NUCLEAR
CHEMICAL
ENGINEERING
Second Edition
Manson Benedict
Professor Emeritus of
Nuclear
Engineering
Massachusetts Institute of Technology
Thomas
H.
Pigford
Professor
of
Nuclear Engineering
University
of
Gal$omia,
Berkeley
Hans Wolfgang Levi
Hahn-Meitner-Institut
fir
Kernforschung Berlin
and
apL
Professor
of
Nuclear Chemistty
Technische Universitat Berlin
McGraw-Hill
Book

Company
New
York
St.
Louis
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Francisco Auckland
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This
book was set in Press Roman by Hemisphere Publishing Corporation.
The
editor was Diane
D.
Heiberg;
the production supervisor was Rosann
E.
Raspini.
Kingsport Press, Inc.
was
printer and binder.
NUCLEAR CHEMICAL ENGINEERING
Copyright
0
1981, 1957 by McCraw-Hill, Inc.

All
rights reserved.
Printed in the United States of America. No part
of
this publication
may
be
reproduced, stored in a retrieval system,
or
transmitted, in any
form
or
by any means, electronic, mechanical, photocopying, recording,
or
otherwise, without the prior written permission
of
the publisher.
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KPKP
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Library
of
Congress
Cataloging
in Publication Data
Benedict, Manson
Nuclear chemical engineering,
(McGraw-Hill series in nuclear engineering)
Includes bibliographies and index.
1. Nuclear engineering. 2. Nuclear chemistry.
I.
Pigford, Thomas
H.,
joint author.
11.
Levi,
Hans
Wolfgang, joint author.
111.
Title.
TK9350.B4 1981 621.48 80-21538
ISBN
0-07-004531-3
PREFACE
The development of nuclear fssion chain reactors for the conversion of mass to energy and the
transmutation of elements has brought into industrial prominence chemical substances and
chemical engineering processes that a few years ago were
no
more than scientific curiosities.

Uranium, formerly used mainly for coloring
glass
and ceramics, has become one of the world’s
most important sources of energy. Thorium, once used mainly in the Welsbach
gas
mantle,
promises to become a nuclear fuel second in importance only to uranium. Zirconium and its
chemical twin hafnium, formerly always produced together, have been separated and have emerged
as structural materials of unique value in reactors. New chemical engineering processes have been
devised to separate these elements, and even more novel processes have been developed for
producing deuterium,
*’’
U,
and the other separated isotopes that have become the fine chemicals
of the nuclear age. The processing of radioactive materials, formerly limited mainly to a few curies
of radium,
is
now concerned with the
millions
of curies of radioactive isotopes of the many
chemical elements that are present in spent fuel discharged from nuclear reactors.
The preceding introduction to the preface of the first edition of this book can still serve as the
theme of this second edition. Since
1957
nuclear power systems have become important
contributors to the energy supply of most industrialized nations.
This
text describes the materials
of special importance in nuclear reactors and the processes that have been developed to
concentrate, purify, separate, and store safely these materials. Because of the growth in nuclear

technology since the first edition appeared and the great amount of published new information,
this second edition is an entirely new book,.following the
first
edition only in its general outline.
Chapter
1
lists the special materials of importance in nuclear technology and outlines the
relationship between nuclear reactors and the chemical production plants associated with them.
Chapter
2
summarizes the aspects of nuclear physics and radioactivity that are pertinent to many
of the processes to be described in later chapters. Chapter
3
describes the changes
in
composition
and reactivity that occur during irradiation of fuel in a nuclear reactor and shows how these
changes determine the material and processing requirements of the reactor’s fuel cycle. Chapter
4
describes the principles of solvent extraction, the chemical engineering unit operation used most
extensively for purifying uranium, thorium, and zirconium and reprocessing irradiated fuel
discharged from reactors.
Chapters
5,
6,
and
7
take up uranium, thorium, and
zirconium
in that order. Each chapter

discusses
the physical and chemical properties
of
the element and
its
compounds, its natural
occurrence, and the processes
used
to extract the element from its ores, purify it, and convert it
to
the forms most useful
in
nuclear technology.
X
iii
xiv
PREFACE
The next four chapters take up processing of the
highly
radioactive materials produced in
reactors. Chapter
8
describes the isotopic composition and radioactive constituents of spent fuel
discharged from representative types of reactors and deals briefly with other radioisotopes
resulting from reactor operation. Chapter
9
describes the physical and chemical properties of the
synthetic actinide elements produced in reactors: protactinium, neptunium, plutonium,
americium, and curium, and their compounds. Chapter
10

describes the radiochemical processes
that have been developed for reprocessing irradiated fuel to recover uranium, plutonium, and other
valuable actinides from
it.
Chapter
11
describes conversion of radioactive wastes from reactor
operation and fuel reprocessing into stable forms suitable for safe, long-term storage, and systems
to be used
for
such storage.
The
last
three chapters deal with separation of stable isotopes. Chapter
12
lists the isotopes of
principal importance in nuclear technology, discusses their natural occurrence, and develops the
chemical engineering principles generally applicable to isotope separation processes. Chapter
13
describes processes useful for separating deuterium and isotopes of other light elements, specifically
distillation, electrolysis, and chemical exchange. Chapter
14
describes processes used for separating
uranium isotopes, specifically
gaseous
diffusion, the
gas
centrifuge, aerodynamic processes,
mass
and thermal diffusion, and laser-based processes.

Four appendixes list fundamental physical constants, conversion tables, nuclide properties,
and radioactivity concentration limits for nuclear plant effluents.
As may be seen from this synopsis,
this
text combines an account of scientific and engineering
principles
with
a description of materials and processes of importance in nuclear chemical
technology. It aims thus to serve both as
a
text for classroom instruction and as
a
source of
information on chemical engineering practice in nuclear industry.
Problems at the end of each chapter may prove useful when the text is used for instruction.
References are provided for readers who wish more details about the topics treated in each
chapter. Extensive use has been made of information from the
Roceedings
of the four
International Conferences
on
the Peaceful Uses of Atomic Energy in Geneva, Switzerland,
sponsored by the United Nations, which are listed as
PIG,
followed by the number of the
conference, in the references at the ends of chapters.
This book was written in a transition period when U.S. engineering and business practice was
changing from English to
SI
units. When the references cited used Enash units, these have been

retained in the text in most cases. Equivalent
SI
values are also provided in many passages, or
conversion factors are given in footnotes.
In
addition, conversion tables are provided in App.
B.
The multiplicity of units is regrettable, but it is unavoidable until the world’s technical literature
has changed over completely to the
SI
system.
In
preparing
this
text the authors have been blessed with assistance from
so
many sources that
not all
can
be mentioned here. We are grateful to our respective institutions, Massachusetts
Institute of Technology, University of California (Berkeley), and Hahn-Meitner-Institut (Berlin),
for the freedom and opportunity to write
this
book. For help with calculations, illustrations, and
typing,
thanks
are due Marjorie Benedict, Ellen
Mandigo,
Mary
BOSCO,

Sue
Thur,
and many others.
Editorial assistance from Judith
B.
Gandy and Lynne Lackenbach is acknowledged with gratitude.
To the many generations of students who used the notes
on
which this book is based and helped
to correct
its
mistakes we are greatly indebted. Among our more recent students we
wish
to
thank
Men Croff, Charles Forsberg,
Saeed
Tajik,
and
Cheh-Suei Yang.
Among
our
American professional colleagues we are greatly indebted to Don Ferguson and
his
associates at Oak Ridge National Laboratory;
Paul
McMurray and others of Exxon Nuclear
Company; James Buckham and Wesley Murbach of Allied General Nuclear Services; James
Duckworth of Nuclear Fuel Services, Inc.; Joseph Megy of Teledyne
Wah

Chang Albany Company;
Paul Vanstrum and Edward Von Halle of Union Carbide Corporation; Lombard
Squires,
John
PREFACE
xv
Proctor, and their associates of
E.
I.
duPont de Nemours and Company; Marvin Miller of
MIT;
and
Donald Olander
of
the University of California (Berkeley). In Germany, we
wish
to
thank Hubert
Eschrich of Eurochemic, Richard Kroebel of Kernforschungszentrum Karlsruhe, Erich Merz of
Kernforschungsanlage Jiilich, Walther Schuller of
Wiederaufarbeitungsanlage
Karlsruhe,
and
Eckhart Ewest of Deutsche Gesellschaft fur Wiederaufarbeitung von Kernbrennstoff.
Assistance provided to one of the authors
(MB)
by a fellowship from the Guggenheim
Foundation
is
acknowledged with gratitude.

Despite the valued assistance the authors have had in preparing this text, it doubtless
still
contains many errors and omissions. We shall be grateful to our readers for calling these to
our
attention.
Manson
Benedict
Thomas
H.
pisford
Hans Wolfgang
Levi
CONTENTS
Preface
Chapter
1
Chemical Engineering Aspects of Nuclear
Power
Introduction
Nuclear Fission
Nuclear Fuels
Nuclear Reactor Types
Fuel Processing
Flow
Sheets
Fuel-Cycle Operations
Fuel Reprocessing
lsotope Separation
Nuclear Fusion
References

Problems
Chapter
2
Nuclear Reactions
1 Nuclides
2
Radioactivity
3
Decaychains
4
Neutron Reactions
5
The Fission Process
6
7
Growth and Decay of Nuclides with Simultaneous Radioactive
Decay, Neutron Absorption, and Continuous Processing
Derivation
of
the Bateman Equation (2.17) by
Laplace Transforms
Nomenclature
References
Problems
Chapter
3
Fuel Cycles for Nuclear Reactors
1
Nuclear Fuels
2 Effects of Irradiation

on
Nuclear Fuels
3
Fuel
and
Poison
Management
Xiii
1
1
2
5
7
10
15
20
22
23
24
25
26
26
27
35
42
53
63
76
78
80

81
84
84
87
90
.iii
CONTENTS
4
5
6
7
Chapter
4
1
2
3
4
5
6
7
Chapter
5
1
2
3
4
5
6
7
8

9
10
Chapter
6
1
2
3
4
5
6
7
8
9
10
Chapter
7
1
2
Fuel Management
in
a Large Pressurized-Water
Reactor
Fuel-Cycle Costs
Hand
Calculation of Fuel-Cycle Performance
Fuel-Cycle hiaterial
Flow
Sheets
Nomenclature
References

Problems
Solvent Extraction
of
Metals
Applications
Extractable Metal-Organic Complexes
Solvent Extraction Principles
Distribution Coefficients
Solvent Requirements
Theory of Countercurrent Equilibrium Extraction
Solvent Extraction Equipment
Nomenclature
References
Problems
Uranium
Uranium Isotopes
Uranium Radioactive Decay Series
Metallic Uranium
Uranium Compounds
Uranium Solution Chemistry
Sources of Uranium
Uranium Resource Estimates
Concentration of Uranium
Uranium Refining
Production
of
Uranium Metal
References
Problems
Thorium

Uses of Thorium
Thorium isotopes
Thorium Radioactivity
Metallic Thorium
Thorium Compounds
Thorium Solution Chemistry
Thorium Resources
Concentration and Extraction of Thorium
Purification of Thorium
Conversion of Thorium Nitrate to Oxide, Fluoride,
Chloride,
or
Metal
References
Problems
Zirconium and Hafnium
Uses of Zirconium and Hafnium
Natural Occurrence
105
113
126
144
151
153
154
157
157
157
160
165

172
173
198
21 1
212
214
216
21 6
217
222
223
229
232
234
236
266
274
280
28 1
283
283
283
285
287
289
293
294
298
30 7
309

315
317
318
318
319
CONTENTS
ix
Chapter
8
1
2
3
4
5
Chapter
9
1
2
3
4
5
6
Chapter
10
1
2
3
4
5
6

7
8
Chapter
11
1
2
3
4
5
Production and Rice
Zirconium and Hafnium Metal and
Alloys
Zirconium and Hafnium Compounds
Extraction of Zirconium
and
Hafnium from Zircon
Separation of Zirconium and Hafnium
Production of Metallic Zirconium and Hafnium
Alternatives for Producing Hafnium-Free Zirconium
from Zircon
References
Problems
Properties
of
Irradiated Fuel
and Other Reactor Materials
Fission-Product Radioactivity
Radioactivity of the Actinides
Effect of Fuel-Cycle Alternatives
on

Properties
of Irradiated Fuel
Radioactivity from Neutron Activation
Neutron Activity in Recycled Fuel
Nomenclature
References
Problems
Plutonium and Other Actinide Elements
General Chemical Properties of the Actinides
Properties of Protactinium
Properties
of
Neptunium
Properties of Plutonium
Properties of Americium
Properties
of
Curium
References
Problems
Fuel Reprocessing
Objectives
of
Reprocessing
Composition of Irradiated Fuel
History of Reprocessing
The Purex Process
Reprocessing Thorium-Based Fuels
Reprocessing LMFBR Fuels
Neptunium Recovery in Reprocessing

Prevention
of
Criticality
in
Reprocessing Plants
References
Problems
Radioactive Waste Management
Introduction
High-Level Waste
Non-High-Level Waste
Special Radioactive Waste
Disposal
of Radioactive Waste
319
320
323
330
333
342
348
348
350
352
352
364
381
39
1
401

404
405
406
401
407
420
424
426
449
45
1
454
45 6
457
45 7
45
7
45
8
466
514
527
537
547
556
563
565
565
567
604

609
61
3
x
CONTENTS
6
Chapter
12
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
Chapter
13
1
2
3
4
5

6
7
8
9
10
11
12
13
14
15
16
Assessment of Long-Term Safety
References
Problems
Stable Isotopes: Uses, Separation Methods,
and Separation Principles
Uses
of Stable Isotopes
Isotope Separation Methods
Terminology
Stage Properties
Types of Cascade
The Simple Cascade
The Recycle Cascade
The Ideal Cascade
Close-Separation Cascade
Separative Capacity, Separative Work, and Separation
Potential
Differential Equation for Separation Potential
Equilibrium Time for Isotope Separation Plants

Squared-off Cascade
Generalized Ideal Cascade
Three-Component Isotope Separation
Nomenclature
References
Problems
Separation
of
Isotopes
of
Hydrogen
and Other Light Elements
Sources of Deuterium
Deuterium Production Processes and Plants
Separation Factors
in
Distillation
Distillation of Hydrogen
Distillation of Water
Electrolysis
Electrolysis and Steam-Hydrogen Exchange
Separation Factors
in
Deuterium Exchange Processes
Number of Theoretical Stages
in
Exchange Columns
Monothermal Exchange Processes
Dual-Temperature Water-Hydrogen Sulfide Exchange
Process

Dual-Temperature Ammonia-Hydrogen Exchange Process
Methylamine-Hydrogen Exchange Processes
Dual-Temperature Water-Hydrogen Exchange Processes
Exchange Processes for Separation of Lithium Isotopes
Exchange Processes for Other Elements
Nomenclature
References
Problems
618
624
626
627
627
629
644
647
65 1
65 3
654
65 8
665
667
674
677
684
685
693
70
1
703

70
5
708
708
710
71 2
717
722
738
749
756
760
7 62
767
792
797
799
800
80 1
804
806
808
CONTENTS
xi
Chapter
14
Uranium
Isotope
Separation
Introduction

Isotopic Content of Uranium
Uranium Enrichment Projects
Gaseous
Diffusion
The
Gas
Centrifuge
Aerodynamic Processes
Mass
Diffusion
Thermal Diffusion
Laser Isotope Separation
Nomenclature
References
Problems
Appendixes
A
Fundamental Physical Constants
B
Conversion Factors
C
Properties of the Nuclides
D
Radioactivity Concentration Limits for Selected
Radionuclides
812
812
813
815
818

847
876
895
906
914
922
925
929
933
933
935
937
979
Index
983
This
text
is
dedicated
to
the authors' wives,
Marjorie Allen Benedict, Catherine Cathey Pigford,
and Ruth Levi,
whose assistance, encouragement,
and
patience
made
this
book possible,
CHAPTER

ONE
CHEMICAL ENGINEERING ASPECTS
OF
NUCLEAR POWER
1
INTRODUCTION
The production of power from controlled nuclear fission of heavy elements is the most
important technical application of nuclear reactions at the present time. This is
so
because the
world’s reserves of energy
in
the nuclear fuels uranium and thorium greatly exceed the energy
reserves in all the coal, oil, and gas in the world
[HI],
because the energy of nuclear fuels is in
a form far more intense and concentrated than in conventional fuels, and because in many parts
of the world power
can
be produced as economically from nuclear fission as from the
combustion of conventional fuels.
The establishment of a nuclear power industry based on fission reactors involves the
production
of
a number of materials that have only recently acquired commercial importance,
notably uranium, thorium, zirconium, and heavy water, and on the operation of a number of
novel chemical engineering processes, including isotope separation, separation of metals by
solvent extraction, and the separation and purification of intensely radioactive materials on a
large scale.
This

text is concerned primarily with methods for producing the special materials
used
in
nuclear fission reactors and with processes for separating isotopes and reclaiming
radioactive fuel discharged from nuclear reactors.
This
chapter gives a brief account of the nuclear fusion reaction and the most important
fdle fuels. It continues with a short description of a typical nuclear power plant and outlines
the characteristics of the principal reactor types proposed for nuclear power generation. It
sketches the principal fuel cycles for nuclear power plants and points out the chemical
engineering processes needed to make these fuel cycles feasible and economical. The chapter
concludes
with
an
outline of another process that may some day become of practical
importance for the production of power: the controlled fusion of light elements. The fusion
process makes
use
of rare isotopes of hydrogen and lithium, which may be produced by isotope
separation methods analogous to
those
used
for
materials for fission reactors.
As
isotope
separation processes are of such importance
in
nuclear chemical engineering, they are discussed
briefly in this chapter and in some detail

in
the last three chapters of
this
book.
1
2
NUCLEAR CHEMICAL ENGINEERING
Neutron
-1)
Uranium235
nudeus
Flpun
1.1
Fission
of
235
U
nucleus by neutron.
2
NUCLEAR
FISSION
The nuclear fwon process utilized
in
today's power-producing reactors is initiated by
interaction between a neutron and a fissile nucleus, such
as
The nucleus then divides into
two fragments, with release of an enormous amount
of
energy and with production

of
several
new neutrons. Under proper conditions, these product neutrons can react with additional
='U
atoms and thus give
rise
to a neutron chain reaction, which continues as long as sufficient ='U
remains to react. Fission of a single nucleus of
='U
is represented pictorially in Fig.
1.1,
and a
fission chain reaction is
shown
in
Fig. 1.2. To keep the rate of the chain reaction constant,
neutrons are allowed
to
leak from a nuclear reactor or are absorbed in boron,
or
other
nonfissionable materials placed in the reactor.
A
steady chain reaction is depicted
in
Fig.
1.3.
The fission of llsU
can
take place in a number of ways, one of which is shown in Fig.

1.4.
The nucleus of '"U, which contains
92
protons and
143
neutrons, divides into
two
fragments,
plus some extra neutrons, in such a manner that the total number of protons and neutrons
in
the product nuclei equals the total number in the reactant neutron and
llsU
nucleus. In the
example of
this
figure, the fission fragments are lUBa, containing
56
protons and
88
neutrons;
%I,
containing
36
protons and
53
neutrons; and three extra neutrons. The fission fragments
are unstable and subsequently undergo radioactive decay.
In
the radioactive decay some of the
neutrons of the nucleus are converted into protons, which remain in the nucleus, and into

electrons, which fly out as beta radiation.
In
this example, four neutrons
in
'"Ba are
successively converted into protons, resulting
in
lUNd
as
end product, and three neutrons in
"KI
are
converted into protons, resulting
in
*'Y
as
end product.
The numbers assigned to each reactant or.end product represent its mass
in
atomc mass
units
(amu).
This
unit is defined as the ratio of the mass of a neutral atom to one-twelfth the
mass
of
an
atom of
"C.
In

the present instance the
mass
of the
products
is less
than
that
of
the
reactants*:
tIn
this text each nuclide, such as uranium-235,
is
referred to by its chemical symbol,
in
this case
='u.
*The mass of the electrons
is
not included
in
this calculation because the electrons emitted
from
the
nucleus
in
radioactive
decay
ultimately return
as

orbital electrons
mounding
the
nucleus
of
a neutral atom.
CHEMICAL ENGINEERING ASPECTS
OF
NUCLEAR POWER
3
Reactants Products Difference
WU
235.043915 '"Nd 143.910039
3 neutrons 3.025995
Neutron 1.008665
*Y
88.90587 1
Total 236.052580 235.841 905 0.210675
A fraction 0.210675/235.043915
=
0.0008963 of the mass of the
='U
atom disappears in
this
fission reaction.
This
reduction
in
mass is a measure of the amount of energy released in
this fssion reaction. The Einstein equation (1.1) expressing the equivalence of energy

and
mass,
AE=czLLn
(1.1)
predicts that when
Am
kilograms of mass disappears,
AE
joules of energy appears in its place.
In this relation,
c
is
the velocity of light, 2.997925
X
10'
m/st The energy released
in
this
fission reaction thus is
(0.0008963) (2.997925
X
=
8.06
X
1013
J/kg
235U
(1.2)
or
3.46

X
10''
Btu/lb.
Energy changes associated with
a
single nuclear event are commonly expressed in terms of
millions of electron volts (MeV), defined as the amount of energy acquired by
an
electronic
charge (1.602
X
lo-''
C)
when accelerated through
a
potential difference of
1,OOO,ooO
V. One
MeV therefore equals 1.602
X
The energy released when one atom of
23'U
undergoes fission in the above reaction is
X
IO6
=
1.602
X
J.
=

196 MeV/atom (1.3)
(8.06
X
IOI3
J/kg)(235.04 g/g-atom)
(1.602
X
IO-l3
J/MeV)(6.023
X
atorns/g-atom)(lOOO g/kg)
TFundamental physical constants are listed in App.
A.
A
table
of
mass and energy
equivalents
is
given in App.
B.
235
U
fission
Figure
1.2
Fission
chain reaction.
4
NUCLEAR CHEMICAL

ENGINEERING
Figure
1.3
Steady fission chain reaction.
Atoms of
235U
may undergo fission
in
a variety of ways, of which the reaction shown
in
Fig.
1.4
is
only one. The average yield of particles and energy from fission of
235U
in
all
possible ways is shown in Fig.
1.5.
In the primary fission reaction
shown
at the top of
this
figure,
'"U
splits into two parts, the radioactive fission products, while at the same time giving
off several fast neutrons
(2.418
on the average) and gamma radiation. One
of

these neutrons is
used to maintain the fission reaction. The remaining neutrons may either be used to bring
about other desired nuclear reactions
or
be lost either through leakage from the reactor or
through capture by elements present in the reactor to produce unwanted
or
waste products.
Following the primary fission reaction, the radioactive fission products undergo radioactive
disintegration, yielding beta particles and delayed gamma rays and ending up as stable fission
products. Since the radioactive fission products have half-lives ranging from fractions of a
Neutron Uranium-235 Barium-144 Krypton-89
1.008665
amu
235.04396
amu
r
Proton
e
+
1
4
Radioactive
Neodymium-144
Qdecay
@
Yttrium89
143.910039
amu
50

0
88905871
am
+
+
40 3
@
Electrons
Figure
1.4
Example
of
fission
of
CHEMICAL ENGINEERING ASPECTS
OF
NUCLEAR POWER
5
7
MeV
1.
Prompt
gama
roys
&&)
I/
Q
+
a
Neutron

23g
%‘
fission
nude Radioactive
fragments
167
MeV
Captured
in
shield
and reactor
Producing-3
to
12
MeV
Q
Used to continue
chain reaction
+
2.418 neutrons
5
MeV
Initial
fission
reaction
I,
7
electrons
8
MeV

Ra+$!;:Aive
Stable fission products
fragments
.c
Delayed
gamma
rays
6
MeV
Later
radioactive
disintegrations
Figure
1.5
Average yields in fission of
”’
U.
second to millions of years, the emission of beta particles and delayed gamma rays takes place
over
a long period of time after a reactor has been shut down, but at a diminishing rate.
The
total energy released in fission is the sum of the energies associated with the different
particles
shown
in this figure, 196 to 205 MeV.
As
up to 5 MeV
of
gamma energy escapes from
a typical power reactor and is not utilized,

a
nominal figure for the energy released in
fission
is
200 MeV. This corresponds to around 35.2 billion Btu of energy per pound or 0.95 MWd ofenergy
per gram of
235U
undergoing fission. In addition, some
=’U
is consumed without undergoing
fission by reacting with neutrons to form
%U.
When this reaction is taken into account, the
energy released is around 29 billion Btu per pound, or
0.78
MWd per
gram
of
U
consumed.
This
is about
2
million times the energy released in the combustion of an equivalent mass of coal.
3
NUCLEAR
FIJELS
In
addition to
=’U,

two
other isotopes can be used as fuel in nuclear fission reactors. These are
plutonium-239,
239F’u,
produced by absorption of neutrons in
238U;
and
u3U,
produced by
absorption
of
neutrons in natural thorium. The reactions by which these isotopes are made are
as
follows:
usU
+
In
-+
u9U
-+
2J9Np
+
e-
+e-
Neutron
J
Beta particles
232n
+
1n

-+
233n
-+
233pa
+
e-
J-
W3U
+e-
Properties of these three fissile fuel nuclides are listed in Table
l.
I.
The number
of
neutrons produced per neutron absorbed by fissile material is less than the
number of neutrons produced per fission because some
of
the neutrons absorbed produce the
higher isotopes
=U,
140Pu,
or
=U
rather
than
causing fission.
6
NUCLEAR
CHEMICAL
ENGINEERING

Tsbk
1.1
Nudearfudn
Isotope
Obtained from
Absorption
of
0.7%
of
neutrons by
natural
uranium Th
Neutrons produced per
Fission
2.418 2.811 2.492
Thermalt neutron absorbed
1.96 1.86 2.2
Absorption
cross
section, b:
Thermalt neutrons
555
1618 41
0
Fast neutrons
1.5
2 2
+In a typical reactor for power production.
The fact that the number of neutrons produced per neutron absorbed exceeds
1.0

for each
fuel indicates that each will support a nuclear chain reaction. Neutrons
in
excess
of
the one
needed to sustain the nuclear chain reaction may be used to produce new and valuable isotopes,
for example, to produce
'"Pu
from
u8U
or from thorium by the reactions cited earlier.
When the number of neutrons produced per neutron absorbed in fissile material is greater
than
2.0,
it is theoretically possible to generate fissile material at a faster rate than it
is
consumed. One neutron is used to maintain the chain reaction, and the second neutron is used
to produce a new atom of fissile material to replace the atom that is consumed by the first
neutron.
This
process
is
known as breeding. The reactions taking place in breeding
u9Pu
from
=U
are
shown
in

Fig.
1.6.
=U
is the only material consumed over all;
u9Pu
is produced
from
lSeU
and then consumed in fission.
Fission
of
'=PU
-,
One neutron continues
chain reaction
Neutron
+
+
Second neutron
is
captured
by238U
to
produce
23%
Later,
2%
form
'39eU
@

radioactively
0
I
decays
to
Atom of
239pU
to
replace
atom consumed
in
fission
Figure
1.6
Breeding
of
?jpPu.
CHEMICAL ENGINEERING ASPECT'S
OF
NUCLEAR POWER
7
In
thermal reactors fueled with plutonium, the number of neutrons produced
per
neutron
absorbed is less
than
2.0
and breeding is impossible. For
luU,

on the other hand,
this
number
is substantially greater than
2.0,
and breeding is practicable in a thermal reactor.
In
fast
reactors, the number of neutrons produced per neutron absorbed is close to the total number
of neutrons produced per fission,
so
that breeding is possible with both and plutonium.
Breeding as here defined is not possible with
%U,
because there
is
no
naturally occurring
isotope from which
='U
can be produced.
A fast reactor is one in which the average speed of neutrons is near that which they have
at the moment of fission, around
15
million m/s. At these
high
speeds the probability of a
neutron's being absorbed by a fissionable atom is low, and the neutron-absorption cross section,
which is a measure of this probability, is small.
A thermal reactor is one in which the neutrons have been slowed down until they are in

thermal equilibrium with reactor materials; in a typical power reactor, thermal neutrons have
speeds around
3000
m/s. At these lower speeds, the neutron-absorption cross sections are much
larger than for fast neutrons.
The critical mass of fissile material required to maintain the fission process is roughly
inversely proportional to the neutron-absorption cross section. Thus the critical mass
is
lowest
for plutonium in thermal reactors, larger for the uranium isotopes in thermal reactors, and
much greater in fast reactors.
For
this reason, as well as others, thermal reactors are the
preferred type except when breeding with plutonium is an objective; then a fast reactor must
be used.
4
NUCLEAR REACTOR TYPES
In addition to classifying nuclear reactors as thermal
or
fast, they may be characterized by their
purpose, by the type of moderator used to slow down neutrons, by the type of coolant, or by
the type of fuel. The principal purposes for which reactors may be used are for research,
testing, production of materials such
as
radioisotopes
or
plutonium, or power generation.
This
text is concerned mainly with power reactors.
The most

effective substances for slowing down neutrons are those elements of low
molecular weight that have low probability of capturing neutrons, namely, hydrogen, deuterium
(the hydrogen isotope
of
atomic mass
2,
chemical symbol
D),
beryllium,
or
carbon. Examples
of moderators containing these elements are light water
(H,O),
heavy water
(D20),
beryllium
oxide, and graphite.
In
many types of thermal power reactors, moderator, fuel, and coolant are kept separate in
the reactor. Figure
1.7
is a schematic diagram of a nuclear power plant utilizing such a reactor.
Table
1.2
lists five examples of reactors with separate moderator, fuel, and coolant and gives
references where more detailed information about these reactors may be obtained. In
this
type
of reactor, fuel and moderator ordinarily remain
in

place
in
the reactor and only coolant flows
through the reactor to remove the heat of fission. Hot coolant flows from the reactor to a
steam generator, where it is cooled by heat exchange with feedwater. The feedwater is
converted to steam, which drives a steam turbine. The steam then is condensed, preheated, and
recirculated as feedwater to the steam generator. Coolant, after being cooled in the steam
generator, is returned to the reactor by the coolant circulator. The steam turbine drives an
electric generator.
When H20 is used as coolant, the same material serves also
as
moderator,
so
,hat the
reactor structure can be simplified. Figure
1.8
is a schematic diagram of a pressurized-water
reactor, in which the coolant and moderator consist of liquid water whose pressure of
150
bar
(2200
Ib/in2) is
so
high that it remains liquid at the highest temperature, around
3W°C
(572'F),
to which it
is
heated
in

the reactor. The main difference in principle from Fig
1.7
is
8
NUCLEAR CHEMICAL ENGINEERING
Steom
Coolant
Genera tor
Steom
Condenser
Condensote
Preheater
Coolont Feed Woter
Circulotor pump
Figure
1.7
Schematic of nuclear power plant with separate fuel, moderator, and coolant.
that there
is
no separation of coolant from moderator in the reactor. The pressurized-water
reactor is one of the two types of power reactor in most common use in the United States.
More information about it
is
given in Chap.
3.
The boiling-water reactor
is
the other type of power reactor in common use in the United
States that uses
HzO

as coolant and moderator. In this type the water in the reactor is at a
lower pressure, around
70
bar
(1000
lb/in2),
so
that it boils and is partially converted to steam
as it flows through the reactor. Coolant leaving the reactor
is
separated into water, which is
recycled, and steam, which is sent directly to the turbine as illustrated in Fig 1.9. Comparison
with
Fig.
1.8
shows that the boiling-water system differs from the pressurized-water system in
having no external steam generator, the reactor itself providing this function.
In a fast-breeder reactor
it
is impractical to use water as coolant because it
is
too effective
a moderator for neutrons. Liquid sodium is the coolant most extensively investigated for fast
Table
1.2
Examples
of
nuclear power reactors
with
separate

fuel,
moderator, and coolant
High-
Advanced temperature Heavy- Heavy-water
Gas-cooled gas-cooled gas-cooled water organic-cooled
reactor reactor reactor reactor reactor
Fuel
form
U
alloy
UOZ
ThCZ+UCz U02
UOZ
Enrichment
Natural
U
2%
U
93%
mU Natural
U
0.7-2%
235
U
Cladding
Mg
alloy
Stainless Graphite Zircaloy
Zircaloy
Moderator

Graphite Graphite
Graphite
D20
Dz0
Coolant
co2
coz
He
D2O
Terphenyl
Control material
B
B
B4
C
B4
C
B4
c
Reference
[Lll
[C21
IS11 IC11
[E21
CHEMICAL ENGINEERING ASPECTS
OF
NUCLEAR WWER
9
is
Generotor

Steam
Primary
Feed
Water
Water Pump Pump
-
Woter
Recirculator
Steam
Primary
Feed
Water
Water Pump Pump
Figure
1.8
Schematic
of
pressurized-water nuclear power plant
H20
Coolant
+
Moderator
Condenser
Condensote
-
-
.
Feed Water
Pump
Figure

1.9
Schematic
of
boiling-water nuclear power plant.
10
NUCLEAR CHEMICAL ENGINEERING
reactors;
helium
gas
has
also
been
proposed, Fast reactors need a higher ratio of fde to fertile
iaotopes
than
thermal reactors to support a chain reaction; a mixture of
20
percent plutonium
and
80
percent
2"U
is
typical
for
a fast-reactor fuel.
Mixed
dioxides
or
mixed monocarbides

are possible
fuel
materials. Although natural boron, which contains around
20
percent of the
strong
neutron-absorbing isotope
'OB,
is
satisfactory for control material in thermal reactors,
concentrated
'OB
is
preferred for some fast reactors.
The molten-salt reactor differs from all reactors thus far described in that it
uses
a liquid
dution of uranium
as
fuel and removes heat from the reactor by circulating hot fuel to an
external heat exchanger.
No
reactor coolant
is
employed other
than
the fuel itself. The
molten-salt breeder reactor
(MSBR)
uses

as
fuel a solution of UF4
in
a solvent salt
consisting
of
mixture of BeF2, 7LiF, and ThF4. Separated
'Li
is required instead of natural lithium because
the
7.5
percent of
6Li
in natural lithium would absorb
so
many neutrons
as
to make breeding
impossible. The
MSBR
is
a thermal reactor that breeds ='U from thorium; neutrons are
thermalized by means of graphite moderator blocks, fwed in the reactor, containing channels
through
which the molten salt flows.
Table
1.3
summarizes the materials used for the principal services
in
pressurized-water

and boiling-water reactors, the high-temperature gas-cooled reactor, fast reactors, and the
molten-salt reactor, and indicates which materials are fwed in each reactor and which
flow
through it.
5
FUEL
PROCESSING
FLOW SHEETS
5.1
Uranium
Fuel
The fuel processing operations to be used in conjunction with a nuclear power reactor and the
amount of nuclear fuel that must be provided depend on the type of reactor and on the extent to
which fissile and fertile constituents in spent fuel discharged from the reactor are to be recovered
for reuse. Figures
1.10
and
1.1
1
outline representative fuel processing flow sheets for uranium-
fueled thermal reactors generating
IO00
MW
of electricity, at
a
capacity factor of
80
percent.
Table
1.3

Materials
for
light-water, fast-breeder,
and
molten-salt reactors
Pressurized- Boiling-
Liquid-metal
Gas-cooled Molten-salt
water
water fast-breeder
fast-breeder breeder
reactor reactor
reactor
reactor reactor+
Fuel
Cladding
Moderator
Coolant
Control
material
Fixed
in
reactor
Circulating
Reference
uoz
,
uoz
,
3.3%

2.6%
=U
zircaloy zircaloy
H20
Hz
0
H20
H10
Hf
or
Ag-In-Cd B4C
Fuel Fuel
Coolant and moderator
IC31 [C31
20%
Puoz-
80%
u8
uoz
Stainless
None
Na
B4C
or
"B4C
Fuel
Coolant
[All
20%
PUOz-

80%
ue
uoz
Stainless
None
He
B4C
or
"B4C
Fuel
coolant
[Ell
71.7
m/o
7LiF
16
m/o
BeFz
12
m/o
ThF4
0.3
m/o
D'UF4
None
Graphite
Fuel
Moderator
Fuel
[Bll

+m/o
=
mole percent.
CHEMICAL ENGINEERING
ASPECTS
OF
NUCLEAR POWER
11
Fuel
Preporation
2%
143
u
+
Notural
Uranium
Conversion
(144
MT
U)
Figure
1.10
Fuel
processing flow sheet
for
100bMWe
heavy-water reactor. Basis:
1
year,
80

percent capacity factor.
The simplest flow sheet, Fig. 1.10, is applicable to heavy-water reactors fueled with natural
uranium containing
0.711
w/o
"'U.t
Feed preparation for this type of reactor consists of
purifying natural uranium concentrates, converting the uranium to U02, and fabricating the
UOz
into fuel elements. In this type of heavy-water reactor, fission of
23sU
initially present in
the feed and fission
of
plutonium formed from will produce about 6800 MWd of heat per
metric ton
(1
MT
=
1000
kg)
of fuel before the
fuel
is
so
depleted
in
fissile material and
so
loaded

with neutron-absorbing fission products that the reactor is no longer critical. Since the heat of
fission is 0.95 MWd/g, complete utiljzation
of
1
MT of fuel would generate
950,000
MWd
of
heat.
In
this type of thermal reactor,
thus,
6800/950,000
=
0.0072
fraction of the natural uranium,
about
0.7
percent, is converted to heat.
As
the efficiency of conversion of heat to electricity in a heavy-water nuclear power plant
is about
30
percent, the rate at which a
1WMW
plant would have to be supplied with natural
uranium
is
or
143 MT of uranium per year.

U308.
In
this unit, the annual uranium consumption of this reactor would be
In commercial transactions uranium concentrates are measured in short tons
(2000
Ib) of
(143 MT ux1.1023 short tonshlTX842 MT u30s/714
MT
U3)
=
187
short
tons
u308
0.995
assuming 99.5 percent uranium recovery
in
conversion.
Spent fuel discharged from this reactor contains about
0.2
w/o plutonium and about
0.3
w/o
usU.
This content of fissile material is
so
low that its recovery is hardly economical,
so
that no recovery step has been shown.
Figure

1.11 shows
three possible fuel processing flow sheets for reactors cooled and
moderated by light water. The specific example shown
is
for a pressurized-water reactor. Fuel
for this type of reactor consists of
UOz
enriched
to
around
3.3
w/o in
=U.
The expected
performance of this type of reactor
is
described
in
some detail in Chap.
3,
Sec.
7.
After
tw/o
=
weight percent.
12
NUCLEAR CHEMICAL ENGINEERING
1.
SPENT FUEL NOT RECYCLED

u3,
a83
%
U-235
246
kg
PU
Fiub
PfOduCh
Enriched UF,,
3.3%
U-235
Natural
R
(201
MT
U)
II.
SPENT FUEL REPROCESSED, URANIUM RECYCLED
Recovered Pu,
"02
244
kg
Recovered UF
0.83% U-238
25.8
MT
U
Natural
(169

MT
U)
IE.
SPENT FUEL REPROCESSED, URANIUM
AhlD
PLUTONIUM RECYCLED
Recovered Pu, 445
kg
(9
MT
U)
cnrii
3.3
A
I
Fission
Product?
Natural
Uronium
Cme
Uranium
148
shoct
~2
Conversion
Yo,
.
Depleted
UFg
y

0.3%
U-235
(IJ4fiUI
Figure
1.11
Fuel processing
flow
sheets for 1000-MWe pressurized-water reactor. Basis:
1
year,
80
percent capacity factor.
CHEMICAL ENGINEERING ASPECTS
OF
NUCLEAR POWER
13
producing
33,000
MWd of heat
per
metric ton, the fuel ceases
to
support the fission chain
reaction and must
be
discharged from the reactor. This spent fuel still contains around
0.83
w/o '%U and about 0.6 w/o fissile plutonium. In part
I
of Fig. 1.11 this spent fuel is stored

without reprocessing, as in the heavy-water reactor example of Fig. 1.10. The annual
consumption of
U308
for the light-water reactor, without reprocessing, is 262 short tons
U308, substantially greater than for the heavy-water reactor.
Under some conditions it is economically attractive
or
environmentally preferable to
reprocess spent fuel
in
order to (1) recover uranium
to
be recycled
to
provide part of the
enriched uranium used in subsequent
lots
of fuel, (2) recover plutonium, and
(3)
reduce
radioactive wastes
to
more compact form. In part I1 of Fig. 1.11 the recovered
0.83
percent
enriched uranium is recycled and the
244
kg of plutonium recovered
per
year is stored for later

use
in
either a light-water reactor
or
a fast-breeder reactor.
This
recycle of uranium to the
isotope separation plant reduces the annual
U308
feed rate
to
220
short tons, still appreciably
greater than for the heavy-water reactor.
In part I11 of Fig. 1.1 1, the recovered uranium is recycled and reenriched and the recovered
plutonium is recycled to provide part of the fissile material in the reactor fuel assemblies. Two
kinds
of
fuel assemblies are used. One kind is the same as used in cases I and
11,
which consist
of
U02
enriched to
3.3
w/o
235U.
The annual feed rate
of
these assemblies is 18.3 MT

of
enriched uranium. The other kind consists of mixed uranium and plutonium dioxides,
in
which
the uranium is in the form of natural
UOz.
Their annual feed rate is 8.9 MT of heavy metal
(uranium plus plutonium), including
445
kg of recycle plutonium. The total annual
U308
feed
rate is 160 short tons, which is less than for the heavy-water reactor of Fig. 1.10.
In part
111
of Fig. 1.1 1, the 160 short tons of
U308
consumed per year corresponds to a
daily feed rate
of
341 kg natural uranium.
As
this pressurized-water nuclear power plant has a
thermal efficiency of 32.5 percent, the fraction of the natural uranium feed converted to
energy is
Even with plutonium recycle, thus, this thermal reactor converts less than 1 percent
of
natural
uranium to energy. This low uranium utilization results from the fact that the conversion ratio
of

238U
to plutonium
in
a thermal reactor is less than unity.
In a fast reactor, on the other hand, the conversion ratio can be greater than unity, and
almost all of the uranium can be converted
to
energy, in principle. Figure 1.12 shows the fuel
processing operations associated with a fast-reactor power plant breeding plutonium from
238
U.
Because
of
the low absorption cross section
of
plutonium for fast neutrons, it is necessary to
use a mixture of about 20 percent plutonium and
80
percent
238U
in the core of such a reactor
and
to
surround the core with a blanket of natural or depleted uranium
to
absorb neutrons
leaking from the core and convert them to plutonium. Two types of fuel elements must be
prepared for a fast-breeder reactor, then, blanket elements fabricated from natural or depleted
uranium, and core elements containing around
20

w/o plutonium. Most fast reactors under
development propose
use
of mixed
Pu02-U02
for core elements; mixed PUC-UC is also being
considered. The core elements of a fast reactor are expected to generate from about 65,000
to
100,000 MWd of heat per metric ton before discharge; as they still contain nearly their original
plutonium content, reprocessing is required. The blanket elements also must be reprocessed for
plutonium recovery. Some savings can
be
effected by reprocessing both types of elements
together, as shown in Fig. 1.12. Uranium recovered in the reprocessing plant can be recycled
to
provide most of the uranium used to prepare core and blanket elements. Plutonium recovered
in the reprocessing plant provides
all
the enrichment needed for core elements, plus the net
production of plutonium from the plant. With good conservation of neutrons
in
the
reactor and
efficient recovery of plutonium in reprocessing and core fabrication, a 1000-MWe fast-reactor
14
NUCLEAR CHEMICAL ENGINEERING
Natural
nr
Net product
Recovered Plutonium, Recycled Plutonium

7-265
kp
A
Natural
or
Depleted
Uranium
1.5
MT
U
1
7.9
MT
U
UOa
+
Neutrons
1
1 1
1
t
Blanket
9.2
u
Reoctor
e
Blanket -Irradiated
uo2 Blanket,
-
7

Preporation
u
+
Pu
-
Recovered Uranium, Recycled
-
+
Figure
1.12
Fuel
processing flow sheet for
1000-MWe
fast-breeder reactor. Basis:
1
year,
80
percent capacity factor.
power plant is expected to breed about
265
kg/year of net plutonium product.
A
fast-reactor
power plant cooled with sodium or helium is expected to have
a
thermal efficiency of
40
percent. If it could convert
100
percent of its uranium feed to heat,

a
1000-We plant would
consume only
of uranium. Because of reprocessing losses and conversion of some uranium to nonfissile
isotopes, the uranium consumption of a practical fast-breeder system is expected
to
be
somewhat greater, perhaps
4
kg/day, or
1.5
MT/uranium/year. This is much less
than
for a
thermal reactor, and could be in the form of the depleted uranium
tailings
from the isotope
separation plant
of
Fig. 1.1
1.
5.2
Thorium
Fuel
Figure 1.13
shows
fuel processing arrangements needed for the
two
types of thorium-fueled
reactors mentioned in Sec.

4.
As
the conversion ratio of the high-temperature gas-cooled reactor
(HTGR)
is slightly less than unity, feed for this reactor consists of thorium plus some highly
enriched
='U
from a uranium isotope separation plant. In the fuel preparation operation
thorium, enriched
UF6,
and uranium recovered from spent fuel and recycled are formed into
fuel elements consisting of the carbides
ThC2
and
UC2
or the oxides
Tho2
and
U02
clad with
graphite. Fuel processing after irradiation consists of burning the carbon out of the fuel,
followed by separation of the mixed oxides by solvent extraction into uranium to be recycled
and radioactive fission products and thorium to be stored. The recycled uranium
is
a mixture of
isotopes, mostly formed by absorption of neutrons in thorium. More detail is given in
Chap.
3.
Fuel processing operations for the molten-salt breeder reactor are simpler in principle than
for the

HTGR
As
the conversion ratio
is
expected to be above unity, no fissile feed is needed