CHAPTER 3

PROPERTIES OF AQUEOUS FUEL SOLUTIONS*

3—1. INTRODUCTION

The chemical and physical properties of aqueous fuel solutions are
important because they affect the design, construction, operation, and
safety of homogeneous reactors in which they are used. This chapter will
discuss primarily those chemical and physical properties, except corrosion,
which are important for reactor design and operation. Special attention
will be given to the properties of solutions of uranyl sulfate, since such
solutions have been the most extensively studied, and at present appear
the most attractive for ultimate usefulness in homogeneous reactors.

Solubility relationships are discussed first, with data for uranyl sulfate
followed by information concerning other fissile and fertile materials.
The effects of radintion on water, the decomposition of water by fission
fragments, the recombination of radiolytic hydrogen-oxygen gas, the de-
composition of peroxide in reactor solutions, and the effects of radiation
on nitrate solutes are then presented. Finally, tables of relevant physical
properties are given for light and heavy water, for uranyl sulfate solutions,
for other solutions of potential reactor interest, and for the hydrogen-
oxygen-steam mixtures which oceur as vapor phases in contact with reactor
solutions,

3-2. SovusiLity REeraTronsuirs oF FissiLE aNp FerrTiLE MATERIALST

3-2.1 General. For the most part, studies of aqueous solutions of fissile
muaterials for use in homogenecous reactors have dealt with hexavalent
uranium salts of the strong mineral acids. Hexavalent uranium in aqueous
solutions appears as the divalent uranyl ion, UOs* ¥, Tetravalent uranium
salts in aqueous solutions are relatively unstable, being oxidized to the
hexavalent condition in the presence of air. Other valence states of uranium
either disproportionate or form very insoluble compounds and have not
been seriously proposed as fuel solutes.

Uranyl salts are generally very soluble in water at relatively low tem-
peratures (below 200°C). At higher temperatures, miscibility gaps appear
in the system. These are manifested by the appearance of a basice salt solid

*By H. F. McDuflie, Oak Ridge National Laboratory.
fTaken from material prepared by C. H. Secoy for the revised AEC Reactor
Handbook,

85
86 PROPERTIES OF AQUEOUS FUEL SOLUTIONS [cHap. 3

phase from dilute solutions and by the appearance of a uranium-rich
second liquid phase from more concentrated solutions. In both the =ult
and the second liquid phase, the uranium-to-sulfate ratio 1s found 1o he
greater than in the system at lower temperature; this suggests that hy-
drolysis of the uranyl ion is responsible for the immiscibility in cuch
instance. Hydrolysis can be repressed effectively by increasing the weid-
ity of the solution or, alternatively, by the addition of a suitable complexing
agent for the uranyl ion. Even the anions of the solute itsell muy be con-
sidered to accomplish this to some degree, since very dilute =olntions
hydrolyze much more extensively than more concentrated solution-.

Primary emphasis has been placed on the study of uranyl sulfare ~olution«
because of the superiority of the sulfate over other anions with respect th
thermal and radiation stability, absorption eross section for ncutrons,
ease of chemical processing, and corrosive properties. Other uranyt ~ui~
which have either been used in reactors or studied for possible use inchule
the nitrate, phosphate, fluoride, chromate, and carbonate. It hu~ foro.
found possible to improve the solubility characteristics of uranyl -
solutions at elevated temperatures by the addition of acids or salt= or th-
chosen anion.

The marked differences between light water and heavy water with re-
speet to moderating ability and thermal neutron absorption cross sectimn.
make solutions in both solvents of interest for reactor use. Generudiv
speaking, the upper temperature limit of solution stability occurs whon
10°C lower in heavy-water solutions than in light-water solutions.

Tetravalent uranium can be stabilized by increasing the reduction po-
tential of the solution. ITowever, tetravalent uranium is more rewdioy
hydrolyzed at clevated temperatures than hexavalent uranium, o
probably cannot be kept in solution except by the use of otherwisc ox-
cessively high coneentrations of acid.

Plutonium, the other fissile material, also forms salts which can be G-
solved in water. The posgibility of using such solutions in aqueots fun-
gencous reactor systems has been examined, and limited experinen:.
studies have been directed toward this goal but without substuntia! ~ -
cess (see Article 6-6.3).

Uranium—238 and thorium, the fertile materials, have been conside
for use in converter or breeder reactor systems. The solubility of v
is such that satisfactory aqueous solutions of uranium can be obtaie oy
use in the conversion of U23% to plutonium. Substantial effort= huve e
made to develop solutions of thorium which could be used us w0 bl
a two-region breeder reactor system.

Thorium appears to be stable in the tetravalent form but hus w ~tronz

dioxide is ultimately formed as the hydrolysis product. Thorium mtrate
3-2] SOLUBILITY RELATIONSHIPS (FISSILE AND FERTILE) 87

400 T

 

T T T T T
Unsaturated Solution [Lp) + Vapor

     
   
  
   
 

 

Solution
350

Two-Liquid Phases

300

250

200 — Unsaturated Solution

150 —

Temperature, °C

50 =\
' U0 25043 Hy0 + Solution

 

=50 |— lce + UD4550 4 3 Hy0 -

! | | I | |

0 1 2 3 4 5 6 7 8 9
m, moles/1000 g HpO

F16. 3-1. Phase diagram for the system U0280,-H20.

 

 

 

and thorium phosphate can be maintained in solution at satisfactory con-
centrations by the use of substantial concentrations of nitric or phosphoric
acids to inhibit hydrolysis. However, in the nitrate system an acceptable
breeding ratio could only be obtained by using N'°. Thorium phosphate
solutions containing the necessary amount of phosphoric acid are extremely
corrosive to all but the noble metals.

Neptunium and protactinium complete the listing of fissile and fertile
materials, since these are intermediates in the production of plutonium and
U233 from U?3® and thorium. Limited exploratory studies of their solu-
bilities have been carried out primarily in connection with the development
of processes for their continuous removal from blanket systems.

3-2.2 Uranyl sulfate. The solubility of uranyl sulfate in water and the
characteristics of the phase relationships at elevated temperatures, dis-
played in the form of a binary system, UO2804-H»0, are shown in
Fig. 3-1 [1]. It is necessary, however, to study the ternary system,
U03-803-H20, in order to understand the hydrolytic precipitation of the
38 PROPERTIES OF AQUEOUS FUEL SOLUTIONS [cuar. 3

UO3-2503-1.5H90
U03-2504-3H70
UO3-2503-6H20
UQO45S04-2H90
UO9504°3H90
UO3-HR0
+ & 3 Polymorphic Forms)
G = G Basic Salt
(5U03- 2503 yHoO)?
K = K Basic Salt
18UO3- 3503 xH20)?

mmg O W >

W

   
 
  
  

   
   
    
  
  

All Scales Are
Wt %

504

 

 

99.25 £

 

 

. y J
3 {c) 175°C UGz  SO3

Fi1c. 3-2. The system UQ3-S03-H20.

basic solid phase, 5—UQ3 - H20, which occurs in very dilute solutions at
clevated temperatures, and the position of the tie lines in the liquid-liquid
miscibility gap. Figure 3-2 shows portions of the ternary isotherms at
25, 100, 175, and 250°C [2,3]. A point of special significance in these dia-
grams is that the solubility of UQg3 in uranyl sulfate solutions decreases
with increasing temperature to the extent that at 250°C excess acid is re-
quired to maintain homogeneity in solutions of low concentration. Excess
acid also has a marked effect on the liquid-liquid miscibility gap, as shown
by the curves in Fig. 3-3 [4]. In very dilute solutions the surface formed
by these curves intersects the surface representing the liquid compositions in
equilibrium with the hydrolytically precipitated solid phase, 8-UQO3; - H30.
Figure 3—4 shows this intersection and three paths on the liquidus surface
at fixed SO3/U0O3 mole ratios [5].

Iigure 3-5 shows, from the data of Jones and Marshall [6], how the
two-liquid-phase separation temperature is lowered when the solvent is
changed from light water to heavy water. Scattered experiments suggest
that the temperatures for solid-phase separation through hydrolytic pre-
cipitation are also somewhat lower in heavy-water systems than in systems
containing light water as the solvent.
3-2] SOLUBILITY RELATIONSHIPS (FISSILE AND FERTILE) 89

 

45
0 | One Phase | © One Phase

at 525°C at 530°C

 

 

 

 

 

¢
g
2
o
@
o
E
- SO ]
U4 _ 4 408
U
325 _
50
4 1
U
300 [— —
$O
4 1000
U
- | | 1 |
0 5 10 15 20 25

Uranium, wt %

Fic. 3-3. Coexistence curves for two liquid phases in the system
U02804-H2804-H0.

Figure 3-6 shows the liquidus composition isotherms from 150 to 290°C
for dilute sulfuric acid solutions saturated with UO3 [7].

Study of the data leads to the following general conclusions with respect
to the stability of uranyl sulfate solutions of reactor interest:

(1) Stoichiometric uranyl sulfate solutions in light and heavy water are
unstable at temperatures of 280°C and above because of hydrolysis.

(2) Stability up to approximately 325°C is provided at uranium concen-
trations up to 2.5 w/o by the addition of a 50 mole 9, excess of sulfuric
acid.

(3) Stability up to as high as 400°C is provided at urantum concentra-
tions above 20 w/o by the addition of a 100 mole 9} excess of sulfuric acid.
90 PROPERTIES OF AQUEOUS FUEL SOLUTIONS [cuap. 3

 

375 .
|

\
\ LI+LH+V

 

350

325 M

a
’

Temperature

X

o

o
|
1

 

 

 

 

250

 

0 0.5 1.0 1.5 20

Uranium, wt %

F16. 3-4. Effect of excess HeSO4 on the phase equilibria in very dilute U02804
solutions.

310

 

 

 

 

[ 7
,
300 —]
W
°. - ]
Q
5 Hzo
:E_’ 290 —]
o
8 [}
E .
"= 280 D20 _|
0 1 2 3 4 5 6

m, moles U02504/1000 gm H40, D40

Fi1a. 3-5. Two-liquid phase region of uranyl sulfate in ordinary and heavy water.
3-2] SOLUBILITY RELATIONSHIPS (FISSILE AND FERTILE) 91

       
    

 

 

 

 

1.0 = | I i : | ' Two-Liquid Phase
- Region
0.1 =~ s R .
— Direct UO3 and SO 4 3
— Analyses —
£ B * Analyses Bosed on ]
c | pH at Room Temperature _]
2 ]
3 =
T
& 001 |— —
g - =
o — ]
U I =
q' - —
o [~ |
v} - .
o™
x —
0.001 +— —
0.0001 |
0.3 0.4 0.5 0.6 0.7 0.8 09 1.0 1.1 1.2 1.3 1.4

Solubility Mole Ratic UO3/Ho504

F1a. 3-6. Solubility of UO3 in H2S04-H20 mixtures.

 

 

 

 

 

400 ——
T TP 1 RERA l
| — —_—
380 |— —
360 }— _]
340 [~ —
o ‘i
s | |
>
$ 320 — —]
v
o L —
E S—
LY
" 300 }— —
— ‘ 0.42 - 0.45 m UO2504 Plus LipgSOy4 |
280 — | 1.3 m U050 Plus LigSOy4 |
| 2.2 mUQ 2S04 Plus Li9504
260 — 1 —
L —
|
sso Lt L LIl AR L 1|
0 0.02 0.05 0.1 0.2 0.5 1.0 2.0 5.0

LigSO4 , m

Fic. 3-7. Second-liquid phase temperature of U02804-Li2S04 solutions. Con-
centrations are uncorrected for loss of water to vapor phase at elevated temper-
atures.
02

PROPERTIES OF AQUEQOUS FUEL SOLUTIONS

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

{cHAP, 3
27°9.91 mole % Excess 13 - 2?325°CT  19.25 mole %
" fSulfuric Acid I f,f._‘ \_320 Excess Sulfuric Acid
¢ 305°C @ o i
it TNIOE Y s
300 1
— 63- i ‘?’: 310
25
S K {
0 5 10 15 20 0 5 10 15 20 0 5 10 15 20
15 U02504 wit % UO950, ,wt % UQ4504 ,wt %
| —340°C~—29.11 mole % Excess 15 |.355°C 3875 mole % Excess --- Temperature Contour Lines {isotherms)
335 Suifuric Acid v 3507 ~Sulfuric Acid v For Appearance of Two Liquid Phases
o / Region of Blye-Green Crystalline Solids
10 330\ E::]Region of Red Sclids
3 t 325x ¢, Reference Conposition of
g { 320 0.1 M CuSO4
% 5 \ 0.1 1 UO2504
o 315 < Reference Compaosition of HRE-2 Fuel
5 ( 0.005 M CuSOy
£ d 0.04 M UO2504
0 5 10 15 20 0 5 10 15 20 {Plus ~ 0.025 M H3504) (55%}
UQL50,4 ,wt % UO950,4 , wt %
uranyl sulfate, and sulfuric acid.

Fre. 3-8. Phase transition temperatures in solutions containing cupric sulfate,

 

 

 

 

 

 

 

0.04 - » * ’
(249) (<249 T(-<249) >\ | 7200 (186) (175)
-282 ’ H 215
220
Z 185
Z s 200
(261) Z
.03 — o ® @ ® 8—“
(249) (249) 7 (215) (197) (185)
-282/  \-282 229
2 3 CuO 503 2H20
2 UD3-503°5H20 7 {+ CuSO 4 Hy0)
€ { + Ni50O4-H;0) é
o 002 — 7 e ° ——t
@ 7. 1240) (211) (186)
2 44 216
v
%
A o A A A 1//
Z
0.01 |—- ’/,,//
“
Second Liquid Phose ///,/
%,
\ "
330 320 310
N\ \ \
0 0.01 0.02 0.03

 

0.04

Cu$SO4, m

Fra. 3-9. The effect of CuSO4 and NiSO4 on phase transition temperatures
0.04 m UO2804; 0.01 m Ha80,).
3-2] SOLUBILITY RELATIONSHIPS (FISSILE AND FERTILE) 93

The addition of lithium sulfate or beryllium sulfate to uranyl sulfate
solutions has been found to elevate the temperatures at which the second
liquid phase appears [8]. Figure 3-7 shows the effect of Li2SO4 additions
on the second liquid phase temperature for three uranyl sulfate solutions.
In very dilute uranyl sulfate solutions excess acid must also be added to
prevent hydrolytic precipitation.-

The solubility relationships in uranyl sulfate solutions containing cupric
copper are also of interest (see Article 3-3.4). Copper sulfate solutions,
like uranyl sulfate, undergo hydrolytic precipitation at elevated tempera-
tures [9]. IEven though the required concentration of cupric 1on may be
quite low, its presence influences the phase relationships. This influence
is most significant in dilute uranyl sulfate solutions. A complete phase
diagram for the four-component system, CuO-UO3-S03-H20, has not
been determined, but regions of special interest have been studied. Fig-
ure 3-8 shows the phase transition temperatures in solutions containing
copper sulfate, uranyl sulfate, and sulfuric acid. The solid phase which
appears at the higher CuSO4 concentrations has been shown to be at least
in part the basic copper sulfate, 3Cu0 - 8O3 - 2H20 [10].

In uranyl sulfate solutions in contact with austenitic stainless steels it
is important to know the effect of the corrosion products upon the solu-
bility relationships. Under most conditions iron and chromium appear as
insoluble hydrolytic products, but nickel appears as o soluble contaminant
of the solution. Studies have been made of the precipitation temperatures
for dilute solutions in the system UQ2804—CuS04-Ni304-H250 -0,
and the solid phases have been identified [11]. Iigure 3-9 summarizes
the information for systems having compositions approximately that of
the fuel solution of the HRE-2. In this preliminary study the tests were
limited to short time intervals (15 minutes or less of exposure to the ele-
vated temperatures).  When solutions containing 0.01 m CuSO. plus
0.01 m Ni8Oy, or 0.02 m CuS0; with no NiSO; were heated for longer
periods of time at 300 to 310°C (Just below the temperature for the forma-
tion of two liquid phases) green solids were deposited. Thus the results
pictured in Iig. 3-9 should be applied to practical situations with con-
siderable reservation until experiments with the exact composition of
interest have been conducted.

3-2.3 Other uranium compounds. Uranyl nitrate. A phase diagram for
the system uranyl nitrate-water [12] is shown in I'ig. 3-10.  Although
uranyl nitrate remains very soluble at the elevated temperatures of -
terest for power-reactor operation, the nitrate group in such solutions
decomposes to yield oxides of nitrogen which appear in the vapor phase.
Although this situation is reversible with the lowering of temperature, it
does introduce corrosion problems with respect to the vapor phase. The
04

PROPERTIES OF AQUEQUS FUEL SOLUTIONS

 

   

  
  
  

 

 

 

 

  
  

 

 

 

 

 

 

400 l
300 — ~ —
—_—— - —_
LA
250 — O Precipitation Limits -
® Vapor Phase Coloration
200 |— —
g
5
5
T 150 — ]
@
a
£
2
100 |— —
50 — 1
0 —t
_50 | | | | | J | | !
0 10 20 30 40 50 60 70 80 90 100
UO,INO3l,, Wi %
F1c. 3-10. The U02(NO3)2-H20 system.
400
Tl T T T sahd 1 Super Critical Fluid ' | T [ T
H——a—A—A—A b A& A g |
350 — ¥ UO2F7-2H20 + Liquid |
— N — .
- Basic Solid _,)(_3:. e §—e—n E
q00 |-2oMtien + e quid | + Liquid I
Saturated UG9F 2+ 2HAO
| Solution / ! 202°<M20%
® + Saturated
/ :
250 |— /0 Solution + o
=
o - 2 :
s r=>9
oN
g 200 f— ~3d
5 K “
- N -
g — Unsaturated Solution 0%
8 150 =7
- e
g /C
= — »
100 — O/
&
[~ CeBA Qrml Data f@'
50 | & Dato oi Dec:.n L w UO9Fp-2H90
© Data of Kunin + Soturated
- Soluti
0 A lce + Saturated Sol'n. o fi euhen
7 *-0-0 040 gt b
- lee 4 UOQFQ‘?HQO
so b L op o Lo by e
0 10 20 30 40 50 60 70 80 90

UOgFy Wt %

[cHAP. 3

F1a. 3-11. Phase equilibria of aqueous solutions of UO3 and HF in stoichiometric

concentrations.
3-2] SOLUBILITY RELATIONSHIPS (FISSILE AND FERTILE) 95
2.0 | T | T 1T T T

 

    
  
 
    

Solid Phase Is
UOz(H2P04)2‘3H20
{25°QC)

| Solid Phase Is
— UOZHPOA'AHZO
{25° — 400°C)

0.5

U, Molarity

0.2

 

 

 

01 2 5 10 20
PO4, Molarity

Fi1c. 3-12. Solubility of UO; in H3PO4 solution.

nitrate ion is, moreover, not completely stable in fissioning solutions;
elemental nitrogen is one of the products of radiation decomposition.
Although uranyl nitrate solutions have proved quite satisfactory in low-
power water-boiler type research reactors, where makeup nitric acid can
be added as needed [13], they do not appear attractive for high-tempera-
ture, high-power aqueous homogeneous reactors.

Uranyl fluoride. Uranyl fluoride is a very attractive fuel solute because
of the low neutron capture cross-section of fluorine. However, at high
temperatures it undergoes hydrolysis, which means that excess HF would
be required to maintain homogeneity. Hydrogen fluoride is also a com-
ponent of the vapor phase. Both liquid and vapor are very corrosive
toward zirconium and titanium, but less corrosive toward stainless steel
(see Article 5-3.3). Iigure 3-11 shows the phase relationships in this
system [14].

Uranium phosphate. Neither hexavalent nor tetravalent uranium phos-
phate is sufficiently soluble in water to be of reactor interest, but both
UO2 and UOgs are quite soluble in moderately strong phosphoric acid.
These solutions have been the subject of considerable study at the Los
Alamos Scientific Laboratory [15]. The solubility of UO3 in phosphoric
acid is illustrated by Fig. 3-12 [16]. Although the solubilities of uranyl
96 PROPERTIES OF AQUEOUS FUEL SOLUTIONS [crAP. 3

140 f
T T 1 1 ] [
Non-Binary
130 |— Solid + Liguid 1 . -]
A+
120 |— ' - & 1
A
- A
noF— g P —
L/’/I B + tiquid
100 |- r

Liquid

 

90—

    

70—

60—

Temperature, °C
|

50 |—

A+ B

A + Lliquid

A = UO,CrO 52 HpO —
B = UC2CrCy X HyO

 

A +lce

-10 | | | | | | i \

Q 10 20 30 40 50 60 70 80 920 100

 

 

 

 

UO,CrO4, wt %

Fi16. 3-13. The system UO2Cr0O4-H:0.

phosphate and uranyl sulfate in water are quite different, their respective
solubilities in concentrated phosphoric acid and coneentrated sulfurie acid
are analogous; in either case temperatures as high as 450°C can be ob-
tained with no phase separation. PPhosphorus has an advantage over sulfur
in possessing a somewhat lower neutron absorption cross section, but both
anions appear to be stable under radiation. Both the phosphate and sul-
fate solutions in concentrated acid at temperatures of 450°C are extremely
corrosive toward most metals and alloys except the noble metals. Attempts
to operate experimental high-temperature reactors wusing uranium
phosphate-phosphoric acid fuel solutions have failed because of catastrophic
corrosion rates due to imperfections in noble metal plating or cladding of
the reactor core and heat-exchanger tubing [17] (see Section 7-5).

Uranyl chromate. Uranyl chromate solutions also suffer from hydrolysis
at elevated temperatures; excess chromic acid is required for stability [18].
This system is, however, not unattractive insofar as corrosion of stainless
and carbon steels is concerned. The conditions of acidity and oxidation-
3-2] SOLUBILITY RELATIONSHIPS (FISSILE AND FERTILE) 97

 

1.000 T | ]

0.800 +—

   
 

Predicted 2375 psi COy

 

 

 

0.600
=z
™
3 ® 875 psi CO9
= o
375 psi COy
0.400 |— @ —
® Slopes:
@ 3.37 moles of Li per mole of U
* [ligCOy Is Solid Phase)
A 4.44 moles of Li per mole of U
® at Isobaric invariant
0.200 | | | | 1 | | |
0 0.02 0.04 0.06 0.08 0.0 0.12 0.14 016 0.8

UO,CO3, m

Fic. 3-14. Variation of Li2COg3 solubility with UO2CO3 concentration at con-
stant CO2 pressure (250°C).

reduction potential determine the valence state of the chromium, but
present knowledge 1s not adequate to specify required conditions for re-
liable behavior at elevated temperatures and under reactor radiation.
Figure 3-13 shows the phase diagram insofar as it has been established.

Uranyl carbonate. Uranium trioxide 1s quite soluble in alkali carbonate
solutions. This solubility can be attributed to the complexing of UO3 or
uranyl ion by the bicarbonate ion to form uranium-containing anions.
In any event, one would not expect the solubility of uranium to be retained
at high temperature unless the carbonate content of the aqueous phase
were kept high. This can be accomplished by retaining an adequately
high partial pressure of COs over the solution. The solubility of UO3 in
Li2CO3 solutions at 250°C has been studied [19], and the significant re-
sults are shown in Figs. 3-14 and 3-15. Referring to Fig. 3-14, we see that
at a constant COs2 pressure the concentration of uranium increases linearly
with the lithium concentration until a limit is reached at the isobaric in-
variant. The uranium concentration cannot be increased further unless
the CO2 pressure is increased. I'igure 3-15 is a projection of the composi-
tions of solutions saturated with respect to lithium and uranium at 250°C
and at a constant total pressure (CO2 4 steam) of 1500 psi. The projection
98 PROPERTIES OF AQUEOUS FUEL SOLUTIONS [cHAP. 3

 

 

10 AV N1 v 20
0 5 10 15 20
U03,wl %

Fig. 3-15. The system Liz0-U03-CO2-Hz0 at 250°C and 1500 psi.

figure gives no information concerning the concentration of carbonate in
the liquid phase. The region in which a single homogeneous liquid phase.
exists is the very narrow shaded region near the H.O apex. Although the
scope of this region is small, there should be no difficulty in maintaining a
homogeneous liquid phase if an adequate pressure of COs2 is kept on the
system and an appropriate composition is selected in preparing the solution.

3-2.4 Solubilities of nonuranium compounds. Thorium. Thorium solu-
tions having concentrations as high as 0.5 m would be useful for one-region
breeder reactors. For the breeder blanket of a two-region reactor con-
centrations of about 6.0 m thorium appear to be optimal, although some-
what lower concentrations would be of interest. At the present time only
thorium nitrate or phosphate solutions in the presence of excess acid have
been demonstrated to have the required solubilities at elevated tempera-
tures; both of these solutions have substantial disadvantages. Thorium
chloride would be expected, by analogy, to show substantial solubility at
elevated temperatures, but this system has not been investigated in
detail. Complex organic salts, such as thorium acetylacetonate, have high
solubilities at relatively low temperatures, but these have not been in-
vestigated for use in aqueous solutions at temperatures above 100°C.

Data from the literature on the solubility of thorium sulfate at low
temperatures both alone and in the presence of other solutes [20] indicate
that such solutions will probably not be satisfactory at elevated tem-
peratures.

Thorium phosphate (or thorium oxide) is very soluble in concentrated
phosphoric acid. Solutions eontaining up to 1100 g Th/liter with PO4/Th
3-2] SOLUBILITY RELATIONSHIPS (FISSILE AND FERTILE) 99

 

     

 

 

 

380 E I
NO3/U = 2.0
340
300 —
NO5/Th = 5.47
o
°_ 260
¢ NO3/Th = 6.65
=2
T
@
e 220 — —
& 220
Al
NO+/Th = 4.0
180 |- 3/ —]
140 |— "
100 | |
0 1.0 2.0 2.5

Thorium, Uranivm, m

Fia. 3-16. Hydrolytic stability of thorium nitrate and uranyl nitrate solutions.

ratios of 5 and 7 could be prepared and appeared to be thermally stable but
had high viscosities. Solutions containing 400 g Th/liter at PO4/Th ratios
of 10 were thermally stable at 250 to 300°C with viscosities little higher
than that of concentrated phosphoric acid. Efforts to improve the prop-
erties of thorium phosphate-phosphoric acid systems by the inclusion of
HF, HNO3z, H280,47, Se047, S04, Li*™, or Mg™ ™, alone or in combination,
have not proved encouraging [21].

The thorium nitrate-water system has been reported [22] as having
considerable solubility up to about 225°C, at which point hydrolytic pre-
cipitation occurs. Further investigation [23] revealed a maximal stability
for the 80 w/o material (to around 255°C). Increasing the acidity of the
solutions (increasing the NO3;~ /Th ratio) suppresses hydrolysis and in-
creases the stability of the solutions as indicated by Figure 3-16, which
shows the precipitation temperatures for various solutions [24]. The
intensity of vapor phase coloration at elevated temperatures (rapidly
reversible) increased as the nitrate/thorium ratio was raised above 4.0.

Plutonium. A considerable investigation of the chemistry of plutonium
in aqueous uranyl sulfate solutions has been directed, not toward the
achievement of solubility, but toward the achievement of znsolubility
in order to provide the basis for continuous processing of a U238 blanket
solution for plutonium production [25] (see Chapter 6).

The possible use of aqueous solutions of plutonium in homogeneous re-
actors has been reviewed by Glanville and Grant [26] in order to determine
TasLeE 3-1

TaE SorvuBiLiITY oF PruroNium CompPoUNDs AT RooM TEMPERATURE

 

PUIII

PuIV

PPyvz

 

 

 

Fluoride
Chloride
Bromide
Bromate
Todate
Perchlorate
Nitrate
Sulfate
Chromate
Phosphate
Carbonate

Oxalate

Benzoate

 

Soluble in presence of fluoride
complexing ions; e.g., Zr
Soluble in water and dilute acids

Soluble in water
No information

1.5mg Pu/literin 0.0017 M KIOj3,
0.17 M HeS0,

Of the order of grams of Pu/liter
in dilute HCI1O,

>7.7 g Pu/liter in 0.9 M HNO;

125 ¢ Pu/liter in 0.1 M H2S04

No information

Max. reported is 3.89 g Pu/liter
m 0.8 M H3POy, 0.9 M HCI
Soluble in 459 K2COj3

0.46 g Pu/liter in 0.5 M H2C20y,
3.7 M H*

Very soluble

 

Soluble in presence of fluoride
complexing ions; e.g., Zr or Al

Of the order of 50 g Pu/liter in
6 M HCI

Of the order of 1g Pu/liter in
5M HBr

>8 g Pu/liter in 1.5 M HoSOy,
0.15 M KBrQ,

Max. reported 1s 94.5 mg Pu/liter
in 0.1 M KIOs, 6 M HNO3

Of the order of grams of Pu/liter
in dilute HC10 4

500 g Pu/liter in 2 M HNO3

> 125 g Pu/liter in 0.1 M H2S0,

Soluble in 10 M HNOj;, 0.25¢
Pu/liter in 0.1 M Na2Crs07,
0.1 M HNOs3

0.55 g Pu/liter in 1 M H2S0,

0.1 g Pu/liter in 0.2 M Na2COj,
0.2 M CH3COONa

Max. reported is >0.244¢g
Pu/liter in 0.1 M H20204,
1 M HXNOg, 1 M HF

Very soluble

 

>40 g Pu/liter in 19 M HF

~350 g Pu/liter

No information

No information

0.6 g Pu/liter in 0.2 M KIO4

Of the order of grams of Pu/liter
in dilute HCIO4

~500 g Pu/liter

No information
No information

>0.7 g Pu/liter in 0.6 M H3POy,,
0.1 M HNO;
>8.4 g Pu/liter in 0.02 M NaoCOs4

Of the order of grams of Pu/liter

No information

 

 

SNOILNTOS TINA SN0ANDV 40 SHILIEIONd 001

€ "dvHD]
3-3] RADIATION EFFECTS 101

which compounds of plutonium appear most worthy of experimental study
as fuel solutes. Table 3-1 summarizes the available low-temperature solu-
bility information for three valence states of plutonium in the presence of
different anions.

Limited experimental work has been performed in which the solubilities
of plutonium carbonates, sulfates, and phosphates have been determined at
temperatures up to 300°C [27]. No substantial solubilitics have been
established at temperatures above 200°C.

Protactinium. No efforts have been made to achieve high solubilities
of protactinium in order to use it as a component of reactor fuel solutions.
Rather, the chemistry of protactinium has been examined in order to
devise processes for removing Pa?3? continuously from thorium breeder
blanket systems. A project was undertaken by the Mound Labora-
tories [28] to separate gram quantities of the longer-lived Pa*1 which
could be used in studies of the chemistry of protactinium.

Considerable information concerning the low-temperature chemical be-
havior of Pa has accumulated as a by-product of the development of
chemical processes for the separation of U??3 from irradiated thorium
materials [29].

Neptunium. Np239is in a class with Pa?33; no efforts have been made to
use it as a fuel solute, but consideration has been given to its formation in
and removal from blanket solutions of U238 [30a]. The chemistry of nep-
tunium has been reviewed by Hindman et al. [30b], and the hydrolytic
behavior has been reviewed by Kraus [30¢]. Continuous separation of
Np23¢ would provide a Pu?® product of high purity by radioactive decay,
whereas plutonium recovered from long-term irradiation of U®*® usually
contains appreciable amounts of Pu?*’, Spectrophotometric cells for use
at elevated temperatures and pressures in the study of the chemistry of
neptunium (and other materials) have recently been developed by Wag-
gener [30d] and have been used to measure the absorption spectra of
dilute neptunium perchlorate in its six-, five-, four-, and three-valence
states, using heavy water as the solvent. Dilute solutions of neptunyl
nitrate in nitric acid have been so studied at temperatures up to 250°C; the
pentavalent state was found to be stable under the test conditions [30e].

3-3. RapiaTioNn ErrecTs*

3-3.1 Introduction. Any aqueous reactor fuel solution will be subjected
to intense fluxes of high-energy radiations. The action of these radiations
both on the water and on the solute is of considerable importance in

*Taken from material prepared by C. J. Hochanadel for The Reactor Handbook.
102 PROPERTIES GF AQUEOUS FUEL SOLUTIONS [crAP. 3

reactor design and operation. Energy will be dissipated in a fuel solution
by the stopping of fast charged particles. These include mainly the fission
recoil particles, the recoil particles such as protons produced by elastic
neutron scattering, and the fast electrons resulting from the absorption of
gamma rays and from the decay of radioactive fission products. The
extent to which each contributes to the total energy absorbed in the fuel
solution depends upon the design of the reactor and the composition of
the solution.

Water is decomposed by all types of high-energy radiations to give
hydrogen, hydrogen peroxide, and oxygen [31]. If the decomposition
products are confined in solution, a radiation-induced back reaction will
occur and, eventually, steady-state concentrations (pressures) of products
will be attained. The rate of decomposition, the rate of the back reaction,
and hence the steady-state concentrations are sensitive to the conditions
of the system, such as the nature of the radiation, the type and concentra-
tion of solutes present, and the temperature. In particular, the addition of
hydrogen suppresses the decomposition of pure water.

The solutes may also be acted upon by direct absorption of the energy
of the radiations (or by transfer of energy from the solvent) and also by
reactions with the intermediate reactive species produced by the decompo-
sition of the water.

3-3.2 Primary and secondary reactions in pure water. The fast charged
particles give up energy to the electronic systems of the molecules of the
medium, thereby producing various excited and ionized states. In liquid
water, the ionized and excited molecules are rapidly transformed into the
free radicals H and OH. These are formed in relatively high concentrations
along the tracks of the fission recoils or other charged particles. As a
result, many of the radicals combine before they can diffuse apart, thereby
producing the stable molecules H20, Hs, and H20s. The primary chemical
species are therefore considered to be H, OH, Hs, and H202; their yields
per 100 ev of energy absorbed are expressed as G(H), G(OH), G(H2), and
G(H202). The primary chemical reaction can be written;:

3 H20—+H+OH+H2+H202 (3—1)

Some minor subtleties emerging from recent studies of the radiolysis of
aqueous solutions are: (a) although stoichiometry demands that G(H) +
2G(Hz) = G(OH) + 2G(H202), the yields of H and OH and also the
yields of Hz and H20» are not necessarily equal to each other [32]; (b) the
yields of He and HOz are lowered by solutes which scavenge the pre-
cursors in the particle tracks [33]; (¢) HO2 may be another “primary”’
chemical species produced in small yield in the particle tracks [34].
3-3] RADIATION EFFECTS 103

 

 

 

 

¥ Wil v 49 v
- vy
3 w3 5 2 32 57
40| E >83 2 3 ¢ 28 {20
& S8 o E £ e
- B—w ™ w o
“« GlH G{Hql —>
30 1.5
>
@
2 g
g 3
5 S
£ 3
L °
S 201 —10 E
5 £
© &
1.0 —0.5
™ =
________ o
0.1 1.0 10 102 103 104 10°

Initial Linear Energy Transfer,
kev/micron

Fic. 3-17. Yields of atomic and molecular hydrogen from the decomposition of
water by various ionizing radiations.

The yields of the primary chemical species depend markedly on the type
of radiation or, more specifically, on the energy transferred to the solution
per unit length along the track of the charged particle. The linear energy
transfer (LET) parameter varies from 5 X 10* kev per micron of path for
fission recoils to 0.2 kev/micron for fast electrons. The yields G(H2) and
G(H202) are largest for radiations such as fission recoils with large LET,
while the yields G(H) and G(OH) are largest for radiations such as fast
electrons with small LET [31]. This is illustrated in Fig. 3-17, where the
yields [35] G{H2) and G(H) are plotted as a funetion of LET.

The free radicals which escape immediate combination and diffuse into
the bulk of the solution may react with solutes present, including the
Ho and H20s. In water with no added solutes, the principal back reactions
of the free radicals are believed to be:

H + H202, — H20+ OH, (3-2)
104 PROPERTIES OF AQUEOQUS FUEL SOLUTIONS [cHAP. 3

OH+4+H;—HyO4+H, (3-3)
OH + Hs03 —> Hs0 4 HOs, (3-4)
9HO2 —»> H03 + O, (3-5)

H+4 0; — HOs, (3-6)

HO: + H203 —> Hy0 4 OH + Os. (3-7)

Reactions (3-2) and (3-3) provide a chain mechanism for the back reaction
of Hy and H20: to reform water [36], thereby leading to steady-state con-
centrations of decomposition products. The steady-state concentration
will depend on the relative yields of molecular products and free radicals
in reaction (3-1). For gamma rays, which produce the free radicals in
high yield and the molecular decomposition products in low yield, the
steady state in pure water is essentially zero. I'or fission recoils, which
produce essentially no free radicals to promote the back reaction, the
steady-state concentration (pressure) is very high (several thousand psi).
Reactions (3—4) and (3-5) provide a mechanism for decomposing H202 to
02, and reactions (3-3), (3-5), (3-6), and (3-7) provide a mechanism for
combining Ho and O2 to form water at higher temperatures [37].
Dissolved materials may be oxidized or reduced. In general, H atoms
usually reduce the solute and OH radicals reoxidize it. Assuming equal
numbers of H and OH, the net result depends on the action of the Hz2053.
The peroxide may act in either way (depending on the oxidation-reduction
potentials) but usually oxidation is favored. In the presence of Oz, the
H-atom may be converted to HO2, which usually acts as an oxidizing agent.

3-3.3 Decomposition of water in uranium solutions. In Table 3-2 are
listed the hydrogen yields from the decomposition of solutions of various
uranyl salts [38]. The yield depends on the type of radiation and on the
solute concentration, but is independent of temperature. Iigure 3-18
shows how the yield of hydrogen produced by fission recoil decomposition,
and by gamma-ray decomposition, decreases with increasing uranium con-
centration. This decrease may result from scavenging of H-atoms in the
particle tracks by the uranyl ions. Decomposition by fission recoil particles
produces mostly Hs (and an equivalent amount of H2O2 plus O2); the
yields G(H) and G(OH) are very small, probably in the range 0 to 0.1
per 100 ev.

In an aqueous homogeneous reactor fuel solution, the water is decom-
posed by fission recoils, proton recoils, and fast electrons. The rate of
hydrogen formation, in moles per liter per minute, can be expressed by
the equation:
3-3] RADIATION EFFECTS 105

 

  

o 98% Fission Recoil Energy

  

GlHg) , molecules/100 ev

0.5 —

v — Rays

 

 

| ! | |
o 1 2 3 4

Concentration, m

 

Fia. 3-18. The effects of uranium concentration and type of radiation on the
initial Hy yield from irradiated UO2S04 solutions.

d(Hz)

—2 = 0.00622[Gy X Wy+ G X Wyt Ge X W, (3-8)

where G 1s the hydrogen yield in molecules per 100 ev absorbed; and W is
the power density in kilowatts per liter. The subscripts f, p, and e refer to
the values for fission recoil particles; protons produced by neutron scatter-
ing, and electrons produced by gamma-ray absorption and by radioactive
decay of fission products. For an operating homogeneous reactor, about
969, of the hydrogen gas produced in solution is due to the fission recoil
particles, 29, to the neutrons, and 29, to the gamma rays. Therefore the
last two terms in Eq. (3-8) are usually neglected. The fraction due to
recoils is usually above 0.96, since part of the energy of the prompt neu-
trons, gamma rays, and radioactive decay escapes from the solution. The
value of G, for a given solute concentration can be obtained from Fig. 3-17
and the value for W, can be calculated from the neutron flux, the concen-
tration of fissionable atoms, the fission cross section, and the kinetic
energy of the fission recoils.

Along with the hydrogen, an equivalent amount of oxidant (either
peroxide or Og) will be formed.
106 PROPERTIES OF AQUEOUS FUEL SOLUTIONS [cHAP. 3

TABLE 3-2

IntTiaL RaTEs orF Hy Gas PropucTioN ¥rOM REACTOR-IRRADIATED
URANIUM SOLUTIONS

 

 

 

 

 

 

 

Concentration Fission ener
Solute &y pH G(H)
¢ U/liter | g U5, litey | 0TI €eT8y

0.399 0.372 0.688 1.61

4.03 3.76 0.957 3.26 1.66

18.6 1.63 0.906 2.90 1.48

38.1 0.274 0.619 0.95

40.7 37.9 0.995 2.42 1.53

102.1 37 .4 0.995 2.00 1.35

105.2 38.9 0.995 0.10* 1.20

108 .4 40.1 0.995 1.35

202.3 0.063 0.273 0.69

U02504 202.5 37.6 0.995 1.61 1.11
203 .4 189.6 0.999 1.11

227.0 1.63 0.906 0.98

310.4 0.096 0.364 0.62

386.0 1.63 0.906 0.80

431.3 37.8 0.995 1.32 0.77

436 .8 3.10 0.949 0.73

477 .2 0.148 0.467 0.56

713.5 33.5 0.995 0.56

796.0 37.4 0.995 1.03 0.49

4.25 3.96 0.959 4.25 1.63

40.1 37.3 0.995 3.32 1.58

118.8 37.1 0.995 2.98 1.36

UOzF, 272.0 37.2 0.995 2.64 1.11
377.0 39.3 0.996 1.35* 0.84

405.7 42.3 0.996 2.41 0.95

4.24 3.95 0.960 1.63

UO2(NO3)2 42 .3 39.4 0.995 2.05 1.5

318.0 2.29 0.932 1.03 0.6

420.1 36.9 0.99%4 0.60 0.55

42 .2 39.3 0.995 1.95 1.45

U(B04)2 92.5 35.1 0.996 0.1 1.25
350.0 32.0 0.995 0.1 0.75

 

 

 

 

 

 

 

 

*pH adjusted by adding acid.
3-3] RADIATION BFFECTS 107

3-3.4 Recombination in uranium solutions. For fission recoil particles
the radiation-induced back reaction of Hs, O2, and H202 is relatively
slow. Also, the thermal recombination rate in the absence of added catalyst
is relatively slow and, as a result, the steady-state pressure of gases is very
high; e.g. at 250°C the pressure is of the order of thousands of psi.

Certain solutes, notably copper salts, have been shown to act as homo-
geneous catalysts {39] in the thermal combination of hydrogen and oxygen
in aqueous uranium solutions. This provides a convenient method for re-
combining the radiolytic hydrogen and oxygen gases.

The reaction rate is first order in hydrogen and in copper concentration,
and independent of the oxygen concentration. The rate-determining step
is the reaction of hydrogen with the catalyst, and the activation energy
is about 24 kcal/mole. For a particular uranium solution, the rate of hy-
drogen removal, in moles/liter/min, can be expressed by

—d(Hy)

g = kou(Cu)(He), (3-9)

 

where k¢, is the catalytic constant in liters/mole/min, and concentrations
are given as moles/liter. Some selected values of kg, are listed in Table 3-3.
Increasing the concentration of uranyl sulfate or of sulfuric acid decreases
the catalytic activity of the copper somewhat, possibly as a result of com-
plexing. Also, the rate of reaction with Dy is about 0.6 that with Ha.

TaBLE 3-3

SELECTED VALUES OF kgy AT SEVERAL TEMPERATURES AND
Urantum ConcEnTRATIONS (Cu= 1073 M)

 

 

 

Uranium Temperature, kcy, 108 keo/S,
concentration, M °C liters/mole/min psi~!/min
0.17 190 4.3 0.28
0.17 220 26.6 2.3
0.17 250 90.0 12.
¢.00 250 83.0* 11.
0.01to0.1 250 133 18.
0.01t00.1 275 330 61
0.01tc0.1 205 850 149

 

 

 

 

 

 

*With 1073 M Cu(Cl0y4)2 plus HCIO4 in concentrations ranging from 0.005 M
t0 0.05 M.
108 PROPERTIES OF AQUEQUS FUEL SOLUTIONS [cHAP. 3

The concentration (solubility) [40] of Hs is related to the partial pres-
sure of hydrogen by the proportionality factor

— PH2
T (Hy'

 

S (3-10)

where Py, is given 1n psi, (H2) in moles/liter, and S in psi/liter/mole.
Equation (3-9) can then be written

—d(Hj)
dt

PHz
S

 

 

= kcy(Cu) (3-11)

At the steady state, the rates of formation and removal of hydrogen are

equal, and from Eqs. (3-8) and (3-11), the steady-state pressure is given by

P _0.0062 X Gy X WfXS‘
(I{Q)SS - (Cu)kcu

 

(3-12)

Application of copper sulfate catalyst to the suppression of gas evolution
during the operation of the Homogeneous Reactor Experiment was dis-
cussed by Visner and Haubenreich [41]. Design calculations for use of
copper catalysts in the HRE-2 and in other reactors have been reported
[42]. The use of internal recombination catalysts in homogeneous reactors
1s also discussed in Article 7-3.7.

3-3.5 Peroxide decomposition in uranium solutiens. The hydrogen per-
oxide produced by decomposition of the water can undergo several sec-
ondary reactions: (a) It can react with uranyl ion to form the slightly
soluble peroxide UO4 according to the reaction

UO:™* + Hy0: == UO4 + 2H™. (3-13)

The UO4 will precipitate if its solubility (~1073 M) is exceeded. (b) It
(or the UO4) can decompose by a radiation-induced reaction via the free
radicals H and OH produced by decomposition of the water. (c) In a re-
actor operating at high temperatures the HoO2 will decompose thermally
at an appreciable rate according to the over-all reaction:

HzOz —_— Hzo + 1/202 (3—14)
and the UO4 will decompose thermally according to the reactions
U0s—> UOs+ 1/202, (3—15)

UO3 -|— 2H* —> U02++ -|- HQO. (3—16)
3-3] RADIATION EFFECTS 109

Studies of the kinetics of the thermal decomposition of peroxide in uranyl
sulfate solutions [43] have shown the rate to be first order with respect to
peroxide concentration in the range from 0.4 to 5 X 1073 M, independent
of uranium concentration in the range from 4.5 X 1073 to 0.65 M, and in-
dependent of the acidity from pH 1.6 to 3.3. Traces of certain ions showed
pronounced catalytic effects. The rate of decomposition could be ex-
pressed by

—d(H202)

o = k(H202) + keat(Cat)(H203), (3-17)

where (H202) represents total peroxide concentration (UO4+ HoO2) in
moles per liter at time ¢, k is the molar rate constant in the absence of
catalyst, kat is the catalytic constant, and (Cat) is the concentration of
catalyst in moles per liter.

Values of k for uranyl sulfate solutions (4.5 X 10723 M to 0.65 M) with
no added catalyst depended upon the adventitious impurities present, and
were in the ranges as listed in Table 3-4. The indicated activation energy
is 25.5 keal/mole. For pure water, the rate constant at 78°C was 4.5 X 10~
per minute.

Catalytic constants, keat, for various ions added are listed in Table 3-5.

The net rate of peroxide formation is the difference between the rate of
production and the rate of decomposition. At the steady state the two rates
are equal. The rate of formation of peroxide in an operating reactor in
terms of moles per liter per minute can be expressed in terms of the yield,
G, in molecules per 100 ev of fission recoil energy, and the average fission
recoll power density of the reactor, W, in kilowatts per liter, by the
equation:

d(H2052)

s 0.0062 X G X W x 0.96. (3-18)

The maximum value of G for the particular solution used is that given
for the Hs yield in Fig. 3-18. During reactor operation the radiation-
induced decomposition of peroxide is negligible, and the rate of decom-
position is essentially the thermal rate given by Eq. (3-17). At the steady
state the peroxide concentration is given by

0.0062 X G X W X 0.96

(H202)ss = i{?+ kcat(Cat)

 

(3-19)
110 PROPERTIES OF AQUEOUS FUEL SOLUTIONS [cHAP. 3

TABLE 3-4

MoLar RaTE CONSTANTS FOR PEROXIDE
DecoMposiTiON IN U02504 SOLUTIONS

 

 

 

 

 

 

 

Temperature, °C k, min—!
53 1.2x1073to 12x 1073
78 1.8 1072t0 14 x 102
100 1.8x 107 t0 12 x 1071
TaBLE 3-5

CaTaLyTic ConsTanTs AT 100° C ror PEROXIDE
DrecoMposiTioON BY Various IoNs ADDED TO
T02504 SoLuTION

 

 

 

Catalyst keat (liters/mole/min)
Fet2 135,000
Rut4 121,000
Agtl! 4,500
Ni+2 2,600
Cut? 1,200
Fet2 (promoted 502,000
by 793 ppm Cu*2)

 

 

 

 

The maximum allowable power density, Wmax, before precipitation of
uranium peroxide occurs is given by

_ (H202)SOI ”C + kcat(cat)]

Woax = =G 0062 X G X096 (3-20)

 

For example, in 0.17 M UO2zS02 solution at 100°C with no added catalyst,
the peroxide solubility is =4 X 1073 M, k = 1 min™!, and G = 1.5, then
Wax = 0.4 kw/liter.

Following reactor shutdown, peroxide formation and decomposition will
result from the delayed neutrons and from the 8~ and # radiation of the
fission products. The yield for peroxide formation will be essentially that
for y-rays, ¢ = 0.46. The yield for radiation-induced decomposition [37]
may be as high as 4.5, but will depend in a complicated way on the amount
of oxygen, hydrogen, and other solutes such as fission products, corrosion
products, etc., present.
3-4] PHYSICAL PROPERTIES 111

3-3.6 Decomposition of water in thoriura solutions. Under radiation
thorium nitrate solutions decompose [44] to give Hg, H209, and Og from
decomposition of the water, and Os, N2, and oxides of nitrogen from de-
composition of the nitrate. Yields of Hz and N2 for several types of radia-
tion, and for several concentrations of thorium nitrate, are given in
Table 3-6. The hydrogen yield decreases with increasing solute concen-
tration, the same as for uranium solutions. The nitrogen yield increases
with increasing nitrate concentration. The Np is presumably formed by
direct action of radiation on the nitrate. The N yield is greater for radia-
tions of greater LET. The N2 yield is independent of temperature, and
little or no radiation-induced back reaction takes place.

Uranyl nitrate solution also decomposes to give N2 in yield comparable
to that for thorium nitrate solution.

TABLE 3-6

Tae ErreEcts oF CONCENTRATION AND TYPE OF RADIATION ON
THE YIELDS OF No AND Hs 1N THE DrcoMrosiTioN oF THORIUM
NITRATE SOLUTIONS

 

 

 

 

G(I) G(N2)
Th(NO2)s, | piggion
molality ook Fission ORNL graphite
recoils i . . Gamma rays
recoils® pile radiation
0.26 1.11 0.002 0.003 0.04x 103
0.55 0.93 0.016 0.003 0.5 x10°3
1.5 0.51 0.047 0.005 1.5 x1073
2.7 0.33 0.063 0.006 1.1 x10°3
7.2 ~0.087 0.16

 

 

 

 

 

 

 

*0.14 molal enriched UO:804 added to make the energy contributed by fission
recoils > 959, of the total energy absorbed.
fEstimated by extrapolation.

3-4. PHYsI1cAL PROPERTIES

3—4.1 Introduction. Knowledge of the physical properties of aqueous
solutions of reactor fuel materials 1s required for nuclear physies calcula-
tions and analysis of reactor performance, for engineering design, and,
ultimately, for effective reactor operation. The scarcity of information
available in 1951 concerning the properties of uranium salt solutions
prompted the Homogeneous Reactor Program at ORNL to sponsor a
physical properties research program at Mound Laboratory beginning in
112 PROPERTIES OF AQUEOUS FUEL SOLUTIONS [cHAP. 3

July 1951 and carrying through December 1954. The progress of this
effort is discussed in regular reports [45] and in a number of topical reports
dealing with techniques, apparatus, and summarized data [46-56).

A number of compilations of physical properties data for aqueous reactor
solutions have appeared, among which are included those of Van Winkle
[57], Tobias [58], and sections by Briggs, Day, Secoy, and Marshall in
The Reactor Handbook [59].

The properties of light and heavy water are discussed by Lottes [60] as
they relate to reactor heat-transfer problems. Other properties of water
are found in standard reference works [61-62].

The remainder of this section is devoted to particular properties of
reactor solutions which are of interest and to some properties of the vapor
phase above reactor solutions which are important for aqueous homo-
geneous reactors.

 

 

 

 

TaBLE 3-7
Li1Quip AND Varor DENsITIES oF D20
Density, g/ce

T, °C

Vapor Liquid
175 0.004 (.989
180 0.005 0.983
190 0.006 0.970
200 (0.007 0.957
210 0.009 0.943
220 0.010 0.929
230 0.013 0.913
240 0.016 0.898
250 0.020 0 881
260 (0.024 (.864
270 0.029 0.847
280 0.034 0.829
290 0.040 0.809
300 0.048 0.787
310 0.058 0.763
320 0.070 0.735
330 0.087 0.705
340 0.105 0.668
350 0.129 0.626
360 0.163 0.573
370 0.248 0.462
371.5 0.363 0.363*

 

 

 

 

 

*Critical point.
3-4] PHYSICAL PROPERTIES 113

3-4.2 Density of heavy water and uranyl sulfate solutions. The density of
heavy water has been measured up to 250°C at Mound Laboratory by a
direct method making use of a Jolly balance [51,54]. An indirect method
for determining the density of heavy-water liquid and vapor has been used
to extend the data up to the eritical temperature [63]. Table 3-7 gives
density values at convenient temperature intervals. The densities of
uranyl sulfate solutions from 20 to 90°C and at concentrations up to 4.0
molal were measured by Jegart, Heiks, and Orban at Mound Laboratory
[47,56]. The densities of uranyl sulfate solutions were measured by
Barnett et al., of Mound Laboratory [35] at temperatures up to 250°C for
concentrations (at room temperature) of 60.6 and 101.0 g U/liter of solu-
tion in light water and for uranium concentrations (room temperature)
of 20.3, 40.4, and 61.2 g/liter in heavy water. Their data are presented in
Table 3-8.

TABLE 3-8

DensiTies oF LicaT- AND HEAVY-WATER SOLUTIONS OF
URANYL SULFATE

 

 

 

 

 

Density, g/ml
T OC UOzSO4 n H20, g U/liter U02804 in D20, g U/liter
60.6 101.0 20.3 O?; 40.4 O?; 61.2
psi psi
30 1.1340 1.300 1.1567 1.1842
1.318 300 | 1.1598 | 280
45 1.0715 1.1275 1.1245 1.1509 1.1781
1.1263 305 | 1.1540 280
60 1.0649 1.1199 1.1170 1.1433 1.1704
1.1198 | 325 1.1472 | 300
75 1.0562 1.1111 1.1120 340 1 1.1388 325 1 1.1628
90 1.0468 1.1008 1.1026 360  1.1297 350 | 1.1545
100 1.0395 1.0952 1.0954 380 | 1.1228 365 | 1.1476
125 1.0203 1.0742 1.0751 | 415 1.1037 | 410 | 1.1278
150 0.99%4 1.0528 1.0505 4580 | 1.0822 480 | 1.1053
175 0.9735 1.0293 1.0221 o560 | 1.0572 530 | 1.0805
200 0.9460 1.0030 0.9920 680 ¢ 1.0280 670 | 1.0540
225 0.9156 0.9745 0.9578 830 : 0.9973 840
250 0.9440 0.9224 | 1090 | 0.9610 { 1080

 

 

 

 

 

 

 

 

 

 

*Overpressure of oxygen gas as indicated. Otherwise solutions were in contact
with their own vapor.
114 PROPERTIES OF AQUEOUS FUEL SOLUTIONS [craP. 3

The densities of light-water solutions of uranyl sulfate were found by
Marshall [64] to fit the relationship

1

%= (78.65/0) — 1.046

+ dn,0,

where d, is the density of the solution in g/cc, U is the weight percent
uranium, and dp,o 1s the density of water at the same temperature and
pressure as the solution. The accuracy of this formula is believed to be:

+ 0.59% from 25 to 300°C for 0-109;, U,

4+ 1.09, from 120 to 250°C for 10-509], U,
+ 2.09, from 25 to 125°C for 10-509, U,
4+ 2.09, from 250 to 280°C for 10-509; U.

The density of heavy-water solutions of uranyl sulfate may be estimated
from an analogous formula:

1

&= F1070) —0.944

~+ dp.o.

The densities of heavy-water solutions of uranyl sulfate reported in Table
3-8 were found to be within 39 of values calculated by this formula.
Similar formulas were devised by Tobilas [58] for application to the
determination of the densities of solutions of mixed solutes such as uranyl
sulfate-beryllium sulfate and uranyl sulfate-lithium sulfate.
Density information for other uranium salts and for thorium nitrate so-
lutions was compiled by Day, Secoy, and Marshall [59].

3—4.3 Viscosity of D3O and uranium solutions. The viscosity of heavy
water was measured from 30 to 250°C by Heiks et al. [564]. Good agree-
ment with four values reported by Hardy and Cottington [65] was found.
The apparatus used has been described by Heiks et al. [51] and the elec-
tronic instrumentation for measuring the time of fall of a plummet con-
taining a radioactive pellet has been deseribed by Rogers et al. [52].

Van Winkle [57] discussed the viscosity of light and heavy water and
the early data for solutions of uranyl sulfate in heavy water.

Heiks and Jegart of Mound Laboratory [50,56] measured the viscosity
of uranyl sulfate solutions in light water over a concentration range of
0.176 to 2.865 molal and over a temperature range of 20 to 90°C by the use
of Ostwald capillary viscometers. Using the falling-body viscometer re-
ferred to above, Barnett et al. of Mound Laboratory have measured the
viscosities of light- and heavy-water solutions of uranyl sulfate at tem-
peratures up to 250°C [55]. Table 3-9 presents a comparison of viscosities
34 PHYSICAL PROPERTIES 115

TABLE 3-9

ViscosiTies oF LicHT- AND HEAVY-WATER SOLUTIONS OF
URANYL SULFATE

 

Viscosity, cp

 

 

 

 

T, °C U0280, in H:0, g U/liter V02804 in D20, g U/liter
20.2 | 59.7 [100.41201.4|301.1|400.7 0 20.3 | 40.4 | 61.2

30 |0.85710.946:1.04 {1.40 [1.97 |3.04 |0.969| — — —
45 10.629|0.701 10.773|1.05 [1.39 [1.98 [0.713|0.773|0.7890.827
60 [0.492]0.536|0.605[0.820:11.06 |1.41 |0.552|0.586|0.6080.638
75 10.4030.436 | 0.483|0.638|0.821!11.10 [0.445|0.467 {0.487(0.513
90 10.329,0.365]0.399|0.517|0.675[0.884 10.365|0.38410.397|0.415
100 [0.292(0.327 1 0.3510.454|0.589|0.76710.323|0.342:0.350|0.365
125 [0.238|0.264 1 0.27310.347|0.44210.543,0.252|0.263|0.269 | 0.287
150 [0.19410.213/0.223]0.275|0.335(0.437:0.208|0.216/0.222|0.233
175 {0.164 |0.17810.188[0.226|0.269[0.360/0.17510.182(0.187 | 0.196
200 (0.142]10.153]0.158(0.192]0.230(0.298|0.151[0.15410.160|0.169
225 0.136|0.141 0.209]0.264(0.135,0.138|0.144 0.151
250 0.12510.130 0.19010.23810.124[0.125(0.133 |0.137

 

 

 

 

 

 

 

 

 

 

 

 

 

of light- and heavy-water solutions for selected temperatures and uranium
concentrations.

The viscosity of uranyl nitrate solutions is discussed by Day and
Secoy [59].

3-4.4 Heat capacity of uranyl sulfate solutions. Van Winkle [57] esti-
mated the heat capacity of dehydrated uranyl sulfate by comparison with
uranyl nitrate and with salts of other metals and has estimated the heat
capacity of solutions of uranyl sulfate in heavy water and in light water.
The effect of temperature on the heat capacity of solutions was assumed
to be the same, percentagewise, as the effect on the heat capacity of the
pure solvent. Table 3-10 shows the influence of uranyl sulfate concentra-
tion upon the heat capacity of solutions at 25 and 250°C. No experimental
measurements have been reported for temperatures above 103°C; values
below this temperature differ from the estimates by as much as 109, [57].

3—4.5 Vapor pressure of uranyl sulfate solutions. The vapor pressures
of uranyl sulfate solutions have been measured over the temperature range
24 to 100°C and at uranium concentrations up to 4.8 molal by Day [66].
The ratio of the vapor pressure of the solution to that of pure water at the
116 PROPERTIES OF AQUEOUS FUEL SOLUTIONS [cHAP. 3
TasLe 3-10
HEsTiMATED HEAT CAPACITIES OF URANYL SULFATE SOLUTIONS

Estimated Heat Capacity, eal/g -°C

 

 

 

 

H20 solutions 150 solutions
U080y, w/o

25°C 250°C 25°C
0 0.998 1.166 1.005
10 0.905 1.070 0.916
20 0.809 0.975 0.830
30 0.714 0.875 0.745
40 0.619 0.770 0.658
50 0.523 0.665 0.568
60 0.428 0.550 0.474
70 0.333 0.425 0.370
80 0.238 0.290 0.252
85.9 (U050, - 3D20) — — 0.174

87.2 (U0;804 - 3H0) 0.170 — —

 

 

 

 

 

 

same temperature is almost independent of temperature but markedly de-
pendent upon the solute concentration over this range. Table 3-11 shows
the effect of uranium concentration upon the ratios at temperatures up to
100°C. Vapor pressure measurements have been made for light-water
solutions of uranyl sulfate at temperatures up to approximately 200°C and
uranium concentrations (room temperature) of 400 and 500 g/liter. Data
have, however, not yet been published [45]. Other vapor pressure meas-
urements are being made at the Oak Ridge National Laboratory. The
vapor pressure lowering in uranyl sulfate solutions containing added
lithium or beryllium sulfate was stated to be approximately in accord
with Raoult’s law [8b].

The vapor pressures of three solutions of UOj3 in phosphoric acid are
shown in Table 3-12 [17], based on values read from the published curve.

3-4.6 Surface tension of uranyl sulfate solutions. The capillary rise
method was adapted to conditions prevailing at elevated temperatures and
pressures for measurements made at Mound Laboratory by Heiks et al.
[51,54]. Briggs [569] has compiled information on the surface tension of
aqueous solutions of uranyl sulfate, including the relationships established
by Van Winkle.
3-4] PHYSICAL PROPERTIES 117

TasLE 3-11

RaTios oF VAPOR PRESSURES OoF URANYL SULFATE
SOLUTIONS TO THAT OF Pure H.0

 

 

 

 

 

 

 

 

 

 

p/po (at given temperature)
U02804, M

25°C 50°C | 80°C 100°C
0.59166 0.990 0.990 0.991 0.993
1.14545 979 981 .984
3.17433 911 920 .935
4.1821 .863 .881 .900

TABLE 3-12

Varor Pressures or UOsz SoLvtions 1N HzPO4*

 

 

 

 

Vapor pressure, 1b/in?
Temperature,
°C 0.76 M UO3in | 0.309 M UOsz in 0.75 M UQOs in
0.56 M H3PO4 2.90 M H3PO, 7.5 M H3PO,4
110 23 21 21
150 66 64 64
200 200 200 200
250 460 500 500
300 1000 1150 1150
350 1700 2100 2100
400 2900 3600 3700
450 4600 2600 6600
500 6600 8000 —

 

 

 

 

 

 

*Plotted and extrapolated from Los Alamos data.

Barnett and Jegart of Mound Laboratory [53] reported the surface
tension of uranyl sulfate solutions in light water at temperatures up to 75°C
and concentrations up to 2.34 molal.

Barnett et al. [55], using the apparatus described by Heiks et al. [51]
for capillary rise technique at elevated temperatures and pressures, meas-
ured the surface tension of light- and heavy-water solutions of uranyl
sulfate at temperatures up to 250°C.
118 PROPERTIES OF AQUEOUS FUEL SOLUTIONS [cHaP, 3

TaBLE 3-13
Tue pH or UO3-H2S504-H20 Sovrutions at 25.00°C*

 

 

 

 

) Molar ratio, U03/S04

Molarity
H2504

0.7006 | 0.8017 | 0.9024 1.0029 1.1042 1,202
0.02432 1.995 2.151 2.425 3.043 3.612 3.823
0.01214 2.259 2.422 2.692 3.267 3.731 3.922
0.004866 2.619 2.784 3.046 3.526 3.874 4038 |
0.003656 2.730 2.900 3.155 3.604 3.913 4074
0.002437 2.905 3.065 3.321 3.708 3.977 4.124
0.001216 3.174 3.348 3.592 3.906 4.093 4,214
0.0004850 3.587 3.715 3.903 4.130 4.250 4 354
0.0003646 3.671 3.822 3.996 4.201 4.304 4.395
0.0002424 3.851 4.004 4.127 4,326 4.411 4.500

 

 

 

 

 

 

 

 

*These values are based on a determined pH of 2.075 for 0.01000 molar HCI,
0.900 molar KCI. For pH’s relative to the analytical concentration of HCl subtract
0.075 pH unit from each of the above values.

TasLE 3-14

Tue pH VaLves or HCl, H2S04, anp UO2S04 SoLuTiON
AS A FuNcTioN oF TEMPERATURE

 

 

 

M 1°C pH 1°C pH

HCI 0.0100 30 2.13 130 217
H2S04 0.00490 30 1.76 180 1.93
" 0.04615 30 1.28 180 1.62

" 0.09740 30 0.96 180 1.21
T02S5047 0.04 30 2.80 150 258
7 0.17 30 2.20 180 2.20

" 1.1 30 1.76 180 1.81
U0:804* + HeS0, 30 1.22 180 1.83

 

 

 

 

 

 

 

*U02804 solution which read pI 2.68 at 30°C, to which was added sufficient
H,30, to bring the pH reading down to 1.22 at 30°C.

tReported values modified and extended by personal communication with M. H.
Lietzke, April 1958.
34] PHYSICAL PROPERTIES 119

 

22 I T T T T T T

20

  
    

Nitrogen

Hydrogen

Henry's Law Constant, K x 1073, psia/mole fraction

 

 

I ! | | I 1 I
0 40 80 120 1460 200 240 280 320
Temperature, °C

 

Fic. 3-19. Solubility of gases in water.

3—4.7 Hydrogen ion concentration (pH). Orban of Mound Laboratory
[48 56] studied the acidity of uranyl sulfate solutions from 25 to 60°C.
Orban has also reported the pH values for uranyl sulfate solutions con-
taining excess UQO3 at temperatures up to 60°C [49,66]. Secoy has com-
piled information concerning the effects of uranium sulfate concentration
upon pH as reported by a number of investigators [59]. Marshall has
utilized the relationship between pH and concentration to determine the
solubility of UOj3 in sulfuric acid at elevated temperatures [7]. Table 3-13
shows the effect of sulfate concentration on pH at 25.00°C for various
ratios of UO3 to sulfate as found by Marshall.

The direct measurement of pH at elevated temperatures and pressures
has been made possible through the development by Ingruber of high-
temperature electrode systems for use in the sulphite pulping process of
the paper industry [68]. Lietzke and Tarrant have demonstrated the
applicability of Ingruber’s electrode system to the measurement of the pH of
120 PROPERTIES OF AQUEOUS FUEL SOLUTIONS [cHAP. 3

\ [ l W f l ( '
T = Initial Temperature ]
I
T = 500°F T=500°F | 1
|

 

M >

o o

o <
\

ho

Ko

o
l

 

 

 

 

T=475°F
2 200 |-
a
£ 80
3
N ]
o~
2 5
s 140 |— 2
e I o
2120 [t £
& | z
T 100 H- ]
< ||
o
& 80 [H |
£ 1\
£ 60 Y ]
\\\
40 j\\ // —
\Q\\\\. T=300°F -/
20 AQ\ o s -
&_._..—; | No Reaction Below Curves | | ‘ |
_— 1 L
0 10 20 30 40 50 60 70 80 90 100

Initial Mole % Qg in H0 Free Gas

F1c. 3-20. Approximate explosion limits of spark-ignited gaseous mixtures of Os
and Hg saturated with H,0 vapor in 13-in. bomb.

solutions of HCl, UO2S04, and H2S04 at temperatures as high as 180°C [69].
Table 3-14 lists the values reported. Lietzke has reported the properties
of the hydrogen electrode-silver chloride electrode system at high tempera-
tures and pressures [70].

3—4.8 Solubility of gases. The solubilities of various gases in water and
in reactor solutions are indicated by Fig. 3-19 [71,72].

Composition and PVT data [73,74]. Dalton’s Law of additive partial
pressures has been determined to correlate the P-PVT relationships of
steam-oxygen and steam-helium mixtures in the pressure range of interest
(up to 2500 psi) to within 19.

For approximate calculations the perfect gas laws can be applied when
dealing with the permanent gases under reactor conditions, but values for
the properties of water vapor and D20 vapor should be obtained from
experimental data or standard tables [62].

The compositions of saturated steam-gas mixtures are predictable in a
similar manner, in that the partial density of each constituent is equal to
that of the pure constituent at its partial pressure.

3—4.9 Reaction limits and pressures. Stephen et al. [75] measured the
effect of temperature and Hz/O2 ratio on the lower reaction limits of the
H2-0O»-steam system in a small 1.5-in.-diameter autoclave. Their investi-
34] PHYSICAL PROPERTIES 121

 

T 1 T 71

50

 

   

Reflected Detonation \
{Caleulated) ’

20 |

Detonation
{Calculated)

l
I
|
l
l
|

P/Pq
T 11

I

 

 

 

 

— Explosion ]
{Calculated) /
2 |- /Observed _|
/
/
-
—-“/
—

1 | | ] | | | | | 1 i | | !

0 12 24 36 48 60 72 84

Mole Percent Knallgas, Vi

Fia. 3-21. Ratio of peak reaction pressure to initial mixture pressure vs. com-
position of Knallgas saturated with water vapor at 100°C. Tube diameter = 0.957
in., spark ignition energy = 180 millijoules.

gation indicated that the composition (mol basis) of the mixture at the
lower limit was not temperature dependent at constant Hz/O2. Figure 3-20
is a plot of Stephan’s data showing variation in Ha/O2 [76].

Syracuse University is investigating [77] further the effects of geometry,
temperature, and method of ignition on the reaction and detonation limits
of Ho0-(2H2+ O2). Reaction pressures are also measured. Iigure 3-21
shows their reported results in a long 0.957-in.-diameter tube at 100°C;
these are typical of results obtained at 300°C. It is interesting to note
that their lower reaction limit is 69, Knallgas,* as compared with 209,
Knallgas observed by Stephan.

*Knallgas is the term describing the 2:1 mixture of hydrogen and oxygen ob-
tained electrolytically and considered as a single gas.
122 PROPERTIES OF AQUEOUS FUEL SOLUTIONS [cHAP. 3

REFERENCES

1. C. H. Skcoy, J. Am. Chem. Soc. 72, 3343 (1950).

2. C. H. Srcoy, Survey of Homogeneous Reactor Chemical Problems, in
Proceedings of the International Conference on the Peaceful Uses of Atomic Energy,
Vol. 9. New York: United Nations, 1956. (P/821, p. 377)

3. J. S. GiLvu et al., in Homogeneous Reactor Project Quarterly Progress Report
for the Period Ending October 31, 1953, USAEC Report ORNL-1658, Oak Ridge
National Laboratory, Feb. 25, 1954. (p. 87)

4. H. W. WricHT et al., in Homogeneous Reactor Program Quarterly Progress
Report for the Period Ending Oct. 1, 1952, USAEC Report ORNL-1424(Del.),
Oak Ridge National Laboratory, Jan. 10, 1953. (p. 108)

5. W. L. MagsuaLL and C, H. Secoy, Oak Ridge National Laboratory, 1954,
Unpublished.

6. E. V. Jo~nEs and W. L. MarsHALL, in Homogeneous Reactor Project Quarterly
Progress Report for the Pertod Ending Mar. 15, 1952, USAEC Report ORNL-1280,
Oak Ridge National Laboratory, July 14, 1952. (p. 180)

7. W. L. Marsuarr, The pH of UOs-HS804+H 0 Miztures at 25°C and Iis
Application to the Delermination of the Solubility of UOs in Sulfuric Acid at
Elevated Temperature, USAEC Report ORNL-1797, Oak Ridge National Lab-
oratory, Nov. 1, 1954.

8. (a) R. 8. GreeLey, Oak Ridge National Laboratory, 1956, personal com-
munication. (b} R. S. GREELEY et al., High Temperature Behavior of Agueous
Uranyl Sulfate— Lithium Sulfate and Uranyl Sulfate—Beryllium Sulfate Solutions,
paper presented at the 2nd winter meeting of the American Nueclear Society,
New York, Oct. 28, 1957.

9. E. Posnsak and G. TuneLL, Am. J. Sci. 218, 1 (1929).

10. F. E. Cuark et al, in Homogeneous Reacltor Project Quarterly Progress
Report for the Period Ending Apr. 30, 1956, USAEC Report ORNL-2096, Oak
Ridge National Laboratory, May 10, 1956. (p. 130)

11. C. H. Secoy et al.,, in Homogeneous Reactor Project Quarterly Progress
Report for the Period Ending July 31, 1957, USAEC Report ORNL-2379, Oak
Ridge National Laboratory, Oct. 10, 1957 (p. 163); F. MosELEY, Core Solution
Stabilsty in the Homogeneous Aqtieous Reactor: The Effect of Corrosion Product
and Copper Concenirations, Report HARD(C)/P-41, Gt. Brit. Atomic Energy
Research Establishment, May 1957.

12. W. L. MarsHALL et al., J. Am. Chem. Soc. 73, 1867 (1951).

13. L. D. P. KiNg, Design and Description of Water Boiler Reactors, in
Proceedings of the International Conference on the Peaceful Uses of Atomic Energy,
Vol. 2. New York: United Nations, 1956. (P/488, p. 372)

14. W. L. MarsHALL et al., J. Am. Chem. Soc. 76, 4279 (1954).

15. B.J. TuaMmer et al., The Properties of Phosphoric Acid Solutions of Urantum
as Fuels for Homogeneous Reactors, USAEC Report LA-2043, Los Alamos Scien-
tific Laboratory, Mar. 6, 1956. D. Froman et al., Los Alamos Power Reactor
Experiments, in Proceedings of the Internaivonal Conference on the Peaceful Uses
of Atomic Energy, Vol. 3. New York: United Nations, 1956 (P/500, p. 283).
L. D. P. KiNag, Los Alamos Power Reactor Experiment and Iis Associated Hazards,
REFERENCES 123

USAEC Report LAMS-1611(Del.), Los Alamos Scientific Laboratory, Deec. 2,
1953; A Brief Description of a One Megawatt Convection-cooled Homogeneous
Reactor—LAPRE II, USAEC Report LA-1942, Los Alamos Scientific Labora-
tory, Apr. 13, 1955. R. P. Hammonp, Los Alamos Homogeneous Reactor Pro-
gram, in HRP Civilian Power Reactor Conference Held at Oak Ridge, March
21-22, 1956, USAEC Report TID-7524, Los Alamos Scientific Laboratory,
March 1957. (pp. 168-176)

16. Figure 3-12 was supplied by R. B. Brigas, Oak Ridge National Labora-
tory. Unpublished.

17. L. D. P. King, Los Alamos Homogencous Reactor Program, in HREP
Civilian Power Reactor Conference Held at Oak Ridge, March 21-22, 19566, USAEC
Report TID-7524, Los Alamos Scientific Laboratory, March 1951. (pp. 177-209)

18. Frank J. LopresT et al., J. Am. Chem. Soc. 77, 4705 (1955).

19. F. J. Loprest et al., Homogeneous Reaclor Project Quarterly Progress
Report for the Period Ending July 31, 1955, USAEC Report ORNL-1943, Oak
Ridge National Laboratory, Aug. 9, 1955 (p. 227). W. L. MaRrsHALL et al,
Homogeneous Reactor Project Quarterly Progress Report for the Period Ending
Jan. 31, 1956, USAEC Report ORNL-2057(Del.), Oak Ridge National Labora-
tory, Apr.17,1956 (p. 131). C. A. BLakE et al., J. Am. Chem. Soc. 78, 5978 (1956).

20. W. C. WacceNER, Oak Ridge National Laboratory, 1953, personal com-
munication.

21. M. H. Lierzge and W. L. MaARrsHALL, Present Status of the Investigation of
Aqueous Solutions Suitable for Use in a Thorium Breeder Blanket, USAEC
Report ORNL-1711, Oak Ridge National Laboratory, June 1954.

22. W. L. MagrsuALL et al., J. Am. Chem. Soc. 73, 4991 (1951). Jou~ R.
FErRRARO et al., J. Am. Chem. Soc. 76, 909 (1954).

23. P. G. Jones and R. G. Sowpen, Thorium Nitrate Solution as a Breeder
Blanket in the H.A.R., Report AERE-C/M-298, Part I. Thermal Stability,
Gt. Brit. Atomic Energy Research Establishment, 1956.

24. W. L. Magrsuart and C. H. Secoy, Preliminary Exploration of the
Th(NO3)4-HNO3-Hz0 System at Elevated Temperature, in Homogeneous Re-
actor Project Quarterly Progress Report for the Period Ending Oct. 31, 1954,
USAEC Report ORNL-1658, Oak Ridge National Laboratory, Feb. 25, 1954.
(pp. 93-96)

25. R. E. Leuze, Chemistry of Plutonium in Uranyl Sulfate Solutions, in
HRP Civilian Power Reactor Conference Held at Oak Ridge National Laboralory,
May 1-2, 1957, USAEC Report TID-7540, Oak Ridge National Laboratory,
July 1957. (pp. 221-231)

26. D. E. GranvitLe and D. W. GranT, Gt. Brit. Atomic Energy Research
Establishment, 1956. Unpublished.

27. Mound Laboratory, Interim Monthly Reports, 1957-1958.

28. L. V. JonEs et al., Isolation of Protactinium-231, in Homogeneous Reactor
Project Quarterly Progress Report for the Period Ending July 21, 1956, USAEC
Report ORNL-2148(Del.), Oak Ridge National Laboratory, Oct. 3, 1956.
(pp. 144-145)

29, A. T. Gresky, Separation of U233 and Thorium from Fission Products
by Solvent Extraction, in Progress in Nuclear Energy—=Series 3; Process Chem-
124 PROPERTIES OF AQUEOUS FUEL SOLUTIONS [cHAP. 3

istry, by F. R. Bruck et al., New York: MeGraw-Hill Book Co., Ine., 1956:
Thorex Process Summary, February 19566, USAEC Report CF-56-2-157, Oak
Ridge National Laboratory, Apr. 25, 1956. A. T. Gresky et al., Laboratory
Development of the Thorex Process: Progress Report, Oct. I, 1952 to Jan. 31, 1958,
USAEC Report ORNL-1518, Oak Ridge National Laboratory, Aug. 4, 1953.
E. D. Arvorn et al., Preliminary Cost Estimation: Chemical Processing and Fuel
Costs for a Thermal Breeder Reactor Power Station, USAEC Report ORNL-1761,
Oak Ridge National Laboratory, Feb. 23, 1955. R. E. ELson, The Chemistry of
Protactinium, in The Actinide Elements, ed. by G. T. SeaBorG and J. J. Karz,
National Nuclear Energy Series, Division IV, Volume 14A. New York: McGraw-
Hill Book Co., Inc., 1953. (Chap. 5, p. 103)

30. D. E. Fercuson, Homogeneous Plutonium Producer Chemical Processing,
in HRP Civilian Power Reactor Conference Held at Oak Ridge March 21-22, 1956,
USAEC Report TID-7524, Oak Ridge National Laboratory (pp. 224-232). J. C.
Hinpman et al, Some Recent Developments in the Chemistry of Neptunium,
in Proceedings of the International Conference on the Peaceful Uses of Atomic
FEmnergy, Vol. 7. New York: United Nations, 1956 (P/736, p. 345). K. A. Kravus,
Hydrolytic Behavior of the Heavy Elements, in Proceedings of the International
Conference on the Peaceful Uses of Atomic Energy, Vol. 7, New York: United
Nations, 1956. (P/731, p. 245). W. C. WagceenEr and R. W. STOUGHTON,
Spectrophotometry of Aqueous Solutions, in Chemistry Division Annual Progress
Report for the Period Ending June 20, 1957, USAEC Report ORNL-2386, Oak
Ridge National Laboratory, (pp. 64-71). W. C. WAGGENER, in Chemistry Divi-
sion Seminar—1958, Oak Ridge National Laboratory (to be published).

31. A. O. ALLEN, J. Phys. & Colloid Chem. 52, 479 (1948). C.J. HoCHANADEL,
Radiatron Stability of Aqueous Fuel Systems, USAEC Report CF-56-11-54, Oak
Ridge National Laboratory, November 1956.

32. F. 5. DaNron and H. C. SBurrow, Trans. Faraday Soc. 49, 1011 (1953).

33. T. J. Sworskl, J. Am. Chem. Soc. 76, 4687 (1954).

34. E. J. Hart, Radiation Research 2, 33 (1955).

35. Selected values from several literature references.

36. A. O. ALLeN et al,, J. Phys. Chem. 56, 575 (1952). C. J. HOCHANADEL,
J. Phys. Chem. 56, 587 (1952).

37. C. J. Hocuanapew, Radiation Induced Reactions in Water, in Proceedings
of the International Conference on the Peaceful Uses of Atomic Energy, Vol. 7.
New York: United Nations, 1956. (P/739, p. 521)

38. J. W. BoyLE et al.,, Nuclear Engineering and Science Congress, Held in
New York in 1955, American Institute of Chemical Engineers (Preprint 222).
H. F. McDurrFie, Aqueous Fuel Solutions, Chap. 4.3, in Reactor Handbook,
Vol. 2, Engineering, USAEC Report AECD-3646, Oak Ridge National Labora-
tory, 1955. (p. 571)

39. H. F. McDurriE et al.,, The Radiation Chemistry of Homogeneous Reactor
Systems: I11—Homogeneous Catalysis of the Hydrogen-Oxygen Reaction, USAEC
Report CF-54-1-122, Oak Ridge National Laboratory, 1954.

40. For solubilities of Hz and O2 in water and in V02804 solutions at high
temperatures see: H. A. H. Pray ct al., Ind. Eng. Chem. 14, 1146 (1952); E. F.
STeEPHAN et al., The Solubilities of Gases in Water and in Aqueous Uranyl Salt
REFERENCES 125

Solutions at Elevated Temperatures and Pressures, USAEC Report BMI-1067
Battelle Memorial Institute 1956; see also See. 4, this chapter.

41, 8. Visner and P. N. HausenreicH, HRE Experiments on Internal
Recombination of Gas with a Homogeneous Catalyst, USAEC Report CF-55-1-166,
Oak Ridge National Laboratory, Jan. 21, 1955.

42, R. E. Avex and M. C. LAWRENCE, Calculation of Effects of Copper Catalyst
in the HRT, USAEC Report CF-56-4-4, Oak Ridge National Laboratory, Apr.
10, 1956.

43. M, D. SiLVvERMAN et al., Ind. Eng. Chem. 48, 1238 (1956).

44, J. W. BoviLe and H. A. MaarmaN, paper presented at the 2nd Annual
Meeting of the American Nuclear Society, Chicago, 1956.

45. M. M. Harring, Mound Laboratory, Uranium Salis Research Progress
Reports, USAEC Report MLM-628, 1951. E. Orsan, Mound Laboratory, Ura-
nium Salts Rescarch Progress Reports, USAEC Reports MLM-663, 1952,
MLM-680, 1952; MLM-710, 1952; MLM-755, 1952; MLM-790, 1952; MLM-825,
1953; MLM-863, 1953; MLM-928, 1953.

46. E. Orsan, Mound Laboratory, 1952. Unpublished.

47. J. 8. JeGarr ct al., The Densities of Uranyl Sulfate Solutions Between 20°
and 90°C, USAEC Report MLM-728, Mound Laboratory, Oct. 10, 1952.

48. E. OrBan, The pH Measurement of Uranyl Sulfate Solutions from 25°
to 60°C, USAEC Report MLM-729, Mound Laboratory, Aug. 1, 1952.

49. E. Orsan, The pHl of Uranyl Sulfate-Urantum Trioxide Solutions, USAEC
Report AECD-3580, Mound Laboratory, Aug. 14, 1952.

50. J. R. Hriks and J. S. JEgart, The Viscosity of Uranyl Sulfate Solutions
20° to 90°C, USAEC Report MLM-788(Rev.), Mound Laboratory, Feb. 22, 1954.

51. J. R. HEixs et al., Apparatus for Determining the Physical Properties of
Solutions at Elevated Temperatures and Pressures, USAEC Report MLM-799,
Mound Laboratory, Jan. 14, 1953.

52. A. J. Rocers ct al.,, An Insirument for the Measurement of the Time of
Fall of @ Plummet tn a Pressure Vessel, USALC Report MLM-805, Mound
Laboratory, Feb. 27, 1952.

53. M. K. Barnert and J. Jeeart, The Surface Tension of Aqueous Uranyl
Sulfate Solutions Between 20° and 75°C, USAEC Report AECD-3581, Mound
Laboratory, Feb. 15, 1953.

54. J. R. Heiks et al., The Physical Properties of Heavy Waler From Room
Temperature to 250°C, USAEC Report MLM-934, Mound Laboratory, Jan. 12,
1954.

55. M. K. BarNETT et al.,, The Density, Viscosity, and Surface Tenston of
Light and Heavy Water Solutions of Uranyl Sulfate at Temperatures to 250°C,
USAEC Report MLM-1021, Mound Laboratory, Dec. 6, 1954.

56. E. OrBaN et al., Physical Properties of Aqueous Uranyl Sulfate Solutions
from 20° to 90°, J. Phys. Chem. 60, 413 (1956).

57. R. Van WiINKLE, Some Physical Properties of UQaS04-D0 Solutions,
USAEC Report CIF-52-1-124, Oak Ridge National Laboratory, Jan. 18, 1952.

58. M. Toszias, Certain Physical Properties of Aqueous Homogeneous Reactor
Malerials, USAEC Report CF-56-11-135, Oak Ridge National Laboratory, Nov.
23, 1956.
126 PROPERTIES OF AQUEOUS FUEL SOLUTIONS [cHAP. 3

59. J. A. LANE et al., Properties of Aqueous Solution Systems, Chap. 4.3 in
The Reactor Handbook, Vol 2, Engineering, USAEC Report AECD-3646, Oak
Ridge National Laboratory, 1955.

60. P. A. Lorres, Physical and Thermodynamic Properties of Light and
Heavy Water, Chap. 1.3 in The Reactor Handbook, Vol. 2, Engineering, USAEC
Report AECD-3646, Argonne National Laboratory, 1955. (pp. 21-42)

61. N. E. Dorsey, Properties of Ordinary Water-Substance, New York:
Reinhold Publishing Corp., 1940.

62. J. H. Keenan and F. G. Kryrs, Thermodynamic Properties of Steam,
New York: John Wiley & Sons, Inc., 1936.

63. G. M. HegEerT et al., The Densities of Heavy-Water Liquid and Saturated
Vapor at Elevated Temperatures, J. Phys. Chem. 62, 431 (1958).

64. W. L. MagrsnaLL, Density—Weight Percent—Molarity Conversion Equa-
tions for Uranyl Sulfate—Water Solutions at 25.0°C and Between 100-300°C,
USAEC Report CF-52-1-93, Oak Ridge National Laboratory, Jan. 15, 1952.

65. R. C. Harpy and R. L. CorriNgTON, J. Research Nat. Bur. Standards 42,
573 (1949).

66. H. O. Day and C. H SEkcoy, Oak Ridge National Laboratory. Unpub-
lished data.

67. W. L. MarsuavL, Oak Ridge National Laboratory, March 1958, personal
communication,

68. O. V. IN6rUBER, The Direct Measurement of pH at Elevated Tempera-~
ture and Pressure during Sulphite Pulping, Pulp Paper Can. 55 (10), 124-131
(1954).

69. M. H. Lierzke and J. R. TArraNT, Preliminary Report on the Model 904
High-temperature pH Meter, USAEC Report CF-57-11-87, Oak Ridge National
Laboratory.

70. M. H. Lierzxke, The Hydrogen Electrode—Silver Chloride Electrode
System at High Temperatures and Pressures, J. Am. Chem. Soc. 77, 1344 (1955).

71. E. F. Steruan et al., The Solubility of Gases in Water and in Agueous
Uranyl Salt Solutions at Elevated Temperatures and Pressures, USAEC Report
BMI-1067, Battelle Memorial Institute, 1956.

72. H. A. Pray et al.,, The Solubility of Hydrogen, Oxygen, Nitrogen, and
Helium in Water at Elevated Temperatures, Ind. Eng. Chem. 44, 1146 (1952).

73. J. A. Luxer and Taomas GNIEWEK, Saturation Composilion of Steam—
Helvum—Water Mixtures PVT Data and Heat Capacity of Superheated Steam—
Helwm Mixtures, USAEC Report AECU-3299, Syracuse University Research
Institute, July 29, 1955.

74. J. A. Luker and TroMas GNIEWEK, Determination of PVT Relationships
and Heat Capacity of Steam—Ozxygen Mixztures, USAEC Report AECU-3300,
Syracuse University Research Institute, Aug. 2, 1955.

75. E. F. STeEPHAN et al., Ignition Reactions in the Hydrogen—Ozygen—Water
System at Elevated Temperatures, USAEC Report BMI-1138, Battelle Memorial
Institute, Oct. 2, 1956.

76. T. W. Levanp, Evaluation of Data from Batlelle Memorial Institute on
Solubility of He and Oy in Solutions of U0pS04 and UOgF s in Water and on
Ezrplosion Limits in Gaseous Miztures of Hg, Op and Hg0 and in Hs, Og, He,
REFERENCES 127

and Hs0, USAEC Report CF-54-8-215, Qak Ridge National Laboratory, Aug.
23, 1954.

77. Irnviy M. MacArEg, Jr., Detonation, Explosion, and Reaction Limits of
Saturated Stoichiometric Hydrogen—Ozygen—Water Mixtures, USAEC Report
AECU-3302, Syracuse University, Research Institute, July 1, 1956. Paur L.
McGinn and James A. LUKER, Detonation Pressures of Stoichiometric Hydrogen—
Ozygen Mixtures Saturated with Waler at High Intiial Temperatures and Pressures,
USAEC Report AECU-3429, Syracuse University, Research Institute, Dec. 3,
1956.