ORNL-CF-56-10-110.txt

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OAK RIDGE NATIONAL LABORATORY
| Operated By
UNION CARBIDE NUCLEAR COMPANY

WO' S i | - | ()l(l“*l. |
DAk oo, TENEAGES | CENTRAL FILES NUMBER
' O6C-10- 1

 

 

 

October 29, 1956 - This document consists of 9  pages.

FUSED SALT POWER REACTOR STUD
Minutes of Discussion Meeting No. 3

Distribution |
L. G. Alexander and J. T. Roberts

 

Copy /? of 22 copies. Series A

For Internal Use Only

Distribution

L. G. Alexander
E. 5. Bettis

D. 8. Billington
D. A. Carrison
R. A, Charpie
5. Jo. Cromer

W. K. Ergen

W. R. Grimes

W. H. Jordan

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L. A. Ma |

H. F, Popp@ndi@k

J. T. Roberts

J. A. Swartout

F. C. VonderLage

A, M. Welnberg

Laboratory Records
- C. R, Library

   

  

. This documem contains Restncfed D atg..qn
Energy Act of 1954, Its. transmltW%e dlsclosure of its contents
in any manner to an unoutg’gg person is prohlblted

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FUSED SALT POWER REACTOR STUD

 

Minutes of Discussion.meetingzm@013

 

October 18, 1956

Present: L. G. Alexander H. G. MacPherson
5. Bettis E. R Mann

A. Carrison J. T. Roberts

J. Cromer | J. A, Swartout
K. Ergen F. C. VonderLage
W. R. Grimes A. M. Weinberg
G. Wo Keilholtz

 

H. G. MacPherson opened the meeting with a summary of the discussion at the
previous meeting.

L. G. Alexander presented the result @f his UNIVAC calculations using the
Eyewash code.

Eleven reactor cases were studied. The reactors were spherical, the total
power was 600 mw, and the fuel consisted of a nearly equi-molal mixture of

NaF and ZrF), , together with additions of up to 10 mol percent ThF) and suffi-
cient UF) to make the assembly critical. Specific heat release rates of

50, 100 and 200 watts/cc were used, to which core radii of 56, 44 and 35
inches correspond. The cores were bounded variously by l=-inch Ni alloy shells,
reflectors of graphite and NaZfF5g and a blanket containing ThF}.

‘In the first five reactor cases, the core had a radius of 56 inches and was
bounded by a l-inch shell of Ni alloy. The ThF) concentration was varied from
zero to 10 mol percent. The results, in terms of breeding ratio and critical
mass, are summarized in Table I and illustrated in Figure 1. It is seen that
substantial breeding ratios can be obtained at the cost of considerable fuel
inventory,; e.g., to obtain a breeding ratio of 0.75 requires a critical mass
of about 250 kg of U-233. From these data, the optimum combination "of breed-
ing ratio and fuel inventory may be selected.as soon as processmng costs and
1nvent@ry interest rates have been established.

The leakage from these cores is substantial, varying from 23 percent for the
reactor containing no thorium, down to an estimated 12 percent for that con-
taining 10 mol percent ThF). This leakage could be reduced by increasing the
radius of the core, but the critical mass would increase sharply. Conversely,
decreasing the radius (and increasing the power density) reduces the critical
mass sharply at the cost of reduced breeding ratio, as shown in Table I,

Cases 211061 and 211071. |

The leakage from these reactors is excessive. A reflector more efficient than
the Ni alloy shell could be used to reduce leakage. The results from two cases
of reactors reflected with graphite are shown in Table I, Cases 21208k and 212095,

 
SECRET
3.

and are there compared with corresponding reactors having Ni alloy shells.

It is seen that although the substitution of the graphite for the Ni had

little effect on the breeding ratio, a substantial decrease in the critical
masses was obtained. Table II gives the breakdown on the neutron balance

for the larger graphite reflected reactor (212084). The fast leakage was
confined to energies above 34 ev and amounted to 0.306 neutrons. The reflec-
tor returned to the core 0.028 neutrons having energies between 0.084 and 34
ev, and 0.153 neutrons at thermal energy (0.084 ev). Somewhat more than half
of these thermal neutrons caused fission, while only about one-fourth were
absorbed in thorium. -This tended to reduce the breeding ratio from the value
of 0.73, observed- in the Ni reflected reactor. However, the saving of thermal
neutrons permitted a reduction in the U-233 concentration, thus increasing the
ratio of absorptions in Th to absorptions in U-233 in the epithermal region.
This increase offset the increase in parasitic captures at thermal energy. It
seems clear that a non-moderating reflector would be better than graphite, pro-
vided the parasitic absorptions are not significantly greater.

 

 

The last two cases studied (213127, 213126) concerned reactors having radii of
L inches, with no thorium in the core, and a blanket 16 inches thick. In the
first case, the blanket contained 10 mol percent ThFl, and in the second, no
thorium. The critical masses were nearly equal, ~ 35 kg U-233. The breeding
ratio for the first case was 0.60. Table II gives a breakdown on the neutron
balances for these two cases. It is seen that the majority of the absorptions
are epithermal, and that the fractions of the absorptions in Th and U; which
are epithermal, are greater than the corresponding fraction for the carrier.
The epithermal leakage is seen to be nearly independent of the presence of the
thorium in the blanket, and the thermal leakage is but little affected. This
is thought to signify that neutrons which are not reflected into the core by
their first blanket-collision have little chance of returning to the core by
subsequent collisions, being absorbed by the thorium in the one case and the
carrier in the other.

The critical mass (35 kg) was small compared to the critical masses obtained
for other cases in this study. The breeding ratio was comparable to that
obtained in a one region reactor containing 200 kg of U-233. Clearly, the
breeding ratio can be increased by decreasing the radius of the core. (See
post meeting remarks in appendix.) -

A discussion of the accuracy of the Eyewash code cross sections followed the
presentation of the UNIVAC results. A. M. Weinberg commented that the non-
experimental cross sections in the code apparently have not been recalculated
using the theory now available and that the absorption cross sections may thus
be significantly in error. H. G. MacPherson suggested that the thorium absorp-
tion cross section was the most important in this respect, since it strongly
influences the flux-energy relationship, as well as the breeding ratio, and
since the U-233 and U-235 cross sections are based on experiment. In connection
with the U-233 cross section, he reported that recent experimental results from
Arco indicate that the low value of’? at 2 ev is a resonance and that’? apparently
does hold up well at intermediate energies, confirming Russian data.

 

D. A. Carrison reported on his multigroup hand caleulations, using the Eyewash e
code groups and cross sections, for bare, homogeneous Li7éBeéU«233dThéF reactorsowgf

 

SECRET
'SECRET

Reactorwcgde No.

 

Power Density, w/cc
Core Radius, ino
| ThF) , mol %
UF),, mol %
Region IT
Thickness, in.
ThF), ; mol %
'B.R., Th/U
C.M., kg
Losses, %

Leakage, %

_211010

SEMESICCORSCORIMEIIIREICSS 000 CROCHERCOECSIRETIIICIRD 00 COSNIRGONRGRDE R 0 SRssntciRears I smeD

0.065

- Ni-Mo

211022
50

56
1.66
0.132

Ni Mo

0.49
118
36

U-233 FUSED SALT POWER REACTORS

211033
50

56
3.35
0.24k

Ni Mo

0.3
226
25

13

TABLE I

50
56
6.80

0.677

Ni=Mo

0.87
609

.19

211071

21208k

mmwmmwm

50
56

10.26

1.13

Ni-Mo

0.93
958
16

12

0.61k

Ni =Mo

0.47
132
36

50
56

-3+35

0.161

146

 

0.6k

107

29

213127 213126
100 100
Iy b

o 0.0

0.
35
31
35

SECRET

 

60 0

35
57
35.
SECRET

 

 

 

 

TABLE IT

U-233 FUSED SALT POWER REACTORS

Region I
(Core)
 Code No. Item Fast Thermal F& T % Fast
212084 Radius=56", ThFL=3.35 mol %, UFhéOalél mol %
23 fissures 0.2902 0.1120 0.4029 2.2
23 captures 0.0240  0.0112 0.0352 58 .2
Th 10,2914  0.0339 0.3253 89.6
NaZrFs5 0.0739  0.0408 0.11h7 6k .
Total Abs. 0.6793  0.1979 0.8772 T7 b
Leakage 0.279%  -0.1528 0.1266 ==
Th/23 0.926 0.275 - —-
213127 Radius=th" , UF)=0.08 mol % .
23 figsures 0.2930  0.0827 . 0.3757 8.0
23 captures 0.0285 0.0082 0.0367 TT- 9
Th - - -
NaZrFs 0.1500  0.0783 0.2283 65 3
Total Abs. O.4715  0.1692 0.6L0T 73.6
Leakage 0.3469  0.0113 0.3582  96.8
Th/23 - - - ea
213126 Radius=44" , UF4=0.078 mol %
23 fissures 0.2976 0.0755 003738 79.7
23 captures 0.0270 -~ 0.0075 0.0345 78.3
Th -- - - -
NaZrFsg 0.1426  0.071+  0.21k40 66.7
Total Abs. 0.4666 0.154k 0.6223 75.0
Leakage 0.3502 0.0178 0.3680 95.2
Th/23 - - -- -

Neutron Balances

Fast

 

0.0027

0.0000

Region II
(Blanket or Reflector)
Thermal F+ T
Graphite
- 0.0870 0.0897
0.0317 0.0317

% Fast

3.0

0.0

Blanket Thicknesss16", ThFLs10 mol %, NaZrF5=30 mol %

0.2176
0.0243
0, 2&19

 

0.0293
0.0145
0.0438
0.0056
0.322

0.2469
0.0388
0.2857
0.0734

o e

88.2
62.6
8h.T7
92.4

= D

Blanket Thicknesss16", ThF)=0 mol %, NaZrFg5=100 mol %

o

0.2480

0.0939

0.0201

0.0000

&= XD

[ g

0.2681

0.0939

92.4

100.0

a0 &R

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SECRET

FIGURE 1.

BREEDING RATIO AND CRITICAL MASS IN A
ONE REGION, HOMOGENEOUS, FUSED SALT, BREEDING REACTOR

 

Fuel: UFy, ThF),, ZrF),, NeF

Diameter: O ft, 4 in.

Shell: 1 inch of Ni-Mo Alloy

. Temperature: 1283°F

Reflector: None

Critical Mass of Burner (TfiFM = 0) = 58 kg U-233

Power: 600 mw of heat at a power demsity of 50 watts/cc
Fisslon Product Poisoning: None

Method of Calculation: Univac Code Murine-Eyewash, 24 groups, 2 regions

 

 

 

    

 

 

 

 

 

 

 

 

 

1.0 20
mm——— O -
| —
Breeding o— —
| Ratia - — ?
008 ‘ » \ // | / -
O
0.6 A L 12
/ | 0
®w
g
/;) | -
| | | o
o . _ - &
E ' 00)'8" / ‘ / \ ‘ . 8 S
o , o4
0%? | T Relative v
o _ / : Critical o
@ i Mass a3
S | .. 1
@ 02| [ 7 ' X
L 6 8 | 10 SECRET

 

" ThF), Concentration in Fuel, Mol %
SECRET o
To

 

For a 10-foot diameter sphere with 1200°F fuel of composition 63 LiF-15 BeFo-
20 ThFL-2.16 UFL, the breeding ratio was 1.05 and the leakage T.6 percent.
Reducing the diameter to 7 feet, changed the UF), concentration to 2.24 mol
percent and decreased the B.R. to 1.00. Leaving the diameter at 10 feet and
reducing the ThF) to 10 percent, reduced the B.R. to 0.98. With a 51 LiF-

45 BeFp-4 ThF) (plus 0.24-0.27 percent UF)) fuel, the B.R. was 0.84 for a
10-foot reactor and 0.64 for a T-foot reactor. Carrison pointed out that
these were not thermal reactors, even though the Li-Be carrier was used.
About half the neutrons are absorbed or leak above 9118 ev for the 20 percent
ThF), fuel, and above 454 ev for the 4 percent fuel. The main advantage of
the Li-Be salts is their lower melting points. He also pointed out that
apparently inelastic scattering by fluorine was not coupensated for in the
Eyewagh ¢y values. A. M. Weinberg commented that the breeding ratios calcu-=
lated are high because captures by Pa-233 are neglected, and that this
would be significant in a high power density reactor.

J. T. Roberts reported on a brief survey of the 1953 MIT and 1956 ORSORT
designs for fast reactors using chloride fused salts. Both studies neglected
the high n,p cross section of Cl-=35. Roberts reported rough estimates of
C1-37 costs obtained from Y-12 and K-25 of $5/gm for chemical exchange and
$1/gm for gaseous diffusion. The chemical exchange cost is based on present
technology. In principle, much better processes could be developed. The
gaseous diffusion cost is based on separation of anhydrous HCl in present K-25
type equipment, assuming the same efficiency. J. A. Swartout commented that
K-25 equipment would not resist attack by HCl. Other than the Cl-37 problem,
development work indicated for chloride salt fast reactors is much the same
as for fluoride reactors (i.e., cross sections, materials of construction,
components) .

H. G. MacPherson reported that L. A. Mann is making a trip to obtain informa-
tion on components, especially heat exchangers.

J. T. Roberts summarized the results of a preliminary look at optimization of
net fuel cost in & single region breeder. Based on multigroup calculations
for a 600 mw, bare, homogeneous , spherical reactor with 50 NaF-U6 ZrF) -k
(Th-+'U233)FM fuel and varying core radius (at constant extermal holdup of
423 ft3)9 and chemical processing cycle time (batch), the minimum fuel cost
reactor was as follows: |

Radius: 6.4 ft

U-233 Inventory: 813 kg

Core Power Density: 19 watts/cc
Processing Cycle: 21.6 yrs
Neutron Balance:

U-233 1,000
Th (average) .762 = B.R. (clean, infinite
F.P, poisons (average) .129 | (B.R. =1.15
Leakage | .26k
Na 013
Zr 014
F | 0,129
’ 2031 &= YZ

SECRET
SECRET

8.
Fuel Cost Breakdown: |

U-233 burn-up 0.595 ) 0.599

Th burn-up 0.00k4 )

U=-233 inventory 0.385 ) |

Th inventory 0.012 ) 0.597

Salt inventory 0.200 )

U-233 processing 0.045 )

Th processing 0.013 ) 0.135

Salt processing 0.077 ) ,

1.33 mils/kwh

The assumptions were: (1) F.P. poisons build up-as in a thermal reactor;

(2) U-233 costs $18.5/gm, thorium, $40/kg, and salt, $7.5/1b; (3) inventory
charges are 4 percent for U-233 and Th, and 12 percent for salt; and (4) pro-
cessing costs $1.85/gm for U-233, $40/kg for Th, and $7.5/1b for salt.

A. M. Weinberg commented that the assumption regarding ¥.P. poisons was too
optimistic for such an intermediate reactor. For a faster rate of poison
build-up, the optimum reactor would be smaller, the processing time shorter,
and the net fuel cost greater. H. G. MacPherson commented on the possibility
of starting up such a reactor on U-235 (whose critical mass was estimated by
Roberts to be 2.26 times that of U-233), and having to withdraw uranium
initially. E. R. Mann and E. S. Bettis commented on the "sleeping giant”
aspects of such a large, low power density reactor from the control point

 

il

of view.
quGo Alexander
0T 20 lend
J. T. Roberts
LGA/JTR/ds
Att.

SECRET
 

APPENDIX A-1

(Post Meeting Remarks by L. G. Alexander)

The breeding ratios in the two region reactors (Case 213127) could be
increased in a number of ways.

APPENDIX A-2

The radius of the core could be decreased; this would require operation
at a higher power density for the same rated power level, but might also
result in a decrease in critical mass. The power density could be held

stant and the leakage increased by utilizing a spherical annulus for
the fuel region. The central void would be filled with some material of
low capture cross section and low moderating power. The critical mass
would increase somewhat, but probably not as fast as the leakage. The
leakage could be increased "internally” by adding ThF) to the core, either
homogeneously or heterogeneously (e.g., in tubes). This would tend to
increase the critical mass, but would also tend to suppress parasitic
absorptions in the carrier in the core.

 

 

A consideration of the various factors involved leads to the conclusion
that the optimum breeding system will probably consist of a two region
reactor having a core containing some thorium (perhaps the maximum per-
missible, from nelting point considerations) and surrounded by a blanket
containing very high concentrations of thorium.

APPENDIX A-3

From a nuclear standpoint, thorium metal is the most desirable blanket
material. Because of its low moderating power, the energies of neutrons
reflected into the core would not be lowered much; of course, parasitic
captures in the blanket would be reduced to a minimum. But, in order to
prevent large reactivity changes due to formation of fissionable material
in the blanket and to prevent losses of Pa by neutron capture, it will be
necessary to process the blanket at a fairly rapid rate. The processing
of metal elements is disadvantageous. In Reactor Case 213127, only 85
percent of the neutrons absorbed in the blanket were captured by thorium,
but the thorium concentration there (10 mol percent ThFu) is probably
already higher than can be maintained in practice. It seems imperative
to achieve a greater thorium density. Some alternate possibilities for
the blanket include 20 mol percent ThF) in Li-Be fluoride melt (being
studied by D. A. Carrison); uranium oxide plasticized by sodium (proposed
by Bulmer et al); uranium oxide or fluoride powder in tubes. It is doubt-
ful that the last two materials can be handled as fluids.

A study of a series of reactor cases along the indicated lines is being
planned.