ORNL-TM-1467.txt

BAARS

AT T

"OAK RIDGE NATIONAL LABO RATORY

g£. A, B

operated by

UNION CARBIDE_CORPORATION |

for the

 

 

N MARIETTA ENERGY SYSTEMS LIBRARIES

AT

3 445k 0251k43 b

 

 

 

 

 

 

 

nuc,ear potential and to identify

' 3 “TThe fi§“§”re

e_ufl“Q ! “—region3 two-fluid system

iuh “Tuel salt separated f%cm the blanket salt by graphite”
céfif?fi“%fi% Teactor T fffiid is trans-"

 

 

 

S

G reactor to'a supercrltlcal steam’cyéle;
“jggwmfluoride volfi%ility processing is employed, which' leads

 

_%Ww low unit processfhg‘costs and economic ‘reactor opera~
“tion asmmw%h@rmal breeder. —The resulting power cost is
' Ls/kwhr for investor-owned utili-

i¢ el Cycle cost 1s70.45 mill/kwhr(e)

_Lthe speciflc fissile inventory 1570.8 keg/Mw(e), and ‘the

m . Development of a Pa=
L région of the MSBR could
0 milis/kwhr(e), a fuel cycle

WT%???E?TL/RWhr(e), a specific fissile inventory

7vqghWW(e),”andra fuel doubling time of 13 years.

 

 

 

 

 

 

“docum enf contams mFormanon of a prehmmary nature
idge Natlona|

 

 

 
 

Thls reporf was prepured as an accounf of Government sponsored work Nexfher fhe Unlfed Sfofes,'

nor the Commission, nor any person acting on behdlf of the Comrrussnorl.

A. Makes any warranty or representation, expressed ot implied, with reSpecf to fhe .accyracy,
comple?eness, or usefulness ‘of the information coniamed in_this report, or that the use of
~any information, opparatus, method, or process disclosed in this report may not infringe

: prlvutely owned rights; or T I oo

B Assumes any liabilities with respect o the use of or for damages resulhng from fhe use of

S wa‘ny mformahon apparaius method, or process disclosed in this report. »

As used m.fhe above, person acting on behulf of the Commlssaon includes any employee or

contrccfor of the Commlssmn, or employee of such cent.rucfor, ta the extent fhaf such employeer

or contractor of the Commission, or emp]oyee of such contractor prepares, dlssemlna*es, or

provides access to, any information pursuant to his employment or contract with the Commission,

or his employment with such contractor.

 

 

 
caER L L

o e obRikg kil

e e e e

INTRA-LABORATORY CORRESPONDENCE

OAK RIDGE NATIONAL LABORATORY

 

April 7, 1966

To: Recipients of ORNL-TM~-1467

 

 

Report No. : ORNL-TM-1467 Classification: Unclassified

Author(s): ___P.R. Kasten, E.S. Bettis, H.F. Bauman, W.L. Carter, et al.

 

Subject: Summary of Molten-Salt Breeder Reactor Design Studies.

 

Request compliance with indicated action:

Please replace the table of contents on page 5 in your copy(ies) of the subject
report with the atfached. It has been prepared on gummed stock for your
convenience.

N.T. Bray, Supervisor

Laboratory Records Department
- Technical Information Division

Y _as.ck

M/

| UCN-430
(3 5.61)
 

 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
ek el i,

cpolay e b Lo MRS .

v e Sl B AR SRR e e i kel SRRSO, | Sl e M oo

AR, A - e

AL - e B e k- L BERLG. | SR ARG AR e Dactiec e MR W o bl T Nl ko ni bl e e B

*

FOREWORD

This memorandum is a partial summary of the molten-salt breeder
reactor studies which will be presented in a forthcoming ORNL report.
The purpose of the present memo is to provide results of these studies
prior to issue of the complete report.,

In utilizing these studies, it should-be emphasized that the cost
estimates tacitly assume the existence of an established industry.

   

 

 

MARTIN MARIETTA ENERGY SYSTEMS LIBRARI

IR Nlllll

} 3 HHEE uesmbua b

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
       

   

   

 

 

 

     

         

 

 

   

1
i
:
i
i
{

 

 

 

RS e L N AR b |

 

 
s s SRR B b L sl IR . R mgg%

s i i B e s o A b b B

cha IR B oo B GGl .

 

vl ae s il

 

e e E
-

 

»

o

 

FOREWORD . . .

~ INTRODUCTION . .

-

MSBR PLANT DESIGN .

‘Flowsheet .

Reactor Design .

Fuel‘Processing‘.

Heat Exchange and Steam Systems .
CAPITAL cosT ESTIMATES .
\;ReaEtorfPefié?R?iant .

Fuel Recycle Plant

f:Analy31s Procedures fE.
“““ Ba31c Assumptlons'.\;

Nuclear Des1gn Analy51s .

REFERENCES .

 

 

-

.

*

CONTENTS

*

-

>

NUCLEAR PERFORMANCE AND FUEL CYCLE ANALYSES

'POWER COST AND FUEL'UTILIZATION CHARACTERISTICS

O o= -1 =

15

15

19
19
19
19
23
23
25
28

30
 

 

3

 

 

 

 

 

  

 

 

A T T T T T YA T — YT T

v i - - S TSN T A T = — Ay R T T T Ty T
oo St TGN 5 D oIS OB ML . L L Lpkd

R i, bl

 

&

INTRODUCTION

Design and evaluation studies have been made of thermal molten-salt

~ breeder reactors (MSBR) in order to assess their economic and nuclear

potential and to identify the important design and development problems.
The reference reactor design presented here contains design problems -

related to molten»salt reactors in general.

The MSBR reference design concept is a two-region, two-fluid system,
with fuel salt separated from the blanket salt by graphite tubes. The
fuel salt consists of uranium fluoride dissolved in a mixture of lithium~
beryllium fluorides, while the blanket salt is a thorium-lithium fluoride
of eutectic composition (about 27 mole % thorium fluoride). The energy
generated in the reactor fluid is transferred to a secondary coolant-salt
circuit, which couples the reactor to a supercritical steam cycle. On-
site fluoride volatility processing is employed, leading to low unit
processing costs and economic operation as a thermal breeder reactor.

MSBR PLANT DESIGN
Flowsheet

Figure 1 gives the flowsheet of the 1000-Mw(e) MSBR power plant.
Fuel flows through the reactor at a rate of about 44,000 gpm (veloclty
of about 15 ft/sec), entering the core at 1000°F and leaving at 1300°F.,
The primary fuel circuit has four loops, each loop having a pump and &
primary heat exchanger. Each of these pumps has a capacity of about
11,000 gpm. The four blanket pumps and heat exchangers, although smaller,
are similar to corresponding components in the fuel system. The blanket

salt enters the reactor vessel at 1150°F and leaves at 1250°F. The

blanket salt pumps have a capa01ty of about 2000 gpm.

Four 1k OOO»gpm coolant pumps circulate the sodium fluoroborate
coolant salt, which enters the shell side of the primary heat exchanger
at 850°F and leaves at 1112°F, After leaving the primary heat exchanger,
the coolant salt is further heated to 1125°F on the shell side of the
blanket heat exchangers. The coolant then circulates through the shell
side of 16 once-through superheaters (four superheaters per pump). In
addition, four 2000-gpm pumps circulate a portion of the coolant through
eight reheaters° :

The steam system flowsheet is essentially that of the new TVA Bull

Run plant, with modifications to increase the rating to 1000 Mw(e) and to

preheat the working fluid to TOO°F prior to entering the heat exchanger—

superheater unit. A supercritical power couversion system is used, which

is appropriate for molten-salt application and takes advantage of the
high-strength struetural alloy employed. Use of a supercritical fluid
system results in an overall plant thermal efficiency of about hs%
Sk, S o R

WEACTCR VESSEL(U
2225 Mut

ST TRE T T BSKo4106

4 AEHEAT ST —
1518.5 1000 3500 — - ! FREVEATLRS 18/

1OQS5 Mui

. )

I $70p -650" l ’
{
i _
| #5¢8.5%-540p -1000*
I

| ~—3600p -100c*14240 raan- : 4

2515 p -10Q,

|
|
) r
A
I !
I I
I |
! | .r—;yc%c ‘866" P R 6FA.
e ’ is -
Fgféogkgrp’fi/’:zif:! -—-o—--—-fi—--—d——--—-r——---—fi—---—--é-——q—---
1 1 :
| t
| |
‘ I
' !
Lo
}
t
|

BANTET SAT PUMPS (4
2,00C gom (Nom/) sa.

E 5|

   
 

 

2272 Mwe
gross

 

 

 

 

 

i ' 295"
|

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   

 

 

 

 

 

 
      

 

 

 

 

!7—5 rase™
1
* BOILER - SUPERMEATER | |
COCLANT SALT P (4]
_HEaE J ! FEHEATER COCLANY . 4,007 gom (Nomi Ex. i
: Sac: pares (&1 : ' GEN.
I(. _Lc;‘ -“J“E_}:’. 1 2 qpm“mm-lm I “ o ' . o o L
l 1 . F3I00F &50°F 165 . J wfiu
{ g - ] ipder F4F
, I\' _._____qi-] i . . . ,\+ r . 13
A - s el A i = — 1
% nsE ! ,H.J, oo, 41
' ! i : l TESZA- IO -FOCT COMDENSER B FEEOWATER
S B ANKEY ST HEAT - . — —_— - ’ SYSTEMS
EXCHANGERS (4] . ’ | (See sream syssem Piowsheat)
. 1 Mwr( Teral) | |
LU b sowsm-sureorEarens
o/ J— ader 16
[ 1™ FUEL SALT MEAT e ! l 19313 Mw L 3500 p - S48 5* -5 46 38
A CHANGERS (4] : ;
T ) 14 Mt (Tolal} 293.5 Mw’ | l .
3 e — ._l
. ' '
\ BOLER FEEDWATER
! A N A SURE POOSTER PUMPS (2)
b
. N__yu730 698" Psn
1 20000 g8 (Nom.) Eace
4.5 Mwe Eoch
BAWET SALT QRAsK TANKS FUEL SALT ORAIN TANKS | oot AT ST oRam TS
2 | i ‘o
i .
PERFORMANCE
NET QuTRUT. LOOO Mwa
: GAOSS GEwERAT.ON L0349 wwe
LECEND ’ . | #FBOOSIER PuMPS 22 e
FuEt —— STATION AUXILIARIES 257 Mwe
BUANKET e = e .
COCLANT o = = — , AEACTOR HEAT INPUT 2223 Mot
e o= NET NEAT RATE 7600 Bresiwn
N —— -
- o an . NET EFFICIENCY C e
» oo,
» P .
. or
-0 Froure reive
. . o
Fig. 1. Molten Salt Breeder Reactor Flow Diagram (700°F Feedwater).
e »
 

Reactor Design

Figure 2 shows a plan view of the MSBR cell arrangement. The reactor
cell is surrounded by four shielded cells containing the superheaters and
reheater units; these cells can be individually isolated for maintenance.
The processing cell, located adjacent to the reactor, is divided into a
high-level and a low=-level activity area.

- Figure 3 shows an elevation view of the reactor and indicates the
position of equipment in the various cells. Figure 4, a plan view of
the reactor cell, shows the location of the reactor, pumps, and fuel and
blanket heat exchangers. Figure 5 is an elevation of the reactor cell.
The Hastelloy N reactor vessel has a side wall thickness of about 1-1/k-
in. and a head thickness of about 2-1/h in.; it is designed to operate
at 1200°F and 150 psi. The plenum chambers, with 1/4-in.-thick walls,
communicate with the external heat exchangers by concentric inlet-outlet
piping. The inner pipe has slip Jjoints to accommodate thermal expansion.
Bypass flow through these slip joints is about 1% of the total flow. As
indicated in Fig. 5, the heat exchangers are suspended from the top of
the cell and are located below the reactor. Each fuel pump has a free
fluid surface and a storage volume vhich permit rapid drainage of fuel
- fluid from the core upon loss of flow. In addition, the fuel salt can
be drained to the dump tanks when the reactor is shut down for an extended
time. The entire reactor cell is kept at high temperature, while cold
"fingers” and thermal insulation surround structural support members and
all special equipment which must be kept at relatively low temperatures.
The control rod drives are located above the core, and the control rods
are inserted into the central region of the core.

The reactor vessel, sbout 1k £t in diameter by about 15 £t high,
contains a 10-ft-diam core assenmbly composed of reentry-type graphite
fuel cells. The graphite tubes are attached to the two plemum chambers
at the bottom of the reactor with graphite-to-metal transition sleeves.
Fuel from the entrance plenum flows up fuel passages in the outer region
of the fuel cell and down through a single central passage to the exit
plenmum. The fuel flows from the exit plenum to the heat exchangers,
then to the pump and back to the reactor. A lml/2mftwthick molten-salt
planket plus a l/haftwthlck graphite reflector surround the core. The
blanket salt also permeates the interstices of the core lattice so
fertile material flows through the core without mixing with the fissile
fael salt.,

The MSBR requires structural 1ntegr1ty of the graphite fuel cell.
In order to reduce the effect of radiation damage, the fuel cells have
been made small to reduce the fast flux gradient across the graphite
wall. Also, the cells are anchored only at one end to permit axial move-
ment. The core volume has been made large in order to reduce the flux
level in the core. In addition, the reactor is designed to permit re-
placement of the entire graphlte core by remote means if required.

Figure 6 shows a cross section of a fuel cello Fuel fluid flows
upward through the small passages and downward through the large central
tae Spnd ol AEERL b B - R AL e

3

 

 

 

  
     
  
      
  
   

   

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

t T L] ~ ORNL DWG. 66-795
; " WASTE GAS M
‘m fl CELL o H
REHEWT STEAM - —- — - W <1
33 T e— (R S ) A
' ' . T PUMPS
0 MP STEAM ——Jt T b1 COOLANT SALT
HP. STEAM —T 7 FUEL HEAT £XCHANGER o 7
FEED Ho — i T ‘ T
i LP. STEAM —-——.m m : rl
" e L A Lot L2
4 L et LT e el R
T B T ;._.
'
20' 1
l | .
ne' o & - o VT R N s b 3
1 .
{r :
‘————-—0‘ el sl el .\ ...r.‘4
O OO ) -
20 8 REHEATERS : 16 SUPERHEATERS |- -]
l !é 8 _ & &8 -D : - S
: - = ——0— 60— N — 8 —8 o
} S — 2NN : : ;’ i
+ U e
BLANKET HEAT EXCHANGER '
LESSING " [oecontammnanol:]  sToraee [Uf| 26
. o foo
¥ CONTROL AREA
o :
b e e e — 18O —— - — e

 

 

Fig. 2. Molten Salt Breeder Reactor — Reactor and Steam Cells-Plan.

0T
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

» » - > -~ 8
_ f— CONTROL ROD DRIVE ORNL DWG. 66-793
. FUEL CIRCULATING PUMP—— / . '
. /
COOLANT SALT SUMPS —— / F—BLAMKET CRCULATING PUMP
CONTROL ROOM LEVEL ———— ),

=
’_l

 

 

ANALYTICAL LAB. LEVEL —

.. .
.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

- GROUND LEVEL —

 

 

 

. ~——SUPERHEATERS

Fig. 3. Molten Salt Breeder Reactor'— Reactor and Steam Cells-Elevation.
 

 

 

 

A e i e, ey Ecall i Sl il i, » bl At -das Bcile b AR BB, ke B . - D Bl et MELRRRERL M W opm e - e e s —_———
'
' E-SK-4091
....4’ - A . .
. - - - A A s A
- < - 4
. . : . . - ’ . . . . . . \
t ; . - -
ar ) ) p . ) : T . :
- -
a 4 ' - -
. * 1 . ' - .
'
g

BLANKET
HEAT EXCHANGER

m

A,

COOLANT SALT
TO SUPER HEATLRS]
AND REHEATERS

 

 

 

 

 

 

 

COOLANT SALT
RETURN LINE

   
 

STEAM GENERATOR CELL

   

 

 

BLANKET
REGION

 
   
 
 

CORE REGION

    

 

 

 

 

 

 

 

 

 

 

 

 

———HEAT EXCHANGER
SUPPORT STRUCTURE

 

 

 

 

 

 

 

 
  

 

 

J‘_/_ o | e . - T L T T L L.
-~ - . / ,’ . ; 4 ) . - '
< T S PRIVARY 1 AT EXCHANGER - 3G FT DIAMETER REACTOR CELL e T e

 

Fig. 4. Molten Salt Breeder Reactor — Reactor Cell-Plan View.

S

¢l
 

       
  

 

=
&
ORNL DWG. 66-79%
PUMP MOTOR : BLANKET PUMP MOTOR
ROD ORIVES
LONSTANT SUPPORT
HENGERS
/V_/;,,—-IO FT DIAMETER OFE
i
|
1
REACTCH VESSIL
;= FUEL ALY DISTRIEUT.ON
PLENL WS
Q FUEL DUMP TANK
WITH COOLING CONS:
FOR AFTER MEAT |
REMIVAL i

 

AL ANKET HENT EXCHANSFR

{
i
PRIMARY HEST EXCHANGER

 

|
i
i
!
{
i

 

 

o . . * :
.~ wl’ e ey . P Tt . ¢ . i
A~ — - —

Z-j--REACYOR CELL. HEATERS

Fig. 5. Molten Salt Breeder Reactor — Reactor Cell-Elevation.
e A MY

o o e e B o Bl M e+ B . ool dndobint s lihiemedtblinn B e A e S bl b ek e S e b

 

 

FUEL MSSAGE (UP) —

14

; |  B-5K-4092

MODERATOR  (GRAMNTE BLANKEY PASSAGE

FUEL PASSAGE {DOWN)

 

 

34 00 FUEL TuaE RATOR  HOLD DOWN NUT  (GRAPHITE)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

‘\
I ! .
P
{ Gt
&
REAC
} TR 3
1 ¥
—4 Al Ak <o . QJ
—
— [ ) - —p X — e
]{.. /t: = -l J
.- SPACER
,,//
ce Tl METAL YO GRAPHTE SLIP-JOINT
by - _/__——’"" -
] FUH [ .o —— METAL TO GRAPMTE BRAZED JOINT
4 ---[ e T
Ao
| m—BRAZED JOINT
[
N
1N |
+ = =
oA ~ r }k ~——FUEL MLET PLEKUM
. 1 . :
I } 1 + Y '
1 J T.,\' )
] A ! 1 . _
i =-e——--FUEL OUTLET PLENUM
i ‘ z ’
‘I’l . e el b W ——— e Y S 3
= ' :
%fi e —r
L T

 

I

Fig. 6. Molten Salt Bréeder Reactor Core Cell.
LR ., - iEEGEE

el

R, B i PR ey e DR e R Rl KK, L .

 

_’via the fuel and fertile stream storage tanks. ¥

  

 

15

passage. The outside diameter of a fuel cell tube is 3.5 in.; there

‘are 534 of these tubes spaced on a 4,8-in. triangular pitch. The tube

assemblies are surrounded by hexagonal blocks of moderator graphite

with blanket salt filling the interstices. The nominal core composition

is T5% graphite, 18% fuel salt, and Th blanket salt by volume.

A summary cf paremeter values chosen for the MSBR design is given
in Teble 1,

Fuel Processing

 

The prlmary obJectives of fuel processing are to purify and recycle
fissile and carrier components, and minimize fissile inventory while

“holding losses to a low value., The fluoride volatility-vacuum distilla-

tion process fulfllls these objectives through simple operatlons.

' The core fuel is convenlently processed by fluoride volatility and

| vacnum distillation. Blanket processing is accompllshed'by fluoride

volatillty alone, and the processing cycle time is short enough to main-
tain a very low concentration of fissile material. The effluent UFg is
absorbed by fuel salt and reduced to UF4 by treatment with hydrogen to
reconstitute a fuelusalt mixture of the desired composition, '

Moltennsalt reactors are inherently sulted to the de31gn of pro-
cessing facilities integral with the reactor plant; these facilities
require only a small amount of cell space adjacent to the reactor cell.
Because all services and equipment available to the reactor are avallable
to the processing plant and shipping and storage charges are eliminated,
integral processing facilities permit signlficant savings in capital and
operating costs. Also, the processing pleant inventory of fissile material
is greatly reduced, resulting in low fuel inventory charges and improved
fuel utillzatlon characterlstlcs for the reactor,

Fod o neghy
i g

The prlncipal steps in core and blanket stream proce331ng of the

MSBR are shown in Fig. 7. A small side stréam of each fluid is continu-

ously withdrewn “from the fuel and blanket circulating loops and circulated

through the process1ng syetem, After processing, the decontaminated

fluids are returned to the reactor at some convenient point--for example,

  
 

Fuel invent esing plant are estimated to

:tif“be about 10% of the reactor system inventory for core processing, and
_less ‘than 1% for blanket processing.

 

terial for all components contacted by molten selt

"in the fuel, blanket, and coolant systems, including the reactor vessel,
'pumps, "heat exchangers, piping and storage tanks, is Hastelloy N.

. The primary heat exchangers are of the tube»and»shell type. Each
shell contains two concentric tube bundles connected in series and
 

 

i
;

ok il b bR e GRS L e

-

e wie o -

R i ey ol s

el

 

 

16

 

: ; -*

 
  
 

 

i Mw i R
‘Thermal
Electrical

L3 thermal efficiency

 

. Plant factor

Dimensions, ft :
Core height 12,5

.Core dlameter ' . 10.0
Blanket thickness ‘ o

Radial 1.5
Axial ’ o 2.0
Reflector thickness o 0.2
., Volumes, £t3

507 Core | 7 S 982
~ Blanket i o T 120

e B S e

    

Volume fractions
: Core ! e | o
Fuel salt ' o 0,169
Fertile salt : 1 0.0735
Moderator S T 0.T575
Blanket o ‘ ‘ oo T
Fertile salt C 1.0

 

Salt volumes, £t°
~ Fuel

 

Coré . . - Coem e dll T W 'flr . 166
e Blanket o . 26
@ Plena o ' o o b7
C Heat exchanger and piping ' 345
Processing : 33

w2 Total : T

   

TSEee UMy R D lEad L

 

" Fertile
% Core
" Blanket T ‘
' T el 'Heat exchanger end piping =~ - 100 e : ¢
wEOT e ene e N e T 5‘/‘7 Storage (Pa decfly) ' 2066 - :
e o Processing o
mf T e My e Doted , S 3383

   

 

‘ Salt compositions, mole §
o ) Fuel
LiF
BeFo '
UFy (fissile
Fertile
LiF
BeFo

ThFy
UFy, (fissile)
Core atom ratios
~ Th/u
c/u
" Fissile i_.nven'tb.i'y, kg

- Fertile i_'n'ventox"y',r 1000 kg '

  

JnEes0 7L vl Processing
Cyele time, days .
~ Volatility process

Rate, £t3/day
. Volatility process

"0 'Unit processing cost, $/ft> :

 

 

 
 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  
 
  

 

 

 

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  
  
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

ki e MR RRGE  f . mbua s | . e R i s Sl i b R LR i s B R L S ol o BNl GURR R R EHES G . GEEEED L. o G GRS B G R R AL L e L 2L
* - 8 » &
s 5 [
UF, RECYCLE TO REACTOR

* i . i

SORBERS o ers | _ ,;

COLD TRAP sor Ia COLD TRAP '

. MNof MgF, -70°C NoF Lo [

100-400%C 2 vooscovc | | MaFa -700c |

; N

UF,
T 1 ‘ { | |
WASTE WASTE
E
STORAGE PROE’;:(EI?ISON J STORAGE
. + . NqF/MgF/FP : NgF/Mng/F?
MAKE-UP 6. ; UF, + MAKE-UP
LiF/BeF ,/ ThF VOLATILE FP VOLATILE F? _ [ o eer
4 2
Y : { ]
BLANKET
CONTINUOUS CONTINUCUS DISTILLATE UF —aUF ,

FERTILE FLUORIDE .| HOLDUP FOR FLUORIDE . VACUUM LiF/3eF 6 T4, - FILTRATION
MAKE-UP VOLATILITY USPENT | FP DECAY VOLATILITY DISTILLATION 2 repuction |, _

FUEL ~ 550°C ~1000°C ~ 56%C- 550-400°C

o~ », —~ "
| 550°C 1.5 DAYS Up oFUF, o Mg
J . 1
. J . ' L_. H,
: — LiF + RARE
LiF/BeF F F
F/BeF/ThF /EP Ty 2 EARTH FP REDUCED METALS
‘ L " Cr, Fe N
DISCARD FOR : WASTE
FP REMOVAL | STORAGE
. 'i .

 

 

 

FERTILE STREAM RECYCLE

 

 

Fuel and Fertile

 

Lif /8ef,, LF, RECYCLE

Stream Processing for the MSBR.

 

 

LT

 
P

 

i, Sl o & R s B R BRI S, . k.

 

18

: ,:':.a::l»:f *"-g:‘-,:;ri-"i’:'.‘ i ?.w e T TR e S e

'attached to fixed tube sheets., The fuel salt flows dovnward in the outer .

section of tubes, enters a plenum at the bottom of the exchanger, and
then flows upward to the pump through the center section of tubes. Enter-
ing at the top, the coolant salt flows on the baffled shell-side of the
exchanger down the central core, under the barrier that separates the two

'sectlons, end UP the outer annular sectlon.

. Since a large temperature difference exists in the two tube seetlons,
the tube sheets at the bottom of the exchanger are not attached to the -

 ‘shell. The design permits differential tube growth between the two sec~
| tions without creating troublesome stress problems. To accomplish this,

the tube sheets are connected at the bottom of the exchanger by a bellows-
type j01nt.' This arrangement essentially a floating plenum, permits

" enough relative motion between the central and outer tube sheets to com-

pensate for difference in tube growth without creating intolerable stresses

S in elther the joznt the tubes, or in the pump.

: The blanket heat exchangers increase the temperature of the coolant
leaving the primary core heat exchangers. Since the coolant-salt tempera-

" ture rise through the blanket exchangers is small and the flow rate is
. relatively high, the exchangers are designed for a single shell-side pass

for the coolant salt, although two-pass flow is retained for the blanket

, salt in the tubes. Stralght tubes with two tube sheets are used.

G The superheater is a U-tube, U-shell exchanger using dlsc and dough- *
" nut baffles with varying spacing. It is a long, slender exchanger having
relatively large baffle spacing. The baffle spacing is established by the o

shell-side pressure drop and by the temperature gradient across the tube
wall, and is greatest in the central portion of the exchanger where the
temperature difference between the fluids is high. The supercritical
fluid enters the tube side of the superheater at TOO°F and 3800 psi and
leaves at 1000°F and 3600 psi°

The reheaters transfer energy from.the coolant salt to the working
fluid before its use in the intermediate pressure turbine. A shell-tube
exchanger is used, producing steam at lOOO °F and 5h0 psi°

- Slnee the freezing temperature of the secondary salt eoolant is about
TOO°F, a high working fluid inlet temperature is required. Preheaters,

- along with prime fluid, are used in raising the temperature of the working
. fluid entering the superheaters° Prime fluid goes through a preheater

, exehanger and leaves at a pressure of 3550 psi and about 870°F. It is

- then injected into the feedwater in a mixing tee, producing fluid at

TOO°F and 3500 psi. The pressure is then increased to about 3800 psi by

‘a pressurlzer (feedwater pump) before the fluid enters the superheatero

[

 
 

AL A ek ke Ml e b Auama -

el MR £ B o el e b b e

]

CAPITAL COoST ESTIMATES

: iy T
e el e SAY

Reactor Power Plant

 

fwbut s1gnificantly dlfferenf yields:
calculations was pr&marlly to find the conditions” fihatlgawe ‘the lowest

'-'hlghest fuel yieldo

* Preliminary estimates of the capital cost of a 1000-Mw(e) molten-
salt breeder reactor power station indicate a direct construction cost
of about $80.4 million. After supplying the indirect cost factors used
in the advanced converter evaluation,’ an estimated total plant cost of
$113.6 million is obtained. A summary of plant costs is given in Table
2., The conceptual design was not sufficiently detailed to permit a com-
pletely reliable estimate; however, the design and estimates were studied
thoroughly enocugh to make meaningful comparisons with previous converter

 reactor plant cost studies. The relatively low capital cost estimate

obtained results from the small physical size of the MSER and the simple
control requirements. The results of the study encourage the belief
that the cost of an MSBR power station wzll be as low as for stations
utillzing other reactor concepts°

The operatlng and maintenance costs of the MSBR were not estimated.

- Based on the ground rules used in reference 1, these costs would be
about o 3 m..ll/kwhr'(e) -

Fuel Recycle Plant

_ The capital costs associated with fuel recycle equipment were obtained
by itemizmng and costing the major process equipment required, and esti-
mating the costs of site, buildings, instrumentation, waste disposal, and
buildlng services as3001ated With fuel reeycle°

Table 3 summarizes the direct constructlon costs, “the indirect costs,
and total costs associated with the 1ntegrated proce581ng facillty having

‘approxlmately the requlred capaclty,_

 The operating and maintenance costs for the fuel recycle facility
include labor, labor overhead, chemicals, utilities, and maintenance
materials. The total annual cost for the capacity considered here (15
12 of fuel salt per day and 105 £t of fertile’ salt per day) is est1»
mated to be $721,230, which is equivalent to about 0.1 mlll/kmhr(e)
A breakdown of these charges is given in Table H

     

The*dbgective of the nuclear design

ng thls cost the

fuel cycle cost “and then, w1thout app ci bly" nc”udm

A T R

 
&

Table 2. Preliminary Cost-Estimate Summary |

IOOO-MW(e) Molten-5Salt Breeder Reactor Power Statlon

 

Federal Power Commission Account

'COsts ($1000)

 

 

 

*¥See Table 3 for these costs, which

.are not included here.

20 _Land and Land nghtéb 360 |
21 Structures and Improvements e
.. 211 Ground Improvements 866
. 212 Buildings and Structures o

.. +1 Reactor building® 4,181
‘v 2 Turbine building, auxiliary building, 2,832
. .. . and feedwater heater space
~7 «3 Offices, shops, and laboratories 1,160
4 Waste disposal building 150
.5 Stack 76
«6 Warehouse . ho
.7 Miscellaneous , 30
: Subtotal Account 212 8,469
Total Account 21 9,335
22  Reactor Plant Equipment o
'_q..221 Reactor Equipment
wroen .1 Reactor vessel 1,610
.2 Control rods 250
.3 Shielding and containment 1,h77
4 Heating-cooling systems and vapor- 1.20
. 3 O
suppression system e
Moderator and reflector 1,089
- +6 Reactor plant crane S 265
Subtotal Account 221 5,801
222 Heat Transfer Systems
-~ +1 Reactor coolant system 6,732
.2 Intermediate cooling system 1,947
«3 Steam generator and reheaters 9,853
“‘Seg.& Coolant supply and treatmentd - 300
.5 Coolant sslt inventory 354
' Subtotal Account 222 19,186
223 RNuclear Fuel Handling and Storage 1,700
' (Drain Tanks) ’
22k Nuclear Fuel Processing and Fabrication *
- (included in Fuel Cycle Costs) -
225 Radloactlve Waste Treatment and Disposal 450
| (off-Gas System) |
226 Instrumentation and Controls 4,500
227 TFeedwater Supply and Treatment 4,051
228 Steam, Condensate, and FW Piping 4,069
229 Other Reactor Plant Equipment (Remote 5,000©
Maintenance)
Total Account 22 Ll 8h7
Continued
 

[N SRS ST~ TP PN P SRR TP T T

o A M L ELE L i

B B K L, DiR P B0 SEEL .o Sk R e L -

i S o G GARAEE.

L ealiioke D

S iffiéderel‘Power Commission,Accoufififf”wq"""'

Table 2 (continued)

i AT R A 1t P g e i et sy el RIS gy B P T e A S AT

" Costs ($1000)

 

23 | Tufbine-Generator'Units

231
232
233

23k

235
236
237

238

o Accessory Flectrical
: Switchgear, Main and Station Service

2k

=) T-3

2k3
2ul;

c  o5

Other Turbine Plant Equipment
L Total.Account 23

Turbine-Generstor Units
Circulating Water System
Condensers and Auxiliaries
Central Lube 0il System

Turbine Plant Instrumentation
Turbine Plant Piping

Auxiliary Equipment for Generator

Switchboards

Station Service Transformers
Auxiliary Generator
Distributed Items

. Total Account 24

25 Miscellaneous

Total Direct Construction Costg
Total Indirect Cost
Total Plent Cost -

19,174
- 1,243
1,690
80

25
2o0f
66
2D
22,523

500
128
169
50
2,000

2,897

800

80,402
181

335
113,503

 

8pstimates are based on 1966 costs; assuming an established molten-
salt nuclear power plant industry.

Land costs are not included in total‘direct construction costs.

MSBR contalnment cost is 1ncluded in Account 221 3

 

dAssumed as $300,000 on the basis of MSRE experience.

 

| The amplerMSBR allowance for remo ehmamntenance may be too high,
and some of the included replacement equipment ‘allowances could more
logically be classified as operating expenses rather than first capltal

costs.

gDoes not 1nc1ude Account 20 Land Costs.

 

 

1ndirect costss

 

 

'1'un plant cost of $160 000 plus 37% for uncertaintlese,

This 1siincluded 1n the
R2

Table 3. Summary of Processing-Plant Costs for lOOO-Mw(e) MSBR

 

8 53,160

&

 

Installed process egquipment
Structures and improvements 556,770
© Waste storage 387,970
Process piping - 155,800
Process instrumentation 272,100
Electrical auxiliaries ~ 84,300
Sampling connections N 20 000
Service and utility piping S o 128 060
Insulation - o 50,510
Radiation monitoring | | - 1002000
* Total direct cost T $2,609,270
Constructlon overhead 182,780

(30% of direct costs) Total construction cost_ _§B3,392 050

Englneerlng and inspection

(25% of total construction cost) | "m§E§LQ£9
o - Subtotal plant cost gkh,ehofioéo”
Contingency (25% of subtotal o "
. plant cost) 1,060,020
o Total plant cost $ 5,300,080

 

N Table h Summary of Operating and Maintenance Charges
o “for Fuel Recycle in a 1000-Mw(e) MSBR

 

$/year)
Direct labor | '_ $ 222,000
. Labor overhead 177,600
i/ Chemicals ~ 1h,6h0
~«..Waste containers - 28,270
oo o Utilities - 80,300
- Maintenance materials

Site 2,500

ciierises Services and utilities | 35,880
| Process equipment | 16020h0

T i:V'I‘otal annual charges - $ 721,23

 

>

&

Ty
e etk

i, R e e Bt 2 o s Mo b MR o B

il L B AR L i K SBBR b ke, i

.fl.-.kmh.a il o

&

;
]
:1
4
A
3

 

 

"

o BasxcfAssum@tlons

LS The total?processing cost is assumea to be a functlon'of the throughput

23

 

Analysis Procedures

Calculation Method. The calculations were performed with OPTIMERC,
a combination of an optimlzatlon code with the MERC multigroup, dlffus1on,

 

equilibrium reactor code. MERC® calculates the nuclear performance, the
‘equilibrium concentrations of the various nuclides, including fission
products, and the fuel cycle cost for a given set of conditions. OPTI-

MERC permits up to twenty reactor parameters to be varied, within limits,
in order to determine an optimum, by the method of steepest ascent. The
designs were optimize& essentially for minimum fuel cycle cost, with

"1esser weight given to maximizing the annual fuel yield. Typical param-
 eters varied were the reactor dimensions, blanket thickness, fractions
- of fuel and fertlle salts in the core, and fuel and fertile stream pro-

cessing rates.

 Several eQuations were included in the code for approximating cer-

o tein capltal and operating costs that vary with the design parameteérs

(for example, capital cost of the reactor vessel, which varies with the
reactor dimensions). These costs were automatically added to the fuel
cycle cost in the optimization routine so that the optimization search
would take into account all known economic factors. However, only the

fuel cycle cost itself is reported in the results.

' Mbdlfied GAM-l —-THERMOS cross section lIbraries were used to com=
pute the broad group cross sections for these calculations. It was
assumed that all nuclides in the reactor system are at their equilibrium
concentrations. To check this assumption, a typical reactor design was
examined to determine the operating time required for the various uranium
isotopes to approach their equilibrium concentrations from a startup with
285y, It was found that 233y and ° 235y were within 95% of their equilib-
rium concentrations in less than two yearso Uranium-234 was within 95%

of equilibrium after eight years, while 236y was within 80% after 10
years. Since the breeding performance depends mainly on the ratio of
238y to #°5U in the fuel, the equilbrium calculation appears to be a
good representation of the 1ifetime performance of these reactors, even
for startup on 22 . |

     
   

to some fractional power called the scale factor.
 

 

 

24

Table'5° Basic Economic Assumptions

 

 

Reactor power, Mw(e) | | 1000
Thermal efficiency, % ' ks
Load factor 0.80

2?1rCost assumptions | S T
J ,.;Value of 233y and 233Pa $/g - | . 1k
.. Value of 225U, $/g | 12

sy, . .Value of thorlum, $/ke 12 ‘
i, JValue of carrier salt, $/kg 26
Capital charge, annual rate, % T

.. Plant | 12
Nondeprec1ating capital, includlng lb
'ifi531le 1nventory

    
 

;1WMMProoe581ng oost $/ft3 salt

“Fuel (at 10 £t3/day processing rate) 228
Blanket (at 100 fts/day processing rate) 8.47

Process1ng cost scale factor (exponent ) 0.k

 

&

 Processing. The processing scheme is that indicated in Fig. T.

A f13511e material loss of 0.1% per pass through proce331ng was assum.ed°
In addition to the basic proce831ng scheme employed results were

also obtained for the case that Pa can be removed directly from the

blanket stream. The improvement in performance under these clrcumstances

is a measure of the incentive to develop Pa removal ability.

Flssion Product Behavmro The. dlSPOSltlon of the various flssion
”products was assumed as shown in Table 6. The behavior of *>°Xe and
other fission gases has a significant influence on nuclear performance.

 

7 A gas-stripping system is provided to remove these gases from the fuel
- galt. However, part of the xenon could diffuse 1nto the moderator’

* graphite. In the calculations reported here, a *°Xe poison fraction
: of 0 005 was assumedo '

   
 

o Corrosion Product Behav1or. The oontrol of corrosion products in o

' molten-salt fuels does not appear to be & 51gniflcant prdblem, and the

effect of corrosion products was néglected in the nuclear calculations.

The proce351ng method oonsidered here can oontrol corr031on product

bulldup 1n the fuel. , fwn e »

 

 

T TR O TpEECT R P oox oron

 

 

o

 
 

i
@

ke

e b G e R R BRARL .

I

¢ oot b, ARG o MR i B odoeri

<2

Table 6. Disposition of Fission Products in MSBR
..Beactor and Processing Systems

Sl A LD g T

 

Elements préEEEt as gases; assumed to be paftI§¢;“Jv“m
absorbed by graphite and partly removed by gas
strlpping (l? poisoning assumed ) : Kr, Xe

$s et T
SRR e S e s ST T

Elements that plats out on metal surfaces, assumed

to be removed instantaneously: Ru, Rb, Pd, Ag, In
Elements that form_volatile fluorides;%asgfiséé&%gfikiuiéeiufi§;'fio;ffio, Te,
be removed in the fluoride volatility process: Te, I
'Elements that form stable fluorides less volatile  §r, Y, Ba, La, Ce,
- than LiF; assumed to be separated by vacuum Pr, N4, Pm, Sm, Eu,

distillation: \ Gd, Tb

Elements that are not separated from the carrier
salt; assumed to be removed only by salt discard: Xb, Cd, Sn, Cs, Zr -

 

Nuclear Design Analysis

The important parameters describing the MSBER design are given in
Table 1. Many of the parameters were basically fixed by the ground rules
for the evaluation or by the engineering design. These include the therm-
al efficiency, plant factor, capital charge rate, maximum fuel velocity,
size of fuel tubes, processing costs and fissile loss rate, and the out-
of-core fuel inventory. The parameters which were optimized by OPTIMERC
were the reactor dimensions, power density, the core composition includ-
ing the C/U and Th/U ratlos, and the processing rates.

_ Nuclear ?erformancec‘ The results of the calculations for the MSER
aes1gn are given in Table T, and tH& héutron balance in Table 8. The
basic design has the inherent advantage of no neutron losses to struc-
tural materials other than the moderatoro Except for some unavoidable
loss of delayed neutrons in the external fuel circuit, there is almost

- 2ero neutron leakage from the reactor because of the thick blanket.
. The neutron losses_to fission products are minimized by the availability
- of rapld an?.” e

 

N‘xpéhs1ve integrated process::.ngo N | e
Tuel Cycle Cost. The components of the fuel cycle cost for the NEER

are given in Table 9. The main components are the fissile inventory and

processing costs. The inventory costs are rather rigid for a given reac-

 

. tor design, since theéy are largely determined by the assumed external

fuel volume, The processing costs are, of course, a function of the pro-
cessing cycle times, one of the chief parameters optimized in this study.
che B kB L ekl h S B Wl el

e i ik L e e e il

 

26

Teble 7. MSBR Performance

»

 

Fuel yield, % per annum

Breeding ratio

Fissile losses in processing,

. atoms/flss1le absorption

o Neutron production per fissile

| absorption (ne)

' Speclfic 1nventory, kg fissile/Mw(e)

- Spec1f1c power, Mw(t)/kg fissile

Power density, core average,

S kw/liter

Gross

g A

In fuei salt

| Neutron flux, core average,

1014 neutrons cm "2 gec

'f_ ‘Thermal
Fast
Fast over lOO kev _

 fThermal flux factors, core, peak/mean

T Radlal
Axial

:fFractlon of f1351ons in fuel stream

iFraction of fissions in thermal i-°??** e

' neutron group

,.%MEan n of 233U

.Mean no of 235U

T o .«g; Lo

u 86

" °1 oh91

0.0057

- 2.221

0.769

2.8

473

6.7

‘\

12.1
3.1

o 2.22
| 1. 37

 o.8T |
-2? 36:88gg;5m-ff3;“

”wMg 221;4“” ‘)

| '1 958* o

 

 

 

7

 
el b ke e cn e

ek e MGG aSEE cchbes ik oo L ik

e b cotenen oAl LD L

ek e bR R B sl L ko Lob el Ml R MR ol

Aol b

- ciad E

o e B L

s o 5w e eeams et S e e T Ty ot e R R LY

b

Neutrons per Fissile Absorptlon

 

 

i_ Material B :f’ Absorbed ”". Absorbed
Cin LRl “ "Total o by Fission

 

Produced

 

256y

'237Np

288y “

Carrier salt
(except ©1i)

6Li

Graphlte

 \135X£

' 1498m

lSlsm

P

Other fission

. products

“Telayea nditrons -+

lost®

    
  
  
 

o019

Ce9119

©0.09%
0.0881

0.0115
0.001k

0.0009
0.0623

0.0030

0.0300

O 0050

 

0.0025

0.8090

0.000k

0.0708

0.0001

00059

2.0233

.. 01721

0.0001

i

0.0185

 

 

 

 

  

 

aDelayed neutrons emitted outside théwébfé}c

    

S bLeakage,{héluaifig'figfitféfié’éfiéOfEéd:ifi'tfiétféfiécféf;"
rs e e S OB, | SE——

b el L -l e L i - 1 n

28

Table Q. Fuel Cycle Cost for MSBR

 

Costs, mills/kvhr(e)

 

Fertile

" Grand

 

Fuel ,.
- Stream Stream Total Total
Fissile inventory® 0.1180 0.032%  0.1504
~ Fertile inventory 0.0000 0.0459 00,0459
- Salt inventory 0.0146  0.0580 0.0726 N
Total 1nventory 0.2690
.,Fertlle replacement 10.0000 0.0185 0.0185
Salt replacement 0.0565 ~ 0.0217 0.0782
TotalrreplacefiE§féJ? o '0.096ffimm
Processing 0.110%  '0.0411  0.1513
Total processing | '0.l5i§$fi$#
. Production credit ©o0.0m8
0.h4452

- Net fuel cycle cost

#Including 23%pPa, 233y, and 235y,

 MSBR Performance with Pa-Removal Scheme.

 

directly from the blanket of the MSBR has a marked effect on fuel yield
and fuel cycle cost. This is due primarily to the marked decrease in

&t TEAE L bk o
bl ew) R E Ay
M o RS Sl i

Pa neutron absorptions when Pa is removed from the blanket region.
simple and inexpensive blanket Pa-removal scheme would give the MSBR

POWER COST AND FUEL UTILIZATION CHARACmWRISTICSV

‘Based on the above, the power cost, spec1f1" fissile 1nventory, ‘and
fuel doubllng time for the MSBR and MSBR (Pa) are summarized in Table 11.

et

 

The ability to”fém5§%5Pa

A

~_ the performance indicated under MSBR (Pa) in Table 10; for comparison,
- the results W1thout Pa removal are also given in the table. -

Table 11 1llustrates the economic advantage of MSBR's as nuclear

pewer plants. Also, the fuel utilization characteristics as measured by
the product of the specific inventory and the square of the doubling time*
are excellent. On this basis the MSBR is comparable to a fast breeder
with a specific inventory of 3 kg/Mw(e) and a doubling time of 10.5 years,
- while the MSBR (Pa) is comparable to the same fast breeder with a doubllng
~ time of 6 years.

%

 
i

 

 

 

A 29¢
2 Table 10. Comparison of MSBR Performance With
j and Without Pa Removal
i - MSBR° MSBR (Pa)
! (Without Pa (With Pa
Process) Removal)
é Fuel yield, % per annum L8 7.95
| Breeding ratio 10491 1,0713
Fuel cycle cost, mills/kwhr 0.k45 0.33
Specific inventory, kg/Mw(e) 0.769 0.681
Specific power, Mw(t)/kg 7 2.89 \ 3.26
Neutron production‘pér fissile
gbsorption (ne) 2,221 2,227
Volume fractions, core
i Fuel ' 0.169 0,169
4 & Fertile 0.0745 ~ 0.0735
i Moderator 0.7565 0.7575
L w
' Salt volumes, £t3
Fuel
Core 166 166
External . . . S5ht o 551
Total T13 TLT
‘Fertile | | .
Total o 3383 . 1317
Core atom ratios , | - S
¢/u . sdio 5800
&
i e HRESEEE

Table 11.

of the MOBR znd

Reactor:

ne MSBR {Fa)

s

- -. ' s T TP S ey & At e d gwede S e .
Power Cost and Fuel Tiilization Characteristios s

i il .,./!"’Wh’”(ff‘)

 

 

MSBR MEBR (Paj
. ' a 3 E 0
Capital cost 1.G5 1.95
» : b 2, .
Operating and maintenance cost 0.20 .20

FHP* cycle cost

Total power‘acst

0.b5
fi, v-r 0

bummn---,---—-------_----‘fiui---‘-l--—wp-‘\b—‘Rmm,¢wufimwa:@_wwxmm

Specmzlc fissile inventory, ke/Mw{e) 077

Fuel doubling time, years

20.5

0,68
12.6

T. ¥. Kerlin, Jr.,

 

Twelve per cent fixed sharge rate, 80% road ?aflhars
1000-N z{e) plant, |

Nominal value used in a vanued cafiver+er evaiuation,

Cobts of on-site 1n+eg

;*c uaed in thlq value%_

ated process

ing plact are

ey
HoPERT Jw“ h

’.

4

o .
e ST, A ST e
LTTaR e

g

 

M. W. Rosenthal et al., A Corsarative Bvaluation of Adwvanced

(AN -3686 (Janusry 1965).

  
 

5ot
T A s e
o, 3 I

 

 

Seott and W. L. Carter, Pralim
ination-Vacuum Distiliabicn System for Regenerating Fusl
»iile Stresms in a Molten Zali Breeder Reactor, ORNL»ZT?Z

   

en, Jr., L. G. Alexanderj ard J
, RNL-?M 847 (April 22, 156},

C. W Crev
The MERC=-1 Equllibr*um Cbifi

Paul R. Kasten.

 

 

;96/} Wth“ngton, D. C,

minary Uesign Study of & o

"Nuclear ¥Fuei Utilization and Economic Incentiw
pager presented at Amerizsn Nuclear Society Meeting, Hovesber 1

 

g5
Gt
o,

1,
‘:‘(n 5 amfl?“v
* T‘ Ju;»{. ¢‘f 2:3

i
R

e
& "‘:»-4
2T
L e e i i i i i e SRS,

RARRGEL oriiibe o o oen e NV

et BRSPS il G

b e -

R

L E

-l henakh L SRR Sl il e

& o ®

9

2 o o

\O QO=1 O\ =0 PO

@

=
O

11.
12.
13.
1k,
15,
16.
17.
18 ar
o,
20,
21-&1,
ho,
43,

45,
L6,
,-I'To

79,
81.
8.

. f,,.'—: 83.

8L,

- 85-99.

100,

101-102._"“"’;

 

Internal Distribution

G. M, Adamson 48, R. E. MacPherson

C. J. Barton 49, H. C. McCurdy

H. Fe Ba‘lman 500 W. Be MCDOHa-ld.

S. E. Beall 51, H. F. McDuffie

C. E, Bettis 52, A. J. Miller

E, S. Bettis - 53. R. L, Moore

F. F. Blankenship 54, M. L. Myers
"Re. J. Braatz ' 55, Re C, Olson

R. B. Briggs 56, A. M. Perry .

S. Cantor - 57T T W, Pickel

R, S. Carlsmith | 58. R. C. Robertson

W, L, Carter 59. A. W. Savolainen

W. H. Cook ' 60. D. Scott

G. A, Cristy | 61, J. H. Shaffer

F. L, Culler = ' 62, M. J, Skinner

D. E. Ferguson 63. I. Spiewak

A, P, Fraas 6k. R. E. Thoma

Wo R. Grimes 65. G..M, Tolson

A. G Grindell = - 66. D. B. Trauger

P. N. Haubenreich 67. A. M. Weinberg

P, R. Kasten 68. J. H. Westsik

Co R. Kennedy 69. J. C. White

S. S. Kirslis 0. G. D, Whitman

J. A, Lane TL-T2, Central Research Library
M. I. Lundin 73~Th. Y=12 Document Reference Section
R. N. Lyon - -T7T. Laboratory Records

H. G, MacPherson 78. Laboratory Records, RC

External Distribution

D. F. Cope, AEC, ORO

C. B. Deering, AEC, ORO

R. G. Garrison, AEC, Washington

R, E, Hoskins, TVA, Chattanooga

W. J. Larkin, AEC, ORO

Se R. Bapirie, AEC, ORO

Division of Technical Informaetion Extension (DTIE)
Research and Development Division, ORO

fReactor Division, ORO

 
 

-

‘&

 

 

 
 

 

 

 

TR T -

 

 
 

A A Tyl i T IR T T

gt

G e N Y SR —