? N R NAlf . hb ,
gl. "~ge. g.) '. j 1 ENERGY ENERGY SRC STRENGTH '
- 6R00PS (MEV) . (MEV/CC*SEC) .
i 4 0 1 5.000000-01 7.28000D 11 .' i ...2, ' 1.00000D 00 1.92000D 12f, ... .t. .. y l 3 1.50000D 00 8.67000D 10',. ~
. (, 4 5
2.000000 00 7.77000D 11~ . '
'2'60000D
. 00 2.57000D 10 '.
! 6 ~3.00000D 00 -.4.220000 07 ', ' (, ~7 -'3.500000 00
- " ' - ' ~ ' '
5.680000 08 '? T . , i j . 8 -4.000000 00 ._ 1.47000D 04 . . . _ . 2' ! g, SOURCE MTL SENSTY(6/CC) 'DIAMETEP ~ LENGTH , j WATER 3.750000 00 e 7 6000D-01 F/1.2500D 01 F / ; i 0 SHIELD j MATERIAL DENSTYtG/CC) THICKNESS I g 1 WATER 1.00000D 00 .1200D 00 F ' l l 2 IRON 3.78450D 00 3.0000D 00 F I 1 , 3 CONCRETE 2.300000 00 1.00000 02 F I l (, l l DISTANCE IN ! , DOSE POINT X-COORDINATE Y-COORDINATE LAST SHIELD j { .. ....... ............. ............. ............. (NO. 3) 'l 9 1 l }{ i
.I 1 6 0000E 00 F 6 2500E 00 F 1.5000D 00 F '
! ; ENERGY DOSE RATE 2 i GROUP BUILDUP (MR/HR) .
~~~---- - - - - - - --------- "
l C ] 1 1.744D 04 2.2410-07 - j 2 1.637D 02 1.449D.03' - l , C 3 6.2680 01 7.8220 03 4 3.1740 01 9.493D-01 j 5 2.223D 01 1.8260-01 l C 6 1 7420 01 9.755D-04 7 1.477D 01 2.7670-02 - < f 8 1.229D 01 1.4870-06 . J C ,
' TOTAL 1.170D 00
} 4 4., .
- , . .
- A ..; . . . .-
8 -
- DISTANCE IN , ,
! DOSE POINT N-Co0RDINATE Y-COORDINATE LAST SHIELD ,
~
(No. 3) ! 0 . 2 7 0000E 00 F- 6 2500E 00 7 2.5000D'00 F # i C - ENERGY DOSE RATE .
; GROUP BUILDUP (MR/HR)
Q ..... .< ... 5.3180 04' 1.1510-09 ]{
. 1 g
2 2.5570 02 2.110D-05 O 3 8.459D 01 2.253D-04 4 3.882D 01 4.1100 02 . i
- 5 2.6100 01 -1.0720-02
- 6 1.9980 01 7.0520-05
]' Q
" ~ -
7 1.6690 01 ^ 2.317D-03 ^- ^
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. _ , ~ . . - . . -. --
.... . ... . . ..s. . .-
I U 1.311D US ').46tu ut 7 . e TOTAL 5.4450 0? / (/. ,. . ' S l p , fJf. 7(7 iUf t H 77/TL H /i. '.'u: G (,,) ' DISTANCE IN /J . yy. 3j DOSE POINT X-CODRDINATE Y. COORDINATE LAST SHIELD (NO. 3) 3 8 0000E 00 F 6.2500E 00 F 3.50000 00 F j ENERGY DOSE RATE l GROUP BUILDUP (MR/HR) 1 1.444D 05 5.424D 12 ' 2 3.8370 02 3.0410 07 3 1.117D 02 6.5360 06 4 4.684n 01 1.8070 03 5 3.030D 01 6.4100-04 6 2.270D 01 5.194D.06 7 1.869D 01 1.979D-04 8 1.518D 01 1.4360 08 TOTAL ?.6570-03 e k O , e l v
( 7 YA l't: Y'/* M MMS M3 16079 MT o& N ' h/&c- 777A/J (X ~77) Q C~ .T//f0 /A $ ( CYL 0 RUN OF 04/10/75 AT 10.48.10
,J 0 h ff
- O' ENERGY EwERGY SRC STRENGTH Mwo-3-1 GROUPS (MEV) (MEV/CC*SEC) g 1 5.000000 01 7.280000 11 2 1.00000D 00 1.92000D 12
(- 3 1.50000D 00 8.67000D 10
- 4 2.000000 00 7.77060D 11 1 5 2.600000 00 2.570n0D 10 I
( 6 3.000000 00 4.220000 07 7 3.5n00nD 00 5.680600 n8 8 4.000000 00 1.470000 04 t SOURCE MTL DENSTY(G/CC) DIAMETER LENGTH ( WATER 3.750000 00 7 60000-01 F 1.2500D 01 F SHIELD MATERIAL DENSTY(G/CC) THICKNESS z 1 WATED 1.00000D 00 1.12000 00 F 2 IRON 3.78450D 00 3.0000D 00 F ( 3 CouCRETE 2.300000 00 1.00000 02 F
. DISTANCE IN
, DOSE POINT X-COORDINATE Y. COORDINATE LAST SHIELD (NO. 3) 1 6 0000F 00 F 1 2500F 01 F 1.50000 00 F ENERGY NOSE RATE
< GROUP RUILDUP (MR/HR) 1 1 744n 04 1.1210 07
. 2 1.6380 02 7.2480 04 3 6.272n 01 3.9130-03 !
4 3.176n 01 4.749D-01 5 2.224n 01 9.1330-02 6 1.743D 01 4.8800 04 , 7 1.478D 01 1 3840 02 l 8 1.229D 01 7.4390 07 l TOTAL, 5.8520 01 DISTANCE IN l DOSE POINT X-COORDINATE Y. COORDINATE LAST SHIELD (NO. 3) 2 7 0000E 00 F 1.2500F 01 F 2.50000 00 F ENERGY NOSE RATE , GROUP BUILDUP (MR/HR) O ( 1 2 5.31dn 04 2.558D 02 5.757D-10 1.0560 05 3 8.4640 01 1.1270 04 4 3.884D 01 2.0560 02 (' 5 6 2.611n 01 1.9990 01
'5.3640 03 3.5280 05
. . . J l'; d - .. ... ....: .. . .
1 l - 7 1.6.700 01 1.1600 03 ~~ ' ~ ~ l O . 6 1.3720 01 7.242D-08 2.724D.0
/[*.,,[r P./4.;c f//[// I TOTAL g foopy 49 , i i 1
, up. 7ty>J ex T-CG u/ gor!r Q l l
i fy- 4 80'3 * / 1 DISTANCE IN l DOSE POINT X CODRDINATE Y. COORDINATE LAST SHIELD (NO. 3) i i i 3 6 0000F 00 F 1 2500E 01 F 3.50000 00 F
- ENERGY DOSE RATE
, GROUP BUILDUP (MR/HR) 1 1.4440 05 2.7120 12 i 2 3.838D 02 1.5210-07 i 3 1.1170 02 3.2690-06 ! 4 4.6860 01 9.0380 04 ! , 5 3.0320 01 3.207D.04 , 6 2.2710 01 2.5990 06 i 7 1.8700 01 9.9050 05 i 8 1.520n 01 7.1910 09
- TOTAL 1.3300 03 1
l l I 4 !O
7, , . ; . . -n
. ,7 -_; - -
{ . *jl . l* .it. 4) f"W7f Sw(r$ 1/3 /00Q Mg'k W
' , {? [ivt L TK +JriR 7Dil f C///C t pen)&
( CYLSO RUN OF 04/08/75 AT 15.56.32 Mh J. ,, i'O= ENERGY ENERGY SRC STRENGTH
- S
{ GROUPS (MEV) (HEV/CC*SEC) ( __ ....... __.......... ............ L,p ygg,3,/ 1 5.00000D-01 7.28000D 11 2 1.00000D 00 1.92000D 12 (. 3 1.500000 00 8.67000D 10 i 4 2.00000D 00 7.77000D 11 5 2.60000D 00 2.57000D 10 ( 6 3.00000D 00 4.220000 07 l 7 3.500000 00 5.68000D 08 8 4.00000D 00 1.47000D 04 j SOURCE NTL DENSTY(G/CC) DIAMFTER LFNGTH . ( WATER 3.75000D 00 7.6000D-01 F 1.25000 01 F SHIELD MATERIAL DENSTYtG/CC) THICKNESS ( 1 WATER 1.00000D 00 1.12000 00 F 2 IRON 2.52300D 00 ' 3.00000 00 F [ 3 COFCRETE 2.300n0D 00 1.0000D 02 F DISTANCE IN ( DOSE POINT X-COORDINATE Y. COORDINATE LAST SHIELD (NO. 3) ("' ........__ ............ ....____.... .___........
\s 1 6.0000F 00 F 6 2500F 00 F 1.50000 00 F ENERGY DOSE RATE
( GROUP BUILDUP (MR/HR)
...... .....__ ...__.... l 1 3.897D 03 9.1670 04
( 2 9.442D 01 9.005D.01 3 4.1250 01 1.5580 00 4 2.243D 01 1.004D 02 ( 5 1.608D 01 1.279D 01 6 1.2790 01 5.186D.02 7 1.089D 01 1.236D 00 ( 8 9.123D 00 5.5990_05 TOTAL 1.170D 02 ( DISTANCE IN ( DOSE POINT X-COnRDINATE Y. COORDINATE LAST SHIELD (NO. 3) ( 2 7 0000F 00 F 6 2500F 00 F 2.5000D 00 F
, ENERGY NOSE RATE
( GPOUP PUILDUP (MR/HR) (~) __.... ....... 1.1950 04 4.668D.06
\) 1
(' 2 3 1.4940 02 5.7030 01 1.3070 02 4.5260-02 4 2.8270 01 4.4110 00 ( 5 1.9460 01 7.6270 01 6 - 1.5120 01 3.806D.03
r - , -- ' - r-- - m s o m 1--- ~ 1.m m -v1
,. 8 1.048D 01 5.5260-06 ,
[e W /cy- p f Jhff TOTAL 5.341D'OQ g g_ y ' Qf3 / dd 77 3/>b pp nmw vnc wein E g
'. DISTANCE IN fJ ~ YIO'3'!
DOSE PDINT x-COORDINATE Y. COORDINATE LAST SHIELD (NO. 3) g 3 8 0000E 00 F 642500E 00 F 3.,b0000 00 F ENERGY DOSE RATE
, GR0up BUILDUP (HR/HR)
- ' 1 3.2;6n 04 2.194D-08 - 2- 2.2620 02 1.878D-04 3 7.6560 01 1.3190-03 - 4 3.486D 01 1.958D-01 5 2.313D 01 4.6120-02 P.8360-04 6 1.758D 01 ! I 7 1.4530 01 9.0780-03 i 8 1.188D 01 5.5460-07 - 1 TOTAL P.5280-01 . me I
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- M##h CA'.C. Nh.~.Il8'T "2 SNEET ur U S CNATUTIM d DATE V// t'hf CHECi'ED 8 bM DATE #[8tI17 _
PGOJECT S U 21I JO B T40. W 79 SUBJECT Fe/ s ,,Se 7,ja
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, _ . _ . . . _ _ . .- . _ . - . _ , . . . . _ =
CYL80 EUN QF 04/14/79 AT 10.49.09 FUEL TRANSFER TUDE-LEA 8 SENSITY ENERGY ENER6Y SOURCE STREN6TH SHIELD NATERIAL GROUPS (NEV) DENSITY THICENESS (NEV/CC-SEC) NUMBER 1 0.500 (S/CC) 2 f.000 7.280008+11 't WATER 1.f20008+12 2 AIR 1.0008+00 1.120D+00 F 3 1.500 1.300D-03 3'000D+00 . F 8.470008+10 3 4 2.000 LEAD 0.0 1.0008+01 F 7.77000D+11 v , 5 2.600 2.570008+10 4 3.000 4.220008+07 7 3.500 5.660008+08 8 4.000 f.470008+04 SOURCE NATERIAL DENSITY (G/CC) WATER SIAMETER LENSTH 3.7508+00 7.600D-01 F 1.250D+01 F 30SE RATES IN TISSUE (NREN/NR) DOSE P81NT 1 X-C00R8 1.0508+01 F Y-C00R8 4.2508+00 F DIST. INTO d.0008+00 F SHIELD NO. 3. 9t0UP I 1.38838+07 2 5.11018+07 3 2.93128+04 4 2.72028+07 5 f.50598+05 . a 1.62778+03 . 7 2.22108+04 . ( 8 5.82918-01 TDTALS f.60928+07 e *
.- _. . . . . . . . . _ . . ... I
, l < l C/.LCULATION SH5Ei g *8M'/TL CALC.Nb.'64"I"2'Sl LEET OF-S!CNATL'T E d DATE Y/ MPf CHECKCD //N' N NCATE 2dd PROJECT I #52 f3 JOB NO. /##7/ f SUBJECT dr/ Inn b" b/c-i
$W&
4 3 i i 3, . . . . . - - . . -- .- - . _ . , .. - . . . . . _ CYLS0
; RUN OF 04/14/79 AT 10.49.45 1 FUEL TRANSFER TUBE-LEAD DENSITT t
i
; ENER6Y ENERGY SOURCE STREN6TH SHIELD MATERIAL SENSITY w GROUPS (MEV) THICKNESS
* (MEV/CC-SEC) NUMBER (6/CC) 1 0.500 7.28000D+11 I 1 WATER
- 1.0009+00^ 1.120D+00 F 2 1.000 1.92000D+12 d
- 2 AIR 1,100D-03 3.000D+00 F 3 1.500 8.67000B+10 ' 1-- t.EAD ------4.105ft--Ot --1.0008+01-F-t 4 2.000 7.77000D+11- '
5 2.600 2.57000D+10 4 3.000 4.22000D+07-7 3.500 5.680008+08 i 8 4.000 1.47000D+04 - SOURCE MATERIAL DENSITY (0/CC) IIAMETER & LENGTH - WATER 3.7505+00 6 7.4009-01 F 1.250D+01 F SOSE RATES IN TISSUE (NREM/NR) DOSE POINT 1
. X-C00RD 1.0508+01 F Y-COORD 6.250D+00 F DIST. INTO 6.0008+00 F BHIELD NO. 3.
SROUP 1 1.27278+01 2 . 1.12039+05 , 3 3.60703+04 i 4 6.2224B+05 5 2.79188+04 ( 6 7 5.62498+01 7.99098+02
, 8 2.14258-02 TOTALS 7.99138+05
-,7; , , , ,
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- l
- ~
CALCUl.ATION SHEET /fo[ 'kg d OF / CALC.1.' .NO-P2 l Ot.TE b Wl79
?d- StiEET[Y
] h CICtLYiUREhllH CHECKED _lf k CATE_ Y ( FROJECT_ A J 2*7 JCS TIO. Ae'07# I b SUBJECT be/ Meerb l ! 3 __ Nh ~ $e V 2- l 5 i ? i ~ i A l l
'CYLSO RUN OF 04/14/79 AT 10.50.,42 l
FUEL TRANSFER TOBE-LEAD SENSITT ! ) ENER6Y ENER6Y SOURCE STRENGTH SHIELD NATERIAL BENSITT THICKNESS GROUPS (MEV) (NEV/CC-SEC) NUMBER (6/CC) 1 0.500 7.28000D+11 1 WATER 1.000D+00 !.1209+00 F l AIR 1.300D-03 3.000D+00 F
- 2 1.000 1.92000D+12 2 l j 3 1.500 0.67000D+10 3 LEAD 8.000D-01 1.000D+01 F j 4 2.000 7.77000D+11 5 2.600 2.57000l+10 6 3.000 4.22000B+07
- 7 3.500 5.680000+0B i 8 4.000 1.47000l+04 SOURCE NATERIAL DENSITY (G/CC) SIAMETER LENGTH UATER 3.750D+00 7.600D-01 F 1.250D+01 F
- . SOSE RATES IN TISSUE (NREN/HR) l
. DOSE POINT 1 . M569+6t4-i -X-COORD l Y-COORD 4.2508+00 F DIST. INTO 6.000D+00 F j SHIELD NO. 3. BROUP
- 1 1.37438-03 -
2 1.f208D+03 l 3 1.18978+03 j } 4 4.87218+04 5 2.5740B+03 4 5.76743400 i , 7 8.4885B+01 8 2.36843-03. TOTALS 5.51968+04 i l
f CALCUL TION SHEET U
'Y OF '
bg CALC. Nh. ~~ #'I"2 SHEET SIGNATURE ,b IA DATE M#M/ f ~ CHECKED Mt/N DATE d23-/ 7 $ } PROJECT IBAV8I2fI JOB NO.# M l SUBJECT fu e/ Me affer l ^ 7~t &e. -4 /2.- l 4 __ f e !, CYLSO RUN OF 04/14/79 AT 10.51.54 a e FUEL TRANSFER TUBE-LEAD BENSITT l : l IH! ELD NATERIAL DENSITT THICENESS i ENER8Y ENER6Y SOURCE STRENGTN NUNBER (6/CC) l OROUPS (NEY) (MEV/CC-SEC) 1.0008+00 1.120D+00 F , I WATER 1 0.500 7.28000D+1I I.300D-03 3.000D+00 F , 2 AIR l 2 1.000 1.920000+12 1.1213+00 1.000D+01 F , 3 LEAD 3 1.500 8.670008+10
- 7 4 2.000 7.77000D+11 I
5 2.600 2.570000+10 l 4 3.000 4.22000D+07 : 4 7 3.500 5.68000D+08 . 8 4.000 1.47000D+04 ' SIAMETER LENGTH SOURCE MATERIAL SENSITY(G/CC) 1.250D+01 F 2 3.7508+00 7.4003-01 F WATER i DOSE RATES IN TISSUE (NREM/NR) DOSE POINT I X-COORD 1.050B+01 F l Y-COOR) 4.250D+00 F i 4.0003400 F DIST. INT 8 ' SHIELD N0. 3. St00P f 1.50859-07 j 3.2734B+01 2 l 3 f.70428+01
- 3.71828+03 4 . '
5 2.3105D+02 l 4 5.75909-91 . 7 8.80209'00 3( 8 2.54298-04 TOTALS 4.08840+03 N _ d
MgY3/h4 i ~^ d CAf.CU(ATION
~
CALC. NO. M'I~ % SHEETSHEET~ o S!CNATURE O DATE Vd M7f ' ~ 4 (8_ CATE Ebf~/7 / CHECKED PROJECT S0YS3 Ef3 _JO3 No /90 7f f l SUBJECT-d'I MensfV I _ Nbe- AeV2-4 CYLSO RUN OF 04/14/79 AT 10.52.44 FUEL TRANSFER TUBE-LEA 9:9ENSITY DENSITY THICKNESS ; SHIELD NATERIAL ENERBY ENERGY SOURCE STREN6TN O (S/CC) (NEV/CC-SEC) NUNDER GROUPS (NEV) 1.000p+00 1.120D+00 F 7.280003411 1 WATER 1 0.500 1.92000B+12 2 AIR I.300D-03 3.000D+00 F 2 1.000 1.442D+00 1.0003+01 F . 0.47000B+10 3 LEAD 3 1.500 - 4 2.000 7.77000D+11
~
5 2.600 2.57000D+10 6 3.000 4.22000D+07 7 3.500 5.68000D+08 8 4.000 1.47000D+04 DIANETER LEN8TN SoutCE NATERIAL DENSITY (6/CC) 3 750D+00 7.600D-01 F 1.2508+01 F WATER DOSE RATES IN TISSUE (NREN/NR) 1 DOSE POINT X-COORD 1.0508+01 F Y-CQQRD 4.2500+00 F DIST. INTO 6.0000+00 F SHIELD NO. 3. St0UP
't 1.72389-11 2 5.65939-01 3 4.9825B+00 4 2.81419+02 -
5 2.05498401 6 5.69058-02 7 f.05058-01 8 2.71428-0:i TOTALS 3.08478+02
~
- t 9f rem e4 e
CALCULATION SHEET #3 ' Vfd-5-2 SHEET OF N CALC. N
~ SIGNATURE ,NfM DATE YM V/7f CHECKEDNAl#M DATE NSP/ 7 f PROJECT Sf#1330 JOB No, M79 SUBJECT e/ Me##M 5}e -$cy2
*^
~~ ~
CYLSO SUN OF 04/14/79 AT 10.53.41 FUEL TRANSFER TUIE-LEAD' 8ENSITY
. 3 SHIELD NATERIAL DENSITY THICKNES$ $
ENERGY ENERGY SOURCE STRENGTH (6/CC) O GROUPS (NEV) (MEV/CC-SEC) NUNDER 1 0.500 7.200003+11 1 UATER 1.0008+00 1.120D*00 F 2 1.000 1.92000D+12 2 AIR 1.300D-03 3.000D+00 F 3 1.500 8.67000D+10 3 LEAD 1.400D+00 1.0008+01 F 4 2.000 --1-17000Nt- -- 5 2.400 2.57000l+10 i* a 3.000 4.220008407 7 3.500 5.68000B+0B 8 4.000 1.47000l+04 SOBRCE NATERIAL DENSITY (S/CC) IIAMETER LENGTH UATER 3.750D+00 7.600B-01 F 1.250D+0! F DOSE RATES IN TI5 SUE (NREN/Ht) BOSE POINT 1 X-COORD 1.050D+01 F Y-COORD 4.250D+00 F DIST. INTO d.0009+00 F SHIELD NO. 3. SROUP 1 2.00339-13 2 7.71988-02 3 1.15600+00 4 7.8943B+01 .
- l. 5 6.23099+00 6 1.81478-02 O TOTALS 7
8 2.94789-91 f.01028-04 8.67208+01
. .e .M
CALCULATION SHfET W [96 OF / CALC. NO.VG-J SHEET SIGNATURE d DATE YdV/79
# /2 ## DATE NM'/ ?S CHECKED PROJECT MNI FJ J03 NO. 4 0 7f SUBJECT be! b##N-7Gls - fer 2-
~
CYLSD RUN OF 04/14/79 AT 10.54.29 FUEL TRANSFER TUBE-LEAD BENSITY SHIELD MATERIAL *)ENSITT THICKNESS ENER6Y ENERGY SOURCE:STRENSTN (NEV/CC-SEC) NUMBER (6/CC) GROUPS (MEV) 1.000D+00 1.120D+00 F O V I 2 0.500 f.000 7.28000D+11 1.920003+12 1 2 UATER AIR 1.3000-03 3.000D+00 F 1.762D+00 1.000D+01 F 3 1.500 8.67000D+10 3 LEAD ~ 4 2.000 7.77000D+11 - 5 2.600 2.57000B+10
- 6 3.000 4.220008+07 .
7 3.500 5.680008+08 - 8 4.000 1.47000B+04 BIAMETER LENGTH SOURCE MATERIAL DENSITY (G/CC) 3.750D+00 7.6009-01 F t.250D+01 F UATER DOBE RATES IN TISSUE (MREN/HR) DOSE POINT 1 X-C00RD 1.050E+01 F Y-COORD - 6.2509+00 F DIST. INTO 6.000D+00 F SHIELD NO. 3. GROUP f # 2.09658-15 2 t.00465-02 3 2.58689-01 4 2.f4f30401 5 1.8312B+00 6 5.62769-03 7 f.32285-02 8 2.90669-06 TOTALS 2.36129+01
-~ . . . . . . _
i l CALCULATION SHEET ff CALC, NIN#-5-2 SMEET ' / OF NQ O ~ j S!GNATUR{N dVbfb .-- DATEN/M7 A l CHECKED 8 4 /#8DATE ' f JOB NOAR979 PROJECT!Y S Lt ! l T&oJ'fv I h SUBJECTb*l
~I.,4 ) e - AWL _
i l ] l CYLSD RUN OF 04/14/79 AT 10.55.25 ' l \ FUEL TRANSFER TUBE-LEAD BENSITY 8 i j .
- i i ENERGY ENERGY SOURCE STRENGTH SHIELD NATERIAL DENSITY
' GROUPS (MEV) THICKNESS i (MEV/CC-SEC) NUMBER (6/CC) I 0.500 7.20000D+1I I ! WATER 1.0008+00 1.1208+00 F 2 1.000 1.920008+12 2 AIR ' 3 1.500 1.3009-03 3.000D+00 F : 8.67000B+10 3 LEAD i 4 2.000 1.9208+00 1.000D+01 F 7.770008+11 l i 5 2.600 2.570008+10 4 3.000 4.22000D+07 ! 7 3.500 5.48000D+08 , ! 8 4.000 ' 1.47000D+04 1 4 l SOURCE MATERIAL DENSITY (0/CC) 8IAMETER LENGTH WATER 3.750D+00
- 7.600D-01 F 1.250D+0! F SOSE RATES IN TISSUE (MREM /NR)
- DOSE POINT 1 X-C00R8 1.0508+0! F
) Y-C00R8 4.2508+00 F i DIST. INTO 4.0008+00 F SHIELD NO. 3. GROUP 1 2.47398-17
- 2 1.37f38-03 I 3 6.01138-02 4 5.9955B+00 5 5.54238-01 2
6 f.79278-03
- i
. 7 3.030f8-02 '
8 f.63678-07
! i TOTALS 6.64338+00 4 .. .-
4
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- PROJECT MNI2 #I JOB NO. O -
- SUBJECT Nd b *5 # # - e "' 2- SHEET / OF SHEET t
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. 23 O" 35 as
EC&FS DEPARTMENT ICCN NO. I CALCULATION SHEET CCN COWERS 10ft CCN NO. CCN . l roject or DCP/FCN N/A Cale No. N-04AO-003
'ubiect Puel Transfer Tine Shicidir e . _._ ._. - _. _ _ _
_ _ _ _ _ _ _ __ _ _ .__ Sheet No 67_ of 117__ _ REV ORIGillATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER D5-31-% C. W. SAYLES D6-03-% y I REVISION 4 TABLE OF CONTENTS SECTION PAGE REVISION 4 TABLE OF CONTENTS 67 l I. PURPOSE . . . 68 I.1 Objective . . . . . 68 I.2 Acceptance Criteria .68 II.
SUMMARY
OF RESULTS . 69 II.1 Analysis Results . 69 II.2 Recommendations . . . 69 III. ASSUMPTIONS . . 71 I I IV. DESIGN INPUT DATA . . 77 V. hETHODOLOGY . . 80 V.1 Fuel Assembly Source Strength Spectrum . 82 V.2 Fuel Assembly Composite Density . . 84 V.3 Dose Rate in Penetration Bldg El.15'0" Corridor Below Fuel Transfer Tube 86 VI. REFERENCES. . 89 VII. NOhENCLATURE 93 VIII. COMPUTATIONS . . 94 VIII.1 Fuel Assembly Source Strength Spectrum . . 94 VIII.2 Fuel Assembly Composite Density . 97 VIII.3 Dose Rate in Penetration Bldg El.15'0" Corridor Below Fuel Transfer Tube .99 IX. COMPUTER CODE INPUT AND OUTPUT FILES . 105 IX.1 SOURCE 2 Code Run . 105 IX.2 SHIELD-SG Code Run 112 SCZ 26 426 REV.0 E94 [ REFERENCE 80121 XXIV 715]
1 EC&FS DEPARTMENT ICCN NO.
"""""~ ~ "-
CALCULATION SHEET
- CCN CONVERSION.
CCN NO. CCN . 1 Project or DCP/FCN N/A Calc No. N 0.iAG 003 l l hiect Fuel TransferTuw Shielding . __ _ _ _ Sheet No _68 of 117_ REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 NARK DRUCKER 05-31 % C. W. SAYLES 06-03-% h l 1 I. PURPOSE i I.1 Objective The objective of Revision 4 to this calculation is to determine the dose rate in the Penetration Building Elevation 15'0" corridor below the fuel transfer tube using as-built dimensions and the , SONGS specific source term based on an assumed discharge burnup level of 60 GWD/ ton. This l calculation will conservatively consider the presence of two spent fuel assemblies, side-by-side, in the fuel transfer tube 72 hours after reactor shutdown. This calculation revision is originated in support of the resolution of Action Request 960300680 and RCTS Commitment #9605010 i Action #4. ' This analysis differs from previous revisions to this calculation in that the previous revisions present parametric evaluations ofproposed concrete shielding modifications to the 1973 design l using a generic source term and the presence of a single fuel assembly in the fuel transfer tube. l I This analysis differs from Calculation N-0480-016 in that Calculation N-0480-016 presents an evaluation of proposed concrete and steel shielding modifications to the 1973 design using a generic source term and the presence of a single fuel assembly in the fuel transfer tube. Due to the historical nature of Revisions 0 through 3, this analysis will not delete portions of the previous revisions. 1 I.2 Acceptance Criteria Movement ofirradiated fuel assemblies in the transfer tube would only occur during the shutdown mode of operation. Radiation Zone Drawings 40023 and 40027 (Reference 2.1) dictate that during shutdown mode the Penetration Building Elevation 15'0" corridor below the fuel transfer tube should not exceed a Zone IV criterion, corresponding to a maximum dose rate of 100 millirem / hour. 1
- - . .,an-- ., m
EC&FS DEPARTMENT ICCNul CALCULATION SHEET CCN CONVERSION-CCN NO. CCN . Project or DCP/FCN N/A Calc No. N. MAO 001 bubiect Fue.I Transfer Tine Rhieldig a _. _ _ _ . . _ _ . _ _ _ ._ _ _ _ _ _ - _ _ _ . _
.. _ Sheet No 39 of_117_
REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05 % C. W. SAYLES 06 % y I II.
SUMMARY
OF RESULTS 11.1 Analysis Results Section VIII.3.3 documents dose rates in and around the fuel transfer tube using the SONGS specific source term based on an assumed discharge burnup level of 60 GWD/ ton. These dose rates are depicted in Figure II-l for various concrete shielding attenuation thicknesses. Per Section VIII.3.3, the current fuel transfer tube and Penetration Building designs with 4'10" of concrete shielding are sufficient to maintain a maximum dose rate of 4.03 millirem / hour in the Penetration Building Elevation 15'0" corridor below the fuel transfer tube should one fuel assembly be present in the tube 72 hours after reactor shutdown. Should two fuel assemblies be present in the tube 72 hours after reactor shutdown, then the current fuel transfer tube and Penetration Building designs are sufficient to maintain a maximum dose rate of 8.06 millirem / hour in the Penetration Building Elevation 15'0" corridor below the fuel transfer tube. Each of these j dose rates is less than the 100 millirem / hour Radiation Zone IV criterion (per Section I.2), thereby indicating the acceptability of the as-built fuel transfer tube and Penetration Building design. II.2 Recommendations This analysis differs from Calculation N-0480-016 in that Calculation N-0480-016 presents an evaluation of proposed concrete and steel shielding modifications to the 1973 design using a generic source term and the presence of a single fuel assembly in the fuel transfer tu'ae. Since no steel was added to the design, it is recommended that Calculation N-0480-016 be voided. 1 1 ( l i SCE 26 426 REV 0 &94 [R.EPTJt.ENCE 301:3-XXIV.7.15]
EC&FS DEPARTMENT ICCN O.
""""""' PAGE _ 0F _
CALCULATION SHEET CCN COWERSION-CCN No. CCN - Project or DCP/FCN N/A Cale No. N G.1RS001 ubiect Fuel Transfer Tu w Rhieldir n . _ _ _ . - _ _ _ _ _ _ _ _ . . _ _ . _ . _ . . . Sheet No ._70. of_117 REV ORIGillATOR DATE 1RE DATE REV ORIGINATOR DATE IRE DATE R 4 E MARK DRUCKER 05 % C. W. SAYLES 06-03 % y i Figure 11-1 DOSE RATES FOR VARIABLE FUEL TRANSFER TUBE CONCRETE SHIELDING THICKNESSES 1E6 l ! I l l l l l l l l 1E4- 4 j y [ -j y
. c ,
. l l l l l 8 l i ! l l l e 1E2-3--
1 ; y ; E i > ' ' i e
~
! l '.
l l l
- 1EO- I l -
L I I
= ! l 1 l l l I e i
- i I l
, 1 E l 4 -i F- -'
i 8 l l l l . . . l l l l l l l L 1 E -j j -j ? -j , l l l l l l l 8 l l 1 1 1 E-6 l l l l l l 0 1 2 3 4 5 6 7 Concrete Shield Thickness (feet)
- 1 Assemblyin Tube -e- 2 Assemblies in Tube 4'10" Thickness - 100 mrem / hour criterion I I sCE 26426 REV 0 EH [ RIFF.RENCE sol 23-XXIV 715)
EC&FS DEI ? RTMENT ICCN NO. CALCULATION SHEET CCN CONVERS!ON. CCN NO. CCN . Project or DCP/FCN N/A Calc No. N-MAO 001 abiect Fuel Transfer Tu w Shieldir e - _ Sheet No _.71 of 117. REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05-31-96 C. W. SAYLES 06-03-96 I IU. ASSUMPTIONS
- 1. Configuration Symmetry Between Unitt 7. and 3 This analysis and its conclusions are assumed to be applicable to both Units 2 and 3. Where possible, references are provided to show that assumptions and design input data are representative of both units.
At the time that this calculation revision is being originated, the latest fuel specification data is provided in documents associated with the Unit 2 Batch L Manufacturing Order (Reference 2.2). The fuel, fuel tube and fuel assembly dimensions for this Unit 2 Batch L serve as the basis for several of this calculation's assumptions and design input data. It is assumed that this data is also representative of Unit 3 fuel batches and previous Unit 2 fuel batches. This assumption is considered reasonable based on a review ofpast Fuel Batch Manufacturing Orders. l I
- 2. Maximum Fuel Burnup This analysis models a reactor core activity profile based on an assumed discharge burnup level of 60 GWD/t. A discharge burnup level of 60 GWD/t represents the maximum anticipated fuel burnup for the peak assembly in the most recently evaluated fuel cycle, as documented in SONGS-2 Cycle 9 design analysis A-SG2-FE-0058 (Reference 1.1).
- 3. Reactor Shutdown Time This analysis assumes that the reactor has been suberitical for at least 72 hours prior to movement ofirradiated fuel in the reactor pressure vessel. This shutdown time is consistent with the fuel movement requirements of Units 2 and 3 Technical Specification LCO 3.9.3 (References 4.1 and 4.2).
- 4. Daughter Product Isotopes
) This analysis considers the dose contributions associated with the reactor core daughter product isotopes formed by 72 hours of decay following reactor shutdown. l l 3CE 2H26 REV C E94 lREPT.KENCE SO123-XXIV-7.15)
EC&FS DEPARTMENT ICCN NO.
"""" NG OL CALCULATION SHEET CCN CONVERS10N:
CCN No. CCN . Project or DCP/FCN N/A Cale No. N nan _not ibiect Tuel Transfer Tuhe Shieltlir a _ _ _ _ _ . _ . _ _ __ __ _ _.
.. _. Sheet No .72. of_117_.
REV ORIGINATOR DATE IRE DATE I REV ORIGINATOR DATE IRE DATE R 6 MARK DRUCKER 0.' % c. W. SAYLES 06-03 96 y I
- 5. Fuel Assembly Radial Peaking Factor The highest dose rate in the Penetration Building Elevation 15'0" corridor below the fuel transfer tube will occur when the fuel assembly with the greatest concentration ofisotopes (typically from the center of the reactor core) is being transferred. The activity profile of this peak fuel assembly can be estimated by scaling the activity profile of the core's average fuel assembly by a radial
- peaking factor. The radial peaking factor for a fuel assembly may be approximated by a conservative estimate of assembly power peaking (Relative Power Density [RPD]).
Per Assumption 2, this analysis models a reactor cure activity profile based on an assumed , discharge burnup level of 60 GWD/t. Past SONGS cycle experience, as documented in the SONGS 2 and SONGS 3 Cycles 6 through 8 Paysics Databooks (References 2.3 and 2.4, Figures 6.7 through 6.12), indicates that RPIJs for high burnup assemblies (once, twice burned) throughout a cycle are all below 1.15. Therefore, it is conservative to assume an assembly radial peaking factor of 1.20 when modeling activity profiles associated with high burnup fuel. l I
- 6. Core Axial Peaking Factor Due to the neutron flux axial shape in the core, the greatest concentration ofisotopes in a fuel i
assembly are at the midheight of the active fuel length. Therefore, the highest dose rate in the Penetration Building Elevation 15'0" corridor below the fuel transfer tube will occur when the center of the fuel assembly being transferred through the tube is directly above the corridor. The activity profile at at the midheight of the active fuel length of a fuel assembly can be estimated by scaling the activity profile of the core's average fuel assembly by the core axial peaking factor. Past SONGS cycle experience, as documented in the SONGS 2 and SONGS 3 Cycles 6 through 8 Physics Databooks (References 2.3 and 2.4, Tables 6.1 through 6.6), indicates that core axial peaking factors throughout a cycle are all below 1.20. Therefore, it is conservative to assume a core axial peaking factor of 1.25. SCE 20426 REY 0 E94 (REPT.RENCE 80121 XXIV 7.15]
EC&FS DEPARTMENT ICCN NO. CALCULATION SHEET CCN CONVERSION. CCN NO. CCN - Project or DCP/FCN N/A Calc No. N 04AO 001
;ubiect Fuel Transfer Tu mShieldir e ~
_ _ . _ _ _ Sheet No . 73. of_117_ REV l ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R , 6 MARK DRUCKER 05 % C. W. SAYLES 06 % y i i ! s i
- 7. Fuel Pellet (Uranium Dioxide) Density ;
i At the time that this calculation revision is being originated, the latest fuel specification data is i provided in documents associated with the Unit 2 Batch L hianufacturing Order (Reference 2.2). Per Assumption I this Unit 2 data is assumed applicable to both Units 2 and 3. Per the Unit 2 Batch L Bill of hiaterials Document S2L-FhiDE-BOh11 (Reference 2.5) and Specification Document SO23-990-10 (Reference 2.6, Section 3.2.2), the density of the sintered uranium dioxide fuel pellets snall be between 94.0 percent and 96.5 percent of theoretical density based on a UO2theoretical density of 10.96 gram /cc. To maximize dose rates it is conservative for this caleclation to minimize self-attenuation of the l
- gamma radiation emitted by the fuel. Therefore, it is conservative to model the minimum fuel pellet density. Based on the preceding data, the fuel pellets in use at SONGS have a minimum 4
density of10.30 gram /cc: q psi = (0.940) x (10.96 gram /cc) = 10.30 gram /cc
- 8. Fuel Rod Cladding Density i
At the time that this calculation revision is being originated, the latest fuel specification data is i provided in documents associated with the Unit 2 Batch L hianufacturing Order (Reference 2.2). Per Assumption 1 this Unit 2 data is assumed applicable to both Units 2 and 3. Per the Unit 2 Batch L Bill of hiaterials Document S2L-FMDE-BOhil (Reference 2.5) and Specification Document S023-990-13 (Reference 2.7, Section 1.1.1), the fuel tube (rod) cladding material is Zircaloy-4. Per Nuclear Fuel Management (Reference 6.1, sheet 193), Zircaloy-4 has a density of 6.44 gram /cc. see 234a arv o am larrrazuce som-xxrva151
EC&FS DEPARTMENT
!Ena"o- =-
CALCULATION SHEET CCN CONVERSION: CCN No. CCN - Project or DCP/FCN N/A Calc No. N-04AO 001
,ubiect Fuel Transfer Tine Shieldir 2 - _ _ _ _ - ___ . - _ _ _ _ _ _ . _ _ _ Sheet No .. _74 . of_117_
REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE !RE DATE R 6 MARK DRUCKER 05-31-% C. W. SAYLES 06 % y I
- 9. Helium Fill Gas (Gap) Density At the time that this calculation revision is being originated, the latest fuel specification data is provided in documents associated with the Unit 2 Batch L Manufacturing Order (Reference 2.2).
Per Assumption 1 this Unit 2 data is assumed applicable to both Units 2 and 3. Per the Unit 2 l Batch L Bill of Materials Document S2L-FMDE-BOMI (Reference 2.5), Fuel Rod Assembly Drawings SO23-990-165 and SO23-990-167 (References 2.8 and 2.9), and Specification Document SO23-990-8 (Reference 2.10, Section 4.6.5), Unit 2 Batch L utilizes pure Helium as the fill gas in the fuel rod assembly gap space, with an initial fill gas pressure of 380 psig at 68 degrees Fahrenheit. Per Assumption 3, this analysis assumes that the reactor has been suberitical for at least 72 hours prior to fuel movement. It is therefore considered reasonable to assume that the fuel assemblies will have cooled down from normal reactor operating temperatures to no more than approximately 300 degrees Fahrenheit. It is also assumed that the fill gas behaves as an ideal gas { > as characterized by the ideal gas law (PV=nRT). As such, at 300 degrees Fahrenheit it is expected that the Helium fill gas would exist at a pressure of approximately 568 psia: nR/V = P/T @ initial conditions = P/T @ some other conditions
. Pu e = (Pm / Tm) x Tue,
- where:
Pm = 380 psig + 14.7 = 394.7 psia Tm = 68 Fahrenheit + 460 = 528 Rankin Tu,,. = 300 Fahrenheit + 460 = 760 Rankm
. Pn,.. = [(394.7 psia) / (528 Rankin)] x (760 Rankin) = 568 psia Per the Handbook of Physical Properties of Liquids and Gases (Reference 6.2, page 527), the density of helium at 300 degrees Fahrenheit and 568 psia (approximately 150 C and 40 bar) is 4.5 kg/m'. Applying volumetric conversion factors yields an equivalent fuel gap space density of 4.5e-3 gram /cc:
S py,, = (4.5 kg/m ) x (1e3 gram /kg) x (1e-6 m'/cm') = 4.5e-3 gram /cc The accuracy of this density (and its underlying temperature assumption) is not of paramount importance because this density will only be used in estimating gamma radiation self-attenuation i ' by the helium in the fuel gap space, and the attenuation provided by this helium should be minimal in comparison to the attenuation provided by the other intervening media (e.g., concrete). SCE M26 REV 0 &H [ REFERENCE SOW-XXIV.715)
EC&FS DEPARTMENT ICCN N& CALCULATION SHEET CCN CONVERSION. CCN NO. CCN - Project or DCP/FCN N/A Calc No. N-0.1RG 003 Subiect hel Transfer Tuie Shieldiro _ _ _ _ _ Sheet No 75. of.117 REV ORIGINATOR DATE IRE DATE l REV ORIGINATOR DATE -IRE DATE R 4 MARK DRUCKER 05 % c. W. sAirLES 06-03-96 h
! l l
l
- 10. Density of Fuel Transfer Tube Water The water in the fuel transfer tube is assumed to be saturated liquid at 10 psig (approximately 25 psia). The assumed saturated nature of the water is based on the maximum allowable water ,
temperature without the water boiling. The assumed pressure is based on the belief that the fuel l transfer tube water is at least 23 feet below the surface of the cpent fuel pool and application of a i 27.7 inch of water per psi conversion factor. The 23 foot submersion depth is consistent with the I minimum water depth that must cover spent fuel per Technical Specification Limiting Condition for Operation 3.9.11 (References 4.1 and 4.2). Per the ash 8 Steam Tables (Reference 6.3, l Table 1 page 192), the specific volume of saturated liquid water at 25 psia is 0.017009 ft'/lbm. ' Taking the inverse of the specific volume and applying volumetric and mass conversion factors l yields an equivalent density of 0.94 gram /cc: 1 p = (1/ u) = 1/ [(0.017009 ft'/lbm) x (28317 cm'/ft') x (0.0022046 lbm/ gram)) { j p = 0.94 gram /cc The accuracy of this density (and its underlying temperature and pressure assumption) is not of paramount importance because this density will only be used in estimating gamma radiation attenuation by the water in the fuel transfer tube, and the attenuation provided by this water should be minimal in comparison to the attenuation provided by the other intervening media (e.g., concrete).
- 11. Density of Air between Fuel Transfer Tube and Concrete Shielding The air present in the open space between the fuel transfer tube and its concrete shielding is assumed to be at a temperature of 120 degrees Fahrenheit. This assumed temperature is reasonable based on the large quantity ~of nearby concrete that acts as a heat sink. This concrete should maintain an air temperature in this space that is cool relative to the water and fuel temperatures in the fuel transfer tube. Per the Handbook of Physical Properties of Liquids and Gases (Reference 6.2, page 588), the specific volume of air at 120 degrees Fahrenheit and 14.7 psia (322 K and 1 bar) is 924.3 dm'/kg. Taking the inverse of the specific volume and applying volumetric and mass conversion factors yields an equivalent density of 1.le-3 gram /cc:
p = (1/ u) = 1/ [(924.3 dm'/kg) x (1e3 cm 3/dm') x (1e-3 kg/ gram)] l p = 1.le-3 gram /cc l I
- -. - - ~ ~ ~ .w,
\
l
I EC&FS DEPAllTMENT CCN NO. CALCULATION SHEET CCN CONVERSION-CCN NO. CCN - _ Project or DCP/FCN N/A Calc No. N-0.tAS-003 i .ubject Fuel Transfer Tu w Shieldir 2 - _ _ _ _ . _ _ _ _ _ _ _ _ . - __ _ _ Sheet No _.76_ of 117_... l REV ORIGINATOR DATE IRE DATE l REV ORIGINATOR DATE IRE DATE R l4 MARK DRUCKER 05 % C. W. SAYLES 06 % h The accuracy of this density (and its underlying temperature assumption) is not of paramount importance because this density will only be used in estimating gamma radiation attenuation by the air space, and the attenuation provided by this air should be minimal in comparison to the attenuation provided by the other intervening media (e.g., concrete).
- 12. Density of Concrete Shielding The dose rates in the areas surrounding the fuel transfer tube are mitigated by radiation shielding provided by the concrete enclosure surrounding the tube. These concrete barriers are assumed to be constructed of ordinary concrete. Per Section 4 of ANSI /ANS-6.4-1977 (Reference 6.4),
ordinary concrete is " Type 04" concrete, which has a theoretical density of 2.35 gram /cc 4 (148 lbm/ft'). l k I l > scE 26 426 arv o t$4 ptertrJNCE SOID XMV.715j
EC&FS DEPARTMENT secN NO. CALCULATION SHEET CON COWERSION: CCN NO. CCN - ( Project or DCP/FCN N/A Calc No. N-AdAn-003
?ubiect F'inel Trancier Inn x Rhieldir o .. . . _ _ . . . _ - - __ . _ _ _ _ _ _ _
__ _ _ _ _ . _ _ Sheet No _JL of _11L_ REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 1%RK DRUCKER 05 % C. W. SAYLES 06 % h I i IV. DESIGN INPUT DATA 1
- 1. Reactor Core Isotope Inventory Per Assumption 2, it is anticipated that burnup levels on the order of 60 GWD/t will occur with future operating cycles. The core inventory for the core noble gas, iodine and particulate isotopes at 60 GWD/ ton and at 105 percent or full power core conditions (i.e.,3560 Mwt)is as shown in Table IV-1. This core inventory is per Calculation N-4097-013 (Reference 1.2, Table 8.1-2).
This core inventory does not address tbe 72 hour decay prior to fuel movement (per Assumption 3). Pe alculation N-4097-013, the core inventory in Table IV-1 was determined by scaling the M ,WD/ ton core inventory at 105 percent offull power core conditions (i.e.,3560 Mwt) by burnup change factors. The burnup change factors are isotope specific scaling factors that allow i an estimation of the impact of an increase in the discharge burnup level from 33 GWD/t to l q , 60 GWD/t. l Table IV-1 l CORE INVENTORY AT 60 GWD/T BURNUP Isotope Activity Isotope l Activity Isotope Activity Isotope Activity (curies) : (curies) (curies) (curies) Br-84 i 2.85e+07 Y-91 i 3.58e+08 I-132 , 1.32e+08 Cs-137 7.89e+06 Br-85 3.99e+07 Y-95 i 1.87e+08 Te-133m 1.07e+08 Xe-138 ! 1.79e+08 Kr-85m
- 3.98e+07 Zr-95 , 1.78e+08 Te-133 . 1.12e+08 Cs-138 ; 2.04e+08 Kr-85 ! 1.75e+06 Nb-95 l 1.81e406 I-133 l 2.02e+08 Cs-140 i 1.81e+08 Kr-87 4 6.47e+07 Mo-99 ; 1.89e+08 Xe-133 j 1.93e+08 La-140 : 1.93e+08 Kr-88 4 9.48e+07 Tc-99m . 2.27e+07 Cs-134 l 2.58e+06 Ba-143 1.60e+08 Rb-88 1.10e+08 Ru-103 . 1.0le+08 Te-134 l 2.12e+08 La-143 ; 1.81e+08 Kr-89 ! 1.41e+08 Ru-106 i 1.15e+07 I-134 2.39e+08 Ce-143 1.81e408 Rb-89 . 1.46e+08 Te-129m 1.07e+07 I-135 l 1.85e+08 Pr-143 i 1.81e+08 Sr-89 : 1.26e+08 Te-129 3.29e+07 Xe-135m -
5.51e+07 Ce-144 1.39e+08 Sr-90 4 1.67e+07 I-129 4.16e+00 Xe-135 r 5.08e+07 Pr-144 l 1.26e+08 Y-90 ; 9.22e+06 I-131 i 8.96e+07 Cs-135 j 2.30e+01 >
- ,f'/?' J' ,';'jii;'l'/
Sr-91 1.78e+08 Xe-131m 6.16e+05 Cs-136 + 1.78e+05 ,
,r.j , ,j Dj;@ff
't
; 1.80e+08 A ': e h';lh W
^
Y-91m - 1.05e+08 Te-132 1.32e+08 Xe-137 . sCE 3 4N REY. O sw (RIFERENCE 301D.XUV4 !$)
EC&FS DEPARTMENT ICCN NO.
" " " " " " - ~ " - -
CALCULATION SHEET CCN CONVERSION-Project or DCP/FCN N/A Calc No. N 04AG.-003
.ubiect Fuel Transfer Tu m Shi+1dir a . _ _ _ _ _ _ _ _ _ - - - _ . _ _ _ _ _ . _ . . _ _ _ _ _ _ _ _ _ _
Sheet No. _.78 of_117_ _ REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05-31-% C. W. SAYLES 06-03-96 y 1
- 2. Number of Fuel Assemblies in the Core This calculation will model 217 fuel assemblies in the reactor coi e. This value is the maximum number of fuel assemblies per core allowed by Units 2 and 3 Techical Specification Design Feature 5.3.1 (References 4.1 and 4.2).
- 3. Number of Fuel Rods in a Fuel Assembly This calculation will model 236 fuel rods in an assembly. This value is the maximum number of fuel rods per assembly allowed by Technical Specifications Design Feature 5.3.1 (References 4.1 and 4.2).
- 4. Fuel Pellet and Fuel Rod Dimensions I
At the time that this calculation revision is being originated, the latest fuel specification data is provided in documents associated with the Unit 2 Batch L Manufacturing Order (Reference 2.2). Per Assumption 1 this Unit 2 data is assumed applicable to both Units 2 and 3. Per the Unit 2 Batch L Bill of Materials Document S2L-FMDE-BOM1 (Reference 2.5) typical filel pellets for this fuel batch are depicted in Drawings SO23-990-166 (Reference 2.11) and SO23-990-168 (Reference 2.12). Per these drawings a typical Unit 2 Batch L fuel pellet has a nominal outer diameter of 0.3255 inches. Per the Unit 2 Batch L Bill of Materials Document S2L-FMDE-BOM1 a typical fuel rod for this fuel batch is depicted in Drawings SO23-990-165 (Reference 2.8), SO23-990-167 ; (Reference 2.9), and SO23-990-198 (Reference 2.13). Per Drawings SO23-990-165 and ' SO23-990-198, a typical Unit 2 Batch L fuel tube has a nominal outer diameter of 0.382 inches I and a nominalinner diameter of 0.3320 inches. Per Drawings SO23-990-165 and SO23-990-167, ' a typical Unit 2 Batch L fuel rod has a nominal active fuel length of 150 inches. j i l l l
$CE 26 06 REV 0 SH [RMRENCE 301&IXrV415]
EC&FS DEPARTMENT ICCN NO.
**""~ " " - " -
CALCULATION SHEET CCN CONVERSION-
. CCN NO. CCN -
goject or DCP/FCN N/A Calc No. N-04AG 001 k,_ ubject FuelTransfer. Tube Shieldir z __ _ _ _ . ___. _ _ ___ . _ _ _ _ _ Sheet No ._79-_ of_11L , REV ORIGINATOR DATE 1RE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05-31-96 c. W. SATLES 06 % y I l
- 5. Fuel Assembly Dimensions At the time that this calculation revision is being originated, the latest fuel specification data is provided in documents associated with the Unit 2 Batch L Manufacturing Order (Reference 2.2).
Per the Unit 2 Batch L Bill of Materials Document S2L-FMDE-BOM1 (Reference 2.5) a typical fuel bundle (i.e., fuel assembly) for this fuel batch is depicted in Drawing SO23-990-164 Revision 0 (Reference 2.14). View A-A of this drawing depicts a cross-section of the fuel assembly. Per View A-A, each fuel assembly has an overall square appearance with 8.030 inches separating the outer edge of a fuel rod on one side from the outer edge of a fuel rod on the opposite side.
- 6. Fuel Transfer Tube Dimensions Per Drawing SO23-207-4-5 (Reference 2.15), the fuel transfer tube is constructed of a 0.25 inch
{ } thick stainless steel 304 pipe, with an outer diameter of 3 feet. Per Drawings 23119 (Reference 2.16),25407 (Reference 2.17, Detail 1) and 25412 (Reference 2.18), the fuel transfer tube is installed with its centerline at plant elevation 28'6"
- 7. Concrete Shielding Below the Fuel Transfer Tube Per Drawings 25412 (Reference 2.18) and 25407 (Reference 2.17, Detail 1), the ceiling of the Penetration Building Elevation 15'0" corridor below the fuel transfer tube is at plant elevation 21'6" Per these same drawings, the floor of the fuel transfer tube enclosure is at plant elevation 26'4" Therefore, there is 4'10" of concrete separating the fuel transfer tube enclosure from the Penetration Building Elevation 15'0" corridor below.
- 8. Concrete Shielding Above the Fuel Transfer Tube Per Drawings 25412 (Reference 2.18) and 25407 (Reference 2.17, Detail 1), the floor of the Penetration Building Elevation 15'0" corridor above the fuel transfer tube is at plant elevation 38'3" Per these same drawings, there is 5'3" of concrete separating this corridor from the fbel transfer tube enclosure below.
l I sce n.ca REV O EH @ETERENCE 5013-XXIV 115)
EC&FS DEPARTMENT iCCN NO. CALCULATION SHEET CCN COWERSiON-CCN NO. CCN . Project or DCP/FCN N/A Cale No. N-0. tan 001 abject Fuel Transfer Tu w Shieldir n . __ _ _ ____ ..__ _ Sheet No _80- oL 117._ _ l REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 6 MARK DRUCKER 05-31-96 C. W. SAYLES 06-03-96 y I l V. METHODOLOGY The objective of Revision 4 to this calculation is to determine the dose rate in the Penetration Building Elevation 15'0" corridor below the fuel transfer tube using. as-built dimensions and the SONGS specific source term. Units 2 and 3 General Arrangement Drawings 10000 and 10050 (Reference 2.19) depict Penetration Building floor plans including the Elevation 15'0" passage (i.e., corridor). Figure V-1 is an excerpt from the Unit 2 drawing showing the "3'0" WIDE PASSAGE AT EL.15'0"", and its surrounding structure. This corridor is also depicted in l Section A of General Arrangement Drawing 10007 (Reference 2.20). Figure V-2 is an excerpt from Section A of this drawing showing the "OPEN" space of the corridor. A direct proportionality exists between dose rate and source term. As such, this calculation will determine the dose rate in the Penetration Building Elevation 15'0" corridor for the source term of a single fuel assembly in the fuel transfer tube. The dose rate will then be doubled to define the I dose rate for the source term of two side-by-side fuel assemblies in the fuel transfer tube. l l The SOURCE 2 Code (Reference 6 5)id to calculate the source strength spectrum of a single fuel assembly. The maliodology used to define the fuel assnbly source strength spectrum with the SOURCE 2 code is addressed in Section V. I. Determination of the dose rate in the Penetration Building Elevation 15'0" corridor requires determination of the average fuel assembly density. The methodology used to determine the average fuel assembly density of a single fuel assembly is addressed in Section V.2. The SHIELD-SG Code (Reference 6.6) is used to calculate the dose rate in the Penetration Building Elevation 15'0" corridor. The methodology used to determine the dose rate with the SHIELD-SG Code is addressed in Section V.3. I I 3CE 2H26 REY 0 E94 [RERRENCE 50m-XXTV 715)
EC&FS DEPARTMENT iccu no, CALCULAT10N SHEET CCN CONVERSION-CCN NO. CCN . l Project or DCP/FCN N/A Calc No. N-0.1RG 001 ibiect Fuel Transfer Tu w Shieldir n - . _ . _ ____ _ . _ . . ___ _ _ ___ Sheet No . 81 of_117 REV ORIGINATOR DATE IRE DATE l REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05 % C. W. SAYLES 06-03-% h , I i i Figure V-1 , UNIT 2 PENETRATION BLDG FLOOR PLAN IN VICINITY OF l l FUEL TRANSFER TUBE (Excerpt from Drawing 10001) l 4mr hb b "" h v\k:
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gQ* 4 SCE 26=326 REV 0 EH [REITRENCE 30123-XXIV 7.15)
EC&FS DEPARTMENT ICCN 3 CALCULATION SHEET CCN CONVERSION CCN No. CCN . Project or DCP/FCN N/A Calc No. N 0.1A0.001 ubiect Fuel Transfer Tu w Shieldir n Sheet No 12_ of .117. _ _ REV ORIGlWATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 6 MARK DRUCKER 05-31 96 C. W. SAYLES 06 % y I I l V.1 Fuel Assembly Source Strength Spectrum Previous revisions of this calculation present parametric evaluations of proposed concrete i shielding modifications to the 1973 design using a generic source term. Revision 4 of this calculation utilizes a SONGS specific source term. Section VIII.1 uses the SOURCE 2 Code (Reference 6.5) to determine a source strength spectrum for a single fuel assembly based on the SONGS specific source term and a User specified energy grouping arrangement. The methodology used to define the fuel assembly source strength spectrum with the SOURCE 2 Code is addressed in this section. The SONGS specific source term represents the activity concentration (curies /cc) of each isotope that is present in a fuel assembly. When input into the SOURCE 2 Code, the code uses its gamma disintegration energy library to manipulate the activity concentration profile into a corresponding source strength spectmm (Mev/cc-sec). 4 g The activity concentration profile is entered into the SOURCE 2 Code as an isotope-specific core activity profile, a multiplier to convert the core activity profile to a fuel assembly activity concentration profile, and a User specified decay time. The isotope-specific core activity profile is 4 per Design Input I which defines the total activity loading of the core for a 60 GWD/ ton discharge burnup level at the time of reactor shutdown. The multiplier converts this activity profile to an activity concentration profile for a single fuel assembly by considering the following parameters in the following equation: M = (RPF x APF) / (N% x V) where: M = SOURCE 2 Code Multiplier (core /cc) , RPF = Radial peaking factor. (unitiess) [per Assumption 5] APF = Axial peaking factor (unitiess) [per Assumption 6] Nmy = Number of fuel assemblies in the core (assemblies / core) [per Design Input 2] V = Volume of a fuel assembly (cc/ assembly) [per Design Inputs 4 and 5 dimensions] Per Assumption 3 this analysis assumes the reactor has been subcritical for at least 72 hours prior to the movement ofirradiated fuel. Since the activity profile is based on Design Input I and its core activity profile at the time of reactor shutdown, the SOURCE 2 Code input will consider a User specified decay time of 72 hours. { } SOURCE 2 will determine the gamma energy source strength spectrum using the SOURCE 2 BASE 10 energy stmeture. The BASE 10 gamma energy structure is shown in Table V-1. Per sa su urv . m mznne som.xxmm
EC&FS DEPARTMENT ICCN NO. CALCULATION SHEET CCN CONVERSION-CCN NO. CCN - _ Project or DCP/FCN N/A Calc No. N-04RG 001 biect Fuel Transfer.Tu w Shieldir a . __ Sheet No 83 of.117.
] REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R l0 MARK DRUCKER 05-31-96 C. W. SAYLES 06-03-96 h
l i l Bechtel Design Guide 3DG-N63-011 (Reference 6.7), the BASE 10 gamma energy stmeture is based on the " SPECTRUM A"in Table 7-2 of ANL-5800, which was established for general gamma shield design purposes. The BASE 10 gamma energy stmeture differs from that of
" SPECTRUM A" in that the BASE 10 stmeture includes additional low and high energy ranges to more accurately address the dose contributions of the low and high energy gamma emitting isotopes. '
The SOURCE 2 Code will be executed on the Nuclear Fuel Management IBM-RISC 6000 workstation. Use of the SOURCE 2 Code on the IBM-RISC 6000 workstation has been verified t and validated as detailed in a Software Installation Report (Reference 6.8). I Section IX.1 of this calculation presents the input and output files associated with the SOURCE 2 j code analysis. Table V-1
" Base 10" GAMMA ENERGY STRUCTURE j l I Energy Ranges Effective Energy Energy Group Number J (Mev/y-disintegration) (Mev/y-disintegration) I l
1 0.0 s E s 0.1 0.1 ~ 2 0.1 < E s 0.4 0.4 l 3 0.4 < E s 0.9 0.8 1 4 0.9 < E s 1.35 1.3 l 1 5 1.35 < E s 1.80 1.7 6 1.80 < E s 2.20 2.18 7 2.20 < E s 2.60 2.5 8 2.60 < E s 3.00 2.8 9 3.00 < E s 5.00 4.0 10 5.00 < E s 15.00 6.2 l I i 3CE 26426 REV 0 3,94 [REIT.RENCE $O121-XXIV.715J
)
EC&FS DEPARTMENT ICCN NO. CALCULATION SHEET CCN COWERSION. cCN No. CCN . I roject or DCP/FCN N/A Calc No. N 0.1AS 003 (Iubiect Fuel Transfer Tu w Shieldir z _ ___._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ Sheet No _84_ oL117_ l REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05 % C. W. SAYLES 06 % V.2 Fuel Assembly Composite Density The single fuel assembly source strength spectrum determined in Section VIII.1 will be input into the SHIELD-SG Code (see Section V.3) to determine the dose rate in the corridor below the fuel transfer tube. However, prior to running the SHIELD-SG Code, Section VIII.2 will determine the composite (or average) fuel assembly density. This parameter is ofinterest in that it l determines the degree of self-shielding afforded by the fuel assembly. Of note, however, is that the corridor dose rate should be negligibly impacted by the calculated average fuel assembly density due to the significant attenuation provided by the several feet of concrete shielding separating the corridor from the fuel assembly (per Design Input 7). The methodology used to determine the average fuel assembly density of a single fuel assembly is addressed in this section. Section VIII.2 will calculate an average density that is representative of a single fuel assembly in ! the fuel transfer tube. This average density will be determined by volume averaging the significant regions of a fuel assembly. These regions include, the uranium dioxide fuel, the helium gap space 4 ( ) between the fuel pellets and the fuel rod cladding, the Zircaloy-4 fuel rod cladding, and the water ~ ' surrounding the fuel rods within the box-like shape of the fuel assembly. To simplify the average I density determination, water will be assumed to exist in the space occupied by the metallic l supports and cross-ties. The omission of the supports and cross-ties from consideration is j conservative in that the resultant average fuel assembly density will be lowered and less . self-shielding will result. Based on this logic, the average density of a fuel assembly is: pay = (pu x VFu) + (py, x VF,,) + (pu x VFu) + (p , x VF ,,) where: p% = Composite (or average) density of a fuel assembly (gram /cc) pu = Density ofuranium dioxide fuelin a fuel assembly (gram /cc) VFw = Volume fraction of uranium dioxide fuel in a fuel assembly (unitless) py, = Density of Helium gas in the gap of a fuel assembly (gram /cc) VF,, = Volume fraction of Helium gas in the gap of a fuel assembly (unitiess) pu = Density of Zircaloy-4 cladding in a fuel assembly (gram /cc) VFu = Volume fraction ofZircaloy-4 cladding in a fuel assembly (unitless) p, = Density of water around fuel rods in a fuel assembly (gram /cc) VF,,, = Volume fraction ofwater around fuel rods in a fuel assembly (unitiess) The fuel, gap, cladding and water densities are addressed in Assumptions 7 through 10. l I 3CE 2&G REY.0 SH @EFERENCE 30W-XXIVo l5)
EC&fS DEPARTMENT ICCN NO.
""'"'""~ '--
CALCULATION SHEET CCN CONVERSION-i CCN NO. CCN - Project or DCP/FCN__N/A Calc No. N-MAG-001 (y ubiect Fuel Transfer Tu w Shieldir 2 - __ _ _. _ _ - _.. _ . Sheet No _85 of .11 L REV ORIGINATOR DATE 1RE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05-31-% C. W. SAYLES 06 % y i For a constant active fuel length, the volume fractions (VFs) may be calculated for each fuel , assembly region by ratioing the region's cross-sectional area to the total fuel assembly cross-sectional area: VFw = Au + A,,,,, VF,, = Ay, + A,,, 9 VFu = Au + A,,,sq VF.,. = A. + A,,, 9 where: A,,,s, = Cross-sectional area of a fuel assembly (cm2) Au = Cross-sectional area ofuranium dioxide fuel in a fuel assembly (cm2) A,y = Cross-sectional area of Helium gas in the gap of a fuel assembly (cm2 ) Au = Cross-sectional area of Zircaloy-4 cladding in a fuel assembly (cm2) A = Cross-sectional area of water around fuel rods in a fuel assembly (cm2 ) l > Per Design Input 5, each fuel assembly has an overall square appearance. Therefore, the total cross-sectional area of a fuel assembly is equivalent to: A,,, y=L ,xL.,,,, where: L ,, 3 = Length of one side of the square fuel assembly (cm) By omitting consideration of supports and cross-ties (and thereby conservatively omitting their steel gamma radiation attenuation characteristic), the cross-sectional area of the water space surrounding the fuel rods within the fuel assembly is: A,,,,, = A + Au + A y, + A ci,a
.A = A muy - [ Au + Ay , + Au ]
The cross-sectional area for the fuel, gap and cladding regions of the fuel assembly are calculated by scaling the cross-sectional area of these regions in each individual fuel rod by the total number of fuel rods in the assembly: 2 Au = (N%) x [(n/4) x Du] Ay,= (Na) x {[(n/4) x Dg ,,,, 2] - [(n/4) x Du2)) ( l 2 Au= (N%) x (((n/4) x D% ] - [(n/4) x Dm 3) 2 rs uu uv . 3% prnect som.xuv.wi
l EC&FS DEPARTMENT ICCN NO. I
""""""~ ~ '- '
CALCULATION SHEET CCN CONVERSION-CCN NO. CCN . oject or DCP/FCN N/A Calc No. N-GaAn ont object Fuel Transfer. Tube Shieldire_ __ Sheet No _.86_ of_.117 __ REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05 % C. W. SAYLES 06 % h I where: N% = Number of fuel rods in a fuel assembly (rods / assembly) [per Design Input 3] Du = Fuel Pellet Diameter (cm) [per Design Input 4] D% = Fuel Rod Inner Diameter (cm) [per Design Input 4] D% = Fuel Rod Outer Diameter (cm) [per Design Input 4] V.3 Dose Rate in Penetration Bldg El.15'0" Corridor Below Fuel Transfer Tube Previous revisions of this calculation present parametric evaluations of proposed concrete shielding modifications to the 1973 design. Revision 4 of this calculation utilizes as-built dimensions in its evaluation. Section VIII.3 uses the SHIELD-SG Code (Reference 6.6) to calculate the dose rates due to the presence of a single fuel assembly in the fuel transfer tube. The methodology used to calculate the dose rates with the SHIELD-SG Code is addressed in this section. l I Bechtel Standard Computer Program NE-650, " SHIELD-SG" is an interactive program designed to solve gamma ray transport problems using the point kernel method (i.e., given an input energy spectmm and source / shield geometry, it calculates the gamma dose rate through the shield). The source / shielding geometry can be any combination of up to fifty geometrical bodies, including orthogonal slabs, right paraMepipeds, and/or right cylinders. Multiple dose points can be specified in a single mn. The SHIELD-SG Code determines the doses in the Penetration Building Elevation 15'0" corridor below the fuel transfer tube as a function of a gamma energy source strength spectrum and geometry. The source strength spectrum is calculated with the SOURCE 2 Code as detailed in Section V. I . Figure V-3 depicts the SHIELD-SG Code model with its cylindrical geometry. Figure V-3 shows cross-sectional and length-wise views of the SHIELD-SG Code model. Figure V-3 is not drawn to scale, and merely serves to orient the user of this calculation. The SHIELD-SG model will consist of a source body (the fuel assembly in the fuel transfer tube), and the shield bodies that surround the fuel assembly. These shield bodies include the water around the fuel assembly, the steel fuel transfer tube cylinder (which will be conservatively omitted from the model), the air gap around the fuel transfer tube, and the concrete shielding. l l scs 2u26 arv o sH [ REFERENCE SO12FXXIV-7.15]
i EC&FS DEPARTMENT icCs a. l
""""""' ~ ~ ' - -
CALCULATION SHEET ! CCN CONVERSION: CCN NO. CCN . hoject or DCP/FCN N/A Calc No. N.O.iA0.001 biect Fuel Transfer Tu >e Shieldirt. _ _ . _ _ _ _ _ Sheet No . 87. of_117_. REV ORIGINATOR DATE IRE DATE REV , ORIGINATOR DATE IRE DATE R
~
6 MARK DRUCKER 05-31-96 C. W. SAYLES 06-03-96 E y 1 To simplify the model into cylindrical bodies, the square cross-sectional area of the fuel assembly will be approximated as a circle with an equivalent cross-sectional area. The radius of the equivalent fuel assembly circular area is: A% = A, 4 nR 2= L2 4 R=L/n 5 In addition, to facilitate evaluation of various shielding thicknesses the concrete shield will be modeled with a thickness greater than that which is actually present. Each cylindrical body will be modeled with a length equivalent to the active fuel height, and the dose points will be positioned at midlength (to maximize the dose rate contribution from the entire fuel assembly length), and at various radial distances from the fuel assembly centerline. The distances will include one dosepoint positioned 4'10" into the concrete shield representing the Penetration Building Elevation 15'0" corridor beneath the fuel transfer tube. Due to the extent of concrete shielding between the source and this dosepoint, this dosepoint will be modeled with the
- ( ) concrete build-up factor coded into the SHIELD-SG Code library.
The fuel assembly source body will be modeled with a composite (average) density as determined
; in Section VIII.2. The shield bodies will be modeled with appropriate densities as addressed in Assumptions 10 through 12.
The SHIELD-SG Code will be executed on the Nuclear Fuel Management IBM-RISC 6000 workstation. Use of the SHIELD-SG Code on the IBM-RISC 6000 workstation has been verified and validated as detailed in a Software Installation Report (Reference 6.9). Section IX.2 of this calculation presents the input and output files associated with the 1 SHIELD-SG Code analysis. sCE M4M REV 0 &H [ REFERENCE 30123-XXIV 415)
EC&FS DEPARTMENT scCN G.
""""""~ PAGE _0F _
CALCULATION SHEET CCN CONVERSION-CCN NO. CCN - Project or DCP/FCN N/A Cale No. N 0.tAMM
,ubiect Puel Transfer Tine Shieldir a _ _ . _ . _ _ _ ._.
.- Sheet No . 88.. of_117_ _
REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05 % C. W. SAYLES 06-03-% y i Figure V-3 SHELD-SG CODE MODEL GEOMETRY (not drawn to scale) 4 BODY 4 - CONCRETE SHIELD 3 BODY 3 - AIR ABOUT TUBE BODY 2-WATERIN TUBE 2 1 BODY 1-FUEL ASSEMBLY SOURCE BODY 2 - WATER IN TUBE BODY 3 - AIR ABOUT TUBE BODY 4 - CONCRETE SHIELD l I I I set 26-426 REv.o sw [ REFERENCE 5012k%IIV USl E
EC&FS DEPARTMENT secN NO. CALCULATION SHEET CCN CONVERSION-CCN NO. CCN . Project or DCP/FCN N/A Calc No. N.04An-not ( \ubiect Fuel Transfer Tu o
)e Shieldir _
Sheet No _89.. o0.117.__ REV ORIGINATOR !RE DATE DATE l REV ORIGINATOR DATE !RE DATE R 4 MARK DRUCKER 05-31-96 C. W. SAYLES 06-03-% l y 1 VI. REFERENCES VI.1 Calculations 1.1 Analysis A-SG2-FE-0058, Revision 01," SONGS-2 Cycle 9 CORD / ROCS /MC Design Models, Depletions, Rodded Cases, and Integrated Files Using ROCS 4.1.7 and MC 4.1.6", issued May 08,1996 1.2 Calculation N-4097-013, Revision 0 including CCN-1, " Spent Fuel Pool Rerack Activities , and Source Strength Spectrums" l l VI.2 Drau ings and Documents 2.1 Radiation Zones - Plan at El. 30'-0" to El. (-)l5'-6" { } a) Unit 2 Drawing 40023, Revision 4 b) Unit 3 Dnwing 40027, Revision 2 2.2 Document SO2.!-990-200, Revision 0, " Manufacturing Order No. 5403-95-2025 Supplemes: 00', dated January 18,1996 2.3 Document M-38907, Revision 21, " Plant Physics Data Book for San Onofre Unit 2 Cycle 7" NOTE: Previous revisions of this document present plant physics data for earlier Unit 2 cycles. 2.4 Document M-38908, Revision.15, " Plant Physics Data Book for San Onofre Unit 3 Cycle 7" NOTE: Previous revisions of this document present plant physics data for earlier Unit 3 cycles. 2.5 ABB Supplier Document S2L-FMDE-BOM1, Revision 00, " Bill of Materials for San Onofre Unit 2 Batch L" issued January 18,1996. NOTE: No SONGS CDM Document number was assigned to this ABB Supplier Document. This ABB Supplier Document is incorporated into Document SO23-990-200 [see Reference 2.2] 2.6 Document SO23-990-10, Revision 3 (ABB Supplier Document 00000-PD-110, Revision 09), " Specification for Uranium Dioxide Fuel Pellets", issued October 06,1994 I l sCE 2HM REY.0 kW lREFTRENCE SO123-3OUV 7 IS)
EC&FS DEPARTMENT ICCN NO.
""""""~ ""~ "
CALCULATION SHEET CCN CONVERSION. l CCN NO. CCN - Project or DCP/FCN N/A Calc No. N-04AO 001 abiect Fuel Transfer Tu ne Shieldir a . -__ __ __ . _ _ __ _ ___ _ _ Sheet No _90 . of 117_ _ REV ORIGINATOR DATE IRE DATE lREV ORIGINATOR DATE IRE DATE R 6 MARK DRUCKER 05-31-96 C. W. SAYLES 06-03-96 y 1 2.7 Document S023-990-13, Revision 1 (ABB Supplier Document 00000-PD-301, Revision 04), " Specification for Zircaloy-4 Fuel Rod Cladding Tubes", issued July 17, 1991 This Document was modified by Document SO23-990-152, Revision 0 (ABB Vendor ) Document SPl 16-04 to 00000-PD-301), " Supplemental Purchase Information for Zircaloy-4 Fuel Rod Cladding Tubes to be Procured from Sandvik Special Metals (SSM) USA", issued August 31,1993 2.8 Drawing SO23-990-165, Revision 0 (ABB Supplier Drawing E-F16-E000-F13, Revision 01)," FUEL ROD ASSY" 2.9 Drawing SO23-990-167, Revision 0 (ABB Supplier Drawing E-F16-E000-F14, Revision 01), "UO2 -Er203 ROD ASSY" { } 2.10 Document SO23-990-8, Revision 5 (ABB Supplier Document 00000-FMDE-0100, Revision 09), " Standard Engineering Specification for Fuel Assemblies", issued January 12,1995 This Document was modified by Document SO23-990-158, Revision 0 ABB Vendor Document SPI l-08 to 00000-FMDE-0100), " Supplemental Purchase Information for Standard Engineering Specification for Fuel Assemblies", issued September 9,1994 2.11 Drawing SO23-990-166, Revision 0 (ABB Supplier Drawing C-F16-E100-F03, Revision 06), " FUEL PELLET" 2.12 Drawing SO23-990-168, Revision 0 (ABB Supplier Drawing C-F16-E100-F04, Revision 06),"UO2 -Er2 03 PELLET" 2.13 Drawing SO23-990-198, Revision 0 (ABB Supplier Drawing C-F16-E200-M07, Revision 01), " TUBE" 2.14 Drawing SO23-990-164, Revision 0 (ABB Supplier Drawing E-FSG-E000-A07, Revision 01)," FUEL BUNDLE ASSEMBLY" 2.15 Drawing SO23-207-4-5, Revision 5 (AhETEK Supplier Drawing 7-030043, Revision E),
" Fuel Transfer Tube Assembly - Pipe Assembly" scE 26.cM REV.0 SH [ REFERENCE SO12FXXIV 715)
f EC&FS DEPARTMENT ICCN NO. CALCULATION SHEET CCN COWERSION' CCN NO. CCN . Project or DCP/FCN N/A Calc No. _&0480 003 ubiect Fuel Transfer Tu m Shieldir 2 - __ _ _ _ _ _ .- _ _ _ _ _ - - . __ _ _ _ ___ Sheet No _91_ of__117_ REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05-31-96 C. W. SAYLES 06-03 96 h I 2.16 Drawing 23119, Revision 7, " Containment Interior Struct. - Reinforced Concrete - Cross Section Facing North" 2.17 Drawing 25407, Revision 24, " Fuel Handling Building - Exterior Wall - West Elevation" l 2.18 Drawing 25412, Revision 21, " Fuel Handling Building - Concrete Sections and Details - Sht. 3" 2.19 Containment Structure Penetration Building Floor Plan, Elevations 9'0" and 15'0" a) Unit 2 Drawing 10001, Revision 16 b) Unit 3 Drawing 10050, Revision 14 2.20 Drawing 10007, Revision 3, " Containment Structure - Penetration Bldg. Unit 2 -Bldg. I Section & Details" l l VL3 Correspondence No Correspondence is referenced in this calculation. VL4 Regulatory Documents 4.1 NUREG/CR-5009 (also numbered PNL-6258), " Assessment of the Use of Extended Burnup Fuel in Light Water Power R.eactors", published February 1988 4.2 SONGS Units 2&3 Updated Final Safety Analysis Report, up to and including Amendment 11, published March 1996 4.3 NUREG-0741, SONGS Unit 2 Technical Specifications, Appendix A to License NPF-10, up to and including Amendment No.128 4.4 NUREG-0741, SONGS Unit 3 Technical Specifications, Appendix A to License NPF-15, up to and including Amendment No. I17 VL5 Site Procedures and Operating Instructions l No Site Procedures or Operating Instructions are referenced in this calculation. sCE 26-426 REV.0 &H (REIT.RENCE 3012% XXIV-715]
EC&FS DEPARTMENT iCCN Q. CALCULATION SHEET CCN CONVERSION: CCN NO. CCN . Project or DCP/FCN N/A Calc No. _N-0480-001 ubiect Fuel Transfer Tu a Shieldir e - _ _ . . __ __ _ __ _ Sheet No 92 . of_ 117___ REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05 % C. W. SAYLES 06 % h I l VI.6 Other Documents l 6.1 Nuclear Fuel Managment. by Harvey W. Graves, Jr., published by John Wiley & Sons, New York,1979 6.2 Handbook of Physical Prooerties of Liauids and Gases. Second Edition. by N. B. Vargaflik, published by Hemisphere Publishing Corporation, Washington,1983 l 6.3 ASME Steam Tables. Sixth Edition, published by the American Society ofMechanical ' Engineers, New York City,1993 6.4 American National Standard ANSI /ANS-6.4-1977 (N403), approved August 8,1977, !
" Guidelines on the Nuclear Analysis and Design of Concrete Radiation Shielding for I Nuclear Power Plants" l
l 6.5 SOURCE 2 Code, Bechtel Standard Computer Program NE-602, Version D2-5, dated ) December 1991 ' 6.6 SHIELD-SG Code, Bechtel Standard Computer Program NE-650, Version D2-10, dated December 1991 6.7 Bechtel Nuclear Engineering Standard 3DG-N61-003 (previously numbered N6.1.3), Revision 0, " Nuclear Data for Radionuclides", dated June 1983 6.8 Software Installation Report. Revison 1, SOURCE 2 (NE-602) Version D2-5, RISC 6000 Computer System - Device ID D026748, Operating System AIX Version 3.2.5, Approved October 27,1995 6.9 Software Installation Report. SHIELD-SG (NE-602) Version D2-10, RISC 6000 Computer System - Device ID D026748, Operating System AIX Version 3.2.5, Approved October 27,1995 l I l scE 2wa uv o sM [REFErdENCE 80171 XXIV-719]
EC&FS DEPARTMENT ICCN NO.
""*"""~ " " ' -
CALCULATION SHEET CCN CONVERSION-CCNNO. CCN . Project or DCP/FCN N/A Calc No. N 0.1A0_001 l bbiect Fuel Transfer Tule Shieldirr . _ . . _ _ _ _ _ . . _ _
._ _ Sheet No 93 of 117 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE !RE DATE R 6 MARK DRUCKER 05-31-96 C. W. SAYLES 06-03-96 h I
VB. NOMENCLATURE A Cross-sectional Area APF Axial Peaking Factor cc cubic centimeters em centimeter dm decimeter D Diameter E Gamma Disintegration Energy , ft foot ' GWD Gigawatt-days I K degrees KeMn kg kilogram l lbm pound-mass l L Length ( q g LCO Limiting Condition for Operation (per the Unit Technical Specifications) l m meter M SOURCE 2 Code Multiplier Mwt Megawatt-thermal n Number of moles (used in Ideal gas Law) N Number (quantity) of an item P Pressure , psia pound-force per square inch, absolute psia pound-force per square inch, gauge PWR Pressurized Water Reactor R Ideal Gas Law Constant R Radius RCTS Regulatory Commitment Tracking System l RPD Relative Power Density RPF Radial Peaking Factor t ton (of uranium) T Temperature UO 2 Uranium Dioxide V Volume VF Volume Fraction y Gamma Particle p Density { }v Specific Volume SCE 26-426 REV O E94 [ REFERENCE SO123-XXIV 415)
l < 4 Control No. 05931 CERTIFIC ATE OF AUTHENTICITY (V 1 i CONTINUATION This is to certify that the microphotographic image appearing on this microform are direct and facsimile reproduction of the original records of the Southem California Edison company and were microfilmed in the regular course ofi.usiness. The microfilming has been performed according to established routine Company l Policy for systems utilization and/or maintenance and preservation of records through the storage of each microformsin protected locations. THE DOCUMENTS CONTAINED ON THIS MICROFORM ARE ORIGINAL RECORDS OF: [ Southern California Edison l l For the NUCLEAR -INDUSTRIAL ENGINEERING & MANAGEMENT SERVICES Department CDM-SONGS - DESIGN CALCULATIONS This microform file is a complete record of the transaction herein recorded. The documents are l arranged on this microform in the following manner: j O O By month in location and Work Order sequence O OrderofPayrollLocationNumber O Alphabeticalorderby O GrievanceFileNumbersequence S Numerical order by Desicn Cal No O Dateorder O Order of Customer Senice Store Number O Other The hardcopy documents used to create this microform have been authorized for destruction after i verification of correctness and acceptability of the microfilming. It is further certified that on the date specified below, the micrographic images appearing on the microform were made at a reduction ratio of 29 : 1 under my direction and control. l The above informat on is deemed r===y is compliance Utilities and licenses -issued March 14,1972. This with the Federal Powr Commuuan Order No. 450 order has subsequently both approved by the Public Regulation-to govern the Preservation of Public Utilities Commission of the State of Califomia on October 29,1974 M M DATE MICROFILMED CAMERA OP'RATOR AWS BLDG . D-2P SONGS dd44 os LDCATION AUT RIZED SIGNATURE p SUPE 1SOR MICROGRAPHICS U
EC&FS DEPARTMENT icc3 no, CALCULATION SHEET CCN CONVERSION-ccu No. CCN - Project or DCP/FCN N/A Calc No. N 04An_ont
.uhiect Fuel Transfer Tule Shieldir n - _ __ _ __ _ __ __. _ _ _ _ Sheet No . 94 of _117_ l REV ORIGINATOR DATE IRE DATE l REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05 % C. W. SAYLES 06 % y I
VDI. COMPUTATIONS The computations in this calculation have been performed using a hand-held calculator. Wordperfect 6.1 was used only as a word processor, for the purpose of documenting the results of this calculation. VHI.1 Fuel Assembly Source Strength Spectrum Previous revisions of this calculation present parametric evaluations of proposed concrete shielding modifications to the 1973 design using a generic source term. Revision 4 of this calculation utilizes a SONGS specific source term. This section uses the SOURCE 2 Code (Reference 6.5) to determine a source strength spectrum for a single fuel assembly based on the SONGS specific source term and a User specified energy grouping arrangement. The methodology used to defme the fuel assembly source strength spectrum with the SOURCE 2 Code q , is addressed in Section V.I. The SONGS specific source term represents the activi ty concentration (curies /cc) of each isotope that is present in a fuel assembly. When input into the SOURCE 2 Code, the code uses its gamma disintegration energy library to manipulate the activity concentration profile into a corresponding source strength spectrum (Mev/cc-sec). The activity concentration profile is entered into the SOURCE 2 Code as an isotope-specific core activity profile, a multiplier to convert the core activity profile to a fuel assembly activity concentration profile, and a User specified decay time. The isotope-specific core activity profile input into the SOURCE 2 Code is per Design Input I and its Table IV-1, which define the total activity loading of the core for a 60 GWD/ ton discharge burnup level at the time of reactor shutdown. The muhiplier input into the SOURCE 2 Code converts the Design Input I core activity profile to an activity concentration profile for a single fuel assembly. Per Section V.1, the SOURCE 2 Code Multiplier is dependent on the fuel assembly volume, the radial and axial peaking factors, and the number of fuel assemblies in the reactor core. Per Design Input 5, each fuel assembly has an overall square appearance with 8.030 inches separating the outer edge of a fuel rod on one side from the the outer edge of a fuel rod on the opposite side. Per Design Input 4, each fuel assembly has an active fuel length of 150 inches. Therefore, the volume of the fuel assembly is: l l sCE 2MM REV 0 M4 PEFERENCE SOID-XXIV.7 IS]
EC&FS DEPARTMENT icey g, CALCULATION SHEET CCN CONVERSION-CCN No. CCN - Project or DCP/FCN N/A Calc No. N-01AO-001 ubiect TuelTransferTube Shieklir 2 _ _ . _ . _ . _ _ _ _ __
. _ . _ . . _ _ . _ . . . _ _ _ _ Sheet No __95 of_117_ _
REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 6 MARK DRUCKER 05 % C. W. SAYLES 06 % h I 1 l V, =W,xL,xH e l 3 l
= (8.030 inches) x (8.030 inches) x (150 inches) = 9672 in' = 1.58e5 cm I The remaining parameters input into the SOURCE 2 Code Multiplier are:
j RPF = 1.20 radial peaking factor (per Assumption 5) APF = 1.25 axial peaking factor (per Assumption 6) N, = 217 fuel assemblies / core (per Design Input 2) Based on these parameters, the SOURCE 2 Code Multiplier has a value of: i M = (RPF x APF) / (N , x V)
= (1.20 x 1.25) / (217 assemblies / core x 1.58e5 cc/ assembly) = 4.37e-8 cc/ core Per Assumption 3 this analysis assumes the reactor has been subcritical for at least 72 hours prior q g to the movement ofirradiated fuel. Since the activity profile is based on Design Input I and its core activity profile at the time of reactor shutdown, the SOURCE 2 Code input will consider a User specified decay time of 72 hours.
Lastly, for reasons presented in Section V.1, the SOURCE 2 Code will determine the gamma energy source strength spectrum using the SOURCE 2 BASE 10 energy structure. The SOURCE 2 Code has been executed on the Nuclear Fuel Management IBM-RISC 6000 workstation. Section IX.1 of this calculation presents the input and output files associated with the SOURCE 2 code analysis. Table VIII.1-1 presents the source strength spectmm as determined with the SOURCE 2 Code for a single fuel assembly after 72 hours of decay. l I scE2w26 arv o aH [ REFERENCE SO123-XXTY+715)
I i EC&FS DEPARTMENT ICCN O. CALCULATION SHEET CCN CONVERSION-CCN No. CCN . i Project or DCP/FCN N/A Calc No. N-0.1An-003
.ubiect Fuel Transfer Tu ie Shieldir n __ _ _ - . . . _ _ -
. _. Sheet No . 96. of._117.._
f REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R ' 4 MARK DRUCKER 05-31-96 C. W. SAYLES 06-03-96 E f y i I ' l l l Table VHI.1-1 ; SOURCE 2 CODE OUTPUT - SOURCE STRENGTH SPECTRUM ; Source Strengths Energy Group Number nerg ges ec e erg One Fuel AsseinMy (Mev/y-disintegration) l (Alev/y-disintegration) at 72 hours decay j j (Mev/cc-second) 1 0.0 s E s 0.1 0.1 9.33e+09 ( 2 0.1 < E s 0.4 0.4 9.99e+10 3 0.4 < E s 0.9 0.8 9.55e+11 ! l 4 0.9 < E s 1.35 1.3 4.92e+10 j 5 1.35 < E s 1.80 1.7 1.63e+11 6 1.80 < E s 2.20 2. I8 1.10e+10 ; { l 7 2.20 < E s 2.60 2.5 1.17e+10 8 2.60 < E s 3.00 2.8 1.71e+08 9 3.00 < E s 5.00 4.0 8.71e+07 l 10 5.00 < E s 15.00 6.2 0.00e+00 TOTAL 0.0 < E s 15.00 - 1.30e+12 ! i u f l
!I
EC&FS DEPARTMENT g so, CALCULATION SHEET CCN CONVERSION: CCN No. CCN - Project or DCP/FCN N/A Calc No. .N_. ann.not abiect Fuel Transfer _Tu a Shieldirn . _ __ _ . . _ _ _ _ _ _ _ _ _ _ _ _ Sheet No _97_. oL 117 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER ,6_-31-% C. W. SAYLES 06-03-% h
)..
i l I l VIIL2 Fuel Assembly Composite Density ! The single fuel assembly source strength spectrum determined in Section VIII.1 will be input into the SHIELD-SG Code (see Section VIII.3) to determine the dose rate in the corridor below the fuel transfer tube. However, prior to running the SHIELD-SG Code, this section will determine the composite (or average) fuel assembly density. This parameter is ofinterest in that it determines the degree of self-shielding afforded by the fuel assembly. Ofnote, however, is that the corridor dose rate should be negligibly impacted by the calculated average fuel assembly density due to the significant attenuation provided by the several feet of concrete shielding separating the corridor from the fuel assembly (per Design Input 7). The methodology used to determine the average fuel assembly density of a single fuel assembly is addressed in Section V.2. This section calculates a composite density that is representative of a single fuel assembly in the foci transfer tube. This average density is determined by volume averaging the significant regions of a fuel assembly. These regions include, the uranium dioxide fuel, the helium gap space between the fuel pellets and the fuel rod cladding, the Zircaloy-4 fuel rod cladding, and the water surrounding the fuel rods within the box-like shape of the fuel assembly. To simplify the average density determination, water will be assumed to exist in the space occupied by the metallic supports and cross-ties. The omission of the supports and cross-ties from consideration is conservative in that the resultant average fuel assembly density will be lowered and less self-shielding will result. The crossweetional area for the fuel, gap and cladding regions of the fuel assembly are calculated by scaling the cross-sectional area of these regions in each individual fuel rod by the total number of fuel rods in the assembly. Based on Design Input 4, the fuel rod and pellet are characterized by the following dimensions: D% = Fuel Pod Outer Diameter = 0.382 inches = 0.970 cm (per Design Input 4) D,w, w = Fuel Rud Inner Diameter = 0.3320 inches = 0.843 cm (per Design Input 4) Du = Fuel Pellet Diameter = 0.3255 inches = 0.827 cm (per Design Input 4) ' Per Design Input 3, there are 236 fuel rods in a fuel assembly. Therefore, the cross-sectional areas defining a fuel assembly are: 2 Au = (236 rods) x [(n/4) x Du ] = (236 rods) x [(n/4) x (0.827 cm)2] = 127 cm 2 A, = (236 rods) x {[(n/4) x D,g,,2] - [(n/4) x Du2)) 2
= (236 rods) x (((n/4) x (0.843 cm)2] - [(n/4) x (0.827 cm)2]) = 4.95 cm m u m n. pm-ce somxuv.wi
EC&FS DEPARTMENT ICCN NO.
" " '" " ~ "
CALCULATION SHEET CCN COWERSiON-CCN NO. CCN - Project or DCP/FCN N/A Calc No. N-04An 001 abiect Fuel Transfer Tu w Shieldir e _ _. ._ _ _ ._. _ ___ Sheet No . 98_ of .117_ i l REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 6 MARK DRUCKER 05-31-96 C. U. SAYLES 06 % h l I l Au = (236 rods) x {[(x/4) x D% 2) _ [(3/4) x p 2)) 2
= (236 rods) x {[(x/4) x (0.970 cm)2] - [(x/4) x (0.843 cm)2]} = 42.7 cm Per Design Input 5, each fuel assembly has overall square appearance with 8.030 inches separating the outer edge of a fuel rod on one side from the the outer edge of a fuel rod on the 4 opposite side. Therefore, the total cross-sectional area of a fuel assembly is:
A% = (8.030 inches) x (8.030 inches) = 64.5 in = 416 cm 2 2 By omitting consideration of supports and cross-ties (as discussed in Section V.2), the cross-sectional area of the water space surrounding the fuel rods within the fuel assembly is: A.,,, =A+A+A,+by
.A = Am33 -[A +A,+Au]
2 2 2 2
= (416 cm ) - [127 cm + 4.95 cm + 42.7 cm ] = 241 cm I I For a constant active fuel length, the volume fractions (VF) may be calculated for each fuel assembly region by ratioing the region's cross-sectional area to the total fuel assembly cross-sectional area:
2 2 VFui = A + A,,,,ms, = (127 cm ) + (416 cm ) = 0.305 VF,, = A ,,+ A,,,,m3, 2
= (4.95 cm ) + (416 cm ) = 0.0119 2
VFu = A u + A.,,,m3, 2
= (42.7 cm ) + (416 cm ) = 0.103 2
2 2 VF.,. =A.+A m3, = (241 cm ) + (416 cm ) = 0.579 Per Assumptions 7 through 10, the material densities associated with the fuel assembly regions are: pui = 10.30 gram /cm' (per Assumption 7) py, = 4.5e-3 gram /cm' (per Assumption 9) pu = 6.44 gram /cm' (per Assumption 8) p = 0.94 gram /cm3 (per Assumption 10) Therefore, the average (or composite) density of a single fuel assembly is: p.,,,ms, = (pui x VFu) + (p y , x VF,,) + (pu x VFu) + (p x VF , )
= [(10.30 gram /cm') x (0.305)] + [(4.5e-3 gram /cm') x (0.00119)] +
+ [(6.44 gram /cm') x (0.103)] + [(0.94 gram /cm') x (0.579)]
i ) = 4.35 gram /cm3 sema m. im numace nom.xxw.m I
EC&FS DEPARTMENT ICCN NO. CALCULATION SHEET CCN CONVERSION: CCN NO. CCN . Project or DCP/FCN N/A Calc No. X-04An nM ibiect Fuel Transfer Tube Shielding _ __ _ _ _ _ _ _ _ _ _ _. . Sheet No . 99 . of.117__ REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 6 MARK DRUCKER 05 % C. W. SAYLES 06 % y I VIII.3 Dose Rate in Penetration Bldg El.15'0" Corridor Below Fuel Transfer Tube Previous revisions of this calculation present parametric evaluations of proposed concrete shielding modifications to the 1973 design. Revision 4 of this calculation utilizes as-built dimensions in its evaluation. This section uses the SHIELD-SG Code (Reference 6.6) to calculate the dose rates due to the presence of a single fuel assembly in the fuel transfer tube. The methodology used to calculate the dose rates with the SHIELD-SG Code is addressed in Section V.3. The SHIELD-SG Code determines the dose rates in the Penetration Building Elevation 15'0" corridor below the fuel transfer tube as a function of a gamma energy source strength spectmm and geometry. The source strength spectrum is calculated with the SOURCE 2 Code in Section VIII.1. Figure V-3 of Section V.3 depicts the SHIELD-SG Code model with its cylindrical geometry. 4 g Detailed information on each body is provided in Section VIII.3.1. The dose points are positioned at various radial distances from the midpoint of the active fuel length. This modeling will maximize the dose rate contribution from the entire feel assembly length. Detailed information on each dose point is provided in Section VIII.3.2 Section VIII.3.3 summarizes the resultant dose rates as determined by the SHIELD-SG code run. VIII.3.1) SHIELD-SG Bodies Figure V-1 of Section V.3 depicts the SHIELD-SG Code model with its cylindrical geometry. The SHIELD-SG model consists of a source body (the fuel assembly in the fuel transfer tube), and the shield bodies that surround the fuel assembly. These shield bodies include the water around the fuel assembly, the steel fuel transfer tube cylinder (which will be conservatively omitted from the model), the air gap around the fuel transfer tube, and the concrete shielding. Detailed information on each body is provided in the following text. s sca 2w26 *rv o am (*nuENCE. 30G-XXIV-11$]
EC&FS DEPAllTMENT ICCN NO.
""""""~ ' ' ' ' i CALCULATION SHEET CCN CONVERSION-CCN No. CCN .
Project or DCP/FCN N/A Cale No. NJMAO004
.ubiect Fuel Transfer Tu w Shieldir e _ _ _ _ _ _ _ . _ . _ ___ _ _
_ _ _ Sheet No __100_ cf _117_ REV ORIGINATOR DATE IRE DATE l REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER Os-31-96 C. W. SAYLES 06 % h 8
-. s Dody 1: Fuel Assembly Body 1 is the source region body. In this calculation the source region is the single fuel assembly passing through the fuel transfer tube. This body has the following characteristics:
Geometry Cylinder (assumed for simplification, see following text) Cylinder Axis X (arbitrary choice) l Cylinder Radius 4.53 inches (determined in following text) Cylinder Length 150 inches (active fuel length per Design Input 4) Material PWR-Core (assumed to best represent fuel characteristics) Density 4.35 gram /cc (determined in Section VIII.1) To simplify the SHIELD-SG input file, Body 1 is modeled as a cylinder. The square cross-sectional area of the fuel assembly is approximated as a circle with an equivalent cross-sectional area. Per Design Input 5, each fuel assembly has an overall square appearance 4
, with 8.030 inches separating the outer edge of a fuel rod on one side from the the outer edge of a fuel rod on the opposite side. Therefore, the equivalent fuel radius of the equivalent fuel assembly circular area is:
3 Am = A, 4 nR = 2 L2 4 R=L/n R = (8.030 inches) / (x"5) = 4.53 inches Body 2: Water in the Fuel Transfer Tube Body 2 is the water surrounding the fuel sssembly within the confines of the fuel transfer tube. This body has the following characteristics. Geometry Cylinder Cylinder Axis X (chosen for consistency with Body 1 model) Cylinder Radius I foot,6 inches (per Design Input 6) Cylinder Length 150 inches (active fuel length per Design Input 4) Material Water Density 0.94 gram /cc (per Assumption 10) l I > scs 2un arv.o m @ETERENCE sold-XXIV-tH)
EC&FS DEPARTMENT ICCN NO. CALCULATION SHEET CCN CONVERSION: CCN NO. CCN - Project or DCP/FCN N/A Calc No. _N-04AG-003 uNect Fuel Transfer Tube Shieldir 2 _ _ . _ _ Sheet No .101_ of 117. REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 6 MARK DRUCCER 05-31-96 C. W. SAYLES 06-03-96 y i Body 3: Air About the Fuel Transfer Tube Body 3 is the air between the fuel transfer tube and the concrete shield. This body has the following characteristics: Geometry Cylinder (assumed for simplification, see following text) Cylinder Axis X (chosen for consistency with Body 1 model) Cylinder Radius 2 feet, 2 inches (determined in following text) Cylinder Length 150 inclies (active fuel length per Design Input 4) Material Air Density 1.1e-3 gram /cc (per Assumption 11) To simplify the SHIELD-SG input file, Body 3 is modeled as a cylinder. The 2'2" outer radius of this region is modeled to ensure equivalency between the modeled air thickness and the air thickness that exists between the fuel transfer tube and the concrete above the Penetration j } Building Elevation 15'0" conidor. Per Design Input 7 the floor of the fuel transfer tube enclosure is at plant elevation 26'4", and per Design Input 6 the centerline of the fuel transfer tube is at plant elevation 28'6" Therefore, the radius of this air shielding is 2'2" (i.e.,28'6" less 26'4"). Body 4: Concrete Shield Body 4 is the concrete shield between the fuel transfer tube and the Penetration Building Elevation 15'0" corridor below. This body has the following characteristics: Geometry Cylinder (assumed for simplification, see following text) Cylinder Axis X (chosen for consistency with Body 1 model) Cylinder Radius 22 feet, 2 inches (determined in following text) Cylinder Length 150 inches (active fuel length per Design Input 4) Material Concrete Densitv 2.35 gram /cc (per Assumption 12) To simplify the SHIELD-SG input file, Body 4 is modeled as a cylinder. The 22'2" outer radius of this region is modeled to arbitrarily place 20 feet of concrete shielding in this fuel transfer tube concrete enclosure. 1
--..----,.,m
EC&FS DEPARTMENT ICCN NO.
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CALCULATION SHEET CCN CONVERSION-CCN NO. CCN - . Project or DCP/FCN N/A Calc No. N 0.1R0.001 1 ubiect Fuel Tmnsfer Tu w Shieldir 2 --_ __ ___ _ _ - _ . _ _ _ _._ __ Sheet No. _ til2._ of .117 _ l ! REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05-31-96 C. W. SAYLES 06-03-96 y I
- VIIL3.2) SIIIELD-SG Dosepoints The SHIELD-SG dosepoints are positioned at various radial distances from the midpoint of the
, active fuel length. This modeling will maximize the dose rate contribution from the entire fuel l assembly length. The distances include dosepoint #3 which is positioned 4'10" into the concrete shield (Body 4). Due to the extent of concrete shielding between the source and this dosepoint, l this dosepoint is modeled with the concrete build-up factor coded into the SHIELD-SG Code l library. Detailed information on each dose point is provided in Table VIII.3-1. i 1 Table VID.3-1 SHIELD-SG CODE DOSEPOINTS 1 X Y Z Buildup I Desepoint *** PI"." Coordinate Coordinate Coordinate Material l 1 6' 3" 4.51" 0'0" Uranium Outer surface of a single fuel assembly 2 6' 3" l'6" O'0" Water Inner surface of the fuel transfer tube l I 3 6'3" 7' 0" 0'0" Concrete 4' 10" into the concrete enclosure 4 6' 3" 2' 2" 0' 0" Air Inner surface of the concrete enclosure 1 5 6' 3" 2' 8" O'0" Concrete 0.5 feet into the concrete enclosure l 6 6' 3" 3' 2" 0'0" Concrete 1.0 feet into the concrete enclosure 7 6' 3" 3' 8" 0' 0" Concrete 1.5 feet into the concrete enclosure 8 6' 3" 4' 2" 0'0" Concrete 2.0 feet into the concrete enclosure 9 6' 3" 4' 8" O' 0" Concrete 2.5 feet into the concrete enclosure 10 6' 3" 5' 2" O' 0" Concrete 3.0 feet into the concrete enclosure 11 6' 3" 5' 8" 0'0" Concrete 3.5 feet into the concrete enclosure 12 6' 3" 6'2" O' 0" Concrete 4.0 feet into the concrete enclosure 13 6' 3" 6' 8" 0' 0" Concrete 4.5 feet into the concrete enclosure 14 6' 3" 7' 2" O'0" Concrete 5.0 feet into the concrete enclosure 15 6' 3" 7' 8" O' 0" Concrete 5.5 feet into the concrete enclosure 16 6' 3" 8' 2" 0' 0" Concrete 6.0 feet into the concrete enclosure 17 6'3" 8' 8" 0'0" Concrete 6.5 feet into the concrete enclosure 18 6' 3" 9' 2" O'0" Concrete 7.0 feet into de concrete enclosure l k SCE 26 026 REV 0 tu l REFERENCE SOID-XXIV.715]
EC&FS DEPARTMENT ICCN O.
""""~
CALCULATION SHEET CCN CONVERSION: CCN NO. CCN - Project or DCP/FCN N/A Calc No. N.O.1An 001 Gubject Fuel Transfer _TtneShieldirm .. _ _ _
. Sheet No _101 of_117 .
REV ORIGINATOR DATE 1RE DATE REV ORIGINATOR DATE l IRE DATE R 4 MARK DRUCKER 05 31-96 C. W. SAYLES 06 % h i VIIL3.3) SHIELD-SG Results l l This section and its Table VIII.3-2 summarize the resultant dose rates as determined by the SHIELD-SG code run based on the input geometry and parameters specified in the preceding subsections.1 o 'HIELD-SG Code has been executed on the Nuclear Fuel Management IBM-RISC 60V aorkstation. Section IX.2 of this calculation presents the input and output files associated with the SHIELD-SG Code analysis. l Table VIII.3-2 presents not only the dose rate results due to a single fuel assembly, but also the dose rates attributed to two fuel assemblies being present side-by-side within the fuel transfer tube. These latter doses are equivalent to twice the dose rate due to the single fuel assembly. Per Design Input 7 the Penetration Building Elevation 15'0" corridor below the fuel transfer tube is separated from the fuel transfer tube by 4'10" of concrete shielding. Therefore, per Table VIII.3-2 Dosepoint 4 the dose rate in the Penetration Building Elevation 15'0" corridor ! ( } below the fuel transfer tube would be 4.03 millirem / hour if one fuel assembly were in transit through the fuel transfer tube, and 8.06 millirem / hour if two fuel assemblies were in transit side-by-side through the fuel transfer tube. Per Design Input 8 the Penetration Building Elevation 15'0" corridor above the fuel transfer tube is separated from the fuel transfer tube by 5'3" of concrete shielding. Therefore, per Table VIII.3-2 Dosepoint 14 the dose rate in the Penetration Building Elevation 15'0" corridor above the fuel transfer tube would be no greater than 2.41 millirem / hour if one fuel assembly were in transit through the fuel transfer tube, and 4.82 millirem / hour if two fuel assemblies were in transit side-by-side through the fuel transfer tube. Of note are the relatively high dose rates adjacent to the fuel transfer tube without the beneSt of concrete shielding (Dosepoints 2 and 4). I I
$CE 26426 l LEY 0 8,94 [ REFERENCE Sold-XXIV ? 15]
EC&FS DEPARTMENT ICCN 3 CALCULATION SHEET CCN CONVERSION-CCN NO. CCN - n Project or DCP/FCN N/A Calc No. N-04AO 001 ( \ C ?ubiect Fuel Transfer.Tn w Shieldir 2 -_ _ . _ . _ _ _ _ _ _ _ _ _ _ - - - _ _ . . _ _ _ _ _ . . - _ _ . Sheet No _104_ of_117 _ REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05-31-96 C. W. SAYLES 06 % h 1 I Table VIII.3-2
; SHIELD-SG CODE OUTPUT DOSE RATES a
Dose Rate with Dose Rate with l
. . One Assemblyin the Two Assemblies in the Dosepoint Description Fuel Transfer Tube FuelTransfer Tube l (rem / hour) (rem / hour) 1 Outer surface of a single fuel assembly 9.31e%6 1.86e%7 j 2 Inner surface of the fuel transfer tube 2.42e+05 4.84e+05
, 3 4' 10"into the concrete enclosure 4.03e-03 8.06e-03 4 Inner surface of the concrete enclosure 1.65e+05 3.30e+05 5 0.5 feet into the concrete enclosure 1.57e+04 3.14e%4 6 1.0 feet into the concrete enclosure 1.97e+03 3.94e+03 j l 7 1.5 feet into the concrete enclosure 2.83e+02 5.66e+02 8 2.0 feet into the concrete enclosure 4.54e+01 9.08e+01 l 9 2.5 feet into the concrete enclosure 7.90e@0 1.58e+01 10 3.0 feet into the concrete enclosure 1.46e+00 2.92e+00 11 3.5 feet into the concrete enclosure 2.80e-01 5.60e-01 12 4.0 feet into the concrete enclosure 5.56e-02 1.11e-01 13 4.5 feet into the concrete enclosure 1.14e-02 2.28e-02 i 14 5.0 feet into the concrete enclosure 2.41e-03 4.82e-03
- 15 5.5 feet into the concrete enclosure 5.22e-04 1.04e-03 16 6.0 feet into the concrete enclosure 1.16e-04 2.32e-04 17 6.5 feet into the concrete enclosure 2.65e-05 5.30e-05 1
18 7.0 feet into the concrete enclosure 6.21e-06 1.24e-05 i scs 2wa arv o sw (RmKENCE SOID-XXIVd15]
EC&FS DEPARTMENT tcCN R CALCULATION SHEET CCN CONVERSION-CCN NO. CCN - _ Project or DCP/FCN N/A Calc No. N.0.iRO 001 y ubiect Fuct Transfer Tu mShieldir a _ _ _ Sheet No _105_ of 117_ REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05 % c. W. SAYLES 06 % y I LX. COMPUTER CODE INPUT AND OUTPUT FILES IX.1 SOURCE 2 Code Run One SOURCE 2 Code mn is required for this calculation. The SOURCE 2 Code analysis was executed on the Nuclear Fuel Management IBM-RISC 6000 workstation. The SOURCE 2 executable used in this analysis was titled and stored as: l l
/scenuc/nfa/ bin / source 2d2-5 created: 10/11/95 at 1:18 pm This section presents the following SOURCE 2 Code input and output files employed in this calculation:
3d-asmby.dat RISC 6000 file size: 2461 bytes created: 05/20/96 at 10:50 am 3d-asmby.out RISC 6000 file size: 10567 bytes created: 05/20/96 at 11:28 am I I NOTE: The SOURCE 2 Code input file enters the input activity data in units of curies / core, and the multiplier in units of core /cm$ assembly. These data units are in disagreement with the default units that print in the first pages of the SOURCE 2 Code output file. However, when the input activity data and multiplier are combined, the resultant units for a fuel assembly are curies /cc, which is in agreement with later pages of the output file. I I _ . . - ,e _ ,w
, EC&FS DEPARTMENT ICCN O.
""""~ PAGE _ 0F _
4 CALCULATION SHEET CCN COMERSIOR CCN NO. CCN . l
'ect or DCP/FCN N/A Cale No. N.O.1AO 003 l l
t ubiect Fuel Transfer Tine Shielding . _ ___ _ __ . . _ _ _ _ _ _ _ _ _ - . _ . _. _ _ _. Sheet No .106 of. _117..___ l REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R E 4 MARK DRUCKER 05 % C. W. SAYLES 06-03-96 y I IX.1.1) SOURCE 2 Code Input File (3d-asmby.dat) SOUR N0480-003 4 MARK DRUCKER SONGS 2,3 FUEL TRANSFER TUBE SHIELDING SINGLE ASSEMBLY SOURCE STRENGTHS (60 GWD/T, 1.2 RFF, 1. 2 5 AF F) i 53 11 0 0 0 1 4.3700E-08 .00000E+00 .00000E+00 i 72.000 i HRS BR--84 2.85000E+07 . BR--85 3.99000E+07 j KR-85M 3.98000E+07 KR--85 1.75000E+06
- KR--87 6.47000E+07 KR--88 9.48000E+07 RB--88 1.10000E+00 .
KR--89 1.41000E+08 RB--89 1.46000E+08 SR--89 1.26000E+08 SR--90 1.67000E+07 Y---90 9.22000E+06 SR--91 1.78000E+08 Y--91M 1.05000E+08 Y---91 3.58000E+08 Y---95 1.87000E+08 l l ZR--95 1.78000E+08 NB--95 1.81000E+08 MO--99 1.89000E+08 TC-99M 2.27000E+07 RU-103 1.01000E+08 RU-106 1.15000E+07 TE129M 1.07000E+07 TE-129 3.29000E+07 I--129 4.16000E+00 I--131 8.96000E+07 XE131M 6.16000E+05 TE-132 1.32000E+08 I--132 1.32000E+08 TE133M 1.07000E+08 TE-133 1.12000E+08 I--133 2.02000E+08 XE-133 1.93000E+08 CS-134 2.58000E+06 TE-134 2.12000E+08 I--134 2.39000E+08 I--135 1.85000E+08 XE135M 5.51000E+07 . XE-135 5.08000E+07 CS-135 2.30000E+01 CS-136 1.78000E+05 XE-137 1.80000E+08 05-137 7.89000E+06 XE-139 1.79000E+08 CS-138 2.04000E+08 CS-140 1.81000E+08 LA-140 1. 93000E+ 08 BA-143 1.60000E+08 LA-143 1.81000E+08 CE-143 1.81000E+08 FR-143 1.81000E+08 g CE-144 1.39000E+08 F FR-144 1.26000E+08 5 SCE 2H26 REV 0 EH (MTERENCE 301& XXIV M5]
EC&FS DEPARTMENT ICCN NO.
"""""* "" - "-~~
CALCULAT10N SHEET CCN CONVERSION: CCN NO. CCN - Project or DCP/FCN N/A Calc No. N MAO-001
.ubiect Fuel Transfer Tine Shieldir o.__.. _ _
..___ __.__ _ _ _ Sheet No _107. of__117.._
REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 E MARK DRUCKER 05-31-% c. W. SAYLES 06 % y I IX.1.2) SOURCE 2 Code Input File (3d-asmby.out) NE602-SOURCE 2(D2-5) [ ( C11972,1991 BECHTEL. SCE AIX VERS. OCT 95] ORIGINATOR: MARK DRUCKER DATE: 05/20/96 CALC: N0480-003 FIV: 4 PROJECT: SONGS 2,3 JOB: CHECK: DATE: / /
SUBJECT:
FUEL TRANSTER TUBE SHIELDING PAGE: 1 CASE TITLE: SINGLE ASSEMBLY SOURCE STRENGTHS [60 GWD/T, 1.2 RPF, 1.25 AFF) MULTIPLICATION FACTOR = 4.3700E-08 ISOTOPE INPUT ACTIVITY (CURIES /CC ) BR--84 2.85000E+07 BR--85 3.99000E+07 KR-85M 3.98000E+07 KR--85 1.75000E+06 KR--87 6.47000E+07 KR--88 9.48000E+07 RB--88 1.10000E+08 KR--89 1.41000E+08 RB--89 1.46000E+08 SR--89 1.26000E+08 SR--90 1.67000E+07 j Y---90 9.22000E+06 SR--91 1.78000E+08 Y--91M 1.05000E+08 Y---91 3'.58000E+08 Y---95 1.87n00E+08 ZR--95 1.78000E+08 NB--95 1.81000E+08 MO--99 1.89000E+08 TC-99M 2.27000E+07 RU-103 1.01000E+08 RU-106 1.15000E+07 TE129M 1.07000E+07 TE-129 3.29000E+07 I--129 4.16000E+00 I--131 8.96000E+07 XE131M 6.16000E+05 TE-132 1.32000E+08 I--132 1.32000E+08 TE133M 1.07000E+08 TE-133 1.12000E+08 I--133 2.02000E+08 XE-133 1.93000E+00 CS-134 2.58000E+06 TE-134 2.12000E+08 I--134 2.39000E+08 I--135 1.85000E+08 XE135M 5.51000E+07 XE-135 5.08000E+07 CS-135 2.30000E+01 CS-136 1.78000E+05 XE-137 1.00000E+08 CS-137 7.89000E+06 71-138 1.79000E+08 CS-138 2.04000E+08 CS-140 l'.81000E+08 a n m . mmm a ~xw.n,,
EC&FS DEPARTMENT ICCN NO. PREUNLCCN D~ PAGE _0F CALCULATION SHEET CCN CONVERSION: l CCN NO. CCN . Project or DCP/FCN N/A Calc No. N 0. TAO 001L D Subiect Fuel Tramfer Tu x Shieldir2 . --_ _ __ ____. ,
. Sheet No _108_ of_117 !
REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05-31-96 C. W. SAYLES 06 % y i NE602-SOURCE 2(D2-5) [ (C)1972,1991 BECHTEL. SCE AIX VERS. OCT 95) ORIGINATOR: MARK DRUCKER CATE: 05/20/96 CALC: N0480-003 REV: 4 FROJECT: SONGS 2,3 JOB: CHECK: DATE: / /
SUBJECT:
FUEL TRANSFER TUBE SHIELDING PAGE- 2 CASE TITLE: SINGLE ASSEMBLY SOURCE STRENGTHS (60 GWD/T, 1.2 RPF, 1.25 APF] MULTIPLICATION FACTOR = 4.3700E-08 ISOTOPE INPUT ACTIVITY (CURIES /CC ) LA-140 1.9300CE+08 BA-143 1.60000E+08 LA-143 1.81000E+08 CE-143 1.8100CE+08 PR-143 1.81000E+08 CE-144 1.39000E+08 l PR-144 1.26000E+08 TOTAL 5.99893E+09 1 l I i
-. - . - n-x . ,n
EC&FS DEPARTMENT ICCN[3.
"""**""~ ~ '
CALCULATION SHEET CCN CONVERSION-l l CCN NO. CCN - Project or DCP/FCN N/A Calc NO. NJMAO 001 l ubiect Fuel Transfer. Tine Shieldir e ---. - _ - _ . . - - . _ _ _ _ - _ . . ._ __ __ .___ -_ Sheet No 109. of_ _117_. l REV ORIGINATOR DATE IRE DATE REV ORIGINATOR IRE DATE DATE R ! E 4 MARK DRUCKER 05-31 96 c. W. SAYLES 06 % y I NE602-SOURCE 2(D2-5) ( (C)1972,1991 BECHTEL. SCE AIX VERS. OCT 95) ORIGINATOR: MARK DRUCKER DATE: 05/20/46 CALC: N0480-003 REV: 4 PROJECT: SONGS 2,3 JOB: CHECK: DATE: / /
SUBJECT:
FUEL TRANSFER TUBE SHIELDING PAGE: 3 CASE TITLE: SINGLE ASSEMBLY SOURCE STRENGTHS (60 GWD/T, 1.2 RPF, 1.25 AFFJ ACTIVITY AS A FUNCTION OF DECAY TIME (CURIES /CC ) ISOTOPE .000 SEC 72.000 HRS
............ _-_...... _ ... ==. - - - _
BR--84 1.24545E+00 0.00000E+00 BR--85 1.74363E+00 0.00000E+00 KR-85M 1.73926E+00 2.55886E-05 KR--85 7.64750E-02 7.64519E-02 KR--87 2.82739E+00 2.58775E-17 KR--88 4.14276E+00 1.09771E-07 RB--88 4.80700E+00 1.22459E-07 KR--89 6.16170E+00 0.00000E+00 RB--89 6.3802CE+00 0.00000E+00 SR--89 5.50620E+00 5.28587E+00 SR--90 7.29790E-01 7.29646E-01 Y---90 4.02914E-01 5.79730E-01 SR--91 7.7786CE+00 4.02784E-02 Y--91M 4.58850E+00 2.51573E-02 Y---91 1.56446E+01 1.51519E+01 Y---95 8.1719CE+00 0.00000E+00 ZR--95 7.7766CE+00 7.53079E+00 NB--95 7. 90 97 0E+ 00 7.89187E+00 MO--99 8.25930E+00 3.87895E+00 TC-99M 9.9199CE-01 3.71036E+00 RU-103 4.41370E+00 4.18656E+00 RU-106 5.02550E-01 4.99707E-01 TE129M 4.6759CE-01 4.3947CE-01 TE-129 1.43773E+00 2.77266E-01 I--129 1. 817 92E-07 1.81905E-07 I--131 3.91552E+00 3.02335E+00 XE131M 2.69192E-02 2.86823E-02 TE-132 5.76840E+00 3.04758E+00 I--132 5.76840E400 3.13931E+00 TE133M 4.67590E+00 1.59020E-23 TE-133 4.89440E+00 3.48695E-24 I--133 8.82740E+00 8.35388E-01 XE-133 8.43410E+00 6.71088E+00 CS-134 1.12746E-01 1.12435E-01 TE-134 9.26440E+00 7.27303E-31 I--134 1.04443E+01 8.00733E-24 I--135 8.00450E+00 4.25895E-03 XE135M 2.40787E+00 6.87257E-04 XE-135 2.21996E+00 8.78600E-02 CS-135 1.00510E-06 1.00794E-06 CS-136 7.7786CE-03 6.62899E-03 XE-137 7.8660CE+00 0.00000E+00 CS-137 3.44793E-01 3.44730E-01 XE-138 7.8223CE+00 0.00000E+00 CS-138 8. 914 8 0E+ 00 0.00000E+00 sCE 26 426 REV.0 894 [tEFERENCE SO1D XXIV.M5)
_ . - . . _ = EC&FS DEPARTMENT ICCN NO. PR"" *C" " ~ PAGE _ 0F _ CALCULATION SHEET CCN CONVERSION. CCN NO. CCN . Project or DCP/FCN N/A Calc No. N-04AO oM j abiect Fuel Transfer Tu w Shieldir 2 -- ._- ._ - . . _ _ _ . _ - - _ . - _ . . _ _ _ _ _ _ _ _ _ _ _ Sheet No .110. of_117_ I REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R E 4 MARK DRUCKER 05 % C. W. SAYLES 06-03-96 y I NE602-SOURCE 2(D2-5) ( ( C) 197 2,1991 BECHTEL. SCE AIX VERS. OCT 95] ORIGINATOR: MARK DRUCKER DATE: 05/20/96 CALC: N0480-003 REV: 4 PROJECT: SONGS 2,3 JOB: CHECK: DATE: / /
SUBJECT:
FUEL TRANSFER TUBE SHIZLDING PAGE: 4 CASE TITLE: SINGLE ASSEMBLY SOURCE STRENGTHS [60 GWD/T, 1.2 RPF, 1.25 APF] ACTIVITY AS A FUNCTION OF DECAY TIME (CURIES /CC ) ISOTOPE .000 SEC 72.000 HRS CS-140 7. 90 97 0E+00 0.00000E+00 LA-140 8.43410E+00 2.44337E+00 BA-143 6.99200E+00 0.00000E+00 LA-143 7.90970E+00 0.00000E+00 CE-143 7. 90 97 0E+00 1.75645E+00 PR-143 7.90970E+00 7.35952E+00 CE-144 6.07430E+00 6.03007E+00 PR-144 5.50620E+00 6.03033E+00 ND-144 0.00000E+00 1.64063E-17 PR144M 0.00000E+00 9.04527E-02 BA-140 0.00000E+00 3.98552E-04 BA137M 0.00000E+00 3.26114E-01 BA136M 0.00000E+00 9.94349E-04 XE133M 0.00000E+00 5.13290E-02 I-133M 0.00000E+00 1.32345E-24 RH-106 0.00000E+00 4.99708E-01 RH103M 0.00000E+00 4.18233E+00 TC--99 0.00000E+00 1.15566E-07 NB-95M 0.00000E+00 3.01241E-02 RB--87 0.00000E+00 8.69680E-15 TOTAL 2.62153E+02 9.64470E+01 SCE 2H26 REY. 0 EH (REITJtENCE. SO123-XXIV 7.15)
EC&FS DEPARTMENT ICCN O.
"""""3' PAGE _ 0F _
I CALCULATION SHEET CCN CONVERSION-CCN NO. CCN . Project or DCP/FCN N/A Calc No. N-0.iRG 001 I
.ubiect Fud Transfer Tine Shieldir m _ _ __ . _ _ ___ _ _ _ . _ _ _ _
_ _ _ _ _ _ _ _ . _ Sheet No _111 of 117. REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R ~ E 4 MARK DRUCKER 05-31-96 C. W. SAYLES 06 % y I NE602-SOURCE 2(D2-5) ( ( C) 1972,1991 BECHTEL. SCE AIX VERS. OCT 951 l ORIGINATOR: MARK DRUCKER DATE: 05/20/96 CALC: N0480-003 REV: 4 PROJECT: SONGS 2,3 JOB: CHECK: DATE: / /
SUBJECT:
FUEL TRANSTER TUBE SHIELDING PAGE: 5 l CASE TITLE: SINGLE ASSEMBLY SOURCE STRENGTHS [60 GWO/T, 1.2 RPF, 1.25 APF) i GAMMA SOURCE STRENGTH AS A FUCTION OF DECAY TIME (MEV/CC-SEC ) i .......... __.._...____...................._____. ..... a ENERGY GROUP .000 SEC 72.000 HRS ! .000 - .100 MEV 2.06726E+10 9.33325E+09
; .100 - .400 MEV 4.68643E+11 9.99057E+10
.400 - .900 MEV 3.56896E+12 9.55036E+11 1
.900 - 1.350 MEV 1.85334E+12 4.91705E+10
, 1.350 - 1.800 MEV 1.52569E+12 1.62602E+11
! 1.800 - 2.200 MEV 7.10533E+11 1.10273E+10 j 2.200 - 2.600 MEV 6.40011E+11 1.17234E+10 1 2.600 - 3.000 MEV 2.09668E+11 1.71322E+08 j 3.000 - 5.000 MEV 3.52861E+11 8.70632E+07 4
5.000 -15.000 MEV 2.20323E+08 0.00000E+00 TOTAL 9.35061E+12 1.29906E+12 sCE 26-426 REV 0 M4 [ REFERENCE Sol 23. XXIV-715)
EC&FS DEPARTMENT ICCN NO. CALCULATION SHEET CCN CONVERSION-CCN No. CCN -
, Project or DCP/FCN N/A Cale No. N-04AO-001
;ubiect Fuel Transfer Tu a RhleMir a _
_ Sheet No .112_ of_117_
~
REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R 4 MARK DRUCKER 05 % C. W. SAYLES 06 % h . l 3 l l IX.2 SHIELD-SG Code Run One SHIELD-SG Code run is required for this calculation. The SHIELD-SG Code analysis was executed on the Nuclear Fuel Management IBM-RISC 6000 workstation. The SHIELD-SG executable used in this analysis was titled and stored as:
/scenuc/nfa/ bin /shieldsgd2-10 created: 10/11/95 at 1:11 pm l
This section presents the following SHIELD-SG Code input and output files employed in this calculation: xfr-tube.in RISC 6000 file size: 2284 bytes created: 05/20/96 at 10:45 am xfr-tube.out RISC 6000 file size: 8395 bytes created: 05/20/96 at 12:20 pm l t I
- ,-. . .,. unaa -n.,a
1 EC&FS DEPARTMENT ICCN O. PREUM. CCN Q~ IAGI - - 0F CALCULATION SHEET CCN CONVERSION: CCN fJ0. CCN. Project or DCP/FCN N/A Cale No. N.naAO-001 abiectTuel Transfer _Tu m Rhieldir e . _ . _ . __ _ _ _ _ . _ _ . ._ _ _ _ _ _ _ _ _ __ _ _ _ Sheet No _ lil_ of _117_ REV ORIGINATOR DATE IRE DATE lREV ORIGINATOR DATE IRE DATE R E 6 MARK DRUCKER 05 % C. W. SAYLES 06-03-96 y i IX.2.1) SHIELD-SG Code Input File (xfr-tube.in) 1 <JDEV xfr-tube.out 0 <JSCR 1 <JOPT 0 <JSUM 1 <JENT N0400-003 4 MARK DRUCKER SONGS 2,3 FUEL TRANSFER TUBE SHIELDING 10 <NUM 0.10 0.40 0.80 1.30 1.70 2.18 2.50 2.80 4.00 6.20 1 <IUN 1 <ICO 1.00E-02 <EPS 0 <NENT 0 <NBTE 5 <NSEG 1 < DOSE SEGMENT # 1
- 6. 3.000 0, 0.451 0. .000
{ } ASSEMBLY OUTER SURFACE 1 < DOSE SEGMENT # 2
- 6. 3.000 1. 6.000 0, .000 WATER TRANSFER TUBE SURFACE 1 < DOSE SEGMENT # 3
- 6. 3.000 7 0.000 0. .000 CONCFITE 4'10" INTO CONC SHIELD 1 < DOSE SEGMENT # 4
- 6. 3.000 2. 2.000 C. .000 AIR ,
ENCLOSFI INNER SURFACE 14 < DOSE SEGMENT # 5
- 6. 3.000 9. 2.000 0. .000 CONCRETE !
DISTANCE INTO SHIELD
< CASE # 1 EVALUATION OF DOSE RATE DUE TO A SINGLE FUEL ASSEMBLY 9.33E+09 9.99E+10 9.55E+11 4.92E+10 1.63E+11 1.10E+10 1.17E+10 1.71E+08 8.71E+07 0.00E+00 1.00E+00 <C2 4 <NBB FUEL ASSEMBLY < BODY # 1 1
INT CYL X
- 0. .000 0. .000 0. .000
- 12. 6.000 C. 4.530 PWR-CORE 4.35E+00 WATER IN TUBE < BODY # 2 f EXT CYL X
sCE 2H26 REV 0194 [HFERENCE 50123 XX!Y.M5]
EC&FS DEPARTMENT ICCN NO.
""""" PAGE _ 0F CALCULATION SHEET CCN CONVERSION-l CCN No. CCN .
Project or DCP/FCN N/A Calc No. N MAMIM hubiect Pnel Transfer Tu n Shieldir1__ . _. - .__ . . _ _ . __ _. _ . _ _ _ . Sheet No 114 of 117_ j REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R I E 4 MARK DRUCKER 05-31-96 C. W. SAYLES 06 % y I
- 0. .000 0. .000 0. .000
- 12. 6.000 1. 6.000 WATER 0.94E+00 AIR ABOUT TUBE < BODY # 3 3
EXT CYL X
- 0. .000 0. .000 0. .000
- 12. 6.000 2. 2.000 AIR 1.10E-03 CONCRETE SHIELD < BODY # 4 4
EXT CYL X
- 0. .000 0. .000 C. .000
- 12. 6.000 22, 2.000 CONCFITE 2.35E+00 l
9 i sca 2W26 REY.0 kH [REFTRENCE sot 2FXXIV4Dj \
r EC&FS DEPARTMENT :::CN NO.
"""""~ PAGE ~ OF -
CALCULATION SHEET CCN CONVERSION: CCN NO. CCN . ; Project or DCP/FCN N/A Calc No. N-0.iAO001
.ubiect Fuel Transfer Tube ShieMir2 __ _. _ . . . _ . . ___ _ Sheet No .115.. of 117.
REV ORIGINATOR DATE 1RE DATE REV ORICINATOR DATE IRE DATE R E 4 MARK DRUCKER 05-31-96 C. W. SAYLES 06-03-96 y I IX.2.2) SHIELD-SG Code Input File (xfr-tube.out) NE650 SHIELD-SG(D2-10) [(C) 1983,1992 BECHTEL. SCE AIX VER., OCT 95 xfr-tube.out ORIGINATOR: MARK DRUCKER DATE: 05/20/96 CALC: N0480-003 REV: 4 PROJECT: SONGS 2,3 JOB: CHECK: DATE: / / CALC
SUBJECT:
WEL TRANSFER TUBE SHIELDING PAGE: 1 CASE TITLE: EVALUATION OF DOSE RATE' DUE TO A SINGLE WEL ASSEMBLY
*" SOURCE ENERGY GROUPS AND EMISSION RATES "* [ SOURCE MULTIFLIER = 1.00CE+00)
GROUP 4 1 2 3 4 5
......... ...._... ........ _....... ..._. __ .._.___. i MEV/ GAMMA .100 .400 .800 1.300 1.700 MEV/CM3-S 9. 3 3E+09 9. 9 9E+10 9. 55E+11 4. 92E+10 1. 63E+11 GROUP # 6 7 8 9 10 MEV/ GAMMA 2.180 2.500 2.800 4.000 6.200 MEV/CM3-S 1.10E+10 1.17E+10 1.71E+08 8.71E+07 0.00E+C0
* *
- BODY DATA * "
BODY COORDINATES X Y Z { } BODY BODY AX MU/ RHO DENSITY -------------- -------------- --------------
# TYPE LOC IS MATERIAL [G/CM3] [FT IN) (ET IN) [FT IN) 1 CYL SRC X FWR-CORE 4.35E+00 1 0 .000] ! 0 .000] [ 0 .000}
2 CYL EXT X WATER 9.40E-01 ( 0 .000) { 0 .000] [ 0 .000) 3 CYL EXT X AIR 1.10E-03 [ 0 .000) 1 0 .000) [ 0 .000) 4 CYL EXT X CONCRETE 2.35E+00 1 0 .000} [ 0 .000} [ 0 .000) BCDY DIMENSIONS DX DY DZ R BODY -------------- -------------- -------------- -------------- BODY
# (FT IN) [FT IN) {fi IN) [ET IN) DESCRIPTION !
I 2 [ 12 6.000) - - - - ( 1 6.000] WATER IN TUBE j 3 ( 12 6.000} - - - - [ 2 2.000) AIR ABOUT TUBE i 4 [ 12 6.000) - - - - [ 22 2.000) CONCRETE SHIELD I
*** DOSE FOINT DATA "*
DOSE POINT COORDINATES XP YP ZF DOSE BUILDUE -------------- -------------- -------------- DOSE POINT POINT MATERIAL [ F* IN) [FT IN] [FT IN} SET DESCRIPTION 1 URANIUM +[ 6 3.000] +{ 0 .4511 [ 0 .000} ASSEMBLY OUTER SURFACE 2 WATER +[ 6 3.000] +( 1 6.000] [ 0 .000) TRANSFER TUBE SURFACE 3 CONCRETE +[ 6 3.000] +1 7 .000] [ 0 .0001 4'10" INTO CONC SHIELD 4 AIR +[ 6 3.000] +[ 2 2.000) [ 0 .000) ENCLOSRE INNER SURFACE 5 CONCPITE + [ 6 3.000) +[ 2 8.0001 [ 0 . A >l DISTANCE INTO SH! ELD 6 CONCRETE +[ 6 3.000] +1 3 2.000} [ 0 .000] DISTANCE INTO SHIELD 7 CONCRETE +[ 6 3.000) +[ 3 8.000) [ 0 .000) DISTANCE INTO SHIELD 8 CONCRETE +[ 6 3.000] +[ 4 2.000] [ 0 .000] DISTANCE INTO SHIELD h l 1 l 1 scs26-426 uv o em (EFEMNCE SO123-XX!Y-715] s
EC&FS DEPARTMENT #CCN NO.
" ' ' " " ~ "-~
CALCULATION SHEET CCN CONVERSION-CCN NO. CCN . Project or DCP/FCN N/A Calc No. N 0.1An 001 hicct Fuel Transfer Tine Shieldir 2 -- - - _ - _ . _ _ - - - - - -_ -____ -_ _ sheet No 116.. of_.117._ REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R E
- 6 MARK DRUCKER 05-31-96 C. W. SAYLES 06 % y 4 I NE650 SHIELD-SG(D2-10) [(C) 1983,1992 BECHTEL. SCE AIX VER., OCT 95 xfr-tube.out ORIGINATOR
- MARK DRUCKER DATE: 05/20/96 CALC: N0480-003 REV: 4 PROJECT: SONGS 2,3 JOB: CHECK: DATE: / /
CALC
SUBJECT:
FUEL TPANSTER TUBE SHIELDING PAGE: 2 CASE TITLE: EVALUATION OF DOSE RATE DUE TO A SINCLE FUEL ASSEMBLY DOSE POINT COORDINATES XP YP ZP DOSi BUILDUP -------------- -------------- -------- --- DOSE POINT POINT MATERIAL [FT INI [ET IN) [FT IN] SET DESCRIPTION 9 CONCRETE +[ 6 3.000] +[ 4 8.000] [ 0 .000] DISTANCE INTO SHIELD 10 CONCRETE +[ 6 3.000) +[ 5 2.000] [ 0 .000) DISTANCE INTO SHIELD 11 CONCRETE +[ 6 3.000] +[ 5 8.000} [ 0 .000] DISTANCE INTO SHIELD 12 CONCRETE +[ 6 3.000] +[ 6 2.000] [ 0 .000) DISTANCE INTO SHIELD 13 CONCRETE *[ 6 3.000) +[ 6 8.000) [ 0 .000) DISTANCE INTO SHI4LD 14 CONCRETE +[ 6 3.000] +! 7 2.000) [ 0 .000) DISTANCE INTO SHIELD 15 CONCRETE +[ 6 3.000} +[ 7 8.003] [ 0 .000) DISTANCE INTO SHIELD 16 CONCRETE +[ 6 3.000] +[ 8 '2.000) [ 0 .000) DISTANCE INTO SHIELD 17 CONCRETE +[ 6 3.0001 +[ 8 8.000) [ 0 .000) DISTANCE I..TO SHIELD 18 CONCRETE +[ 6 3.000] +[ 9 2.000) [ 0 .000) DISTANCE INTO SHIELD
*** DOSE EQUIVALENT RATES [ REM /HR] [CONNERGENCE CRITERION = .010]
GRP DOSE POINT NUMBERS
# 1 2 3 4 L 6 i 1 5.83E+03 6.61E+01 9.11E-25 4.14E+01 2.12E-02 5.70E-05 2 3.46E+05 7.40E+03 7.99E-11 4.94E+03 1.37E+02 5.38E+00 3 6.74E+06 1.68E+05 4.13E-06 1.14E+05 8.24E+03 7.00E+02 4 4.46E+05 1.28E+04 3.98E-05 8.79E+03 1.09E+03 1.50E+02 5 1.56E+06 4.72E+04 1.32E-03 3.25E+04 5.16E+03 8.87E+02 6 1.05E+05 3.44E+03 6.44E-04 2.37E+03 4.63E+02 9.61E+01 7
1.10E+05 3.76E+03 1.81E-03 2.60E+03 5.60E+02 1.28E+02 8 1.57E+03 5.4eE+01 5.15E-05 3.00E+01 8.7EE+00 2.13E+00 9 7.04E+02 2.71E+01 1.60E-04 1.85E+01 5.16E+00 1.50E+00 10 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 TOTAL 9. 31E+ C6 2. 4 2E+ 05 4. 03E-03 1. 65E+ 05 1. 57E+ 04 1. 97E+ 0 3 GRP DOSE POINT NUMBERS
# 7 8 9 10 11 12 1 1.51E-07 3.94E-10 1.03E-12 2.69E-15 7.02E-18 1.83E-20 1 2 2.09E-01 8.07E-03 3.12E-04 1.20E-05 4.65E-07 1.80E-08 I 3 5.91E+01 4.97E+00 4.19E-01 3.53E-02 2.98E-03 2.52E-04 4 2.07E+01 2.85E+00 3.95E-01 5.47E-02 7.60E-03 1.06E-03 5 1.53E+02 2.63E+01 4.56E+00 7.93E-01 1.38E-01 2.41E-02 6 2.00E+01 4.20E+00 8.85E-01 1.87E-01 3.97E-02 8.45E-03 7 2.93E+01 6.76E+00 1.57E+00 3.66E-01 8.56E-02 2.01E-02 8 5.23E-01 1.29E-01 3.22E-02 8.05E-03 2.02E-03 5.09E-04 )
9 4.44E-01 1.33E-01 4.01E-02 1.22E-02 3.72E-03 1.14E-03 ' 10 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 TOTAL 2. 8 3E+ 02 4. 5 4 E+01 7. 90E+00 1. 4 6E+ 00 2. 80E-01 5. 56E-02 i i SCE 26 426 REY.0 W [ REFERENCE Sol 3XXIYJ 15]
c EC&FS DEPARTMENT iCCN O.
""""""' PAGE OF -
CALCULATION SHEET CCN CONVERSION-CCN NO. CCN . Project or DCP/FCN N/A Calc No. N-0. tan-001 ibiect Fuel Transfer Tu m Shieldir n _. __ _ _ _ _ _ _ _ _ __ _ _ Sheet Na _117 of _117__ REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R E 4 MARK DRUCKER 05-31-96 C. W. SAYLES 06 % y i NE650 SHIELD-SGID2-10) ((C) 1983,1992 BECHTEL. SCE AIX VER., OCT 95 xfr-tube.out ORIGINATOR: MARK DRUCKER DATE: 05/20/96 CALC: N0480-C01 REV: 4 PROJECT: SONGS 2,3 JOB: CHECK: DATE: / / CALC
SUBJECT:
FUEL TRANSFER TUBE SHIELDING EAGE: 3 CASE TITLE: EVALUATION OF DOSE RATE DUE TO A SINGLE FUEL ASSEMBLY GRP DOSE POINT NUMBERS
# 13 14 15 16 17 18 1 4.80E-23 1.25E-25 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2 6.97E-10 2.71E-11 1.05E-12 4.09E-14 1.59E-15 6.20E-17 3 2.14E-05 1.82E-06 1.54E-07 1.31E-08 1.12E-09 9.56E-11 4 1.48E-04 2.06E-05 2.89E-06 4.05E-07 5.69E-08 8.00E-09 5 4.23E-03 7.41E-04 1.30E-04 2.29E-05 4.04E-06 7.14E-07 6 1.80E-03 3.85E-04 8.25E-05 1.77E-05 3.81E-06 8.19E-07 7 4.73E-03 1.12E-03 2.64E-04 6.25E-05 1.48E-05 3.52E-06 8 1.29E-04 3.26E-05 8.27E-06 2.10E-06 5. 35E-07 1. 37E-07 9 3.51E-04 1.00E-04 3.35E-05 1.04E-05 3.23E-06 1.01E-06 10 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 TOTAL 1.14E-02 2.41E-03 5.22E-04 1.16E-04 2.65E-05 6.21E-06 1
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